Nuclear Resonance Fluorescence for Materials Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quiter, Brian; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX's photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

MCNP/X TRANSPORT IN THE TABULAR REGIME

DOE Office of Scientific and Technical Information (OSTI.GOV)

HUGHES, H. GRADY

2007-01-08

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

Evaluation of an alternative shielding materials for F-127 transport package

NASA Astrophysics Data System (ADS)

Gual, Maritza R.; Mesquita, Amir Z.; Pereira, Cláubia

2018-03-01

Lead is used as radiation shielding material for the Nordion's F-127 source shipping container is used for transport and storage of the GammaBeam -127's cobalt-60 source of the Nuclear Technology Development Center (CDTN) located in Belo Horizonte, Brazil. As an alternative, Th, Tl and WC have been evaluated as radiation shielding material. The goal is to check their behavior regarding shielding and dosing. Monte Carlo MCNPX code is used for the simulations. In the MCNPX calculation was used one cylinder as exclusion surface instead one sphere. Validation of MCNPX gamma doses calculations was carried out through comparison with experimental measurements. The results show that tungsten carbide WC is better shielding material for γ-ray than lead shielding.

Introducing single-crystal scattering and optical potentials into MCNPX: Predicting neutron emission from a convoluted moderator

DOE PAGES

Gallmeier, F. X.; Iverson, E. B.; Lu, W.; ...

2016-01-08

Neutron transport simulation codes are an indispensable tool used for the design and construction of modern neutron scattering facilities and instrumentation. It has become increasingly clear that some neutron instrumentation has started to exploit physics that is not well-modelled by the existing codes. Particularly, the transport of neutrons through single crystals and across interfaces in MCNP(X), Geant4 and other codes ignores scattering from oriented crystals and refractive effects, and yet these are essential ingredients for the performance of monochromators and ultra-cold neutron transport respectively (to mention but two examples). In light of these developments, we have extended the MCNPX codemore » to include a single-crystal neutron scattering model and neutron reflection/refraction physics. Furthermore, we have also generated silicon scattering kernels for single crystals of definable orientation with respect to an incoming neutron beam. As a first test of these new tools, we have chosen to model the recently developed convoluted moderator concept, in which a moderating material is interleaved with layers of perfect crystals to provide an exit path for neutrons moderated to energies below the crystal s Bragg cut off at locations deep within the moderator. Studies of simple cylindrical convoluted moderator systems of 100 mm diameter and composed of polyethylene and single crystal silicon were performed with the upgraded MCNPX code and reproduced the magnitude of effects seen in experiments compared to homogeneous moderator systems. Applying different material properties for refraction and reflection, and by replacing the silicon in the models with voids, we show that the emission enhancements seen in recent experiments are primarily caused by the transparency of the silicon/void layers. Finally the convoluted moderator experiments described by Iverson et al. were simulated and we find satisfactory agreement between the measurement and the results of simulations performed using the tools we have developed.« less

Simulation of neutron production using MCNPX+MCUNED.

PubMed

Erhard, M; Sauvan, P; Nolte, R

2014-10-01

In standard MCNPX, the production of neutrons by ions cannot be modelled efficiently. The MCUNED patch applied to MCNPX 2.7.0 allows to model the production of neutrons by light ions down to energies of a few kiloelectron volts. This is crucial for the simulation of neutron reference fields. The influence of target properties, such as the diffusion of reactive isotopes into the target backing or the effect of energy and angular straggling, can be studied efficiently. In this work, MCNPX/MCUNED calculations are compared with results obtained with the TARGET code for simulating neutron production. Furthermore, MCUNED incorporates more effective variance reduction techniques and a coincidence counting tally. This allows the simulation of a TCAP experiment being developed at PTB. In this experiment, 14.7-MeV neutrons will be produced by the reaction T(d,n)(4)He. The neutron fluence is determined by counting alpha particles, independently of the reaction cross section. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Simulation of the Mg(Ar) ionization chamber currents by different Monte Carlo codes in benchmark gamma fields

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Liu, Yuan-Hao; Nievaart, Sander; Chen, Yen-Fu; Wu, Shu-Wei; Chou, Wen-Tsae; Jiang, Shiang-Huei

2011-10-01

High energy photon (over 10 MeV) and neutron beams adopted in radiobiology and radiotherapy always produce mixed neutron/gamma-ray fields. The Mg(Ar) ionization chambers are commonly applied to determine the gamma-ray dose because of its neutron insensitive characteristic. Nowadays, many perturbation corrections for accurate dose estimation and lots of treatment planning systems are based on Monte Carlo technique. The Monte Carlo codes EGSnrc, FLUKA, GEANT4, MCNP5, and MCNPX were used to evaluate energy dependent response functions of the Exradin M2 Mg(Ar) ionization chamber to a parallel photon beam with mono-energies from 20 keV to 20 MeV. For the sake of validation, measurements were carefully performed in well-defined (a) primary M-100 X-ray calibration field, (b) primary 60Co calibration beam, (c) 6-MV, and (d) 10-MV therapeutic beams in hospital. At energy region below 100 keV, MCNP5 and MCNPX both had lower responses than other codes. For energies above 1 MeV, the MCNP ITS-mode greatly resembled other three codes and the differences were within 5%. Comparing to the measured currents, MCNP5 and MCNPX using ITS-mode had perfect agreement with the 60Co, and 10-MV beams. But at X-ray energy region, the derivations reached 17%. This work shows us a better insight into the performance of different Monte Carlo codes in photon-electron transport calculation. Regarding the application of the mixed field dosimetry like BNCT, MCNP with ITS-mode is recognized as the most suitable tool by this work.

Nuclear Resonance Fluorescence for Materials Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quiter, Brian J.; Ludewigt, Bernhard; Mozin, Vladimir

This paper discusses the use of nuclear resonance fluorescence (NRF) techniques for the isotopic and quantitative assaying of radioactive material. Potential applications include age-dating of an unknown radioactive source, pre- and post-detonation nuclear forensics, and safeguards for nuclear fuel cycles Examples of age-dating a strong radioactive source and assaying a spent fuel pin are discussed. The modeling work has ben performed with the Monte Carlo radiation transport computer code MCNPX, and the capability to simulate NRF has bee added to the code. Discussed are the limitations in MCNPX?s photon transport physics for accurately describing photon scattering processes that are importantmore » contributions to the background and impact the applicability of the NRF assay technique.« less

Effects of micro-sized and nano-sized WO3 on mass attenauation coefficients of concrete by using MCNPX code.

PubMed

Tekin, H O; Singh, V P; Manici, T

2017-03-01

In the present work the effect of tungsten oxide (WO 3 ) nanoparticles on mass attenauation coefficients of concrete has been investigated by using MCNPX (version 2.4.0). The validation of generated MCNPX simulation geometry has been provided by comparing the results with standard XCOM data for mass attenuation coefficients of concrete. A very good agreement between XCOM and MCNPX have been obtained. The validated geometry has been used for definition of nano-WO 3 and micro-WO 3 into concrete sample. The mass attenuation coefficients of pure concrete and WO 3 added concrete with micro-sized and nano-sized have been compared. It was observed that shielding properties of concrete doped with WO 3 increased. The results of mass attenauation coefficients also showed that the concrete doped with nano-WO 3 significanlty improve shielding properties than micro-WO 3 . It can be concluded that addition of nano-sized particles can be considered as another mechanism to reduce radiation dose. Copyright © 2016 Elsevier Ltd. All rights reserved.

Comparison of a 3-D multi-group SN particle transport code with Monte Carlo for intracavitary brachytherapy of the cervix uteri.

PubMed

Gifford, Kent A; Wareing, Todd A; Failla, Gregory; Horton, John L; Eifel, Patricia J; Mourtada, Firas

2009-12-03

A patient dose distribution was calculated by a 3D multi-group S N particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs-137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi-group S N particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within +/- 3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than +/- 1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs-137 CT-based patient geometry. Our data showed that a three-group cross-section set is adequate for Cs-137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations.

Comparison of a 3D multi‐group SN particle transport code with Monte Carlo for intercavitary brachytherapy of the cervix uteri

PubMed Central

Wareing, Todd A.; Failla, Gregory; Horton, John L.; Eifel, Patricia J.; Mourtada, Firas

2009-01-01

A patient dose distribution was calculated by a 3D multi‐group SN particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs‐137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi‐group SN particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within ±3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than ±1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs‐137 CT‐based patient geometry. Our data showed that a three‐group cross‐section set is adequate for Cs‐137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations. PACS number: 87.53.Jw

Benchmarking Heavy Ion Transport Codes FLUKA, HETC-HEDS MARS15, MCNPX, and PHITS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ronningen, Reginald Martin; Remec, Igor; Heilbronn, Lawrence H.

Powerful accelerators such as spallation neutron sources, muon-collider/neutrino facilities, and rare isotope beam facilities must be designed with the consideration that they handle the beam power reliably and safely, and they must be optimized to yield maximum performance relative to their design requirements. The simulation codes used for design purposes must produce reliable results. If not, component and facility designs can become costly, have limited lifetime and usefulness, and could even be unsafe. The objective of this proposal is to assess the performance of the currently available codes PHITS, FLUKA, MARS15, MCNPX, and HETC-HEDS that could be used for designmore » simulations involving heavy ion transport. We plan to access their performance by performing simulations and comparing results against experimental data of benchmark quality. Quantitative knowledge of the biases and the uncertainties of the simulations is essential as this potentially impacts the safe, reliable and cost effective design of any future radioactive ion beam facility. Further benchmarking of heavy-ion transport codes was one of the actions recommended in the Report of the 2003 RIA R&D Workshop".« less

Simulation of a beam rotation system for a spallation source

NASA Astrophysics Data System (ADS)

Reiss, Tibor; Reggiani, Davide; Seidel, Mike; Talanov, Vadim; Wohlmuther, Michael

2015-04-01

With a nominal beam power of nearly 1 MW on target, the Swiss Spallation Neutron Source (SINQ), ranks among the world's most powerful spallation neutron sources. The proton beam transport to the SINQ target is carried out exclusively by means of linear magnetic elements. In the transport line to SINQ the beam is scattered in two meson production targets and as a consequence, at the SINQ target entrance the beam shape can be described by Gaussian distributions in transverse x and y directions with tails cut short by collimators. This leads to a highly nonuniform power distribution inside the SINQ target, giving rise to thermal and mechanical stresses. In view of a future proton beam intensity upgrade, the possibility of homogenizing the beam distribution by means of a fast beam rotation system is currently under investigation. Important aspects which need to be studied are the impact of a rotating proton beam on the resulting neutron spectra, spatial flux distributions and additional—previously not present—proton losses causing unwanted activation of accelerator components. Hence a new source description method was developed for the radiation transport code MCNPX. This new feature makes direct use of the results from the proton beam optics code TURTLE. Its advantage to existing MCNPX source options is that all phase space information and correlations of each primary beam particle computed with TURTLE are preserved and transferred to MCNPX. Simulations of the different beam distributions together with their consequences in terms of neutron production are presented in this publication. Additionally, a detailed description of the coupling method between TURTLE and MCNPX is provided.

Implementation of a small-angle scattering model in MCNPX for very cold neutron reflector studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grammer, Kyle B.; Gallmeier, Franz X.

Current neutron moderator media do not sufficiently moderate neutrons below the cold neutron regime into the very cold neutron (VCN) regime that is desirable for some physics applications. Nesvizhevsky et al [1] have demonstrated that nanodiamond powder efficiently reflect VCN via small angle scattering. He suggests that these effects could be exploited to boost the neutron output of a VCN moderator. Simulation studies of nanoparticle reflectors are being investigated as part of the development of a VCN source option for the SNS second target station. We are pursuing an expansion of the MCNPX code by implementation of an analytical small-anglemore » scattering function [2], which is adaptable by scattering particle sizes, distributions, and packing fractions in order to supplement currently existing scattering kernels. The analytical model and preliminary studies using MCNPX will be discussed.« less

Measurement of Thick Target Neutron Yields at 0-Degree Bombarded With 140-MeV, 250-MeV And 350-MeV Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iwamoto, Yosuke; /JAERI, Kyoto; Taniguchi, Shingo

Neutron energy spectra at 0{sup o} produced from stopping-length graphite, aluminum, iron and lead targets bombarded with 140, 250 and 350 MeV protons were measured at the neutron TOF course in RCNP of Osaka University. The neutron energy spectra were obtained by using the time-of-flight technique in the energy range from 10 MeV to incident proton energy. To compare the experimental results, Monte Carlo calculations with the PHITS and MCNPX codes were performed using the JENDL-HE and the LA150 evaluated nuclear data files, the ISOBAR model implemented in PHITS, and the LAHET code in MCNPX. It was found that thesemore » calculated results at 0{sup o} generally agreed with the experimental results in the energy range above 20 MeV except for graphite at 250 and 350 MeV.« less

Overview of Recent Radiation Transport Code Comparisons for Space Applications

NASA Astrophysics Data System (ADS)

Townsend, Lawrence

Recent advances in radiation transport code development for space applications have resulted in various comparisons of code predictions for a variety of scenarios and codes. Comparisons among both Monte Carlo and deterministic codes have been made and published by vari-ous groups and collaborations, including comparisons involving, but not limited to HZETRN, HETC-HEDS, FLUKA, GEANT, PHITS, and MCNPX. In this work, an overview of recent code prediction inter-comparisons, including comparisons to available experimental data, is presented and discussed, with emphases on those areas of agreement and disagreement among the various code predictions and published data.

Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations

PubMed Central

Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

2014-01-01

High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by 241Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy-1 for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy-1 achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production. PMID:24600167

Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations.

PubMed

Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

2014-01-01

High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.

A New Approach in Coal Mine Exploration Using Cosmic Ray Muons

NASA Astrophysics Data System (ADS)

Darijani, Reza; Negarestani, Ali; Rezaie, Mohammad Reza; Fatemi, Syed Jalil; Akhond, Ahmad

2016-08-01

Muon radiography is a technique that uses cosmic ray muons to image the interior of large scale geological structures. The muon absorption in matter is the most important parameter in cosmic ray muon radiography. Cosmic ray muon radiography is similar to X-ray radiography. The main aim in this survey is the simulation of the muon radiography for exploration of mines. So, the production source, tracking, and detection of cosmic ray muons were simulated by MCNPX code. For this purpose, the input data of the source card in MCNPX code were extracted from the muon energy spectrum at sea level. In addition, the other input data such as average density and thickness of layers that were used in this code are the measured data from Pabdana (Kerman, Iran) coal mines. The average thickness and density of these layers in the coal mines are from 2 to 4 m and 1.3 gr/c3, respectively. To increase the spatial resolution, a detector was placed inside the mountain. The results indicated that using this approach, the layers with minimum thickness about 2.5 m can be identified.

The COsmic-ray Soil Moisture Interaction Code (COSMIC) for use in data assimilation

NASA Astrophysics Data System (ADS)

Shuttleworth, J.; Rosolem, R.; Zreda, M.; Franz, T.

2013-08-01

Soil moisture status in land surface models (LSMs) can be updated by assimilating cosmic-ray neutron intensity measured in air above the surface. This requires a fast and accurate model to calculate the neutron intensity from the profiles of soil moisture modeled by the LSM. The existing Monte Carlo N-Particle eXtended (MCNPX) model is sufficiently accurate but too slow to be practical in the context of data assimilation. Consequently an alternative and efficient model is needed which can be calibrated accurately to reproduce the calculations made by MCNPX and used to substitute for MCNPX during data assimilation. This paper describes the construction and calibration of such a model, COsmic-ray Soil Moisture Interaction Code (COSMIC), which is simple, physically based and analytic, and which, because it runs at least 50 000 times faster than MCNPX, is appropriate in data assimilation applications. The model includes simple descriptions of (a) degradation of the incoming high-energy neutron flux with soil depth, (b) creation of fast neutrons at each depth in the soil, and (c) scattering of the resulting fast neutrons before they reach the soil surface, all of which processes may have parameterized dependency on the chemistry and moisture content of the soil. The site-to-site variability in the parameters used in COSMIC is explored for 42 sample sites in the COsmic-ray Soil Moisture Observing System (COSMOS), and the comparative performance of COSMIC relative to MCNPX when applied to represent interactions between cosmic-ray neutrons and moist soil is explored. At an example site in Arizona, fast-neutron counts calculated by COSMIC from the average soil moisture profile given by an independent network of point measurements in the COSMOS probe footprint are similar to the fast-neutron intensity measured by the COSMOS probe. It was demonstrated that, when used within a data assimilation framework to assimilate COSMOS probe counts into the Noah land surface model at the Santa Rita Experimental Range field site, the calibrated COSMIC model provided an effective mechanism for translating model-calculated soil moisture profiles into aboveground fast-neutron count when applied with two radically different approaches used to remove the bias between data and model.

CEM2k and LAQGSM Codes as Event-Generators for Space Radiation Shield and Cosmic Rays Propagation Applications

NASA Technical Reports Server (NTRS)

Mashnik, S. G.; Gudima, K. K.; Sierk, A. J.; Moskalenko, I. V.

2002-01-01

Space radiation shield applications and studies of cosmic ray propagation in the Galaxy require reliable cross sections to calculate spectra of secondary particles and yields of the isotopes produced in nuclear reactions induced both by particles and nuclei at energies from threshold to hundreds of GeV per nucleon. Since the data often exist in a very limited energy range or sometimes not at all, the only way to obtain an estimate of the production cross sections is to use theoretical models and codes. Recently, we have developed improved versions of the Cascade-Exciton Model (CEM) of nuclear reactions: the codes CEM97 and CEM2k for description of particle-nucleus reactions at energies up to about 5 GeV. In addition, we have developed a LANL version of the Quark-Gluon String Model (LAQGSM) to describe reactions induced both by particles and nuclei at energies up to hundreds of GeVhucleon. We have tested and benchmarked the CEM and LAQGSM codes against a large variety of experimental data and have compared their results with predictions by other currently available models and codes. Our benchmarks show that CEM and LAQGSM codes have predictive powers no worse than other currently used codes and describe many reactions better than other codes; therefore both our codes can be used as reliable event-generators for space radiation shield and cosmic ray propagation applications. The CEM2k code is being incorporated into the transport code MCNPX (and several other transport codes), and we plan to incorporate LAQGSM into MCNPX in the near future. Here, we present the current status of the CEM2k and LAQGSM codes, and show results and applications to studies of cosmic ray propagation in the Galaxy.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

NASA Astrophysics Data System (ADS)

Hosseini, Seyed Abolfazl; Afrakoti, Iman Esmaili Paeen

2017-04-01

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.

SU-E-T-493: Accelerated Monte Carlo Methods for Photon Dosimetry Using a Dual-GPU System and CUDA.

PubMed

Liu, T; Ding, A; Xu, X

2012-06-01

To develop a Graphics Processing Unit (GPU) based Monte Carlo (MC) code that accelerates dose calculations on a dual-GPU system. We simulated a clinical case of prostate cancer treatment. A voxelized abdomen phantom derived from 120 CT slices was used containing 218×126×60 voxels, and a GE LightSpeed 16-MDCT scanner was modeled. A CPU version of the MC code was first developed in C++ and tested on Intel Xeon X5660 2.8GHz CPU, then it was translated into GPU version using CUDA C 4.1 and run on a dual Tesla m 2 090 GPU system. The code was featured with automatic assignment of simulation task to multiple GPUs, as well as accurate calculation of energy- and material- dependent cross-sections. Double-precision floating point format was used for accuracy. Doses to the rectum, prostate, bladder and femoral heads were calculated. When running on a single GPU, the MC GPU code was found to be ×19 times faster than the CPU code and ×42 times faster than MCNPX. These speedup factors were doubled on the dual-GPU system. The dose Result was benchmarked against MCNPX and a maximum difference of 1% was observed when the relative error is kept below 0.1%. A GPU-based MC code was developed for dose calculations using detailed patient and CT scanner models. Efficiency and accuracy were both guaranteed in this code. Scalability of the code was confirmed on the dual-GPU system. © 2012 American Association of Physicists in Medicine.

Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

NASA Astrophysics Data System (ADS)

Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

2018-03-01

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

Calculation of Viscous Effects on Ship Wave Resistance Using Axisymmetric Boundary Layer Approaches

DTIC Science & Technology

1985-05-13

Layers in Pressure Gradients," NSRDC Report 3308, April 1970. 38. Garcia, J.M. and Zazurca, J.A.A., " Calculo de la Resistencia Viscosa de un Buque a...none USERS 21 ABSTRACT SECURITY CLASSIFICATION UNCLASSIFIED 22a NAME OF RESPONSIBLE INDIVIDUAL Henry T. Wang 22b TELEPHONE (Include Area Code...theory. Since then, calculation of the resistance due to the waves generated by a surface ship advancing at constant forward speed has been an area of

MCNPX Cosmic Ray Shielding Calculations with the NORMAN Phantom Model

NASA Technical Reports Server (NTRS)

James, Michael R.; Durkee, Joe W.; McKinney, Gregg; Singleterry Robert

2008-01-01

The United States is planning manned lunar and interplanetary missions in the coming years. Shielding from cosmic rays is a critical aspect of manned spaceflight. These ventures will present exposure issues involving the interplanetary Galactic Cosmic Ray (GCR) environment. GCRs are comprised primarily of protons (approx.84.5%) and alpha-particles (approx.14.7%), while the remainder is comprised of massive, highly energetic nuclei. The National Aeronautics and Space Administration (NASA) Langley Research Center (LaRC) has commissioned a joint study with Los Alamos National Laboratory (LANL) to investigate the interaction of the GCR environment with humans using high-fidelity, state-of-the-art computer simulations. The simulations involve shielding and dose calculations in order to assess radiation effects in various organs. The simulations are being conducted using high-resolution voxel-phantom models and the MCNPX[1] Monte Carlo radiation-transport code. Recent advances in MCNPX physics packages now enable simulated transport over 2200 types of ions of widely varying energies in large, intricate geometries. We report here initial results obtained using a GCR spectrum and a NORMAN[3] phantom.

Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

PubMed

Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

2016-03-08

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes.

Neutron displacement cross-sections for tantalum and tungsten at energies up to 1 GeV

NASA Astrophysics Data System (ADS)

Broeders, C. H. M.; Konobeyev, A. Yu.; Villagrasa, C.

2005-06-01

The neutron displacement cross-section has been evaluated for tantalum and tungsten at energies from 10 -5 eV up to 1 GeV. The nuclear optical model, the intranuclear cascade model combined with the pre-equilibrium and evaporation models were used for the calculations. The number of defects produced by recoil atoms nuclei in materials was calculated by the Norgett, Robinson, Torrens model and by the approach combining calculations using the binary collision approximation model and the results of the molecular dynamics simulation. The numerical calculations were done using the NJOY code, the ECIS96 code, the MCNPX code and the IOTA code.

Electronics Devices and Materials

DTIC Science & Technology

2008-03-17

Molecular -bea epitaxy MCNPX ............... Software code Misse6 ................. Satellite expected to carry ORMatE-I Misse7...patterning using electron beam lithography), spaces (class 1000 clean benches), and skills (appropriate mix of skilled technicians and professionals...34 Process samples for various projects such as Antimode Base High Electron Mobility Transistors ( HEMT ) and Double Heterojuction Bipolar Transistors

Proton Lateral Broadening Distribution Comparisons Between GRNTRN, MCNPX, and Laboratory Beam Measurements

NASA Technical Reports Server (NTRS)

Mertens, Christopher J.; Moyers, Michael F.; Walker, Steven A.; Tweed, John

2010-01-01

Recent developments in NASA s deterministic High charge (Z) and Energy TRaNsport (HZETRN) code have included lateral broadening of primary ion beams due to small-angle multiple Coulomb scattering, and coupling of the ion-nuclear scattering interactions with energy loss and straggling. This new version of HZETRN is based on Green function methods, called GRNTRN, and is suitable for modeling transport with both space environment and laboratory boundary conditions. Multiple scattering processes are a necessary extension to GRNTRN in order to accurately model ion beam experiments, to simulate the physical and biological-effective radiation dose, and to develop new methods and strategies for light ion radiation therapy. In this paper we compare GRNTRN simulations of proton lateral broadening distributions with beam measurements taken at Loma Linda University Proton Therapy Facility. The simulated and measured lateral broadening distributions are compared for a 250 MeV proton beam on aluminum, polyethylene, polystyrene, bone substitute, iron, and lead target materials. The GRNTRN results are also compared to simulations from the Monte Carlo MCNPX code for the same projectile-target combinations described above.

Digital pile-up rejection for plutonium experiments with solution-grown stilbene

NASA Astrophysics Data System (ADS)

Bourne, M. M.; Clarke, S. D.; Paff, M.; DiFulvio, A.; Norsworthy, M.; Pozzi, S. A.

2017-01-01

A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of 241Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding.

Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes.

PubMed

Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

2013-08-21

The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX's MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application.

A scintillator-based approach to monitor secondary neutron production during proton therapy.

PubMed

Clarke, S D; Pryser, E; Wieger, B M; Pozzi, S A; Haelg, R A; Bashkirov, V A; Schulte, R W

2016-11-01

The primary objective of this work is to measure the secondary neutron field produced by an uncollimated proton pencil beam impinging on different tissue-equivalent phantom materials using organic scintillation detectors. Additionally, the Monte Carlo code mcnpx-PoliMi was used to simulate the detector response for comparison to the measured data. Comparison of the measured and simulated data will validate this approach for monitoring secondary neutron dose during proton therapy. Proton beams of 155- and 200-MeV were used to irradiate a variety of phantom materials and secondary particles were detected using organic liquid scintillators. These detectors are sensitive to fast neutrons and gamma rays: pulse shape discrimination was used to classify each detected pulse as either a neutron or a gamma ray. The mcnpx-PoliMi code was used to simulate the secondary neutron field produced during proton irradiation of the same tissue-equivalent phantom materials. An experiment was performed at the Loma Linda University Medical Center proton therapy research beam line and corresponding models were created using the mcnpx-PoliMi code. The authors' analysis showed agreement between the simulations and the measurements. The simulated detector response can be used to validate the simulations of neutron and gamma doses on a particular beam line with or without a phantom. The authors have demonstrated a method of monitoring the neutron component of the secondary radiation field produced by therapeutic protons. The method relies on direct detection of secondary neutrons and gamma rays using organic scintillation detectors. These detectors are sensitive over the full range of biologically relevant neutron energies above 0.5 MeV and allow effective discrimination between neutron and photon dose. Because the detector system is portable, the described system could be used in the future to evaluate secondary neutron and gamma doses on various clinical beam lines for commissioning and prospective data collection in pediatric patients treated with proton therapy.

Comparison of TG‐43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes

PubMed Central

Zaker, Neda; Sina, Sedigheh; Koontz, Craig; Meigooni1, Ali S.

2016-01-01

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross‐sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross‐sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code — MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low‐energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg PMID:27074460

Cellular dosimetry calculations for Strontium-90 using Monte Carlo code PENELOPE.

PubMed

Hocine, Nora; Farlay, Delphine; Boivin, Georges; Franck, Didier; Agarande, Michelle

2014-11-01

To improve risk assessments associated with chronic exposure to Strontium-90 (Sr-90), for both the environment and human health, it is necessary to know the energy distribution in specific cells or tissue. Monte Carlo (MC) simulation codes are extremely useful tools for calculating deposition energy. The present work was focused on the validation of the MC code PENetration and Energy LOss of Positrons and Electrons (PENELOPE) and the assessment of dose distribution to bone marrow cells from punctual Sr-90 source localized within the cortical bone part. S-values (absorbed dose per unit cumulated activity) calculations using Monte Carlo simulations were performed by using PENELOPE and Monte Carlo N-Particle eXtended (MCNPX). Cytoplasm, nucleus, cell surface, mouse femur bone and Sr-90 radiation source were simulated. Cells are assumed to be spherical with the radii of the cell and cell nucleus ranging from 2-10 μm. The Sr-90 source is assumed to be uniformly distributed in cell nucleus, cytoplasm and cell surface. The comparison of S-values calculated with PENELOPE to MCNPX results and the Medical Internal Radiation Dose (MIRD) values agreed very well since the relative deviations were less than 4.5%. The dose distribution to mouse bone marrow cells showed that the cells localized near the cortical part received the maximum dose. The MC code PENELOPE may prove useful for cellular dosimetry involving radiation transport through materials other than water, or for complex distributions of radionuclides and geometries.

Benchmarking of Neutron Production of Heavy-Ion Transport Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald M.; Heilbronn, Lawrence

Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models andmore » codes and additional benchmarking are required.« less

Benchmarking of Heavy Ion Transport Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald M.; Heilbronn, Lawrence

Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in designing and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models andmore » codes and additional benchmarking are required.« less

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Brachytherapy treatment simulation of strontium-90 and ruthenium-106 plaques on small size posterior uveal melanoma using MCNPX code

NASA Astrophysics Data System (ADS)

Barbosa, N. A.; da Rosa, L. A. R.; Facure, A.; Braz, D.

2014-02-01

Concave eye applicators with 90Sr/90Y and 106Ru/106Rh beta-ray sources are usually used in brachytherapy for the treatment of superficial intraocular tumors as uveal melanoma with thickness up to 5 mm. The aim of this work consisted in using the Monte Carlo code MCNPX to calculate the 3D dose distribution on a mathematical model of the human eye, considering 90Sr/90Y and 160Ru/160Rh beta-ray eye applicators, in order to treat a posterior uveal melanoma with a thickness 3.8 mm from the choroid surface. Mathematical models were developed for the two ophthalmic applicators, CGD produced by BEBIG Company and SIA.6 produced by the Amersham Company, with activities 1 mCi and 4.23 mCi respectively. They have a concave form. These applicators' mathematical models were attached to the eye model and the dose distributions were calculated using the MCNPX *F8 tally. The average doses rates were determined in all regions of the eye model. The *F8 tally results showed that the deposited energy due to the applicator with the radionuclide 106Ru/106Rh is higher in all eye regions, including tumor. However the average dose rate in the tumor region is higher for the applicator with 90Sr/90Y, due to its high activity. Due to the dosimetric characteristics of these applicators, the PDD value for 3 mm water is 73% for the 106Ru/106Rh applicator and 60% for 90Sr/90Y applicator. For a better choice of the applicator type and radionuclide it is important to know the thickness of the tumor and its location.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zieb, Kristofer James Ekhart; Hughes, Henry Grady III; Xu, X. George

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. Here, this article discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

Monte Carlo Simulation of a Segmented Detector for Low-Energy Electron Antineutrinos

NASA Astrophysics Data System (ADS)

Qomi, H. Akhtari; Safari, M. J.; Davani, F. Abbasi

2017-11-01

Detection of low-energy electron antineutrinos is of importance for several purposes, such as ex-vessel reactor monitoring, neutrino oscillation studies, etc. The inverse beta decay (IBD) is the interaction that is responsible for detection mechanism in (organic) plastic scintillation detectors. Here, a detailed study will be presented dealing with the radiation and optical transport simulation of a typical segmented antineutrino detector withMonte Carlo method using MCNPX and FLUKA codes. This study shows different aspects of the detector, benefiting from inherent capabilities of the Monte Carlo simulation codes.

Shielding Analyses for VISION Beam Line at SNS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popova, Irina; Gallmeier, Franz X

2014-01-01

Full-scale neutron and gamma transport analyses were performed to design shielding around the VISION beam line, instrument shielding enclosure, beam stop, secondary shutter including a temporary beam stop for the still closed neighboring beam line to meet requirement is to achieve dose rates below 0.25 mrem/h at 30 cm from the shielding surface. The beam stop and the temporary beam stop analyses were performed with the discrete ordinate code DORT additionally to Monte Carlo analyses with the MCNPX code. Comparison of the results is presented.

GUINEVERE experiment: Kinetic analysis of some reactivity measurement methods by deterministic and Monte Carlo codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bianchini, G.; Burgio, N.; Carta, M.

The GUINEVERE experiment (Generation of Uninterrupted Intense Neutrons at the lead Venus Reactor) is an experimental program in support of the ADS technology presently carried out at SCK-CEN in Mol (Belgium). In the experiment a modified lay-out of the original thermal VENUS critical facility is coupled to an accelerator, built by the French body CNRS in Grenoble, working in both continuous and pulsed mode and delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target. The modified lay-out of the facility consists of a fast subcritical core made of 30% U-235 enriched metallic Uranium in a lead matrix. Severalmore » off-line and on-line reactivity measurement techniques will be investigated during the experimental campaign. This report is focused on the simulation by deterministic (ERANOS French code) and Monte Carlo (MCNPX US code) calculations of three reactivity measurement techniques, Slope ({alpha}-fitting), Area-ratio and Source-jerk, applied to a GUINEVERE subcritical configuration (namely SC1). The inferred reactivity, in dollar units, by the Area-ratio method shows an overall agreement between the two deterministic and Monte Carlo computational approaches, whereas the MCNPX Source-jerk results are affected by large uncertainties and allow only partial conclusions about the comparison. Finally, no particular spatial dependence of the results is observed in the case of the GUINEVERE SC1 subcritical configuration. (authors)« less

Proton Dose Assessment to the Human Eye Using Monte Carlo N-Particle Transport Code (MCNPX)

DTIC Science & Technology

2006-08-01

current treatments are applied using an infrared diode laser 10 (projecting a spot size of 2-3 mm), used for about 1 minute per exposure. The laser heats...1983. Shultis J, Faw R. An MCNP Primer. Available at: http:// ww2 .mne.ksu.edu/-jks/MCNPprmr.pdf. Accessed 3 January 2006. Stys P, Lopachin R

A high-fidelity Monte Carlo evaluation of CANDU-6 safety parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Y.; Hartanto, D.

2012-07-01

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium) reactor have been evaluated by using a modified MCNPX code. For accurate analysis of the parameters, the DBRC (Doppler Broadening Rejection Correction) scheme was implemented in MCNPX in order to account for the thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted by using the MCNPX and the FTC value is evaluated for several burnup points including the mid-burnupmore » representing a near-equilibrium core. The Doppler effect has been evaluated by using several cross section libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR value is also evaluated at mid-burnup conditions to characterize safety features of equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, huge number of neutron histories are considered in this work and the standard deviation of the k-inf values is only 0.5{approx}1 pcm. It has been found that the FTC is significantly enhanced by accounting for the Doppler broadening of scattering resonance and the PCR are clearly improved. (authors)« less

Internal dosimetry with the Monte Carlo code GATE: validation using the ICRP/ICRU female reference computational model

NASA Astrophysics Data System (ADS)

Villoing, Daphnée; Marcatili, Sara; Garcia, Marie-Paule; Bardiès, Manuel

2017-03-01

The purpose of this work was to validate GATE-based clinical scale absorbed dose calculations in nuclear medicine dosimetry. GATE (version 6.2) and MCNPX (version 2.7.a) were used to derive dosimetric parameters (absorbed fractions, specific absorbed fractions and S-values) for the reference female computational model proposed by the International Commission on Radiological Protection in ICRP report 110. Monoenergetic photons and electrons (from 50 keV to 2 MeV) and four isotopes currently used in nuclear medicine (fluorine-18, lutetium-177, iodine-131 and yttrium-90) were investigated. Absorbed fractions, specific absorbed fractions and S-values were generated with GATE and MCNPX for 12 regions of interest in the ICRP 110 female computational model, thereby leading to 144 source/target pair configurations. Relative differences between GATE and MCNPX obtained in specific configurations (self-irradiation or cross-irradiation) are presented. Relative differences in absorbed fractions, specific absorbed fractions or S-values are below 10%, and in most cases less than 5%. Dosimetric results generated with GATE for the 12 volumes of interest are available as supplemental data. GATE can be safely used for radiopharmaceutical dosimetry at the clinical scale. This makes GATE a viable option for Monte Carlo modelling of both imaging and absorbed dose in nuclear medicine.

Benchmarking of neutron production of heavy-ion transport codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, I.; Ronningen, R. M.; Heilbronn, L.

Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondarymore » neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)« less

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

2011-01-14

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity ofmore » the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development time, nor concerns related to the use of Pu in measurement systems. This report discusses basic NRF measurement concepts, i.e., backscatter and transmission methods, and photon source and {gamma}-ray detector options in Section 2. An analytical model for calculating NRF signal strengths is presented in Section 3 together with enhancements to the MCNPX code and descriptions of modeling techniques that were drawn upon in the following sections. Making extensive use of the model and MCNPX simulations, the capabilities of the backscatter and transmission methods based on bremsstrahlung or quasi-monoenergetic photon sources were analyzed as described in Sections 4 and 5. A recent transmission experiment is reported on in Appendix A. While this experiment was not directly part of this project, its results provide an important reference point for our analytical estimates and MCNPX simulations. Used fuel radioactivity calculations, the enhancements to the MCNPX code, and details of the MCNPX simulations are documented in the other appendices.« less

Review of heavy charged particle transport in MCNP6.2

NASA Astrophysics Data System (ADS)

Zieb, K.; Hughes, H. G.; James, M. R.; Xu, X. G.

2018-04-01

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. This paper discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models' theories are included as well.

Review of Heavy Charged Particle Transport in MCNP6.2

DOE PAGES

Zieb, Kristofer James Ekhart; Hughes, Henry Grady III; Xu, X. George; ...

2018-01-05

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. Here, this article discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

Estimation of dose rates at the entrance surface for exposure scenarios of total body irradiation using MCNPX code

NASA Astrophysics Data System (ADS)

Cunha, J. S.; Cavalcante, F. R.; Souza, S. O.; Souza, D. N.; Santos, W. S.; Carvalho Júnior, A. B.

2017-11-01

One of the main criteria that must be held in Total Body Irradiation (TBI) is the uniformity of dose in the body. In TBI procedures the certification that the prescribed doses are absorbed in organs is made with dosimeters positioned on the patient skin. In this work, we modelled TBI scenarios in the MCNPX code to estimate the entrance dose rate in the skin for comparison and validation of simulations with experimental measurements from literature. Dose rates were estimated simulating an ionization chamber laterally positioned on thorax, abdomen, leg and thigh. Four exposure scenarios were simulated: ionization chamber (S1), TBI room (S2), and patient represented by hybrid phantom (S3) and water stylized phantom (S4) in sitting posture. The posture of the patient in experimental work was better represented by S4 compared with hybrid phantom, and this led to minimum and maximum percentage differences of 1.31% and 6.25% to experimental measurements for thorax and thigh regions, respectively. As for all simulations reported here the percentage differences in the estimated dose rates were less than 10%, we considered that the obtained results are consistent with experimental measurements and the modelled scenarios are suitable to estimate the absorbed dose in organs during TBI procedure.

MCNP6 Fission Multiplicity with FMULT Card

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilcox, Trevor; Fensin, Michael Lorne; Hendricks, John S.

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

Performance revaluation of a N-type coaxial HPGe detector with front edges crystal using MCNPX.

PubMed

Azli, Tarek; Chaoui, Zine-El-Abidine

2015-03-01

The MCNPX code was used to determine the efficiency of a N-type HPGe detector after two decades of operation. Accounting for the roundedness of the crystal`s front edges and an inhomogeneous description of the detector's dead layers were shown to achieve better agreement between measurements and simulation efficiency determination. The calculations were experimentally verified using point sources in the energy range from 50keV to 1400keV, and an overall uncertainty less than 2% was achieved. In order to use the detector for different matrices and geometries in radioactivity, the suggested model was validated by changing the counting geometry and by using multi-gamma disc sources. The introduced simulation approach permitted the revaluation of the performance of an HPGe detector in comparison of its initial condition, which is a useful tool for precise determination of the thickness of the inhomogeneous dead layer. Copyright © 2014 Elsevier Ltd. All rights reserved.

Effect of the multiple scattering of electrons in Monte Carlo simulation of LINACS.

PubMed

Vilches, Manuel; García-Pareja, Salvador; Guerrero, Rafael; Anguiano, Marta; Lallena, Antonio M

2008-01-01

Results obtained from Monte Carlo simulations of the transport of electrons in thin slabs of dense material media and air slabs with different widths are analyzed. Various general purpose Monte Carlo codes have been used: PENELOPE, GEANT3, GEANT4, EGSNRC, MCNPX. Non-negligible differences between the angular and radial distributions after the slabs have been found. The effects of these differences on the depth doses measured in water are also discussed.

McStas event logger: Definition and applications

NASA Astrophysics Data System (ADS)

Bergbäck Knudsen, Erik; Bryndt Klinkby, Esben; Kjær Willendrup, Peter

2014-02-01

Functionality is added to the McStas neutron ray-tracing code, which allows individual neutron states before and after a scattering to be temporarily stored, and analysed. This logging mechanism has multiple uses, including studies of longitudinal intensity loss in neutron guides and guide coating design optimisations. Furthermore, the logging method enables the cold/thermal neutron induced gamma background along the guide to be calculated from the un-reflected neutron, using a recently developed MCNPX-McStas interface.

Preliminary investigation of parasitic radioisotope production using the LANL IPF secondary neutron flux

NASA Astrophysics Data System (ADS)

Engle, J. W.; Kelsey, C. T.; Bach, H.; Ballard, B. D.; Fassbender, M. E.; John, K. D.; Birnbaum, E. R.; Nortier, F. M.

2012-12-01

In order to ascertain the potential for radioisotope production and material science studies using the Isotope Production Facility at Los Alamos National Lab, a two-pronged investigation has been initiated. The Monte Carlo for Neutral Particles eXtended (MCNPX) code has been used in conjunction with the CINDER 90 burnup code to predict neutron flux energy distributions as a result of routine irradiations and to estimate yields of radioisotopes of interest for hypothetical irradiation conditions. A threshold foil activation experiment is planned to study the neutron flux using measured yields of radioisotopes, quantified by HPGe gamma spectroscopy, from representative nuclear reactions with known thresholds up to 50 MeV.

64Cu, a powerful positron emitter for immunoimaging and theranostic: Production via natZnO and natZnO-NPs.

PubMed

Karimi, Zahra; Sadeghi, Mahdi; Mataji-Kojouri, Naimeddin

2018-07-01

64 Cu is one of the most beneficial radionuclide that can be used as a theranostic agent in Positron Emission Tomography (PET) imaging. In this current work, 64 Cu was produced with zinc oxide nanoparticles ( nat ZnONPs) and zinc oxide powder ( nat ZnO) via the 64 Zn(n,p) 64 Cu reaction in Tehran Research Reactor (TRR) and the activity values were compared with each other. The theoretical activity of 64 Cu also was calculated with MCNPX-2.6 and the cross sections of this reaction were calculated by using TALYS-1.8, EMPIRE-3.2.2 and ALICE/ASH nuclear codes and were compared with experimental values. Transmission Electronic Microscopy (TEM), Scanning Electronic Microscopy (SEM) and X-Ray Diffraction (XRD) analysis were used for samples characterizations. From these results, it's concluded that 64 Cu activity value with nanoscale target was achieved more than the bulk state target and had a good adaptation with the MCNPX result. Copyright © 2018 Elsevier Ltd. All rights reserved.

Utilizing Radioisotope Power Systems for Human Lunar Exploration

NASA Technical Reports Server (NTRS)

Schreiner, Timothy M.

2005-01-01

The Vision for Space Exploration has a goal of sending crewed missions to the lunar surface as early as 2015 and no later than 2020. The use of nuclear power sources could aid in assisting crews in exploring the surface and performing In-Situ Resource Utilization (ISRU) activities. Radioisotope Power Systems (RPS) provide constant sources of electrical power and thermal energy for space applications. RPSs were carried on six of the crewed Apollo missions to power surface science packages, five of which still remain on the lunar surface. Future RPS designs may be able to play a more active role in supporting a long-term human presence. Due to its lower thermal and radiation output, the planned Stirling Radioisotope Generator (SRG) appears particularly attractive for manned applications. The MCNPX particle transport code has been used to model the current SRG design to assess its use in proximity with astronauts operating on the surface. Concepts of mobility and ISRU infrastructure were modeled using MCNPX to analyze the impact of RPSs on crewed mobility systems. Strategies for lowering the radiation dose were studied to determine methods of shielding the crew from the RPSs.

Development of self-interrogation neutron resonance densitometry (sinrd) to measure the fissile content in nuclear fuel

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne Marie

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: (1) SINRD provides absolute measurements of burnup independent of the operator's declaration. (2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3o from LWR spent LEU and MOX fuel. (3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. (4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. (5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

Simulation of a complete X-ray digital radiographic system for industrial applications.

PubMed

Nazemi, E; Rokrok, B; Movafeghi, A; Choopan Dastjerdi, M H

2018-05-19

Simulating X-ray images is of great importance in industry and medicine. Using such simulation permits us to optimize parameters which affect image's quality without the limitations of an experimental procedure. This study revolves around a novel methodology to simulate a complete industrial X-ray digital radiographic system composed of an X-ray tube and a computed radiography (CR) image plate using Monte Carlo N Particle eXtended (MCNPX) code. In the process of our research, an industrial X-ray tube with maximum voltage of 300 kV and current of 5 mA was simulated. A 3-layer uniform plate including a polymer overcoat layer, a phosphor layer and a polycarbonate backing layer was also defined and simulated as the CR imaging plate. To model the image formation in the image plate, at first the absorbed dose was calculated in each pixel inside the phosphor layer of CR imaging plate using the mesh tally in MCNPX code and then was converted to gray value using a mathematical relationship determined in a separate procedure. To validate the simulation results, an experimental setup was designed and the images of two step wedges created out of aluminum and steel were captured by the experiments and compared with the simulations. The results show that the simulated images are in good agreement with the experimental ones demonstrating the ability of the proposed methodology for simulating an industrial X-ray imaging system. Copyright © 2018 Elsevier Ltd. All rights reserved.

A new cubic phantom for PET/CT dosimetry: Experimental and Monte Carlo characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belinato, Walmir; Silva, Rogerio M.V.; Souza, Divanizia N.

In recent years, positron emission tomography (PET) associated with multidetector computed tomography (MDCT) has become a diagnostic technique widely disseminated to evaluate various malignant tumors and other diseases. However, during PET/CT examinations, the doses of ionizing radiation experienced by the internal organs of patients may be substantial. To study the doses involved in PET/CT procedures, a new cubic phantom of overlapping acrylic plates was developed and characterized. This phantom has a deposit for the placement of the fluorine-18 fluoro-2-deoxy-D-glucose ({sup 18}F-FDG) solution. There are also small holes near the faces for the insertion of optically stimulated luminescence dosimeters (OSLD). Themore » holes for OSLD are positioned at different distances from the {sup 18}F-FDG deposit. The experimental results were obtained in two PET/CT devices operating with different parameters. Differences in the absorbed doses were observed in OSLD measurements due to the non-orthogonal positioning of the detectors inside the phantom. This phantom was also evaluated using Monte Carlo simulations, with the MCNPX code. The phantom and the geometrical characteristics of the equipment were carefully modeled in the MCNPX code, in order to develop a new methodology form comparison of experimental and simulated results, as well as to allow the characterization of PET/CT equipments in Monte Carlo simulations. All results showed good agreement, proving that this new phantom may be applied for these experiments. (authors)« less

Accelerator shield design of KIPT neutron source facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Z.; Gohar, Y.

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generatedmore » by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary at less than 0.5-mrem/hr. The shield configuration and parameters of the accelerator building have been determined and are presented in this paper. (authors)« less

DHS Summary Report -- Robert Weldon

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weldon, Robert A.

This summer I worked on benchmarking the Lawrence Livermore National Laboratory fission multiplicity capability used in the Monte Carlo particle transport code MCNPX. This work involved running simulations and then comparing the simulation results with experimental experiments. Outlined in this paper is a brief description of the work completed this summer, skills and knowledge gained, and how the internship has impacted my planning for the future. Neutron multiplicity counting is a neutron detection technique that leverages the multiplicity emissions of neutrons from fission to identify various actinides in a lump of material. The identification of individual actinides in lumps ofmore » material crossing our boarders, especially U-235 and Pu-239, is a key component for maintaining the safety of the country from nuclear threats. Several multiplicity emission options from spontaneous and induced fission already existed in MCNPX 2.4.0. These options can be accessed through use of the 6th entry on the PHYS:N card. Lawrence Livermore National Laboratory (LLNL) developed a physics model for the simulation of neutron and gamma ray emission from fission and photofission that was included in MCNPX 2.7.B as an undocumented feature and then was documented in MCNPX 2.7.C. The LLNL multiplicity capability provided a different means for MCNPX to simulate neutron and gamma-ray distributions for neutron induced, spontaneous and photonuclear fission reactions. The original testing on the model for implementation into MCNPX was conducted by Gregg McKinney and John Hendricks. The model is an encapsulation of measured data of neutron multiplicity distributions from Gwin, Spencer, and Ingle, along with the data from Zucker and Holden. One of the founding principles of MCNPX was that it would have several redundant capabilities, providing the means of testing and including various physics packages. Though several multiplicity sampling methodologies already existed within MCNPX, the LLNL fission multiplicity was included to provide a separate capability for computing multiplicity as well as including several new features not already included in MCNPX. These new features include: (1) prompt gamma emission/multiplicity from neutron-induced fission; (2) neutron multiplicity and gamma emission/multiplicity from photofission; and (3) an option to enforce energy correlation for gamma neutron multiplicity emission. These new capabilities allow correlated signal detection for identifying presence of special nuclear material (SNM). Therefore, these new capabilities help meet the missions of the Domestic Nuclear Detection Office (DNDO), which is tasked with developing nuclear detection strategies for identifying potential radiological and nuclear threats, by providing new simulation capability for detection strategies that leverage the new available physics in the LLNL multiplicity capability. Two types of tests were accomplished this summer to test the default LLNL neutron multiplicity capability: neutron-induced fission tests and spontaneous fission tests. Both cases set the 6th entry on the PHYS:N card to 5 (i.e. use LLNL multiplicity). The neutron-induced fission tests utilized a simple 0.001 cm radius sphere where 0.0253 eV neutrons were released at the sphere center. Neutrons were forced to immediately collide in the sphere and release all progeny from the sphere, without further collision, using the LCA card, LCA 7j -2 (therefore density and size of the sphere were irrelevant). Enough particles were run to ensure that the average error of any specific multiplicity did not exceed 0.36%. Neutron-induced fission multiplicities were computed for U-233, U-235, Pu-239, and Pu-241. The spontaneous fission tests also used the same spherical geometry, except: (1) the LCA card was removed; (2) the density of the sphere was set to 0.001 g/cm3; and (3) instead of emitting a thermal neutron, the PAR keyword was set to PAR=SF. The purpose of the small density was to ensure that the spontaneous fission neutrons would not further interact and induce fissions (i.e. the mean free path greatly exceeded the size of the sphere). Enough particles were run to ensure that the average error of any specific spontaneous multiplicity did not exceed 0.23%. Spontaneous fission multiplicities were computed for U-238, Pu-238, Pu-240, Pu-242, Cm-242, and Cm-244. All of the computed results were compared against experimental results compiled by Holden at Brookhaven National Laboratory.« less

A new response matrix for a 6LiI scintillator BSS system

NASA Astrophysics Data System (ADS)

Lacerda, M. A. S.; Méndez-Villafañe, R.; Lorente, A.; Ibañez, S.; Gallego, E.; Vega-Carrillo, H. R.

2017-10-01

A new response matrix was calculated for a Bonner Sphere Spectrometer (BSS) with a 6 LiI(Eu) scintillator, using the Monte Carlo N-Particle radiation transport code MCNPX. Responses were calculated for 6 spheres and the bare detector, for energies varying from 1.059E(-9) MeV to 105.9 MeV, with 20 equal-log(E)-width bins per energy decade, totalizing 221 energy groups. A comparison was done among the responses obtained in this work and other published elsewhere, for the same detector model. The calculated response functions were inserted in the response input file of the MAXED code and used to unfold the total and direct neutron spectra generated by the 241Am-Be source of the Universidad Politécnica de Madrid (UPM). These spectra were compared with those obtained using the same unfolding code with the Mares and Schraube matrix response.

Analysis and evaluation for consumer goods containing NORM in Korea.

PubMed

Jang, Mee; Chung, Kun Ho; Lim, Jong Myoung; Ji, Young Yong; Kim, Chang Jong; Kang, Mun Ja

2017-08-01

We analyzed the consumer goods containing NORM by ICP-MS and evaluated the external dose. To evaluate the external dose, we assumed the small room model as irradiation scenario and calculated the specific effective dose rate using MCNPX code. The external doses for twenty goods are less than 1 mSv considering the specific effective dose rates and usage quantities. However, some of them have relatively high dose and the activity concentration limits are necessary as a screening tool. Copyright © 2017 Elsevier Ltd. All rights reserved.

Assessment of radiation-induced secondary cancer risk in the Brazilian population from left-sided breast-3D-CRT using MCNPX.

PubMed

Mendes, Bruno Melo; Trindade, Bruno Machado; Fonseca, Telma Cristina Ferreira; de Campos, Tarcisio Passos Ribeiro

2017-12-01

The aim of this work was to simulate a 6MV conventional breast 3D conformational radiation therapy (3D-CRT) with physical wedges (50 Gy/25#) in the left breast, calculate the mean absorbed dose in the body organs using robust models and computational tools and estimate the secondary cancer-incidence risk to the Brazilian population. The VW female phantom was used in the simulations. Planning target volume (PTV) was defined in the left breast. The 6MV parallel-opposed fields breast-radiotherapy (RT) protocol was simulated with MCNPx code. The absorbed doses were evaluated in all the organs. The secondary cancer-incidence risk induced by radiotherapy was calculated for different age groups according to the BEIR VII methodology. RT quality indexes indicated that the protocol was properly simulated. Significant absorbed dose values in red bone marrow, RBM (0.8 Gy) and stomach (0.6 Gy) were observed. The contralateral breast presented the highest risk of incidence of a secondary cancer followed by leukaemia, lung and stomach. The risk of a secondary cancer-incidence by breast-RT, for the Brazilian population, ranged between 2.2-1.7% and 0.6-0.4%. RBM and stomach, usually not considered as OAR, presented high second cancer incidence risks of 0.5-0.3% and 0.4-0.1%, respectively. This study may be helpful for breast-RT risk/benefit assessment. Advances in knowledge: MCNPX-dosimetry was able to provide the scatter radiation and dose for all body organs in conventional breast-RT. It was found a relevant risk up to 2.2% of induced-cancer from breast-RT, considering the whole thorax organs and Brazilian cancer-incidence.

Intercomparison of Monte Carlo radiation transport codes to model TEPC response in low-energy neutron and gamma-ray fields.

PubMed

Ali, F; Waker, A J; Waller, E J

2014-10-01

Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Calculations vs. measurements of remnant dose rates for SNS spent structures

NASA Astrophysics Data System (ADS)

Popova, I. I.; Gallmeier, F. X.; Trotter, S.; Dayton, M.

2018-06-01

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction. Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.

Calculations vs. measurements of remnant dose rates for SNS spent structures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popova, Irina I.; Gallmeier, Franz X.; Trotter, Steven M.

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction.more » Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.« less

Investigating a cyclotron HM-30 based neutron source for BNCT of deep-seated tumors by using shifting method

NASA Astrophysics Data System (ADS)

Suharyana; Riyatun; Octaviana, E. F.

2016-11-01

We have successfully proposed a simulation of a neutron beam-shaping assembly using MCNPX Code. This simulation study deals with designing a compact, optimized, and geometrically simple beam shaping assembly for a neutron source based on a proton cyclotron for BNCT purpose. Shifting method was applied in order to lower the fast neutron energy to the epithermal energy range by choosing appropriate materials. Based on a set of MCNPX simulations, it has been found that the best materials for beam shaping assembly are 3 cm Ni layered with 7 cm Pb as the reflector and 13 cm AlF3 the moderator. Our proposed beam shaping assembly configuration satisfies 2 of 5 of the IAEA criteria, namely the epithermal neutron flux 1.25 × 109 n.cm-2 s-1 and the gamma dose over the epithermal neutron flux is 0.18×10 -13 Gy.cm 2 n -1. However, the ratio of the fast neutron dose rate over neutron epithermal flux is still too high. We recommended that the shifting method must be accompanied by the filter method to reduce the fast neutron flux.

CT-based MCNPX dose calculations for gynecology brachytherapy employing a Henschke applicator

NASA Astrophysics Data System (ADS)

Yu, Pei-Chieh; Nien, Hsin-Hua; Tung, Chuan-Jong; Lee, Hsing-Yi; Lee, Chung-Chi; Wu, Ching-Jung; Chao, Tsi-Chian

2017-11-01

The purpose of this study is to investigate the dose perturbation caused by the metal ovoid structures of a Henschke applicator using Monte Carlo simulation in a realistic phantom. The Henschke applicator has been widely used for gynecologic patients treated by brachytherapy in Taiwan. However, the commercial brachytherapy planning system (BPS) did not properly evaluate the dose perturbation caused by its metal ovoid structures. In this study, Monte Carlo N-Particle Transport Code eXtended (MCNPX) was used to evaluate the brachytherapy dose distribution of a Henschke applicator embedded in a Plastic water phantom and a heterogeneous patient computed tomography (CT) phantom. The dose comparison between the MC simulations and film measurements for a Plastic water phantom with Henschke applicator were in good agreement. However, MC dose with the Henschke applicator showed significant deviation (-80.6%±7.5%) from those without Henschke applicator. Furthermore, the dose discrepancy in the heterogeneous patient CT phantom and Plastic water phantom CT geometries with Henschke applicator showed 0 to -26.7% dose discrepancy (-8.9%±13.8%). This study demonstrates that the metal ovoid structures of Henschke applicator cannot be disregard in brachytherapy dose calculation.

MCNP SIMULATION OF THE HP(10) ENERGY RESPONSE OF A BRAZILIAN TLD ALBEDO NEUTRON INDIVIDUAL DOSEMETER, FROM THERMAL TO 20 MeV.

PubMed

Freitas, B M; Martins, M M; Pereira, W W; da Silva, A X; Mauricio, C L P

2016-09-01

The Brazilian Instituto de Radioproteção e Dosimetria (IRD) runs a neutron individual monitoring system with a home-made TLD albedo dosemeter. It has already been characterised and calibrated in some reference fields. However, the complete energy response of this dosemeter is not known, and the calibration factors for all monitored workplace neutron fields are difficult to be obtained experimentally. Therefore, to overcome such difficulties, Monte Carlo simulations have been used. This paper describes the simulation of the HP(10) neutron response of the IRD TLD albedo dosemeter using the MCNPX transport code, for energies from thermal to 20 MeV. The validation of the MCNPX modelling is done comparing the simulated results with the experimental measurements for ISO standard neutron fields of (241)Am-Be, (252)Cf, (241)Am-B and (252)Cf(D2O) and also for (241)Am-Be source moderated with paraffin and silicone. Bare (252)Cf are used for normalisation. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Comparison of Measured Leakage Current Distributions with Calculated Damage Energy Distributions in HgCdTe

NASA Technical Reports Server (NTRS)

Marshall, C. J.; Ladbury, R.; Marshall, P. W.; Reed, R. A.; Howe, C.; Weller, B.; Mendenhall, M.; Waczynski, A.; Jordan, T. M.; Fodness, B.

2006-01-01

This paper presents a combined Monte Carlo and analytic approach to the calculation of the pixel-to-pixel distribution of proton-induced damage in a HgCdTe sensor array and compares the results to measured dark current distributions after damage by 63 MeV protons. The moments of the Coulombic, nuclear elastic and nuclear inelastic damage distribution were extracted from Monte Carlo simulations and combined to form a damage distribution using the analytic techniques first described in [I]. The calculations show that the high energy recoils from the nuclear inelastic reactions (calculated using the Monte Car10 code MCNPX [2]) produce a pronounced skewing of the damage energy distribution. The nuclear elastic component (also calculated using the MCNPX) has a negligible effect on the shape of the damage distribution. The Coulombic contribution was calculated using MRED [3,4], a Geant4 [4,5] application. The comparison with the dark current distribution strongly suggests that mechanisms which are not linearly correlated with nonionizing damage produced according to collision kinematics are responsible for the observed dark current increases. This has important implications for the process of predicting the on-orbit dark current response of the HgCdTe sensor array.

Comparison between experimental data and Monte-Carlo simulations of neutron production in spallation reactions of 0.7-1.5 GeV protons on a thick, lead target

NASA Astrophysics Data System (ADS)

Krása, A.; Majerle, M.; Krízek, F.; Wagner, V.; Kugler, A.; Svoboda, O.; Henzl, V.; Henzlová, D.; Adam, J.; Caloun, P.; Kalinnikov, V. G.; Krivopustov, M. I.; Stegailov, V. I.; Tsoupko-Sitnikov, V. M.

2006-05-01

Relativistic protons with energies 0.7-1.5 GeV interacting with a thick, cylindrical, lead target, surrounded by a uranium blanket and a polyethylene moderator, produced spallation neutrons. The spatial and energetic distributions of the produced neutron field were measured by the Activation Analysis Method using Al, Au, Bi, and Co radio-chemical sensors. The experimental yields of isotopes induced in the sensors were compared with Monte-Carlo calculations performed with the MCNPX 2.4.0 code.

MicroCT parameters for multimaterial elements assessment

NASA Astrophysics Data System (ADS)

de Araújo, Olga M. O.; Silva Bastos, Jaqueline; Machado, Alessandra S.; dos Santos, Thaís M. P.; Ferreira, Cintia G.; Rosifini Alves Claro, Ana Paula; Lopes, Ricardo T.

2018-03-01

Microtomography is a non-destructive testing technique for quantitative and qualitative analysis. The investigation of multimaterial elements with great difference of density can result in artifacts that degrade image quality depending on combination of additional filter. The aim of this study is the selection of parameters most appropriate for analysis of bone tissue with metallic implant. The results show the simulation with MCNPX code for the distribution of energy without additional filter, with use of aluminum, copper and brass filters and their respective reconstructed images showing the importance of the choice of these parameters in image acquisition process on computed microtomography.

Electron Accelerator Shielding Design of KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Zhaopeng; Gohar, Yousry

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biologicalmore » dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, similar to 0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper. Copyright (C) 2016, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.« less

Simulation of Charge Collection in Diamond Detectors Irradiated with Deuteron-Triton Neutron Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milocco, Alberto; Trkov, Andrej; Pillon, Mario

2011-12-13

Diamond-based neutron spectrometers exhibit outstanding properties such as radiation hardness, low sensitivity to gamma rays, fast response and high-energy resolution. They represent a very promising application of diamonds for plasma diagnostics in fusion devices. The measured pulse height spectrum is obtained from the collection of helium and beryllium ions produced by the reactions on {sup 12}C. An original code is developed to simulate the production and the transport of charged particles inside the diamond detector. The ion transport methodology is based on the well-known TRIM code. The reactions of interest are triggered using the ENDF/B-VII.0 nuclear data for the neutronmore » interactions on carbon. The model is implemented in the TALLYX subroutine of the MCNP5 and MCNPX codes. Measurements with diamond detectors in a {approx}14 MeV neutron field have been performed at the FNG (Rome, Italy) and IRMM (Geel, Belgium) facilities. The comparison of the experimental data with the simulations validates the proposed model.« less

Design Analysis of SNS Target StationBiological Shielding Monoligh with Proton Power Uprate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bekar, Kursat B.; Ibrahim, Ahmad M.

2017-05-01

This report documents the analysis of the dose rate in the experiment area outside the Spallation Neutron Source (SNS) target station shielding monolith with proton beam energy of 1.3 GeV. The analysis implemented a coupled three dimensional (3D)/two dimensional (2D) approach that used both the Monte Carlo N-Particle Extended (MCNPX) 3D Monte Carlo code and the Discrete Ordinates Transport (DORT) two dimensional deterministic code. The analysis with proton beam energy of 1.3 GeV showed that the dose rate in continuously occupied areas on the lateral surface outside the SNS target station shielding monolith is less than 0.25 mrem/h, which compliesmore » with the SNS facility design objective. However, the methods and codes used in this analysis are out of date and unsupported, and the 2D approximation of the target shielding monolith does not accurately represent the geometry. We recommend that this analysis is updated with modern codes and libraries such as ADVANTG or SHIFT. These codes have demonstrated very high efficiency in performing full 3D radiation shielding analyses of similar and even more difficult problems.« less

SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cho, S; Shin, E H; Kim, J

2015-06-15

Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using themore » production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shin, H; Yoon, D; Jung, J

Purpose: The purpose of this study is to suggest a tumor monitoring technique using prompt gamma rays emitted during the reaction between an antiproton and a boron particle, and to verify the increase of the therapeutic effectiveness of the antiproton boron fusion therapy using Monte Carlo simulation code. Methods: We acquired the percentage depth dose of the antiproton beam from a water phantom with and without three boron uptake regions (region A, B, and C) using F6 tally of MCNPX. The tomographic image was reconstructed using prompt gamma ray events from the reaction between the antiproton and boron during themore » treatment from 32 projections (reconstruction algorithm: MLEM). For the image reconstruction, we were performed using a 80 × 80 pixel matrix with a pixel size of 5 mm. The energy window was set as a 10 % energy window. Results: The prompt gamma ray peak for imaging was observed at 719 keV in the energy spectrum using the F8 tally fuction (energy deposition tally) of the MCNPX code. The tomographic image shows that the boron uptake regions were successfully identified from the simulation results. In terms of the receiver operating characteristic curve analysis, the area under the curve values were 0.647 (region A), 0.679 (region B), and 0.632 (region C). The SNR values increased as the tumor diameter increased. The CNR indicated the relative signal intensity within different regions. The CNR values also increased as the different of BURs diamter increased. Conclusion: We confirmed the feasibility of tumor monitoring during the antiproton therapy as well as the superior therapeutic effect of the antiproton boron fusion therapy. This result can be beneficial for the development of a more accurate particle therapy.« less

High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yasin, Zafar, E-mail: zafar.yasin@eli-np.ro; Matei, Catalin; Ur, Calin A.

The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKAmore » and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.« less

Detector photon response and absorbed dose and their applications to rapid triage techniques

NASA Astrophysics Data System (ADS)

Voss, Shannon Prentice

As radiation specialists, one of our primary objectives in the Navy is protecting people and the environment from the effects of ionizing and non-ionizing radiation. Focusing on radiological dispersal devices (RDD) will provide increased personnel protection as well as optimize emergency response assets for the general public. An attack involving an RDD has been of particular concern because it is intended to spread contamination over a wide area and cause massive panic within the general population. A rapid method of triage will be necessary to segregate the unexposed and slightly exposed from those needing immediate medical treatment. Because of the aerosol dispersal of the radioactive material, inhalation of the radioactive material may be the primary exposure route. The primary radionuclides likely to be used in a RDD attack are Co-60, Cs-137, Ir-192, Sr-90 and Am-241. Through the use of a MAX phantom along with a few Simulink MATLAB programs, a good anthropomorphic phantom was created for use in MCNPX simulations that would provide organ doses from internally deposited radionuclides. Ludlum model 44-9 and 44-2 detectors were used to verify the simulated dose from the MCNPX code. Based on the results, acute dose rate limits were developed for emergency response personnel that would assist in patient triage.

Comparison of Measured Dark Current Distributions with Calculated Damage Energy Distributions in HgCdTe

NASA Technical Reports Server (NTRS)

Marshall, C. J.; Marshall, P. W.; Howe, C. L.; Reed, R. A.; Weller, R. A.; Mendenhall, M.; Waczynski, A.; Ladbury, R.; Jordan, T. M.

2007-01-01

This paper presents a combined Monte Carlo and analytic approach to the calculation of the pixel-to-pixel distribution of proton-induced damage in a HgCdTe sensor array and compares the results to measured dark current distributions after damage by 63 MeV protons. The moments of the Coulombic, nuclear elastic and nuclear inelastic damage distributions were extracted from Monte Carlo simulations and combined to form a damage distribution using the analytic techniques first described in [1]. The calculations show that the high energy recoils from the nuclear inelastic reactions (calculated using the Monte Carlo code MCNPX [2]) produce a pronounced skewing of the damage energy distribution. While the nuclear elastic component (also calculated using the MCNPX) contributes only a small fraction of the total nonionizing damage energy, its inclusion in the shape of the damage across the array is significant. The Coulombic contribution was calculated using MRED [3-5], a Geant4 [4,6] application. The comparison with the dark current distribution strongly suggests that mechanisms which are not linearly correlated with nonionizing damage produced according to collision kinematics are responsible for the observed dark current increases. This has important implications for the process of predicting the on-orbit dark current response of the HgCdTe sensor array.

Active-Interrogation Measurements of Induced-Fission Neutrons from Low-Enriched Uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. L. Dolan; M. J. Marcath; M. Flaska

2012-07-01

Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a sample’s mass and enrichment. Using MCNPX-PoliMi, amore » system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5” by 5” EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.« less

Monte Carlo simulations of adult and pediatric computed tomography exams: Validation studies of organ doses with physical phantoms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, Daniel J.; Lee, Choonsik; Tien, Christopher

2013-01-15

Purpose: To validate the accuracy of a Monte Carlo source model of the Siemens SOMATOM Sensation 16 CT scanner using organ doses measured in physical anthropomorphic phantoms. Methods: The x-ray output of the Siemens SOMATOM Sensation 16 multidetector CT scanner was simulated within the Monte Carlo radiation transport code, MCNPX version 2.6. The resulting source model was able to perform various simulated axial and helical computed tomographic (CT) scans of varying scan parameters, including beam energy, filtration, pitch, and beam collimation. Two custom-built anthropomorphic phantoms were used to take dose measurements on the CT scanner: an adult male and amore » 9-month-old. The adult male is a physical replica of University of Florida reference adult male hybrid computational phantom, while the 9-month-old is a replica of University of Florida Series B 9-month-old voxel computational phantom. Each phantom underwent a series of axial and helical CT scans, during which organ doses were measured using fiber-optic coupled plastic scintillator dosimeters developed at University of Florida. The physical setup was reproduced and simulated in MCNPX using the CT source model and the computational phantoms upon which the anthropomorphic phantoms were constructed. Average organ doses were then calculated based upon these MCNPX results. Results: For all CT scans, good agreement was seen between measured and simulated organ doses. For the adult male, the percent differences were within 16% for axial scans, and within 18% for helical scans. For the 9-month-old, the percent differences were all within 15% for both the axial and helical scans. These results are comparable to previously published validation studies using GE scanners and commercially available anthropomorphic phantoms. Conclusions: Overall results of this study show that the Monte Carlo source model can be used to accurately and reliably calculate organ doses for patients undergoing a variety of axial or helical CT examinations on the Siemens SOMATOM Sensation 16 scanner.« less

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

Remanent Activation in the Mini-SHINE Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Micklich, Bradley J.

2015-04-16

Argonne National Laboratory is assisting SHINE Medical Technologies in developing a domestic source of the medical isotope 99Mo through the fission of low-enrichment uranium in a uranyl sulfate solution. In Phase 2 of these experiments, electrons from a linear accelerator create neutrons by interacting in a depleted uranium target, and these neutrons are used to irradiate the solution. The resulting neutron and photon radiation activates the target, the solution vessels, and a shielded cell that surrounds the experimental apparatus. When the experimental campaign is complete, the target must be removed into a shielding cask, and the experimental components must bemore » disassembled. The radiation transport code MCNPX and the transmutation code CINDER were used to calculate the radionuclide inventories of the solution, the target assembly, and the shielded cell, and to determine the dose rates and shielding requirements for selected removal scenarios for the target assembly and the solution vessels.« less

A multi-detector neutron spectrometer with nearly isotropic response for environmental and workplace monitoring

NASA Astrophysics Data System (ADS)

Gómez-Ros, J. M.; Bedogni, R.; Moraleda, M.; Delgado, A.; Romero, A.; Esposito, A.

2010-01-01

This communication describes an improved design for a neutron spectrometer consisting of 6Li thermoluminescent dosemeters located at selected positions within a single moderating polyethylene sphere. The spatial arrangement of the dosemeters has been designed using the MCNPX Monte Carlo code to calculate the response matrix for 56 log-equidistant energies from 10 -9 to 100 MeV, looking for a configuration that permits to obtain a nearly isotropic response for neutrons in the energy range from thermal to 20 MeV. The feasibility of the proposed spectrometer and the isotropy of its response have been evaluated by simulating exposures to different reference and workplace neutron fields. The FRUIT code has been used for unfolding purposes. The results of the simulations as well as the experimental tests confirm the suitability of the prototype for environmental and workplace monitoring applications.

Using computational modeling to compare X-ray tube Practical Peak Voltage for Dental Radiology

NASA Astrophysics Data System (ADS)

Holanda Cassiano, Deisemar; Arruda Correa, Samanda Cristine; de Souza, Edmilson Monteiro; da Silva, Ademir Xaxier; Pereira Peixoto, José Guilherme; Tadeu Lopes, Ricardo

2014-02-01

The Practical Peak Voltage-PPV has been adopted to measure the voltage applied to an X-ray tube. The PPV was recommended by the IEC document and accepted and published in the TRS no. 457 code of practice. The PPV is defined and applied to all forms of waves and is related to the spectral distribution of X-rays and to the properties of the image. The calibration of X-rays tubes was performed using the MCNPX Monte Carlo code. An X-ray tube for Dental Radiology (operated from a single phase power supply) and an X-ray tube used as a reference (supplied from a constant potential power supply) were used in simulations across the energy range of interest of 40 kV to 100 kV. Results obtained indicated a linear relationship between the tubes involved.

Correlated prompt fission data in transport simulations

DOE PAGES

Talou, P.; Vogt, R.; Randrup, J.; ...

2018-01-24

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Correlated prompt fission data in transport simulations

NASA Astrophysics Data System (ADS)

Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

2018-01-01

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. This review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Finally, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.

Correlated prompt fission data in transport simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talou, P.; Vogt, R.; Randrup, J.

Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n -n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ raysmore » from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation, safeguards, nuclear energy, and defense programs. Here, this review provides an overview of the topic, starting from theoretical considerations of the fission process, with a focus on correlated signatures. It then explores the status of experimental correlated fission data and current efforts to address some of the known shortcomings. Numerical simulations employing the FREYA and CGMF codes are compared to experimental data for a wide range of correlated fission quantities. The inclusion of those codes into the MCNP6.2 and MCNPX - PoliMi transport codes is described and discussed in the context of relevant applications. The accuracy of the model predictions and their sensitivity to model assumptions and input parameters are discussed. Lastly, a series of important experimental and theoretical questions that remain unanswered are presented, suggesting a renewed effort to address these shortcomings.« less

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

PubMed

Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

2018-01-17

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers. © The Author(s) 2018. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Radiation induced currents in mineral-insulated cables and in pick-up coils: model calculations and experimental verification in the BR1 reactor

NASA Astrophysics Data System (ADS)

Vermeeren, Ludo; Leysen, Willem; Brichard, Benoit

2018-01-01

Mineral-insulated (MI) cables and Low-Temperature Co-fired Ceramic (LTCC) magnetic pick-up coils are intended to be installed in various position in ITER. The severe ITER nuclear radiation field is expected to lead to induced currents that could perturb diagnostic measurements. In order to assess this problem and to find mitigation strategies models were developed for the calculation of neutron-and gamma-induced currents in MI cables and in LTCC coils. The models are based on calculations with the MCNPX code, combined with a dedicated model for the drift of electrons stopped in the insulator. The gamma induced currents can be easily calculated with a single coupled photon-electron MCNPX calculation. The prompt neutron induced currents requires only a single coupled neutron-photon-electron MCNPX run. The various delayed neutron contributions require a careful analysis of all possibly relevant neutron-induced reaction paths and a combination of different types of MCNPX calculations. The models were applied for a specific twin-core copper MI cable, for one quad-core copper cable and for silver conductor LTCC coils (one with silver ground plates in order to reduce the currents and one without such silver ground plates). Calculations were performed for irradiation conditions (neutron and gamma spectra and fluxes) in relevant positions in ITER and in the Y3 irradiation channel of the BR1 reactor at SCK•CEN, in which an irradiation test of these four test devices was carried out afterwards. We will present the basic elements of the models and show the results of all relevant partial currents (gamma and neutron induced, prompt and various delayed currents) in BR1-Y3 conditions. Experimental data will be shown and analysed in terms of the respective contributions. The tests were performed at reactor powers of 350 kW and 1 MW, leading to thermal neutron fluxes of 1E11 n/cm2s and 3E11 n/cm2s, respectively. The corresponding total radiation induced currents are ranging from 1 to 7 nA only, putting a challenge on the acquisition system and on the data analysis. The detailed experimental results will be compared with the corresponding values predicted by the model. The overall agreement between the experimental data and the model predictions is fairly good, with very consistent data for the main delayed current components, while the lower amplitude delayed currents and some of the prompt contributions show some minor discrepancies.

Design and feasibility of a multi-detector neutron spectrometer for radiation protection applications based on thermoluminescent 6LiF:Ti,Mg (TLD-600) detectors

NASA Astrophysics Data System (ADS)

Lis, M.; Gómez-Ros, J. M.; Bedogni, R.; Delgado, A.

2008-01-01

The design of a neutron detector with spectrometric capability based on thermoluminescent (TL) 6LiF:Ti,Mg (TLD-600) dosimeters located along three perpendicular axis within a single polyethylene (PE) sphere has been analyzed. The neutron response functions have been calculated in the energy range from 10 -8 to 100 MeV with the Monte Carlo (MC) code MCNPX 2.5 and their shape and behaviour have been used to discuss a suitable configuration for an actual instrument. The feasibility of such a device has been preliminary evaluated by the simulation of exposure to 241Am-Be, bare 252Cf and Fe-PE moderated 252Cf sources. The expected accuracy in the evaluation of energy quantities has been evaluated using the unfolding code FRUIT. The obtained results together with additional calculations performed using MAXED and GRAVEL codes show the spectrometric capability of the proposed design for radiation protection applications, especially in the range 1 keV-20 MeV.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahanani, Nursinta Adi, E-mail: sintaadi@batan.go.id; Natsir, Khairina, E-mail: sintaadi@batan.go.id; Hartini, Entin, E-mail: sintaadi@batan.go.id

Data processing software packages such as VSOP and MCNPX are softwares that has been scientifically proven and complete. The result of VSOP and MCNPX are huge and complex text files. In the analyze process, user need additional processing like Microsoft Excel to show informative result. This research develop an user interface software for output of VSOP and MCNPX. VSOP program output is used to support neutronic analysis and MCNPX program output is used to support burn-up analysis. Software development using iterative development methods which allow for revision and addition of features according to user needs. Processing time with this softwaremore » 500 times faster than with conventional methods using Microsoft Excel. PYTHON is used as a programming language, because Python is available for all major operating systems: Windows, Linux/Unix, OS/2, Mac, Amiga, among others. Values that support neutronic analysis are k-eff, burn-up and mass Pu{sup 239} and Pu{sup 241}. Burn-up analysis used the mass inventory values of actinide (Thorium, Plutonium, Neptunium and Uranium). Values are visualized in graphical shape to support analysis.« less

Development and validation of MCNPX-based Monte Carlo treatment plan verification system

PubMed Central

Jabbari, Iraj; Monadi, Shahram

2015-01-01

A Monte Carlo treatment plan verification (MCTPV) system was developed for clinical treatment plan verification (TPV), especially for the conformal and intensity-modulated radiotherapy (IMRT) plans. In the MCTPV, the MCNPX code was used for particle transport through the accelerator head and the patient body. MCTPV has an interface with TiGRT planning system and reads the information which is needed for Monte Carlo calculation transferred in digital image communications in medicine-radiation therapy (DICOM-RT) format. In MCTPV several methods were applied in order to reduce the simulation time. The relative dose distribution of a clinical prostate conformal plan calculated by the MCTPV was compared with that of TiGRT planning system. The results showed well implementation of the beams configuration and patient information in this system. For quantitative evaluation of MCTPV a two-dimensional (2D) diode array (MapCHECK2) and gamma index analysis were used. The gamma passing rate (3%/3 mm) of an IMRT plan was found to be 98.5% for total beams. Also, comparison of the measured and Monte Carlo calculated doses at several points inside an inhomogeneous phantom for 6- and 18-MV photon beams showed a good agreement (within 1.5%). The accuracy and timing results of MCTPV showed that MCTPV could be used very efficiently for additional assessment of complicated plans such as IMRT plan. PMID:26170554

Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

DOE PAGES

Fensin, M. L.; Galloway, J. D.; James, M. R.

2015-04-11

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

Modeling study of a proposed field calibration source using K-40 and high-Z targets for sodium iodide detectors

DOE PAGES

Rogers, Jeremy; Marianno, Craig; Kallenbach, Gene; ...

2016-06-01

Calibration sources based on the primordial isotope potassium-40 ( 40K) have reduced controls on the source’s activity due to its terrestrial ubiquity and very low specific activity. Potassium–40’s beta emissions and 1,460.8 keV gamma ray can be used to induce K-shell fluorescence x rays in high-Z metals between 60 and 80 keV. A gamma ray calibration source that uses potassium chloride salt and a high-Z metal to create a two-point calibration for a sodium iodide field gamma spectroscopy instrument is thus proposed. The calibration source was designed in collaboration with the Sandia National Laboratory using the Monte Carlo N-Particle eXtendedmore » (MCNPX) transport code. Two methods of x-ray production were explored. First, a thin high-Z layer (HZL) was interposed between the detector and the potassium chloride-urethane source matrix. Second, bismuth metal powder was homogeneously mixed with a urethane binding agent to form a potassium chloride-bismuth matrix (KBM). The bismuth-based source was selected as the development model because it is inexpensive, nontoxic, and outperforms the high-Z layer method in simulation. As a result, based on the MCNPX studies, sealing a mixture of bismuth powder and potassium chloride into a thin plastic case could provide a light, inexpensive field calibration source.« less

Narrow beam neutron dosimetry.

PubMed

Ferenci, M Sutton

2004-01-01

Organ and effective doses have been estimated for male and female anthropomorphic mathematical models exposed to monoenergetic narrow beams of neutrons with energies from 10(-11) to 1000 MeV. Calculations were performed for anterior-posterior, posterior-anterior, left-lateral and right-lateral irradiation geometries. The beam diameter used in the calculations was 7.62 cm and the phantoms were irradiated at a height of 1 m above the ground. This geometry was chosen to simulate an accidental scenario (a worker walking through the beam) at Flight Path 30 Left (FP30L) of the Weapons Neutron Research (WNR) Facility at Los Alamos National Laboratory. The calculations were carried out using the Monte Carlo transport code MCNPX 2.5c.

MCNPX CALCULATIONS OF SPECIFIC ABSORBED FRACTIONS IN SOME ORGANS OF THE HUMAN BODY DUE TO APPLICATION OF 133Xe, 99mTc and 81mKr RADIONUCLIDES.

PubMed

Jovanovic, Z; Krstic, D; Nikezic, D; Ros, J M Gomez; Ferrari, P

2018-03-01

Monte Carlo simulations were performed to evaluate treatment doses with wide spread used radionuclides 133Xe, 99mTc and 81mKr. These different radionuclides are used in perfusion or ventilation examinations in nuclear medicine and as indicators for cardiovascular and pulmonary diseases. The objective of this work was to estimate the specific absorbed fractions in surrounding organs and tissues, when these radionuclides are incorporated in the lungs. For this purpose a voxel thorax model has been developed and compared with the ORNL phantom. All calculations and simulations were performed by means of the MCNP5/X code.

Neutron H*(10) estimation and measurements around 18MV linac.

PubMed

Cerón Ramírez, Pablo Víctor; Díaz Góngora, José Antonio Irán; Paredes Gutiérrez, Lydia Concepción; Rivera Montalvo, Teodoro; Vega Carrillo, Héctor René

2016-11-01

Thermoluminescent dosimetry, analytical techniques and Monte Carlo calculations were used to estimate the dose of neutron radiation in a treatment room with a linear electron accelerator of 18MV. Measurements were carried out through neutron ambient dose monitors which include pairs of thermoluminescent dosimeters TLD 600 ( 6 LiF: Mg, Ti) and TLD 700 ( 7 LiF: Mg, Ti), which were placed inside a paraffin spheres. The measurements has allowed to use NCRP 151 equations, these expressions are useful to find relevant dosimetric quantities. In addition, photoneutrons produced by linac head were calculated through MCNPX code taking into account the geometry and composition of the linac head principal parts. Copyright © 2016 Elsevier Ltd. All rights reserved.

Measurement of thermal neutrons reflection coefficients for two-layer reflectors.

PubMed

Azimkhani, S; Zolfagharpour, F; Ziaie, F

2018-05-01

In this research, thermal neutrons albedo coefficients and relative number of excess counts have been measured experimentally for different thicknesses of two-layer reflectors by using 241 Am-Be neutron source (5.2Ci) and BF 3 detector. Our used reflectors consist of two-layer which are combinations of water, graphite, polyethylene, and lead materials. Experimental results reveal that thermal neutron reflection coefficients slightly increased by addition of the second layer. The maximum value of growth for thermal neutrons albedo is obtained for lead-polyethylene compound (0.72 ± 0.01). Also, there is suitable agreement between the experimental values and simulation results by using MCNPX code. Copyright © 2018 Elsevier Ltd. All rights reserved.

New thermal neutron calibration channel at LNMRI/IRD

NASA Astrophysics Data System (ADS)

Astuto, A.; Patrão, K. C. S.; Fonseca, E. S.; Pereira, W. W.; Lopes, R. T.

2016-07-01

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241Am-Be sources with 0.6 TBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence in the central chamber and H*(10) at 50 cm from the front face with the polyethylene filter.

Spallation yield of neutrons produced in thick lead target bombarded with 250 MeV protons

NASA Astrophysics Data System (ADS)

Chen, L.; Ma, F.; Zhanga, X. Y.; Ju, Y. Q.; Zhang, H. B.; Ge, H. L.; Wang, J. G.; Zhou, B.; Li, Y. Y.; Xu, X. W.; Luo, P.; Yang, L.; Zhang, Y. B.; Li, J. Y.; Xu, J. K.; Liang, T. J.; Wang, S. L.; Yang, Y. W.; Gu, L.

2015-01-01

The neutron yield from thick target of Pb irradiated with 250 MeV protons has been studied experimentally. The neutron production was measured with the water-bath gold method. The thermal neutron distributions in the water were determined according to the measured activities of Au foils. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data. It was found out that the Au foils with cadmium cover significantly changed the spacial distribution of the thermal neutron field. The corrected neutron yield was deduced to be 2.23 ± 0.19 n/proton by considering the influence of the Cd cover on the thermal neutron flux.

Extraterrestrial Studies Using Nuclear Interactions

NASA Technical Reports Server (NTRS)

Reedy, Robert C.

2003-01-01

Cosmogenic nuclides were used to study the recent histories of the aubrite Norton County and the pallasite Brenham using calculated production rates. Calculations were done of the rates for making cosmogenic noble-gas isotopes in the Jovian satellite Europa by the interactions of galactic cosmic rays and especially trapped Jovian protons. Cross sections for the production of cosmogenic nuclides were reported and plans made to measure additional cross sections. A new code, MCNPX, was used to numerically simulate the interactions of cosmic rays with matter and the subsequent production of cosmogenic nuclides. A review was written about studies of extraterrestrial matter using cosmogenic radionuclides. Several other projects were done. Results are reviewed here with references to my recent publications for details.

Benchmark Analysis of Pion Contribution from Galactic Cosmic Rays

NASA Technical Reports Server (NTRS)

Aghara, Sukesh K.; Blattnig, Steve R.; Norbury, John W.; Singleterry, Robert C., Jr.

2008-01-01

Shielding strategies for extended stays in space must include a comprehensive resolution of the secondary radiation environment inside the spacecraft induced by the primary, external radiation. The distribution of absorbed dose and dose equivalent is a function of the type, energy and population of these secondary products. A systematic verification and validation effort is underway for HZETRN, which is a space radiation transport code currently used by NASA. It performs neutron, proton and heavy ion transport explicitly, but it does not take into account the production and transport of mesons, photons and leptons. The question naturally arises as to what is the contribution of these particles to space radiation. The pion has a production kinetic energy threshold of about 280 MeV. The Galactic cosmic ray (GCR) spectra, coincidentally, reaches flux maxima in the hundreds of MeV range, corresponding to the pion production threshold. We present results from the Monte Carlo code MCNPX, showing the effect of lepton and meson physics when produced and transported explicitly in a GCR environment.

Radionuclide production and dose rate estimation during the commissioning of the W-Ta spallation target

NASA Astrophysics Data System (ADS)

Yu, Q. Z.; Liang, T. J.

2018-06-01

China Spallation Neutron Source (CSNS) is intended to begin operation in 2018. CSNS is an accelerator-base multidisciplinary user facility. The pulsed neutrons are produced by a 1.6GeV short-pulsed proton beam impinging on a W-Ta spallation target, at a beam power of100 kW and a repetition rate of 25 Hz. 20 neutron beam lines are extracted for the neutron scattering and neutron irradiation research. During the commissioning and maintenance scenarios, the gamma rays induced from the W-Ta target can cause the dose threat to the personal and the environment. In this paper, the gamma dose rate distributions for the W-Ta spallation are calculated, based on the engineering model of the target-moderator-reflector system. The shipping cask is analyzed to satisfy the dose rate limit that less than 2 mSv/h at the surface of the shipping cask. All calculations are performed by the Monte carlo code MCNPX2.5 and the activation code CINDER’90.

Coupled Neutron Transport for HZETRN

NASA Technical Reports Server (NTRS)

Slaba, Tony C.; Blattnig, Steve R.

2009-01-01

Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.

Interactive three-dimensional visualization and creation of geometries for Monte Carlo calculations

NASA Astrophysics Data System (ADS)

Theis, C.; Buchegger, K. H.; Brugger, M.; Forkel-Wirth, D.; Roesler, S.; Vincke, H.

2006-06-01

The implementation of three-dimensional geometries for the simulation of radiation transport problems is a very time-consuming task. Each particle transport code supplies its own scripting language and syntax for creating the geometries. All of them are based on the Constructive Solid Geometry scheme requiring textual description. This makes the creation a tedious and error-prone task, which is especially hard to master for novice users. The Monte Carlo code FLUKA comes with built-in support for creating two-dimensional cross-sections through the geometry and FLUKACAD, a custom-built converter to the commercial Computer Aided Design package AutoCAD, exists for 3D visualization. For other codes, like MCNPX, a couple of different tools are available, but they are often specifically tailored to the particle transport code and its approach used for implementing geometries. Complex constructive solid modeling usually requires very fast and expensive special purpose hardware, which is not widely available. In this paper SimpleGeo is presented, which is an implementation of a generic versatile interactive geometry modeler using off-the-shelf hardware. It is running on Windows, with a Linux version currently under preparation. This paper describes its functionality, which allows for rapid interactive visualization as well as generation of three-dimensional geometries, and also discusses critical issues regarding common CAD systems.

Evaluation of SNS Beamline Shielding Configurations using MCNPX Accelerated by ADVANTG

DOE Office of Scientific and Technical Information (OSTI.GOV)

Risner, Joel M; Johnson, Seth R.; Remec, Igor

2015-01-01

Shielding analyses for the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory pose significant computational challenges, including highly anisotropic high-energy sources, a combination of deep penetration shielding and an unshielded beamline, and a desire to obtain well-converged nearly global solutions for mapping of predicted radiation fields. The majority of these analyses have been performed using MCNPX with manually generated variance reduction parameters (source biasing and cell-based splitting and Russian roulette) that were largely based on the analyst's insight into the problem specifics. Development of the variance reduction parameters required extensive analyst time, and was often tailored to specific portionsmore » of the model phase space. We previously applied a developmental version of the ADVANTG code to an SNS beamline study to perform a hybrid deterministic/Monte Carlo analysis and showed that we could obtain nearly global Monte Carlo solutions with essentially uniform relative errors for mesh tallies that cover extensive portions of the model with typical voxel spacing of a few centimeters. The use of weight window maps and consistent biased sources produced using the FW-CADIS methodology in ADVANTG allowed us to obtain these solutions using substantially less computer time than the previous cell-based splitting approach. While those results were promising, the process of using the developmental version of ADVANTG was somewhat laborious, requiring user-developed Python scripts to drive much of the analysis sequence. In addition, limitations imposed by the size of weight-window files in MCNPX necessitated the use of relatively coarse spatial and energy discretization for the deterministic Denovo calculations that we used to generate the variance reduction parameters. We recently applied the production version of ADVANTG to this beamline analysis, which substantially streamlined the analysis process. We also tested importance function collapsing (in space and energy) capabilities in ADVANTG. These changes, along with the support for parallel Denovo calculations using the current version of ADVANTG, give us the capability to improve the fidelity of the deterministic portion of the hybrid analysis sequence, obtain improved weight-window maps, and reduce both the analyst and computational time required for the analysis process.« less

Analysis of a Neutronic Experiment on a Simulated Mercury Spallation Neutron Target Assembly Bombarded by Giga-Electron-Volt Protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maekawa, Fujio; Meigo, Shin-ichiro; Kasugai, Yoshimi

2005-05-15

A neutronic benchmark experiment on a simulated spallation neutron target assembly was conducted by using the Alternating Gradient Synchrotron at Brookhaven National Laboratory and was analyzed to investigate the prediction capability of Monte Carlo simulation codes used in neutronic designs of spallation neutron sources. The target assembly consisting of a mercury target, a light water moderator, and a lead reflector was bombarded by 1.94-, 12-, and 24-GeV protons, and the fast neutron flux distributions around the target and the spectra of thermal neutrons leaking from the moderator were measured in the experiment. In this study, the Monte Carlo particle transportmore » simulation codes NMTC/JAM, MCNPX, and MCNP-4A with associated cross-section data in JENDL and LA-150 were verified based on benchmark analysis of the experiment. As a result, all the calculations predicted the measured quantities adequately; calculated integral fluxes of fast and thermal neutrons agreed approximately within {+-}40% with the experiments although the overall energy range encompassed more than 12 orders of magnitude. Accordingly, it was concluded that these simulation codes and cross-section data were adequate for neutronics designs of spallation neutron sources.« less

Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

NASA Astrophysics Data System (ADS)

Sabaibang, S.; Lekchaum, S.; Tipayakul, C.

2015-05-01

This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

Radiography simulation on single-shot dual-spectrum X-ray for cargo inspection system.

PubMed

Gil, Youngmi; Oh, Youngdo; Cho, Moohyun; Namkung, Won

2011-02-01

We propose a method to identify materials in the dual energy X-ray (DeX) inspection system. This method identifies materials by combining information on the relative proportions T of high-energy and low-energy X-rays transmitted through the material, and the ratio R of the attenuation coefficient of the material when high-energy are used to that when low energy X-rays are used. In Monte Carlo N-Particle Transport Code (MCNPX) simulations using the same geometry as that of the real container inspection system, this T vs. R method successfully identified tissue-equivalent plastic and several metals. In further simulations, the single-shot mode of operating the accelerator led to better distinguishing of materials than the dual-shot system. Copyright © 2010 Elsevier Ltd. All rights reserved.

Absolute Calibration of Image Plate for electrons at energy between 100 keV and 4 MeV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, H; Back, N L; Eder, D C

2007-12-10

The authors measured the absolute response of image plate (Fuji BAS SR2040) for electrons at energies between 100 keV to 4 MeV using an electron spectrometer. The electron source was produced from a short pulse laser irradiated on the solid density targets. This paper presents the calibration results of image plate Photon Stimulated Luminescence PSL per electrons at this energy range. The Monte Carlo radiation transport code MCNPX results are also presented for three representative incident angles onto the image plates and corresponding electron energies depositions at these angles. These provide a complete set of tools that allows extraction ofmore » the absolute calibration to other spectrometer setting at this electron energy range.« less

Evaluation of RayXpert® for shielding design of medical facilities

NASA Astrophysics Data System (ADS)

Derreumaux, Sylvie; Vecchiola, Sophie; Geoffray, Thomas; Etard, Cécile

2017-09-01

In a context of growing demands for expert evaluation concerning medical, industrial and research facilities, the French Institute for radiation protection and nuclear safety (IRSN) considered necessary to acquire new software for efficient dimensioning calculations. The selected software is RayXpert®. Before using this software in routine, exposure and transmission calculations for some basic configurations were validated. The validation was performed by the calculation of gamma dose constants and tenth value layers (TVL) for usual shielding materials and for radioisotopes most used in therapy (Ir-192, Co-60 and I-131). Calculated values were compared with results obtained using MCNPX as a reference code and with published values. The impact of different calculation parameters, such as the source emission rays considered for calculation and the use of biasing techniques, was evaluated.

Multi-resolution voxel phantom modeling: a high-resolution eye model for computational dosimetry

NASA Astrophysics Data System (ADS)

Caracappa, Peter F.; Rhodes, Ashley; Fiedler, Derek

2014-09-01

Voxel models of the human body are commonly used for simulating radiation dose with a Monte Carlo radiation transport code. Due to memory limitations, the voxel resolution of these computational phantoms is typically too large to accurately represent the dimensions of small features such as the eye. Recently reduced recommended dose limits to the lens of the eye, which is a radiosensitive tissue with a significant concern for cataract formation, has lent increased importance to understanding the dose to this tissue. A high-resolution eye model is constructed using physiological data for the dimensions of radiosensitive tissues, and combined with an existing set of whole-body models to form a multi-resolution voxel phantom, which is used with the MCNPX code to calculate radiation dose from various exposure types. This phantom provides an accurate representation of the radiation transport through the structures of the eye. Two alternate methods of including a high-resolution eye model within an existing whole-body model are developed. The accuracy and performance of each method is compared against existing computational phantoms.

Development of a technique using MCNPX code for determination of nitrogen content of explosive materials using prompt gamma neutron activation analysis method

NASA Astrophysics Data System (ADS)

Nasrabadi, M. N.; Bakhshi, F.; Jalali, M.; Mohammadi, A.

2011-12-01

Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma 10.8 MeV following radioactive neutron capture by 14N nuclei. We aimed to study the feasibility of using field-portable prompt gamma neutron activation analysis (PGNAA) along with improved nuclear equipment to detect and identify explosives, illicit substances or landmines. A 252Cf radio-isotopic source was embedded in a cylinder made of high-density polyethylene (HDPE) and the cylinder was then placed in another cylindrical container filled with water. Measurements were performed on high nitrogen content compounds such as melamine (C3H6N6). Melamine powder in a HDPE bottle was placed underneath the vessel containing water and the neutron source. Gamma rays were detected using two NaI(Tl) crystals. The results were simulated with MCNP4c code calculations. The theoretical calculations and experimental measurements were in good agreement indicating that this method can be used for detection of explosives and illicit drugs.

Neutron-induced reaction cross-sections of 93Nb with fast neutron based on 9Be(p,n) reaction

NASA Astrophysics Data System (ADS)

Naik, H.; Kim, G. N.; Kim, K.; Zaman, M.; Nadeem, M.; Sahid, M.

2018-02-01

The cross-sections of the 93Nb (n , 2 n)92mNb, 93Nb (n , 3 n)91mNb and 93Nb (n , 4 n)90Nb reactions with the average neutron energies of 14.4 to 34.0 MeV have been determined by using an activation and off-line γ-ray spectrometric technique. The fast neutrons were produced using the 9Be (p , n) reaction with the proton energies of 25-, 35- and 45-MeV from the MC-50 Cyclotron at the Korea Institute of Radiological and Medical Sciences (KIRAMS). The neutron flux-weighted average cross-sections of the 93Nb(n , xn ; x = 2- 4) reactions were also obtained from the mono-energetic neutron-induced reaction cross-sections of 93Nb calculated using the TALYS 1.8 code, and the neutron flux spectrum based on the MCNPX 2.6.0 code. The present results for the 93Nb(n , xn ; x = 2- 4) reactions are compared with the calculated neutron flux-weighted average values and found to be in good agreement.

Shot-by-shot spectrum model for rod-pinch, pulsed radiography machines

NASA Astrophysics Data System (ADS)

Wood, Wm M.

2018-02-01

A simplified model of bremsstrahlung production is developed for determining the x-ray spectrum output of a rod-pinch radiography machine, on a shot-by-shot basis, using the measured voltage, V(t), and current, I(t). The motivation for this model is the need for an agile means of providing shot-by-shot spectrum prediction, from a laptop or desktop computer, for quantitative radiographic analysis. Simplifying assumptions are discussed, and the model is applied to the Cygnus rod-pinch machine. Output is compared to wedge transmission data for a series of radiographs from shots with identical target objects. Resulting model enables variation of parameters in real time, thus allowing for rapid optimization of the model across many shots. "Goodness of fit" is compared with output from LSP Particle-In-Cell code, as well as the Monte Carlo Neutron Propagation with Xrays ("MCNPX") model codes, and is shown to provide an excellent predictive representation of the spectral output of the Cygnus machine. Improvements to the model, specifically for application to other geometries, are discussed.

Study of neutron spectra in a water bath from a Pb target irradiated by 250 MeV protons

NASA Astrophysics Data System (ADS)

Li, Yan-Yan; Zhang, Xue-Ying; Ju, Yong-Qin; Ma, Fei; Zhang, Hong-Bin; Chen, Liang; Ge, Hong-Lin; Wan, Bo; Luo, Peng; Zhou, Bin; Zhang, Yan-Bin; Li, Jian-Yang; Xu, Jun-Kui; Wang, Song-Lin; Yang, Yong-Wei; Yang, Lei

2015-04-01

Spallation neutrons were produced by the irradiation of Pb with 250 MeV protons. The Pb target was surrounded by water which was used to slow down the emitted neutrons. The moderated neutrons in the water bath were measured by using the resonance detectors of Au, Mn and In with a cadmium (Cd) cover. According to the measured activities of the foils, the neutron flux at different resonance energies were deduced and the epithermal neutron spectra were proposed. Corresponding results calculated with the Monte Carlo code MCNPX were compared with the experimental data to check the validity of the code. The comparison showed that the simulation could give a good prediction for the neutron spectra above 50 eV, while the finite thickness of the foils greatly effected the experimental data in low energy. It was also found that the resonance detectors themselves had great impact on the simulated energy spectra. Supported by National Natural Science Foundation and Strategic Priority Research Program of the Chinese Academy of Sciences (11305229, 11105186, 91226107, 91026009, XDA03030300)

SU-E-T-409: Evaluation of Tissue Composition Effect On Dose Distribution in Radiotherapy with 6 MV Photon Beam of a Medical Linac

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghorbani, M; Tabatabaei, Z; Noghreiyan, A Vejdani

Purpose: The aim of this study is to evaluate soft tissue composition effect on dose distribution for various soft tissues and various depths in radiotherapy with 6 MV photon beam of a medical linac. Methods: A phantom and Siemens Primus linear accelerator were simulated using MCNPX Monte Carlo code. In a homogeneous cubic phantom, six types of soft tissue and three types of tissue-equivalent materials were defined separately. The soft tissues were muscle (skeletal), adipose tissue, blood (whole), breast tissue, soft tissue (9-component) and soft tissue (4-component). The tissue-equivalent materials included: water, A-150 tissue-equivalent plastic and perspex. Photon dose relativemore » to dose in 9-component soft tissue at various depths on the beam’s central axis was determined for the 6 MV photon beam. The relative dose was also calculated and compared for various MCNPX tallies including,F8, F6 and,F4. Results: The results of the relative photon dose in various materials relative to dose in 9-component soft tissue and using different tallies are reported in the form of tabulated data. Minor differences between dose distributions in various soft tissues and tissue-equivalent materials were observed. The results from F6 and F4 were practically the same but different with,F8 tally. Conclusion: Based on the calculations performed, the differences in dose distributions in various soft tissues and tissue-equivalent materials are minor but they could be corrected in radiotherapy calculations to upgrade the accuracy of the dosimetric calculations.« less

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mashnik, Stepan Georgievich; Kerby, Leslie Marie

2015-05-22

MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 usingmore » CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.« less

Monte carlo simulations of the n_TOF lead spallation target with the Geant4 toolkit: A benchmark study

NASA Astrophysics Data System (ADS)

Lerendegui-Marco, J.; Cortés-Giraldo, M. A.; Guerrero, C.; Quesada, J. M.; Meo, S. Lo; Massimi, C.; Barbagallo, M.; Colonna, N.; Mancussi, D.; Mingrone, F.; Sabaté-Gilarte, M.; Vannini, G.; Vlachoudis, V.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bacak, M.; Balibrea, J.; Bečvář, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brown, A.; Caamaño, M.; Calviño, F.; Calviani, M.; Cano-Ott, D.; Cardella, R.; Casanovas, A.; Cerutti, F.; Chen, Y. H.; Chiaveri, E.; Cortés, G.; Cosentino, L.; Damone, L. A.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Göbel, K.; Gómez-Hornillos, M. B.; García, A. R.; Gawlik, A.; Gilardoni, S.; Glodariu, T.; Gonçalves, I. F.; González, E.; Griesmayer, E.; Gunsing, F.; Harada, H.; Heinitz, S.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Kalamara, A.; Kavrigin, P.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Kurtulgil, D.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lonsdale, S. J.; Macina, D.; Marganiec, J.; Martínez, T.; Masi, A.; Mastinu, P.; Mastromarco, M.; Maugeri, E. A.; Mazzone, A.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Musumarra, A.; Negret, A.; Nolte, R.; Oprea, A.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Radeck, D.; Rauscher, T.; Reifarth, R.; Rout, P. C.; Rubbia, C.; Ryan, J. A.; Saxena, A.; Schillebeeckx, P.; Schumann, D.; Smith, A. G.; Sosnin, N. V.; Stamatopoulos, A.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tassan-Got, L.; Valenta, S.; Variale, V.; Vaz, P.; Ventura, A.; Vlastou, R.; Wallner, A.; Warren, S.; Woods, P. J.; Wright, T.; Žugec, P.

2017-09-01

Monte Carlo (MC) simulations are an essential tool to determine fundamental features of a neutron beam, such as the neutron flux or the γ-ray background, that sometimes can not be measured or at least not in every position or energy range. Until recently, the most widely used MC codes in this field had been MCNPX and FLUKA. However, the Geant4 toolkit has also become a competitive code for the transport of neutrons after the development of the native Geant4 format for neutron data libraries, G4NDL. In this context, we present the Geant4 simulations of the neutron spallation target of the n_TOF facility at CERN, done with version 10.1.1 of the toolkit. The first goal was the validation of the intra-nuclear cascade models implemented in the code using, as benchmark, the characteristics of the neutron beam measured at the first experimental area (EAR1), especially the neutron flux and energy distribution, and the time distribution of neutrons of equal kinetic energy, the so-called Resolution Function. The second goal was the development of a Monte Carlo tool aimed to provide useful calculations for both the analysis and planning of the upcoming measurements at the new experimental area (EAR2) of the facility.

Induced radioactivity studies of the shielding and beamline equipment of the high intensity proton accelerator facility at PSI

NASA Astrophysics Data System (ADS)

Otiougova, Polina; Bergmann, Ryan; Kiselev, Daniela; Talanov, Vadim; Wohlmuther, Michael

2017-09-01

The Paul Scherrer Institute (PSI) is the largest national research center in Switzerland. Its multidisciplinary research is dedicated to a wide ↓eld in natural science and technology as well as particle physics. The High Intensity Proton Accelerator Facility (HIPA) has been in operation at PSI since 1974. It includes an 870 keV Cockroft-Walton pre-accelerator, a 72 MeV injector cyclotron as well as a 590 MeV ring cyclotron. The experimental facilities, the meson production graphite targets, Target E and Target M, and the spallation target stations (SINQ and UCN) are used for material research and particle physics. In order to ful↓ll the request of the regulatory authorities and to be reported to the regulators, the expected radioactive waste and nuclide inventory after an anticipated ↓nal shutdown in the far future has to be estimated. In this contribution, calculations for the 20 m long beamline between Target E and the 590 MeV beam dump of HIPA are presented. The ↓rst step in the calculations was determining spectra and spatial particle distributions around the beamlines using the Monte-Carlo particle transport code MCNPX2.7.0 [1]. To perform the analysis of the MCNPX output and to determine the radionuclide inventory as well as the speci↓c activity of the nuclides, an activation script [2] using the FISPACT10 code with the cross sections from the European Activation File (EAF2010) [3] was applied. The speci↓c activity values were compared to the currently existing Swiss exemption limits (LE) [4] as well as to the Swiss liberation limits (LL) [5], becoming e↑ective in the near future. The obtained results were used to estimate the total volume of the radioactive waste produced at HIPA and have to be reported to the Swiss regulatory authorities. The comparison of the performed calculations to measurements is discussed as well. Note to the reader: the pdf file has been changed on September 22, 2017.

In-Situ Assays Using a New Advanced Mathematical Algorithm - 12400

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oginni, B.M.; Bronson, F.L.; Field, M.B.

2012-07-01

Current mathematical efficiency modeling software for in-situ counting, such as the commercially available In-Situ Object Calibration Software (ISOCS), typically allows the description of measurement geometries via a list of well-defined templates which describe regular objects, such as boxes, cylinder, or spheres. While for many situations, these regular objects are sufficient to describe the measurement conditions, there are occasions in which a more detailed model is desired. We have developed a new all-purpose geometry template that can extend the flexibility of current ISOCS templates. This new template still utilizes the same advanced mathematical algorithms as current templates, but allows the extensionmore » to a multitude of shapes and objects that can be placed at any location and even combined. In addition, detectors can be placed anywhere and aimed at any location within the measurement scene. Several applications of this algorithm to in-situ waste assay measurements, as well as, validations of this template using Monte Carlo calculations and experimental measurements are studied. Presented in this paper is a new template of the mathematical algorithms for evaluating efficiencies. This new template combines all the advantages of the ISOCS and it allows the use of very complex geometries, it also allows stacking of geometries on one another in the same measurement scene and it allows the detector to be placed anywhere in the measurement scene and pointing in any direction. We have shown that the template compares well with the previous ISOCS software within the limit of convergence of the code, and also compare well with the MCNPX and measured data within the joint uncertainties for the code and the data. The new template agrees with ISOCS to within 1.5% at all energies. It agrees with the MCNPX to within 10% at all energies and it agrees with most geometries within 5%. It finally agrees with measured data to within 10%. This mathematical algorithm can now be used for quickly and accurately evaluating efficiencies for wider range of gamma-ray spectroscopy applications. (authors)« less

Radiation Environment Inside Spacecraft

NASA Technical Reports Server (NTRS)

O'Neill, Patrick

2015-01-01

Dr. Patrick O'Neill, NASA Johnson Space Center, will present a detailed description of the radiation environment inside spacecraft. The free space (outside) solar and galactic cosmic ray and trapped Van Allen belt proton spectra are significantly modified as these ions propagate through various thicknesses of spacecraft structure and shielding material. In addition to energy loss, secondary ions are created as the ions interact with the structure materials. Nuclear interaction codes (FLUKA, GEANT4, HZTRAN, MCNPX, CEM03, and PHITS) transport free space spectra through different thicknesses of various materials. These "inside" energy spectra are then converted to Linear Energy Transfer (LET) spectra and dose rate - that's what's needed by electronics systems designers. Model predictions are compared to radiation measurements made by instruments such as the Intra-Vehicular Charged Particle Directional Spectrometer (IV-CPDS) used inside the Space Station, Orion, and Space Shuttle.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Purwaningsih, Anik

Dosimetric data for a brachytherapy source should be known before it used for clinical treatment. Iridium-192 source type H01 was manufactured by PRR-BATAN aimed to brachytherapy is not yet known its dosimetric data. Radial dose function and anisotropic dose distribution are some primary keys in brachytherapy source. Dose distribution for Iridium-192 source type H01 was obtained from the dose calculation formalism recommended in the AAPM TG-43U1 report using MCNPX 2.6.0 Monte Carlo simulation code. To know the effect of cavity on Iridium-192 type H01 caused by manufacturing process, also calculated on Iridium-192 type H01 if without cavity. The result ofmore » calculation of radial dose function and anisotropic dose distribution for Iridium-192 source type H01 were compared with another model of Iridium-192 source.« less

SESAME: a software tool for the numerical dosimetric reconstruction of radiological accidents involving external sources and its application to the accident in Chile in December 2005.

PubMed

Huet, C; Lemosquet, A; Clairand, I; Rioual, J B; Franck, D; de Carlan, L; Aubineau-Lanièce, I; Bottollier-Depois, J F

2009-01-01

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. This dose distribution can be assessed by physical dosimetric reconstruction methods. Physical dosimetric reconstruction can be achieved using experimental or numerical techniques. This article presents the laboratory-developed SESAME--Simulation of External Source Accident with MEdical images--tool specific to dosimetric reconstruction of radiological accidents through numerical simulations which combine voxel geometry and the radiation-material interaction MCNP(X) Monte Carlo computer code. The experimental validation of the tool using a photon field and its application to a radiological accident in Chile in December 2005 are also described.

Study on a new design of Tehran Research Reactor for radionuclide production based on fast neutrons using MCNPX code.

PubMed

Zandi, Nadia; Afarideh, Hossein; Aboudzadeh, Mohammad Reza; Rajabifar, Saeed

2018-02-01

The aim of this work is to increase the magnitude of the fast neutron flux inside the flux trap where radionuclides are produced. For this purpose, three new designs of the flux trap are proposed and the obtained fast and thermal neutron fluxes compared with each other. The first and second proposed designs were a sealed cube contained air and D 2 O, respectively. The results of calculated production yield all indicated the superiority of the latter by a factor of 55% in comparison to the first proposed design. The third proposed design was based on changing the surrounding of the sealed cube by locating two fuel plates near that. In this case, the production yield increased up to 70%. Copyright © 2017. Published by Elsevier Ltd.

Fluence-to-Absorbed Dose Conversion Coefficients for Use in Radiological Protection of Embryo and Foetus Against External Exposure to Muons from 20MeV to 50GeV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen Jing

2008-08-07

This study used the Monte-Carlo code MCNPX to determine mean absorbed doses to the embryo and foetus when the mother is exposed to external muon fields. Monoenergetic muons ranging from 20 MeV to 50 GeV were considered. The irradiation geometries include anteroposterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT), isotropic (ISO), and top-down (TOP). At each of these irradiation geometries, absorbed doses to the foetal body were calculated for the embryo of 8 weeks and the foetus of 3, 6 or 9 months, respectively. Muon fluence-to-absorbed-dose conversion coefficients were derived for the four prenatal ages. Since such conversion coefficients aremore » yet unknown, the results presented here fill a data gap.« less

Feasibility study for a realistic training dedicated to radiological protection improvement

NASA Astrophysics Data System (ADS)

Courageot, Estelle; Reinald, Kutschera; Gaillard-Lecanu, Emmanuelle; Sylvie, Jahan; Riedel, Alexandre; Therache, Benjamin

2014-06-01

Any personnel involved in activities within the controlled area of a nuclear facility must be provided with appropriate radiological protection training. An evident purpose of this training is to know the regulation dedicated to workplaces where ionizing radiation may be present, in order to properly carry out the radiation monitoring, to use suitable protective equipments and to behave correctly if unexpected working conditions happen. A major difficulty of this training consist in having the most realistic reading from the monitoring devices for a given exposure situation, but without using real radioactive sources. A new approach is developed at EDF R&D for radiological protection training. This approach combines different technologies, in an environment representative of the workplace but geographically separated from the nuclear power plant: a training area representative of a workplace, a Man Machine Interface used by the trainer to define the source configuration and the training scenario, a geolocalization system, fictive radiation monitoring devices and a particle transport code able to calculate in real time the dose map due to the virtual sources. In a first approach, our real-time particles transport code, called Moderato, used only an attenuation low in straight line. To improve the realism further, we would like to switch a code based on the Monte Carlo transport of particles method like Geant 4 or MCNPX instead of Moderato. The aim of our study is the evaluation of the code in our application, in particular, the possibility to keep a real time response of our architecture.

Pulsation, Mass Loss and the Upper Mass Limit

NASA Astrophysics Data System (ADS)

Klapp, J.; Corona-Galindo, M. G.

1990-11-01

RESUMEN. La existencia de estrellas con masas en exceso de 100 M0 ha sido cuestionada por mucho tiempo. Lfmites superiores para la masa de 100 M0 han sido obtenidos de teorfas de pulsaci6n y formaci6n estelar. En este trabajo nosotros primero investigamos la estabilidad radial de estrellas masivas utilizando la aproximaci6n clasica cuasiadiabatica de Ledoux, la aproximaci6n cuasiadiabatica de Castor y un calculo completamente no-adiabatico. Hemos encontrado que los tres metodos de calculo dan resultados similares siempre y cuando una pequefia regi6n de las capas externas de la estrella sea despreciada para la aproximaci6n clasica. La masa crftica para estabilidad de estrellas masivas ha sido encontrada en acuerdo a trabajos anteriores. Explicamos Ia discrepancia entre este y trabajos anteriores por uno de los autores. Discunmos calculos no-lineales y perdida de masa con respecto a) lfmite superior de masa. The existence of stars with masses in excess of 100 M0 has been questioned for a very long time. Upper mass limits of 100 Me have been obtained from pulsation and star formation theories. In this work we first investigate the radial stability of massive stars using the classical Ledoux's quasiadiabatic approximation. the Castor quasiadiabatic approximation and a fully nonadiabatic calculation. We have found that the three methods of calculation give similar results provided that a small region in outer layers of the star be neglected for the classical approximation. The critical mass for stability of massive stars is found to be in agreement with previous work. We explain the reason for the discrepancy between this and previous work by one of the authors. We discuss non-linear calculations and mass loss with regard to the upper mass limit. Key words: STARS-MASS FUNCTION - STARS-MASS LOSS - STARS-PULSATION

Dual-resolution dose assessments for proton beamlet using MCNPX 2.6.0

NASA Astrophysics Data System (ADS)

Chao, T. C.; Wei, S. C.; Wu, S. W.; Tung, C. J.; Tu, S. J.; Cheng, H. W.; Lee, C. C.

2015-11-01

The purpose of this study is to access proton dose distribution in dual resolution phantoms using MCNPX 2.6.0. The dual resolution phantom uses higher resolution in Bragg peak, area near large dose gradient, or heterogeneous interface and lower resolution in the rest. MCNPX 2.6.0 was installed in Ubuntu 10.04 with MPI for parallel computing. FMesh1 tallies were utilized to record the energy deposition which is a special designed tally for voxel phantoms that converts dose deposition from fluence. 60 and 120 MeV narrow proton beam were incident into Coarse, Dual and Fine resolution phantoms with pure water, water-bone-water and water-air-water setups. The doses in coarse resolution phantoms are underestimated owing to partial volume effect. The dose distributions in dual or high resolution phantoms agreed well with each other and dual resolution phantoms were at least 10 times more efficient than fine resolution one. Because the secondary particle range is much longer in air than in water, the dose of low density region may be under-estimated if the resolution or calculation grid is not small enough.

Cross-correlation measurements with the EJ-299-33 plastic scintillator

NASA Astrophysics Data System (ADS)

Bourne, Mark M.; Whaley, Jeff; Dolan, Jennifer L.; Polack, John K.; Flaska, Marek; Clarke, Shaun D.; Tomanin, Alice; Peerani, Paolo; Pozzi, Sara A.

2015-06-01

New organic-plastic scintillation compositions have demonstrated pulse-shape discrimination (PSD) of neutrons and gamma rays. We present cross-correlation measurements of 252Cf and mixed uranium-plutonium oxide (MOX) with the EJ-299-33 plastic scintillator. For comparison, equivalent measurements were performed with an EJ-309 liquid scintillator. Offline, digital PSD was applied to each detector. These measurements show that EJ-299-33 sacrifices a factor of 5 in neutron-neutron efficiency relative to EJ-309, but could still utilize the difference in neutron-neutron efficiency and neutron single-to-double ratio to distinguish 252Cf from MOX. These measurements were modeled with MCNPX-PoliMi, and MPPost was used to convert the detailed collision history into simulated cross-correlation distributions. MCNPX-PoliMi predicted the measured 252Cf cross-correlation distribution for EJ-309 to within 10%. Greater photon uncertainty in the MOX sample led to larger discrepancy in the simulated MOX cross-correlation distribution. The modeled EJ-299-33 plastic also gives reasonable agreement with measured cross-correlation distributions, although the MCNPX-PoliMi model appears to under-predict the neutron detection efficiency.

USE OF TRANSPORTABLE RADIATION DETECTION INSTRUMENTS TO ASSESS INTERNAL CONTAMINATION FROM INTAKES OF RADIONUCLIDES PART I: FIELD TESTS AND MONTE CARLO SIMULATIONS

PubMed Central

Anigstein, Robert; Erdman, Michael C.; Ansari, Armin

2017-01-01

The detonation of a radiological dispersion device or other radiological incidents could result in the dispersion of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure photon radiation from radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for further assessments. Computer simulations and experimental measurements are required for these instruments to be used for assessing intakes of radionuclides. Count rates from calibrated sources of 60Co, 137Cs, and 241Am were measured on three instruments: a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal, a thyroid probe using a 5.08 × 5.08-cm NaI(Tl) crystal, and a portal monitor incorporating two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators. Computer models of the instruments and of the calibration sources were constructed, using engineering drawings and other data provided by the manufacturers. Count rates on the instruments were simulated using the Monte Carlo radiation transport code MCNPX. The computer simulations were within 16% of the measured count rates for all 20 measurements without using empirical radionuclide-dependent scaling factors, as reported by others. The weighted root-mean-square deviations (differences between measured and simulated count rates, added in quadrature and weighted by the variance of the difference) were 10.9% for the survey meter, 4.2% for the thyroid probe, and 0.9% for the portal monitor. These results validate earlier MCNPX models of these instruments that were used to develop calibration factors that enable these instruments to be used for assessing intakes and committed doses from several gamma-emitting radionuclides. PMID:27115229

Use of Transportable Radiation Detection Instruments to Assess Internal Contamination From Intakes of Radionuclides Part I: Field Tests and Monte Carlo Simulations.

PubMed

Anigstein, Robert; Erdman, Michael C; Ansari, Armin

2016-06-01

The detonation of a radiological dispersion device or other radiological incidents could result in the dispersion of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure photon radiation from radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for further assessments. Computer simulations and experimental measurements are required for these instruments to be used for assessing intakes of radionuclides. Count rates from calibrated sources of Co, Cs, and Am were measured on three instruments: a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal, a thyroid probe using a 5.08 × 5.08-cm NaI(Tl) crystal, and a portal monitor incorporating two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators. Computer models of the instruments and of the calibration sources were constructed, using engineering drawings and other data provided by the manufacturers. Count rates on the instruments were simulated using the Monte Carlo radiation transport code MCNPX. The computer simulations were within 16% of the measured count rates for all 20 measurements without using empirical radionuclide-dependent scaling factors, as reported by others. The weighted root-mean-square deviations (differences between measured and simulated count rates, added in quadrature and weighted by the variance of the difference) were 10.9% for the survey meter, 4.2% for the thyroid probe, and 0.9% for the portal monitor. These results validate earlier MCNPX models of these instruments that were used to develop calibration factors that enable these instruments to be used for assessing intakes and committed doses from several gamma-emitting radionuclides.

Proton radiography and fluoroscopy of lung tumors: A Monte Carlo study using patient-specific 4DCT phantoms

PubMed Central

Han, Bin; Xu, X. George; Chen, George T. Y.

2011-01-01

Purpose: Monte Carlo methods are used to simulate and optimize a time-resolved proton range telescope (TRRT) in localization of intrafractional and interfractional motions of lung tumor and in quantification of proton range variations. Methods: The Monte Carlo N-Particle eXtended (MCNPX) code with a particle tracking feature was employed to evaluate the TRRT performance, especially in visualizing and quantifying proton range variations during respiration. Protons of 230 MeV were tracked one by one as they pass through position detectors, patient 4DCT phantom, and finally scintillator detectors that measured residual ranges. The energy response of the scintillator telescope was investigated. Mass density and elemental composition of tissues were defined for 4DCT data. Results: Proton water equivalent length (WEL) was deduced by a reconstruction algorithm that incorporates linear proton track and lateral spatial discrimination to improve the image quality. 4DCT data for three patients were used to visualize and measure tumor motion and WEL variations. The tumor trajectories extracted from the WEL map were found to be within ∼1 mm agreement with direct 4DCT measurement. Quantitative WEL variation studies showed that the proton radiograph is a good representation of WEL changes from entrance to distal of the target. Conclusions:MCNPX simulation results showed that TRRT can accurately track the motion of the tumor and detect the WEL variations. Image quality was optimized by choosing proton energy, testing parameters of image reconstruction algorithm, and comparing to ground truth 4DCT. The future study will demonstrate the feasibility of using the time resolved proton radiography as an imaging tool for proton treatments of lung tumors. PMID:21626923

MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes

NASA Astrophysics Data System (ADS)

Fonseca, T. C. F.; Mendes, B. M.; Lacerda, M. A. S.; Silva, L. A. C.; Paixão, L.; Bastos, F. M.; Ramirez, J. V.; Junior, J. P. R.

2017-11-01

The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm2. This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results.

An Improved Neutron Transport Algorithm for Space Radiation

NASA Technical Reports Server (NTRS)

Heinbockel, John H.; Clowdsley, Martha S.; Wilson, John W.

2000-01-01

A low-energy neutron transport algorithm for use in space radiation protection is developed. The algorithm is based upon a multigroup analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. This analysis is accomplished by solving a realistic but simplified neutron transport test problem. The test problem is analyzed by using numerical and analytical procedures to obtain an accurate solution within specified error bounds. Results from the test problem are then used for determining mean values associated with rescattering terms that are associated with a multigroup solution of the straight-ahead Boltzmann equation. The algorithm is then coupled to the Langley HZETRN code through the evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for a water and an aluminum-water shield-target configuration is then compared with LAHET and MCNPX Monte Carlo code calculations for the same shield-target configuration. The algorithm developed showed a great improvement in results over the unmodified HZETRN solution. In addition, a two-directional solution of the evaporation source showed even further improvement of the fluence near the front of the water target where diffusion from the front surface is important.

Neutron spectra due (13)N production in a PET cyclotron.

PubMed

Benavente, J A; Vega-Carrillo, H R; Lacerda, M A S; Fonseca, T C F; Faria, F P; da Silva, T A

2015-05-01

Monte Carlo and experimental methods have been used to characterize the neutron radiation field around PET (Positron Emission Tomography) cyclotrons. In this work, the Monte Carlo code MCNPX was used to estimate the neutron spectra, the neutron fluence rates and the ambient dose equivalent (H*(10)) in seven locations around a PET cyclotron during (13)N production. In order to validate these calculations, H*(10) was measured in three sites and were compared with the calculated doses. All the spectra have two peaks, one above 0.1MeV due to the evaporation neutrons and another in the thermal region due to the room-return effects. Despite the relatively large difference between the measured and calculated H*(10) for one point, the agreement was considered good, compared with that obtained for (18)F production in a previous work. Copyright © 2015 Elsevier Ltd. All rights reserved.

Shielding and activation calculations around the reactor core for the MYRRHA ADS design

NASA Astrophysics Data System (ADS)

Ferrari, Anna; Mueller, Stefan; Konheiser, J.; Castelliti, D.; Sarotto, M.; Stankovskiy, A.

2017-09-01

In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions.

Radiation damage calculations for the SINQ Target 5

NASA Astrophysics Data System (ADS)

Wechsler, Monroe S.; Lu, Wei; Dai, Yong

2003-03-01

Calculations are underway of radiation damage (production of displacements, helium, and hydrogen) at Target 5 of the SINQ spallation neutron source at the Paul Scherrer Institute in Switzerland. The target is bombarded by 575-MeV protons, and the spallation-neutron-producing target material is liquid lead. The calculations employ the Monte Carlo code MCNPX (version 2.3.0). The peak proton and neutron fluxes at the aluminum-alloy entrance window are determined to be about 1.9E14 protons/cm2s per mA of incident proton current and 2.4E13 neutrons/cm2s per mA. For a beam exposure of 10 Ahr, the peak damage sustained at the entrance window due to protons and neutrons combined is calculated to be 7.8 dpa, 2000 appmHe, and 4000 appmH. The significance of the damage results for the entrance window and components within Target 5 will be discussed.

Microtron MT 25 as a source of neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kralik, M.; Solc, J.; Chvatil, D.

2012-08-15

The objective was to describe Microtron MT25 as a source of neutrons generated by bremsstrahlung induced photonuclear reactions in U and Pb targets. Bremsstrahlung photons were produced by electrons accelerated at energy 21.6 MeV. Spectral fluence of the generated neutrons was calculated with MCNPX code and then experimentally determined at two positions by means of a Bonner spheres spectrometer in which the detector of thermal neutrons was replaced by activation Mn tablets or track detectors CR-39 with a {sup 10}B radiator. The measured neutron spectral fluence and the calculated anisotropy served for the estimation of neutron yield from the targetsmore » and for the determination of ambient dose equivalent rate at the place of measurement. Microtron MT25 is intended as one of the sources for testing neutron sensitive devices which will be sent into the space.« less

Cancer risk coefficient for patient undergoing kyphoplasty surgery using Monte Carlo method

NASA Astrophysics Data System (ADS)

Santos, Felipe A.; Santos, William S.; Galeano, Diego C.; Cavalcante, Fernanda R.; Silva, Ademir X.; Souza, Susana O.; Júnior, Albérico B. Carvalho

2017-11-01

Kyphoplasty surgery is widely used for pain relief in patients with vertebral compression fracture (VCF). For this surgery, an X-ray emitter that provides real-time imaging is employed to guide the medical instruments and the surgical cement used to fill and strengthen the vertebra. Equivalent and effective doses related to high temporal resolution equipment has been studied to assess the damage and more recently cancer risk. For this study, a virtual scenario was prepared using MCNPX code and a pair of UF family simulators. Two projections with seven tube voltages for each one were simulated. The organ in the abdominal region were those who had higher cancer risk because they receive the primary beam. The risk of lethal cancer is on average 20% higher in AP projection than in LL projection. This study aims at estimating the risk of cancer in organs and the risk of lethal cancer for patient submitted to kyphoplasty surgery.

A Monte Carlo model for photoneutron generation by a medical LINAC

NASA Astrophysics Data System (ADS)

Sumini, M.; Isolan, L.; Cucchi, G.; Sghedoni, R.; Iori, M.

2017-11-01

For an optimal tuning of the radiation protection planning, a Monte Carlo model using the MCNPX code has been built, allowing an accurate estimate of the spectrometric and geometrical characteristics of photoneutrons generated by a Varian TrueBeam Stx© medical linear accelerator. We considered in our study a device working at the reference energy for clinical applications of 15 MV, stemmed from a Varian Clinac©2100 modeled starting from data collected thanks to several papers available in the literature. The model results were compared with neutron and photon dose measurements inside and outside the bunker hosting the accelerator obtaining a complete dose map. Normalized neutron fluences were tallied in different positions at the patient plane and at different depths. A sensitivity analysis with respect to the flattening filter material were performed to enlighten aspects that could influence the photoneutron production.

Application of thin-film breakdown counters for characterization of neutron field of the VESUVIO instrument at the ISIS spallation source

NASA Astrophysics Data System (ADS)

Smirnov, A. N.; Pietropaolo, A.; Prokofiev, A. V.; Rodionova, E. E.; Frost, C. D.; Ansell, S.; Schooneveld, E. M.; Gorini, G.

2012-09-01

The high-energy neutron field of the VESUVIO instrument at the ISIS facility has been characterized using the technique of thin-film breakdown counters (TFBC). The technique utilizes neutron-induced fission reactions of natU and 209Bi with detection of fission fragments by TFBCs. Experimentally determined count rates of the fragments are ≈50% higher than those calculated using spectral neutron flux simulated with the MCNPX code. This work is a part of the project to develop ChipIr, a new dedicated facility for the accelerated testing of electronic components and systems for neutron-induced single event effects in the new Target Station 2 at ISIS. The TFBC technique has shown to be applicable for on-line monitoring of the neutron flux in the neutron energy range 1-800 MeV at the position of the device under test (DUT).

Experiments and Monte Carlo modeling of a higher resolution Cadmium Zinc Telluride detector for safeguards applications

NASA Astrophysics Data System (ADS)

Borella, Alessandro

2016-09-01

The Belgian Nuclear Research Centre is engaged in R&D activity in the field of Non Destructive Analysis on nuclear materials, with focus on spent fuel characterization. A 500 mm3 Cadmium Zinc Telluride (CZT) with enhanced resolution was recently purchased. With a full width at half maximum of 1.3% at 662 keV, the detector is very promising in view of its use for applications such as determination of uranium enrichment and plutonium isotopic composition, as well as measurement on spent fuel. In this paper, I report about the work done with such a detector in terms of its characterization. The detector energy calibration, peak shape and efficiency were determined from experimental data. The data included measurements with calibrated sources, both in a bare and in a shielded environment. In addition, Monte Carlo calculations with the MCNPX code were carried out and benchmarked with experiments.

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

Feasibility study of using laser-generated neutron beam for BNCT.

PubMed

Kasesaz, Y; Rahmani, F; Khalafi, H

2015-09-01

The feasibility of using a laser-accelerated proton beam to produce a neutron source, via (p,n) reaction, for Boron Neutron Capture Therapy (BNCT) applications has been studied by MCNPX Monte Carlo code. After optimization of the target material and its thickness, a Beam Shaping Assembly (BSA) has been designed and optimized to provide appropriate neutron beam according to the recommended criteria by International Atomic Energy Agency. It was found that the considered laser-accelerated proton beam can provide epithermal neutron flux of ∼2×10(6) n/cm(2) shot. To achieve an appropriate epithermal neutron flux for BNCT treatment, the laser must operate at repetition rates of 1 kHz, which is rather ambitious at this moment. But it can be used in some BNCT researches field such as biological research. Copyright © 2015 Elsevier Ltd. All rights reserved.

A qualitative analysis of the neutron population in fresh and spent fuel assemblies during simulated interrogation using the differential die-away technique

DOE PAGES

Tobin, Stephen J.; Lundkvist, Niklas; Goodsell, Alison V.; ...

2015-12-01

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

Influence of clouds on the cosmic radiation dose rate on aircraft.

PubMed

Pazianotto, Maurício T; Federico, Claudio A; Cortés-Giraldo, Miguel A; Pinto, Marcos Luiz de A; Gonçalez, Odair L; Quesada, José Manuel M; Carlson, Brett V; Palomo, Francisco R

2014-10-01

Flight missions were made in Brazilian territory in 2009 and 2011 with the aim of measuring the cosmic radiation dose rate incident on aircraft in the South Atlantic Magnetic Anomaly and to compare it with Monte Carlo simulations. During one of these flights, small fluctuations were observed in the vicinity of the aircraft with formation of Cumulonimbus clouds. Motivated by these observations, in this work, the authors investigated the relationship between the presence of clouds and the neutron flux and dose rate incident on aircraft using computational simulation. The Monte Carlo simulations were made using the MCNPX and Geant4 codes, considering the incident proton flux at the top of the atmosphere and its propagation and neutron production through several vertically arranged slabs, which were modelled according to the ISO specifications. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Replacing a 252Cf source with a neutron generator in a shuffler - a conceptual design performed with MCNPX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schear, Melissa A; Tobin, Stephen J

2009-01-01

The {sup 252}Cf shuffler has been widely used in nuclear safeguards and radioactive waste management to assay fissile isotopes, such as {sup 235}U or {sup 239}Pu, present in a variety of samples, ranging from small cans of uranium waste to metal samples weighing several kilograms. Like other non-destructive assay instruments, the shuffler uses an interrogating neutron source to induce fissions in the sample. Although shufflers with {sup 252}Cf sources have been reliably used for several decades, replacing this isotopic source with a neutron generator presents some distinct advantages. Neutron generators can be run in a continuous or pulsed mode, andmore » may be turned off, eliminating the need for shielding and a shuffling mechanism in the shuffler. There is also essentially no dose to personnel during installation, and no reliance on the availability of {sup 252}Cf. Despite these advantages, the more energetic neutrons emitted from the neutron generator (141 MeV for D-T generators) present some challenges for certain material types. For example when the enrichment of a uranium sample is unknown, the fission of {sup 238}U is generally undesirable. Since measuring uranium is one of the main uses of a shuffler, reducing the delayed neutron contribution from {sup 238}U is desirable. Hence, the shuffler hardware must be modified to accommodate a moderator configuration near the source to tailor the interrogating spectrum in a manner which promotes sub-threshold fissions (below 1 MeV) but avoids the over-moderation of the interrogating neutrons so as to avoid self-shielding. In this study, where there are many material and geometry combinations, the Monte Carlo N-Particle eXtended (MCNPX) transport code was used to model, design, and optimize the moderator configuration within the shuffler geometry. The code is then used to evaluate and compare the assay performances of both the modified shuffler and the current {sup 252}Cf shuffler designs for different test samples. The matrix effect and the non-uniformity of the interrogating flux are investigated and quantified in each case. The modified geometry proposed by this study can serve s a guide in retrofitting shufflers that are already in use.« less

Incertidumbre en los valores de la fuerza de oscilador de las lineas espectrales y su influencia en el calculo de abundancias

NASA Astrophysics Data System (ADS)

Marchiano, P. E.; Di Rocco, H. O.; Cruzado, A.

In this paper; we compare oscillator strengths values from XeII lines theoretically obtained with values obtained via experimentation. We put forward a working hypothesis aimed at reaching the best possible agreement; and applying it to the study of abundances of other elements in the atmospheres of chemically peculiar stars. FULL TEXT IN SPANISH

Radiation environment at LEO orbits: MC simulation and experimental data.

NASA Astrophysics Data System (ADS)

Zanini, Alba; Borla, Oscar; Damasso, Mario; Falzetta, Giuseppe

The evaluations of the different components of the radiation environment in spacecraft, both in LEO orbits and in deep space is of great importance because the biological effect on humans and the risk for instrumentation strongly depends on the kind of radiation (high or low LET). That is important especially in view of long term manned or unmanned space missions, (mission to Mars, solar system exploration). The study of space radiation field is extremely complex and not completely solved till today. Given the complexity of the radiation field, an accurate dose evaluation should be considered an indispensable part of any space mission. Two simulation codes (MCNPX and GEANT4) have been used to assess the secondary radiation inside FO-TON M3 satellite and ISS. The energy spectra of primary radiation at LEO orbits have been modelled by using various tools (SPENVIS, OMERE, CREME96) considering separately Van Allen protons, the GCR protons and the GCR alpha particles. This data are used as input for the two MC codes and transported inside the spacecraft. The results of two calculation meth-ods have been compared. Moreover some experimental results previously obtained on FOTON M3 satellite by using TLD, Bubble dosimeter and LIULIN detector are considered to check the performances of the two codes. Finally the same experimental device are at present collecting data on the ISS (ASI experiment BIOKIS -nDOSE) and at the end of the mission the results will be compared with the calculation.

Spallation neutron production and the current intra-nuclear cascade and transport codes

NASA Astrophysics Data System (ADS)

Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

Dose conversion coefficients for neutron exposure to the lens of the human eye.

PubMed

Manger, R P; Bellamy, M B; Eckerman, K F

2012-03-01

Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 × 10(-9) to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

Designing an extended energy range single-sphere multi-detector neutron spectrometer

NASA Astrophysics Data System (ADS)

Gómez-Ros, J. M.; Bedogni, R.; Moraleda, M.; Esposito, A.; Pola, A.; Introini, M. V.; Mazzitelli, G.; Quintieri, L.; Buonomo, B.

2012-06-01

This communication describes the design specifications for a neutron spectrometer consisting of 31 thermal neutron detectors, namely Dysprosium activation foils, embedded in a 25 cm diameter polyethylene sphere which includes a 1 cm thick lead shell insert that degrades the energy of neutrons through (n,xn) reactions, thus allowing to extension of the energy range of the response up to hundreds of MeV neutrons. The new spectrometer, called SP2 (SPherical SPectrometer), relies on the same detection mechanism as that of the Bonner Sphere Spectrometer, but with the advantage of determining the whole neutron spectrum in a single exposure. The Monte Carlo transport code MCNPX was used to design the spectrometer in terms of sphere diameter, number and position of the detectors, position and thickness of the lead shell, as well as to obtain the response matrix for the final configuration. This work focuses on evaluating the spectrometric capabilities of the SP2 design by simulating the exposure of SP2 in neutron fields representing different irradiation conditions (test spectra). The simulated SP2 readings were then unfolded with the FRUIT unfolding code, in the absence of detailed pre-information, and the unfolded spectra were compared with the known test spectra. The results are satisfactory and allowed approving the production of a prototypal spectrometer.

Multifunction Radar for Airborne Applications.

DTIC Science & Technology

1986-07-01

de rapport ( doo roSarvre o ntn orcp tion To de 0 & TR. Pour chaque valour de To, on calculo de la m~me faqon quo prdcdmment PS lo rIpporr - pour...senseur primordial dfalts les aviotts militaires. (in certain ttimbreC de caracteristiulues imptirtantes donne anl radar Ia supuirioritui sur Ics...muitcoroliugiques 1ironillard. nuagesi. 1estimation de distance et lestimatitin Douppler. - la sonplesse de at ]’orientatiiin ilectritnique des

Long Period Seismic Waves

DTIC Science & Technology

1976-08-01

Geoffsica, TPHM. No. 5 , p. 161. Vargas, Freddy (To he published in 1976) 1 .-DTSCRP1TNACTON DE EVENTO«; NATHDALE«; Y ARTTFTCT ALES. 2.- CALCULO DEL...seismic risk, bv de - fininn relative weiqht of maximum MM intensity at a pivon distance ponulation density, area feolupy and attenuation of intensity wit...Population densitv, area peolopv and attenuation of intensitv with distance, is presented topether with a map anplvinp theorv to Bo- livia. ^«^a

TH-C-BRD-02: Analytical Modeling and Dose Calculation Method for Asymmetric Proton Pencil Beams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelover, E; Wang, D; Hill, P

2014-06-15

Purpose: A dynamic collimation system (DCS), which consists of two pairs of orthogonal trimmer blades driven by linear motors has been proposed to decrease the lateral penumbra in pencil beam scanning proton therapy. The DCS reduces lateral penumbra by intercepting the proton pencil beam near the lateral boundary of the target in the beam's eye view. The resultant trimmed pencil beams are asymmetric and laterally shifted, and therefore existing pencil beam dose calculation algorithms are not capable of trimmed beam dose calculations. This work develops a method to model and compute dose from trimmed pencil beams when using the DCS.more » Methods: MCNPX simulations were used to determine the dose distributions expected from various trimmer configurations using the DCS. Using these data, the lateral distribution for individual beamlets was modeled with a 2D asymmetric Gaussian function. The integral depth dose (IDD) of each configuration was also modeled by combining the IDD of an untrimmed pencil beam with a linear correction factor. The convolution of these two terms, along with the Highland approximation to account for lateral growth of the beam along the depth direction, allows a trimmed pencil beam dose distribution to be analytically generated. The algorithm was validated by computing dose for a single energy layer 5×5 cm{sup 2} treatment field, defined by the trimmers, using both the proposed method and MCNPX beamlets. Results: The Gaussian modeled asymmetric lateral profiles along the principal axes match the MCNPX data very well (R{sup 2}≥0.95 at the depth of the Bragg peak). For the 5×5 cm{sup 2} treatment plan created with both the modeled and MCNPX pencil beams, the passing rate of the 3D gamma test was 98% using a standard threshold of 3%/3 mm. Conclusion: An analytical method capable of accurately computing asymmetric pencil beam dose when using the DCS has been developed.« less

The radiation fields around a proton therapy facility: A comparison of Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Ottaviano, G.; Picardi, L.; Pillon, M.; Ronsivalle, C.; Sandri, S.

2014-02-01

A proton therapy test facility with a beam current lower than 10 nA in average, and an energy up to 150 MeV, is planned to be sited at the Frascati ENEA Research Center, in Italy. The accelerator is composed of a sequence of linear sections. The first one is a commercial 7 MeV proton linac, from which the beam is injected in a SCDTL (Side Coupled Drift Tube Linac) structure reaching the energy of 52 MeV. Then a conventional CCL (coupled Cavity Linac) with side coupling cavities completes the accelerator. The linear structure has the important advantage that the main radiation losses during the acceleration process occur to protons with energy below 20 MeV, with a consequent low production of neutrons and secondary radiation. From the radiation protection point of view the source of radiation for this facility is then almost completely located at the final target. Physical and geometrical models of the device have been developed and implemented into radiation transport computer codes based on the Monte Carlo method. The scope is the assessment of the radiation field around the main source for supporting the safety analysis. For the assessment independent researchers used two different Monte Carlo computer codes named FLUKA (FLUktuierende KAskade) and MCNPX (Monte Carlo N-Particle eXtended) respectively. Both are general purpose tools for calculations of particle transport and interactions with matter, covering an extended range of applications including proton beam analysis. Nevertheless each one utilizes its own nuclear cross section libraries and uses specific physics models for particle types and energies. The models implemented into the codes are described and the results are presented. The differences between the two calculations are reported and discussed pointing out disadvantages and advantages of each code in the specific application.

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

Dose conversion coefficients for neutron exposure to the lens of the human eye

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manger, Ryan P; Bellamy, Michael B; Eckerman, Keith F

Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10{sup -9} to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematicalmore » phantom in the MCNPX transport calculations.« less

Preliminary calibration of the ACP safeguards neutron counter

NASA Astrophysics Data System (ADS)

Lee, T. H.; Kim, H. D.; Yoon, J. S.; Lee, S. Y.; Swinhoe, M.; Menlove, H. O.

2007-10-01

The Advanced Spent Fuel Conditioning Process (ACP), a kind of pyroprocess, has been developed at the Korea Atomic Energy Research Institute (KAERI). Since there is no IAEA safeguards criteria for this process, KAERI has developed a neutron coincidence counter to make it possible to perform a material control and accounting (MC&A) for its ACP materials for the purpose of a transparency in the peaceful uses of nuclear materials at KAERI. The test results of the ACP Safeguards Neutron Counter (ASNC) show a satisfactory performance for the Doubles count measurement with a low measurement error for its cylindrical sample cavity. The neutron detection efficiency is about 21% with an error of ±1.32% along the axial direction of the cavity. Using two 252Cf neutron sources, we obtained various parameters for the Singles and Doubles rates for the ASNC. The Singles, Doubles, and Triples rates for a 252Cf point source were obtained by using the MCNPX code and the results for the ft8 cap multiplicity tally option with the values of ɛ, fd, and ft measured with a strong source most closely match the measurement results to within a 1% error. A preliminary calibration curve for the ASNC was generated by using the point model equation relationship between 244Cm and 252Cf and the calibration coefficient for the non-multiplying sample is 2.78×10 5 (Doubles counts/s/g 244Cm). The preliminary calibration curves for the ACP samples were also obtained by using an MCNPX simulation. A neutron multiplication influence on an increase of the Doubles rate for a metal ingot and UO2 powder is clearly observed. These calibration curves will be modified and complemented, when hot calibration samples become available. To verify the validity of this calibration curve, a measurement of spent fuel standards for a known 244Cm mass will be performed in the near future.

SU-E-T-523: On the Radiobiological Impact of Lateral Scatter in Proton Beams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heuvel, F Van den; Deruysscher, D

2014-06-01

Introduction: In proton therapy, justified concern has been voiced with respect to an increased efficiency in cell kill at the distal end of the Bragg peak. This coupled with range uncertainty is a counter indication to use the Bragg peak to define the border of a treated volume with a critical organ. An alternative is to use the lateral edge of the proton beam, obtaining more robust plans. We investigate the spectral and biological effects of the lateral scatter . Methods: A general purpose Monte Carlo simulation engine (MCNPX 2.7c) installed on a Scientific Linux cluster, calculated the dose depositionmore » spectrum of protons, knock on electrons and generated neutrons for a proton beam with maximal kinetic energy of 200MeV. Around the beam at different positions in the beam direction the spectrum is calculated in concentric rings of thickness 1cm. The deposited dose is converted to a double strand break map using an analytical expression.based on micro dosimetric calculations using a phenomenological Monte Carlo code (MCDS). A strict version of RBE is defined as the ratio of generation of double strand breaks in the different modalities. To generate the reference a Varian linac was modelled in MCNPX and the generated electron dose deposition spectrum was used . Results: On a pristine point source 200MeV beam the RBE before the Bragg peak was of the order of 1.1, increasing to 1.7 right behind the Bragg peak. When using a physically more realistic beam of 10cm diameter the effect was smaller. Both the lateral dose and RBE increased with increasing beam depth, generating a dose deposition with mixed biological effect. Conclusions: The dose deposition in proton beams need to be carefully examined because the biological effect will be different depending on the treatment geometry. Deeply penetrating proton beams generate more biologically effective lateral scatter.« less

Proton flux and radiation dose from galactic cosmic rays in the lunar regolith and implications for organic synthesis at the poles of the Moon and Mercury

NASA Astrophysics Data System (ADS)

Crites, S. T.; Lucey, P. G.; Lawrence, D. J.

2013-11-01

Galactic cosmic rays are a potential energy source to stimulate organic synthesis from simple ices. The recent detection of organic molecules at the polar regions of the Moon by LCROSS (Colaprete, A. et al. [2010]. Science 330, 463-468, http://dx.doi.org/10.1126/science.1186986), and possibly at the poles of Mercury (Paige, D.A. et al. [2013]. Science 339, 300-303, http://dx.doi.org/10.1126/science.1231106), introduces the question of whether the organics were delivered by impact or formed in situ. Laboratory experiments show that high energy particles can cause organic production from simple ices. We use a Monte Carlo particle scattering code (MCNPX) to model and report the flux of GCR protons at the surface of the Moon and report radiation dose rates and absorbed doses at the Moon’s surface and with depth as a result of GCR protons and secondary particles, and apply scaling factors to account for contributions to dose from heavier ions. We compare our results with dose rate measurements by the Cosmic Ray Telescope for the Effects of Radiation (CRaTER) experiment on Lunar Reconnaissance Orbiter (Schwadron, N.A. et al. [2012]. J. Geophys. Res. 117, E00H13, http://dx.doi.org/10.1029/2011JE003978) and find them in good agreement, indicating that MCNPX can be confidently applied to studies of radiation dose at and within the surface of the Moon. We use our dose rate calculations to conclude that organic synthesis is plausible well within the age of the lunar polar cold traps, and that organics detected at the poles of the Moon may have been produced in situ. Our dose rate calculations also indicate that galactic cosmic rays can induce organic synthesis within the estimated age of the dark deposits at the pole of Mercury that may contain organics.

Simulation of image detectors in radiology for determination of scatter-to-primary ratios using Monte Carlo radiation transport code MCNP/MCNPX.

PubMed

Smans, Kristien; Zoetelief, Johannes; Verbrugge, Beatrijs; Haeck, Wim; Struelens, Lara; Vanhavere, Filip; Bosmans, Hilde

2010-05-01

The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values. For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.

An Eye Model for Computational Dosimetry Using A Multi-Scale Voxel Phantom

NASA Astrophysics Data System (ADS)

Caracappa, Peter F.; Rhodes, Ashley; Fiedler, Derek

2014-06-01

The lens of the eye is a radiosensitive tissue with cataract formation being the major concern. Recently reduced recommended dose limits to the lens of the eye have made understanding the dose to this tissue of increased importance. Due to memory limitations, the voxel resolution of computational phantoms used for radiation dose calculations is too large to accurately represent the dimensions of the eye. A revised eye model is constructed using physiological data for the dimensions of radiosensitive tissues, and is then transformed into a high-resolution voxel model. This eye model is combined with an existing set of whole body models to form a multi-scale voxel phantom, which is used with the MCNPX code to calculate radiation dose from various exposure types. This phantom provides an accurate representation of the radiation transport through the structures of the eye. Two alternate methods of including a high-resolution eye model within an existing whole body model are developed. The accuracy and performance of each method is compared against existing computational phantoms.

Shielding properties of the ordinary concrete loaded with micro- and nano-particles against neutron and gamma radiations.

PubMed

Mesbahi, Asghar; Ghiasi, Hosein

2018-06-01

The shielding properties of ordinary concrete doped with some micro and nano scaled materials were studied in the current study. Narrow beam geometry was simulated using MCNPX Monte Carlo code and the mass attenuation coefficient of ordinary concrete doped with PbO 2 , Fe 2 O 3 , WO 3 and H 4 B (Boronium) in both nano and micro scales was calculated for photon and neutron beams. Mono-energetic beams of neutrons (100-3000 keV) and photons (142-1250 keV) were used for calculations. The concrete doped with nano-sized particles showed higher neutron removal cross section (7%) and photon attenuation coefficient (8%) relative to micro-particles. Application of nano-sized material in the composition of new concretes for dual protection against neutrons and photons are recommended. For further studies, the calculation of attenuation coefficients of these nano-concretes against higher energies of neutrons and photons and different particles are suggested. Copyright © 2018 Elsevier Ltd. All rights reserved.

124Sb-Be photo-neutron source for BNCT: Is it possible?

NASA Astrophysics Data System (ADS)

Golshanian, Mohadeseh; Rajabi, Ali Akbar; Kasesaz, Yaser

2016-11-01

In this research a computational feasibility study has been done on the use of 124SbBe photo-neutron source for Boron Neutron Capture Therapy (BNCT) using MCNPX Monte Carlo code. For this purpose, a special beam shaping assembly has been designed to provide an appropriate epithermal neutron beam suitable for BNCT. The final result shows that using 150 kCi of 124Sb, the epithermal neutron flux at the designed beam exit is 0.23×109 (n/cm2 s). In-phantom dose analysis indicates that treatment time for a brain tumor is about 40 min which is a reasonable time. This high activity 124Sb could be achieved using three 50 kCi rods of 124Sb which can be produced in a research reactor. It is clear, that as this activity is several hundred times the activity of a typical cobalt radiotherapy source, issues related to handling, safety and security must be addressed.

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

NOTE: Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool

NASA Astrophysics Data System (ADS)

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-01

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool.

PubMed

Courageot, Estelle; Sayah, Rima; Huet, Christelle

2010-05-07

Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

Angular distribution measurements of photo-neutron yields produced by 2.0 GeV electrons incident on thick targets.

PubMed

Lee, Hee-Seock; Ban, Syuichi; Sanami, Toshiya; Takahashi, Kazutoshi; Sato, Tatsuhiko; Shin, Kazuo; Chung, Chinwha

2005-01-01

A study of differential photo-neutron yields by irradiation with 2 GeV electrons has been carried out. In this extension of a previous study in which measurements were made at an angle of 90 degrees relative to incident electrons, the differential photo-neutron yield was obtained at two other angles, 48 degrees and 140 degrees, to study its angular characteristics. Photo-neutron spectra were measured using a pulsed beam time-of-flight method and a BC418 plastic scintillator. The reliable range of neutron energy measurement was 8-250 MeV. The neutron spectra were measured for 10 Xo-thick Cu, Sn, W and Pb targets. The angular distribution characteristics, together with the previous results for 90 degrees, are presented in the study. The experimental results are compared with Monte Carlo calculation results. The yields predicted by MCNPX 2.5 tend to underestimate the measured ones. The same trend holds for the comparison results using the EGS4 and PICA3 codes.

In field application of differential Die-Away time technique for detecting gram quantities of fissile materials

NASA Astrophysics Data System (ADS)

Remetti, Romolo; Gandolfo, Giada; Lepore, Luigi; Cherubini, Nadia

2017-10-01

In the frame of Chemical, Biological, Radiological, and Nuclear defense European activities, the ENEA, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development, is proposing the Neutron Active Interrogation system (NAI), a device designed to find transuranic-based Radioactive Dispersal Devices hidden inside suspected packages. It is based on Differential Die-Away time Analysis, an active neutron technique targeted in revealing the presence of fissile material through detection of induced fission neutrons. Several Monte Carlo simulations, carried out by MCNPX code, and the development of ad-hoc design methods, have led to the realization of a first prototype based on a 14 MeV d-t neutron generator coupled with a tailored moderating structure, and an array of helium-3 neutron detectors. The complete system is characterized by easy transportability, light weight, and real-time response. First results have shown device's capability to detect gram quantities of fissile materials.

Thermal neutron calibration channel at LNMRI/IRD.

PubMed

Astuto, A; Salgado, A P; Leite, S P; Patrão, K C S; Fonseca, E S; Pereira, W W; Lopes, R T

2014-10-01

The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four (241)Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

IMPROVEMENTS IN THE THERMAL NEUTRON CALIBRATION UNIT, TNF2, AT LNMRI/IRD.

PubMed

Astuto, A; Fernandes, S S; Patrão, K C S; Fonseca, E S; Pereira, W W; Lopes, R T

2018-02-21

The standard thermal neutron flux unit, TNF2, in the Brazilian National Ionizing Radiation Metrology Laboratory was rebuilt. Fluence is still achieved by moderating of four 241Am-Be sources with 0.6 TBq each. The facility was again simulated and redesigned with graphite core and paraffin added graphite blocks surrounding it. Simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The resulting neutron fluence quality in terms of intensity, spectrum and cadmium ratio was evaluated. After this step, the system was assembled based on the results obtained from the simulations and measurements were performed with equipment existing in LNMRI/IRD and by simulated equipment. This work focuses on the characterization of a central chamber point and external points around the TNF2 in terms of neutron spectrum, fluence and ambient dose equivalent, H*(10). This system was validated with spectra measurements, fluence and H*(10) to ensure traceability.

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

PubMed

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

Spent nuclear fuel/water interface behavior: Alpha dose rate profile determination for model surfaces and microcracks by using Monte-Carlo methods

NASA Astrophysics Data System (ADS)

Tribet, M.; Mougnaud, S.; Jégou, C.

2017-05-01

This work aims to better understand the nature and evolution of energy deposits at the UO2/water reactional interface subjected to alpha irradiation, through an original approach based on Monte-Carlo-type simulations, using the MCNPX code. Such an approach has the advantage of describing the energy deposit profiles on both sides of the interface (UO2 and water). The calculations have been performed on simple geometries, with data from an irradiated UOX fuel (burnup of 47 GWd.tHM-1 and 15 years of alpha decay). The influence of geometric parameters such as the diameter and the calculation steps at the reactional interface are discussed, and the exponential laws to be used in practice are suggested. The case of cracks with various different apertures (from 5 to 35 μm) has also been examined and these calculations have also enabled new information on the mean range of radiolytic species in cracks, and thus on the local chemistry.

Computational Fluid Dynamics.

DTIC Science & Technology

1986-06-01

this area concentrated first on the compu- tation of foldpoints by means of one of the many possible forms of augmented equations. One approach in this...this area is, of course, the deter- mination of the principal characteristic features of the solution manifold of the given parametrized equation. This...into Spanish and appeared in Metodos Numericos Para Calculo y Diseno en Ingeniera 1 (1985), 67-80. [5] J. P. Fink and W. C. Rheinboldt, "Folds on the

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lafleur, Adrienne M.; Ulrich, Timothy J. II; Menlove, Howard O.

Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio wasmore » more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.« less

Ion Exchange and Its Connection to Industry. II. Calculation Methods for Installations; EL CAMBIO DE ION Y SU RELACION CON LA INDUSTRIA. II. METHODOS DE CALCULO DE INSTALACIONES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hueda, A.U.; Perez, B.L.; Jodra, L.G.

1960-01-01

A presentation is made of calculation methods for ionexchange installations based on kinetic considerations and similarity with other unit operations. Factors to be experimentally obtained as well as difficulties which may occur in its determination are also given. Calculation procedures most commonly used in industry are enclosed and explain with numerical resolution of a problem of water demineralization. (auth)

Cost-Effective Optimization of Rubble-Mound Breakwater Cross Sections.

DTIC Science & Technology

1986-02-01

of a Conference Held in London, London, England, p 20. Iribarren, Cavanilles R. 1938. "Una formula para el calculo de los diques de escollera," M...NAME AND ADDRESS 10. PROGRAM ELEMENT PROJECT, TASK" AREA & WORK UNIT NUMBERS US Army Engineer Waterways Experiment Station Civil Works Research...purpose as a wave barrier. A breakwater protecting a harbor entrance and mooring area from wave attack might serve q 6 to divert currents and longshore

Topics in computational physics

NASA Astrophysics Data System (ADS)

Monville, Maura Edelweiss

Computational Physics spans a broad range of applied fields extending beyond the border of traditional physics tracks. Demonstrated flexibility and capability to switch to a new project, and pick up the basics of the new field quickly, are among the essential requirements for a computational physicist. In line with the above mentioned prerequisites, my thesis described the development and results of two computational projects belonging to two different applied science areas. The first project is a Materials Science application. It is a prescription for an innovative nano-fabrication technique that is built out of two other known techniques. The preliminary results of the simulation of this novel nano-patterning fabrication method show an average improvement, roughly equal to 18%, with respect to the single techniques it draws on. The second project is a Homeland Security application aimed at preventing smuggling of nuclear material at ports of entry. It is concerned with a simulation of an active material interrogation system based on the analysis of induced photo-nuclear reactions. This project consists of a preliminary evaluation of the photo-fission implementation in the more robust radiation transport Monte Carlo codes, followed by the customization and extension of MCNPX, a Monte Carlo code developed in Los Alamos National Laboratory, and MCNP-PoliMi. The final stage of the project consists of testing the interrogation system against some real world scenarios, for the purpose of determining the system's reliability, material discrimination power, and limitations.

Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Henzl, Vladimir; Swinhoe, Martyn Thomas

2015-01-14

Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDAmore » instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.« less

The Coast Artillery Journal. Volume 59, Number 6, December 1923

DTIC Science & Technology

1923-12-01

Remarques sur Ie tir Fusant.-F-12, January, 1923. RegIa de Calculo Para Muelles Cilindricos.-Spa-2, June, 1923. Terrain Reduit pour Exercices de tir Fictif au...I to V, on a front Brussels-Metz, pivoting on the fortified area Metz- Thionville. The inner flank of the wheel was to be covered by the VI and VII...rapidly on the fron t Beauraing-Gedinne- Paliseul- Fay- des -Veneurs-Cuignon (5th Army) and Tetaigne-Margut-Quincy (4th Army)." Fearing the German advance

Stopping power and dose calculations with analytical and Monte Carlo methods for protons and prompt gamma range verification

NASA Astrophysics Data System (ADS)

Usta, Metin; Tufan, Mustafa Çağatay; Aydın, Güral; Bozkurt, Ahmet

2018-07-01

In this study, we have performed the calculations stopping power, depth dose, and range verification for proton beams using dielectric and Bethe-Bloch theories and FLUKA, Geant4 and MCNPX Monte Carlo codes. In the framework, as analytical studies, Drude model was applied for dielectric theory and effective charge approach with Roothaan-Hartree-Fock charge densities was used in Bethe theory. In the simulations different setup parameters were selected to evaluate the performance of three distinct Monte Carlo codes. The lung and breast tissues were investigated are considered to be related to the most common types of cancer throughout the world. The results were compared with each other and the available data in literature. In addition, the obtained results were verified with prompt gamma range data. In both stopping power values and depth-dose distributions, it was found that the Monte Carlo values give better results compared with the analytical ones while the results that agree best with ICRU data in terms of stopping power are those of the effective charge approach between the analytical methods and of the FLUKA code among the MC packages. In the depth dose distributions of the examined tissues, although the Bragg curves for Monte Carlo almost overlap, the analytical ones show significant deviations that become more pronounce with increasing energy. Verifications with the results of prompt gamma photons were attempted for 100-200 MeV protons which are regarded important for proton therapy. The analytical results are within 2%-5% and the Monte Carlo values are within 0%-2% as compared with those of the prompt gammas.

Monte Carlo calculations of positron emitter yields in proton radiotherapy.

PubMed

Seravalli, E; Robert, C; Bauer, J; Stichelbaut, F; Kurz, C; Smeets, J; Van Ngoc Ty, C; Schaart, D R; Buvat, I; Parodi, K; Verhaegen, F

2012-03-21

Positron emission tomography (PET) is a promising tool for monitoring the three-dimensional dose distribution in charged particle radiotherapy. PET imaging during or shortly after proton treatment is based on the detection of annihilation photons following the ß(+)-decay of radionuclides resulting from nuclear reactions in the irradiated tissue. Therapy monitoring is achieved by comparing the measured spatial distribution of irradiation-induced ß(+)-activity with the predicted distribution based on the treatment plan. The accuracy of the calculated distribution depends on the correctness of the computational models, implemented in the employed Monte Carlo (MC) codes that describe the interactions of the charged particle beam with matter and the production of radionuclides and secondary particles. However, no well-established theoretical models exist for predicting the nuclear interactions and so phenomenological models are typically used based on parameters derived from experimental data. Unfortunately, the experimental data presently available are insufficient to validate such phenomenological hadronic interaction models. Hence, a comparison among the models used by the different MC packages is desirable. In this work, starting from a common geometry, we compare the performances of MCNPX, GATE and PHITS MC codes in predicting the amount and spatial distribution of proton-induced activity, at therapeutic energies, to the already experimentally validated PET modelling based on the FLUKA MC code. In particular, we show how the amount of ß(+)-emitters produced in tissue-like media depends on the physics model and cross-sectional data used to describe the proton nuclear interactions, thus calling for future experimental campaigns aiming at supporting improvements of MC modelling for clinical application of PET monitoring. © 2012 Institute of Physics and Engineering in Medicine

Comparison of depth-dose distributions of proton therapeutic beams calculated by means of logical detectors and ionization chamber modeled in Monte Carlo codes

NASA Astrophysics Data System (ADS)

Pietrzak, Robert; Konefał, Adam; Sokół, Maria; Orlef, Andrzej

2016-08-01

The success of proton therapy depends strongly on the precision of treatment planning. Dose distribution in biological tissue may be obtained from Monte Carlo simulations using various scientific codes making it possible to perform very accurate calculations. However, there are many factors affecting the accuracy of modeling. One of them is a structure of objects called bins registering a dose. In this work the influence of bin structure on the dose distributions was examined. The MCNPX code calculations of Bragg curve for the 60 MeV proton beam were done in two ways: using simple logical detectors being the volumes determined in water, and using a precise model of ionization chamber used in clinical dosimetry. The results of the simulations were verified experimentally in the water phantom with Marcus ionization chamber. The average local dose difference between the measured relative doses in the water phantom and those calculated by means of the logical detectors was 1.4% at first 25 mm, whereas in the full depth range this difference was 1.6% for the maximum uncertainty in the calculations less than 2.4% and for the maximum measuring error of 1%. In case of the relative doses calculated with the use of the ionization chamber model this average difference was somewhat greater, being 2.3% at depths up to 25 mm and 2.4% in the full range of depths for the maximum uncertainty in the calculations of 3%. In the dose calculations the ionization chamber model does not offer any additional advantages over the logical detectors. The results provided by both models are similar and in good agreement with the measurements, however, the logical detector approach is a more time-effective method.

Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

NASA Astrophysics Data System (ADS)

Tsinganis, A.; Kokkoris, M.; Vlastou, R.; Kalamara, A.; Stamatopoulos, A.; Kanellakopoulos, A.; Lagoyannis, A.; Axiotis, M.

2017-09-01

Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of 234U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on `microbulk' Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The 235U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the `Demokritos' National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

Analytical functions for beta and gamma absorbed fractions of iodine-131 in spherical and ellipsoidal volumes.

PubMed

Mowlavi, Ali Asghar; Fornasier, Maria Rossa; Mirzaei, Mohammd; Bregant, Paola; de Denaro, Mario

2014-10-01

The beta and gamma absorbed fractions in organs and tissues are the important key factors of radionuclide internal dosimetry based on Medical Internal Radiation Dose (MIRD) approach. The aim of this study is to find suitable analytical functions for beta and gamma absorbed fractions in spherical and ellipsoidal volumes with a uniform distribution of iodine-131 radionuclide. MCNPX code has been used to calculate the energy absorption from beta and gamma rays of iodine-131 uniformly distributed inside different ellipsoids and spheres, and then the absorbed fractions have been evaluated. We have found the fit parameters of a suitable analytical function for the beta absorbed fraction, depending on a generalized radius for ellipsoid based on the radius of sphere, and a linear fit function for the gamma absorbed fraction. The analytical functions that we obtained from fitting process in Monte Carlo data can be used for obtaining the absorbed fractions of iodine-131 beta and gamma rays for any volume of the thyroid lobe. Moreover, our results for the spheres are in good agreement with the results of MIRD and other scientific literatures.

Characterization of a tin-loaded liquid scintillator for gamma spectroscopy and neutron detection

NASA Astrophysics Data System (ADS)

Wen, Xianfei; Harvey, Taylor; Weinmann-Smith, Robert; Walker, James; Noh, Young; Farley, Richard; Enqvist, Andreas

2018-07-01

A tin-loaded liquid scintillator has been developed for gamma spectroscopy and neutron detection. The scintillator was characterized in regard to energy resolution, pulse shape discrimination, neutron light output function, and timing resolution. The loading of tin into scintillators with low effective atomic number was demonstrated to provide photopeaks with acceptable energy resolution. The scintillator was shown to have reasonable neutron/gamma discrimination capability based on the charge comparison method. The effect on the discrimination quality of the total charge integration time and the initial delay time for tail charge integration was studied. To obtain the neutron light output function, the time-of-flight technique was utilized with a 252Cf source. The light output function was validated with the MCNPX-PoliMi code by comparing the measured and simulated pule height spectra. The timing resolution of the developed scintillator was also evaluated. The tin-loading was found to have negligible impact on the scintillation decay times. However, a relatively large degradation of timing resolution was observed due to the reduced light yield.

INVESTIGATION OF THE INFLUENCE OF THE POSITION INSIDE A SMALL AIRCRAFT ON THE COSMIC-RADIATION-INDUCED DOSE.

PubMed

Prado, A C M; Pazianotto, M T; Gonçalez, O L; Dos Santos, L R; Caldeira, A D; Pereira, H H C; Hubert, G; Federico, C A

2017-11-01

This article report the measurements on-board a small aircraft at the same altitude and around the same geographic coordinates. The measurements of Ambient Dose Equivalent Rate (H*(10)) were performed in several positions inside the aircraft, close and far from the pilot location and the discrimination between neutron and non-neutron components. The results show that the neutrons are attenuated close to fuel depots and the non-neutron component appears to have the opposite behavior inside the aircraft. These experimental results are also confronted with results from Monte Carlo simulation, obtained with the MCNPX code, using a simplified model of the Learjet-type aircraft and a modeling of the standard atmosphere, which reproduces the real energy and angular distribution of the particles. The Monte Carlo simulation agreed with the experimental measurements and shows that the total H*(10) presents small variation (around 1%) between the positions inside aircraft, although the neutron spectra present significant variations. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Evaluation of the Effect of Source Geometry on the Output of Miniature X-ray Tube for Electronic Brachytherapy through Simulation

PubMed Central

Barati, B.; Zabihzadeh, M.; Tahmasebi Birgani, M.J.; Chegini, N.; Fatahiasl, J.; Mirr, I.

2018-01-01

Objective: The use of miniature X-ray source in electronic brachytherapy is on the rise so there is an urgent need to acquire more knowledge on X-ray spectrum production and distribution by a dose. The aim of this research was to investigate the influence of target thickness and geometry at the source of miniature X-ray tube on tube output. Method: Five sources were simulated based on problems each with a specific geometric structure and conditions using MCNPX code. Tallies proportional to the output were used to calculate the results for the influence of source geometry on output. Results: The results of this work include the size of the optimal thickness of 5 miniature sources, energy spectrum of the sources per 50 kev and also the axial and transverse dose of simulated sources were calculated based on these thicknesses. The miniature source geometric was affected on the output x-ray tube. Conclusion: The result of this study demonstrates that hemispherical-conical, hemispherical and truncated-conical miniature sources were determined as the most suitable tools. PMID:29732338

Effect of the size of experimental channels of the lead slowing-down spectrometer SVZ-100 (Institute for Nuclear Research, Moscow) on the moderation constant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Latysheva, L. N.; Bergman, A. A.; Sobolevsky, N. M., E-mail: sobolevs@inr.ru

Lead slowing-down (LSD) spectrometers have a low energy resolution (about 30%), but their luminosity is 10{sup 3} to 10{sup 4} times higher than that of time-of-flight (TOF) spectrometers. A high luminosity of LSD spectrometers makes it possible to use them to measure neutron cross section for samples of mass about several micrograms. These features specify a niche for the application of LSD spectrometers in measuring neutron cross sections for elements hardly available in macroscopic amounts-in particular, for actinides. A mathematical simulation of the parameters of SVZ-100 LSD spectrometer of the Institute for Nuclear Research (INR, Moscow) is performed in themore » present study on the basis of the MCNPX code. It is found that the moderation constant, which is the main parameter of LSD spectrometers, is highly sensitive to the size and shape of detecting volumes in calculations and, hence, to the real size of experimental channels of the LSD spectrometer.« less

Three-dimensional Monte Carlo calculation of some nuclear parameters

NASA Astrophysics Data System (ADS)

Günay, Mehtap; Şeker, Gökmen

2017-09-01

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Contributions of each isotope in some fluids on neutronic performance in a fusion-fission hybrid reactor: a Monte Carlo method

NASA Astrophysics Data System (ADS)

Günay, M.; Şarer, B.; Kasap, H.

2014-08-01

In the present investigation, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa ferritic steel structural material and 99-95 % Li20Sn80-1-5 % SFG-Pu, 99-95 % Li20Sn80-1-5 % SFG-PuF4, 99-95 % Li20Sn80-1-5 % SFG-PuO2 the molten salt-heavy metal mixtures, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. The contributions of each isotope in fluids on the nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional analyses were performed by using Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

Advances in the computation of the Sjöstrand, Rossi, and Feynman distributions

DOE PAGES

Talamo, A.; Gohar, Y.; Gabrielli, F.; ...

2017-02-01

This study illustrates recent computational advances in the application of the Sjöstrand (area), Rossi, and Feynman methods to estimate the effective multiplication factor of a subcritical system driven by an external neutron source. The methodologies introduced in this study have been validated with the experimental results from the KUKA facility of Japan by Monte Carlo (MCNP6 and MCNPX) and deterministic (ERANOS, VARIANT, and PARTISN) codes. When the assembly is driven by a pulsed neutron source generated by a particle accelerator and delayed neutrons are at equilibrium, the Sjöstrand method becomes extremely fast if the integral of the reaction rate frommore » a single pulse is split into two parts. These two integrals distinguish between the neutron counts during and after the pulse period. To conclude, when the facility is driven by a spontaneous fission neutron source, the timestamps of the detector neutron counts can be obtained up to the nanosecond precision using MCNP6, which allows obtaining the Rossi and Feynman distributions.« less

Monte Carlo simulations in multi-detector CT (MDCT) for two PET/CT scanner models using MASH and FASH adult phantoms

NASA Astrophysics Data System (ADS)

Belinato, W.; Santos, W. S.; Paschoal, C. M. M.; Souza, D. N.

2015-06-01

The combination of positron emission tomography (PET) and computed tomography (CT) has been extensively used in oncology for diagnosis and staging of tumors, radiotherapy planning and follow-up of patients with cancer, as well as in cardiology and neurology. This study determines by the Monte Carlo method the internal organ dose deposition for computational phantoms created by multidetector CT (MDCT) beams of two PET/CT devices operating with different parameters. The different MDCT beam parameters were largely related to the total filtration that provides a beam energetic change inside the gantry. This parameter was determined experimentally with the Accu-Gold Radcal measurement system. The experimental values of the total filtration were included in the simulations of two MCNPX code scenarios. The absorbed organ doses obtained in MASH and FASH phantoms indicate that bowtie filter geometry and the energy of the X-ray beam have significant influence on the results, although this influence can be compensated by adjusting other variables such as the tube current-time product (mAs) and pitch during PET/CT procedures.

Using the Monte Carlo technique to calculate dose conversion coefficients for medical professionals in interventional radiology

NASA Astrophysics Data System (ADS)

Santos, W. S.; Carvalho, A. B., Jr.; Hunt, J. G.; Maia, A. F.

2014-02-01

The objective of this study was to estimate doses in the physician and the nurse assistant at different positions during interventional radiology procedures. In this study, effective doses obtained for the physician and at points occupied by other workers were normalised by air kerma-area product (KAP). The simulations were performed for two X-ray spectra (70 kVp and 87 kVp) using the radiation transport code MCNPX (version 2.7.0), and a pair of anthropomorphic voxel phantoms (MASH/FASH) used to represent both the patient and the medical professional at positions from 7 cm to 47 cm from the patient. The X-ray tube was represented by a point source positioned in the anterior posterior (AP) and posterior anterior (PA) projections. The CC can be useful to calculate effective doses, which in turn are related to stochastic effects. With the knowledge of the values of CCs and KAP measured in an X-ray equipment, at a similar exposure, medical professionals will be able to know their own effective dose.

Two-dimensional dosimetry of radiotherapeutical proton beams using thermoluminescence foils.

PubMed

Czopyk, L; Klosowski, M; Olko, P; Swakon, J; Waligorski, M P R; Kajdrowicz, T; Cuttone, G; Cirrone, G A P; Di Rosa, F

2007-01-01

In modern radiation therapy such as intensity modulated radiation therapy or proton therapy, one is able to cover the target volume with improved dose conformation and to spare surrounding tissue with help of modern measurement techniques. Novel thermoluminescence dosimetry (TLD) foils, developed from the hot-pressed mixture of LiF:Mg,Cu,P (MCP TL) powder and ethylene-tetrafluoroethylene (ETFE) copolymer, have been applied for 2-D dosimetry of radiotherapeutical proton beams at INFN Catania and IFJ Krakow. A TLD reader with 70 mm heating plate and CCD camera was used to read the 2-D emission pattern of irradiated foils. The absorbed dose profiles were evaluated, taking into account correction factors specific for TLD such as dose and energy response. TLD foils were applied for measuring of dose distributions within an eye phantom and compared with predictions obtained from the MCNPX code and Eclipse Ocular Proton Planning (Varian Medical Systems) clinical radiotherapy planning system. We demonstrate the possibility of measuring 2-D dose distributions with point resolution of about 0.5 x 0.5 mm(2).

An Improved Elastic and Nonelastic Neutron Transport Algorithm for Space Radiation

NASA Technical Reports Server (NTRS)

Clowdsley, Martha S.; Wilson, John W.; Heinbockel, John H.; Tripathi, R. K.; Singleterry, Robert C., Jr.; Shinn, Judy L.

2000-01-01

A neutron transport algorithm including both elastic and nonelastic particle interaction processes for use in space radiation protection for arbitrary shield material is developed. The algorithm is based upon a multiple energy grouping and analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. The algorithm is then coupled to the Langley HZETRN code through a bidirectional neutron evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for an aluminum water shield-target configuration is then compared with MCNPX and LAHET Monte Carlo calculations for the same shield-target configuration. With the Monte Carlo calculation as a benchmark, the algorithm developed in this paper showed a great improvement in results over the unmodified HZETRN solution. In addition, a high-energy bidirectional neutron source based on a formula by Ranft showed even further improvement of the fluence results over previous results near the front of the water target where diffusion out the front surface is important. Effects of improved interaction cross sections are modest compared with the addition of the high-energy bidirectional source terms.

HDRK-Man: a whole-body voxel model based on high-resolution color slice images of a Korean adult male cadaver

NASA Astrophysics Data System (ADS)

Kim, Chan Hyeong; Hyoun Choi, Sang; Jeong, Jong Hwi; Lee, Choonsik; Chung, Min Suk

2008-08-01

A Korean voxel model, named 'High-Definition Reference Korean-Man (HDRK-Man)', was constructed using high-resolution color photographic images that were obtained by serially sectioning the cadaver of a 33-year-old Korean adult male. The body height and weight, the skeletal mass and the dimensions of the individual organs and tissues were adjusted to the reference Korean data. The resulting model was then implemented into a Monte Carlo particle transport code, MCNPX, to calculate the dose conversion coefficients for the internal organs and tissues. The calculated values, overall, were reasonable in comparison with the values from other adult voxel models. HDRK-Man showed higher dose conversion coefficients than other models, due to the facts that HDRK-Man has a smaller torso and that the arms of HDRK-Man are shifted backward. The developed model is believed to adequately represent average Korean radiation workers and thus can be used for more accurate calculation of dose conversion coefficients for Korean radiation workers in the future.

Depth profile of production yields of natPb(p, xn) 206,205,204,203,202,201Bi nuclear reactions

NASA Astrophysics Data System (ADS)

Mokhtari Oranj, Leila; Jung, Nam-Suk; Kim, Dong-Hyun; Lee, Arim; Bae, Oryun; Lee, Hee-Seock

2016-11-01

Experimental and simulation studies on the depth profiles of production yields of natPb(p, xn) 206,205,204,203,202,201Bi nuclear reactions were carried out. Irradiation experiments were performed at the high-intensity proton linac facility (KOMAC) in Korea. The targets, irradiated by 100-MeV protons, were arranged in a stack consisting of natural Pb, Al, Au foils and Pb plates. The proton beam intensity was determined by activation analysis method using 27Al(p, 3p1n)24Na, 197Au(p, p1n)196Au, and 197Au(p, p3n)194Au monitor reactions and also by Gafchromic film dosimetry method. The yields of produced radio-nuclei in the natPb activation foils and monitor foils were measured by HPGe spectroscopy system. Monte Carlo simulations were performed by FLUKA, PHITS/DCHAIN-SP, and MCNPX/FISPACT codes and the calculated data were compared with the experimental results. A satisfactory agreement was observed between the present experimental data and the simulations.

Compact DD generator-based neutron activation analysis (NAA) system to determine fluorine in human bone in vivo: a feasibility study.

PubMed

Mostafaei, Farshad; Blake, Scott P; Liu, Yingzi; Sowers, Daniel A; Nie, Linda H

2015-10-01

The subject of whether fluorine (F) is detrimental to human health has been controversial for many years. Much of the discussion focuses on the known benefits and detriments to dental care and problems that F causes in bone structure at high doses. It is therefore advantageous to have the means to monitor F concentrations in the human body as a method to directly assess exposure. F accumulates in the skeleton making bone a useful biomarker to assess long term cumulative exposure to F. This study presents work in the development of a non-invasive method for the monitoring of F in human bone. The work was based on the technique of in vivo neutron activation analysis (IVNAA). A compact deuterium-deuterium (DD) generator was used to produce neutrons. A moderator/reflector/shielding assembly was designed and built for human hand irradiation. The gamma rays emitted through the (19)F(n,γ)(20)F reaction were measured using a HPGe detector. This study was undertaken to (i) find the feasibility of using DD system to determine F in human bone, (ii) estimate the F minimum detection limit (MDL), and (iii) optimize the system using the Monte Carlo N-Particle eXtended (MCNPX) code in order to improve the MDL of the system. The F MDL was found to be 0.54 g experimentally with a neutron flux of 7   ×   10(8) n s(-1) and an optimized irradiation, decay, and measurement time scheme. The numbers of F counts from the experiment were found to be close to the (MCNPX) simulation results with the same irradiation and detection parameters. The equivalent dose to the irradiated hand and the effective dose to the whole body were found to be 0.9 mSv and 0.33 μSv, respectively. Based on these results, it is feasible to develop a compact DD generator based IVNAA system to measure bone F in a population with moderate to high F exposure.

SU-E-CAMPUS-I-02: Estimation of the Dosimetric Error Caused by the Voxelization of Hybrid Computational Phantoms Using Triangle Mesh-Based Monte Carlo Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C; Badal, A

Purpose: Computational voxel phantom provides realistic anatomy but the voxel structure may result in dosimetric error compared to real anatomy composed of perfect surface. We analyzed the dosimetric error caused from the voxel structure in hybrid computational phantoms by comparing the voxel-based doses at different resolutions with triangle mesh-based doses. Methods: We incorporated the existing adult male UF/NCI hybrid phantom in mesh format into a Monte Carlo transport code, penMesh that supports triangle meshes. We calculated energy deposition to selected organs of interest for parallel photon beams with three mono energies (0.1, 1, and 10 MeV) in antero-posterior geometry. Wemore » also calculated organ energy deposition using three voxel phantoms with different voxel resolutions (1, 5, and 10 mm) using MCNPX2.7. Results: Comparison of organ energy deposition between the two methods showed that agreement overall improved for higher voxel resolution, but for many organs the differences were small. Difference in the energy deposition for 1 MeV, for example, decreased from 11.5% to 1.7% in muscle but only from 0.6% to 0.3% in liver as voxel resolution increased from 10 mm to 1 mm. The differences were smaller at higher energies. The number of photon histories processed per second in voxels were 6.4×10{sup 4}, 3.3×10{sup 4}, and 1.3×10{sup 4}, for 10, 5, and 1 mm resolutions at 10 MeV, respectively, while meshes ran at 4.0×10{sup 4} histories/sec. Conclusion: The combination of hybrid mesh phantom and penMesh was proved to be accurate and of similar speed compared to the voxel phantom and MCNPX. The lowest voxel resolution caused a maximum dosimetric error of 12.6% at 0.1 MeV and 6.8% at 10 MeV but the error was insignificant in some organs. We will apply the tool to calculate dose to very thin layer tissues (e.g., radiosensitive layer in gastro intestines) which cannot be modeled by voxel phantoms.« less

Modelling of aircrew radiation exposure during solar particle events

NASA Astrophysics Data System (ADS)

Al Anid, Hani Khaled

In 1990, the International Commission on Radiological Protection recognized the occupational exposure of aircrew to cosmic radiation. In Canada, a Commercial and Business Aviation Advisory Circular was issued by Transport Canada suggesting that action should be taken to manage such exposure. In anticipation of possible regulations on exposure of Canadian-based aircrew in the near future, an extensive study was carried out at the Royal Military College of Canada to measure the radiation exposure during commercial flights. The radiation exposure to aircrew is a result of a complex mixed-radiation field resulting from Galactic Cosmic Rays (GCRs) and Solar Energetic Particles (SEPs). Supernova explosions and active galactic nuclei are responsible for GCRs which consist of 90% protons, 9% alpha particles, and 1% heavy nuclei. While they have a fairly constant fluence rate, their interaction with the magnetic field of the Earth varies throughout the solar cycles, which has a period of approximately 11 years. SEPs are highly sporadic events that are associated with solar flares and coronal mass ejections. This type of exposure may be of concern to certain aircrew members, such as pregnant flight crew, for which the annual effective dose is limited to 1 mSv over the remainder of the pregnancy. The composition of SEPs is very similar to GCRs, in that they consist of mostly protons, some alpha particles and a few heavy nuclei, but with a softer energy spectrum. An additional factor when analysing SEPs is the effect of flare anisotropy. This refers to the way charged particles are transported through the Earth's magnetosphere in an anisotropic fashion. Solar flares that are fairly isotropic produce a uniform radiation exposure for areas that have similar geomagnetic shielding, while highly anisotropic events produce variable exposures at different locations on the Earth. Studies of neutron monitor count rates from detectors sharing similar geomagnetic shielding properties show a very different response during anisotropic events, leading to variations in aircrew radiation doses that may be significant for dose assessment. To estimate the additional exposure due to solar flares, a model was developed using a Monte-Carlo radiation transport code, MCNPX. The model transports an extrapolated particle spectrum based on satellite measurements through the atmosphere using the MCNPX analysis. This code produces the estimated flux at a specific altitude where radiation dose conversion coefficients are applied to convert the particle flux into effective and ambient dose-equivalent rates. A cut-off rigidity model accounts for the shielding effects of the Earth's magnetic field. Comparisons were made between the model predictions and actual flight measurements taken with various types of instruments used to measure the mixed radiation field during Ground Level Enhancements 60 and 65. An anisotropy analysis that uses neutron monitor responses and the pitch angle distribution of energetic solar particles was used to identify particle anisotropy for a solar event in December 2006. In anticipation of future commercial use, a computer code has been developed to implement the radiation dose assessment model for routine analysis. Keywords: Radiation Dosimetry, Radiation Protection, Space Physics.

Monte Carlo Analysis of Pion Contribution to Absorbed Dose from Galactic Cosmic Rays

NASA Technical Reports Server (NTRS)

Aghara, S.K.; Battnig, S.R.; Norbury, J.W.; Singleterry, R.C.

2009-01-01

Accurate knowledge of the physics of interaction, particle production and transport is necessary to estimate the radiation damage to equipment used on spacecraft and the biological effects of space radiation. For long duration astronaut missions, both on the International Space Station and the planned manned missions to Moon and Mars, the shielding strategy must include a comprehensive knowledge of the secondary radiation environment. The distribution of absorbed dose and dose equivalent is a function of the type, energy and population of these secondary products. Galactic cosmic rays (GCR) comprised of protons and heavier nuclei have energies from a few MeV per nucleon to the ZeV region, with the spectra reaching flux maxima in the hundreds of MeV range. Therefore, the MeV - GeV region is most important for space radiation. Coincidentally, the pion production energy threshold is about 280 MeV. The question naturally arises as to how important these particles are with respect to space radiation problems. The space radiation transport code, HZETRN (High charge (Z) and Energy TRaNsport), currently used by NASA, performs neutron, proton and heavy ion transport explicitly, but it does not take into account the production and transport of mesons, photons and leptons. In this paper, we present results from the Monte Carlo code MCNPX (Monte Carlo N-Particle eXtended), showing the effect of leptons and mesons when they are produced and transported in a GCR environment.

Optimization of radioactive sources to achieve the highest precision in three-phase flow meters using Jaya algorithm.

PubMed

Roshani, G H; Karami, A; Khazaei, A; Olfateh, A; Nazemi, E; Omidi, M

2018-05-17

Gamma ray source has very important role in precision of multi-phase flow metering. In this study, different combination of gamma ray sources (( 133 Ba- 137 Cs), ( 133 Ba- 60 Co), ( 241 Am- 137 Cs), ( 241 Am- 60 Co), ( 133 Ba- 241 Am) and ( 60 Co- 137 Cs)) were investigated in order to optimize the three-phase flow meter. Three phases were water, oil and gas and the regime was considered annular. The required data was numerically generated using MCNP-X code which is a Monte-Carlo code. Indeed, the present study devotes to forecast the volume fractions in the annular three-phase flow, based on a multi energy metering system including various radiation sources and also one NaI detector, using a hybrid model of artificial neural network and Jaya Optimization algorithm. Since the summation of volume fractions is constant, a constraint modeling problem exists, meaning that the hybrid model must forecast only two volume fractions. Six hybrid models associated with the number of used radiation sources are designed. The models are employed to forecast the gas and water volume fractions. The next step is to train the hybrid models based on numerically obtained data. The results show that, the best forecast results are obtained for the gas and water volume fractions of the system including the ( 241 Am- 137 Cs) as the radiation source. Copyright © 2018 Elsevier Ltd. All rights reserved.

Dosimetric factors for diagnostic nuclear medicine procedures in a non-reference pregnant phantom.

PubMed

Rafat-Motavalli, Laleh; Miri Hakimabad, Hashem; Hoseinian Azghadi, Elie

2018-05-01

This study was evaluated the impact of using non-reference fetal models on the fetal radiation dose from diagnostic radionuclide administration. The 6 month pregnant phantoms including fetal models at 10th and 90th growth percentiles were constructed at either end of the normal range around the 50th percentile and implemented in the Monte Carlo N-Particle code version MCNPX 2.6. The code have been used then to evaluate the 99mTc S factors of interested target organs as the most common used radionuclide in nuclear medicine procedures. Substantial variations were observed in the S factors between the 10th/90th percentile phantoms from the 50th percentile phantom, with the greatest difference being 38.6 %. When the source organs were in close proximity to, or inside the fetal body, the 99mTc S factors presented strong statistical correlations with fetal body habitus. The trends observed in the S factors and the differences between various percentiles were justified by the source organs' masses, and chord length distributions (CLDs). The results of this study showed that fetal body habitus had a considerable effect on fetal dose (on average up to 8.4%) if constant fetal biokinetic data was considered for all fetal weight percentiles. However, an almost smaller variation on fetal dose (up to 5.3%) was obtained if the available biokinetic data for the reference fetus was scaled by fetal mass. © 2018 IOP Publishing Ltd.

Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

2003-06-01

The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

Calculation of the Effective Cross Sections of the Reaction X(n,p)Y with Neutrons in the Energy Range 2-5 Mev; CALCULO DE SECCIONES EFICACES X(n,p)Y CON NEUTRONES DE ENERGIAS ENTRE 2-5 Mev

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapaport, J.; Trier, A.

1960-05-01

Parallel with experimental work to measure the ic neutrons between 2 and 3.6 Mev, it was necessary to estimate the theoretical behavior of these cross sections. The statistical theory of Blatt and Weisskopf was used in the calculation. The theoretical results obtained for squarewell and diffuse-well development are compared with the experimental results. (J.S.R.)

An MCNPX2.7.0 study of Bragg peak degradation owing to density heterogeneity patterns for a CGMH therapeutic proton beam

NASA Astrophysics Data System (ADS)

Chao, Tsi-Chian; Tsai, Yi-Chun; Chen, Shih-Kuan; Wu, Shu-Wei; Tung, Chuan-Jong; Hong, Ji-Hong; Wang, Chun-Chieh; Lee, Chung-Chi

2017-08-01

The purpose of this study was to investigate the density heterogeneity pattern as a factor affecting Bragg peak degradation, including shifts in Bragg peak depth (ZBP), distal range (R80 and R20), and distal fall-off (R80-R20) using Monte Carlo N-Particles, eXtension (MCNPX). Density heterogeneities of different patterns with increasing complexity were placed downstream of commissioned proton beams at the Proton and Radiation Therapy Centre of Chang Gung Memorial Hospital, including one 150 MeV wobbling broad beam (10×10 cm2) and one 150 MeV proton pencil beam (FWHM of cross-plane=2.449 cm, FWHM of in-plane=2.256 cm). MCNPX 2.7.0 was used to model the transport and interactions of protons and secondary particles in density heterogeneity patterns and water using its repeated structure geometry. Different heterogeneity patterns were inserted into a 21×21×20 cm3 phantom. Mesh tally was used to track the dose distribution when the proton beam passed through the different density heterogeneity patterns. The results show that different heterogeneity patterns do cause different Bragg peak degradations owing to multiple Coulomb scattering (MCS) occurring in the density heterogeneities. A trend of increasing R20 and R80-R20 with increasing geometry complexity was observed. This means that Bragg peak degradation is mainly caused by the changes to the proton spectrum owing to MCS in the density heterogeneities. In contrast, R80 did not change considerably with different heterogeneity patterns, which indicated that the energy spectrum has only minimum effects on R80. Bragg peak degradation can occur both for a broad proton beam and a pencil beam, but is less significant for the broad beam.

Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Plaschy, M.; Murphy, M.; Jatuff, F.

2006-07-01

The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of themore » PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k{sub eff} (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)« less

Proton induced activity in graphite - comparison between measurement and simulation

NASA Astrophysics Data System (ADS)

Kiselev, Daniela; Bergmann, Ryan; Schumann, Dorothea; Talanov, Vadim; Wohlmuther, Michael

2018-06-01

The Paul Scherrer Institut (PSI) operates the Meson production target stations E and M with 590 MeV protons at currents of up to 2.4 mA. Both targets consist of polycrystalline graphite and rotate with 1 Hz due to the high power deposition (40 kW at 2 mA) in Target E. The graphite wheel is regularly exchanged and disposed as radioactive waste after a maximum of 3 to 4 years in operation, which corresponds to about 30 to 40 Ah of proton fluence. For disposal, the nuclide inventory of the long-lived isotopes (T1/2 > 60 d) has to be calculated and reported to the authorities. Measurements of gamma emitters, as well as 3H, 10Be and 14C, were carried out using different techniques. The measured specific activities are compared to Monte Carlo particle transport simulations performed with MCNPX2.7.0 using the BERTINI-DRESNER-RAL (default model in MCNPX2.7.0) and INCL4.6/ABLA07 as nuclear reaction models.

Methods for the Calculation of Settling Tanks for Batch Experiments; METODOS DE CALCULO DE ESPESADORES POR ENSAYOS DISCONTINUOS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gasos, P.; Perea, C.P.; Jodra, L.G.

1957-01-01

>In order to calculate settling tanks, some tests on batch sedimentation were made, and with the data obtained the dimensions of the settling tank were found. The mechanism of sedimentation is first briefly described, and then the factors involved in the calculation of the dimensions and the sedimentation velocity are discussed. The Cloe and Clevenger method and the Kynch method were investigated experimentally and compared. The application of the calculations are illustrated. It is shown that the two methods gave markedly different results. (J.S.R.)

Path Toward a Unified Geometry for Radiation Transport

NASA Astrophysics Data System (ADS)

Lee, Kerry

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex CAD models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN (high charge and energy transport code developed by NASA LaRC), are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The work-flow for doing radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats.

Photon mass attenuation coefficients of a silicon resin loaded with WO3, PbO, and Bi2O3 Micro and Nano-particles for radiation shielding

NASA Astrophysics Data System (ADS)

Verdipoor, Khatibeh; Alemi, Abdolali; Mesbahi, Asghar

2018-06-01

Novel shielding materials for photons based on silicon resin and WO3, PbO, and Bi2O3 Micro and Nano-particles were designed and their mass attenuation coefficients were calculated using Monte Carlo (MC) method. Using lattice cards in MCNPX code, micro and nanoparticles with sizes of 100 nm and 1 μm was designed inside a silicon resin matrix. Narrow beam geometry was simulated to calculate the attenuation coefficients of samples against mono-energetic beams of Co60 (1.17 and 1.33 MeV), Cs137 (663.8 KeV), and Ba133 (355.9 KeV). The shielding samples made of nanoparticles had higher mass attenuation coefficients, up to 17% relative to those made of microparticles. The superiority of nano-shields relative to micro-shields was dependent on the filler concentration and the energy of photons. PbO, and Bi2O3 nanoparticles showed higher attenuation compared to WO3 nanoparticles in studied energies. Fabrication of novel shielding materials using PbO, and Bi2O3 nanoparticles is recommended for application in radiation protection against photon beams.

A neutron diagnostic for high current deuterium beams.

PubMed

Rebai, M; Cavenago, M; Croci, G; Dalla Palma, M; Gervasini, G; Ghezzi, F; Grosso, G; Murtas, F; Pasqualotto, R; Cippo, E Perelli; Tardocchi, M; Tollin, M; Gorini, G

2012-02-01

A neutron diagnostic for high current deuterium beams is proposed for installation on the spectral shear interferometry for direct electric field reconstruction (SPIDER, Source for Production of Ion of Deuterium Extracted from RF plasma) test beam facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission on the beam dump surface by placing a detector in close contact, right behind the dump. CNESM uses gas electron multiplier detectors equipped with a cathode that also serves as neutron-proton converter foil. The cathode is made of a thin polythene film and an aluminium film; it is designed for detection of neutrons of energy >2.2 MeV with an incidence angle < 45°. CNESM was designed on the basis of simulations of the different steps from the deuteron beam interaction with the beam dump to the neutron detection in the nGEM. Neutron scattering was simulated with the MCNPX code. CNESM on SPIDER is a first step towards the application of this diagnostic technique to the MITICA beam test facility, where it will be used to resolve the horizontal profile of the beam intensity.

Monte carlo simulations of Yttrium reaction rates in Quinta uranium target

NASA Astrophysics Data System (ADS)

Suchopár, M.; Wagner, V.; Svoboda, O.; Vrzalová, J.; Chudoba, P.; Tichý, P.; Kugler, A.; Adam, J.; Závorka, L.; Baldin, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Solnyshkin, A.; Tsoupko-Sitnikov, V.; Tyutyunnikov, S.; Bielewicz, M.; Kilim, S.; Strugalska-Gola, E.; Szuta, M.

2017-03-01

The international collaboration Energy and Transmutation of Radioactive Waste (E&T RAW) performed intensive studies of several simple accelerator-driven system (ADS) setups consisting of lead, uranium and graphite which were irradiated by relativistic proton and deuteron beams in the past years at the Joint Institute for Nuclear Research (JINR) in Dubna, Russia. The most recent setup called Quinta, consisting of natural uranium target-blanket and lead shielding, was irradiated by deuteron beams in the energy range between 1 and 8 GeV in three accelerator runs at JINR Nuclotron in 2011 and 2012 with yttrium samples among others inserted inside the setup to measure the neutron flux in various places. Suitable activation detectors serve as one of possible tools for monitoring of proton and deuteron beams and for measurements of neutron field distribution in ADS studies. Yttrium is one of such suitable materials for monitoring of high energy neutrons. Various threshold reactions can be observed in yttrium samples. The yields of isotopes produced in the samples were determined using the activation method. Monte Carlo simulations of the reaction rates leading to production of different isotopes were performed in the MCNPX transport code and compared with the experimental results obtained from the yttrium samples.

Monte Carlo modeling of a conventional X-ray computed tomography scanner for gel dosimetry purposes.

PubMed

Hayati, Homa; Mesbahi, Asghar; Nazarpoor, Mahmood

2016-01-01

Our purpose in the current study was to model an X-ray CT scanner with the Monte Carlo (MC) method for gel dosimetry. In this study, a conventional CT scanner with one array detector was modeled with use of the MCNPX MC code. The MC calculated photon fluence in detector arrays was used for image reconstruction of a simple water phantom as well as polyacrylamide polymer gel (PAG) used for radiation therapy. Image reconstruction was performed with the filtered back-projection method with a Hann filter and the Spline interpolation method. Using MC results, we obtained the dose-response curve for images of irradiated gel at different absorbed doses. A spatial resolution of about 2 mm was found for our simulated MC model. The MC-based CT images of the PAG gel showed a reliable increase in the CT number with increasing absorbed dose for the studied gel. Also, our results showed that the current MC model of a CT scanner can be used for further studies on the parameters that influence the usability and reliability of results, such as the photon energy spectra and exposure techniques in X-ray CT gel dosimetry.

Preliminary Monte Carlo calculations for the UNCOSS neutron-based explosive detector

NASA Astrophysics Data System (ADS)

Eleon, C.; Perot, B.; Carasco, C.

2010-07-01

The goal of the FP7 UNCOSS project (Underwater Coastal Sea Surveyor) is to develop a non destructive explosive detection system based on the associated particle technique, in view to improve the security of coastal area and naval infrastructures where violent conflicts took place. The end product of the project will be a prototype of a complete coastal survey system, including a neutron-based sensor capable of confirming the presence of explosives on the sea bottom. A 3D analysis of prompt gamma rays induced by 14 MeV neutrons will be performed to identify elements constituting common military explosives, such as C, N and O. This paper presents calculations performed with the MCNPX computer code to support the ongoing design studies performed by the UNCOSS collaboration. Detection efficiencies, time and energy resolutions of the possible gamma-ray detectors are compared, which show NaI(Tl) or LaBr 3(Ce) scintillators will be suitable for this application. The effect of neutron attenuation and scattering in the seawater, influencing the counting statistics and signal-to-noise ratio, are also studied with calculated neutron time-of-flight and gamma-ray spectra for an underwater TNT target.

A neutron diagnostic for high current deuterium beams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rebai, M.; Perelli Cippo, E.; Cavenago, M.

2012-02-15

A neutron diagnostic for high current deuterium beams is proposed for installation on the spectral shear interferometry for direct electric field reconstruction (SPIDER, Source for Production of Ion of Deuterium Extracted from RF plasma) test beam facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission on the beam dump surface by placing a detector in close contact, right behind the dump. CNESM uses gas electron multiplier detectors equipped with a cathode that also serves as neutron-proton converter foil. The cathode is made of a thinmore » polythene film and an aluminium film; it is designed for detection of neutrons of energy >2.2 MeV with an incidence angle < 45 deg. CNESM was designed on the basis of simulations of the different steps from the deuteron beam interaction with the beam dump to the neutron detection in the nGEM. Neutron scattering was simulated with the MCNPX code. CNESM on SPIDER is a first step towards the application of this diagnostic technique to the MITICA beam test facility, where it will be used to resolve the horizontal profile of the beam intensity.« less

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek J.; Diamond D.; Cuadra, A.

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

Organic Scintillator for Real-Time Neutron Dosimetry

DOE PAGES

Beyer, Kyle A.; Di Fulvio, Angela; Stolarczyk, Liliana; ...

2017-11-15

We have developed a radiation detector based on an organic scintillator for spectrometry and dosimetry of out-of-field secondary neutrons from clinical proton beams. The detector consists of an EJ-299-34 crystalline organic scintillator, coupled by fiber optic cable to a silicon photomultiplier (SiPM). Proof of concept measurements were taken with 137Cs and 252Cf, and corresponding simulations were performed in MCNPX-PoliMi. Despite its small size, the detector is able to discriminate between neutron and gamma-rays via pulse shape discrimination. We simulated the response function of the detector to monoenergetic neutrons in the 100 keV–0 MeV range using MCNPX-PoliMi. The measured unfolded 252Cfmore » neutron spectrum is in good agreement with the theoretical Watt fission spectrum. We determined the ambient dose equivalent by folding the spectrum with the fluence-to-ambient dose conversion coefficient, with a 1.4% deviation from theory. Some preliminary proton beam experiments were preformed at the Bronowice Cyclotron Center patient treatment facility using a clinically relevant proton pencil beam for brain tumor and craino-spinal treatment directed at a child phantom.« less

Organic Scintillator for Real-Time Neutron Dosimetry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, Kyle A.; Di Fulvio, Angela; Stolarczyk, Liliana

We have developed a radiation detector based on an organic scintillator for spectrometry and dosimetry of out-of-field secondary neutrons from clinical proton beams. The detector consists of an EJ-299-34 crystalline organic scintillator, coupled by fiber optic cable to a silicon photomultiplier (SiPM). Proof of concept measurements were taken with 137Cs and 252Cf, and corresponding simulations were performed in MCNPX-PoliMi. Despite its small size, the detector is able to discriminate between neutron and gamma-rays via pulse shape discrimination. We simulated the response function of the detector to monoenergetic neutrons in the 100 keV–0 MeV range using MCNPX-PoliMi. The measured unfolded 252Cfmore » neutron spectrum is in good agreement with the theoretical Watt fission spectrum. We determined the ambient dose equivalent by folding the spectrum with the fluence-to-ambient dose conversion coefficient, with a 1.4% deviation from theory. Some preliminary proton beam experiments were preformed at the Bronowice Cyclotron Center patient treatment facility using a clinically relevant proton pencil beam for brain tumor and craino-spinal treatment directed at a child phantom.« less

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hursin, M.; Koeberl, O.; Perret, G.

2012-07-01

High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivitymore » Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)« less

A boundary-representation method for designing whole-body radiation dosimetry models: pregnant females at the ends of three gestational periods—RPI-P3, -P6 and -P9

NASA Astrophysics Data System (ADS)

Xu, X. George; Taranenko, Valery; Zhang, Juying; Shi, Chengyu

2007-12-01

Fetuses are extremely radiosensitive and the protection of pregnant females against ionizing radiation is of particular interest in many health and medical physics applications. Existing models of pregnant females relied on simplified anatomical shapes or partial-body images of low resolutions. This paper reviews two general types of solid geometry modeling: constructive solid geometry (CSG) and boundary representation (BREP). It presents in detail a project to adopt the BREP modeling approach to systematically design whole-body radiation dosimetry models: a pregnant female and her fetus at the ends of three gestational periods of 3, 6 and 9 months. Based on previously published CT images of a 7-month pregnant female, the VIP-Man model and mesh organ models, this new set of pregnant female models was constructed using 3D surface modeling technologies instead of voxels. The organ masses were adjusted to agree with the reference data provided by the International Commission on Radiological Protection (ICRP) and previously published papers within 0.5%. The models were then voxelized for the purpose of performing dose calculations in identically implemented EGS4 and MCNPX Monte Carlo codes. The agreements of the fetal doses obtained from these two codes for this set of models were found to be within 2% for the majority of the external photon irradiation geometries of AP, PA, LAT, ROT and ISO at various energies. It is concluded that the so-called RPI-P3, RPI-P6 and RPI-P9 models have been reliably defined for Monte Carlo calculations. The paper also discusses the needs for future research and the possibility for the BREP method to become a major tool in the anatomical modeling for radiation dosimetry.

Path Toward a Unifid Geometry for Radiation Transport

NASA Technical Reports Server (NTRS)

Lee, Kerry; Barzilla, Janet; Davis, Andrew; Zachmann

2014-01-01

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex computer-aided design (CAD) models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN [high charge and energy transport code developed by NASA Langley Research Center (LaRC)], are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit-specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The workflow for achieving radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats

Quantitative NDA measurements of advanced reprocessing product materials containing uranium, neptunium, plutonium, and americium

NASA Astrophysics Data System (ADS)

Goddard, Braden

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (alpha,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

Validation of Monte Carlo simulation of neutron production in a spallation experiment

DOE PAGES

Zavorka, L.; Adam, J.; Artiushenko, M.; ...

2015-02-25

A renewed interest in experimental research on Accelerator-Driven Systems (ADS) has been initiated by the global attempt to produce energy from thorium as a safe(r), clean(er) and (more) proliferation-resistant alternative to the uranium-fuelled thermal nuclear reactors. The ADS research has been actively pursued at the Joint Institute for Nuclear Research (JINR), Dubna, since decades. Most recently, the emission of fast neutrons was experimentally investigated at the massive (m = 512 kg) natural uranium spallation target QUINTA. The target has been irradiated with the relativistic deuteron beams of energy from 0.5 AGeV up to 4 AGeV at the JINR Nuclotron acceleratormore » in numerous experiments since 2011. Neutron production inside the target was studied through the gamma-ray spectrometry measurement of natural uranium activation detectors. Experimental reaction rates for (n,γ), (n,f) and (n,2n) reactions in uranium have provided valuable information about the neutron distribution over a wide range of energies up to some GeV. The experimental data were compared to the predictions of Monte Carlo simulations using the MCNPX 2.7.0 code. In conclusion, the results are presented and potential sources of partial disagreement are discussed later in this work.« less

Simulation of internal contamination screening with dose rate meters

NASA Astrophysics Data System (ADS)

Fonseca, T. C. F.; Mendes, B. M.; Hunt, J. G.

2017-11-01

Assessing the intake of radionuclides after an accident in a nuclear power plant or after the intentional release of radionuclides in public places allows dose calculations and triage actions to be carried out for members of the public and for emergency response teams. Gamma emitters in the lung, thyroid or the whole body may be detected and quantified by making dose rate measurements at the surface of the internally contaminated person. In an accident scenario, quick measurements made with readily available portable equipment are a key factor for success. In this paper, the Monte Carlo program Visual Monte Carlo (VMC) and MCNPx code are used in conjunction with voxel phantoms to calculate the dose rate at the surface of a contaminated person due to internally deposited radionuclides. A whole body contamination with 137Cs and a thyroid contamination with 131I were simulated and the calibration factors in kBq per μSv/h were calculated. The calculated calibration factors were compared with real data obtained from the Goiania accident in the case of 137Cs and the Chernobyl accident in terms of the 131I. The close comparison of the calculated and real measurements indicates that the method may be applied to other radionuclides. Minimum detectable activities are discussed.

Occupational exposures during abdominal fluoroscopically guided interventional procedures for different patient sizes - A Monte Carlo approach.

PubMed

Santos, William S; Belinato, Walmir; Perini, Ana P; Caldas, Linda V E; Galeano, Diego C; Santos, Carla J; Neves, Lucio P

2018-01-01

In this study we evaluated the occupational exposures during an abdominal fluoroscopically guided interventional radiology procedure. We investigated the relation between the Body Mass Index (BMI), of the patient, and the conversion coefficient values (CC) for a set of dosimetric quantities, used to assess the exposure risks of medical radiation workers. The study was performed using a set of male and female virtual anthropomorphic phantoms, of different body weights and sizes. In addition to these phantoms, a female and a male phantom, named FASH3 and MASH3 (reference virtual anthropomorphic phantoms), were also used to represent the medical radiation workers. The CC values, obtained as a function of the dose area product, were calculated for 87 exposure scenarios. In each exposure scenario, three phantoms, implemented in the MCNPX 2.7.0 code, were simultaneously used. These phantoms were utilized to represent a patient and medical radiation workers. The results showed that increasing the BMI of the patient, adjusted for each patient protocol, the CC values for medical radiation workers decrease. It is important to note that these results were obtained with fixed exposure parameters. Copyright © 2017 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Mitigating the Effects of the Space Radiation Environment: A Novel Approach of Using Graded-Z Materials

NASA Technical Reports Server (NTRS)

Atwell, William; Rojdev, Kristina; Aghara, Sukesh; Sriprisan, Sirikul

2013-01-01

In this paper we present a novel space radiation shielding approach using various material lay-ups, called "Graded-Z" shielding, which could optimize cost, weight, and safety while mitigating the radiation exposures from the trapped radiation and solar proton environments, as well as the galactic cosmic radiation (GCR) environment, to humans and electronics. In addition, a validation and verification (V&V) was performed using two different high energy particle transport/dose codes (MCNPX & HZETRN). Inherently, we know that materials having high-hydrogen content are very good space radiation shielding materials. Graded-Z material lay-ups are very good trapped electron mitigators for medium earth orbit (MEO) and geostationary earth orbit (GEO). In addition, secondary particles, namely neutrons, are produced as the primary particles penetrate a spacecraft, which can have deleterious effects to both humans and electronics. The use of "dopants," such as beryllium, boron, and lithium, impregnated in other shielding materials provides a means of absorbing the secondary neutrons. Several examples of optimized Graded-Z shielding layups that include the use of composite materials are presented and discussed in detail. This parametric shielding study is an extension of some earlier pioneering work we (William Atwell and Kristina Rojdev) performed in 20041 and 20092.

Neutron yield and induced radioactivity: a study of 235-MeV proton and 3-GeV electron accelerators.

PubMed

Hsu, Yung-Cheng; Lai, Bo-Lun; Sheu, Rong-Jiun

2016-01-01

This study evaluated the magnitude of potential neutron yield and induced radioactivity of two new accelerators in Taiwan: a 235-MeV proton cyclotron for radiation therapy and a 3-GeV electron synchrotron serving as the injector for the Taiwan Photon Source. From a nuclear interaction point of view, neutron production from targets bombarded with high-energy particles is intrinsically related to the resulting target activation. Two multi-particle interaction and transport codes, FLUKA and MCNPX, were used in this study. To ensure prediction quality, much effort was devoted to the associated benchmark calculations. Comparisons of the accelerators' results for three target materials (copper, stainless steel and tissue) are presented. Although the proton-induced neutron yields were higher than those induced by electrons, the maximal neutron production rates of both accelerators were comparable according to their respective beam outputs during typical operation. Activation products in the targets of the two accelerators were unexpectedly similar because the primary reaction channels for proton- and electron-induced activation are (p,pn) and (γ,n), respectively. The resulting residual activities and remnant dose rates as a function of time were examined and discussed. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

NASA Astrophysics Data System (ADS)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Evaluation of Am–Li neutron spectra data for active well type neutron multiplicity measurements of uranium

DOE PAGES

Goddard, Braden; Croft, Stephen; Lousteau, Angela; ...

2016-05-25

Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with ( α,n) neutrons from an Am-Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am-Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured spectra ofmore » various Am-Li sources to each other. This paper examines fourteen different Am-Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am-Li spectra produce similar doubles and triples count rates. Finally, the singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis.« less

SU-E-T-656: Quantitative Analysis of Proton Boron Fusion Therapy (PBFT) in Various Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, D; Jung, J; Shin, H

2015-06-15

Purpose: Three alpha particles are concomitant of proton boron interaction, which can be used in radiotherapy applications. We performed simulation studies to determine the effectiveness of proton boron fusion therapy (PBFT) under various conditions. Methods: Boron uptake regions (BURs) of various widths and densities were implemented in Monte Carlo n-particle extended (MCNPX) simulation code. The effect of proton beam energy was considered for different BURs. Four simulation scenarios were designed to verify the effectiveness of integrated boost that was observed in the proton boron reaction. In these simulations, the effect of proton beam energy was determined for different physical conditions,more » such as size, location, and boron concentration. Results: Proton dose amplification was confirmed for all proton beam energies considered (< 96.62%). Based on the simulation results for different physical conditions, the threshold for the range in which proton dose amplification occurred was estimated as 0.3 cm. Effective proton boron reaction requires the boron concentration to be equal to or greater than 14.4 mg/g. Conclusion: We established the effects of the PBFT with various conditions by using Monte Carlo simulation. The results of our research can be used for providing a PBFT dose database.« less

Prompt gamma neutron activation analysis of toxic elements in radioactive waste packages.

PubMed

Ma, J-L; Carasco, C; Perot, B; Mauerhofer, E; Kettler, J; Havenith, A

2012-07-01

The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R&D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages. In particular, the amount of toxic elements present in radioactive waste packages must be assessed before they can be accepted in repository facilities in order to avoid pollution of underground water reserves. To this aim, the Nuclear Measurement Laboratory of CEA-Cadarache has started to study the performances of Prompt Gamma Neutron Activation Analysis (PGNAA) for elements showing large capture cross sections such as mercury, cadmium, boron, and chromium. This paper reports a comparison between Monte Carlo calculations performed with the MCNPX computer code using the ENDF/B-VII.0 library and experimental gamma rays measured in the REGAIN PGNAA cell with small samples of nickel, lead, cadmium, arsenic, antimony, chromium, magnesium, zinc, boron, and lithium to verify the validity of a numerical model and gamma-ray production data. The measurement of a ∼20kg test sample of concrete containing toxic elements has also been performed, in collaboration with Forschungszentrum Jülich, to validate the model in view of future performance studies for dense and large LL-MA waste packages. Copyright © 2012 Elsevier Ltd. All rights reserved.

MCNP Simulation Benchmarks for a Portable Inspection System for Narcotics, Explosives, and Nuclear Material Detection

NASA Astrophysics Data System (ADS)

Alfonso, Krystal; Elsalim, Mashal; King, Michael; Strellis, Dan; Gozani, Tsahi

2013-04-01

MCNPX simulations have been used to guide the development of a portable inspection system for narcotics, explosives, and special nuclear material (SNM) detection. The system seeks to address these threats to national security by utilizing a high-yield, compact neutron source to actively interrogate the threats and produce characteristic signatures that can then be detected by radiation detectors. The portability of the system enables rapid deployment and proximity to threats concealed in small spaces. Both dD and dT electronic neutron generators (ENG) were used to interrogate ammonium nitrate fuel oil (ANFO) and cocaine hydrochloride, and the detector response of NaI, CsI, and LaBr3 were compared. The effect of tungsten shielding on the neutron flux in the gamma ray detectors was investigated, while carbon, beryllium, and polyethylene ENG moderator materials were optimized by determining the reaction rate density in the threats. In order to benchmark the modeling results, experimental measurements are compared with MCNPX simulations. In addition, the efficiency and die-away time of a portable differential die-away analysis (DDAA) detector using 3He proportional counters for SNM detection has been determined.

ORGANIC SCINTILLATOR FOR REAL-TIME NEUTRON DOSIMETRY.

PubMed

Beyer, Kyle A; Di Fulvio, Angela; Stolarczyk, Liliana; Parol, Wiktor; Mojzeszek, Natalia; Kopéc, Renata; Clarke, Shaun D; Pozzi, Sara A

2017-11-15

We developed a radiation detector based on an organic scintillator for spectrometry and dosimetry of out-of-field secondary neutrons from clinical proton beams. The detector consists of an EJ-299-34 crystalline organic scintillator, coupled by fiber optic cable to a silicon photomultiplier (SiPM). Proof of concept measurements were taken with 137Cs and 252Cf, and corresponding simulations were performed in MCNPX-PoliMi. Despite its small size, the detector is able to discriminate between neutron and gamma-rays via pulse shape discrimination. We simulated the response function of the detector to monoenergetic neutrons in the 100 keV-0 MeV range using MCNPX-PoliMi. The measured unfolded 252Cf neutron spectrum is in good agreement with the theoretical Watt fission spectrum. We determined the ambient dose equivalent by folding the spectrum with the fluence-to-ambient dose conversion coefficient, with a 1.4% deviation from theory. Some preliminary proton beam experiments were preformed at the Bronowice Cyclotron Center patient treatment facility using a clinically relevant proton pencil beam for brain tumor and craino-spinal treatment directed at a child phantom. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Fine-resolution voxel S values for constructing absorbed dose distributions at variable voxel size.

PubMed

Dieudonné, Arnaud; Hobbs, Robert F; Bolch, Wesley E; Sgouros, George; Gardin, Isabelle

2010-10-01

This article presents a revised voxel S values (VSVs) approach for dosimetry in targeted radiotherapy, allowing dose calculation for any voxel size and shape of a given SPECT or PET dataset. This approach represents an update to the methodology presented in MIRD pamphlet no. 17. VSVs were generated in soft tissue with a fine spatial sampling using the Monte Carlo (MC) code MCNPX for particle emissions of 9 radionuclides: (18)F, (90)Y, (99m)Tc, (111)In, (123)I, (131)I, (177)Lu, (186)Re, and (201)Tl. A specific resampling algorithm was developed to compute VSVs for desired voxel dimensions. The dose calculation was performed by convolution via a fast Hartley transform. The fine VSVs were calculated for cubic voxels of 0.5 mm for electrons and 1.0 mm for photons. Validation studies were done for (90)Y and (131)I VSV sets by comparing the revised VSV approach to direct MC simulations. The first comparison included 20 spheres with different voxel sizes (3.8-7.7 mm) and radii (4-64 voxels) and the second comparison a hepatic tumor with cubic voxels of 3.8 mm. MC simulations were done with MCNPX for both. The third comparison was performed on 2 clinical patients with the 3D-RD (3-Dimensional Radiobiologic Dosimetry) software using the EGSnrc (Electron Gamma Shower National Research Council Canada)-based MC implementation, assuming a homogeneous tissue-density distribution. For the sphere model study, the mean relative difference in the average absorbed dose was 0.20% ± 0.41% for (90)Y and -0.36% ± 0.51% for (131)I (n = 20). For the hepatic tumor, the difference in the average absorbed dose to tumor was 0.33% for (90)Y and -0.61% for (131)I and the difference in average absorbed dose to the liver was 0.25% for (90)Y and -1.35% for (131)I. The comparison with the 3D-RD software showed an average voxel-to-voxel dose ratio between 0.991 and 0.996. The calculation time was below 10 s with the VSV approach and 50 and 15 h with 3D-RD for the 2 clinical patients. This new VSV approach enables the calculation of absorbed dose based on a SPECT or PET cumulated activity map, with good agreement with direct MC methods, in a faster and more clinically compatible manner.

Meson Production and Space Radiation

NASA Astrophysics Data System (ADS)

Norbury, John; Blattnig, Steve; Norman, Ryan; Aghara, Sukesh

Protecting astronauts from the harmful effects of space radiation is an important priority for long duration space flight. The National Council on Radiation Protection (NCRP) has recently recommended that pion and other mesons should be included in space radiation transport codes, especially in connection with the Martian atmosphere. In an interesting accident of nature, the galactic cosmic ray spectrum has its peak intensity near the pion production threshold. The Boltzmann transport equation is structured in such a way that particle production cross sec-tions are multiplied by particle flux. Therefore, the peak of the incident flux of the galactic cosmic ray spectrum is more important than other regions of the spectrum and cross sections near the peak are enhanced. This happens with pion cross sections. The MCNPX Monte-Carlo transport code now has the capability of transporting heavy ions, and by using a galactic cosmic ray spectrum as input, recent work has shown that pions contribute about twenty percent of the dose from galactic cosmic rays behind a shield of 20 g/cm2 aluminum and 30 g/cm2 water. It is therefore important to include pion and other hadron production in transport codes designed for space radiation studies, such as HZETRN. The status of experimental hadron production data for energies relevant to space radiation will be reviewed, as well as the predictive capa-bilities of current theoretical hadron production cross section and space radiation transport models. Charged pions decay into muons and neutrinos, and neutral pions decay into photons. An electromagnetic cascade is produced as these particles build up in a material. The cascade and transport of pions, muons, electrons and photons will be discussed as they relate to space radiation. The importance of other hadrons, such as kaons, eta mesons and antiprotons will be considered as well. Efficient methods for calculating cross sections for meson production in nucleon-nucleon and nucleus-nucleus reactions will be presented. The NCRP has also recom-mended that more attention should be paid to neutron and light ion transport. The coupling of neutrons, light ions, mesons and other hadrons will be discussed.

Simulations of GCR interactions within planetary bodies using GEANT4

NASA Astrophysics Data System (ADS)

Mesick, K.; Feldman, W. C.; Stonehill, L. C.; Coupland, D. D. S.

2017-12-01

On planetary bodies with little to no atmosphere, Galactic Cosmic Rays (GCRs) can hit the body and produce neutrons primarily through nuclear spallation within the top few meters of the surfaces. These neutrons undergo further nuclear interactions with elements near the planetary surface and some will escape the surface and can be detected by landed or orbiting neutron radiation detector instruments. The neutron leakage signal at fast neutron energies provides a measure of average atomic mass of the near-surface material and in the epithermal and thermal energy ranges is highly sensitive to the presence of hydrogen. Gamma-rays can also escape the surface, produced at characteristic energies depending on surface composition, and can be detected by gamma-ray instruments. The intra-nuclear cascade (INC) that occurs when high-energy GCRs interact with elements within a planetary surface to produce the leakage neutron and gamma-ray signals is highly complex, and therefore Monte Carlo based radiation transport simulations are commonly used for predicting and interpreting measurements from planetary neutron and gamma-ray spectroscopy instruments. In the past, the simulation code that has been widely used for this type of analysis is MCNPX [1], which was benchmarked against data from the Lunar Neutron Probe Experiment (LPNE) on Apollo 17 [2]. In this work, we consider the validity of the radiation transport code GEANT4 [3], another widely used but open-source code, by benchmarking simulated predictions of the LPNE experiment to the Apollo 17 data. We consider the impact of different physics model options on the results, and show which models best describe the INC based on agreement with the Apollo 17 data. The success of this validation then gives us confidence in using GEANT4 to simulate GCR-induced neutron leakage signals on Mars in relevance to a re-analysis of Mars Odyssey Neutron Spectrometer data. References [1] D.B. Pelowitz, Los Alamos National Laboratory, LA-CP-05-0369, 2005. [2] G.W. McKinney et al, Journal of Geophysics Research, 111, E06004, 2006. [3] S. Agostinelli et al, Nuclear Instrumentation and Methods A, 506, 2003.

Study of a new design of p-N semiconductor detector array for nuclear medicine imaging by monte carlo simulation codes.

PubMed

Hajizadeh-Safar, M; Ghorbani, M; Khoshkharam, S; Ashrafi, Z

2014-07-01

Gamma camera is an important apparatus in nuclear medicine imaging. Its detection part is consists of a scintillation detector with a heavy collimator. Substitution of semiconductor detectors instead of scintillator in these cameras has been effectively studied. In this study, it is aimed to introduce a new design of P-N semiconductor detector array for nuclear medicine imaging. A P-N semiconductor detector composed of N-SnO2 :F, and P-NiO:Li, has been introduced through simulating with MCNPX monte carlo codes. Its sensitivity with different factors such as thickness, dimension, and direction of emission photons were investigated. It is then used to configure a new design of an array in one-dimension and study its spatial resolution for nuclear medicine imaging. One-dimension array with 39 detectors was simulated to measure a predefined linear distribution of Tc(99_m) activity and its spatial resolution. The activity distribution was calculated from detector responses through mathematical linear optimization using LINPROG code on MATLAB software. Three different configurations of one-dimension detector array, horizontal, vertical one sided, and vertical double-sided were simulated. In all of these configurations, the energy windows of the photopeak were ± 1%. The results show that the detector response increases with an increase of dimension and thickness of the detector with the highest sensitivity for emission photons 15-30° above the surface. Horizontal configuration array of detectors is not suitable for imaging of line activity sources. The measured activity distribution with vertical configuration array, double-side detectors, has no similarity with emission sources and hence is not suitable for imaging purposes. Measured activity distribution using vertical configuration array, single side detectors has a good similarity with sources. Therefore, it could be introduced as a suitable configuration for nuclear medicine imaging. It has been shown that using semiconductor P-N detectors such as P-NiO:Li, N-SnO2 :F for gamma detection could be possibly applicable for design of a one dimension array configuration with suitable spatial resolution of 2.7 mm for nuclear medicine imaging.

Effect of photon energy spectrum on dosimetric parameters of brachytherapy sources.

PubMed

Ghorbani, Mahdi; Mehrpouyan, Mohammad; Davenport, David; Ahmadi Moghaddas, Toktam

2016-06-01

The aim of this study is to quantify the influence of the photon energy spectrum of brachytherapy sources on task group No. 43 (TG-43) dosimetric parameters. Different photon spectra are used for a specific radionuclide in Monte Carlo simulations of brachytherapy sources. MCNPX code was used to simulate 125I, 103Pd, 169Yb, and 192Ir brachytherapy sources. Air kerma strength per activity, dose rate constant, radial dose function, and two dimensional (2D) anisotropy functions were calculated and isodose curves were plotted for three different photon energy spectra. The references for photon energy spectra were: published papers, Lawrence Berkeley National Laboratory (LBNL), and National Nuclear Data Center (NNDC). The data calculated by these photon energy spectra were compared. Dose rate constant values showed a maximum difference of 24.07% for 103Pd source with different photon energy spectra. Radial dose function values based on different spectra were relatively the same. 2D anisotropy function values showed minor differences in most of distances and angles. There was not any detectable difference between the isodose contours. Dosimetric parameters obtained with different photon spectra were relatively the same, however it is suggested that more accurate and updated photon energy spectra be used in Monte Carlo simulations. This would allow for calculation of reliable dosimetric data for source modeling and calculation in brachytherapy treatment planning systems.

Effect of photon energy spectrum on dosimetric parameters of brachytherapy sources

PubMed Central

Ghorbani, Mahdi; Davenport, David

2016-01-01

Abstract Aim The aim of this study is to quantify the influence of the photon energy spectrum of brachytherapy sources on task group No. 43 (TG-43) dosimetric parameters. Background Different photon spectra are used for a specific radionuclide in Monte Carlo simulations of brachytherapy sources. Materials and methods MCNPX code was used to simulate 125I, 103Pd, 169Yb, and 192Ir brachytherapy sources. Air kerma strength per activity, dose rate constant, radial dose function, and two dimensional (2D) anisotropy functions were calculated and isodose curves were plotted for three different photon energy spectra. The references for photon energy spectra were: published papers, Lawrence Berkeley National Laboratory (LBNL), and National Nuclear Data Center (NNDC). The data calculated by these photon energy spectra were compared. Results Dose rate constant values showed a maximum difference of 24.07% for 103Pd source with different photon energy spectra. Radial dose function values based on different spectra were relatively the same. 2D anisotropy function values showed minor differences in most of distances and angles. There was not any detectable difference between the isodose contours. Conclusions Dosimetric parameters obtained with different photon spectra were relatively the same, however it is suggested that more accurate and updated photon energy spectra be used in Monte Carlo simulations. This would allow for calculation of reliable dosimetric data for source modeling and calculation in brachytherapy treatment planning systems. PMID:27247558

Study of the counting efficiency of a WBC setup by using a computational 3D human body library in sitting position based on polygonal mesh surfaces.

PubMed

Fonseca, T C Ferreira; Bogaerts, R; Lebacq, A L; Mihailescu, C L; Vanhavere, F

2014-04-01

A realistic computational 3D human body library, called MaMP and FeMP (Male and Female Mesh Phantoms), based on polygonal mesh surface geometry, has been created to be used for numerical calibration of the whole body counter (WBC) system of the nuclear power plant (NPP) in Doel, Belgium. The main objective was to create flexible computational models varying in gender, body height, and mass for studying the morphology-induced variation of the detector counting efficiency (CE) and reducing the measurement uncertainties. First, the counting room and an HPGe detector were modeled using MCNPX (Monte Carlo radiation transport code). The validation of the model was carried out for different sample-detector geometries with point sources and a physical phantom. Second, CE values were calculated for a total of 36 different mesh phantoms in a seated position using the validated Monte Carlo model. This paper reports on the validation process of the in vivo whole body system and the CE calculated for different body heights and weights. The results reveal that the CE is strongly dependent on the individual body shape, size, and gender and may vary by a factor of 1.5 to 3 depending on the morphology aspects of the individual to be measured.

Commissioning dosimetry and in situ dose mapping of a semi-industrial Cobalt-60 gamma-irradiation facility using Fricke and Ceric-cerous dosimetry system and comparison with Monte Carlo simulation data

NASA Astrophysics Data System (ADS)

Mortuza, Md Firoz; Lepore, Luigi; Khedkar, Kalpana; Thangam, Saravanan; Nahar, Arifatun; Jamil, Hossen Mohammad; Bandi, Laxminarayan; Alam, Md Khorshed

2018-03-01

Characterization of a 90 kCi (3330 TBq), semi-industrial, cobalt-60 gamma irradiator was performed by commissioning dosimetry and in-situ dose mapping experiments with Ceric-cerous and Fricke dosimetry systems. Commissioning dosimetry was carried out to determine dose distribution pattern of absorbed dose in the irradiation cell and products. To determine maximum and minimum absorbed dose, overdose ratio and dwell time of the tote boxes, homogeneous dummy product (rice husk) with a bulk density of 0.13 g/cm3 were used in the box positions of irradiation chamber. The regions of minimum absorbed dose of the tote boxes were observed in the lower zones of middle plane and maximum absorbed doses were found in the middle position of front plane. Moreover, as a part of dose mapping, dose rates in the wall positions and some selective strategic positions were also measured to carry out multiple irradiation program simultaneously, especially for low dose research irradiation program. In most of the cases, Monte Carlo simulation data, using Monte Carlo N-Particle eXtended code version MCNPX 2.7., were found to be in congruence with experimental values obtained from Ceric-cerous and Fricke dosimetry; however, in close proximity positions from the source, the dose rate variation between chemical dosimetry and MCNP was higher than distant positions.

Monte Carlo calculations of the cellular S-values for α-particle-emitting radionuclides incorporated into the nuclei of cancer cells of the MDA-MB231, MCF7 and PC3 lines.

PubMed

Rojas-Calderón, E L; Ávila, O; Ferro-Flores, G

2018-05-01

S-values (dose per unit of cumulated activity) for alpha particle-emitting radionuclides and monoenergetic alpha sources placed in the nuclei of three cancer cell models (MCF7, MDA-MB231 breast cancer cells and PC3 prostate cancer cells) were obtained by Monte Carlo simulation. The MCNPX code was used to calculate the fraction of energy deposited in the subcellular compartments due to the alpha sources in order to obtain the S-values. A comparison with internationally accepted S-values reported by the MIRD Cellular Committee for alpha sources in three sizes of spherical cells was also performed leading to an agreement within 4% when an alpha extended source uniformly distributed in the nucleus is simulated. This result allowed to apply the Monte Carlo Methodology to evaluate S-values for alpha particles in cancer cells. The calculation of S-values for nucleus, cytoplasm and membrane of cancer cells considering their particular geometry, distribution of the radionuclide source and chemical composition by means of Monte Carlo simulation provides a good approach for dosimetry assessment of alpha emitters inside cancer cells. Results from this work provide information and tools that may help researchers in the selection of appropriate radiopharmaceuticals in alpha-targeted cancer therapy and improve its dosimetry evaluation. Copyright © 2018 Elsevier Ltd. All rights reserved.

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perret, G.; Pattupara, R. M.; Girardin, G.

2012-07-01

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fastmore » Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)« less

Neutron production from beam-modifying devices in a modern double scattering proton therapy beam delivery system.

PubMed

Pérez-Andújar, Angélica; Newhauser, Wayne D; Deluca, Paul M

2009-02-21

In this work the neutron production in a passive beam delivery system was investigated. Secondary particles including neutrons are created as the proton beam interacts with beam shaping devices in the treatment head. Stray neutron exposure to the whole body may increase the risk that the patient develops a radiogenic cancer years or decades after radiotherapy. We simulated a passive proton beam delivery system with double scattering technology to determine the neutron production and energy distribution at 200 MeV proton energy. Specifically, we studied the neutron absorbed dose per therapeutic absorbed dose, the neutron absorbed dose per source particle and the neutron energy spectrum at various locations around the nozzle. We also investigated the neutron production along the nozzle's central axis. The absorbed doses and neutron spectra were simulated with the MCNPX Monte Carlo code. The simulations revealed that the range modulation wheel (RMW) is the most intense neutron source of any of the beam spreading devices within the nozzle. This finding suggests that it may be helpful to refine the design of the RMW assembly, e.g., by adding local shielding, to suppress neutron-induced damage to components in the nozzle and to reduce the shielding thickness of the treatment vault. The simulations also revealed that the neutron dose to the patient is predominated by neutrons produced in the field defining collimator assembly, located just upstream of the patient.

The light output and the detection efficiency of the liquid scintillator EJ-309.

PubMed

Pino, F; Stevanato, L; Cester, D; Nebbia, G; Sajo-Bohus, L; Viesti, G

2014-07-01

The light output response and the neutron and gamma-ray detection efficiency are determined for liquid scintillator EJ-309. The light output function is compared to those of previous studies. Experimental efficiency results are compared to predictions from GEANT4, MCNPX and PENELOPE Monte Carlo simulations. The differences associated with the use of different light output functions are discussed. Copyright © 2014 Elsevier Ltd. All rights reserved.

Effective dose to immuno-PET patients due to metastable impurities in cyclotron produced zirconium-89

NASA Astrophysics Data System (ADS)

Alfuraih, Abdulrahman; Alzimami, Khalid; Ma, Andy K.; Alghamdi, Ali; Al Jammaz, Ibrahim

2014-11-01

Immuno-PET is a nuclear medicine technique that combines positron emission tommography (PET) with radio-labeled monoclonal antibodies (mAbs) for tumor characterization and therapy. Zirconium-89 (89Zr) is an emerging radionuclide for immuno-PET imaging. Its long half-life (78.4 h) gives ample time for the production, the administering and the patient uptake of the tagged radiopharmaceutical. Furthermore, the nuclides will remain in the tumor cells after the mAbs are catabolized so that time series studies are possible without incurring further administration of radiopharmarceuticals. 89Zr can be produced in medical cyclotrons by bombarding an yttrium-89 (89Y) target with a proton beam through the 89Y(p,n)89Zr reaction. In this study, we estimated the effective dose to the head and neck cancer patients undergoing 89Zr-based immune-PET procedures. The production of 89Zr and the impurities from proton irradiation of the 89Y target in a cyclotron was calculated with the Monte Carlo code MCNPX and the nuclear reaction code TALYS. The cumulated activities of the Zr isotopes were derived from real patient data in literature and the effective doses were estimated using the MIRD specific absorbed fraction formalism. The estimated effective dose from 89Zr is 0.5±0.2 mSv/MBq. The highest organ dose is 1.8±0.2 mSv/MBq in the liver. These values are in agreement with those reported in literature. The effective dose from 89mZr is about 0.2-0.3% of the 89Zr dose in the worst case. Since the ratio of 89mZr to 89Zr depends on the cooling time as well as the irradiation details, contaminant dose estimation is an important aspect in optimizing the cyclotron irradiation geometry, energy and time.

Assessment of Mean Glandular Dose in Mammography System with Different Anode-Filter Combinations Using MCNP Code.

PubMed

Gholamkar, Lida; Mowlavi, Ali Asghar; Sadeghi, Mahdi; Athari, Mitra

2016-10-01

X-ray mammography is one of the general methods for early detection of breast cancer. Since glandular tissue in the breast is sensitive to radiation and it increases the risk of cancer, the given dose to the patient is very important in mammography. The aim of this study was to determine the average absorbed dose of X-ray radiation in the glandular tissue of the breast during mammography examinations as well as investigating factors that influence the mean glandular dose (MGD). One of the precise methods for determination of MGD absorbed by the breast is Monte Carlo simulation method which is widely used to assess the dose. We studied some different X-ray sources and exposure factors that affect the MGD. "Midi-future" digital mammography system with amorphous-selenium detector was simulated using the Monte Carlo N-particle extended (MCNPX) code. Different anode/filter combinations such as tungsten/silver (W/Ag), tungsten/rhodium (W/Rh), and rhodium/aluminium (Rh/Al) were simulated in this study. The voltage of X-ray tube ranged from 24 kV to 32 kV with 2 kV intervals and the breast phantom thickness ranged from 3 to 8 cm, and glandular fraction g varied from 10% to 100%. MGD was measured for different anode/filter combinations and the effects of changing tube voltage, phantom thickness, combination and glandular breast tissue on MGD were studied. As glandular g and X-ray tube voltage increased, the breast dose increased too, and the increase of breast phantom thickness led to the decrease of MGD. The obtained results for MGD were consistent with the result of Boone et al. that was previously reported. By comparing the results, we saw that W/Rh anode/filter combination is the best choice in breast mammography imaging because of the lowest delivered dose in comparison with W/Ag and Rh/Al. Moreover, breast thickness and g value have significant effects on MGD.

Dosimetric verification of small fields in the lung using lung-equivalent polymer gel and Monte Carlo simulation.

PubMed

Gharehaghaji, Nahideh; Dadgar, Habib Alah

2018-01-01

The main purpose of this study was evaluate a polymer-gel-dosimeter (PGD) for three-dimensional verification of dose distributions in the lung that is called lung-equivalent gel (LEG) and then to compare its result with Monte Carlo (MC) method. In the present study, to achieve a lung density for PGD, gel is beaten until foam is obtained, and then sodium dodecyl sulfate is added as a surfactant to increase the surface tension of the gel. The foam gel was irradiated with 1 cm × 1 cm field size in the 6 MV photon beams of ONCOR SIEMENS LINAC, along the central axis of the gel. The LEG was then scanned on a 1.5 Tesla magnetic resonance imaging scanner after irradiation using a multiple-spin echo sequence. Least-square fitting the pixel values from 32 consecutive images using a single exponential decay function derived the R2 relaxation rates. Moreover, 6 and 18 MV photon beams of ONCOR SIEMENS LINAC are simulated using MCNPX MC Code. The MC model is used to calculate the depth dose water and low-density water resembling the soft tissue and lung, respectively. Percentages of dose reduction in the lung region relative to homogeneous phantom for 6 MV photon beam were 44.6%, 39%, 13%, and 7% for 0.5 cm × 0.5 cm, 1 cm × 1 cm, 2 cm × 2 cm, and 3 cm × 3 cm fields, respectively. For 18 MV photon beam, the results were found to be 82%, 69%, 46%, and 25.8% for the same field sizes, respectively. Preliminary results show good agreement between depth dose measured with the LEG and the depth dose calculated using MCNP code. Our study showed that the dose reduction with small fields in the lung was very high. Thus, inaccurate prediction of absorbed dose inside the lung and also lung/soft-tissue interfaces with small photon beams may lead to critical consequences for treatment outcome.

Differential Die-Away Instrument: Report on Benchmark Measurements and Comparison with Simulation for the Effects of Neutron Poisons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2015-03-30

In this report, new experimental data and MCNPX simulation results of the differential die-away (DDA) instrument response to the presence of neutron absorbers are evaluated. In our previous fresh nuclear fuel experiments and simulations, no neutron absorbers or poisons were included in the fuel definition. These new results showcase the capability of the DDA instrument to acquire data from a system that better mimics spent nuclear fuel.

Fluence correction factors for graphite calorimetry in a low-energy clinical proton beam: I. Analytical and Monte Carlo simulations.

PubMed

Palmans, H; Al-Sulaiti, L; Andreo, P; Shipley, D; Lühr, A; Bassler, N; Martinkovič, J; Dobrovodský, J; Rossomme, S; Thomas, R A S; Kacperek, A

2013-05-21

The conversion of absorbed dose-to-graphite in a graphite phantom to absorbed dose-to-water in a water phantom is performed by water to graphite stopping power ratios. If, however, the charged particle fluence is not equal at equivalent depths in graphite and water, a fluence correction factor, kfl, is required as well. This is particularly relevant to the derivation of absorbed dose-to-water, the quantity of interest in radiotherapy, from a measurement of absorbed dose-to-graphite obtained with a graphite calorimeter. In this work, fluence correction factors for the conversion from dose-to-graphite in a graphite phantom to dose-to-water in a water phantom for 60 MeV mono-energetic protons were calculated using an analytical model and five different Monte Carlo codes (Geant4, FLUKA, MCNPX, SHIELD-HIT and McPTRAN.MEDIA). In general the fluence correction factors are found to be close to unity and the analytical and Monte Carlo codes give consistent values when considering the differences in secondary particle transport. When considering only protons the fluence correction factors are unity at the surface and increase with depth by 0.5% to 1.5% depending on the code. When the fluence of all charged particles is considered, the fluence correction factor is about 0.5% lower than unity at shallow depths predominantly due to the contributions from alpha particles and increases to values above unity near the Bragg peak. Fluence correction factors directly derived from the fluence distributions differential in energy at equivalent depths in water and graphite can be described by kfl = 0.9964 + 0.0024·zw-eq with a relative standard uncertainty of 0.2%. Fluence correction factors derived from a ratio of calculated doses at equivalent depths in water and graphite can be described by kfl = 0.9947 + 0.0024·zw-eq with a relative standard uncertainty of 0.3%. These results are of direct relevance to graphite calorimetry in low-energy protons but given that the fluence correction factor is almost solely influenced by non-elastic nuclear interactions the results are also relevant for plastic phantoms that consist of carbon, oxygen and hydrogen atoms as well as for soft tissues.

Angular distributions of absorbed dose of Bremsstrahlung and secondary electrons induced by 18-, 28- and 38-MeV electron beams in thick targets.

PubMed

Takada, Masashi; Kosako, Kazuaki; Oishi, Koji; Nakamura, Takashi; Sato, Kouichi; Kamiyama, Takashi; Kiyanagi, Yoshiaki

2013-03-01

Angular distributions of absorbed dose of Bremsstrahlung photons and secondary electrons at a wide range of emission angles from 0 to 135°, were experimentally obtained using an ion chamber with a 0.6 cm(3) air volume covered with or without a build-up cap. The Bremsstrahlung photons and electrons were produced by 18-, 28- and 38-MeV electron beams bombarding tungsten, copper, aluminium and carbon targets. The absorbed doses were also calculated from simulated photon and electron energy spectra by multiplying simulated response functions of the ion chambers, simulated with the MCNPX code. Calculated-to-experimental (C/E) dose ratios obtained are from 0.70 to 1.57 for high-Z targets of W and Cu, from 15 to 135° and the C/E range from 0.6 to 1.4 at 0°; however, the values of C/E for low-Z targets of Al and C are from 0.5 to 1.8 from 0 to 135°. Angular distributions at the forward angles decrease with increasing angles; on the other hand, the angular distributions at the backward angles depend on the target species. The dependences of absorbed doses on electron energy and target thickness were compared between the measured and simulated results. The attenuation profiles of absorbed doses of Bremsstrahlung beams at 0, 30 and 135° were also measured.

Leakage of radioactive materials from particle accelerator facilities by non-radiation disasters like fire and flooding and its environmental impacts

NASA Astrophysics Data System (ADS)

Lee, A.; Jung, N. S.; Mokhtari Oranj, L.; Lee, H. S.

2018-06-01

The leakage of radioactive materials generated at particle accelerator facilities is one of the important issues in the view of radiation safety. In this study, fire and flooding at particle accelerator facilities were considered as the non-radiation disasters which result in the leakage of radioactive materials. To analyse the expected effects at each disaster, the case study on fired and flooded particle accelerator facilities was carried out with the property investigation of interesting materials presented in the accelerator tunnel and the activity estimation. Five major materials in the tunnel were investigated: dust, insulators, concrete, metals and paints. The activation levels on the concerned materials were calculated using several Monte Carlo codes (MCNPX 2.7+SP-FISPACT 2007, FLUKA 2011.4c and PHITS 2.64+DCHAIN-SP 2001). The impact weight to environment was estimated for the different beam particles (electron, proton, carbon and uranium) and the different beam energies (100, 430, 600 and 1000 MeV/nucleon). With the consideration of the leakage path of radioactive materials due to fire and flooding, the activation level of selected materials, and the impacts to the environment were evaluated. In the case of flooding, dust, concrete and metal were found as a considerable object. In the case of fire event, dust, insulator and paint were the major concerns. As expected, the influence of normal fire and flooding at electron accelerator facilities would be relatively low for both cases.

Monte Carlo Determination of Dosimetric Parameters of a New (125)I Brachytherapy Source According to AAPM TG-43 (U1) Protocol.

PubMed

Baghani, Hamid Reza; Lohrabian, Vahid; Aghamiri, Mahmoud Reza; Robatjazi, Mostafa

2016-03-01

(125)I is one of the important sources frequently used in brachytherapy. Up to now, several different commercial models of this source type have been introduced to the clinical radiation oncology applications. Recently, a new source model, IrSeed-125, has been added to this list. The aim of the present study is to determine the dosimetric parameters of this new source model based on the recommendations of TG-43 (U1) protocol using Monte Carlo simulation. The dosimetric characteristics of Ir-125 including dose rate constant, radial dose function, 2D anisotropy function and 1D anisotropy function were determined inside liquid water using MCNPX code and compared to those of other commercially available iodine sources. The dose rate constant of this new source was found to be 0.983+0.015 cGyh-1U-1 that was in good agreement with the TLD measured data (0.965 cGyh-1U-1). The 1D anisotropy function at 3, 5, and 7 cm radial distances were obtained as 0.954, 0.953 and 0.959, respectively. The results of this study showed that the dosimetric characteristics of this new brachytherapy source are comparable with those of other commercially available sources. Furthermore, the simulated parameters were in accordance with the previously measured ones. Therefore, the Monte Carlo calculated dosimetric parameters could be employed to obtain the dose distribution around this new brachytherapy source based on TG-43 (U1) protocol.

Construction of boundary-surface-based Chinese female astronaut computational phantom and proton dose estimation

PubMed Central

Sun, Wenjuan; JIA, Xianghong; XIE, Tianwu; XU, Feng; LIU, Qian

2013-01-01

With the rapid development of China's space industry, the importance of radiation protection is increasingly prominent. To provide relevant dose data, we first developed the Visible Chinese Human adult Female (VCH-F) phantom, and performed further modifications to generate the VCH-F Astronaut (VCH-FA) phantom, incorporating statistical body characteristics data from the first batch of Chinese female astronauts as well as reference organ mass data from the International Commission on Radiological Protection (ICRP; both within 1% relative error). Based on cryosection images, the original phantom was constructed via Non-Uniform Rational B-Spline (NURBS) boundary surfaces to strengthen the deformability for fitting the body parameters of Chinese female astronauts. The VCH-FA phantom was voxelized at a resolution of 2 × 2 × 4 mm3for radioactive particle transport simulations from isotropic protons with energies of 5000–10 000 MeV in Monte Carlo N-Particle eXtended (MCNPX) code. To investigate discrepancies caused by anatomical variations and other factors, the obtained doses were compared with corresponding values from other phantoms and sex-averaged doses. Dose differences were observed among phantom calculation results, especially for effective dose with low-energy protons. Local skin thickness shifts the breast dose curve toward high energy, but has little impact on inner organs. Under a shielding layer, organ dose reduction is greater for skin than for other organs. The calculated skin dose per day closely approximates measurement data obtained in low-Earth orbit (LEO). PMID:23135158

Electrical Properties of MWCNT/HDPE Composite-Based MSM Structure Under Neutron Irradiation

NASA Astrophysics Data System (ADS)

Kasani, H.; Khodabakhsh, R.; Taghi Ahmadi, M.; Rezaei Ochbelagh, D.; Ismail, Razali

2017-04-01

Because of their low cost, low energy consumption, high performance, and exceptional electrical properties, nanocomposites containing carbon nanotubes are suitable for use in many applications such as sensing systems. In this research work, a metal-semiconductor-metal (MSM) structure based on a multiwall carbon nanotube/high-density polyethylene (MWCNT/HDPE) nanocomposite is introduced as a neutron sensor. Scanning electron microscopy, Fourier-transform infrared, and infrared spectroscopy techniques were used to characterize the morphology and structure of the fabricated device. Current-voltage ( I- V) characteristic modeling showed that the device can be assumed to be a reversed-biased Schottky diode, if the voltage is high enough. To estimate the depletion layer length of the Schottky contact, impedance spectroscopy was employed. Therefore, the real and imaginary parts of the impedance of the MSM system were used to obtain electrical parameters such as the carrier mobility and dielectric constant. Experimental observations of the MSM structure under irradiation from an americium-beryllium (Am-Be) neutron source showed that the current level in the device decreased significantly. Subsequently, current pulses appeared in situ I- V and current-time ( I- t) curve measurements when increasing voltage was applied to the MSM system. The experimentally determined depletion region length as well as the space-charge-limited current mechanism for carrier transport were compared with the range for protons calculated using Monte Carlo n-particle extended (MCNPX) code, yielding the maximum energy of recoiled protons detectable by the device.

Lunar Surface Reactor Shielding Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kang, Shawn; McAlpine, William; Lipinski, Ronald

A nuclear reactor system could provide power to support long term human exploration of the moon. Such a system would require shielding to protect astronauts from its emitted radiations. Shielding studies have been performed for a Gas Cooled Reactor system because it is considered to be the most suitable nuclear reactor system available for lunar exploration, based on its tolerance of oxidizing lunar regolith and its good conversion efficiency. The goals of the shielding studies were to determine a material shielding configuration that reduces the dose (rem) to the required level in order to protect astronauts, and to estimate themore » mass of regolith that would provide an equivalent protective effect if it were used as the shielding material. All calculations were performed using MCNPX, a Monte Carlo transport code. Lithium hydride must be kept between 600 K and 700 K to prevent excessive swelling from large amounts of gamma or neutron irradiation. The issue is that radiation damage causes separation of the lithium and the hydrogen, resulting in lithium metal and hydrogen gas. The proposed design uses a layer of B4C to reduce the combined neutron and gamma dose to below 0.5Grads before the LiH is introduced. Below 0.5Grads the swelling in LiH is small (less than about 1%) for all temperatures. This approach causes the shield to be heavier than if the B4C were replaced by LiH, but it makes the shield much more robust and reliable.« less

The influence of patient size on dose conversion coefficients: a hybrid phantom study for adult cardiac catheterization

NASA Astrophysics Data System (ADS)

Johnson, Perry; Lee, Choonsik; Johnson, Kevin; Siragusa, Daniel; Bolch, Wesley E.

2009-06-01

In this study, the influence of patient size on organ and effective dose conversion coefficients (DCCs) was investigated for a representative interventional fluoroscopic procedure—cardiac catheterization. The study was performed using hybrid phantoms representing an underweight, average and overweight American adult male. Reference body sizes were determined using the NHANES III database and parameterized based on standing height and total body mass. Organ and effective dose conversion coefficients were calculated for anterior-posterior, posterior-anterior, left anterior oblique and right anterior oblique projections using the Monte Carlo code MCNPX 2.5.0 with the metric dose area product being used as the normalization factor. Results show body size to have a clear influence on DCCs which increased noticeably when body size decreased. It was also shown that if patient size is neglected when choosing a DCC, the organ and effective dose will be underestimated to an underweight patient and will be overestimated to an underweight patient, with errors as large as 113% for certain projections. Results were further compared with those published for a KTMAN-2 Korean patient-specific tomographic phantom. The published DCCs aligned best with the hybrid phantom which most closely matched in overall body size. These results highlighted the need for and the advantages of phantom-patient matching, and it is recommended that hybrid phantoms be used to create a more diverse library of patient-dependent anthropomorphic phantoms for medical dose reconstruction.

A methodology for efficiency optimization of betavoltaic cell design using an isotropic planar source having an energy dependent beta particle distribution.

PubMed

Theirrattanakul, Sirichai; Prelas, Mark

2017-09-01

Nuclear batteries based on silicon carbide betavoltaic cells have been studied extensively in the literature. This paper describes an analysis of design parameters, which can be applied to a variety of materials, but is specific to silicon carbide. In order to optimize the interface between a beta source and silicon carbide p-n junction, it is important to account for the specific isotope, angular distribution of the beta particles from the source, the energy distribution of the source as well as the geometrical aspects of the interface between the source and the transducer. In this work, both the angular distribution and energy distribution of the beta particles are modeled using a thin planar beta source (e.g., H-3, Ni-63, S-35, Pm-147, Sr-90, and Y-90) with GEANT4. Previous studies of betavoltaics with various source isotopes have shown that Monte Carlo based codes such as MCNPX, GEANT4 and Penelope generate similar results. GEANT4 is chosen because it has important strengths for the treatment of electron energies below one keV and it is widely available. The model demonstrates the effects of angular distribution, the maximum energy of the beta particle and energy distribution of the beta source on the betavoltaic and it is useful in determining the spatial profile of the power deposition in the cell. Copyright © 2017. Published by Elsevier Ltd.

Re-186 and Sm-153 dosimetry based on scintigraphic imaging data in skeletal metastasis palliative treatment and Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Andreou, M.; Lagopati, N.; Lyra, M.

2011-09-01

Optimum treatment planning of patients suffering from painful skeletal metastases requires accurate calculations concerning absorbed dose in metastatic lesions and critical organs, such as red marrow. Delivering high doses to tumor cells while limiting radiation dose to normal tissue, is the key for successful palliation treatment. The aim of this study is to compare the dosimetric calculations, obtained by Monte Carlo (MC) simulation and the MIRDOSE model, in therapeutic schemes of skeleton metastatic lesions, with Rhenium-186 (Sn) -HEDP and Samarium-153 -EDTMP. A bolus injection of 1295 MBq (35mCi) Re-186- HEDP was infused in 11 patients with multiple skeletal metastases. The administered dose for the 8 patients who received Sm-153 was 1 mCi /kg. Planar scintigraphic images for the two groups of patients were obtained, 24 h, 48 h and 72 h post injection, by an Elscint Apex SPX gamma camera. The images were processed, utilizing ROI quantitative methods, to determine residence times and radionuclide uptakes. Dosimetric calculations were performed using the patient specific scintigraphic data by the MIRDOSE3 code of MIRD. Also, MCNPX was employed, simulating the distribution of the radioisotope in the ROI and calculating the absorbed doses in the metastatic lesion, and in critical organs. Summarizing, there is a good agreement between the results, derived from the two pathways, the patient specific and the mathematical, with a deviation of less than 9% for planar scintigraphic data compared to MC, for both radiopharmaceuticals.

Neutron production from beam-modifying devices in a modern double scattering proton therapy beam delivery system

PubMed Central

Pérez-Andújar, Angélica; Newhauser, Wayne D; DeLuca, Paul M

2014-01-01

In this work the neutron production in a passive beam delivery system was investigated. Secondary particles including neutrons are created as the proton beam interacts with beam shaping devices in the treatment head. Stray neutron exposure to the whole body may increase the risk that the patient develops a radiogenic cancer years or decades after radiotherapy. We simulated a passive proton beam delivery system with double scattering technology to determine the neutron production and energy distribution at 200 MeV proton energy. Specifically, we studied the neutron absorbed dose per therapeutic absorbed dose, the neutron absorbed dose per source particle and the neutron energy spectrum at various locations around the nozzle. We also investigated the neutron production along the nozzle's central axis. The absorbed doses and neutron spectra were simulated with the MCNPX Monte Carlo code. The simulations revealed that the range modulation wheel (RMW) is the most intense neutron source of any of the beam spreading devices within the nozzle. This finding suggests that it may be helpful to refine the design of the RMW assembly, e.g., by adding local shielding, to suppress neutron-induced damage to components in the nozzle and to reduce the shielding thickness of the treatment vault. The simulations also revealed that the neutron dose to the patient is predominated by neutrons produced in the field defining collimator assembly, located just upstream of the patient. PMID:19147903

Organ and effective dose conversion coefficients for a sitting female hybrid computational phantom exposed to monoenergetic protons in idealized irradiation geometries.

PubMed

Alves, M C; Santos, W S; Lee, Choonsik; Bolch, Wesley E; Hunt, John G; Carvalho Júnior, A B

2014-12-21

The conversion coefficients (CCs) relate protection quantities, mean absorbed dose (DT) and effective dose (E), with physical radiation field quantities, such as fluence (Φ). The calculation of CCs through Monte Carlo simulations is useful for estimating the dose in individuals exposed to radiation. The aim of this work was the calculation of conversion coefficients for absorbed and effective doses per fluence (DT/ Φ and E/Φ) using a sitting and standing female hybrid phantom (UFH/NCI) exposure to monoenergetic protons with energy ranging from 2 MeV to 10 GeV. The radiation transport code MCNPX was used to develop exposure scenarios implementing the female UFH/NCI phantom in sitting and standing postures. Whole-body irradiations were performed using the recommended irradiation geometries by ICRP publication 116 (AP, PA, RLAT, LLAT, ROT and ISO). In most organs, the conversion coefficients DT/Φ were similar for both postures. However, relative differences were significant for organs located in the abdominal region, such as ovaries, uterus and urinary bladder, especially in the AP, RLAT and LLAT geometries. Anatomical differences caused by changing the posture of the female UFH/NCI phantom led an attenuation of incident protons with energies below 150 MeV by the thigh of the phantom in the sitting posture, for the front-to-back irradiation, and by the arms and hands of the phantom in the standing posture, for the lateral irradiation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Han, S; Ji, Y; Kim, K

Purpose: A diagnostics Multileaf Collimator (MLC) was designed for diagnostic radiography dose reduction. Monte Carlo simulation was used to evaluate efficiency of shielding material for producing leaves of Multileaf collimator. Material & Methods: The general radiography unit (Rex-650R, Listem, Korea) was modeling with Monte Carlo simulation (MCNPX, LANL, USA) and we used SRS-78 program to calculate the energy spectrum of tube voltage (80, 100, 120 kVp). The shielding materials was SKD 11 alloy tool steel that is composed of 1.6% carbon(C), 0.4% silicon (Si), 0.6% manganese (Mn), 5% chromium (Cr), 1% molybdenum (Mo), and vanadium (V). The density of itmore » was 7.89 g/m3. We simulated leafs diagnostic MLC using SKD 11 with general radiography unit. We calculated efficiency of diagnostic MLC using tally6 card of MCNPX depending on energy. Results: The diagnostic MLC consisted of 25 individual metal shielding leaves on both sides, with dimensions of 10 × 0.5 × 0.5 cm3. The leaves of MLC were controlled by motors positioned on both sides of the MLC. According to energy (tube voltage), the shielding efficiency of MLC in Monte Carlo simulation was 99% (80 kVp), 96% (100 kVp) and 93% (120 kVp). Conclusion: We certified efficiency of diagnostic MLC fabricated from SKD11 alloy tool steel. Based on the results, the diagnostic MLC was designed. We will make the diagnostic MLC for dose reduction of diagnostic radiography.« less

Proton Therapy At Siteman Cancer Center: The State Of The Art

NASA Astrophysics Data System (ADS)

Bloch, Charles

2011-06-01

Barnes-Jewish Hospital is on the verge of offering proton radiation therapy to its patients. Those treatments will be delivered from the first Monarch 250, a state-of-the-art cyclotron produced by Still River Systems, Inc., Littleton, MA. The accelerator is the world's first superconducting synchrocyclotron, with a field-strength of 10 tesla, providing the smallest accelerator for high-energy protons currently available. On May 14, 2010 it was announced that the first production unit had successfully extracted 250 MeV protons. That unit is scheduled for delivery to the Siteman Cancer Center, an NCI-designated Comprehensive Cancer Center at Washington University School of Medicine. At a weight of 20 tons and with a diameter of less than 2 meters the compact cyclotron will be mounted on a gantry, another first for proton therapy systems. The single-energy system includes 3 contoured scatterers and 14 different range modulators to provide 24 distinct beam delivery configurations. This allows proton fields up to 25 cm in diameter, with a maximum range from 5.5 to 32 cm and spread-out-Bragg-peak extent up to 20 cm. Monte Carlo simulations have been run using MCNPX to simulate the clinical beam properties. Those calculations have been used to commission a commercial treatment planning system prior to final clinical measurements. MCNPX was also used to calculate the neutron background generated by protons in the scattering system and patient. Additional details of the facility and current status will be presented.

Development of a neutron measurement system in unified non-destructive assay for the PRIDE facility

NASA Astrophysics Data System (ADS)

Seo, Hee; Park, Se-Hwan; Won, Byung-Hee; Ahn, Seong-Kyu; Shin, Hee-Sung; Na, Sang-Ho; Song, Dae-Yong; Kim, Ho-Dong; Lee, Seung Kyu

2013-12-01

The Korea Atomic Energy Research Institute (KAERI) has made an effort to develop pyroprocessing technology to resolve an on-going problem in Korea, i.e., the management of spent nuclear fuels. To this end, a test-bed facility for pyroprocessing, called PRIDE (PyRoprocessing Integrated inactive DEmonstration facility), is being constructed at KAERI. The main objective of PRIDE is to evaluate the performance of the unit processes, remote operation, maintenance, and proliferation resistance. In addition, integrating all unit processes into a one-step process is also one of the main goals. PRIDE can also provide a good opportunity to test safeguards instrumentations for a pyroprocessing facility such as nuclear material accounting devices, surveillance systems, radiation monitoring systems, and process monitoring systems. In the present study, a non-destructive assay (NDA) system for the testing of nuclear material accountancy of PRIDE was designed by integrating three different NDA techniques, i.e., neutron, gamma-ray, and mass measurements. The developed neutron detection module consists of 56 3He tubes and 16 AMPTEK A111 signal processing circuits. The amplifiers were matched in terms of the gain and showed good uniformity after a gain-matching procedure (%RSD=0.37%). The axial and the radial efficiency distributions within the cavity were then measured using a 252Cf neutron source and were compared with the MCNPX calculation results. The measured efficiency distributions showed excellent agreement with the calculations, which confirmed the accuracy of the MCNPX model of the system.

Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Botta, F; Di Dia, A; Pedroli, G

The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK),more » quantifying the energy deposition all around a point isotropic source, is often the one.Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10–3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I, 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8·RCSDA and 0.9·RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8·X90 and 0.9·X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9·RCSDA and 0.9·X90 for electrons and isotopes, respectively.Results: Concerning monoenergetic electrons, within 0.8·RCSDA (where 90%–97% of the particle energy is deposed), fluka and penelope agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The discrepancies between fluka and the other codes are of the same order of magnitude than those observed when comparing the other codes among them, which can be referred to the different simulation algorithms. When considering the beta spectra, discrepancies notably reduce: within 0.9·X90, fluka and penelope differ for less than 1% in water and less than 2% in bone with any of the isotopes here considered. Complete data of fluka DPKs are given as Supplementary Material as a tool to perform dosimetry by analytical point kernel convolution.Conclusions: fluka provides reliable results when transporting electrons in the low energy range, proving to be an adequate tool for nuclear medicine dosimetry.« less

Active-Interrogation Measurements of Fast Neutrons from Induced Fission in Low-Enriched Uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. L. Dolan; M. J. Marcath; M. Flaska

2014-02-01

A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre (JRC) in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutron to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials.

Fan-beam intensity modulated proton therapy.

PubMed

Hill, Patrick; Westerly, David; Mackie, Thomas

2013-11-01

This paper presents a concept for a proton therapy system capable of delivering intensity modulated proton therapy using a fan beam of protons. This system would allow present and future gantry-based facilities to deliver state-of-the-art proton therapy with the greater normal tissue sparing made possible by intensity modulation techniques. A method for producing a divergent fan beam of protons using a pair of electromagnetic quadrupoles is described and particle transport through the quadrupole doublet is simulated using a commercially available software package. To manipulate the fan beam of protons, a modulation device is developed. This modulator inserts or retracts acrylic leaves of varying thickness from subsections of the fan beam. Each subsection, or beam channel, creates what effectively becomes a beam spot within the fan area. Each channel is able to provide 0-255 mm of range shift for its associated beam spot, or stop the beam and act as an intensity modulator. Results of particle transport simulations through the quadrupole system are incorporated into the MCNPX Monte Carlo transport code along with a model of the range and intensity modulation device. Several design parameters were investigated and optimized, culminating in the ability to create topotherapy treatment plans using distal-edge tracking on both phantom and patient datasets. Beam transport calculations show that a pair of electromagnetic quadrupoles can be used to create a divergent fan beam of 200 MeV protons over a distance of 2.1 m. The quadrupole lengths were 30 and 48 cm, respectively, with transverse field gradients less than 20 T/m, which is within the range of water-cooled magnets for the quadrupole radii used. MCNPX simulations of topotherapy treatment plans suggest that, when using the distal edge tracking delivery method, many delivery angles are more important than insisting on narrow beam channel widths in order to obtain conformal target coverage. Overall, the sharp distal falloff of a proton depth-dose distribution was found to provide sufficient control over the dose distribution to meet objectives, even with coarse lateral resolution and channel widths as large as 2 cm. Treatment plans on both phantom and patient data show that dose conformity suffers when treatments are delivered from less than approximately ten angles. Treatment time for a sample prostate delivery is estimated to be on the order of 10 min, and neutron production is estimated to be comparable to that found for existing collimated systems. Fan beam proton therapy is a method of delivering intensity modulated proton therapy which may be employed as an alternative to magnetic scanning systems. A fan beam of protons can be created by a set of quadrupole magnets and modified by a dual-purpose range and intensity modulator. This can be used to deliver inversely planned treatments, with spot intensities optimized to meet user defined dose objectives. Additionally, the ability of a fan beam delivery system to effectively treat multiple beam spots simultaneously may provide advantages as compared to spot scanning deliveries.

131I activity quantification of gamma camera planar images.

PubMed

Barquero, Raquel; Garcia, Hugo P; Incio, Monica G; Minguez, Pablo; Cardenas, Alexander; Martínez, Daniel; Lassmann, Michael

2017-02-07

A procedure to estimate the activity in target tissues in patients during the therapeutic administration of 131 I radiopharmaceutical treatment for thyroid conditions (hyperthyroidism and differentiated thyroid cancer) using a gamma camera (GC) with a high energy (HE) collimator, is proposed. Planar images are acquired for lesions of different sizes r, and at different distances d, in two HE GC systems. Defining a region of interest (ROI) on the image of size r, total counts n g are measured. Sensitivity S (cps MBq -1 ) in each acquisition is estimated as the product of the geometric G and the intrinsic efficiency η 0 . The mean fluence of 364 keV photons arriving at the ROI per disintegration G, is calculated with the MCNPX code, simulating the entire GC and the HE collimator. Intrinsic efficiency η 0 is estimated from a calibration measurement of a plane reference source of 131 I in air. Values of G and S for two GC systems-Philips Skylight and Siemens e-cam-are calculated. The total range of possible sensitivity values in thyroidal imaging in the e-cam and skylight GC measure from 7 cps MBq -1 to 35 cps MBq -1 , and from 6 cps MBq -1 to 29 cps MBq -1 , respectively. These sensitivity values have been verified with the SIMIND code, with good agreement between them. The results have been validated with experimental measurements in air, and in a medium with scatter and attenuation. The counts in the ROI can be produced by direct, scatter and penetration photons. The fluence value for direct photons is constant for any r and d values, but scatter and penetration photons show different values related to specific r and d values, resulting in the large sensitivity differences found. The sensitivity in thyroidal GC planar imaging is strongly dependent on uptake size, and distance from the GC. An individual value for the acquisition sensitivity of each lesion can significantly alleviate the level of uncertainty in the measurement of thyroid uptake activity for each patient.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

131I activity quantification of gamma camera planar images

NASA Astrophysics Data System (ADS)

Barquero, Raquel; Garcia, Hugo P.; Incio, Monica G.; Minguez, Pablo; Cardenas, Alexander; Martínez, Daniel; Lassmann, Michael

2017-02-01

A procedure to estimate the activity in target tissues in patients during the therapeutic administration of 131I radiopharmaceutical treatment for thyroid conditions (hyperthyroidism and differentiated thyroid cancer) using a gamma camera (GC) with a high energy (HE) collimator, is proposed. Planar images are acquired for lesions of different sizes r, and at different distances d, in two HE GC systems. Defining a region of interest (ROI) on the image of size r, total counts n g are measured. Sensitivity S (cps MBq-1) in each acquisition is estimated as the product of the geometric G and the intrinsic efficiency η 0. The mean fluence of 364 keV photons arriving at the ROI per disintegration G, is calculated with the MCNPX code, simulating the entire GC and the HE collimator. Intrinsic efficiency η 0 is estimated from a calibration measurement of a plane reference source of 131I in air. Values of G and S for two GC systems—Philips Skylight and Siemens e-cam—are calculated. The total range of possible sensitivity values in thyroidal imaging in the e-cam and skylight GC measure from 7 cps MBq-1 to 35 cps MBq-1, and from 6 cps MBq-1 to 29 cps MBq-1, respectively. These sensitivity values have been verified with the SIMIND code, with good agreement between them. The results have been validated with experimental measurements in air, and in a medium with scatter and attenuation. The counts in the ROI can be produced by direct, scatter and penetration photons. The fluence value for direct photons is constant for any r and d values, but scatter and penetration photons show different values related to specific r and d values, resulting in the large sensitivity differences found. The sensitivity in thyroidal GC planar imaging is strongly dependent on uptake size, and distance from the GC. An individual value for the acquisition sensitivity of each lesion can significantly alleviate the level of uncertainty in the measurement of thyroid uptake activity for each patient.

Assessment of Mean Glandular Dose in Mammography System with Different Anode-Filter Combinations Using MCNP Code

PubMed Central

Gholamkar, Lida; Mowlavi, Ali Asghar; Sadeghi, Mahdi; Athari, Mitra

2016-01-01

Background X-ray mammography is one of the general methods for early detection of breast cancer. Since glandular tissue in the breast is sensitive to radiation and it increases the risk of cancer, the given dose to the patient is very important in mammography. Objectives The aim of this study was to determine the average absorbed dose of X-ray radiation in the glandular tissue of the breast during mammography examinations as well as investigating factors that influence the mean glandular dose (MGD). One of the precise methods for determination of MGD absorbed by the breast is Monte Carlo simulation method which is widely used to assess the dose. Materials and Methods We studied some different X-ray sources and exposure factors that affect the MGD. “Midi-future” digital mammography system with amorphous-selenium detector was simulated using the Monte Carlo N-particle extended (MCNPX) code. Different anode/filter combinations such as tungsten/silver (W/Ag), tungsten/rhodium (W/Rh), and rhodium/aluminium (Rh/Al) were simulated in this study. The voltage of X-ray tube ranged from 24 kV to 32 kV with 2 kV intervals and the breast phantom thickness ranged from 3 to 8 cm, and glandular fraction g varied from 10% to 100%. Results MGD was measured for different anode/filter combinations and the effects of changing tube voltage, phantom thickness, combination and glandular breast tissue on MGD were studied. As glandular g and X-ray tube voltage increased, the breast dose increased too, and the increase of breast phantom thickness led to the decrease of MGD. The obtained results for MGD were consistent with the result of Boone et al. that was previously reported. Conclusion By comparing the results, we saw that W/Rh anode/filter combination is the best choice in breast mammography imaging because of the lowest delivered dose in comparison with W/Ag and Rh/Al. Moreover, breast thickness and g value have significant effects on MGD. PMID:27895876

SU-E-T-753: Three-Dimensional Dose Distributions of Incident Proton Particle in the Polymer Gel Dosimeter and the Radiochromic Gel Dosimeter: A Simulation Study with MCNP Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, M; Kim, G; Ji, Y

Purpose: The purpose of this study is to estimate the three-dimensional dose distributions in the polymer and the radiochromic gel dosimeter, and to identify the detectability of both gel dosimeters by comparing with the water phantom in case of irradiating the proton particles. Methods: The normoxic polymer gel and the LCV micelle radiochromic gel were used in this study. The densities of polymer and the radiochromic gel dosimeter were 1.024 and 1.005 g/cm{sup 3}, respectively. The dose distributions of protons in the polymer and radiochromic gel were simulated using Monte Carlo radiation transport code (MCNPX, Los Alamos National Laboratory). Themore » shape of phantom irradiated by proton particles was a hexahedron with the dimension of 12.4 × 12.4 × 15.0 cm{sup 3}. The energies of proton beam were 50, 80, and 140 MeV energies were directed to top of the surface of phantom. The cross-sectional view of proton dose distribution in both gel dosimeters was estimated with the water phantom and evaluated by the gamma evaluation method. In addition, the absorbed dose(Gy) was also calculated for evaluating the proton detectability. Results: The evaluation results show that dose distributions in both gel dosimeters at intermediated section and Bragg-peak region are similar with that of the water phantom. At entrance section, however, inconsistencies of dose distribution are represented, compared with water. The relative absorbed doses in radiochromic and polymer gel dosimeter were represented to be 0.47 % and 2.26 % difference, respectively. These results show that the radiochromic gel dosimeter was better matched than the water phantom in the absorbed dose evaluation. Conclusion: The polymer and the radiochromic gel dosimeter show similar characteristics in dose distributions for the proton beams at intermediate section and Bragg-peak region. Moreover the calculated absorbed dose in both gel dosimeters represents similar tendency by comparing with that in water phantom.« less

Accelerator driven neutron source design via beryllium target and 208Pb moderator for boron neutron capture therapy in alternative treatment strategy by Monte Carlo method.

PubMed

Khorshidi, Abdollah

2017-01-01

The reactor has increased its area of application into medicine especially boron neutron capture therapy (BNCT); however, accelerator-driven neutron sources can be used for therapy purposes. The present study aimed to discuss an alternative method in BNCT functions by a small cyclotron with low current protons based on Karaj cyclotron in Iran. An epithermal neutron spectrum generator was simulated with 30 MeV proton energy for BNCT purposes. A low current of 300 μA of the proton beam in spallation target concept via 9Be target was accomplished to model neutron spectrum using 208Pb moderator around the target. The graphite reflector and dual layer collimator were planned to prevent and collimate the neutrons produced from proton interactions. Neutron yield per proton, energy distribution, flux, and dose components in the simulated head phantom were estimated by MCNPX code. The neutron beam quality was investigated by diverse filters thicknesses. The maximum epithermal flux transpired using Fluental, Fe, Li, and Bi filters with thicknesses of 7.4, 3, 0.5, and 4 cm, respectively; as well as the epithermal to thermal neutron flux ratio was 161. Results demonstrated that the induced neutrons from a low energy and low current proton may be effective in tumor therapy using 208Pb moderator with average lethargy and also graphite reflector with low absorption cross section to keep the generated neutrons. Combination of spallation-based BNCT and proton therapy can be especially effective, if a high beam intensity cyclotron becomes available.

Dose conversion coefficients for monoenergetic electrons incident on a realistic human eye model with different lens cell populations

NASA Astrophysics Data System (ADS)

Nogueira, P.; Zankl, M.; Schlattl, H.; Vaz, P.

2011-11-01

The radiation-induced posterior subcapsular cataract has long been generally accepted to be a deterministic effect that does not occur at doses below a threshold of at least 2 Gy. Recent epidemiological studies indicate that the threshold for cataract induction may be much lower or that there may be no threshold at all. A thorough study of this subject requires more accurate dose estimates for the eye lens than those available in ICRP Publication 74. Eye lens absorbed dose per unit fluence conversion coefficients for electron irradiation were calculated using a geometrical model of the eye that takes into account different cell populations of the lens epithelium, together with the MCNPX Monte Carlo radiation transport code package. For the cell population most sensitive to ionizing radiation—the germinative cells—absorbed dose per unit fluence conversion coefficients were determined that are up to a factor of 4.8 higher than the mean eye lens absorbed dose conversion coefficients for electron energies below 2 MeV. Comparison of the results with previously published values for a slightly different eye model showed generally good agreement for all electron energies. Finally, the influence of individual anatomical variability was quantified by positioning the lens at various depths below the cornea. A depth difference of 2 mm between the shallowest and the deepest location of the germinative zone can lead to a difference between the resulting absorbed doses of up to nearly a factor of 5000 for electron energy of 0.7 MeV.

Dose conversion coefficients for monoenergetic electrons incident on a realistic human eye model with different lens cell populations.

PubMed

Nogueira, P; Zankl, M; Schlattl, H; Vaz, P

2011-11-07

The radiation-induced posterior subcapsular cataract has long been generally accepted to be a deterministic effect that does not occur at doses below a threshold of at least 2 Gy. Recent epidemiological studies indicate that the threshold for cataract induction may be much lower or that there may be no threshold at all. A thorough study of this subject requires more accurate dose estimates for the eye lens than those available in ICRP Publication 74. Eye lens absorbed dose per unit fluence conversion coefficients for electron irradiation were calculated using a geometrical model of the eye that takes into account different cell populations of the lens epithelium, together with the MCNPX Monte Carlo radiation transport code package. For the cell population most sensitive to ionizing radiation-the germinative cells-absorbed dose per unit fluence conversion coefficients were determined that are up to a factor of 4.8 higher than the mean eye lens absorbed dose conversion coefficients for electron energies below 2 MeV. Comparison of the results with previously published values for a slightly different eye model showed generally good agreement for all electron energies. Finally, the influence of individual anatomical variability was quantified by positioning the lens at various depths below the cornea. A depth difference of 2 mm between the shallowest and the deepest location of the germinative zone can lead to a difference between the resulting absorbed doses of up to nearly a factor of 5000 for electron energy of 0.7 MeV.

Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

NASA Astrophysics Data System (ADS)

Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

2016-08-01

Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

Measurement and simulation of the TRR BNCT beam parameters

NASA Astrophysics Data System (ADS)

Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser; Golshanian, Mohadeseh; Ghods, Hossein; Ezzati, Arsalan; Keyvani, Mehdi; Haddadi, Mohammad

2016-09-01

Recently, the configuration of the Tehran Research Reactor (TRR) thermal column has been modified and a proper thermal neutron beam for preclinical Boron Neutron Capture Therapy (BNCT) has been obtained. In this study, simulations and experimental measurements have been carried out to identify the BNCT beam parameters including the beam uniformity, the distribution of the thermal neutron dose, boron dose, gamma dose in a phantom and also the Therapeutic Gain (TG). To do this, the entire TRR structure including the reactor core, pool, the thermal column and beam tubes have been modeled using MCNPX Monte Carlo code. To measure in-phantom dose distribution a special head phantom has been constructed and foil activation techniques and TLD700 dosimeter have been used. The results show that there is enough uniformity in TRR thermal BNCT beam. TG parameter has the maximum value of 5.7 at the depth of 1 cm from the surface of the phantom, confirming that TRR thermal neutron beam has potential for being used in treatment of superficial brain tumors. For the purpose of a clinical trial, more modifications need to be done at the reactor, as, for example design, and construction of a treatment room at the beam exit which is our plan for future. To date, this beam is usable for biological studies and animal trials. There is a relatively good agreement between simulation and measurement especially within a diameter of 10 cm which is the dimension of usual BNCT beam ports. This relatively good agreement enables a more precise prediction of the irradiation conditions needed for future experiments.

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

Assessment of Neutron Contamination Originating from the Presence of Wedge and Block in Photon Beam Radiotherapy.

PubMed

Bahreyni Toossi, M T; Khajetash, B; Ghorbani, M

2018-03-01

One of the main causes of induction of secondary cancer in radiation therapy is neutron contamination received by patients during treatment. Objective: In the present study the impact of wedge and block on neutron contamination production is investigated. The evaluations are conducted for a 15 MV Siemens Primus linear accelerator. Simulations were performed using MCNPX Monte Carlo code. 30˚, 45˚ and 60˚ wedges and a cerrobend block with dimensions of 1.5 × 1.5 × 7 cm 3 were simulated. The investigation were performed in the 10 × 10 cm 2 field size at source to surface distance of 100 cm for depth of 0.5, 2, 3 and 4 cm in a water phantom. Neutron dose was calculated using F4 tally with flux to dose conversion factors and F6 tally. Results showed that the presence of wedge increases the neutron contamination when the wedge factor was considered. In addition, 45˚ wedge produced the most amount of neutron contamination. If the block is in the center of the field, the cerrobend block caused less neutron contamination than the open field due to absorption of neutrons and photon attenuation. The results showed that neutron contamination is less in steeper depths. The results for two tallies showed practically equivalent results. Wedge causes neutron contamination hence should be considered in therapeutic protocols in which wedge is used. In terms of clinical aspects, the results of this study show that superficial tissues such as skin will tolerate more neutron contamination than the deep tissues.

A macroscopic and microscopic study of radon exposure using Geant4 and MCNPX to estimate dose rates and DNA damage

NASA Astrophysics Data System (ADS)

van den Akker, Mary Evelyn

Radon is considered the second-leading cause of lung cancer after smoking. Epidemiological studies have been conducted in miner cohorts as well as general populations to estimate the risks associated with high and low dose exposures. There are problems with extrapolating risk estimates to low dose exposures, mainly that the dose-response curve at low doses is not well understood. Calculated dosimetric quantities give average energy depositions in an organ or a whole body, but morphological features of an individual can affect these values. As opposed to human phantom models, Computed Tomography (CT) scans provide unique, patient-specific geometries that are valuable in modeling the radiological effects of the short-lived radon progeny sources. Monte Carlo particle transport code Geant4 was used with the CT scan data to model radon inhalation in the main bronchial bifurcation. The equivalent dose rates are near the lower bounds of estimates found in the literature, depending on source volume. To complement the macroscopic study, simulations were run in a small tissue volume in Geant4-DNA toolkit. As an expansion of Geant4 meant to simulate direct physical interactions at the cellular level, the particle track structure of the radon progeny alphas can be analyzed to estimate the damage that can occur in sensitive cellular structures like the DNA molecule. These estimates of DNA double strand breaks are lower than those found in Geant4-DNA studies. Further refinements of the microscopic model are at the cutting edge of nanodosimetry research.

Neutron-neutron angular correlations in spontaneous fission of 252Cf and 240Pu

NASA Astrophysics Data System (ADS)

Verbeke, J. M.; Nakae, L. F.; Vogt, R.

2018-04-01

Background: Angular anisotropy has been observed between prompt neutrons emitted during the fission process. Such an anisotropy arises because the emitted neutrons are boosted along the direction of the parent fragment. Purpose: To measure the neutron-neutron angular correlations from the spontaneous fission of 252Cf and 240Pu oxide samples using a liquid scintillator array capable of pulse-shape discrimination. To compare these correlations to simulations combining the Monte Carlo radiation transport code MCNPX with the fission event generator FREYA. Method: Two different analysis methods were used to study the neutron-neutron correlations with varying energy thresholds. The first is based on setting a light output threshold while the second imposes a time-of-flight cutoff. The second method has the advantage of being truly detector independent. Results: The neutron-neutron correlation modeled by FREYA depends strongly on the sharing of the excitation energy between the two fragments. The measured asymmetry enabled us to adjust the FREYA parameter x in 240Pu, which controls the energy partition between the fragments and is so far inaccessible in other measurements. The 240Pu data in this analysis was the first available to quantify the energy partition for this isotope. The agreement between data and simulation is overall very good for 252Cf(sf ) and 240Pu(sf ) . Conclusions: The asymmetry in the measured neutron-neutron angular distributions can be predicted by FREYA. The shape of the correlation function depends on how the excitation energy is partitioned between the two fission fragments. Experimental data suggest that the lighter fragment is disproportionately excited.

Fluence-to-absorbed-dose conversion coefficients for neutron beams from 0.001 eV to 100 GeV calculated for a set of pregnant female and fetus models

NASA Astrophysics Data System (ADS)

Taranenko, Valery; Xu, X. George

2008-03-01

Protection of fetuses against external neutron exposure is an important task. This paper reports a set of absorbed dose conversion coefficients for fetal and maternal organs for external neutron beams using the RPI-P pregnant female models and the MCNPX code. The newly developed pregnant female models represent an adult female with a fetus including its brain and skeleton at the end of each trimester. The organ masses were adjusted to match the reference values within 1%. For the 3 mm cubic voxel size, the models consist of 10-15 million voxels for 35 organs. External monoenergetic neutron beams of six standard configurations (AP, PA, LLAT, RLAT, ROT and ISO) and source energies 0.001 eV-100 GeV were considered. The results are compared with previous data that are based on simplified anatomical models. The differences in dose depend on source geometry, energy and gestation periods: from 20% up to 140% for the whole fetus, and up to 100% for the fetal brain. Anatomical differences are primarily responsible for the discrepancies in the organ doses. For the first time, the dependence of mother organ doses upon anatomical changes during pregnancy was studied. A maximum of 220% increase in dose was observed for the placenta in the nine months model compared to three months, whereas dose to the pancreas, small and large intestines decreases by 60% for the AP source for the same models. Tabulated dose conversion coefficients for the fetus and 27 maternal organs are provided.

Internal charging analysis tools, NUMIT 2.0 and 3D NUMIT, and those applications on Europa Clipper and Juno missions

NASA Astrophysics Data System (ADS)

Kim, W.; Chinn, J. Z.; Katz, I.; Jun, I.; Garrett, H. B.

2016-12-01

One of the major concerns in the spacecraft design due to natural space environment interaction is the internal charging in dielectric materials and floating conductors, especially for missions encountering a high radiation environment such as NASA's Juno and proposed Europa Clipper Missions. Sufficiently energetic electrons can penetrate the spacecraft structure or electronics chassis and stop within dielectrics and floating conductors. Electrons can accumulate in dielectrics over time due to the dielectrics' very low conductivity. If the electric field resulting from a charge buildup becomes higher than the breakdown threshold of the dielectric, discharge may occur, potentially damaging near-by sensitive electronics. Indeed, numerous spacecraft anomalies and failures have been attributed to this phenomenon, referred to as internal electrostatic discharge (iESD). Therefore, accurate assessment of the risk of iESD for a given space environment and dielectric geometry is important for spacecraft reliability. To evaluate the risk of iESD, we developed a general three dimensional internal charge analyses method, 3D NUMIT by combining a Monte Carlo radiation transport simulation tool such as MCNPX or GEANT4 and a commercial FEA software such as COMSOL. Also for a simple and fast internal charging assessment, we significantly improved the widely used one dimensional internal charging assessment code, NUMIT and named NUMIT 2.0. We will show the new features of NUMIT 2.0 and the capability of 3D NUMIT with several examples of applications of those tools to iESD assessments on Juno and Europa Clipper Missions.

Study on Characteristic of Temperature Coefficient of Reactivity for Plutonium Core of Pebbled Bed Reactor

NASA Astrophysics Data System (ADS)

Zuhair; Suwoto; Setiadipura, T.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

As a part of the solution searching for possibility to control the plutonium, a current effort is focused on mechanisms to maximize consumption of plutonium. Plutonium core solution is a unique case in the high temperature reactor which is intended to reduce the accumulation of plutonium. However, the safety performance of the plutonium core which tends to produce a positive temperature coefficient of reactivity should be examined. The pebble bed inherent safety features which are characterized by a negative temperature coefficient of reactivity must be maintained under any circumstances. The purpose of this study is to investigate the characteristic of temperature coefficient of reactivity for plutonium core of pebble bed reactor. A series of calculations with plutonium loading varied from 0.5 g to 1.5 g per fuel pebble were performed by the MCNPX code and ENDF/B-VII library. The calculation results show that the k eff curve of 0.5 g Pu/pebble declines sharply with the increase in fuel burnup while the greater Pu loading per pebble yields k eff curve declines slighter. The fuel with high Pu content per pebble may reach long burnup cycle. From the temperature coefficient point of view, it is concluded that the reactor containing 0.5 g-1.25 g Pu/pebble at high burnup has less favorable safety features if it is operated at high temperature. The use of fuel with Pu content of 1.5 g/pebble at high burnup should be considered carefully from core safety aspect because it could affect transient behavior into a fatal accident situation.

Determination of the Secondary Neutron Flux at the Massive Natural Uranium Spallation Target

NASA Astrophysics Data System (ADS)

Zeman, M.; Adam, J.; Baldin, A. A.; Furman, W. I.; Gustov, S. A.; Katovsky, K.; Khushvaktov, J.; Mar`in, I. I.; Novotny, F.; Solnyshkin, A. A.; Tichy, P.; Tsoupko-Sitnikov, V. M.; Tyutyunnikov, S. I.; Vespalec, R.; Vrzalova, J.; Wagner, V.; Zavorka, L.

The flux of secondary neutrons generated in collisions of the 660 MeV proton beam with the massive natural uranium spallation target was investigated using a set of monoisotopic threshold activation detectors. Sandwiches made of thin high-purity Al, Co, Au, and Bi metal foils were installed in different positions across the whole spallation target. The gamma-ray activity of products of (n,xn) and other studied reactions was measured offline with germanium semiconductor detectors. Reaction yields of radionuclides with half-life exceeding 100 min and with effective neutron energy thresholds between 3.6 MeV and 186 MeV provided us with information about the spectrum of spallation neutrons in this energy region and beyond. The experimental neutron flux was determined using the measured reaction yields and cross-sections calculated with the TALYS 1.8 nuclear reaction program and INCL4-ABLA event generator of MCNP6. Neutron spectra in the region of activation sandwiches were also modeled with the radiation transport code MCNPX 2.7. Neutron flux based on excitation functions from TALYS provides a reasonable description of the neutron spectrum inside the spallation target and is in good agreement with Monte-Carlo predictions. The experimental flux that uses INCL4 cross-sections rather underestimates the modeled spectrum in the whole region of interest, but the agreement within few standard deviations was reached as well. The paper summarizes basic principles of the method for determining the spectrum of high-energy neutrons without employing the spectral adjustment routines and points out to the need for model improvements and precise cross-section measurements.

Feasibility study of 235U and 239Pu characterization in radioactive waste drums using neutron-induced fission delayed gamma rays

NASA Astrophysics Data System (ADS)

Nicol, T.; Pérot, B.; Carasco, C.; Brackx, E.; Mariani, A.; Passard, C.; Mauerhofer, E.; Collot, J.

2016-10-01

This paper reports a feasibility study of 235U and 239Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of 235U and 239Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to 137Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of 235U or 239Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.

Pu-239 organ specific dosimetric model applied to non-human biota

NASA Astrophysics Data System (ADS)

Kaspar, Matthew Jason

There are few locations throughout the world, like the Maralinga nuclear test site located in south western Australia, where sufficient plutonium contaminate concentration levels exist that they can be utilized for studies of the long-term radionuclide accumulation in non-human biota. The information obtained will be useful for the potential human users of the site while also keeping with international efforts to better understand doses to non-human biota. In particular, this study focuses primarily on a rabbit sample set collected from the population located within the site. Our approach is intended to employ the same dose and dose rate methods selected by the International Commission on Radiological Protection and adapted by the scientific community for similar research questions. These models rely on a series of simplifying assumptions on biota and their geometry; in particular; organisms are treated as spherical and ellipsoidal representations displaying the animal mass and volume. These simplifications assume homogeneity of all animal tissues. In collaborative efforts between Colorado State University and the Australian Nuclear Science and Technology Organisation (ANSTO), we are expanding current knowledge on radionuclide accumulation in specific organs causing organ-specific dose rates, such as Pu-239 accumulating in bone, liver, and lungs. Organ-specific dose models have been developed for humans; however, little has been developed for the dose assessment to biota, in particular rabbits. This study will determine if it is scientifically valid to use standard software, in particular ERICA Tool, as a means to determine organ-specific dosimetry due to Pu-239 accumulation in organs. ERICA Tool is normally applied to whole organisms as a means to determine radiological risk to whole ecosystems. We will focus on the aquatic model within ERICA Tool, as animal organs, like aquatic organisms, can be assumed to lie within an infinite uniform medium. This model would scientifically be valid for radionuclides emitting short-range radiation, as with Pu-239, where the energy is deposited locally. Two MCNPX models have been created and evaluated against ERICA Tool's aquatic model. One MCNPX model replicates ERICA Tool's intrinsic assumptions while the other uses a more realistic animal model adopted by ICRP Publication 108 and ERICA Tool for the organs "infinite" surrounding universe. In addition, the role of model geometry will be analyzed by focusing on four geometry sets for the same organ, including a spherical geometry. ERICA Tool will be compared to MCNPX results within and between each organ geometry set. In addition, the organ absorbed dose rate will be calculated for six rabbits located on the Maralinga nuclear test site as a preliminary test for further investigation. Data in all cases will be compared using percent differences and Student's t-test with respect to ERICA Tool's results and the overall average organ mean absorbed dose rate.

Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Botta, F.; Mairani, A.; Battistoni, G.

Purpose: The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, fluka Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, fluka has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernelmore » (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. Methods: fluka DPKs have been calculated in both water and compact bone for monoenergetic electrons (10{sup -3} MeV) and for beta emitting isotopes commonly used for therapy ({sup 89}Sr, {sup 90}Y, {sup 131}I, {sup 153}Sm, {sup 177}Lu, {sup 186}Re, and {sup 188}Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. fluka outcomes have been compared to penelope v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (etran, geant4, mcnpx) has been done. Maximum percentage differences within 0.8{center_dot}R{sub CSDA} and 0.9{center_dot}R{sub CSDA} for monoenergetic electrons (R{sub CSDA} being the continuous slowing down approximation range) and within 0.8{center_dot}X{sub 90} and 0.9{center_dot}X{sub 90} for isotopes (X{sub 90} being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9{center_dot}R{sub CSDA} and 0.9{center_dot}X{sub 90} for electrons and isotopes, respectively. Results: Concerning monoenergetic electrons, within 0.8{center_dot}R{sub CSDA} (where 90%-97% of the particle energy is deposed), fluka and penelope agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The discrepancies between fluka and the other codes are of the same order of magnitude than those observed when comparing the other codes among them, which can be referred to the different simulation algorithms. When considering the beta spectra, discrepancies notably reduce: within 0.9{center_dot}X{sub 90}, fluka and penelope differ for less than 1% in water and less than 2% in bone with any of the isotopes here considered. Complete data of fluka DPKs are given as Supplementary Material as a tool to perform dosimetry by analytical point kernel convolution. Conclusions: fluka provides reliable results when transporting electrons in the low energy range, proving to be an adequate tool for nuclear medicine dosimetry.« less

Terrestrial Carbon Fluxes from Deforestation in the Brazilian Amazon and Cerrado Regions Predicted from MODIS Satellite Data and Ecosystem Modeling

NASA Astrophysics Data System (ADS)

Klooster, S.; Potter, C.; Genovese, V.

2008-12-01

The NASA-CASA (Carnegie Ames Stanford Approach) simulation model based on satellite observations of monthly vegetation cover from the Moderate Resolution Imaging Spectroradiometer (MODIS) was used to estimate tropical forest and savanna (Cerrado) carbon pools for the Brazilian Amazon region over the period 2000-2004. Adjustments for mean age of forest stands were carried out across the region, resulting in a new mapping of aboveground biomass pools based on MODIS satellite data. Yearly maps of newly deforested lands from the Brazilian PRODES (Programa de calculo do desflorestamento da Amazonia ) project were combined with these NASA-CASA biomass predictions to generate seasonal budgets of potential carbon and nitrogen trace gas losses from biomass burning events. Simulations of plant residue and soil carbon decomposition were conducted in the NASA-CASA model during and following deforestation events to track the fate of aboveground biomass pools that were cut and burned each year across the region.

Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station

DOE PAGES

Gallmeier, F. X.; Lu, W.; Riemer, B. W.; ...

2016-06-14

We identified candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared tomore » the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm 2 to 20 × 20 mm 2. Furthermore, this increase in brightness has the potential to translate to an increase of beam intensity at the instruments’ sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. Our first effort decoupled group moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.« less

Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station

NASA Astrophysics Data System (ADS)

Gallmeier, F. X.; Lu, W.; Riemer, B. W.; Zhao, J. K.; Herwig, K. W.; Robertson, J. L.

2016-06-01

Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm2 to 20 × 20 mm2. This increase in brightness has the potential to translate to an increase of beam intensity at the instruments' sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.

Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallmeier, F. X.; Lu, W.; Riemer, B. W.

We identified candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared tomore » the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm 2 to 20 × 20 mm 2. Furthermore, this increase in brightness has the potential to translate to an increase of beam intensity at the instruments’ sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. Our first effort decoupled group moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.« less

Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gallmeier, F. X., E-mail: gallmeierfz@ornl.gov; Lu, W.; Riemer, B. W.

Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis comparedmore » to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm{sup 2} to 20 × 20 mm{sup 2}. This increase in brightness has the potential to translate to an increase of beam intensity at the instruments’ sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.« less

A comparison study on various low energy sources in interstitial prostate brachytherapy

PubMed Central

Bakhshabadi, Mahdi; Ghorbani, Mahdi; Knaup, Courtney; Meigooni, Ali S.

2016-01-01

Purpose Low energy sources are routinely used in prostate brachytherapy. 125I is one of the most commonly used sources. Low energy 131Cs source was introduced recently as a brachytherapy source. The aim of this study is to compare dose distributions of 125I, 103Pd, and 131Cs sources in interstitial brachytherapy of prostate. Material and methods ProstaSeed 125I brachytherapy source was simulated using MCNPX Monte Carlo code. Additionally, two hypothetical sources of 103Pd and 131Cs were simulated with the same geometry as the ProstaSeed 125I source, while having their specific emitted gamma spectra. These brachytherapy sources were simulated with distribution of forty-eight seeds in a phantom including prostate. The prostate was considered as a sphere with radius of 1.5 cm. Absolute and relative dose rates were obtained in various distances from the source along the transverse and longitudinal axes inside and outside the tumor. Furthermore, isodose curves were plotted around the sources. Results Analyzing the initial dose profiles for various sources indicated that with the same time duration and air kerma strength, 131Cs delivers higher dose to tumor. However, relative dose rate inside the tumor is higher and outside the tumor is lower for the 103Pd source. Conclusions The higher initial absolute dose in cGy/(h.U) of 131Cs brachytherapy source is an advantage of this source over the others. The higher relative dose inside the tumor and lower relative dose outside the tumor for the 103Pd source are advantages of this later brachytherapy source. Based on the total dose the 125I source has advantage over the others due to its longer half-life. PMID:26985200

A comparison study on various low energy sources in interstitial prostate brachytherapy.

PubMed

Bakhshabadi, Mahdi; Ghorbani, Mahdi; Khosroabadi, Mohsen; Knaup, Courtney; Meigooni, Ali S

2016-02-01

Low energy sources are routinely used in prostate brachytherapy. (125)I is one of the most commonly used sources. Low energy (131)Cs source was introduced recently as a brachytherapy source. The aim of this study is to compare dose distributions of (125)I, (103)Pd, and (131)Cs sources in interstitial brachytherapy of prostate. ProstaSeed (125)I brachytherapy source was simulated using MCNPX Monte Carlo code. Additionally, two hypothetical sources of (103)Pd and (131)Cs were simulated with the same geometry as the ProstaSeed (125)I source, while having their specific emitted gamma spectra. These brachytherapy sources were simulated with distribution of forty-eight seeds in a phantom including prostate. The prostate was considered as a sphere with radius of 1.5 cm. Absolute and relative dose rates were obtained in various distances from the source along the transverse and longitudinal axes inside and outside the tumor. Furthermore, isodose curves were plotted around the sources. Analyzing the initial dose profiles for various sources indicated that with the same time duration and air kerma strength, (131)Cs delivers higher dose to tumor. However, relative dose rate inside the tumor is higher and outside the tumor is lower for the (103)Pd source. The higher initial absolute dose in cGy/(h.U) of (131)Cs brachytherapy source is an advantage of this source over the others. The higher relative dose inside the tumor and lower relative dose outside the tumor for the (103)Pd source are advantages of this later brachytherapy source. Based on the total dose the (125)I source has advantage over the others due to its longer half-life.

Influence of eye size and beam entry angle on dose to non-targeted tissues of the eye during stereotactic x-ray radiosurgery of AMD

NASA Astrophysics Data System (ADS)

Cantley, Justin L.; Hanlon, Justin; Chell, Erik; Lee, Choonsik; Smith, W. Clay; Bolch, Wesley E.

2013-10-01

Age-related macular degeneration is a leading cause of vision loss for the elderly population of industrialized nations. The IRay® Radiotherapy System, developed by Oraya® Therapeutics, Inc., is a stereotactic low-voltage irradiation system designed to treat the wet form of the disease. The IRay System uses three robotically positioned 100 kVp collimated photon beams to deliver an absorbed dose of up to 24 Gy to the macula. The present study uses the Monte Carlo radiation transport code MCNPX to assess absorbed dose to six non-targeted tissues within the eye—total lens, radiosensitive tissues of the lens, optic nerve, distal tip of the central retinal artery, non-targeted portion of the retina, and the ciliary body--all as a function of eye size and beam entry angle. The ocular axial length was ranged from 20 to 28 mm in 2 mm increments, with the polar entry angle of the delivery system varied from 18° to 34° in 2° increments. The resulting data showed insignificant variations in dose for all eye sizes. Slight variations in the dose to the optic nerve and the distal tip of the central retinal artery were noted as the polar beam angle changed. An increase in non-targeted retinal dose was noted as the entry angle increased, while the dose to the lens, sensitive volume of the lens, and ciliary body decreased as the treatment polar angle increased. Polar angles of 26° or greater resulted in no portion of the sensitive volume of the lens receiving an absorbed dose of 0.5 Gy or greater. All doses to non-targeted structures reported in this study were less than accepted thresholds for post-procedure complications.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Selvi, Marco

For all experiments dealing with the rare event searches (neutrino, dark matter, neutrino-less double-beta decay), the reduction of the radioactive background is one of the most important and difficult tasks. There are basically two types of background, electron recoils and nuclear recoils. The electron recoil background is mostly from the gamma rays through the radioactive decay. The nuclear recoil background is from neutrons from spontaneous fission, (α, n) reactions and muoninduced interactions (spallations, photo-nuclear and hadronic interaction). The external gammas and neutrons from the muons and laboratory environment, can be reduced by operating the detector at deep underground laboratories andmore » by placing active or passive shield materials around the detector. The radioactivity of the detector materials also contributes to the background; in order to reduce it a careful screening campaign is mandatory to select highly radio-pure materials. In this review I present the status of current Monte Carlo simulations aimed to estimate and reproduce the background induced by gamma and neutron radioactivity of the materials and the shield of rare event search experiment. For the electromagnetic background a good level of agreement between the data and the MC simulation has been reached by the XENON100 and EDELWEISS experiments, using the GEANT4 toolkit. For the neutron background, a comparison between the yield of neutrons from spontaneous fission and (α, n) obtained with two dedicated softwares, SOURCES-4A and the one developed by Mei-Zhang-Hime, show a good overall agreement, with total yields within a factor 2 difference. The energy spectra from SOURCES-4A are in general smoother, while those from MZH presents sharp peaks. The neutron propagation through various materials has been studied with two MC codes, GEANT4 and MCNPX, showing a reasonably good agreement, inside 50% discrepancy.« less

Monte Carlo simulations used to calculate the energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy.

PubMed

Hocine, Nora; Meignan, Michel; Masset, Hélène

2018-04-01

To better understand the risks of cumulative medical X-ray investigations and the possible causal role of contrast agent on the coronary artery wall, the correlation between iodinated contrast media and the increase of energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy is investigated. The calculations of energy deposition have been performed using Monte Carlo (MC) simulation codes, namely PENetration and Energy LOss of Positrons and Electrons (PENELOPE) and Monte Carlo N-Particle eXtended (MCNPX). Exposure of a cylinder phantom, artery and a metal stent (AISI 316L) to several X-ray photon beams were simulated. For the energies used in cardiac imaging the energy deposited in the coronary artery lumen increases with the quantity of iodine. Monte Carlo calculations indicate a strong dependence of the energy enhancement factor (EEF) on photon energy and iodine concentration. The maximum value of EEF is equal to 25; this factor is showed for 83 keV and for 400 mg Iodine/mL. No significant impact of the stent is observed on the absorbed dose in the artery for incident X-ray beams with mean energies of 44, 48, 52 and 55 keV. A strong correlation was shown between the increase in the concentration of iodine and the energy deposited in the coronary artery lumen for the energies used in cardiac imaging and over the energy range between 44 and 55 keV. The data provided by this study could be useful for creating new medical imaging protocols to obtain better diagnostic information with a lower level of radiation exposure.

The capability of radial basis function to forecast the volume fractions of the annular three-phase flow of gas-oil-water.

PubMed

Roshani, G H; Karami, A; Salehizadeh, A; Nazemi, E

2017-11-01

The problem of how to precisely measure the volume fractions of oil-gas-water mixtures in a pipeline remains as one of the main challenges in the petroleum industry. This paper reports the capability of Radial Basis Function (RBF) in forecasting the volume fractions in a gas-oil-water multiphase system. Indeed, in the present research, the volume fractions in the annular three-phase flow are measured based on a dual energy metering system including the 152 Eu and 137 Cs and one NaI detector, and then modeled by a RBF model. Since the summation of volume fractions are constant (equal to 100%), therefore it is enough for the RBF model to forecast only two volume fractions. In this investigation, three RBF models are employed. The first model is used to forecast the oil and water volume fractions. The next one is utilized to forecast the water and gas volume fractions, and the last one to forecast the gas and oil volume fractions. In the next stage, the numerical data obtained from MCNP-X code must be introduced to the RBF models. Then, the average errors of these three models are calculated and compared. The model which has the least error is picked up as the best predictive model. Based on the results, the best RBF model, forecasts the oil and water volume fractions with the mean relative error of less than 0.5%, which indicates that the RBF model introduced in this study ensures an effective enough mechanism to forecast the results. Copyright © 2017 Elsevier Ltd. All rights reserved.

Technical Note: A proposal of air ventilation system design criteria for a clinical room in a heavy-ion medical facility.

PubMed

Kum, Oyeon

2018-06-01

An optimized air ventilation system design for a treatment room in Heavy-ion Medical Facility is an important issue in the aspects of nuclear safety because the activated air produced in a treatment room can directly affect the medical staff and the general public in the radiation-free area. Optimized design criteria of air ventilation system for a clinical room in 430 MeV/u carbon ion beam medical accelerator facility was performed by using a combination of MCNPX2.7.0 and CINDER'90 codes. Effective dose rate and its accumulated effective dose by inhalation and residual gamma were calculated for a normal treatment scenario (2 min irradiation for one fraction) as a function of decay time. Natural doses around the site were measured before construction and used as reference data. With no air ventilation system, the maximum effective dose rate was about 3 μSv/h (total dose of 90 mSv/y) and minimum 0.2 μSv/h (total dose of 6 mSv/y), which are over the legal limits for medical staff and for the general public. Although inhalation dose contribution was relatively small, it was considered seriously because of its long-lasting effects in the body. The integrated dose per year was 1.8 mSv/y in the radiation-free area with the 20-min rate of air ventilation system. An optimal air ventilation rate of 20 min is proposed for a clinical room, which also agrees with the best mechanical design value. © 2018 American Association of Physicists in Medicine.

Evaluation of equivalent and effective dose by KAP for patient and orthopedic surgeon in vertebral compression fracture surgery

NASA Astrophysics Data System (ADS)

Santos, Felipe A.; Galeano, Diego C.; Santos, William S.; Silva, Ademir X.; Souza, Susana O.; Carvalho Júnior, Albérico B.

2017-03-01

Clinical scenarios were virtually modeled to estimate both the equivalent and effective doses normalized by KAP (Kerma Area Product) to vertebra compression fracture surgery in patient and surgeon. This surgery is known as kyphoplasty and involves the use of X-ray equipment, the C-arm, which provides real-time images to assist the surgeon in conducting instruments inserted into the patient and in the delivery of surgical cement into the fractured vertebra. The radiation transport code used was MCNPX (Monte Carlo N-Particle eXtended) and a pair of UFHADM (University of Florida Hybrid ADult Male) virtual phantoms. The developed scenarios allowed us to calculate a set of equivalent dose (HT) and effective dose (E) for patients and surgeons. In additional, the same scenario was calculated KAP in the tube output and was used for calculating conversion coefficients (E/KAP and HT/KAP). From the knowledge of the experimental values of KAP and the results presented in this study, it is possible to estimate absolute values of effective doses for different exposure conditions. In this work, we developed scenarios with and without the surgical table with the purpose of comparison with the existing data in the literature. The absence of the bed in the scenario promoted a percentage absolute difference of 56% in the patient effective doses in relation to scenarios calculated with a bed. Regarding the surgeon, the use of the personal protective equipment (PPE) reduces between 75% and 79% the effective dose and the use of the under table shield (UTS) reduces the effective dose of between 3% and 7%. All these variations emphasize the importance of the elaboration of virtual scenarios that approach the actual clinical conditions generating E/KAP and HT/KAP closer to the actual values.

Review of Monte Carlo simulations for backgrounds from radioactivity

NASA Astrophysics Data System (ADS)

Selvi, Marco

2013-08-01

For all experiments dealing with the rare event searches (neutrino, dark matter, neutrino-less double-beta decay), the reduction of the radioactive background is one of the most important and difficult tasks. There are basically two types of background, electron recoils and nuclear recoils. The electron recoil background is mostly from the gamma rays through the radioactive decay. The nuclear recoil background is from neutrons from spontaneous fission, (α, n) reactions and muoninduced interactions (spallations, photo-nuclear and hadronic interaction). The external gammas and neutrons from the muons and laboratory environment, can be reduced by operating the detector at deep underground laboratories and by placing active or passive shield materials around the detector. The radioactivity of the detector materials also contributes to the background; in order to reduce it a careful screening campaign is mandatory to select highly radio-pure materials. In this review I present the status of current Monte Carlo simulations aimed to estimate and reproduce the background induced by gamma and neutron radioactivity of the materials and the shield of rare event search experiment. For the electromagnetic background a good level of agreement between the data and the MC simulation has been reached by the XENON100 and EDELWEISS experiments, using the GEANT4 toolkit. For the neutron background, a comparison between the yield of neutrons from spontaneous fission and (α, n) obtained with two dedicated softwares, SOURCES-4A and the one developed by Mei-Zhang-Hime, show a good overall agreement, with total yields within a factor 2 difference. The energy spectra from SOURCES-4A are in general smoother, while those from MZH presents sharp peaks. The neutron propagation through various materials has been studied with two MC codes, GEANT4 and MCNPX, showing a reasonably good agreement, inside 50% discrepancy.

The effect of energy spectrum change on DNA damage in and out of field in 10-MV clinical photon beams.

PubMed

Ezzati, A O; Xiao, Y; Sohrabpour, M; Studenski, M T

2015-01-01

The aim of this study was to quantify the DNA damage induced in a clinical megavoltage photon beam at various depths in and out of the field. MCNPX was used to simulate 10 × 10 and 20 × 20 cm(2) 10-MV photon beams from a clinical linear accelerator. Photon and electron spectra were collected in a water phantom at depths of 2.5, 12.5 and 22.5 cm on the central axis and at off-axis points out to 10 cm. These spectra were used as an input to a validated microdosimetric Monte Carlo code, MCDS, to calculate the RBE of induced DSB in DNA at points in and out of the primary radiation field at three depths. There was an observable difference in the energy spectra for photons and electrons for points in the primary radiation field and those points out of field. In the out-of-field region, the mean energy for the photon and electron spectra decreased by a factor of about six and three from the in-field mean energy, respectively. Despite the differences in spectra and mean energy, the change in RBE was <1 % from the in-field region to the out-of-field region at any depth. There was no significant change in RBE regardless of the location in the phantom. Although there are differences in both the photon and electron spectra, these changes do not correlate with a change in RBE in a clinical MV photon beam as the electron spectra are dominated by electrons with energies >20 keV.

Comparison of the Light Charged Particles on Scatter Radiation Dose in Thyroid Hadron Therapy

PubMed Central

Azizi, M; Mowlavi, AA

2014-01-01

Background: Hadron therapy is a novel technique of cancer radiation therapy which employs charged particles beams, 1H and light ions in particular. Due to their physical and radiobiological properties, they allow one to obtain a more conformal treatment, sparing better the healthy tissues located in proximity of the tumor and allowing a higher control of the disease. Objective: As it is well known, these light particles can interact with nuclei in the tissue, and produce the different secondary particles such as neutron and photon. These particles can damage specially the critical organs behind of thyroid gland. Methods: In this research, we simulated neck geometry by MCNPX code and calculated the light particles dose at distance of 2.14 cm in thyroid gland, for different particles beam: 1H, 2H, 3He, and 4He. Thyroid treatment is important because the spine and vertebrae is situated right behind to the thyroid gland on the posterior side. Results: The results show that 2H has the most total flux for photon and neutron, 1.944E-3 and 1.7666E-2, respectively. Whereas 1H and 3He have best conditions, 8.88609E-4 and 1.35431E-3 for photon, 4.90506E-4 and 4.34057E-3 for neutron, respectively. The same calculation has obtained for energy depositions for these particles. Conclusion: In this research, we investigated that which of these light particles can deliver the maximum dose to the normal tissues and the minimum dose to the tumor. By comparing these results for the mentioned light particles, we find out 1H and 3He is the best therapy choices for thyroid glands whereas 2H is the worst. PMID:25505774

External dose-rate conversion factors of radionuclides for air submersion, ground surface contamination and water immersion based on the new ICRP dosimetric setting.

PubMed

Yoo, Song Jae; Jang, Han-Ki; Lee, Jai-Ki; Noh, Siwan; Cho, Gyuseong

2013-01-01

For the assessment of external doses due to contaminated environment, the dose-rate conversion factors (DCFs) prescribed in Federal Guidance Report 12 (FGR 12) and FGR 13 have been widely used. Recently, there were significant changes in dosimetric models and parameters, which include the use of the Reference Male and Female Phantoms and the revised tissue weighting factors, as well as the updated decay data of radionuclides. In this study, the DCFs for effective and equivalent doses were calculated for three exposure settings: skyshine, groundshine and water immersion. Doses to the Reference Phantoms were calculated by Monte Carlo simulations with the MCNPX 2.7.0 radiation transport code for 26 mono-energy photons between 0.01 and 10 MeV. The transport calculations were performed for the source volume within the cut-off distances practically contributing to the dose rates, which were determined by a simplified calculation model. For small tissues for which the reduction of variances are difficult, the equivalent dose ratios to a larger tissue (with lower statistical errors) nearby were employed to make the calculation efficient. Empirical response functions relating photon energies, and the organ equivalent doses or the effective doses were then derived by the use of cubic-spline fitting of the resulting doses for 26 energy points. The DCFs for all radionuclides considered important were evaluated by combining the photon emission data of the radionuclide and the empirical response functions. Finally, contributions of accompanied beta particles to the skin equivalent doses and the effective doses were calculated separately and added to the DCFs. For radionuclides considered in this study, the new DCFs for the three exposure settings were within ±10 % when compared with DCFs in FGR 13.

A feasibility study to calculate unshielded fetal doses to pregnant patients in 6-MV photon treatments using Monte Carlo methods and anatomically realistic phantoms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bednarz, Bryan; Xu, X. George

2008-07-15

A Monte Carlo-based procedure to assess fetal doses from 6-MV external photon beam radiation treatments has been developed to improve upon existing techniques that are based on AAPM Task Group Report 36 published in 1995 [M. Stovall et al., Med. Phys. 22, 63-82 (1995)]. Anatomically realistic models of the pregnant patient representing 3-, 6-, and 9-month gestational stages were implemented into the MCNPX code together with a detailed accelerator model that is capable of simulating scattered and leakage radiation from the accelerator head. Absorbed doses to the fetus were calculated for six different treatment plans for sites above the fetusmore » and one treatment plan for fibrosarcoma in the knee. For treatment plans above the fetus, the fetal doses tended to increase with increasing stage of gestation. This was due to the decrease in distance between the fetal body and field edge with increasing stage of gestation. For the treatment field below the fetus, the absorbed doses tended to decrease with increasing gestational stage of the pregnant patient, due to the increasing size of the fetus and relative constant distance between the field edge and fetal body for each stage. The absorbed doses to the fetus for all treatment plans ranged from a maximum of 30.9 cGy to the 9-month fetus to 1.53 cGy to the 3-month fetus. The study demonstrates the feasibility to accurately determine the absorbed organ doses in the mother and fetus as part of the treatment planning and eventually in risk management.« less

The use of MCNP and gamma spectrometry in supporting the evaluation of NORM in Libyan oil pipeline scale

NASA Astrophysics Data System (ADS)

Habib, Ahmed S.; Bradley, D. A.; Regan, P. H.; Shutt, A. L.

2010-07-01

The accumulation of scales in production pipes is a common problem in the oil industry, reducing fluid flow and also leading to costly remedies and disposal issues. Typical materials found in such scale are sulphates and carbonates of calcium and barium, or iron sulphide. Radium arising from the uranium/thorium present in oil-bearing rock formations may replace the barium or calcium in these salts to form radium salts. This creates what is known as technologically enhanced naturally occurring radioactive material (TENORM or simply NORM). NORM is a serious environmental and health and safety issue arising from commercial oil and gas extraction operations. Whilst a good deal has been published on the characterisation and measurement of radioactive scales from offshore oil production, little information has been published regarding NORM associated with land-based facilities such as that of the Libyan oil industry. The ongoing investigation described in this paper concerns an assessment of NORM from a number of land based Libyan oil fields. A total of 27 pipe scale samples were collected from eight oil fields, from different locations in Libya. The dose rates, measured using a handheld survey meter positioned on sample surfaces, ranged from 0.1-27.3 μSv h -1. In the initial evaluations of the sample activity, use is being made of a portable HPGe based spectrometry system. To comply with the prevailing safety regulations of the University of Surrey, the samples are being counted in their original form, creating a need for correction of non-homogeneous sample geometries. To derive a detection efficiency based on the actual sample geometries, a technique has been developed using a Monte Carlo particle transport code (MCNPX). A preliminary activity determination has been performed using an HPGe portable detector system.

Effect of tissue composition on dose distribution in brachytherapy with various photon emitting sources

PubMed Central

Ghorbani, Mahdi; Salahshour, Fateme; Haghparast, Abbas; Knaup, Courtney

2014-01-01

Purpose The aim of this study is to compare the dose in various soft tissues in brachytherapy with photon emitting sources. Material and methods 103Pd, 125I, 169Yb, 192Ir brachytherapy sources were simulated with MCNPX Monte Carlo code, and their dose rate constant and radial dose function were compared with the published data. A spherical phantom with 50 cm radius was simulated and the dose at various radial distances in adipose tissue, breast tissue, 4-component soft tissue, brain (grey/white matter), muscle (skeletal), lung tissue, blood (whole), 9-component soft tissue, and water were calculated. The absolute dose and relative dose difference with respect to 9-component soft tissue was obtained for various materials, sources, and distances. Results There was good agreement between the dosimetric parameters of the sources and the published data. Adipose tissue, breast tissue, 4-component soft tissue, and water showed the greatest difference in dose relative to the dose to the 9-component soft tissue. The other soft tissues showed lower dose differences. The dose difference was also higher for 103Pd source than for 125I, 169Yb, and 192Ir sources. Furthermore, greater distances from the source had higher relative dose differences and the effect can be justified due to the change in photon spectrum (softening or hardening) as photons traverse the phantom material. Conclusions The ignorance of soft tissue characteristics (density, composition, etc.) by treatment planning systems incorporates a significant error in dose delivery to the patient in brachytherapy with photon sources. The error depends on the type of soft tissue, brachytherapy source, as well as the distance from the source. PMID:24790623

Application of a color scanner for 60Co high dose rate brachytherapy dosimetry with EBT radiochromic film

PubMed Central

Ghorbani, Mahdi; Toossi, Mohammad Taghi Bahreyni; Mowlavi, Ali Asghar; Roodi, Shahram Bayani; Meigooni, Ali Soleimani

2012-01-01

Background. The aim of this study is to evaluate the performance of a color scanner as a radiochromic film reader in two dimensional dosimetry around a high dose rate brachytherapy source. Materials and methods A Microtek ScanMaker 1000XL film scanner was utilized for the measurement of dose distribution around a high dose rate GZP6 60Co brachytherapy source with GafChromic® EBT radiochromic films. In these investigations, the non-uniformity of the film and scanner response, combined, as well as the films sensitivity to scanner’s light source was evaluated using multiple samples of films, prior to the source dosimetry. The results of these measurements were compared with the Monte Carlo simulated data using MCNPX code. In addition, isodose curves acquired by radiochromic films and Monte Carlo simulation were compared with those provided by the GZP6 treatment planning system. Results Scanning of samples of uniformly irradiated films demonstrated approximately 2.85% and 4.97% nonuniformity of the response, respectively in the longitudinal and transverse directions of the film. Our findings have also indicated that the film response is not affected by the exposure to the scanner’s light source, particularly in multiple scanning of film. The results of radiochromic film measurements are in good agreement with the Monte Carlo calculations (4%) and the corresponding dose values presented by the GZP6 treatment planning system (5%). Conclusions The results of these investigations indicate that the Microtek ScanMaker 1000XL color scanner in conjunction with GafChromic EBT film is a reliable system for dosimetric evaluation of a high dose rate brachytherapy source. PMID:23411947

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Increased dose near the skin due to electromagnetic surface beacon transponder.

PubMed

Ahn, Kang-Hyun; Manger, Ryan; Halpern, Howard J; Aydogan, Bulent

2015-05-08

The purpose of this study was to evaluate the increased dose near the skin from an electromagnetic surface beacon transponder, which is used for localization and tracking organ motion. The bolus effect due to the copper coil surface beacon was evaluated with radiographic film measurements and Monte Carlo simulations. Various beam incidence angles were evaluated for both 6 MV and 18 MV experimentally. We performed simulations using a general-purpose Monte Carlo code MCNPX (Monte Carlo N-Particle) to supplement the experimental data. We modeled the surface beacon geometry using the actual mass of the glass vial and copper coil placed in its L-shaped polyethylene terephthalate tubing casing. Film dosimetry measured factors of 2.2 and 3.0 enhancement in the surface dose for normally incident 6 MV and 18 MV beams, respectively. Although surface dose further increased with incidence angle, the relative contribution from the bolus effect was reduced at the oblique incidence. The enhancement factors were 1.5 and 1.8 for 6 MV and 18 MV, respectively, at an incidence angle of 60°. Monte Carlo simulation confirmed the experimental results and indicated that the epidermal skin dose can reach approximately 50% of the dose at dmax at normal incidence. The overall effect could be acceptable considering the skin dose enhancement is confined to a small area (~ 1 cm2), and can be further reduced by using an opposite beam technique. Further clinical studies are justified in order to study the dosimetric benefit versus possible cosmetic effects of the surface beacon. One such clinical situation would be intact breast radiation therapy, especially large-breasted women.

Estimate of S-values for children due to six positron emitting radionuclides used in PET examinations

NASA Astrophysics Data System (ADS)

Belinato, Walmir; Santos, William S.; Perini, Ana P.; Neves, Lucio P.; Caldas, Linda V. E.; Souza, Divanizia N.

2017-11-01

Positron emission tomography (PET) has revolutionized the diagnosis of cancer since its conception. When combined with computed tomography (CT), PET/CT performed in children produces highly accurate diagnoses from images of regions affected by malignant tumors. Considering the high risk to children when exposed to ionizing radiation, a dosimetric study for PET/CT procedures is necessary. Specific absorbed fractions (SAF) were determined for monoenergetic photons and positrons, as well as the S-values for six positron emitting radionuclides (11C, 13N, 18F, 68Ga, 82Rb, 15O), and 22 source organs. The study was performed for six pediatric anthropomorphic hybrid models, including the newborn and 1 year hermaphrodite, 5 and 10-year-old male and female, using the Monte Carlo N-Particle eXtended code (MCNPX, version 2.7.0). The results of the SAF in source organs and S-values for all organs showed to be inversely related to the age of the phantoms, which includes the variation of body weight. The results also showed that radionuclides with higher energy peak emission produces larger auto absorbed S-values due to local dose deposition by positron decay. The S-values for the source organs are considerably larger due to the interaction of tissue with non-penetrating particles (electrons and positrons) and present a linear relationship with the phantom body masses. The results of the S-values determined for positron-emitting radionuclides can be used to assess the radiation dose delivered to pediatric patients subjected to PET examination in clinical settings. The novelty of this work is associated with the determination of auto absorbed S-values, in six new pediatric virtual anthropomorphic phantoms, for six emitting positrons, commonly employed in PET exams.

Construction of Chinese adult male phantom library and its application in the virtual calibration of in vivo measurement.

PubMed

Chen, Yizheng; Qiu, Rui; Li, Chunyan; Wu, Zhen; Li, Junli

2016-03-07

In vivo measurement is a main method of internal contamination evaluation, particularly for large numbers of people after a nuclear accident. Before the practical application, it is necessary to obtain the counting efficiency of the detector by calibration. The virtual calibration based on Monte Carlo simulation usually uses the reference human computational phantom, and the morphological difference between the monitored personnel with the calibrated phantom may lead to the deviation of the counting efficiency. Therefore, a phantom library containing a wide range of heights and total body masses is needed. In this study, a Chinese reference adult male polygon surface (CRAM_S) phantom was constructed based on the CRAM voxel phantom, with the organ models adjusted to match the Chinese reference data. CRAM_S phantom was then transformed to sitting posture for convenience in practical monitoring. Referring to the mass and height distribution of the Chinese adult male, a phantom library containing 84 phantoms was constructed by deforming the reference surface phantom. Phantoms in the library have 7 different heights ranging from 155 cm to 185 cm, and there are 12 phantoms with different total body masses in each height. As an example of application, organ specific and total counting efficiencies of Ba-133 were calculated using the MCNPX code, with two series of phantoms selected from the library. The influence of morphological variation on the counting efficiency was analyzed. The results show only using the reference phantom in virtual calibration may lead to an error of 68.9% for total counting efficiency. Thus the influence of morphological difference on virtual calibration can be greatly reduced using the phantom library with a wide range of masses and heights instead of a single reference phantom.

Study of two different radioactive sources for prostate brachytherapy treatment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pereira Neves, Lucio; Perini, Ana Paula; Souza Santos, William de

In this study we evaluated two radioactive sources for brachytherapy treatments. Our main goal was to quantify the absorbed doses on organs and tissues of an adult male patient, submitted to a brachytherapy treatment with two radioactive sources. We evaluated a {sup 192}Ir and a {sup 125}I radioactive sources. The {sup 192}Ir radioactive source is a cylinder with 0.09 cm in diameter and 0.415 cm long. The {sup 125}I radioactive source is also a cylinder, with 0.08 cm in diameter and 0.45 cm long. To evaluate the absorbed dose distribution on the prostate, and other organs and tissues of anmore » adult man, a male virtual anthropomorphic phantom MASH, coupled in the radiation transport code MCNPX 2.7.0, was employed.We simulated 75, 90 and 102 radioactive sources of {sup 125}I and one of {sup 192}Ir, inside the prostate, as normally used in these treatments, and each treatment was simulated separately. As this phantom was developed in a supine position, the displacement of the internal organs of the chest, compression of the lungs and reduction of the sagittal diameter were all taken into account. For the {sup 192}Ir, the higher doses values were obtained for the prostate and surrounding organs, as the colon, gonads and bladder. Considering the {sup 125}I sources, with photons with lower energies, the doses to organs that are far from the prostate were lower. All values for the dose rates are in agreement with those recommended for brachytherapy treatments. Besides that, the new seeds evaluated in this work present usefulness as a new tool in prostate brachytherapy treatments, and the methodology employed in this work may be applied for other radiation sources, or treatments. (authors)« less

Organ dose conversion coefficients for pediatric reference computational phantoms in external photon radiation fields

NASA Astrophysics Data System (ADS)

Chang, Lienard A.

In the event of a radiological accident or attack, it is important to estimate the organ doses to those exposed. In general, it is difficult to measure organ dose directly in the field and therefore dose conversion coefficients (DCC) are needed to convert measurable values such as air kerma to organ dose. Previous work on these coefficients has been conducted mainly for adults with a focus on radiation protection workers. Hence, there is a large gap in the literature for pediatric values. This study coupled a Monte Carlo N-Particle eXtended (MCNPX) code with International Council of Radiological Protection (ICRP)-adopted University of Florida and National Cancer Institute pediatric reference phantoms to calculate a comprehensive list of dose conversion coefficients (mGy/mGy) to convert air-kerma to organ dose. Parameters included ten phantoms (newborn, 1-year, 5-year, 10-year, 15-year old male and female), 28 organs over 33 energies between 0.01 and 20 MeV in six (6) irradiation geometries relevant to a child who might be exposed to a radiological release: anterior-posterior (AP), posterior-anterior (PA), right-lateral (RLAT), left-lateral (LLAT), rotational (ROT), and isotropic (ISO). Dose conversion coefficients to the red bone marrow over 36 skeletal sites were also calculated. It was hypothesized that the pediatric organ dose conversion coefficients would follow similar trends to the published adult values as dictated by human anatomy, but be of a higher magnitude. It was found that while the pediatric coefficients did yield similar patterns to that of the adult coefficients, depending on the organ and irradiation geometry, the pediatric values could be lower or higher than that of the adult coefficients.

Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit

NASA Astrophysics Data System (ADS)

El-Jaby, Samy; Richardson, Richard B.

2015-07-01

Occupational exposures from ionizing radiation are currently regulated for airline travel (<20 km) and for missions to low-Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit.

Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit.

PubMed

El-Jaby, Samy; Richardson, Richard B

2015-07-01

Occupational exposures from ionizing radiation are currently regulated for airline travel (<20 km) and for missions to low-Earth orbit (∼300-400 km). Aircrew typically receive between 1 and 6 mSv of occupational dose annually, while aboard the International Space Station, the area radiation dose equivalent measured over just 168 days was 106 mSv at solar minimum conditions. It is anticipated that space tourism vehicles will reach suborbital altitudes of approximately 100 km and, therefore, the annual occupational dose to flight crew during repeated transits is expected to fall somewhere between those observed for aircrew and astronauts. Unfortunately, measurements of the radiation environment at the high altitudes reached by suborbital vehicles are sparse, and modelling efforts have been similarly limited. In this paper, preliminary MCNPX radiation transport code simulations are developed of the secondary neutron flux profile in air from surface altitudes up to low Earth orbit at solar minimum conditions and excluding the effects of spacecraft shielding. These secondary neutrons are produced by galactic cosmic radiation interacting with Earth's atmosphere and are among the sources of radiation that can pose a health risk. Associated estimates of the operational neutron ambient dose equivalent, used for radiation protection purposes, and the neutron effective dose equivalent that is typically used for estimates of stochastic health risks, are provided in air. Simulations show that the neutron radiation dose rates received at suborbital altitudes are comparable to those experienced by aircrew flying at 7 to 14 km. We also show that the total neutron dose rate tails off beyond the Pfotzer maximum on ascension from surface up to low Earth orbit. Crown Copyright © 2015. Published by Elsevier Ltd. All rights reserved.

Construction of Chinese adult male phantom library and its application in the virtual calibration of in vivo measurement

NASA Astrophysics Data System (ADS)

Chen, Yizheng; Qiu, Rui; Li, Chunyan; Wu, Zhen; Li, Junli

2016-03-01

In vivo measurement is a main method of internal contamination evaluation, particularly for large numbers of people after a nuclear accident. Before the practical application, it is necessary to obtain the counting efficiency of the detector by calibration. The virtual calibration based on Monte Carlo simulation usually uses the reference human computational phantom, and the morphological difference between the monitored personnel with the calibrated phantom may lead to the deviation of the counting efficiency. Therefore, a phantom library containing a wide range of heights and total body masses is needed. In this study, a Chinese reference adult male polygon surface (CRAM_S) phantom was constructed based on the CRAM voxel phantom, with the organ models adjusted to match the Chinese reference data. CRAMS phantom was then transformed to sitting posture for convenience in practical monitoring. Referring to the mass and height distribution of the Chinese adult male, a phantom library containing 84 phantoms was constructed by deforming the reference surface phantom. Phantoms in the library have 7 different heights ranging from 155 cm to 185 cm, and there are 12 phantoms with different total body masses in each height. As an example of application, organ specific and total counting efficiencies of Ba-133 were calculated using the MCNPX code, with two series of phantoms selected from the library. The influence of morphological variation on the counting efficiency was analyzed. The results show only using the reference phantom in virtual calibration may lead to an error of 68.9% for total counting efficiency. Thus the influence of morphological difference on virtual calibration can be greatly reduced using the phantom library with a wide range of masses and heights instead of a single reference phantom.

Monte Carlo study of neutron-ambient dose equivalent to patient in treatment room.

PubMed

Mohammadi, A; Afarideh, H; Abbasi Davani, F; Ghergherehchi, M; Arbabi, A

2016-12-01

This paper presents an analytical method for the calculation of the neutron ambient dose equivalent H* (10) regarding patients, whereby the different concrete types that are used in the surrounding walls of the treatment room are considered. This work has been performed according to a detailed simulation of the Varian 2300C/D linear accelerator head that is operated at 18MV, and silver activation counter as a neutron detector, for which the Monte Carlo MCNPX 2.6 code is used, with and without the treatment room walls. The results show that, when compared to the neutrons that leak from the LINAC, both the scattered and thermal neutrons are the major factors that comprise the out-of field neutron dose. The scattering factors for the limonite-steel, magnetite-steel, and ordinary concretes have been calculated as 0.91±0.09, 1.08±0.10, and 0.371±0.01, respectively, while the corresponding thermal factors are 34.22±3.84, 23.44±1.62, and 52.28±1.99, respectively (both the scattering and thermal factors are for the isocenter region); moreover, the treatment room is composed of magnetite-steel and limonite-steel concretes, so the neutron doses to the patient are 1.79 times and 1.62 times greater than that from an ordinary concrete composition. The results also confirm that the scattering and thermal factors do not depend on the details of the chosen linear accelerator head model. It is anticipated that the results of the present work will be of great interest to the manufacturers of medical linear accelerators. Copyright © 2016. Published by Elsevier Ltd.

The feasibility study of 177Lu production in Miniature Neutron Source Reactors using a multi-stage approach in Isfahan, Iran.

PubMed

Golabian, A; Hosseini, M A; Ahmadi, M; Soleimani, B; Rezvanifard, M

2018-01-01

Miniature neutron source reactors (MNSRs) are among the safest and economic research reactors with potentials to be used for neutron studies. This manuscript explores the feasibility of 177 Lu production in Isfahan MNSR reactor using direct production route. In this study, to assess the specific activity of the produced radioisotope, a simulation was carried out through the MCNPX2.6 code. The simulation was validated by irradiating a lutetium disc-like (99.98 chemical purity) at the thermal neutron flux of 5 × 10 11 ncm 2 s -1 and an irradiation time of 4min. After the spectrometry of the irradiated sample, the experimental results of 177 Lu production were compared with the simulation results. In addition, factor from the simulation was extracted by replacing it in the related equations in order to calculate specific activity through a multi-stage approach, and by using different irradiation techniques. The results showed that the simulation technique designed in this study is in agreement with the experimental approach (with a difference of approximately 3%). It was also found that the maximum 177 Lu production at the maximum flux and irradiation time allows access to 723.5mCi/g after 27 cycles. Furthermore, the comparison of irradiation techniques showed that increasing the irradiation time is more effective in 177 Lu production efficiency than increasing the number of irradiation cycles. In a way that increasing the irradiation time would postpone the saturation of the productions. On the other hand, it was shown that the choice of an appropriate irradiation technique for 177 Lu production can be economically important in term of the effective fuel consumption in the reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

External dose-rate conversion factors of radionuclides for air submersion, ground surface contamination and water immersion based on the new ICRP dosimetric setting

PubMed Central

Yoo, Song Jae; Jang, Han-Ki; Lee, Jai-Ki; Noh, Siwan; Cho, Gyuseong

2013-01-01

For the assessment of external doses due to contaminated environment, the dose-rate conversion factors (DCFs) prescribed in Federal Guidance Report 12 (FGR 12) and FGR 13 have been widely used. Recently, there were significant changes in dosimetric models and parameters, which include the use of the Reference Male and Female Phantoms and the revised tissue weighting factors, as well as the updated decay data of radionuclides. In this study, the DCFs for effective and equivalent doses were calculated for three exposure settings: skyshine, groundshine and water immersion. Doses to the Reference Phantoms were calculated by Monte Carlo simulations with the MCNPX 2.7.0 radiation transport code for 26 mono-energy photons between 0.01 and 10 MeV. The transport calculations were performed for the source volume within the cut-off distances practically contributing to the dose rates, which were determined by a simplified calculation model. For small tissues for which the reduction of variances are difficult, the equivalent dose ratios to a larger tissue (with lower statistical errors) nearby were employed to make the calculation efficient. Empirical response functions relating photon energies, and the organ equivalent doses or the effective doses were then derived by the use of cubic-spline fitting of the resulting doses for 26 energy points. The DCFs for all radionuclides considered important were evaluated by combining the photon emission data of the radionuclide and the empirical response functions. Finally, contributions of accompanied beta particles to the skin equivalent doses and the effective doses were calculated separately and added to the DCFs. For radionuclides considered in this study, the new DCFs for the three exposure settings were within ±10 % when compared with DCFs in FGR 13. PMID:23542764

An analytical approach to γ-ray self-shielding effects for radioactive bodies encountered nuclear decommissioning scenarios.

PubMed

Gamage, K A A; Joyce, M J

2011-10-01

A novel analytical approach is described that accounts for self-shielding of γ radiation in decommissioning scenarios. The approach is developed with plutonium-239, cobalt-60 and caesium-137 as examples; stainless steel and concrete have been chosen as the media for cobalt-60 and caesium-137, respectively. The analytical methods have been compared MCNPX 2.6.0 simulations. A simple, linear correction factor relates the analytical results and the simulated estimates. This has the potential to greatly simplify the estimation of self-shielding effects in decommissioning activities. Copyright © 2011 Elsevier Ltd. All rights reserved.

Radiation Generation from Ultra Intense Laser Plasma Interactions with Solid Density Plasmas for Active Interrogation of Nuclear Materials

NASA Astrophysics Data System (ADS)

Zulick, Calvin Andrew

The development of short pulse high power lasers has led to interest in laser based particle accelerators. Laser produced plasmas have been shown to support quasi-static TeV/m acceleration gradients which are more than four orders of magnitude stronger than conventional accelerators. These high gradients have the potential to allow compact particle accelerators for active interrogation of nuclear material. In order to better understand this application, several experiments have been conducted at the HERCULES and Lambda Cubed lasers as the Center for Ultrafast Optical Science at the University of Michigan. Electron acceleration and bremsstrahlung generation were studied on the Lambda Cubed laser. The scaling of the intensity, angular, and material dependence of bremsstrahlung radiation from an intense (I > 10 18 W/cm2 ) laser-solid interaction has been characterized at energies between 100 keV and 1 MeV. These were the first high resolution (lambda / d lambda > 100) measurements of bremsstrahlung photons from a relativistic laser plasma interaction. The electron populations and bremsstrahlung temperatures were modeled in the particle-in-cell code OSIRIS and the Monte Carlo code MCNPX and were in good agreement with the experimental results. Proton acceleration was studied on the HERCULES laser. The effect of three dimensional perturbations of electron sheaths on proton acceleration was investigated through the use of foil, grid, and wire targets. Hot electron density, as measured with an imaging Cu Kalpha crystal, increased as the target surface area was reduced and was correlated to an increase in the temperature of the accelerated proton beam. Additionally, experiments at the HERCULES laser facility have produced directional neutron beams with energies up to 16.8 (+/-0.3) MeV using (d,n) and (p,n) reactions. Efficient (d,n) reactions required the selective acceleration of deuterons through the introduction of a deuterated plastic or cryogenically frozen D2O layer on the surface of a thin film target. The measured neutron yield was up to 1.0 (+/-0.5) x 107 neutrons/sr with a flux 6.2 (+/-3.7) times higher in the forward direction than at 90 degrees . This demonstrated that femtosecond lasers are capable of providing a time averaged neutron flux equivalent to commercial DD generators with the advantage of a directional beam with picosecond bunch duration.

Exposure to 137Cs deposited in soil – A Monte Carlo study

NASA Astrophysics Data System (ADS)

da Silveira, Lucas M.; Pereira, Marco A. M.; Neves, Lucio P.; Perini, Ana P.; Belinato, Walmir; Caldas, Linda V. E.; Santos, William S.

2018-03-01

In the event of an environmental contamination with radioactive materials, one of the most dangerous materials is 137Cs. In order to evaluate the radiation doses involved in an environmental contamination of soil, with 137Cs, we carried out a computational dosimetric study. We determined the radiation conversion coefficients (CC) for effective (E) and equivalent (H T) doses, using a male and a female anthropomorphic phantoms. These phantoms were coupled with the MCNPX (2.7.0) Monte Carlo simulation software, for three different types of soil. The highest CC[H T] values were for the gonads and skin (male) and bone marrow and skin (female). We found no difference for the different types of soil.

A plastic scintillator-based 2D thermal neutron mapping system for use in BNCT studies.

PubMed

Ghal-Eh, N; Green, S

2016-06-01

In this study, a scintillator-based measurement instrument is proposed which is capable of measuring a two-dimensional map of thermal neutrons within a phantom based on the detection of 2.22MeV gamma rays generated via nth+H→D+γ reaction. The proposed instrument locates around a small rectangular water phantom (14cm×15cm×20cm) used in Birmingham BNCT facility. The whole system has been simulated using MCNPX 2.6. The results confirm that the thermal flux peaks somewhere between 2cm and 4cm distance from the system entrance which is in agreement with previous studies. Copyright © 2016 Elsevier Ltd. All rights reserved.

Analysis of 238Pu and 56Fe Evaluated Data for Use in MYRRHA

NASA Astrophysics Data System (ADS)

Díez, C. J.; Cabellos, O.; Martínez, J. S.; Stankovskiy, A.; Van den Eynde, G.; Schillebeeckx, P.; Heyse, J.

2014-04-01

A sensitivity analysis on the multiplication factor, keff, to the cross section data has been carried out for the MYRRHA critical configuration in order to show the most relevant reactions. With these results, a further analysis on the 238Pu and 56Fe cross sections has been performed, comparing the evaluations provided in the JEFF-3.1.2 and ENDF/B-VII.1 libraries for these nuclides. Then, the effect in MYRRHA of the differences between evaluations are analysed, presenting the source of the differences. With these results, recommendations for the 56Fe and 238Pu evaluations are suggested. These calculations have been performed with SCALE6.1 and MCNPX-2.7e.

Absolute measurements of fast neutrons using yttrium.

PubMed

Roshan, M V; Springham, S V; Rawat, R S; Lee, P; Krishnan, M

2010-08-01

Yttrium is presented as an absolute neutron detector for pulsed neutron sources. It has high sensitivity for detecting fast neutrons. Yttrium has the property of generating a monoenergetic secondary radiation in the form of a 909 keV gamma-ray caused by inelastic neutron interaction. It was calibrated numerically using MCNPX and does not need periodic recalibration. The total yttrium efficiency for detecting 2.45 MeV neutrons was determined to be f(n) approximately 4.1x10(-4) with an uncertainty of about 0.27%. The yttrium detector was employed in the NX2 plasma focus experiments and showed the neutron yield of the order of 10(8) neutrons per discharge.

Calculation of Purification Systems of Hydrocarbonmoderated Reactors; CALCULO DE SISTEMAS DE PURIFICATION DE REACTORES MODERADOS FOR HIDROCARBUROS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santos, A.A.

1958-01-01

culation of Purification Systems of Hydrocarbonmoderated Reactors). Agustin Alonso Santos. 1958. 23p. As as introduction to the calculation of the purification systems of bydrocarbon-moderated reactors, the effects of heat and radiation on the polyphenols are considered. The chemical, physical, and nuclear properties are tabulated. The formation velocity of the polymers and gases, pyrolysis, effects of heat on the polymer, and the activity accumulated in the moderator ars discussed. The calculation is based on the hypetheses that the radiation catalyzes the formation of polymers, the velocity of the polymerization reaction is constant, the polymer concentration is maintained at a limit whichmore » does not adversely affect the heat transfer properties, the velocity of the separation of polymers in the distillation column is in proportion to their concentration in the hydrocarbon and the pyrolysis causes gaseous products. Formulas are derived expressing the purified flow and the activities accumulated in the distillation residues. The results are applied to the parification system of the Organic Moderated Reactor Experiment (J.S.R.)« less

Cross correlation calculations and neutron scattering analysis for a portable solid state neutron detection system

NASA Astrophysics Data System (ADS)

Saltos, Andrea

In efforts to perform accurate dosimetry, Oakes et al. [Nucl. Intrum. Mehods. (2013)] introduced a new portable solid state neutron rem meter based on an adaptation of the Bonner sphere and the position sensitive long counter. The system utilizes high thermal efficiency neutron detectors to generate a linear combination of measurement signals that are used to estimate the incident neutron spectra. The inversion problem associated to deduce dose from the counts in individual detector elements is addressed by applying a cross-correlation method which allows estimation of dose with average errors less than 15%. In this work, an evaluation of the performance of this system was extended to take into account new correlation techniques and neutron scattering contribution. To test the effectiveness of correlations, the Distance correlation, Pearson Product-Moment correlation, and their weighted versions were performed between measured spatial detector responses obtained from nine different test spectra, and the spatial response of Library functions generated by MCNPX. Results indicate that there is no advantage of using the Distance Correlation over the Pearson Correlation, and that weighted versions of these correlations do not increase their performance in evaluating dose. Both correlations were proven to work well even at low integrated doses measured for short periods of time. To evaluate the contribution produced by room-return neutrons on the dosimeter response, MCNPX was used to simulate dosimeter responses for five isotropic neutron sources placed inside different sizes of rectangular concrete rooms. Results show that the contribution of scattered neutrons to the response of the dosimeter can be significant, so that for most cases the dose is over predicted with errors as large as 500%. A possible method to correct for the contribution of room-return neutrons is also assessed and can be used as a good initial estimate on how to approach the problem.

Monte Carlo design and simulation of a grid-type multi-layer pixel collimator for radiotherapy: Feasibility study

NASA Astrophysics Data System (ADS)

Yoon, Do-Kun; Jung, Joo-Young; Suh, Tae Suk

2014-05-01

In order to confirm the possibility of field application of a different type collimator with a multileaf collimator (MLC), we constructed a grid-type multi-layer pixel collimator (GTPC) by using a Monte Carlo n-particle simulation (MCNPX). In this research, a number of factors related to the performance of the GPTC were evaluated using simulated output data of a basic MLC model. A layer was comprised of a 1024-pixel collimator (5.0 × 5.0 mm2) which could operate individually as a grid-type collimator (32 × 32). A 30-layer collimator was constructed for a specific portal form to pass radiation through the opening and closing of each pixel cover. The radiation attenuation level and the leakage were compared between the GTPC modality simulation and MLC modeling (tungsten, 17.50 g/cm3, 5.0 × 70.0 × 160.0 mm3) currently used for a radiation field. Comparisons of the portal imaging, the lateral dose profile from a virtual water phantom, the dependence of the performance on the increase in the number of layers, the radiation intensity modulation verification, and the geometric error between the GTPC and the MLC were done using the MCNPX simulation data. From the simulation data, the intensity modulation of the GTPC showed a faster response than the MLC's (29.6%). In addition, the agreement between the doses that should be delivered to the target region was measured as 97.0%, and the GTPC system had an error below 0.01%, which is identical to that of MLC. A Monte Carlo simulation of the GTPC could be useful for verification of application possibilities. Because the line artifact is caused by the grid frame and the folded cover, a lineal dose transfer type is chosen for the operation of this system. However, the result of GTPC's performance showed that the methods of effective intensity modulation and the specific geometric beam shaping differed with the MLC modality.

Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

NASA Astrophysics Data System (ADS)

Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

2013-11-01

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

Delta-ray Production in MCNP 6.2.0

NASA Astrophysics Data System (ADS)

Anderson, C.; McKinney, G.; Tutt, J.; James, M.

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. This paper discusses the MCNP6 heavy charged-particle implementation and provides production results for several benchmark/test problems.

Exploiting Fission Chain Reaction Dynamics to Image Fissile Materials

NASA Astrophysics Data System (ADS)

Chapman, Peter Henry

Radiation imaging is one potential method to verify nuclear weapons dismantlement. The neutron coded aperture imager (NCAI), jointly developed by Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL), is capable of imaging sources of fast (e.g., fission spectrum) neutrons using an array of organic scintillators. This work presents a method developed to discriminate between non-multiplying (i.e., non-fissile) neutron sources and multiplying (i.e., fissile) neutron sources using the NCAI. This method exploits the dynamics of fission chain-reactions; it applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions. TCPH analyzes the neutron energy deposited in the organic scintillator vs. the apparent neutron time-of-flight. Energy deposition is estimated from light output, and time-of-flight is estimated from the time between the neutron interaction and the immediately preceding gamma interaction. Neutrons that deposit more energy than can be accounted for by their apparent time-of-flight are identified as fission chain-reaction neutrons, and the image is reconstructed using only these neutron detection events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility (DAF) in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252Cf source from the image while preserving the fissileWGPu source. TCPH analysis was also applied to additional scenes in which theWGPu and 252Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252Cf source and retain the fissileWGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, solid spheres ofWGPu are retained at a 1sigma level for neutron multiplications M -˜ 3.0 and above, while hollowWGPu spheres are retained for M -˜ 2.7 and above.

Assessment of the dose distribution inside a cardiac cath lab using TLD measurements and Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Baptista, M.; Teles, P.; Cardoso, G.; Vaz, P.

2014-11-01

Over the last decade, there was a substantial increase in the number of interventional cardiology procedures worldwide, and the corresponding ionizing radiation doses for both the medical staff and patients became a subject of concern. Interventional procedures in cardiology are normally very complex, resulting in long exposure times. Also, these interventions require the operator to work near the patient and, consequently, close to the primary X-ray beam. Moreover, due to the scattered radiation from the patient and the equipment, the medical staff is also exposed to a non-uniform radiation field that can lead to a significant exposure of sensitive body organs and tissues, such as the eye lens, the thyroid and the extremities. In order to better understand the spatial variation of the dose and dose rate distributions during an interventional cardiology procedure, the dose distribution around a C-arm fluoroscopic system, in operation in a cardiac cath lab at Portuguese Hospital, was estimated using both Monte Carlo (MC) simulations and dosimetric measurements. To model and simulate the cardiac cath lab, including the fluoroscopic equipment used to execute interventional procedures, the state-of-the-art MC radiation transport code MCNPX 2.7.0 was used. Subsequently, Thermo-Luminescent Detector (TLD) measurements were performed, in order to validate and support the simulation results obtained for the cath lab model. The preliminary results presented in this study reveal that the cardiac cath lab model was successfully validated, taking into account the good agreement between MC calculations and TLD measurements. The simulated results for the isodose curves related to the C-arm fluoroscopic system are also consistent with the dosimetric information provided by the equipment manufacturer (Siemens). The adequacy of the implemented computational model used to simulate complex procedures and map dose distributions around the operator and the medical staff is discussed, in view of the optimization principle (and the associated ALARA objective), one of the pillars of the international system of radiological protection.

Preliminary Analysis of the Multisphere Neutron Spectrometer

NASA Technical Reports Server (NTRS)

Goldhagen, P.; Kniss, T.; Wilson, J. W.; Singleterry, R. C.; Jones, I. W.; VanSteveninck, W.

2003-01-01

Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from galactic cosmic radiation. Crews of future high-speed commercial aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the Atmospheric Ionizing Radiation (AIR) Project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (thermal to greater than 10 GeV), total neutron fluence rate, and neutron effective dose and dose equivalent rates and their dependence on altitude and geomagnetic cutoff. The measured cosmic-ray neutron spectra have almost no thermal neutrons, a large "evaporation" peak near 1 MeV and a second broad peak near 100 MeV which contributes about 69% of the neutron effective dose. At high altitude, geomagnetic latitude has very little effect on the shape of the spectrum, but it is the dominant variable affecting neutron fluence rate, which was 8 times higher at the northernmost measurement location than it was at the southernmost. The shape of the spectrum varied only slightly with altitude from 21 km down to 12 km (56 - 201 grams per square centimeter atmospheric depth), but was significantly different on the ground. In all cases, ambient dose equivalent was greater than effective dose for cosmic-ray neutrons.

Evaluation of the medical and occupational shielding in cerebral angiography using Monte Carlo simulations and virtual anthropomorphic phantoms

NASA Astrophysics Data System (ADS)

Santos, William S.; Neves, Lucio P.; Perini, Ana P.; Caldas, Linda V. E.; Maia, Ana F.

2015-12-01

Cerebral angiography exams may provide valuable diagnostic information for the patients with suspect of cerebral diseases, but it may also deliver high doses of radiation to the patients and medical staff. In order to evaluate the medical and occupational expositions from different irradiation conditions, Monte Carlo (MC) simulations were employed. Virtual anthropomorphic phantoms (MASH) were used to represent the patient and the physician inside a typical fluoroscopy room, also simulated in details, incorporated in the MCNPX 2.7.0 MC code. The evaluation was carried out by means of dose conversion coefficients (CCs) for equivalent (H) and effective (E) doses normalized by the air kerma-area product (KAP). The CCs for the surface entrance dose of the patient (ESD) and equivalent dose for the eyes of the medical staff were determined, because CA exams present higher risks for those organs. The tube voltage was 80 kVp, and Al filters with thicknesses of 2.5 mm, 3.5 mm and 4.0 mm were positioned in the beams. Two projections were simulated: posterior-anterior (PA) and right-lateral (RLAT). In all situations there was an increase of the CC values with the increase of the Al filtration. The highest dose was obtained for a RLAT projection with a 4.0 mm Al filter. In this projection, the ESD/KAP and E/KAP values to patient were 11 (14%) mGy/Gy cm2 and 0.12 (0.1%) mSv/Gy cm2, respectively. For the physician, the use of shield lead glass suspended and lead curtain attached to the surgical table resulted in a significant reduction of the CCs. The use of MC simulations proved to be a very important tool in radiation protection dosimetry, and specifically in this study several parameters could be evaluated, which would not be possible experimentally.

Determination of the quenching correction factors for plastic scintillation detectors in therapeutic high-energy proton beams

PubMed Central

Wang, L L W; Perles, L A; Archambault, L; Sahoo, N; Mirkovic, D; Beddar, S

2013-01-01

The plastic scintillation detectors (PSD) have many advantages over other detectors in small field dosimetry due to its high spatial resolution, excellent water equivalence and instantaneous readout. However, in proton beams, the PSDs will undergo a quenching effect which makes the signal level reduced significantly when the detector is close to Bragg peak where the linear energy transfer (LET) for protons is very high. This study measures the quenching correction factor (QCF) for a PSD in clinical passive-scattering proton beams and investigates the feasibility of using PSDs in depth-dose measurements in proton beams. A polystyrene based PSD (BCF-12, ϕ0.5mm×4mm) was used to measure the depth-dose curves in a water phantom for monoenergetic unmodulated proton beams of nominal energies 100, 180 and 250 MeV. A Markus plane-parallel ion chamber was also used to get the dose distributions for the same proton beams. From these results, the QCF as a function of depth was derived for these proton beams. Next, the LET depth distributions for these proton beams were calculated by using the MCNPX Monte Carlo code, based on the experimentally validated nozzle models for these passive-scattering proton beams. Then the relationship between the QCF and the proton LET could be derived as an empirical formula. Finally, the obtained empirical formula was applied to the PSD measurements to get the corrected depth-dose curves and they were compared to the ion chamber measurements. A linear relationship between QCF and LET, i.e. Birks' formula, was obtained for the proton beams studied. The result is in agreement with the literature. The PSD measurements after the quenching corrections agree with ion chamber measurements within 5%. PSDs are good dosimeters for proton beam measurement if the quenching effect is corrected appropriately. PMID:23128412

A study on the optimum fast neutron flux for boron neutron capture therapy of deep-seated tumors.

PubMed

Rasouli, Fatemeh S; Masoudi, S Farhad

2015-02-01

High-energy neutrons, named fast neutrons which have a number of undesirable biological effects on tissue, are a challenging problem in beam designing for Boron Neutron Capture Therapy, BNCT. In spite of this fact, there is not a widely accepted criterion to guide the beam designer to determine the appropriate contribution of fast neutrons in the spectrum. Although a number of researchers have proposed a target value for the ratio of fast neutron flux to epithermal neutron flux, it can be shown that this criterion may not provide the optimum treatment condition. This simulation study deals with the determination of the optimum contribution of fast neutron flux in the beam for BNCT of deep-seated tumors. Since the dose due to these high-energy neutrons damages shallow tissues, delivered dose to skin is considered as a measure for determining the acceptability of the designed beam. To serve this purpose, various beam shaping assemblies that result in different contribution of fast neutron flux are designed. The performances of the neutron beams corresponding to such configurations are assessed in a simulated head phantom. It is shown that the previously used criterion, which suggests a limit value for the contribution of fast neutrons in beam, does not necessarily provide the optimum condition. Accordingly, it is important to specify other complementary limits considering the energy of fast neutrons. By analyzing various neutron spectra, two limits on fast neutron flux are proposed and their validity is investigated. The results show that considering these limits together with the widely accepted IAEA criteria makes it possible to have a more realistic assessment of sufficiency of the designed beam. Satisfying these criteria not only leads to reduction of delivered dose to skin, but also increases the advantage depth in tissue and delivered dose to tumor during the treatment time. The Monte Carlo Code, MCNP-X, is used to perform these simulations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Monte-Carlo based assessment of MAGIC, MAGICAUG, PAGATUG and PAGATAUG polymer gel dosimeters for ovaries and uterus organ dosimetry in brachytherapy, nuclear medicine and Tele-therapy.

PubMed

Adinehvand, Karim; Rahatabad, Fereidoun Nowshiravan

2018-06-01

Calculation of 3D dose distribution during radiotherapy and nuclear medicine helps us for better treatment of sensitive organs such as ovaries and uterus. In this research, we investigate two groups of normoxic dosimeters based on meta-acrylic acid (MAGIC and MAGICAUG) and polyacrylamide (PAGATUG and PAGATAUG) for brachytherapy, nuclear medicine and Tele-therapy in their sensitive and critical role as organ dosimeters. These polymer gel dosimeters are compared with soft tissue while irradiated by different energy photons in therapeutic applications. This comparison has been simulated by Monte-Carlo based MCNPX code. ORNL phantom-Female has been used to model the critical organs of kidneys, ovaries and uterus. Right kidney is proposed to be the source of irradiation and another two organs are exposed to this irradiation. Effective atomic numbers of soft tissue, MAGIC, MAGICAUG, PAGATUG and PAGATAUG are 6.86, 7.07, 6.95, 7.28, and 7.07 respectively. Results show the polymer gel dosimeters are comparable to soft tissue for using in nuclear medicine and Tele-therapy. Differences between gel dosimeters and soft tissue are defined as the dose responses. This difference is less than 4.1%, 22.6% and 71.9% for Tele-therapy, nuclear medicine and brachytherapy respectively. The results approved that gel dosimeters are the best choice for ovaries and uterus in nuclear medicine and Tele-therapy respectively. Due to the slight difference between the effective atomic numbers of these polymer gel dosimeters and soft tissue, these polymer gels are not suitable for brachytherapy since the dependence of photon interaction to atomic number, for low energy brachytherapy, had been so effective. Also this dependence to atomic number, decrease for photoelectric and increase for Compton. Therefore polymer gel dosimeters are not a good alternative to soft tissue replacement in brachytherapy. Copyright © 2018 Elsevier B.V. All rights reserved.

Determination of the quenching correction factors for plastic scintillation detectors in therapeutic high-energy proton beams

NASA Astrophysics Data System (ADS)

Wang, L. L. W.; Perles, L. A.; Archambault, L.; Sahoo, N.; Mirkovic, D.; Beddar, S.

2012-12-01

Plastic scintillation detectors (PSDs) have many advantages over other detectors in small field dosimetry due to their high spatial resolution, excellent water equivalence and instantaneous readout. However, in proton beams, the PSDs undergo a quenching effect which makes the signal level reduced significantly when the detector is close to the Bragg peak where the linear energy transfer (LET) for protons is very high. This study measures the quenching correction factor (QCF) for a PSD in clinical passive-scattering proton beams and investigates the feasibility of using PSDs in depth-dose measurements in proton beams. A polystyrene-based PSD (BCF-12, ϕ0.5 mm × 4 mm) was used to measure the depth-dose curves in a water phantom for monoenergetic unmodulated proton beams of nominal energies 100, 180 and 250 MeV. A Markus plane-parallel ion chamber was also used to get the dose distributions for the same proton beams. From these results, the QCF as a function of depth was derived for these proton beams. Next, the LET depth distributions for these proton beams were calculated by using the MCNPX Monte Carlo code, based on the experimentally validated nozzle models for these passive-scattering proton beams. Then the relationship between the QCF and the proton LET could be derived as an empirical formula. Finally, the obtained empirical formula was applied to the PSD measurements to get the corrected depth-dose curves and they were compared to the ion chamber measurements. A linear relationship between the QCF and LET, i.e. Birks' formula, was obtained for the proton beams studied. The result is in agreement with the literature. The PSD measurements after the quenching corrections agree with ion chamber measurements within 5%. PSDs are good dosimeters for proton beam measurement if the quenching effect is corrected appropriately.

Mean glandular dose coefficients (D(g)N) for x-ray spectra used in contemporary breast imaging systems.

PubMed

Nosratieh, Anita; Hernandez, Andrew; Shen, Sam Z; Yaffe, Martin J; Seibert, J Anthony; Boone, John M

2015-09-21

To develop tables of normalized glandular dose coefficients D(g)N for a range of anode-filter combinations and tube voltages used in contemporary breast imaging systems. Previously published mono-energetic D(g)N values were used with various spectra to mathematically compute D(g)N coefficients. The tungsten anode spectra from TASMICS were used; molybdenum and rhodium anode-spectra were generated using MCNPX Monte Carlo code. The spectra were filtered with various thicknesses of Al, Rh, Mo or Cu. An initial half value layer (HVL) calculation was made using the anode and filter material. A range of the HVL values was produced with the addition of small thicknesses of polymethyl methacrylate (PMMA) as a surrogate for the breast compression paddle, to produce a range of HVL values at each tube voltage. Using a spectral weighting method, D(g)N coefficients for the generated spectra were calculated for breast glandular densities of 0%, 12.5%, 25%, 37.5%, 50% and 100% for a range of compressed breast thicknesses from 3 to 8 cm. Eleven tables of normalized glandular dose (D(g)N) coefficients were produced for the following anode/filter combinations: W + 50 μm Ag, W + 500 μm Al, W + 700 μm Al, W + 200 μm Cu, W + 300 μm Cu, W + 50 μm Rh, Mo + 400 μm Cu, Mo + 30 μm Mo, Mo + 25 μm Rh, Rh + 400 μm Cu and Rh + 25 μm Rh. Where possible, these results were compared to previously published D(g)N values and were found to be on average less than 2% different than previously reported values.Over 200 pages of D(g)N coefficients were computed for modeled x-ray system spectra that are used in a number of new breast imaging applications. The reported values were found to be in excellent agreement when compared to published values.

Mean Glandular dose coefficients (DgN) for x-ray spectra used in contemporary breast imaging systems

PubMed Central

Nosratieh, Anita; Hernandez, Andrew; Shen, Sam Z.; Yaffe, Martin J.; Seibert, J. Anthony; Boone, John M.

2015-01-01

Purpose To develop tables of normalized glandular dose coefficients DgN for a range of anode–filter combinations and tube voltages used in contemporary breast imaging systems. Methods Previously published mono-energetic DgN values were used with various spectra to mathematically compute DgN coefficients. The tungsten anode spectra from TASMICS were used; Molybdenum and Rhodium anode-spectra were generated using MCNPx Monte Carlo code. The spectra were filtered with various thicknesses of Al, Rh, Mo or Cu. An initial HVL calculation was made using the anode and filter material. A range of the HVL values was produced with the addition of small thicknesses of polymethyl methacrylate (PMMA) as a surrogate for the breast compression paddle, to produce a range of HVL values at each tube voltage. Using a spectral weighting method, DgN coefficients for the generated spectra were calculated for breast glandular densities of 0%, 12.5%, 25%, 37.5%, 50% and 100% for a range of compressed breast thicknesses from 3 to 8 cm. Results Eleven tables of normalized glandular dose (DgN) coefficients were produced for the following anode/filter combinations: W + 50 μm Ag, W + 500 μm Al, W + 700 μm Al, W + 200 μm Cu, W + 300 μm Cu, W + 50 μm Rh, Mo + 400 μm Cu, Mo + 30 μm Mo, Mo + 25 μm Rh, Rh + 400 μm Cu and Rh + 25 μm Rh. Where possible, these results were compared to previously published DgN values and were found to be on average less than 2% different than previously reported values. Conclusion Over 200-pages of DgN coefficients were computed for modeled x-ray system spectra that are used in a number of new breast imaging applications. The reported values were found to be in excellent agreement when compared to published values. PMID:26348995

SU-E-T-235: Monte Carlo Analysis of the Dose Enhancement in the Scalp of Patients Due to Titanium Plate Backscatter During Post-Operative Radiotherapy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardin, M; Elson, H; Lamba, M

2014-06-01

Purpose: To quantify the clinically observed dose enhancement adjacent to cranial titanium fixation plates during post-operative radiotherapy. Methods: Irradiation of a titanium burr hole cover was simulated using Monte Carlo code MCNPX for a 6 MV photon spectrum to investigate backscatter dose enhancement due to increased production of secondary electrons within the titanium plate. The simulated plate was placed 3 mm deep in a water phantom, and dose deposition was tallied for 0.2 mm thick cells adjacent to the entrance and exit sides of the plate. These results were compared to a simulation excluding the presence of the titanium tomore » calculate relative dose enhancement on the entrance and exit sides of the plate. To verify simulated results, two titanium burr hole covers (Synthes, Inc. and Biomet, Inc.) were irradiated with 6 MV photons in a solid water phantom containing GafChromic MD-55 film. The phantom was irradiated on a Varian 21EX linear accelerator at multiple gantry angles (0–180 degrees) to analyze the angular dependence of the backscattered radiation. Relative dose enhancement was quantified using computer software. Results: Monte Carlo simulations indicate a relative difference of 26.4% and 7.1% on the entrance and exit sides of the plate respectively. Film dosimetry results using a similar geometry indicate a relative difference of 13% and -10% on the entrance and exit sides of the plate respectively. Relative dose enhancement on the entrance side of the plate decreased with increasing gantry angle from 0 to 180 degrees. Conclusion: Film and simulation results demonstrate an increase in dose to structures immediately adjacent to cranial titanium fixation plates. Increased beam obliquity has shown to alleviate dose enhancement to some extent. These results are consistent with clinically observed effects.« less

Dosimetric characterizations of GZP6 60Co high dose rate brachytherapy sources: application of superimposition method

PubMed Central

Bahreyni Toossi, Mohammad Taghi; Ghorbani, Mahdi; Mowlavi, Ali Asghar; Meigooni, Ali Soleimani

2012-01-01

Background Dosimetric characteristics of a high dose rate (HDR) GZP6 Co-60 brachytherapy source have been evaluated following American Association of Physicists in MedicineTask Group 43U1 (AAPM TG-43U1) recommendations for their clinical applications. Materials and methods MCNP-4C and MCNPX Monte Carlo codes were utilized to calculate dose rate constant, two dimensional (2D) dose distribution, radial dose function and 2D anisotropy function of the source. These parameters of this source are compared with the available data for Ralstron 60Co and microSelectron192Ir sources. Besides, a superimposition method was developed to extend the obtained results for the GZP6 source No. 3 to other GZP6 sources. Results The simulated value for dose rate constant for GZP6 source was 1.104±0.03 cGyh-1U-1. The graphical and tabulated radial dose function and 2D anisotropy function of this source are presented here. The results of these investigations show that the dosimetric parameters of GZP6 source are comparable to those for the Ralstron source. While dose rate constant for the two 60Co sources are similar to that for the microSelectron192Ir source, there are differences between radial dose function and anisotropy functions. Radial dose function of the 192Ir source is less steep than both 60Co source models. In addition, the 60Co sources are showing more isotropic dose distribution than the 192Ir source. Conclusions The superimposition method is applicable to produce dose distributions for other source arrangements from the dose distribution of a single source. The calculated dosimetric quantities of this new source can be introduced as input data to the GZP6 treatment planning system (TPS) and to validate the performance of the TPS. PMID:23077455

A methodology to develop computational phantoms with adjustable posture for WBC calibration

NASA Astrophysics Data System (ADS)

Ferreira Fonseca, T. C.; Bogaerts, R.; Hunt, John; Vanhavere, F.

2014-11-01

A Whole Body Counter (WBC) is a facility to routinely assess the internal contamination of exposed workers, especially in the case of radiation release accidents. The calibration of the counting device is usually done by using anthropomorphic physical phantoms representing the human body. Due to such a challenge of constructing representative physical phantoms a virtual calibration has been introduced. The use of computational phantoms and the Monte Carlo method to simulate radiation transport have been demonstrated to be a worthy alternative. In this study we introduce a methodology developed for the creation of realistic computational voxel phantoms with adjustable posture for WBC calibration. The methodology makes use of different software packages to enable the creation and modification of computational voxel phantoms. This allows voxel phantoms to be developed on demand for the calibration of different WBC configurations. This in turn helps to study the major source of uncertainty associated with the in vivo measurement routine which is the difference between the calibration phantoms and the real persons being counted. The use of realistic computational phantoms also helps the optimization of the counting measurement. Open source codes such as MakeHuman and Blender software packages have been used for the creation and modelling of 3D humanoid characters based on polygonal mesh surfaces. Also, a home-made software was developed whose goal is to convert the binary 3D voxel grid into a MCNPX input file. This paper summarizes the development of a library of phantoms of the human body that uses two basic phantoms called MaMP and FeMP (Male and Female Mesh Phantoms) to create a set of male and female phantoms that vary both in height and in weight. Two sets of MaMP and FeMP phantoms were developed and used for efficiency calibration of two different WBC set-ups: the Doel NPP WBC laboratory and AGM laboratory of SCK-CEN in Mol, Belgium.

A methodology to develop computational phantoms with adjustable posture for WBC calibration.

PubMed

Fonseca, T C Ferreira; Bogaerts, R; Hunt, John; Vanhavere, F

2014-11-21

A Whole Body Counter (WBC) is a facility to routinely assess the internal contamination of exposed workers, especially in the case of radiation release accidents. The calibration of the counting device is usually done by using anthropomorphic physical phantoms representing the human body. Due to such a challenge of constructing representative physical phantoms a virtual calibration has been introduced. The use of computational phantoms and the Monte Carlo method to simulate radiation transport have been demonstrated to be a worthy alternative. In this study we introduce a methodology developed for the creation of realistic computational voxel phantoms with adjustable posture for WBC calibration. The methodology makes use of different software packages to enable the creation and modification of computational voxel phantoms. This allows voxel phantoms to be developed on demand for the calibration of different WBC configurations. This in turn helps to study the major source of uncertainty associated with the in vivo measurement routine which is the difference between the calibration phantoms and the real persons being counted. The use of realistic computational phantoms also helps the optimization of the counting measurement. Open source codes such as MakeHuman and Blender software packages have been used for the creation and modelling of 3D humanoid characters based on polygonal mesh surfaces. Also, a home-made software was developed whose goal is to convert the binary 3D voxel grid into a MCNPX input file. This paper summarizes the development of a library of phantoms of the human body that uses two basic phantoms called MaMP and FeMP (Male and Female Mesh Phantoms) to create a set of male and female phantoms that vary both in height and in weight. Two sets of MaMP and FeMP phantoms were developed and used for efficiency calibration of two different WBC set-ups: the Doel NPP WBC laboratory and AGM laboratory of SCK-CEN in Mol, Belgium.

Passive Safety Features Evaluation of KIPT Neutron Source Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Zhaopeng; Gohar, Yousry

2016-06-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assemblymore » is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.« less

Accuracy evaluation of distance inverse square law in determining virtual electron source location in Siemens Primus linac.

PubMed

Douk, Hamid Shafaei; Aghamiri, Mahmoud Reza; Ghorbani, Mahdi; Farhood, Bagher; Bakhshandeh, Mohsen; Hemmati, Hamid Reza

2018-01-01

The aim of this study is to evaluate the accuracy of the inverse square law (ISL) method for determining location of virtual electron source ( S Vir ) in Siemens Primus linac. So far, different experimental methods have presented for determining virtual and effective electron source location such as Full Width at Half Maximum (FWHM), Multiple Coulomb Scattering (MCS), and Multi Pinhole Camera (MPC) and Inverse Square Law (ISL) methods. Among these methods, Inverse Square Law is the most common used method. Firstly, Siemens Primus linac was simulated using MCNPX Monte Carlo code. Then, by using dose profiles obtained from the Monte Carlo simulations, the location of S Vir was calculated for 5, 7, 8, 10, 12 and 14 MeV electron energies and 10 cm × 10 cm, 15 cm × 15 cm, 20 cm × 20 cm and 25 cm × 25 cm field sizes. Additionally, the location of S Vir was obtained by the ISL method for the mentioned electron energies and field sizes. Finally, the values obtained by the ISL method were compared to the values resulted from Monte Carlo simulation. The findings indicate that the calculated S Vir values depend on beam energy and field size. For a specific energy, with increase of field size, the distance of S Vir increases for most cases. Furthermore, for a special applicator, with increase of electron energy, the distance of S Vir increases for most cases. The variation of S Vir values versus change of field size in a certain energy is more than the variation of S Vir values versus change of electron energy in a certain field size. According to the results, it is concluded that the ISL method can be considered as a good method for calculation of S Vir location in higher electron energies (14 MeV).

Monte-Carlo Application for Nondestructive Nuclear Waste Analysis

NASA Astrophysics Data System (ADS)

Carasco, C.; Engels, R.; Frank, M.; Furletov, S.; Furletova, J.; Genreith, C.; Havenith, A.; Kemmerling, G.; Kettler, J.; Krings, T.; Ma, J.-L.; Mauerhofer, E.; Neike, D.; Payan, E.; Perot, B.; Rossbach, M.; Schitthelm, O.; Schumann, M.; Vasquez, R.

2014-06-01

Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum, neutron flux distribution. The validation of the measurements simulations with Mont-Carlo transport codes for the design, optimization and data analysis of further P&DGNAA facilities is performed in collaboration with LMN CEA Cadarache. The performance of the prompt gamma neutron activation analysis (PGNAA) for the nondestructive determination of actinides in small samples is investigated. The quantitative determination of actinides relies on the precise knowledge of partial neutron capture cross sections. Up to today these cross sections are not very accurate for analytical purpose. The goal of the TANDEM (Trans-uranium Actinides' Nuclear Data - Evaluation and Measurement) Collaboration is the evaluation of these cross sections. Cross sections are measured using prompt gamma activation analysis facilities in Budapest and Munich. Geant4 is used to optimally design the detection system with Compton suppression. Furthermore, for the evaluation of the cross sections it is strongly needed to correct the results to the self-attenuation of the prompt gammas within the sample. In the framework of cooperation RWTH Aachen University, Forschungszentrum Jülich and the Siemens AG will study the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA). The system is based on a 14 MeV neutron source and an advanced detector system (a-Si flat panel) linked to an exclusive converter/scintillator for fast neutrons. For shielding and radioprotection studies the codes MCNPX and Geant4 were used. The two codes were benchmarked in processing time and accuracy in the neutron and gamma fluxes. Also the detector response was simulated with Geant4 to optimize components of the system.

Cosmogenic nuclide production rates as a function of latitude and altitude calculated via a physics based model and excitation functions

NASA Astrophysics Data System (ADS)

Argento, D.; Reedy, R. C.; Stone, J. O.

2012-12-01

Cosmogenic nuclides have been used to develop a set of tools critical to the quantification of a wide range of geomorphic and climatic processes and events (Dunai 2010). Having reliable absolute measurement methods has had great impact on research constraining ice age extents as well as providing important climatic data via well constrained erosion rates, etc. Continuing to improve CN methods is critical for these sciences. While significant progress has been made in the last two decades to reduce uncertainties (Dunai 2010; Gosse & Phillips 2001), numerous aspects still need to be refined in order to achieve the analytic resolution desired by glaciologists and geomorphologists. In order to investigate the finer details of the radiation responsible for cosmogenic nuclide production, we have developed a physics based model which models the radiation cascade of primary and secondary cosmic-rays through the atmosphere. In this study, a Monte Carlo method radiation transport code, MCNPX, is used to model the galactic cosmic-ray (GCR) radiation impinging on the upper atmosphere. Beginning with a spectrum of high energy protons and alpha particles at the top of the atmosphere, the code tracks the primary and resulting secondary particles through a model of the Earth's atmosphere and into the lithosphere. Folding the neutron and proton flux results with energy dependent cross sections for nuclide production provides production rates for key cosmogenic nuclides (Argento et al. 2012, in press; Reedy 2012, in press). Our initial study for high latitude shows that nuclides scale at different rates for each nuclide (Argento 2012, in press). Furthermore, the attenuation length for each of these nuclide production rates increases with altitude, and again, they increase at different rates. This has the consequence of changing the production rate ratio as a function of altitude. The earth's geomagnetic field differentially filters low energy cosmic-rays by deflecting them away. This effect is strongest at the equator. This filtering reduces the total number of particles, and also biases the spectrum towards the higher energies. This effect is known to generally increase the attenuation length of production with altitude at the same time reducing the overall production rate. Our model now extends from high latitude to the equator. We expect the production rates, attenuation lengths and production ratios to also be functions of latitude. Our radiation model results are being analyzed and nuclide production rate results will be presented at the conference.

Delta-ray Production in MCNP 6.2.0

DOE PAGES

Anderson, Casey Alan; McKinney, Gregg Walter; Tutt, James Robert; ...

2017-10-26

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. As a result, this paper discusses the MCNP6more » heavy charged-particle implementation and provides production results for several benchmark/test problems.« less

Measurement and simulation of the cross sections for nuclide production in {sup nat}W and {sup 181}Ta targets irradiated with 0.04- to 2.6-GeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Titarenko, Yu. E., E-mail: Yury.Titarenko@itep.ru; Batyaev, V. F.; Titarenko, A. Yu.

The cross sections for nuclide production in thin {sup nat}Wand {sup 181}Ta targets irradiated by 0.04-2.6-GeV protons have been measured by direct {gamma} spectrometry using two {gamma} spectrometers with the resolutions of 1.8 and 1.7 keV in the {sup 60}Co 1332-keV {gamma} line. As a result, 1895 yields of radioactive residual product nuclei have been obtained. The {sup 27}Al(p, x){sup 22}Na reaction has been used as a monitor reaction. The experimental data have been compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

Measurement and simulation of the cross sections for nuclide production in {sup 93}Nb and {sup nat}Ni targets irradiated with 0.04- to 2.6-GeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Titarenko, Yu. E., E-mail: Yury.Titarenko@itep.ru; Batyaev, V. F.; Titarenko, A. Yu.

The cross sections for nuclide production in thin {sup 93}Nb and {sup nat}Ni targets irradiated by 0.04- to 2.6-GeV protons have been measured by direct {gamma} spectrometry using two {gamma} spectrometers with the resolutions of 1.8 and 1.7 keV in the {sup 60}Co 1332-keV {gamma} line. As a result, 1112 yields of radioactive residual nuclei have been obtained. The {sup 27}Al(p, x){sup 22}Na reaction has been used as a monitor reaction. The experimental data have been compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

Measurement and simulation of the cross sections for nuclide production in {sup 56}Fe and {sup nat}Cr targets irradiated with 0.04- to 2.6-GeV protons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Titarenko, Yu. E., E-mail: Yury.Titarenko@itep.ru; Batyaev, V. F.; Titarenko, A. Yu.

The cross sections for nuclide production in thin {sup 56}Fe and {sup nat}Cr targets irradiated by 0.04-2.6-GeV protons are measured by direct {gamma} spectrometry using two {gamma} spectrometers with the resolutions of 1.8 and 1.7 keV for the {sup 60}Co 1332-keV {gamma} line. As a result, 649 yields of radioactive residual product nuclei have been obtained. The {sup 27}Al(p, x){sup 22}Na reaction has been used as a monitor reaction. The experimental data are compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

Experimental demonstration of a compact epithermal neutron source based on a high power laser

NASA Astrophysics Data System (ADS)

Mirfayzi, S. R.; Alejo, A.; Ahmed, H.; Raspino, D.; Ansell, S.; Wilson, L. A.; Armstrong, C.; Butler, N. M. H.; Clarke, R. J.; Higginson, A.; Kelleher, J.; Murphy, C. D.; Notley, M.; Rusby, D. R.; Schooneveld, E.; Borghesi, M.; McKenna, P.; Rhodes, N. J.; Neely, D.; Brenner, C. M.; Kar, S.

2017-07-01

Epithermal neutrons from pulsed-spallation sources have revolutionised neutron science allowing scientists to acquire new insight into the structure and properties of matter. Here, we demonstrate that laser driven fast (˜MeV) neutrons can be efficiently moderated to epithermal energies with intrinsically short burst durations. In a proof-of-principle experiment using a 100 TW laser, a significant epithermal neutron flux of the order of 105 n/sr/pulse in the energy range of 0.5-300 eV was measured, produced by a compact moderator deployed downstream of the laser-driven fast neutron source. The moderator used in the campaign was specifically designed, by the help of MCNPX simulations, for an efficient and directional moderation of the fast neutron spectrum produced by a laser driven source.

A method to calculate the gamma ray detection efficiency of a cylindrical NaI (Tl) crystal

NASA Astrophysics Data System (ADS)

Ahmadi, S.; Ashrafi, S.; Yazdansetad, F.

2018-05-01

Given a wide range application of NaI(Tl) detector in industrial and medical sectors, computation of the related detection efficiency in different distances of a radioactive source, especially for calibration purposes, is the subject of radiation detection studies. In this work, a 2in both in radius and height cylindrical NaI (Tl) scintillator was used, and by changing the radial, axial, and diagonal positions of an isotropic 137Cs point source relative to the detector, the solid angles and the interaction probabilities of gamma photons with the detector's sensitive area have been calculated. The calculations present the geometric and intrinsic efficiency as the functions of detector's dimensions and the position of the source. The calculation model is in good agreement with experiment, and MCNPX simulation.

Entrance surface dose distribution and organ dose assessment for cone-beam computed tomography using measurements and Monte Carlo simulations with voxel phantoms

NASA Astrophysics Data System (ADS)

Baptista, M.; Di Maria, S.; Vieira, S.; Vaz, P.

2017-11-01

Cone-Beam Computed Tomography (CBCT) enables high-resolution volumetric scanning of the bone and soft tissue anatomy under investigation at the treatment accelerator. This technique is extensively used in Image Guided Radiation Therapy (IGRT) for pre-treatment verification of patient position and target volume localization. When employed daily and several times per patient, CBCT imaging may lead to high cumulative imaging doses to the healthy tissues surrounding the exposed organs. This work aims at (1) evaluating the dose distribution during a CBCT scan and (2) calculating the organ doses involved in this image guiding procedure for clinically available scanning protocols. Both Monte Carlo (MC) simulations and measurements were performed. To model and simulate the kV imaging system mounted on a linear accelerator (Edge™, Varian Medical Systems) the state-of-the-art MC radiation transport program MCNPX 2.7.0 was used. In order to validate the simulation results, measurements of the Computed Tomography Dose Index (CTDI) were performed, using standard PMMA head and body phantoms, with 150 mm length and a standard pencil ionizing chamber (IC) 100 mm long. Measurements for head and pelvis scanning protocols, usually adopted in clinical environment were acquired, using two acquisition modes (full-fan and half fan). To calculate the organ doses, the implemented MC model of the CBCT scanner together with a male voxel phantom ("Golem") was used. The good agreement between the MCNPX simulations and the CTDIw measurements (differences up to 17%) presented in this work reveals that the CBCT MC model was successfully validated, taking into account the several uncertainties. The adequacy of the computational model to map dose distributions during a CBCT scan is discussed in order to identify ways to reduce the total CBCT imaging dose. The organ dose assessment highlights the need to evaluate the therapeutic and the CBCT imaging doses, in a more balanced approach, and the importance of improving awareness regarding the increased risk, arising from repeated exposures.

SU-E-T-764: Track Repeating Algorithm for Proton Therapy Applied to Intensity Modulated Proton Therapy for Head-And-Neck Patients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yepes, P; Mirkovic, D; Mohan, R

Purpose: To determine the suitability of fast Monte Carlo techniques for dose calculation in particle therapy based on track-repeating algorithm for Intensity Modulated Proton Therapy, IMPT. The application of this technique will make possible detailed retrospective studies of large cohort of patients, which may lead to a better determination of Relative Biological Effects from the analysis of patient data. Methods: A cohort of six head-and-neck patients treated at the University of Texas MD Anderson Cancer Center with IMPT were utilized. The dose distributions were calculated with the standard Treatment Plan System, TPS, MCNPX, GEANT4 and FDC, a fast track-repeating algorithmmore » for proton therapy for the verification and the patient plans. FDC is based on a GEANT4 database of trajectories of protons in a water. The obtained dose distributions were compared to each other utilizing the g-index criteria for 3mm-3% and 2mm-2%, for the maximum spatial and dose differences. The γ-index was calculated for voxels with a dose at least 10% of the maximum delivered dose. Dose Volume Histograms are also calculated for the various dose distributions. Results: Good agreement between GEANT4 and FDC is found with less than 1% of the voxels with a γ-index larger than 1 for 2 mm-2%. The agreement between MCNPX with FDC is within the requirements of clinical standards, even though it is slightly worse than the comparison with GEANT4.The comparison with TPS yielded larger differences, what is also to be expected because pencil beam algorithm do not always performed well in highly inhomogeneous areas like head-and-neck. Conclusion: The good agreement between a track-repeating algorithm and a full Monte Carlo for a large cohort of patients and a challenging, site like head-and-neck, opens the path to systematic and detailed studies of large cohorts, which may yield better understanding of biological effects.« less

A Bonner Sphere Spectrometer based on a large 6LiI(Eu) scintillator: Calibration in reference monoenergetic fields

NASA Astrophysics Data System (ADS)

Bedogni, R.; Pola, A.; Costa, M.; Monti, V.; Thomas, D. J.

2018-07-01

A Bonner Sphere spectrometer employing a large, 11 mm diameter × 3 mm thickness, 6LiI(Eu) scintillator (LL-BSS), was assembled. The purpose was to produce a BSS similar in sensitivity to those based on 3He sensors, but using alternative sensors. With respect to the traditional BSS based on the 4 mm (diameter) × 4 mm (height) 6LiI(Eu), this new BSS is a factor of 3 more sensitive. LL-BSS response matrix, determined with MCNPX, was experimentally evaluated with monoenergetic reference neutron fields of 144 keV, 565 keV and 1.2 MeV available at NPL (Teddington, UK). The results of the experiment confirmed the correctness of the response matrix within an overall uncertainty lower than ±2%.

Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

PubMed

Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

2015-05-01

The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. Copyright © 2015 Elsevier Ltd. All rights reserved.

Corrigendum to "Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit".

PubMed

El-Jaby, Samy

2016-06-01

A recent paper published in Life Sciences in Space Research (El-Jaby and Richardson, 2015) presented estimates of the secondary neutron ambient and effective dose equivalent rates, in air, from surface altitudes up to suborbital altitudes and low Earth orbit. These estimates were based on MCNPX (LANL, 2011) (Monte Carlo N-Particle eXtended) radiation transport simulations of galactic cosmic radiation passing through Earth's atmosphere. During a recent review of the input decks used for these simulations, a systematic error was discovered that is addressed here. After reassessment, the neutron ambient and effective dose equivalent rates estimated are found to be 10 to 15% different, though, the essence of the conclusions drawn remains unchanged. Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

Design of a Neutron Temporal Diagnostic for measuring DD or DT burn histories at the NIF

NASA Astrophysics Data System (ADS)

Lahmann, B.; Frenje, J. A.; Sio, H.; Petrasso, R. D.; Bradley, D. K.; Le Pape, S.; MacKinnon, A. J.; Isumi, N.; Macphee, A.; Zayas, C.; Spears, B. K.; Hermann, H.; Hilsabeck, T. J.; Kilkenny, J. D.

2015-11-01

The DD or DT burn history in Inertial Confinement Fusion (ICF) implosions provides essential information about implosion performance and helps to constrain numerical modeling. The capability of measuring this burn history is thus important for the NIF in its pursuit of ignition. Currently, the Gamma Reaction History (GRH) diagnostic is the only system capable of measuring the burn history for DT implosions with yields greater than ~ 1e14. To complement GRH, a new NIF Neutron Temporal Diagnostic (NTD) is being designed for measuring the DD or DT burn history with yields greater than ~ 1e10. A traditional scintillator-based design and a pulse-dilation-based design are being considered. Using MCNPX simulations, both designs have been optimized, validated and contrasted for various types of implosions at the NIF. This work was supported in part by the U.S. DOE, LLNL and LLE.

U-238 fission and Pu-239 production in subcritical assembly

NASA Astrophysics Data System (ADS)

Grab, Magdalena; Wojciechowski, Andrzej

2018-04-01

The project touches upon an issue of U-238 fission reactions and Pu-239 production reactions in subcritical assembly. The experiment took place in November 2014 at the Dzhelepov Laboratory of Nuclear Problems (JINR, Dubna) using PHASOTRON.Data of this experiment were analyzed in Laboratory of Information Technologies (LIT). Four MCNPX models were considered for simulation: Bertini/Dresnen, Bertini/Abla, INCL4/Drensnen, INCL4/Abla. The main goal of the project was to compare the experimental data and simulation results. We obtain a good agreement of experimental data and computation results especially for detectors placed besides the assembly axis. In addition, the U-238 fission reactions are more probable to be observed in the region of a higher particle energy spectrum, located closer to the assembly axis and the particle beam as well and vice versa Pu-239 production reactions were dominant in the peripheral region of geometry.

Computational study of radiation doses at UNLV accelerator facility

NASA Astrophysics Data System (ADS)

Hodges, Matthew; Barzilov, Alexander; Chen, Yi-Tung; Lowe, Daniel

2017-09-01

A Varian K15 electron linear accelerator (linac) has been considered for installation at University of Nevada, Las Vegas (UNLV). Before experiments can be performed, it is necessary to evaluate the photon and neutron spectra as generated by the linac, as well as the resulting dose rates within the accelerator facility. A computational study using MCNPX was performed to characterize the source terms for the bremsstrahlung converter. The 15 MeV electron beam available in the linac is above the photoneutron threshold energy for several materials in the linac assembly, and as a result, neutrons must be accounted for. The angular and energy distributions for bremsstrahlung flux generated by the interaction of the 15 MeV electron beam with the linac target were determined. This source term was used in conjunction with the K15 collimators to determine the dose rates within the facility.

Photonuclear production of yttrium-88 - A high energy gamma emitter for hydrocarbon extraction applications.

PubMed

Dale, Daniel S; Starovoitova, Valeriia N; Forest, Tony A; Oliphant, Emily

2018-05-05

The use of fracing has risen over the past decade and revolutionized energy production in the US. However, there is still an impetus for further optimization of the extraction of oil and natural gas from vast shale reservoirs. In this work, we discuss photonuclear production of yttrium-88 as a promising radiotracer for fracing operations. Single neutron knock-out from natural monoisotopic yttrium-89 is an inexpensive process resulting in high activity of 88 Y with minimal impurities. MCNPX simulations were performed to estimate the 88 Y yield. Irradiations of natural yttrium using a 32 MeV electron linac equipped with a tungsten bremsstrahlung converter were done to benchmark the simulations. Activities of 88 Y, 87g Y, and 87m Y were measured and found to be in good agreement with the predictions. Copyright © 2018 Elsevier Ltd. All rights reserved.

Cosmogenic Secondary Radiation from a Nearby Supernova

NASA Astrophysics Data System (ADS)

Overholt, Andrew

2017-01-01

Increasing evidence has been found for multiple supernovae within 100 pc of the solar system. Supernovae produce large amounts of cosmic rays which upon striking Earth's atmosphere, produce a cascade of secondary particles. Among these cosmic ray secondaries are neutrons and muons, which penetrate far within the atmosphere to sea level and even below sea level. Muons and neutrons are both forms of ionizing radiation which have been linked to increases in cancer, congenital malformations, and other maladies. This work focuses on the impact of muons, as they penetrate into ocean water to impact the lowest levels of the aquatic food chain. We have used monte carlo simulations (CORSIKA, MCNPx, and FLUKA) to determine the ionizing radiation dose due to cosmic ray secondaries. This information shows that although most astrophysical events do not supply the necessary radiation flux to prove dangerous; there may be other impacts such as an increase to mutation rate.

The effect of X-ray scattering by water in the irradiation of cell cultures for the dosimetric characterization of a new prototype of IORT (Intra-Operative Radiation Therapy) device: Monte Carlo simulation and experimental validation.

PubMed

Ceccolini, E; Ferrari, P; Castelluccio, D M; Mostacci, D; Sumini, M

2013-10-01

The electron beam emitted backward by plasma focus devices is being considered as a radiation source for Intra-Operative Radiation Therapy (IORT) applications. Radiobiological investigations have been conducted to assess the potential of this new prototype of IORT device. A standard x-ray beam, ISO-H60, was used for comparison, irradiating cell cultures in a holder filled with an aqueous solution. The influence of scattering by the culture water and by the walls of the holder was investigated to determine their influence on the dose delivered to the cell culture. MCNPX simulations were run and experimental measurements conducted. The effect of scattering by the holder was found to be negligible; scattering by the culture water was determined to give an increase in dose of the order of 10%.

Computational lymphatic node models in pediatric and adult hybrid phantoms for radiation dosimetry

NASA Astrophysics Data System (ADS)

Lee, Choonsik; Lamart, Stephanie; Moroz, Brian E.

2013-03-01

We developed models of lymphatic nodes for six pediatric and two adult hybrid computational phantoms to calculate the lymphatic node dose estimates from external and internal radiation exposures. We derived the number of lymphatic nodes from the recommendations in International Commission on Radiological Protection (ICRP) Publications 23 and 89 at 16 cluster locations for the lymphatic nodes: extrathoracic, cervical, thoracic (upper and lower), breast (left and right), mesentery (left and right), axillary (left and right), cubital (left and right), inguinal (left and right) and popliteal (left and right), for different ages (newborn, 1-, 5-, 10-, 15-year-old and adult). We modeled each lymphatic node within the voxel format of the hybrid phantoms by assuming that all nodes have identical size derived from published data except narrow cluster sites. The lymph nodes were generated by the following algorithm: (1) selection of the lymph node site among the 16 cluster sites; (2) random sampling of the location of the lymph node within a spherical space centered at the chosen cluster site; (3) creation of the sphere or ovoid of tissue representing the node based on lymphatic node characteristics defined in ICRP Publications 23 and 89. We created lymph nodes until the pre-defined number of lymphatic nodes at the selected cluster site was reached. This algorithm was applied to pediatric (newborn, 1-, 5-and 10-year-old male, and 15-year-old males) and adult male and female ICRP-compliant hybrid phantoms after voxelization. To assess the performance of our models for internal dosimetry, we calculated dose conversion coefficients, called S values, for selected organs and tissues with Iodine-131 distributed in six lymphatic node cluster sites using MCNPX2.6, a well validated Monte Carlo radiation transport code. Our analysis of the calculations indicates that the S values were significantly affected by the location of the lymph node clusters and that the values increased for smaller phantoms due to the shorter inter-organ distances compared to the bigger phantoms. By testing sensitivity of S values to random sampling and voxel resolution, we confirmed that the lymph node model is reasonably stable and consistent for different random samplings and voxel resolutions.

TU-D-209-07: Monte Carlo Assessment of Dose to the Lens of the Eye of Radiologist Using Realistic Phantoms and Eyeglass Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X; Lin, H; Gao, Y

Purpose: To study how eyeglass design features and postures of the interventional radiologist affect the radiation dose to the lens of the eye. Methods: A mesh-based deformable phantom, consisting of an ultra-fine eye model, was used to simulate postures of a radiologist in fluoroscopically guided interventional procedure (facing the patient, 45 degree to the left, and 45 degree to the right). Various eyewear design features were studied, including the shape, lead-equivalent thickness, and separation from the face. The MCNPX Monte Carlo code was used to simulate the X-ray source used for the transcatheter arterial chemoembolization procedure (The X-ray tube ismore » located 35 cm from the ground, emitting X-rays toward to the ceiling; Field size is 40cm X 40cm; X-ray tube voltage is 90 kVp). Experiments were also performed using dosimeter placed on a physical phantom behind eyeglasses. Results: Without protective eyewear, the radiologist’s eye lens can receive an annual dose equivalent of about 80 mSv. When wearing a pair of lead eyeglasses with lead-equivalent of 0.5-mm Pb, the annual dose equivalent of the eye lens is reduced to 31.47 mSv, but both exceed the new ICRP limit of 20 mSv. A face shield with a lead-equivalent of 0.125-mm Pb in the shape of a semi-cylinder (13cm in radius and 20-cm in height) would further reduce the exposure to the lens of the eye. Examination of postures and eyeglass features reveal surprising information, including that the glass-to-eye separation also plays an important role in the dose to the eye lens from scattered X-ray from underneath and the side. Results are in general agreement with measurements. Conclusion: There is an urgent need to further understand the relationship between the radiation environment and the radiologist’s eyewear and posture in order to provide necessary protection to the interventional radiologists under newly reduced dose limits.« less

Neutron spectroscopy with scintillation detectors using wavelets

NASA Astrophysics Data System (ADS)

Hartman, Jessica

The purpose of this research was to study neutron spectroscopy using the EJ-299-33A plastic scintillator. This scintillator material provided a novel means of detection for fast neutrons, without the disadvantages of traditional liquid scintillation materials. EJ-299-33A provided a more durable option to these materials, making it less likely to be damaged during handling. Unlike liquid scintillators, this plastic scintillator was manufactured from a non-toxic material, making it safer to use, as well as easier to design detectors. The material was also manufactured with inherent pulse shape discrimination abilities, making it suitable for use in neutron detection. The neutron spectral unfolding technique was developed in two stages. Initial detector response function modeling was carried out through the use of the MCNPX Monte Carlo code. The response functions were developed for a monoenergetic neutron flux. Wavelets were then applied to smooth the response function. The spectral unfolding technique was applied through polynomial fitting and optimization techniques in MATLAB. Verification of the unfolding technique was carried out through the use of experimentally determined response functions. These were measured on the neutron source based on the Van de Graff accelerator at the University of Kentucky. This machine provided a range of monoenergetic neutron beams between 0.1 MeV and 24 MeV, making it possible to measure the set of response functions of the EJ-299-33A plastic scintillator detector to neutrons of specific energies. The response of a plutonium-beryllium (PuBe) source was measured using the source available at the University of Nevada, Las Vegas. The neutron spectrum reconstruction was carried out using the experimentally measured response functions. Experimental data was collected in the list mode of the waveform digitizer. Post processing of this data focused on the pulse shape discrimination analysis of the recorded response functions to remove the effects of photons and allow for source characterization based solely on the neutron response. The unfolding technique was performed through polynomial fitting and optimization techniques in MATLAB, and provided an energy spectrum for the PuBe source.

Consideration of the ICRP 2006 revised tissue weighting factors on age-dependent values of the effective dose for external photons

NASA Astrophysics Data System (ADS)

Lee, Choonsik; Lee, Choonik; Han, Eun Young; Bolch, Wesley E.

2007-01-01

The effective dose recommended by the International Commission on Radiological Protection (ICRP) is the sum of organ equivalent doses weighted by corresponding tissue weighting factors, wT. ICRP is in the process of revising its 1990 recommendations on the effective dose where new values of organs and tissue weighting factors have been proposed and published in draft form for consultation by the radiological protection community. In its 5 June 2006 draft recommendations, new organs and tissues have been introduced in the effective dose which do not exist within the 1987 Oak Ridge National Laboratory (ORNL) phantom series (e.g., salivary glands). Recently, the investigators at University of Florida have updated the series of ORNL phantoms by implementing new organ models and adopting organ-specific elemental composition and densities. In this study, the effective dose changes caused by the transition from the current recommendation of ICRP Publication 60 to the 2006 draft recommendations were investigated for external photon irradiation across the range of ICRP reference ages (newborn, 1-year, 5-year, 10-year, 15-year and adult) and for six idealized irradiation geometries: anterior-posterior (AP), posterior-anterior (PA), left-lateral (LLAT), right-lateral (RLAT), rotational (ROT) and isotropic (ISO). Organ-absorbed doses were calculated by implementing the revised ORNL phantoms in the Monte Carlo radiation transport code, MCNPX2.5, after which effective doses were calculated under the 1990 and draft 2006 evaluation schemes of the ICRP. Effective doses calculated under the 2006 draft scheme were slightly higher than estimated under ICRP Publication 60 methods for all irradiation geometries exclusive of the AP geometry where an opposite trend was observed. The effective doses of the adult phantom were more greatly affected by the change in tissue weighting factors than that seen within the paediatric members of the phantom series. Additionally, dose conversion coefficients for newly identified radiosensitive organs—salivary glands, gall bladder, heart and prostate—were reported, as well as the brain, which was originally considered in ICRP Publication 60 as a member of the remainder category of the effective dose.

Development of the two Korean adult tomographic computational phantoms for organ dosimetry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Choonsik; Lee, Choonik; Park, Sang-Hyun

2006-02-15

Following the previously developed Korean tomographic phantom, KORMAN, two additional whole-body tomographic phantoms of Korean adult males were developed from magnetic resonance (MR) and computed tomography (CT) images, respectively. Two healthy male volunteers, whose body dimensions were fairly representative of the average Korean adult male, were recruited and scanned for phantom development. Contiguous whole body MR images were obtained from one subject exclusive of the arms, while whole-body CT images were acquired from the second individual. A total of 29 organs and tissues and 19 skeletal sites were segmented via image manipulation techniques such as gray-level thresholding, region growing, andmore » manual drawing, in which each of segmented image slice was subsequently reviewed by an experienced radiologist for anatomical accuracy. The resulting phantoms, the MR-based KTMAN-1 (Korean Typical MAN-1) and the CT-based KTMAN-2 (Korean Typical MAN-2), consist of 300x150x344 voxels with a voxel resolution of 2x2x5 mm{sup 3} for both phantoms. Masses of segmented organs and tissues were calculated as the product of a nominal reference density, the prevoxel volume, and the cumulative number of voxels defining each organs or tissue. These organs masses were then compared with those of both the Asian and the ICRP reference adult male. Organ masses within both KTMAN-1 and KTMAN-2 showed differences within 40% of Asian and ICRP reference values, with the exception of the skin, gall bladder, and pancreas which displayed larger differences. The resulting three-dimensional binary file was ported to the Monte Carlo code MCNPX2.4 to calculate organ doses following external irradiation for illustrative purposes. Colon, lung, liver, and stomach absorbed doses, as well as the effective dose, for idealized photon irradiation geometries (anterior-posterior and right lateral) were determined, and then compared with data from two other tomographic phantoms (Asian and Caucasian), and stylized ORNL phantom. The armless KTMAN-1 can be applied to dosimetry for computed tomography or lateral x-ray examination, while the whole body KTMAN-2 can be used for radiation protection dosimetry.« less

Thermal limits on MV x-ray production by bremsstrahlung targets in the context of novel linear accelerators.

PubMed

Wang, Jinghui; Trovati, Stefania; Borchard, Philipp M; Loo, Billy W; Maxim, Peter G; Fahrig, Rebecca

2017-12-01

To study the impact of target geometrical and linac operational parameters, such as target material and thickness, electron beam size, repetition rate, and mean current on the ability of the radiotherapy treatment head to deliver high-dose-rate x-ray irradiation in the context of novel linear accelerators capable of higher repetition rates/duty cycle than conventional clinical linacs. The depth dose in a water phantom without a flattening filter and heat deposition in an x-ray target by 10 MeV pulsed electron beams were calculated using the Monte-Carlo code MCNPX, and the transient temperature behavior of the target was simulated by ANSYS. Several parameters that affect both the dose distribution and temperature behavior were investigated. The target was tungsten with a thickness ranging from 0 to 3 mm and a copper heat remover layer. An electron beam with full width at half maximum (FWHM) between 0 and3 mm and mean current of 0.05-2 mA was used as the primary beam at repetition rates of 100, 200, 400, and 800 Hz. For a 10 MeV electron beam with FWHM of 1 mm, pulse length of 5 μs, by using a thin tungsten target with thickness of 0.2 mm instead of 1 mm, and by employing a high repetition rate of 800 Hz instead of 100 Hz, the maximum dose rate delivered can increase two times from 0.57 to 1.16 Gy/s. In this simple model, the limiting factor on dose rate is the copper heat remover's softening temperature, which was considered to be 500°C in our study. A high dose rate can be obtained by employing thin targets together with high repetition rate electron beams enabled by novel linac designs, whereas the benefit of thin targets is marginal at conventional repetition rates. Next generation linacs used to increase dose rate need different target designs compared to conventional linacs. © 2017 American Association of Physicists in Medicine.

SU-E-J-104: Single Photon Image From PET with Insertable SPECT Collimator for Boron Neutron Capture Therapy: A Feasibility Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jung, J; Yoon, D; Suh, T

2014-06-01

Purpose: The aim of our proposed system is to confirm the feasibility of extraction of two types of images from one positron emission tomography (PET) module with an insertable collimator for brain tumor treatment during the BNCT. Methods: Data from the PET module, neutron source, and collimator was entered in the Monte Carlo n-particle extended (MCNPX) source code. The coincidence events were first compiled on the PET detector, and then, the events of the prompt gamma ray were collected after neutron emission by using a single photon emission computed tomography (SPECT) collimator on the PET. The obtaining of full widthmore » at half maximum (FWHM) values from the energy spectrum was performed to collect effective events for reconstructed image. In order to evaluate the images easily, five boron regions in a brain phantom were used. The image profiles were extracted from the region of interest (ROI) of a phantom. The image was reconstructed using the ordered subsets expectation maximization (OSEM) reconstruction algorithm. The image profiles and the receiver operating characteristic (ROC) curve were compiled for quantitative analysis from the two kinds of reconstructed image. Results: The prompt gamma ray energy peak of 478 keV appeared in the energy spectrum with a FWHM of 41 keV (6.4%). On the basis of the ROC curve in Region A to Region E, the differences in the area under the curve (AUC) of the PET and SPECT images were found to be 10.2%, 11.7%, 8.2% (center, Region C), 12.6%, and 10.5%, respectively. Conclusion: We attempted to acquire the PET and SPECT images simultaneously using only PET without an additional isotope. Single photon images were acquired using an insertable collimator on a PET detector. This research was supported by the Leading Foreign Research Institute Recruitment Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Science, Information and Communication Technologies (ICT) and Future Planning (MSIP)(Grant No.2009 00420) and the Radiation Technology R and D program (Grant No.2013M2A2A7043498), Republic of Korea.« less

Evaluation of S-values and dose distributions for {sup 90}Y, {sup 131}I, {sup 166}Ho, and {sup 188}Re in seven lobes of the rat liver

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xie Tianwu; Liu Qian; Zaidi, Habib

2012-03-15

Purpose: Rats have been widely used in radionuclide therapy research for the treatment of hepatocellular carcinoma (HCC). This has created the need to assess rat liver absorbed radiation dose. In most dose estimation studies, the rat liver is considered as a homogeneous integrated target organ with a tissue composition assumed to be similar to that of human liver tissue. However, the rat liver is composed of several lobes having different anatomical and chemical characteristics. To assess the overall impact on rat liver dose calculation, the authors use a new voxel-based rat model with identified suborgan regions of the liver. Methods:more » The liver in the original cryosectional color images was manually segmented into seven individual lobes and subsequently integrated into a voxel-based computational rat model. Photon and electron particle transport was simulated using the MCNPX Monte Carlo code to calculate absorbed fractions and S-values for {sup 90}Y, {sup 131}I, {sup 166}Ho, and {sup 188}Re for the seven liver lobes. The effect of chemical composition on organ-specific absorbed dose was investigated by changing the chemical composition of the voxel filling liver material. Radionuclide-specific absorbed doses at the voxel level were further assessed for a small spherical hepatic tumor. Results: The self-absorbed dose for different liver lobes varied depending on their respective masses. A maximum difference of 3.5% was observed for the liver self-absorbed fraction between rat and human tissues for photon energies below 100 keV. {sup 166}Ho and {sup 188}Re produce a uniformly distributed high dose in the tumor and relatively low absorbed dose for surrounding tissues. Conclusions: The authors evaluated rat liver radiation doses from various radionuclides used in HCC treatments using a realistic computational rat model. This work contributes to a better understanding of all aspects influencing radiation transport in organ-specific radiation dose evaluation for preclinical therapy studies, from tissue composition to organ morphology and activity distribution.« less

Organ dose calculations by Monte Carlo modeling of the updated VCH adult male phantom against idealized external proton exposure

NASA Astrophysics Data System (ADS)

Zhang, Guozhi; Liu, Qian; Zeng, Shaoqun; Luo, Qingming

2008-07-01

The voxel-based visible Chinese human (VCH) adult male phantom has offered a high-quality test bed for realistic Monte Carlo modeling in radiological dosimetry simulations. The phantom has been updated in recent effort by adding newly segmented organs, revising walled and smaller structures as well as recalibrating skeletal marrow distributions. The organ absorbed dose against external proton exposure was calculated at a voxel resolution of 2 × 2 × 2 mm3 using the MCNPX code for incident energies from 20 MeV to 10 GeV and for six idealized irradiation geometries: anterior-posterior (AP), posterior-anterior (PA), left-lateral (LLAT), right-lateral (RLAT), rotational (ROT) and isotropic (ISO), respectively. The effective dose on the VCH phantom was derived in compliance with the evaluation scheme for the reference male proposed in the 2007 recommendations of the International Commission on Radiological Protection (ICRP). Algorithm transitions from the revised radiation and tissue weighting factors are accountable for approximately 90% and 10% of effective dose discrepancies in proton dosimetry, respectively. Results are tabulated in terms of fluence-to-dose conversion coefficients for practical use and are compared with data from other models available in the literature. Anatomical variations between various computational phantoms lead to dose discrepancies ranging from a negligible level to 100% or more at proton energies below 200 MeV, corresponding to the spatial geometric locations of individual organs within the body. Doses show better agreement at higher energies and the deviations are mostly within 20%, to which the organ volume and mass differences should be of primary responsibility. The impact of body size on dose distributions was assessed by dosimetry of a scaled-up VCH phantom that was resized in accordance with the height and total mass of the ICRP reference man. The organ dose decreases with the directionally uniform enlargement of voxels. Potential pathways to improve the VCH phantom have also been briefly addressed. This work pertains to VCH-based systematic multi-particle dose investigations and will contribute to comparative dosimetry studies of ICRP standardized voxel phantoms in the near future.

SU-F-T-159: Monte Carlo Simulation Studies of Three-Dimensional Dose Distribution for Polymer Gel Dosimeter and Radiochromic Gel Dosimeter in a Proton Beam

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, M; Kim, G; Jung, H

Purpose: The purpose of this simulation study is to evaluate the proton detectability of gel dosimeters, and estimate the three-dimensional dose distribution of protons in the radiochromic gel and polymer gel dosimeter compared with the dose distribution in water. Methods: The commercial composition ratios of normoxic polymer gel and LCV micelle radiochromic gel were included in this simulation study. The densities of polymer and radiochromic gel were 1.024 and 1.005 g/cm3, respectively. The 50, 80 and 140 MeV proton beam energies were selected. The dose distributions of protons in the polymer and radiochromic gel were simulated using Monte Carlo radiationmore » transport code (MCNPX 2.7.0, Los Alamos Laboratory). The water equivalent depth profiles and the dose distributions of two gel dosimeters were compared for the water. Results: In case of irradiating 50, 80 and 140 MeV proton beam to water phantom, the reference Bragg-peak depths are represented at 2.22, 5.18 and 13.98 cm, respectively. The difference in the water equivalent depth is represented to about 0.17 and 0.37 cm in the radiochromic gel and polymer gel dosimeter, respectively. The proton absorbed doses in the radiochromic gel dosimeter are calculated to 2.41, 3.92 and 6.90 Gy with increment of incident proton energies. In the polymer gel dosimeter, the absorbed doses are calculated to 2.37, 3.85 and 6.78 Gy with increment of incident proton energies. The relative absorbed dose in radiochromic gel (about 0.47 %) is similar to that of water than the relative absorbed dose of polymer gel (about 2.26 %). In evaluating the proton dose distribution, we found that the dose distribution of both gel dosimeters matched that of water in most cases. Conclusion: As the dosimetry device, the radiochromic gel dosimeter has the potential particle detectability and is feasible to use for quality assurance of proton beam therapy beam.« less

First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Antoni, R.; Passard, C.; Perot, B.

2015-07-01

The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ({sup 3}He proportional counter inside the measurement cavity).more » A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the {sup 239}Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and {sup 235}U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach and the industrial feasibility of the method, which will be implemented on the industrial station for the measurement of historical wastes. (authors)« less

Gadolinium-148 and other spallation production cross section measurements for accelerator target facilities

NASA Astrophysics Data System (ADS)

Kelley, Karen Corzine

At the Los Alamos Neutron Science Center accelerator complex, protons are accelerated to 800 MeV and directed to two tungsten targets, Target 4 at the Weapons Neutron Research facility and the 1L target at the Lujan Center. The Department of Energy requires hazard classification analyses to be performed on these targets and places limits on certain radionuclide inventories in the targets to avoid characterizing the facilities as "nuclear facilities." Gadolinium-148 is a radionuclide created from the spallation of tungsten. Allowed isotopic inventories are particularly low for this isotope because it is an alpha-particle emitter with a 75-year half-life. The activity level of Gadolinium-148 is low, but it encompasses almost two-thirds of the total dose burden for the two tungsten targets based on present yield estimates. From a hazard classification standpoint, this severely limits the lifetime of these tungsten targets. The cross section is not well-established experimentally and this is the motivation for measuring the Gadolinium-148 production cross section from tungsten. In a series of experiments at the Weapons Neutron Research facility, Gadolinium-148 production was measured for 600- and 800-MeV protons on tungsten, tantalum, and gold. These experiments used 3 mum thin tungsten, tantalum, and gold foils and 10 mum thin aluminum activation foils. In addition, spallation yields were determined for many short-lived and long-lived spallation products with these foils using gamma and alpha spectroscopy and compared with predictions of the Los Alamos National Laboratory codes CEM2k+GEM2 and MCNPX. The cumulative Gadolinium-148 production cross section measured from tantalum, tungsten, and gold for incident 600-MeV protons were 15.2 +/- 4.0, 8.31 +/- 0.92, and 0.591 +/- 0.155, respectively. The average production cross sections measured at 800 MeV were 28.6 +/- 3.5, 19.4 +/- 1.8, and 3.69 +/- 0.50 for tantalum, tungsten, and gold, respectively. These cumulative measurements compared best with Bertini and were within a factor of two to three of CEM2k+GEM2.

Development of a synthetic single crystal diamond dosimeter for dose measurement of clinical proton beams

NASA Astrophysics Data System (ADS)

Moignier, Cyril; Tromson, Dominique; de Marzi, Ludovic; Marsolat, Fanny; García Hernández, Juan Carlos; Agelou, Mathieu; Pomorski, Michal; Woo, Romuald; Bourbotte, Jean-Michel; Moignau, Fabien; Lazaro, Delphine; Mazal, Alejandro

2017-07-01

The scope of this work was to develop a synthetic single crystal diamond dosimeter (SCDD-Pro) for accurate relative dose measurements of clinical proton beams in water. Monte Carlo simulations were carried out based on the MCNPX code in order to investigate and reduce the dose curve perturbation caused by the SCDD-Pro. In particular, various diamond thicknesses were simulated to evaluate the influence of the active volume thickness (e AV) as well as the influence of the addition of a front silver resin (250 µm in thickness in front of the diamond crystal) on depth-dose curves. The simulations indicated that the diamond crystal alone, with a small e AV of just 5 µm, already affects the dose at Bragg peak position (Bragg peak dose) by more than 2% with respect to the Bragg peak dose deposited in water. The optimal design that resulted from the Monte Carlo simulations consists of a diamond crystal of 1 mm in width and 150 µm in thickness with the front silver resin, enclosed by a water-equivalent packaging. This design leads to a deviation between the Bragg peak dose from the full detector modeling and the Bragg peak dose deposited in water of less than 1.2%. Based on those optimizations, an SCDD-Pro prototype was built and evaluated in broad passive scattering proton beams. The experimental evaluation led to probed SCDD-Pro repeatability, dose rate dependence and linearity, that were better than 0.2%, 0.4% (in the 1.0-5.5 Gy min-1 range) and 0.4% (for dose higher than 0.05 Gy), respectively. The depth-dose curves in the 90-160 MeV energy range, measured with the SCDD-Pro without applying any correction, were in good agreement with those measured using a commercial IBA PPC05 plane-parallel ionization chamber, differing by less than 1.6%. The experimental results confirmed that this SCDD-Pro is suitable for measurements with standard electrometers and that the depth-dose curve perturbation is negligible, with no energy dependence and no significant dose rate dependence.

Development of a synthetic single crystal diamond dosimeter for dose measurement of clinical proton beams.

PubMed

Moignier, Cyril; Tromson, Dominique; de Marzi, Ludovic; Marsolat, Fanny; García Hernández, Juan Carlos; Agelou, Mathieu; Pomorski, Michal; Woo, Romuald; Bourbotte, Jean-Michel; Moignau, Fabien; Lazaro, Delphine; Mazal, Alejandro

2017-07-07

The scope of this work was to develop a synthetic single crystal diamond dosimeter (SCDD-Pro) for accurate relative dose measurements of clinical proton beams in water. Monte Carlo simulations were carried out based on the MCNPX code in order to investigate and reduce the dose curve perturbation caused by the SCDD-Pro. In particular, various diamond thicknesses were simulated to evaluate the influence of the active volume thickness (e AV ) as well as the influence of the addition of a front silver resin (250 µm in thickness in front of the diamond crystal) on depth-dose curves. The simulations indicated that the diamond crystal alone, with a small e AV of just 5 µm, already affects the dose at Bragg peak position (Bragg peak dose) by more than 2% with respect to the Bragg peak dose deposited in water. The optimal design that resulted from the Monte Carlo simulations consists of a diamond crystal of 1 mm in width and 150 µm in thickness with the front silver resin, enclosed by a water-equivalent packaging. This design leads to a deviation between the Bragg peak dose from the full detector modeling and the Bragg peak dose deposited in water of less than 1.2%. Based on those optimizations, an SCDD-Pro prototype was built and evaluated in broad passive scattering proton beams. The experimental evaluation led to probed SCDD-Pro repeatability, dose rate dependence and linearity, that were better than 0.2%, 0.4% (in the 1.0-5.5 Gy min -1 range) and 0.4% (for dose higher than 0.05 Gy), respectively. The depth-dose curves in the 90-160 MeV energy range, measured with the SCDD-Pro without applying any correction, were in good agreement with those measured using a commercial IBA PPC05 plane-parallel ionization chamber, differing by less than 1.6%. The experimental results confirmed that this SCDD-Pro is suitable for measurements with standard electrometers and that the depth-dose curve perturbation is negligible, with no energy dependence and no significant dose rate dependence.

Use of Transportable Radiation Detection Instruments to Assess Internal Contamination from Intakes of Radionuclides Part II: Calibration Factors and ICAT Computer Program.

PubMed

Anigstein, Robert; Olsher, Richard H; Loomis, Donald A; Ansari, Armin

2016-12-01

The detonation of a radiological dispersion device or other radiological incidents could result in widespread releases of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure radiation from gamma-emitting radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for in vitro assessments. The present study derived sets of calibration factors for four instruments: the Ludlum Model 44-2 gamma scintillator, a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal; the Captus 3000 thyroid uptake probe, which contains a 5.08 × 5.08-cm NaI(Tl) crystal; the Transportable Portal Monitor Model TPM-903B, which contains two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators; and a generic instrument, such as an ionization chamber, that measures exposure rates. The calibration factors enable these instruments to be used for assessing inhaled or ingested intakes of any of four radionuclides: Co, I, Cs, and Ir. The derivations used biokinetic models embodied in the DCAL computer software system developed by the Oak Ridge National Laboratory and Monte Carlo simulations using the MCNPX radiation transport code. The three physical instruments were represented by MCNP models that were developed previously. The affected individuals comprised children of five ages who were represented by the revised Oak Ridge National Laboratory pediatric phantoms, and adult men and adult women represented by the Adult Reference Computational Phantoms described in Publication 110 of the International Commission on Radiological Protection. These calibration factors can be used to calculate intakes; the intakes can be converted to committed doses by the use of tabulated dose coefficients. These calibration factors also constitute input data to the ICAT computer program, an interactive Microsoft Windows-based software package that estimates intakes of radionuclides and cumulative and committed effective doses, based on measurements made with these instruments. This program constitutes a convenient tool for assessing intakes and doses without consulting tabulated calibration factors and dose coefficients.

USE OF TRANSPORTABLE RADIATION DETECTION INSTRUMENTS TO ASSESS INTERNAL CONTAMINATION FROM INTAKES OF RADIONUCLIDES PART II: CALIBRATION FACTORS AND ICAT COMPUTER PROGRAM

PubMed Central

Anigstein, Robert; Olsher, Richard H.; Loomis, Donald A.; Ansari, Armin

2017-01-01

The detonation of a radiological dispersion device or other radiological incidents could result in widespread releases of radioactive materials and intakes of radionuclides by affected individuals. Transportable radiation monitoring instruments could be used to measure radiation from gamma-emitting radionuclides in the body for triaging individuals and assigning priorities to their bioassay samples for in vitro assessments. The present study derived sets of calibration factors for four instruments: the Ludlum Model 44-2 gamma scintillator, a survey meter containing a 2.54 × 2.54-cm NaI(Tl) crystal; the Captus 3000 thyroid uptake probe, which contains a 5.08 × 5.08-cm NaI(Tl) crystal; the Transportable Portal Monitor Model TPM-903B, which contains two 3.81 × 7.62 × 182.9-cm polyvinyltoluene plastic scintillators; and a generic instrument, such as an ionization chamber, that measures exposure rates. The calibration factors enable these instruments to be used for assessing inhaled or ingested intakes of any of four radionuclides: 60Co, 131I, 137Cs, and 192Ir. The derivations used biokinetic models embodied in the DCAL computer software system developed by the Oak Ridge National Laboratory and Monte Carlo simulations using the MCNPX radiation transport code. The three physical instruments were represented by MCNP models that were developed previously. The affected individuals comprised children of five ages who were represented by the revised Oak Ridge National Laboratory pediatric phantoms, and adult men and adult women represented by the Adult Reference Computational Phantoms described in Publication 110 of the International Commission on Radiological Protection. These calibration factors can be used to calculate intakes; the intakes can be converted to committed doses by the use of tabulated dose coefficients. These calibration factors also constitute input data to the ICAT computer program, an interactive Microsoft Windows-based software package that estimates intakes of radionuclides and cumulative and committed effective doses, based on measurements made with these instruments. This program constitutes a convenient tool for assessing intakes and doses without consulting tabulated calibration factors and dose coefficients. PMID:27798478

Study of variation of materials patients room's door related of neutron flux iradiation

NASA Astrophysics Data System (ADS)

Nirmalasari, Yuliana Dian; Suparmi, A.; Sardjono, Y.

2017-08-01

The treatment chamber of patients has been simulating with MCNPX Code. Optimation of simulation design of Irradiation chamber is corresponding to ISO standards for 30 MeV cyclotron generator. The simulation has used the variation of door's materials that was applied at treatment room's door. The variation of materials was Stainless Steel 202 and Pb, the thickness Pb and stainless steel 202 with the thickness were 2 cm, respectively. Neutron flux that was radiated to stainless steel 202 in the sequence was 3.34195 × 105 n . Cm-2 s-1 and 8.41568 × 104 n . Cm-2 s-1, while for Pb was 4.01349 × 105 n . Cm-2 s-1 and 2.58058 × 104 n . Cm-2 s-1. The further, neutron flux that was radiated to Pb and stainless steel 202 with the thickness were 4 cm in sequence was 4.00601 × 105 n . Cm-2 s-1 and 1.71713 × 104 n . Cm-2 s-1 for Pb, while for SS 202 was 3.09925 × 105 n . Cm-2 s-1. From this ratio we concluded that material Pb absorbed higher neutron flux than material Stainless Steel 202. On the other hand, the cost of Pb was more expensive than Stainless Steel 202. In addition, the material Stainless Steel 202 was obtaine more easily than the material Pb. There fore to overcome the economics problem, can try to build the door with stainless still 202 sheet and Pb sheet together. The further, the neutron dose with 2 cm of thickness was 7.69603 × 10-2 Gy and 2.10623 × 10-2 Gy for SS 202, while for Pb was 4.19444 × 10-2 Gy and 1.50581 × 10-2 Gy. While the neutron dose with 4 cm of thickness for SS 202 was 9.39602 × 10-2 Gy and for Pb was 4.46541 × 10-2 Gy and 1.50502 × 10-2 Gy. We recommend that this simulation should be further optimized.

Design of photon converter and photoneutron target for High power electron accelerator based BNCT.

PubMed

Rahmani, Faezeh; Seifi, Samaneh; Anbaran, Hossein Tavakoli; Ghasemi, Farshad

2015-12-01

An electron accelerator, ILU-14, with current of 10 mA and 100 kW in power has been considered as one of the options for neutron source in Boron Neutron Capture Therapy (BNCT). The final design of neutron target has been obtained using MCNPX to optimize the neutron production. Tungsten in strip shape and D2O in cylindrical form have been proposed as the photon converter and the photoneutron target, respectively. In addition calculation of heat deposition in the photon target design has been considered to ensure mechanical stability of target. The results show that about 8.37×10(12) photoneutron/s with average energy of 615 keV can be produced by this neutron source design. In addition, using an appropriate beam shaping assembly an epithermal neutron flux of the order of 1.24×10(8) cm(-2) s(-1) can be obtained for BNCT applications. Copyright © 2015 Elsevier Ltd. All rights reserved.

Design of the Next Generation Target at the Lujan Neutron Scattering Center, LANSCE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferres, Laurent

Los Alamos National Laboratory (LANL) supports scientific research in many diverse fields such as biology, chemistry, and nuclear science. The Laboratory was established in 1943 during the Second World War to develop nuclear weapons. Today, LANL is one of the largest laboratories dedicated to nuclear defense and operates an 800 MeV proton linear accelerator for basic and applied research including: production of high- and low-energy neutrons beams, isotope production for medical applications and proton radiography. This accelerator is located at the Los Alamos Neutron Science Center (LANSCE). The work performed involved the redesign of the target for the low-energy neutronmore » source at the Lujan Neutron Scattering Center, which is one of the facilities built around the accelerator. The redesign of the target involves modeling various arrangements of the moderator-reflector-shield for the next generation neutron production target. This is done using Monte Carlo N-Particle eXtended (MCNPX), and ROOT analysis framework, a C++ based-software, to analyze the results.« less

Experimental studies of the medical radioisotopes production in neutron spectra generated by 660 MeV protons and 1-8 GeV deuterons in massive uranium target

NASA Astrophysics Data System (ADS)

Zhadan, A.; Sotnikov, V.; Adam, J.; Solnyshkin, A.; Tyutyunnikov, S.; Voronko, V.; Zhivkov, P.; Zavorka, L.

2017-06-01

The possibility of medical radionuclide 64,67Cu production in spallation neutron spectrum induced by proton and deuteron beams has been studied. Experiments were performed on a massive natural uranium target at the accelerators Phasotron and Nuclotron JINR, Dubna. The main disadvantage of this method is a high 64Cu/67Cu ratio in the final product at EOB. Significantly reduce 64Cu/67Cu ratio is only possible if you use zinc target enriched with 68Zn or 67Zn. The MCNPX simulation of 67,64Cu production and definition of the theoretical limit of the specific activity of 67,64Cu by irradiation of natural zinc and zinc enriched by the 68 isotope were performed. The neutron flux density shouldnot be less than 5.1013 n/cm2/s if we want to obtain high specific activity (>200 GBq/mg) of 67Cu.

A Rotating Scatter Mask for Inexpensive Gamma-Ray Imaging in Orphan Source Search: Simulation Results

NASA Astrophysics Data System (ADS)

FitzGerald, Jack G. M.

2015-02-01

The Rotating Scatter Mask (RSM) system is an inexpensive retrofit that provides imaging capabilities to scintillating detectors. Unlike traditional collimator systems that primarily absorb photons in order to form an image, this system primarily scatters the photons. Over a single rotation, there is a unique, smooth response curve for each defined source position. Testing was conducted using MCNPX simulations. Image reconstruction was performed using a chi-squared reconstruction technique. A simulated 100 uCi, Cs-137 source at 10 meters was detected after a single, 50-second rotation when a uniform terrestrial background was present. A Cs-137 extended source was also tested. The RSM field-of-view is 360 degrees azimuthally as well as 54 degrees above and 54 degrees below the horizontal plane. Since the RSM is built from polyethylene, the overall cost and weight of the system is low. The system was designed to search for lost or stolen radioactive material, also known as the orphan source problem.

Fast neutron response of 6Li-depleted CLYC detectors up to 20 MeV

NASA Astrophysics Data System (ADS)

D`Olympia, N.; Chowdhury, P.; Jackson, E. G.; Lister, C. J.

2014-11-01

The response of 6Li-depleted Cs2LiYCl6 (CLYC) to high-energy neutrons has been investigated using a pair of 1 in.×1 in. crystals. These are the first two detectors of their kind, which will comprise a 16-element array for studies in fast neutron spectroscopy. Their thermal neutron response has been compared with standard CLYC crystals with a 6Li enrichment of 95%, demonstrating excellent suppression of the overwhelming thermal neutron background. The response to mono-energetic neutrons over a range of 0.5 to 20 MeV was tested. From this, the response function, energy resolution, and pulse-shape discrimination up to 20 MeV were characterized. Detailed Monte Carlo investigations with MCNPX have been used to show that the dominant reaction mechanisms contributing to the observed response are 35Cl(n,p) and 35Cl(n,α). Preliminary investigations have also demonstrated the possibility for separating events from these two reactions.

Monte Carlo Simulation of a 12 MeV Cargo Container Inspection System

NASA Astrophysics Data System (ADS)

Ozcan, Ibrahim; Chandler, Katherine; Spaulding, Randy; Farfan, Eduardo

2007-05-01

After the terrorist events of 9/11, border security has become one of the most important issues in national security due to the large number of cargo containers entering the country. Screening of all cargo containers for nuclear materials should be performed during border inspections. The technical aspects of inspecting cargo containers using electron accelerators have been studied previously. However, the radiological protection aspects involved in these studies have not been fully considered. This screening process may accidentally harm operators, workers, and bystanders; as well as stowaways hiding inside the containers. In this research project, external doses were estimated at various locations near the inspection system. A 12-MeV linear accelerator (LINAC) was used in the experiment. The relationship between the various locations and doses were determined in this simulation. The simulation was performed using MCNPX. To cite this abstract, use the following reference: http://meetings.aps.org/link/BAPS.2007.NWS07.B2.8

Assessment of PCXMC for patients with different body size in chest and abdominal x ray examinations: a Monte Carlo simulation study.

PubMed

Borrego, David; Lowe, Erin M; Kitahara, Cari M; Lee, Choonsik

2018-03-21

A PC Program for x ray Monte Carlo (PCXMC) has been used to calculate organ doses in patient dosimetry and for the exposure assessment in epidemiological studies of radiogenic health related risks. This study compared the dosimetry from using the built-in stylized phantoms in the PCXMC to that of a newer hybrid phantom library with improved anatomical realism. We simulated chest and abdominal x ray projections for 146 unique body size computational phantoms, 77 males and 69 females, with different combinations of height (125-180 cm) and weight (20-140 kg) using the built-in stylized phantoms in the PCXMC version 2.0.1.4 and the hybrid phantom library using the Monte Carlo N-particle eXtended transport code 2.7 (MCNPX). Unfortunately, it was not possible to incorporate the hybrid phantom library into the PCXMC. We compared 14 organ doses, including dose to the active bone marrow, to evaluate differences between the built-in stylized phantoms in the PCXMC and the hybrid phantoms (Cristy and Eckerman 1987 Technical Report ORNL/TM-8381/V1, Oak Ridge National Laboratory, Eckerman and Ryman 1993 Technical Report 12 Oak Ridge, TN, Geyer et al 2014 Phys. Med. Biol. 59 5225-42). On average, organ doses calculated using the built-in stylized phantoms in the PCXMC were greater when compared to the hybrid phantoms. This is most prominent in AP abdominal exams by an average factor of 2.4-, 2.8-, and 2.8-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. For chest exams, organ doses are greater by an average factor of 1.1-, 1.4-, and 1.2-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. The PCXMX, due to its ease of use, is often selected to support dosimetry in epidemiological studies; however, it uses simplified models of the human anatomy that fail to account for variations in body morphometry for increasing weight. For epidemiological studies that use PCXMC dosimetry, associations between radiation-related disease risks and organ doses may be underestimated, and to a greater degree in pediatric, especially obese pediatric, compared to adult patients.

Computational assessment of effective dose and patient specific doses for kilovoltage stereotactic radiosurgery of wet age-related macular degeneration

NASA Astrophysics Data System (ADS)

Hanlon, Justin Mitchell

Age-related macular degeneration (AMD) is a leading cause of vision loss and a major health problem for people over the age of 50 in industrialized nations. The current standard of care, ranibizumab, is used to help slow and in some cases stabilize the process of AMD, but requires frequent invasive injections into the eye. Interest continues for stereotactic radiosurgery (SRS), an option that provides a non-invasive treatment for the wet form of AMD, through the development of the IRay(TM) (Oraya Therapeutics, Inc., Newark, CA). The goal of this modality is to destroy choroidal neovascularization beneath the pigment epithelium via delivery of three 100 kVp photon beams entering through the sclera and overlapping on the macula delivering up to 24 Gy of therapeutic dose over a span of approximately 5 minutes. The divergent x-ray beams targeting the fovea are robotically positioned and the eye is gently immobilized by a suction-enabled contact lens. Device development requires assessment of patient effective dose, reference patient mean absorbed doses to radiosensitive tissues, and patient specific doses to the lens and optic nerve. A series of head phantoms, including both reference and patient specific, was derived from CT data and employed in conjunction with the MCNPX 2.5.0 radiation transport code to simulate treatment and evaluate absorbed doses to potential tissues-at-risk. The reference phantoms were used to evaluate effective dose and mean absorbed doses to several radiosensitive tissues. The optic nerve was modeled with changeable positions based on individual patient variability seen in a review of head CT scans gathered. Patient specific phantoms were used to determine the effect of varying anatomy and gaze. The results showed that absorbed doses to the non-targeted tissues were below the threshold levels for serious complications; specifically the development of radiogenic cataracts and radiation induced optic neuropathy (RON). The effective dose determined (0.29 mSv) is comparable to diagnostic procedures involving the head, such as an x-ray or CT scan. Thus, the computational assessment performed indicates that a previously established therapeutic dose can be delivered effectively to the macula with IRay(TM) without the potential for secondary complications.

Organ doses for reference adult male and female undergoing computed tomography estimated by Monte Carlo simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Choonsik; Kim, Kwang Pyo; Long, Daniel

2011-03-15

Purpose: To develop a computed tomography (CT) organ dose estimation method designed to readily provide organ doses in a reference adult male and female for different scan ranges to investigate the degree to which existing commercial programs can reasonably match organ doses defined in these more anatomically realistic adult hybrid phantomsMethods: The x-ray fan beam in the SOMATOM Sensation 16 multidetector CT scanner was simulated within the Monte Carlo radiation transport code MCNPX2.6. The simulated CT scanner model was validated through comparison with experimentally measured lateral free-in-air dose profiles and computed tomography dose index (CTDI) values. The reference adult malemore » and female hybrid phantoms were coupled with the established CT scanner model following arm removal to simulate clinical head and other body region scans. A set of organ dose matrices were calculated for a series of consecutive axial scans ranging from the top of the head to the bottom of the phantoms with a beam thickness of 10 mm and the tube potentials of 80, 100, and 120 kVp. The organ doses for head, chest, and abdomen/pelvis examinations were calculated based on the organ dose matrices and compared to those obtained from two commercial programs, CT-EXPO and CTDOSIMETRY. Organ dose calculations were repeated for an adult stylized phantom by using the same simulation method used for the adult hybrid phantom. Results: Comparisons of both lateral free-in-air dose profiles and CTDI values through experimental measurement with the Monte Carlo simulations showed good agreement to within 9%. Organ doses for head, chest, and abdomen/pelvis scans reported in the commercial programs exceeded those from the Monte Carlo calculations in both the hybrid and stylized phantoms in this study, sometimes by orders of magnitude. Conclusions: The organ dose estimation method and dose matrices established in this study readily provides organ doses for a reference adult male and female for different CT scan ranges and technical parameters. Organ doses from existing commercial programs do not reasonably match organ doses calculated for the hybrid phantoms due to differences in phantom anatomy, as well as differences in organ dose scaling parameters. The organ dose matrices developed in this study will be extended to cover different technical parameters, CT scanner models, and various age groups.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deyglun, C.; Simony, B.; Perot, B.

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microsecondsmore » are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Safigholi, H; Soliman, A; Song, W

Purpose: To evaluate various shielding materials such as Gold (Au), Osmium (Os), Tantalum (Ta), and Tungsten (W) based alloys for use with a novel intensity modulation capable direction modulated brachytherapy (DMBT) tandem applicator for image guided cervical cancer HDR brachytherapy. Methods: The novel MRI-compatible DMBT tandem, made from nonmagnetic tungsten-alloy rod with diameter of 5.4 mm, has 6 symmetric peripheral holes of 1.3 mm diameter with 2.05 mm distance from the center for a high degree intensity modulation capacity. The 0.3 mm thickness of bio-compatible plastic tubing wraps the tandem. MCNPX was used for Monte Carlo simulations of the shieldsmore » and the mHDR Ir-192 V2 source. MC-generated 3D dose matrices of different shielding materials of Au, Os, Ta, and W with 1 mm3 resolution were imported into an in-house-coded inverse optimization planning system to evaluate 19 clinical patient plans. Prescription dose was 15Gy. All plans were normalized to receive the same HRCTV D90. Results: In general, the plan qualities for various shielding materials were similar. The OAR D2cc for bladder was very similar for Au, Os, and Ta with 11.64±2.30Gy. For W, it was very close 11.65±2.30Gy. The sigmoid D2cc was 9.82±2.46Gy for Au and Os while it was 9.84±2.48Gy for Ta and W. The rectum D2cc was 7.44±3.06Gy for Au, 7.43±3.07Gy for Os, 7.48±3.05Gy for Ta, and 7.47±3.05Gy for W. The HRCTV D98 and V100 were very close with 16.37±1.87 Gy and 97.37±1.93 Gy, on average, respectively. Conclusion: Various MRI-compatible shielding alloys were investigated for the DMBT tandem applicator. The clinical plan qualities were not significantly different among these various alloys, however. Therefore, the candidate metals (or in combination) can be used to select best alloys for MRI image guided cervical cancer brachytherapy using the novel DMBT applicator that is capable of unprecedented level of intensity modulation.« less

THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

2012-06-19

Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to amore » particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence of Gd{sub 2}O{sub 3} burnable poison on the measurement of fresh pressurized water reactor fuel. To empirically determine the response function over the range of historical and future use we have considered enrichments up to 5 wt% {sup 235}U/{sup tot}U and Gd weight fractions of up to 10 % Gd/UO{sub 2}. Parameterized correction factors are presented.« less

Monte Carlo modeling of a 6 and 18 MV Varian Clinac medical accelerator for in-field and out-of-field dose calculations: development and validation

PubMed Central

Bednarz, Bryan; Xu, X George

2012-01-01

There is a serious and growing concern about the increased risk of radiation-induced second cancers and late tissue injuries associated with radiation treatment. To better understand and to more accurately quantify non-target organ doses due to scatter and leakage radiation from medical accelerators, a detailed Monte Carlo model of the medical linear accelerator is needed. This paper describes the development and validation of a detailed accelerator model of the Varian Clinac operating at 6 and 18 MV beam energies. Over 100 accelerator components have been defined and integrated using the Monte Carlo code MCNPX. A series of in-field and out-of-field dose validation studies were performed. In-field dose distributions calculated using the accelerator models were tuned to match measurement data that are considered the de facto ‘gold standard’ for the Varian Clinac accelerator provided by the manufacturer. Field sizes of 4 cm × 4 cm, 10 cm × 10 cm, 20 cm × 20 cm and 40 cm × 40 cm were considered. The local difference between calculated and measured dose on the percent depth dose curve was less than 2% for all locations. The local difference between calculated and measured dose on the dose profile curve was less than 2% in the plateau region and less than 2 mm in the penumbra region for all locations. Out-of-field dose profiles were calculated and compared to measurement data for both beam energies for field sizes of 4 cm × 4 cm, 10 cm × 10 cm and 20 cm × 20 cm. For all field sizes considered in this study, the average local difference between calculated and measured dose for the 6 and 18 MV beams was 14 and 16%, respectively. In addition, a method for determining neutron contamination in the 18 MV operating model was validated by comparing calculated in-air neutron fluence with reported calculations and measurements. The average difference between calculated and measured neutron fluence was 20%. As one of the most detailed accelerator models for both in-field and out-of-field dose calculations, the model will be combined with anatomically realistic computational patient phantoms into a computational framework to calculate non-target organ doses to patients from various radiation treatment plans. PMID:19141879

The development, validation and application of a multi-detector CT (MDCT) scanner model for assessing organ doses to the pregnant patient and the fetus using Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Gu, J.; Bednarz, B.; Caracappa, P. F.; Xu, X. G.

2009-05-01

The latest multiple-detector technologies have further increased the popularity of x-ray CT as a diagnostic imaging modality. There is a continuing need to assess the potential radiation risk associated with such rapidly evolving multi-detector CT (MDCT) modalities and scanning protocols. This need can be met by the use of CT source models that are integrated with patient computational phantoms for organ dose calculations. Based on this purpose, this work developed and validated an MDCT scanner using the Monte Carlo method, and meanwhile the pregnant patient phantoms were integrated into the MDCT scanner model for assessment of the dose to the fetus as well as doses to the organs or tissues of the pregnant patient phantom. A Monte Carlo code, MCNPX, was used to simulate the x-ray source including the energy spectrum, filter and scan trajectory. Detailed CT scanner components were specified using an iterative trial-and-error procedure for a GE LightSpeed CT scanner. The scanner model was validated by comparing simulated results against measured CTDI values and dose profiles reported in the literature. The source movement along the helical trajectory was simulated using the pitch of 0.9375 and 1.375, respectively. The validated scanner model was then integrated with phantoms of a pregnant patient in three different gestational periods to calculate organ doses. It was found that the dose to the fetus of the 3 month pregnant patient phantom was 0.13 mGy/100 mAs and 0.57 mGy/100 mAs from the chest and kidney scan, respectively. For the chest scan of the 6 month patient phantom and the 9 month patient phantom, the fetal doses were 0.21 mGy/100 mAs and 0.26 mGy/100 mAs, respectively. The paper also discusses how these fetal dose values can be used to evaluate imaging procedures and to assess risk using recommendations of the report from AAPM Task Group 36. This work demonstrates the ability of modeling and validating an MDCT scanner by the Monte Carlo method, as well as assessing fetal and organ doses by combining the MDCT scanner model and the pregnant patient phantom.

An intrinsically safe facility for forefront research and training on nuclear technologies — General description of the system

NASA Astrophysics Data System (ADS)

Mansani, L.; Bruzzone, M.; Frambati, S.; Reale, M.

2014-04-01

In the framework of research on generation-IV reactors, it is very important to have infrastructures specifically dedicated to the study of fundamental parameters in dynamics and kinetics of future fast-neutron reactors. Among various options pursued by international groups, Italy focused on lead-cooled reactors, which guarantee minimal neutron slowdown and capture and efficient cooling. In this paper it is described the design of a the low-power prototype generator, LEADS, that could be used within research facilities such as the National Laboratory of Legnaro of the INFN. The LEADS has a high safety standard in order to be used as a training facility, but it has also a good flexibility so as to allow a wide range of measurements and experiments. A high safety standard is achieved by limiting the reactor power to less than few hundred kW and the neutron multiplication factor k eff to less than 0.95 (a limiting value for spent fuel pool), by using a pure-uranium fuel (no plutonium) and by using solid lead as a diffuser. The proposed core is therefore intrinsically subcritical and has to be driven by an external neutron source generated by a proton beam impinging in a target. Preliminary simulations, performed with the MCNPX code indicated, for a 0.75mA continuous proton beam current at 70MeV proton energy, a reactor power of about 190kW when using a beryllium converter. The enriched-uranium fuel elements are immersed in a solid-lead matrix and contained within a steel vessel. The system is cooled by helium gas, which is transparent to neutrons and does not undergo activation. The gas is pumped by a compressor through specific holes at the entrance of the active volume with a temperature which varies according to the operating conditions and a pressure of about 1.1MPa. The hot gas coming out of the vessel is cooled by an external helium-water heat exchanger. The beryllium converter is cooled by its dedicated helium gas cooling system. After shutdown, the decay is completely dissipated by conduction through the lead reflector and steel vessel, and then evacuated by irradiation from the vessel surface to the external ambient air.

Assessment of PCXMC for patients with different body size in chest and abdominal x ray examinations: a Monte Carlo simulation study

NASA Astrophysics Data System (ADS)

Borrego, David; Lowe, Erin M.; Kitahara, Cari M.; Lee, Choonsik

2018-03-01

A PC Program for x ray Monte Carlo (PCXMC) has been used to calculate organ doses in patient dosimetry and for the exposure assessment in epidemiological studies of radiogenic health related risks. This study compared the dosimetry from using the built-in stylized phantoms in the PCXMC to that of a newer hybrid phantom library with improved anatomical realism. We simulated chest and abdominal x ray projections for 146 unique body size computational phantoms, 77 males and 69 females, with different combinations of height (125–180 cm) and weight (20–140 kg) using the built-in stylized phantoms in the PCXMC version 2.0.1.4 and the hybrid phantom library using the Monte Carlo N-particle eXtended transport code 2.7 (MCNPX). Unfortunately, it was not possible to incorporate the hybrid phantom library into the PCXMC. We compared 14 organ doses, including dose to the active bone marrow, to evaluate differences between the built-in stylized phantoms in the PCXMC and the hybrid phantoms (Cristy and Eckerman 1987 Technical Report ORNL/TM-8381/V1, Oak Ridge National Laboratory, Eckerman and Ryman 1993 Technical Report 12 Oak Ridge, TN, Geyer et al 2014 Phys. Med. Biol. 59 5225–42). On average, organ doses calculated using the built-in stylized phantoms in the PCXMC were greater when compared to the hybrid phantoms. This is most prominent in AP abdominal exams by an average factor of 2.4-, 2.8-, and 2.8-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. For chest exams, organ doses are greater by an average factor of 1.1-, 1.4-, and 1.2-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. The PCXMX, due to its ease of use, is often selected to support dosimetry in epidemiological studies; however, it uses simplified models of the human anatomy that fail to account for variations in body morphometry for increasing weight. For epidemiological studies that use PCXMC dosimetry, associations between radiation-related disease risks and organ doses may be underestimated, and to a greater degree in pediatric, especially obese pediatric, compared to adult patients.

Revision of orthovoltage chest wall treatment using Monte Carlo simulations.

PubMed

Zeinali-Rafsanjani, B; Faghihi, R; Mosleh-Shirazi, M A; Mosalaei, A; Hadad, K

2017-01-01

Given the high local control rates observed in breast cancer patients undergoing chest wall irradiation by kilovoltage x-rays, we aimed to revisit this treatment modality by accurate calculation of dose distributions using Monte Carlo simulation. The machine components were simulated using the MCNPX code. This model was used to assess the dose distribution of chest wall kilovoltage treatment in different chest wall thicknesses and larger contour or fat patients in standard and mid sternum treatment plans. Assessments were performed at 50 and 100 cm focus surface distance (FSD) and different irradiation angles. In order to evaluate different plans, indices like homogeneity index, conformity index, the average dose of heart, lung, left anterior descending artery (LAD) and percentage target coverage (PTC) were used. Finally, the results were compared with the indices provided by electron therapy which is a more routine treatment of chest wall. These indices in a medium chest wall thickness in standard treatment plan at 50 cm FSD and 15 degrees tube angle was as follows: homogeneity index 2.57, conformity index 7.31, average target dose 27.43 Gy, average dose of heart, lung and LAD, 1.03, 2.08 and 1.60 Gy respectively and PTC 11.19%. Assessments revealed that dose homogeneity in planning target volume (PTV) and conformity between the high dose region and PTV was poor. To improve the treatment indices, the reference point was transferred from the chest wall skin surface to the center of PTV. The indices changed as follows: conformity index 7.31, average target dose 60.19 Gy, the average dose of heart, lung and LAD, 3.57, 6.38 and 5.05 Gy respectively and PTC 55.24%. Coverage index of electron therapy was 89% while it was 22.74% in the old orthovoltage method and also the average dose of the target was about 50 Gy but in the given method it was almost 30 Gy. The results of the treatment study show that the optimized standard and mid sternum treatment for different chest wall thicknesses is with 50 cm FSD and zero (vertical) tube angle, while in large contour patients, it is with 100 cm FSD and zero tube angle. Finally, chest wall kilovoltage and electron therapies were compared, which revealed that electron therapy produces a better dose distribution than kilovoltage therapy.

Skeletal dosimetry in a voxel-based rat phantom for internal exposures to photons and electrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xie Tianwu; Han Dao; Liu Yang

2010-05-15

Purpose: The skeleton makes a significant contribution to the whole body absorbed dose evaluation of rats, since the bone marrow and bone surface in the skeleton express high radiosensitivity and are considered to be important dose-limiting tissues. The bone marrow can be categorized as red bone marrow (RBM) and yellow bone marrow (YBM). It is important to investigate the bone marrow in skeletal dosimetry. Methods: Cryosectional color images of the skeleton of a 156 g rat were segmented into mineral bone (including cortical bone and trabecular bone), RBM, and YBM. These three tissue types were identified at 40 different bonemore » sites and integrated into a previously developed voxel-based rat computational phantom. Photon and electron skeletal absorbed fractions were then calculated using the MCNPX Monte Carlo code. Results: Absorbed fraction (AF) and specific absorbed fraction (SAF) for mineral bone, RBM, and YBM at the 40 different bone sites were established for monoenergetic photon and electron sources placed in 18 organs and seven bone sites. Discrete photon energy was varied from 0.01 to 5.0 MeV in 21 discrete steps, while 21 discrete electron energies were studied, from 0.1 to 10.0 MeV. The trends and values found were consistent with the results of other researchers [M. G. Stabin, T. E. Peterson, G. E. Holburn, and M. A. Emmons, ''Voxel-based mouse and rat models for internal dose calculations,'' J. Nucl. Med. 47, 655-659 (2006)]. S-factors for the radionuclides {sup 169}Er, {sup 143}Pr, {sup 89}Sr, {sup 32}P, and {sup 90}Y, located in 18 organs and seven bone sites for the skeleton, were calculated and are provided in detail. Conclusions: For internal dose calculations, the AF data reveal that the mineral bone in the rat skeletal system is responsible for significant attenuation of gamma rays, especially at low energies. The photon SAF curves of RBM show that, for photon energies greater than 0.6 MeV, there is an increase in secondary photons emitted from the mineral bone as photon energy increases. The SAF values calculated in this study can also be used to determine the absorbed dose to the skeletal system of rats. The S-factors generated here will be useful in preclinical targeted radiotherapy experiments.« less

Xenon-induced power oscillations in a generic small modular reactor

NASA Astrophysics Data System (ADS)

Kitcher, Evans Damenortey

As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a preliminary investigation into excess reactivity control options for the SMR design was conducted confirming the generally held notion that existing PWR control mechanisms can be used in iPWR SMRs with similar effectiveness. With the desire to operate the SMR under the boron free coolant condition, erbium oxide fuel integral burnable absorber rods were identified as a possible means to retain the dispersed absorber effect of soluble boron in the reactor coolant in replacement.

Multiplication and Presence of Shielding Material from Time-Correlated Pulse-Height Measurements of Subcritical Plutonium Assemblies

DOE PAGES

Monterial, Mateusz; Marleau, Peter; Paff, Marc; ...

2017-01-20

Here, we present the results from the first measurements of the Time-Correlated Pulse-Height (TCPH) distributions from 4.5 kg sphere of α-phase weapons-grade plutonium metal in five configurations: bare, reflected by 1.27 cm and 2.54 cm of tungsten, and 2.54 cm and 7.62 cm of polyethylene. A new method for characterizing source multiplication and shielding configuration is also demonstrated. The method relies on solving for the underlying fission chain timing distribution that drives the spreading of the measured TCPH distribution. We found that a gamma distribution fits the fission chain timing distribution well and that the fit parameters correlate with bothmore » multiplication (rate parameter) and shielding material types (shape parameter). The source-to-detector distance was another free parameter that we were able to optimize, and proved to be the most well constrained parameter. MCNPX-PoliMi simulations were used to complement the measurements and help illustrate trends in these parameters and their relation to multiplication and the amount and type of material coupled to the subcritical assembly.« less

Investigation of non-uniformity and inclusions in 6LiInSe2 utilizing laser induced breakdown spectroscopy (LIBS)

NASA Astrophysics Data System (ADS)

Wiggins, Brenden; Tupitsyn, Eugene; Bhattacharya, Pijush; Rowe, Emmanuel; Lukosi, Eric; Chvala, Ondrej; Burger, Arnold; Stowe, Ashley

2013-09-01

Impurity analysis and compositional distribution studies have been conducted on a crystal of LiInSe2, a compound semiconductor which recently has been shown to respond to ionizing radiation. IR microscopy and laser induced breakdown spectroscopy (LIBS) revealed the presence of inclusions within the crystal lattice. These precipitates were revealed to be alkali and alkaline earth elemental impurities with non-uniform spatial distribution in the crystal. LIBS compositional maps correlate the presence of these impurities with visual color differences in the crystal as well as a significant shift of the band gap. Further, LIBS revealed variation in the ratio of I-III-VI2 elemental constituents throughout the crystal. Analysis of compositional variation and impurities will aid in discerning optimal synthesis and crystal growth parameters to maximize the mobility-lifetime product and charge collection efficiency in the LiInSe2 crystal. Preliminary charge trapping calculations have also been conducted with the Monte Carlo N-particle eXtended (MCNPx) package indicating preferential trapping of holes during irradiation with thermal neutrons.

A study on leakage radiation dose at ELV-4 electron accelerator bunker

NASA Astrophysics Data System (ADS)

Chulan, Mohd Rizal Md; Yahaya, Redzuwan; Ghazali, Abu BakarMhd

2014-09-01

Shielding is an important aspect in the safety of an accelerator and the most important aspects of a bunker shielding is the door. The bunker's door should be designed properly to minimize the leakage radiation and shall not exceed the permitted limit of 2.5μSv/hr. In determining the leakage radiation dose that passed through the door and gaps between the door and the wall, 2-dimensional manual calculations are often used. This method is hard to perform because visual 2-dimensional is limited and is also very difficult in the real situation. Therefore estimation values are normally performed. In doing so, the construction cost would be higher because of overestimate or underestimate which require costly modification to the bunker. Therefore in this study, two methods are introduced to overcome the problem such as simulation using MCNPX Version 2.6.0 software and manual calculation using 3-dimensional model from Autodesk Inventor 2010 software. The values from the two methods were eventually compared to the real values from direct measurements using Ludlum Model 3 with Model 44-9 probe survey meter.

Photonuclear activation of pure isotopic mediums.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grohman, Mark A.; Lukosi, Eric Daniel

2010-06-01

This work simulated the response of idealized isotopic U-235, U-238, Th-232, and Pu-239 mediums to photonuclear activation with various photon energies. These simulations were conducted using MCNPX version 2.6.0. It was found that photon energies between 14-16 MeV produce the highest response with respect to neutron production rates from all photonuclear reactions. In all cases, Pu-239 responds the highest, followed by U-238. Th-232 produces more overall neutrons at lower photon energies then U-235 when material thickness is above 3.943 centimeters. The time it takes each isotopic material to reach stable neutron production rates in time is directly proportional to themore » material thickness and stopping power of the medium, where thicker mediums take longer to reach stable neutron production rates and thinner media display a neutron production plateau effect, due to the lack of significant attenuation of the activating photons in the isotopic mediums. At this time, no neutron sensor system has time resolutions capable of verifying these simulations, but various indirect methods are possible and should be explored for verification of these results.« less

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

Multiplication and Presence of Shielding Material from Time-Correlated Pulse-Height Measurements of Subcritical Plutonium Assemblies

NASA Astrophysics Data System (ADS)

Monterial, Mateusz; Marleau, Peter; Paff, Marc; Clarke, Shaun; Pozzi, Sara

2017-04-01

We present the results from the first measurements of the Time-Correlated Pulse-Height (TCPH) distributions from 4.5 kg sphere of α-phase weapons-grade plutonium metal in five configurations: bare, reflected by 1.27 cm and 2.54 cm of tungsten, and 2.54 cm and 7.62 cm of polyethylene. A new method for characterizing source multiplication and shielding configuration is also demonstrated. The method relies on solving for the underlying fission chain timing distribution that drives the spreading of the measured TCPH distribution. We found that a gamma distribution fits the fission chain timing distribution well and that the fit parameters correlate with both multiplication (rate parameter) and shielding material types (shape parameter). The source-to-detector distance was another free parameter that we were able to optimize, and proved to be the most well constrained parameter. MCNPX-PoliMi simulations were used to complement the measurements and help illustrate trends in these parameters and their relation to multiplication and the amount and type of material coupled to the subcritical assembly.

Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

DOE PAGES

Navarro, Jorge; Ring, Terry A.; Nigg, David W.

2015-03-01

A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibrationmore » curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.« less

Study of 232Th(n, γ) and 232Th(n,f) reaction rates in a graphite moderated spallation neutron field produced by 1.6 GeV deuterons on lead target

NASA Astrophysics Data System (ADS)

Asquith, N. L.; Hashemi-Nezhad, S. R.; Westmeier, W.; Zhuk, I.; Tyutyunnikov, S.; Adam, J.

2015-02-01

The Gamma-3 assembly of the Joint Institute for Nuclear Research (JINR), Dubna, Russia is designed to emulate the neutron spectrum of a thermal Accelerator Driven System (ADS). It consists of a lead spallation target surrounded by reactor grade graphite. The target was irradiated with 1.6 GeV deuterons from the Nuclotron accelerator and the neutron capture and fission rate of 232Th in several locations within the assembly were experimentally measured. 232Th is a proposed fuel for envisaged Accelerator Driven Systems and these two reactions are fundamental to the performance and feasibility of 232Th in an ADS. The irradiation of the Gamma-3 assembly was also simulated using MCNPX 2.7 with the INCL4 intra-nuclear cascade and ABLA fission/evaporation models. Good agreement between the experimentally measured and calculated reaction rates was found. This serves as a good validation for the computational models and cross section data used to simulate neutron production and transport of spallation neutrons within a thermal ADS.

Internal photon and electron dosimetry of the newborn patient—a hybrid computational phantom study

NASA Astrophysics Data System (ADS)

Wayson, Michael; Lee, Choonsik; Sgouros, George; Treves, S. Ted; Frey, Eric; Bolch, Wesley E.

2012-03-01

Estimates of radiation absorbed dose to organs of the nuclear medicine patient are a requirement for administered activity optimization and for stochastic risk assessment. Pediatric patients, and in particular the newborn child, represent that portion of the patient population where such optimization studies are most crucial owing to the enhanced tissue radiosensitivities and longer life expectancies of this patient subpopulation. In cases where whole-body CT imaging is not available, phantom-based calculations of radionuclide S values—absorbed dose to a target tissue per nuclear transformation in a source tissue—are required for dose and risk evaluation. In this study, a comprehensive model of electron and photon dosimetry of the reference newborn child is presented based on a high-resolution hybrid-voxel phantom from the University of Florida (UF) patient model series. Values of photon specific absorbed fraction (SAF) were assembled for both the reference male and female newborn using the radiation transport code MCNPX v2.6. Values of electron SAF were assembled in a unique and time-efficient manner whereby the collisional and radiative components of organ dose--for both self- and cross-dose terms—were computed separately. Dose to the newborn skeletal tissues were assessed via fluence-to-dose response functions reported for the first time in this study. Values of photon and electron SAFs were used to assemble a complete set of S values for some 16 radionuclides commonly associated with molecular imaging of the newborn. These values were then compared to those available in the OLINDA/EXM software. S value ratios for organ self-dose ranged from 0.46 to 1.42, while similar ratios for organ cross-dose varied from a low of 0.04 to a high of 3.49. These large discrepancies are due in large part to the simplistic organ modeling in the stylized newborn model used in the OLINDA/EXM software. A comprehensive model of internal dosimetry is presented in this study for the newborn nuclear medicine patient based upon the UF hybrid computational phantom. Photon dose response functions, photon and electron SAFs, and tables of radionuclide S values for the newborn child--both male and female--are given in a series of four electronic annexes available at stacks.iop.org/pmb/57/1433/mmedia. These values can be applied to optimization studies of image quality and stochastic risk for this most vulnerable class of pediatric patients.

RPI-AM and RPI-AF, a pair of mesh-based, size-adjustable adult male and female computational phantoms using ICRP-89 parameters and their calculations for organ doses from monoenergetic photon beams

NASA Astrophysics Data System (ADS)

Zhang, Juying; Hum Na, Yong; Caracappa, Peter F.; Xu, X. George

2009-10-01

This paper describes the development of a pair of adult male and adult female computational phantoms that are compatible with anatomical parameters for the 50th percentile population as specified by the International Commission on Radiological Protection (ICRP). The phantoms were designed entirely using polygonal mesh surfaces—a Boundary REPresentation (BREP) geometry that affords the ability to efficiently deform the shape and size of individual organs, as well as the body posture. A set of surface mesh models, from Anatomium™ 3D P1 V2.0, including 140 organs (out of 500 available) was adopted to supply the basic anatomical representation at the organ level. The organ masses were carefully adjusted to agree within 0.5% relative error with the reference values provided in the ICRP Publication 89. The finalized phantoms have been designated the RPI adult male (RPI-AM) and adult female (RPI-AF) phantoms. For the purposes of organ dose calculations using the MCNPX Monte Carlo code, these phantoms were subsequently converted to voxel formats. Monoenergetic photons between 10 keV and 10 MeV in six standard external photon source geometries were considered in this study: four parallel beams (anterior-posterior, posterior-anterior, left lateral and right lateral), one rotational and one isotropic. The results are tabulated as fluence-to-organ-absorbed-dose conversion coefficients and fluence-to-effective-dose conversion coefficients and compared against those derived from the ICRP computational phantoms, REX and REGINA. A general agreement was found for the effective dose from these two sets of phantoms for photon energies greater than about 300 keV. However, for low-energy photons and certain individual organs, the absorbed doses exhibit profound differences due to specific anatomical features. For example, the position of the arms affects the dose to the lung by more than 20% below 300 keV in the lateral source directions, and the vertical position of the testes affects the dose by more than 80% below 150 keV in the PA source direction. The deformability and adjustability of organs and posture in the RPI adult phantoms may prove useful not only for average workers or patients for radiation protection purposes, but also in studies involving anatomical and posture variability that is important in future radiation protection dosimetry.

MCNPX evaluation of gamma spectrometry results in high radon concentration areas.

PubMed

Thinová, L; Solc, J

2014-07-01

The radon concentration in underground workplaces may reach tens of thousands of Bq m(-3). A simple MCNPXTM Monte Carlo (MC) model of a cave was developed to estimate the influence of radon on the in situ gamma spectrometry results in various geometries and radon concentrations. The detector total count rate was obtained as the sum of the individual count rates due to 214Bi in the air, radon in the walls and deposition of radon daughters on surfaces. The MC model was then modified and used in the natural conditions of the Mladeč Caves, Czech Republic. The content of 226Ra was calculated from laboratory gamma spectrometry measurements, and the concentrations of unattached and attached 214Bi were measured using the FRITRA4 device (SMM-Prague). We present a comparison of the experimental results with results calculated by the MCNPXTM model of the Gamma Surveyor spectrometry probe (GF Instruments) with a 3″×3″ NaI(Tl) detector and a 2″×2″ BGO detector. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

NASA Astrophysics Data System (ADS)

de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

Monte Carlo estimation of radiation dose in organs of female and male adult phantoms due to FDG-F18 absorbed in the lungs

NASA Astrophysics Data System (ADS)

Belinato, Walmir; Santos, William S.; Silva, Rogério M. V.; Souza, Divanizia N.

2014-03-01

The determination of dose conversion factors (S values) for the radionuclide fluorodeoxyglucose (18F-FDG) absorbed in the lungs during a positron emission tomography (PET) procedure was calculated using the Monte Carlo method (MCNPX version 2.7.0). For the obtained dose conversion factors of interest, it was considered a uniform absorption of radiopharmaceutical by the lung of a healthy adult human. The spectrum of fluorine was introduced in the input data file for the simulation. The simulation took place in two adult phantoms of both sexes, based on polygon mesh surfaces called FASH and MASH with anatomy and posture according to ICRP 89. The S values for the 22 internal organs/tissues, chosen from ICRP No. 110, for the FASH and MASH phantoms were compared with the results obtained from a MIRD V phantoms called ADAM and EVA used by the Committee on Medical Internal Radiation Dose (MIRD). We observed variation of more than 100% in S values due to structural anatomical differences in the internal organs of the MASH and FASH phantoms compared to the mathematical phantom.

Comparison between proton boron fusion therapy (PBFT) and boron neutron capture therapy (BNCT): a monte carlo study.

PubMed

Jung, Joo-Young; Yoon, Do-Kun; Barraclough, Brendan; Lee, Heui Chang; Suh, Tae Suk; Lu, Bo

2017-06-13

The aim of this study is to compare between proton boron fusion therapy (PBFT) and boron neutron capture therapy (BNCT) and to analyze dose escalation using a Monte Carlo simulation. We simulated a proton beam passing through the water with a boron uptake region (BUR) in MCNPX. To estimate the interaction between neutrons/protons and borons by the alpha particle, the simulation yielded with a variation of the center of the BUR location and proton energies. The variation and influence about the alpha particle were observed from the percent depth dose (PDD) and cross-plane dose profile of both the neutron and proton beams. The peak value of the maximum dose level when the boron particle was accurately labeled at the region was 192.4% among the energies. In all, we confirmed that prompt gamma rays of 478 keV and 719 keV were generated by the nuclear reactions in PBFT and BNCT, respectively. We validated the dramatic effectiveness of the alpha particle, especially in PBFT. The utility of PBFT was verified using the simulation and it has a potential for application in radiotherapy.

A model of tungsten anode x-ray spectra.

PubMed

Hernández, G; Fernández, F

2016-08-01

A semiempirical model for x-ray production in tungsten thick-targets was evaluated using a new characterization of electron fluence. Electron fluence is modeled taking into account both the energy and angular distributions, each of them adjusted to Monte Carlo simulated data. Distances were scaled by the CSDA range to reduce the energy dependence. Bremsstrahlung production was found by integrating the cross section with the fluence in a 1D penetration model. Characteristic radiation was added using a semiempirical law whose validity was checked. The results were compared the experimental results of Bhat et al., with the SpekCalc numerical tool, and with mcnpx simulation results from the work of Hernandez and Boone. The model described shows better agreement with the experimental results than the SpekCalc predictions in the sense of area between the spectra. A general improvement of the predictions of half-value layers is also found. The results are also in good agreement with the simulation results in the 50-640 keV energy range. A complete model for x-ray production in thick bremsstrahlung targets has been developed, improving the results of previous works and extending the energy range covered to the 50-640 keV interval.

Project Report on Development of a Safeguards Approach for Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert Bean

The Idaho National Laboratory has undertaken an effort to develop a standard safeguards approach for international commercial pyroprocessing facilities. This report details progress for the fiscal year 2010 effort. A component by component diversion pathway analysis has been performed, and has led to insight on the mitigation needs and equipment development needed for a valid safeguards approach. The effort to develop an in-hot cell detection capability led to the digital cloud chamber, and more importantly, the significant potential scientific breakthrough of the inverse spectroscopy algorithm, including the ability to identify energy and spatial location of gamma ray emitting sources withmore » a single, non-complex, stationary radiation detector system. Curium measurements were performed on historical and current samples at the FCF to attempt to determine the utility of using gross neutron counting for accountancy measurements. A solid cost estimate of equipment installation at FCF has been developed to guide proposals and cost allocations to use FCF as a test bed for safeguards measurement demonstrations. A combined MATLAB and MCNPX model has been developed to perform detector placement calculations around the electrorefiner. Early harvesting has occurred wherein the project team has been requested to provide pyroprocessing technology and safeguards short courses.« less

Comparison between proton boron fusion therapy (PBFT) and boron neutron capture therapy (BNCT): a Monte Carlo study

PubMed Central

Barraclough, Brendan; Lee, Heui Chang; Suh, Tae Suk; Lu, Bo

2017-01-01

The aim of this study is to compare between proton boron fusion therapy (PBFT) and boron neutron capture therapy (BNCT) and to analyze dose escalation using a Monte Carlo simulation. We simulated a proton beam passing through the water with a boron uptake region (BUR) in MCNPX. To estimate the interaction between neutrons/protons and borons by the alpha particle, the simulation yielded with a variation of the center of the BUR location and proton energies. The variation and influence about the alpha particle were observed from the percent depth dose (PDD) and cross-plane dose profile of both the neutron and proton beams. The peak value of the maximum dose level when the boron particle was accurately labeled at the region was 192.4% among the energies. In all, we confirmed that prompt gamma rays of 478 keV and 719 keV were generated by the nuclear reactions in PBFT and BNCT, respectively. We validated the dramatic effectiveness of the alpha particle, especially in PBFT. The utility of PBFT was verified using the simulation and it has a potential for application in radiotherapy. PMID:28427153

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE PAGES

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-06-13

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

Simulations of Multi-Gamma Coincidences From Neutron-Induced Fission in Special Nuclear Materials

NASA Astrophysics Data System (ADS)

Kane, Steven; Gozani, Tsahi; King, Michael J.; Kwong, John; Brown, Craig; Gary, Charles; Firestone, Murray I.; Nikkel, James A.; McKinsey, Daniel N.

2013-04-01

A study is presented on the detection of illicit special nuclear materials (SNM) in cargo containers using a conceptual neutron-based inspection system with xenon-doped liquefied argon (LAr(Xe)) scintillation detectors for coincidence gamma-ray detection. For robustness, the system is envisioned to exploit all fission signatures, namely both prompt and delayed neutron and gamma emissions from fission reactions induced in SNM. However, this paper focuses exclusively on the analysis of the prompt gamma ray emissions. The inspection system probes a container using neutrons produced either by (d, D) or (d, T) in pulsed form or from an associated particle neutron generator to exploit the associated particle imaging (API) technique, thereby achieving background reduction and imaging. Simulated signal and background estimates were obtained in MCNPX (2.7) for a 2 kg sphere of enriched uranium positioned at the center of a 1m × 1m × 1m container, which is filled uniformly with wood or iron cargos at 0.1 g/cc or 0.4 g/cc. Detection time estimates are reported assuming probabilities of detection of 95% and false alarm of 0.5%.

Assessment of physician and patient (child and adult) equivalent doses during renal angiography by Monte Carlo method.

PubMed

Karimian, A; Nikparvar, B; Jabbari, I

2014-11-01

Renal angiography is one of the medical imaging methods in which patient and physician receive high equivalent doses due to long duration of fluoroscopy. In this research, equivalent doses of some radiosensitive tissues of patient (adult and child) and physician during renal angiography have been calculated by using adult and child Oak Ridge National Laboratory phantoms and Monte Carlo method (MCNPX). The results showed, in angiography of right kidney in a child and adult patient, that gall bladder with the amounts of 2.32 and 0.35 mSv, respectively, has received the most equivalent dose. About the physician, left hand, left eye and thymus absorbed the most amounts of doses, means 0.020 mSv. In addition, equivalent doses of the physician's lens eye, thyroid and knees were 0.023, 0.007 and 7.9E-4 mSv, respectively. Although these values are less than the reported thresholds by ICRP 103, it should be noted that these amounts are related to one examination. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Generation of the neutron response function of an NE213 scintillator for fusion applications

NASA Astrophysics Data System (ADS)

Binda, F.; Eriksson, J.; Ericsson, G.; Hellesen, C.; Conroy, S.; Nocente, M.; Sundén, E. Andersson; JET Contributors

2017-09-01

In this work we present a method to evaluate the neutron response function of an NE213 liquid scintillator. This method is particularly useful when the proton light yield function of the detector has not been measured, since it is based on a proton light yield function taken from literature, MCNPX simulations, measurements of gamma-rays from a calibration source and measurements of neutrons from fusion experiments with ohmic plasmas. The inclusion of the latter improves the description of the proton light yield function in the energy range of interest (around 2.46 MeV). We apply this method to an NE213 detector installed at JET, inside the radiation shielding of the magnetic proton recoil (MPRu) spectrometer, and present the results from the calibration along with some examples of application of the response function to perform neutron emission spectroscopy (NES) of fusion plasmas. We also investigate how the choice of the proton light yield function affects the NES analysis, finding that the result does not change significantly. This points to the fact that the method for the evaluation of the neutron response function is robust and gives reliable results.

Testing the Delayed Gamma Capability in MCNP6

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weldon, Robert A.; Fensin, Michael L.; McKinney, Gregg W.

The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy.more » Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented in this paper is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems. We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% ( 28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Finally, hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.« less

HPGe detector shielding optimization with MCNPX for the MEDINA PGNAA cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nicol, T.; Perot, B.; Carasco, C.

2015-07-01

Radioactive waste repositories must guarantee the non-toxicity of the waste in the long term, not only regarding radioactivity but also regarding other environmental contamination such as toxic chemicals. Analytical methods already exist for chemical characterization (ICP-MS, ICP-AES...) but they are based on test sampling. A possible alternative, for waste packages with an appropriate gamma radiation level, is to use Prompt Gamma Neutron Activation Analysis (PGNAA), a non-destructive measurement technique sensitive to several toxic chemicals. In view of the characterization of radioactive wastes in Germany and France, collaboration between the CEA Cadarache (France) and the Forschungszentrum Juelich (Germany) was initiated amore » few years ago. FZJ holds a PGNAA graphite cell called MEDINA (Multi Element Detection based on Instrumental Neutron Activation), allowing the characterization of 225 L drums. Fast neutrons are emitted from a D-T pulsed 14 MeV neutron generator and thermalized in graphite to induced radiative captures in the waste materials. Prompt capture gamma rays are detected using a 104% relative efficiency n-type HPGe. However, HPGe crystal is sensitive to fast neutron damage and to thermal neutron activation. A thermal neutron shield made of lithium fluorine and lithium carbonate is already used around the detector. In order to further decrease the current of fast and thermal neutrons coming into the crystal without penalizing MEDINA sensitivity (by decreasing the thermal neutron flux and neutron die away time of the cell, the gamma detection efficiency, or increasing the gamma background), some configurations based on easy-to-implement modifications of MEDINA have been simulated with MCNPX with a model of the cell already validated by experiments. Results show that fast and thermal neutron incoming current in the HPGe could easily be reduced by about a factor of 2 by additional quantities of graphite and by replacing lithium carbonate by lithium fluorine with a higher {sup 6}Li concentration. In addition, these modifications slightly increase the thermal neutron flux in the cell without deteriorating the neutron die away time, and reduce the gamma background about a factor of 2 during the neutron pulse but 5 times less after it. More important changes have also been tested, such as the addition of polyethylene and lead between the neutron generator and the HPGe detector, which is more effective regarding neutron shielding but decreases the neutron die away time, partly compensated by a larger initial thermal neutron flux. Concerning gamma background, hydrogen capture gamma ray (2.23 MeV) is increased due to the presence of polyethylene but lead around the HPGe decreases the total gamma background. In conclusion, simple modifications are possible to improve detector shielding and life time before thermal annealing of the crystal, without reducing MEDINA cell performances. Some of these modifications will be tested in the coming months. (authors)« less

Conceptuaciones de los estudiantes de las facultades de educacion y ciencias naturales de la Universidad de Puerto Rico, recinto de Rio Piedras, acerca de la ciencia y la pseudociencia

NASA Astrophysics Data System (ADS)

Reyes Medina, Hector A.

Esta investigacion describe las conceptuaciones de los estudiantes de tercer ano o mas a nivel de bachillerato de los programas de Educacion en Ciencia y Ciencias Naturales de la Universidad de Puerto Rico, Recinto de Rio Piedras, acerca de lo establecido en la literatura para distinguir el conocimiento cientifico de las creencias pseudocientificas. Este estudio se guio por un diseno tipo encuesta transversal que permitio conocer de manera consistente las conceptuaciones de los estudiantes encuestados acerca de la Ciencia y la Pseudociencia. Ademas, permitio desarrollar inferencias estadisticas relacionadas a la poblacion de estudio, sus conceptuaciones y su inclinacion teorica en torno al Realismo y al Racionalismo cientifico moderados. El instrumento utilizado fue el Cuestionario acerca de las concepciones de la ciencia y la pseudocienca en estudiantes universitarios, Reyes (2015). Este cuestionario fue validado mediante la recopilacion de diversas fuentes de evidencias, entre estas se encuentran las evidencias basadas en el contenido, el proceso de respuesta, la estructura interna y de constructo. Tambien, se calculo el Alfa de Crombach para la escala total y para cada componente y se realizo un analisis de factores que demostro la presencia de seis componentes claramente definidos de acuerdo a lo esperado sobre las caracteristicas originales del instrumento. Las estadisticas utilizadas fueron descriptivas. Participaron 302 alumnos, de las facultades de educacion y ciencias naturales. Se encontro que las conceptuaciones de los estudiantes de ambas facultades se inclinan en un 66.2% a favor con lo establecido en el modelo teorico en torno al Realismo y al Racionalismo cientifico moderados. Sin embargo, aun hay un 33.8% de los estudiantes de ambas facultades que poseen conceptuaciones distintas al modelo teorico propuesto.

Introduccion a la hidraulica de aguas subterraneas : un texto programado para auto-ensenanza

USGS Publications Warehouse

Bennett, Gordon D.

1987-01-01

Este ' texto programado esta diseflado para ayudarle a comprender la teoria de la hidniulica de aguas subterraneas por medio de la auto-enseflanza. La instrucci6n programada es un enfoque a una materia, un metodo de aprender;que no elimina el esfuerzo mental del proceso de aprendizaje. Algunas secciones de este programa necesitan solamente ser leidas; otras tendrian que ser elaboradas con lapiz y papel. Algunas preguntas pueden ser contestadas directamente; otras requieren calculos. A medida que se avanza en el texto, tendra que consultar frecuentemente textos o referencias sobre matematicas, mecanica de fluidos e hidrologia. En cada una de las ocho partes del texto, inicie el programa de instrucci6n leyendo la Secci6n 1. Elija una respuesta a la pregunta al final de la secci6n y dirijase a la nueva secci6n indicada al lado de la respuesta escogida. Si su respuesta fue correcta, pase a la secci6n que contiene materia nueva y otra pregunta, y proceda tal como en la Secci6n 1. Si su respuesta no fue correcta, dirijase a la secci6n que contiene explicaciones adicionales sobre el tema anterior y que le indica volver a la pregunta inicial e intentar de nuevo. En este caso, valdra Ia pena repasar el material de la secci6n anterior. Continue de esta man era en el programa hasta que llegue a Ia secci6n que indica el final de la parte. Observe que aunque las secciones estan en orden numerico en cada una de las ocho partes, por lo general, usted no procedeni en secuencia numerica (Secci6n 1 ala Secci6n 2, etc.) de principia a fin.

Production of α-particle emitting 211At using 45 MeV α-beam

NASA Astrophysics Data System (ADS)

Kim, Gyehong; Chun, Kwonsoo; Park, Sung Ho; Kim, Byungil

2014-06-01

Among the α-particle emitting radionuclides, 211At is considered to be a promising radionuclide for targeted cancer therapy due to its decay properties. The range of alpha particles produced by the decay of 211At are less than 70 µm in water with a linear energy transfer between 100 and 130 keV µm-1, which are about the maximum relative biological effectiveness for heavy ions. It is important to note that at the present time, only a few of cyclotrons routinely produce 211At. The direct production method is based on the nuclear reactions 209Bi(α,2n)211At. Production of the radionuclide 211At was carried out using the MC-50 cyclotron at the Korea Institute of Radiological and Medical Sciences (KIRAMS). To ensure high beam current, the α-beam was extracted with an initial energy of 45 MeV, which was degraded to obtain the appropriate α-beam energy. The calculations of beam energy degradation were performed utilizing the MCNPX. Alumina-baked targets were prepared by heating the bismuth metal powder onto a circular cavity in a furnace. When using an Eα, av of 29.17 MeV, the very small contribution of 210At confirms the right choice of the irradiation energy to obtain a pure production of 211At isotope.

Effect of external shielding for neutrons during radiotherapy for prostate cancer, considering the 2300 CD linear accelerator and voxel phantom

NASA Astrophysics Data System (ADS)

Thalhofer, J. L.; Roque, H. S.; Rebello, W. F.; Correa, S. A.; Silva, A. X.; Souza, E. M.; Batita, D. V. S.; Sandrini, E. S.

2014-02-01

Photoneutron production occurs when high energy photons, greater than 6.7 MeV, interact with linear accelerator head structures. In Brazil, the National Cancer Institute, one of the centers of reference in cancer treatment, uses radiation at 4 angles (0°, 90°, 180° and 270°) as treatment protocol for prostate cancer. With the objective of minimizing the dose deposited in the patient due to photoneutrons, this study simulated radiotherapy treatment using MCNPX, considering the most realistic environment; simulating the radiotherapy room, the Linac 2300 head, the MAX phantom and the treatment protocol with the accelerator operating at 18 MV. In an attempt to reduce the dose deposited by photoneutrons, an external shielding was added to the Linac 2300. Results show that the equivalent dose due to photoneutrons deposited in the patient diminished. The biggest reduction was seen in bone structures, such as the tibia and fibula, and mandible, at approximately 75%. Besides that, organs such as the brain, pancreas, small intestine, lungs and thyroid revealed a reduction of approximately 60%. It can be concluded that the shielding developed by our research group is efficient in neutron shielding, reducing the dose for the patient, and thus, the risk of secondary cancer, and increasing patient survival rates.

A model of tungsten anode x-ray spectra

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hernández, G.; Fernández, F., E-mail: fdz@usal.es

2016-08-15

Purpose: A semiempirical model for x-ray production in tungsten thick-targets was evaluated using a new characterization of electron fluence. Methods: Electron fluence is modeled taking into account both the energy and angular distributions, each of them adjusted to Monte Carlo simulated data. Distances were scaled by the CSDA range to reduce the energy dependence. Bremsstrahlung production was found by integrating the cross section with the fluence in a 1D penetration model. Characteristic radiation was added using a semiempirical law whose validity was checked. The results were compared the experimental results of Bhat et al., with the SpekCalc numerical tool, andmore » with MCNPX simulation results from the work of Hernandez and Boone. Results: The model described shows better agreement with the experimental results than the SpekCalc predictions in the sense of area between the spectra. A general improvement of the predictions of half-value layers is also found. The results are also in good agreement with the simulation results in the 50–640 keV energy range. Conclusions: A complete model for x-ray production in thick bremsstrahlung targets has been developed, improving the results of previous works and extending the energy range covered to the 50–640 keV interval.« less

X-ray sterilization of insects and microorganisms for cultural heritage applications

NASA Astrophysics Data System (ADS)

Borgognoni, F.; Vadrucci, M.; Bazzano, G.; Ferrari, P.; Massa, S.; Moretti, R.; Calvitti, M.; Ronsivalle, C.; Moriani, A.; Picardi, L.

2017-09-01

The APAM (Development of Particle Accelerators and Medical Applications) Laboratory of the ENEA Frascati Research Center is engaged in the preservation of cultural heritage as part of the COBRA (Sviluppo e diffusione di metodi, tecnologie e strumenti avanzati per la COnservazione dei Beni culturali, basati sull'applicazione di Radiazioni e di tecnologie Abilitanti) project addressed to the transfer of innovative technologies and methodologies from research to small and medium enterprises involved in the restorative measures. This work aims to demonstrate the effectiveness of ionizing radiation on the disinfection of biodegraded art objects. The conventional methods for the disinfestation of works of art, using chemicals toxic to humans and environment, might cause some damage to the treated material even on micrometric scale (i. e. either cellulose degradation). Ionizing radiations interact with the infesting biological material causing an irreversible DNA degradation. For this reason, they are certainly suitable for removal treatments of both macro organisms and bacterial colonies. A 4.8 MeV electron linear accelerator, normally dedicated to the characterization of dose detectors and radiographies, has been employed to produce Bremsstrahlung X-rays through a lead converter. The spectral fluence of the radiation source has been calculated using the Monte Carlo MCNPX code. The dosimetric characterization of the radiation field has been made using radiochromic films sensitive in the dose range of our interest (from 50 to 500 Gy) calibrated with a Markus ionization chamber. The irradiation of the artifact prototypes are made within a lead shielded room at a variable distance from the X-rays source. Samples subjected to irradiation consist of a soil bacterium, Agrobacterium rhizogenes, and an insect, Stegobium paniceum, that are found as wall paintings invasive coloniser and as a pest of books, wood works and paintings, respectively. Tests of irradiation have been performed on pest organisms as well as on woods mock-ups to evaluate potential damage to the material during the sterilization. The growing capacity of the treated bacterial cells re-cultured at the end of the treatment was evaluated on the bacterial sample and resulted to strongly inhibit cell growth during post-irradiation incubation, so that after incubation periods at 28 °C, no significant cell growth was observed. The induced levels of insect mortality and sterility vs absorbed dose and operative conditions have been also evaluated, demonstrating the induction of full sterility since the lower dose and 40% mortality by two days after the higher dose treatment. The experiments proved the ability to efficaciously treat objects of cultural heritage with X-rays in order to prevent the increase of the biodeterioration without damaging the materials: in fact, mechanical tests on both irradiated and not irradiated woods have demonstrated the absence of any induced degradation after the radiation exposition.

Tungsten anode spectral model using interpolating cubic splines: unfiltered x-ray spectra from 20 kV to 640 kV.

PubMed

Hernandez, Andrew M; Boone, John M

2014-04-01

Monte Carlo methods were used to generate lightly filtered high resolution x-ray spectra spanning from 20 kV to 640 kV. X-ray spectra were simulated for a conventional tungsten anode. The Monte Carlo N-Particle eXtended radiation transport code (MCNPX 2.6.0) was used to produce 35 spectra over the tube potential range from 20 kV to 640 kV, and cubic spline interpolation procedures were used to create piecewise polynomials characterizing the photon fluence per energy bin as a function of x-ray tube potential. Using these basis spectra and the cubic spline interpolation, 621 spectra were generated at 1 kV intervals from 20 to 640 kV. The tungsten anode spectral model using interpolating cubic splines (TASMICS) produces minimally filtered (0.8 mm Be) x-ray spectra with 1 keV energy resolution. The TASMICS spectra were compared mathematically with other, previously reported spectra. Using pairedt-test analyses, no statistically significant difference (i.e., p > 0.05) was observed between compared spectra over energy bins above 1% of peak bremsstrahlung fluence. For all energy bins, the correlation of determination (R(2)) demonstrated good correlation for all spectral comparisons. The mean overall difference (MOD) and mean absolute difference (MAD) were computed over energy bins (above 1% of peak bremsstrahlung fluence) and over all the kV permutations compared. MOD and MAD comparisons with previously reported spectra were 2.7% and 9.7%, respectively (TASMIP), 0.1% and 12.0%, respectively [R. Birch and M. Marshall, "Computation of bremsstrahlung x-ray spectra and comparison with spectra measured with a Ge(Li) detector," Phys. Med. Biol. 24, 505-517 (1979)], 0.4% and 8.1%, respectively (Poludniowski), and 0.4% and 8.1%, respectively (AAPM TG 195). The effective energy of TASMICS spectra with 2.5 mm of added Al filtration ranged from 17 keV (at 20 kV) to 138 keV (at 640 kV); with 0.2 mm of added Cu filtration the effective energy was 9 keV at 20 kV and 169 keV at 640 kV. Ranging from 20 kV to 640 kV, 621 x-ray spectra were produced and are available at 1 kV tube potential intervals. The spectra are tabulated at 1 keV intervals. TASMICS spectra were shown to be largely equivalent to published spectral models and are available in spreadsheet format for interested users by emailing the corresponding author (JMB). © 2014 American Association of Physicists in Medicine.

Error Analysis of non-TLD HDR Brachytherapy Dosimetric Techniques

NASA Astrophysics Data System (ADS)

Amoush, Ahmad

The American Association of Physicists in Medicine Task Group Report43 (AAPM-TG43) and its updated version TG-43U1 rely on the LiF TLD detector to determine the experimental absolute dose rate for brachytherapy. The recommended uncertainty estimates associated with TLD experimental dosimetry include 5% for statistical errors (Type A) and 7% for systematic errors (Type B). TG-43U1 protocol does not include recommendation for other experimental dosimetric techniques to calculate the absolute dose for brachytherapy. This research used two independent experimental methods and Monte Carlo simulations to investigate and analyze uncertainties and errors associated with absolute dosimetry of HDR brachytherapy for a Tandem applicator. An A16 MicroChamber* and one dose MOSFET detectors† were selected to meet the TG-43U1 recommendations for experimental dosimetry. Statistical and systematic uncertainty analyses associated with each experimental technique were analyzed quantitatively using MCNPX 2.6‡ to evaluate source positional error, Tandem positional error, the source spectrum, phantom size effect, reproducibility, temperature and pressure effects, volume averaging, stem and wall effects, and Tandem effect. Absolute dose calculations for clinical use are based on Treatment Planning System (TPS) with no corrections for the above uncertainties. Absolute dose and uncertainties along the transverse plane were predicted for the A16 microchamber. The generated overall uncertainties are 22%, 17%, 15%, 15%, 16%, 17%, and 19% at 1cm, 2cm, 3cm, 4cm, and 5cm, respectively. Predicting the dose beyond 5cm is complicated due to low signal-to-noise ratio, cable effect, and stem effect for the A16 microchamber. Since dose beyond 5cm adds no clinical information, it has been ignored in this study. The absolute dose was predicted for the MOSFET detector from 1cm to 7cm along the transverse plane. The generated overall uncertainties are 23%, 11%, 8%, 7%, 7%, 9%, and 8% at 1cm, 2cm, 3cm, and 4cm, 5cm, 6cm, and 7cm, respectively. The Nucletron Freiburg flap applicator is used with the Nucletron remote afterloader HDR machine to deliver dose to surface cancers. Dosimetric data for the Nucletron 192Ir source were generated using Monte Carlo simulation and compared with the published data. Two dimensional dosimetric data were calculated at two source positions; at the center of the sphere of the applicator and between two adjacent spheres. Unlike the TPS dose algorithm, The Monte Carlo code developed for this research accounts for the applicator material, secondary electrons and delta particles, and the air gap between the skin and the applicator. *Standard Imaging, Inc., Middleton, Wisconsin USA † OneDose MOSFET, Sicel Technologies, Morrisville NC ‡ Los Alamos National Laboratory, NM USA

Tungsten anode spectral model using interpolating cubic splines: Unfiltered x-ray spectra from 20 kV to 640 kV

PubMed Central

Hernandez, Andrew M.; Boone, John M.

2014-01-01

Purpose: Monte Carlo methods were used to generate lightly filtered high resolution x-ray spectra spanning from 20 kV to 640 kV. Methods: X-ray spectra were simulated for a conventional tungsten anode. The Monte Carlo N-Particle eXtended radiation transport code (MCNPX 2.6.0) was used to produce 35 spectra over the tube potential range from 20 kV to 640 kV, and cubic spline interpolation procedures were used to create piecewise polynomials characterizing the photon fluence per energy bin as a function of x-ray tube potential. Using these basis spectra and the cubic spline interpolation, 621 spectra were generated at 1 kV intervals from 20 to 640 kV. The tungsten anode spectral model using interpolating cubic splines (TASMICS) produces minimally filtered (0.8 mm Be) x-ray spectra with 1 keV energy resolution. The TASMICS spectra were compared mathematically with other, previously reported spectra. Results: Using paired t-test analyses, no statistically significant difference (i.e., p > 0.05) was observed between compared spectra over energy bins above 1% of peak bremsstrahlung fluence. For all energy bins, the correlation of determination (R2) demonstrated good correlation for all spectral comparisons. The mean overall difference (MOD) and mean absolute difference (MAD) were computed over energy bins (above 1% of peak bremsstrahlung fluence) and over all the kV permutations compared. MOD and MAD comparisons with previously reported spectra were 2.7% and 9.7%, respectively (TASMIP), 0.1% and 12.0%, respectively [R. Birch and M. Marshall, “Computation of bremsstrahlung x-ray spectra and comparison with spectra measured with a Ge(Li) detector,” Phys. Med. Biol. 24, 505–517 (1979)], 0.4% and 8.1%, respectively (Poludniowski), and 0.4% and 8.1%, respectively (AAPM TG 195). The effective energy of TASMICS spectra with 2.5 mm of added Al filtration ranged from 17 keV (at 20 kV) to 138 keV (at 640 kV); with 0.2 mm of added Cu filtration the effective energy was 9 keV at 20 kV and 169 keV at 640 kV. Conclusions: Ranging from 20 kV to 640 kV, 621 x-ray spectra were produced and are available at 1 kV tube potential intervals. The spectra are tabulated at 1 keV intervals. TASMICS spectra were shown to be largely equivalent to published spectral models and are available in spreadsheet format for interested users by emailing the corresponding author (JMB). PMID:24694149

Tungsten anode spectral model using interpolating cubic splines: Unfiltered x-ray spectra from 20 kV to 640 kV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hernandez, Andrew M.; Boone, John M., E-mail: john.boone@ucdmc.ucdavis.edu

Purpose: Monte Carlo methods were used to generate lightly filtered high resolution x-ray spectra spanning from 20 kV to 640 kV. Methods: X-ray spectra were simulated for a conventional tungsten anode. The Monte Carlo N-Particle eXtended radiation transport code (MCNPX 2.6.0) was used to produce 35 spectra over the tube potential range from 20 kV to 640 kV, and cubic spline interpolation procedures were used to create piecewise polynomials characterizing the photon fluence per energy bin as a function of x-ray tube potential. Using these basis spectra and the cubic spline interpolation, 621 spectra were generated at 1 kV intervalsmore » from 20 to 640 kV. The tungsten anode spectral model using interpolating cubic splines (TASMICS) produces minimally filtered (0.8 mm Be) x-ray spectra with 1 keV energy resolution. The TASMICS spectra were compared mathematically with other, previously reported spectra. Results: Using pairedt-test analyses, no statistically significant difference (i.e., p > 0.05) was observed between compared spectra over energy bins above 1% of peak bremsstrahlung fluence. For all energy bins, the correlation of determination (R{sup 2}) demonstrated good correlation for all spectral comparisons. The mean overall difference (MOD) and mean absolute difference (MAD) were computed over energy bins (above 1% of peak bremsstrahlung fluence) and over all the kV permutations compared. MOD and MAD comparisons with previously reported spectra were 2.7% and 9.7%, respectively (TASMIP), 0.1% and 12.0%, respectively [R. Birch and M. Marshall, “Computation of bremsstrahlung x-ray spectra and comparison with spectra measured with a Ge(Li) detector,” Phys. Med. Biol. 24, 505–517 (1979)], 0.4% and 8.1%, respectively (Poludniowski), and 0.4% and 8.1%, respectively (AAPM TG 195). The effective energy of TASMICS spectra with 2.5 mm of added Al filtration ranged from 17 keV (at 20 kV) to 138 keV (at 640 kV); with 0.2 mm of added Cu filtration the effective energy was 9 keV at 20 kV and 169 keV at 640 kV. Conclusions: Ranging from 20 kV to 640 kV, 621 x-ray spectra were produced and are available at 1 kV tube potential intervals. The spectra are tabulated at 1 keV intervals. TASMICS spectra were shown to be largely equivalent to published spectral models and are available in spreadsheet format for interested users by emailing the corresponding author (JMB)« less

Estudio numerico y experimental del proceso de soldeo MIG sobre la aleacion 6063--T5 utilizando el metodo de Taguchi

NASA Astrophysics Data System (ADS)

Meseguer Valdenebro, Jose Luis

Electric arc welding processes represent one of the most used techniques on manufacturing processes of mechanical components in modern industry. The electric arc welding processes have been adapted to current needs, becoming a flexible and versatile way to manufacture. Numerical results in the welding process are validated experimentally. The main numerical methods most commonly used today are three: finite difference method, finite element method and finite volume method. The most widely used numerical method for the modeling of welded joints is the finite element method because it is well adapted to the geometric and boundary conditions in addition to the fact that there is a variety of commercial programs which use the finite element method as a calculation basis. The content of this thesis shows an experimental study of a welded joint conducted by means of the MIG welding process of aluminum alloy 6063-T5. The numerical process is validated experimentally by applying the method of finite element through the calculation program ANSYS. The experimental results in this paper are the cooling curves, the critical cooling time t4/3, the weld bead geometry, the microhardness obtained in the welded joint, and the metal heat affected zone base, process dilution, critical areas intersected between the cooling curves and the curve TTP. The numerical results obtained in this thesis are: the thermal cycle curves, which represent both the heating to maximum temperature and subsequent cooling. The critical cooling time t4/3 and thermal efficiency of the process are calculated and the bead geometry obtained experimentally is represented. The heat affected zone is obtained by differentiating the zones that are found at different temperatures, the critical areas intersected between the cooling curves and the TTP curve. In order to conclude this doctoral thesis, an optimization has been conducted by means of the Taguchi method for welding parameters in order to obtain an improvement on mechanical properties in aluminum metal joint. Los procesos de soldadura por arco electrico representan unas de las tecnicas mas utilizadas en los procesos de fabricacion de componentes mecanicos en la industria moderna. Los procesos de soldeo por arco se han adaptado a las necesidades actuales, haciendose un modo de fabricacion flexible y versatil. Los resultados obtenidos numericamente en el proceso de soldadura son validados experimentalmente. Los principales metodos numericos mas empleados en la actualidad son tres, metodo por diferencias finitas, metodos por elementos finitos y metodo por volumenes finitos. El metodo numerico mas empleado para el modelado de uniones soldadas, es el metodo por elementos finitos, debido a que presenta una buena adaptacion a las condiciones geometricas y de contorno ademas de que existe una diversidad de programas comerciales que utilizan el metodo por elementos finitos como base de calculo. Este trabajo de investigacion presenta un estudio experimental de una union soldada mediante el proceso MIG de la aleacion de aluminio 6063-T5. El metodo numerico se valida experimentalmente aplicando el metodo de los elementos finitos con el programa de calculo ANSYS. Los resultados experimentales obtenidos son: las curvas de enfriamiento, el tiempo critico de enfriamiento t4/3, geometria del cordon, microdurezas obtenidas en la union soldada, zona afectada termicamente y metal base, dilucion del proceso, areas criticas intersecadas entre las curvas de enfriamiento y la curva TTP. Los resultados numericos son: las curvas del ciclo termico, que representan tanto el calentamiento hasta alcanzar la temperatura maxima y un posterior enfriamiento. Se calculan el tiempo critico de enfriamiento t4/3, el rendimiento termico y se representa la geometria del cordon obtenida experimentalmente. La zona afectada termicamente se obtiene diferenciando las zonas que se encuentran a diferentes temperaturas, las areas criticas intersecadas entre las curvas de enfriamiento y la curva TTP. Para finalizar el trabajo de investigacion se ha realizado una optimizacion, con la aplicacion del metodo de Taguchi, de los parametros de soldeo con el objetivo de obtener una mejora sustancial en las propiedades mecanicas de las uniones metalicas de aluminio.

Measurement and simulation of a Compton suppression system for safeguards application

NASA Astrophysics Data System (ADS)

Lee, Seung Kyu; Seo, Hee; Won, Byung-Hee; Lee, Chaehun; Shin, Hee-Sung; Na, Sang-Ho; Song, Dae-Yong; Kim, Ho-Dong; Park, Geun-Il; Park, Se-Hwan

2015-11-01

Plutonium (Pu) contents in spent nuclear fuels, recovered uranium (U) or uranium/transuranium (U/TRU) products must be measured in order to secure the safeguardability of a pyroprocessing facility. Self-induced X-Ray fluorescence (XRF) and gamma-ray spectroscopy are useful techniques for determining Pu-to-U ratios and Pu isotope ratios of spent fuel. Photon measurements of spent nuclear fuel by using high-resolution spectrometers such as high-purity germanium (HPGe) detectors show a large continuum background in the low-energy region, which is due in large part to Compton scattering of energetic gamma rays. This paper proposes a Compton suppression system for reducing of the Compton continuum background. In the present study, the system was configured by using an HPGe main detector and a BGO (bismuth germanate: Bi4Ge3O12) guard detector. The system performances for gamma-ray measurement and XRF were evaluated by means of Monte Carlo simulations and measurements of the radiation source. The Monte Carlo N-Particle eXtended (MCNPX) simulations were performed using the same geometry as for the experiments, and considered, for exact results, the production of secondary electrons and photons. As a performance test of the Compton suppression system, the peak-to-Compton ratio, which is a figure of merit to evaluate the gamma-ray detection, was enhanced by a factor of three or more when the Compton suppression system was used.

Calculation of organ doses in x-ray examinations of premature babies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smans, Kristien; Tapiovaara, Markku; Cannie, Mieke

Lung disease represents one of the most life-threatening conditions in prematurely born children. In the evaluation of the neonatal chest, the primary and most important diagnostic study is the chest radiograph. Since prematurely born children are very sensitive to radiation, those radiographs may lead to a significant radiation detriment. Knowledge of the radiation dose is therefore necessary to justify the exposures. To calculate doses in the entire body and in specific organs, computational models of the human anatomy are needed. Using medical imaging techniques, voxel phantoms have been developed to achieve a representation as close as possible to the anatomicalmore » properties. In this study two voxel phantoms, representing prematurely born babies, were created from computed tomography- and magnetic resonance images: Phantom 1 (1910 g) and Phantom 2 (590 g). The two voxel phantoms were used in Monte Carlo calculations (MCNPX) to assess organ doses. The results were compared with the commercially available software package PCXMC in which the available mathematical phantoms can be downsized toward the prematurely born baby. The simple phantom-scaling method used in PCXMC seems to be sufficient to calculate doses for organs within the radiation field. However, one should be careful in specifying the irradiation geometry. Doses in organs that are wholly or partially outside the primary radiation field depend critically on the irradiation conditions and the phantom model.« less

First Industrial Tests of a Matrix Monitor Correction for the Differential Die-away Technique of Historical Waste Drums

DOE Office of Scientific and Technical Information (OSTI.GOV)

Antoni, Rodolphe; Passard, Christian; Perot, Bertrand

2015-07-01

The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA NC La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT). In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (LMN) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor, namely a 3He proportional counter located insidemore » the measurement cavity. After feasibility studies performed with LMN's PROMETHEE 6 laboratory measurement cell and with MCNPX simulations, this paper presents first experimental tests performed on the industrial ACC (hulls and nozzles compaction facility) measurement system. A calculation vs. experiment benchmark has been carried out by performing dedicated calibration measurements with a representative drum and {sup 235}U samples. The comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach and the industrial feasibility of the method, which will be implemented on the industrial station for the measurement of historical wastes. (authors)« less

Nonspherical Radiation Driven Wind Models Applied to Be Stars

NASA Astrophysics Data System (ADS)

Arauxo, F. X.

1990-11-01

ABSTRACT. In this work we present a model for the structure of a radiatively driven wind in the meridional plane of a hot star. Rotation effects and simulation of viscous forces were included in the motion equations. The line radiation force is considered with the inclusion of the finite disk correction in self-consistent computations which also contain gravity darkening as well as distortion of the star by rotation. An application to a typical BlV star leads to mass-flux ratios between equator and pole of the order of 10 and mass loss rates in the range 5.l0 to Mo/yr. Our envelope models are flattened towards the equator and the wind terminal velocities in that region are rather high (1000 Km/s). However, in the region near the star the equatorial velocity field is dominated by rotation. RESUMEN. Se presenta un modelo de la estructura de un viento empujado radiativamente en el plano meridional de una estrella caliente. Se incluyeron en las ecuaciones de movimiento los efectos de rotaci6n y la simulaci6n de fuerzas viscosas. Se consider6 la fuerza de las lineas de radiaci6n incluyendo la correcci6n de disco finito en calculos autoconsistentes los cuales incluyen oscurecimiento gravitacional asi como distorsi6n de la estrella por rotaci6n. La aplicaci6n a una estrella tipica BlV lleva a cocientes de flujo de masa entre el ecuador y el polo del orden de 10 de perdida de masa en el intervalo 5.l0 a 10 Mo/ano. Nuestros modelos de envolvente estan achatados hacia el ecuador y las velocidads terminales del viento en esa regi6n son bastante altas (1000 Km/s). Sin embargo, en la regi6n cercana a la estrella el campo de velocidad ecuatorial esta dominado por la rotaci6n. Key words: STARS-BE -- STARS-WINDS

Generating code adapted for interlinking legacy scalar code and extended vector code

DOEpatents

Gschwind, Michael K

2013-06-04

Mechanisms for intermixing code are provided. Source code is received for compilation using an extended Application Binary Interface (ABI) that extends a legacy ABI and uses a different register configuration than the legacy ABI. First compiled code is generated based on the source code, the first compiled code comprising code for accommodating the difference in register configurations used by the extended ABI and the legacy ABI. The first compiled code and second compiled code are intermixed to generate intermixed code, the second compiled code being compiled code that uses the legacy ABI. The intermixed code comprises at least one call instruction that is one of a call from the first compiled code to the second compiled code or a call from the second compiled code to the first compiled code. The code for accommodating the difference in register configurations is associated with the at least one call instruction.

Neutron dosimetry in organs of an adult human phantom using linacs with multileaf collimator in radiotherapy treatments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martinez-Ovalle, S. A.; Barquero, R.; Gomez-Ros, J. M.

Purpose: To calculate absorbed doses due to neutrons in 87 organs/tissues for anthropomorphic phantoms, irradiated in position supine (head first into the gantry) with orientations anteroposterior (AP) and right-left (RLAT) with a 18 MV accelerator. Conversion factors from monitor units to {mu}Gy per neutron in organs, equivalent doses in organs/tissues, and effective doses, which permit to quantify stochastic risks, are estimated. Methods: MAX06 and FAX06 phantoms were modeled with MCNPX and irradiated with a 18 MV Varian Clinac 2100C/D accelerator whose geometry included a multileaf collimator. Two actual fields of a pelvic treatment were simulated using electron-photon-neutron coupled transport. Absorbedmore » doses due to neutrons were estimated from kerma. Equivalent doses were estimated using the radiation weighting factor corresponding to an average incident neutron energy 0.47 MeV. Statistical uncertainties associated to absorbed doses, as calculated by MCNPX, were also obtained. Results: Largest doses were absorbed in shallowest (with respect to the neutron pathway) organs. In {mu}GyMU{sup -1}, values of 2.66 (for penis) and 2.33 (for testes) were found in MAX06, and 1.68 (for breasts), 1.05 (for lenses of eyes), and 0.94 (for sublingual salivary glands) in FAX06, in AP orientation. In RLAT, the largest doses were found for bone tissues (leg) just at the entrance of the beam in the body (right side in our case). Values, in {mu}GyMU{sup -1}, of 1.09 in upper leg bone right spongiosa, for MAX06, and 0.63 in mandible spongiosa, for FAX06, were found. Except for gonads, liver, and stomach wall, equivalent doses found for FAX06 were, in both orientations, higher than for MAX06. Equivalent doses in AP are higher than in RLAT for all organs/tissues other than brain and liver. Effective doses of 12.6 and 4.1 {mu}SvMU{sup -1} were found for AP and RLAT, respectively. The organs/tissues with larger relative contributions to the effective dose were testes and breasts, in AP, and breasts and red marrow, in RLAT. Equivalent and effective doses obtained for MAX06/FAX06 were smaller (between 2 and 20 times) than those quoted for the mathematical phantoms ADAM/EVA in ICRP-74. Conclusions: The new calculations of conversion coefficients for neutron irradiation in AP and RLAT irradiation geometries show a reduction in the values of effective dose by factors 7 (AP) and 6 (RLAT) with respect to the old data obtained with mathematical phantoms. The existence of tissues or anatomical regions with maximum absorbed doses, such as penis, lens of eyes, fascia (part of connective tissue), etc., organs/tissues that classic mathematical phantoms did not include because they were not considered for the study of stochastic effects, has been revealed. Absorbed doses due to photons, obtained following the same simulation methodology, are larger than those due to neutrons, reaching values 100 times larger as the primary beam is approached. However, for organs far from the treated volume, absorbed photon doses can be up to three times smaller than neutron ones. Calculations using voxel phantoms permitted to know the organ dose conversion coefficients per MU due to secondary neutrons in the complete anatomy of a patient.« less

Bandwidth efficient coding for satellite communications

NASA Technical Reports Server (NTRS)

Lin, Shu; Costello, Daniel J., Jr.; Miller, Warner H.; Morakis, James C.; Poland, William B., Jr.

1992-01-01

An error control coding scheme was devised to achieve large coding gain and high reliability by using coded modulation with reduced decoding complexity. To achieve a 3 to 5 dB coding gain and moderate reliability, the decoding complexity is quite modest. In fact, to achieve a 3 dB coding gain, the decoding complexity is quite simple, no matter whether trellis coded modulation or block coded modulation is used. However, to achieve coding gains exceeding 5 dB, the decoding complexity increases drastically, and the implementation of the decoder becomes very expensive and unpractical. The use is proposed of coded modulation in conjunction with concatenated (or cascaded) coding. A good short bandwidth efficient modulation code is used as the inner code and relatively powerful Reed-Solomon code is used as the outer code. With properly chosen inner and outer codes, a concatenated coded modulation scheme not only can achieve large coding gains and high reliability with good bandwidth efficiency but also can be practically implemented. This combination of coded modulation and concatenated coding really offers a way of achieving the best of three worlds, reliability and coding gain, bandwidth efficiency, and decoding complexity.

A genetic scale of reading frame coding.

PubMed

Michel, Christian J

2014-08-21

The reading frame coding (RFC) of codes (sets) of trinucleotides is a genetic concept which has been largely ignored during the last 50 years. A first objective is the definition of a new and simple statistical parameter PrRFC for analysing the probability (efficiency) of reading frame coding (RFC) of any trinucleotide code. A second objective is to reveal different classes and subclasses of trinucleotide codes involved in reading frame coding: the circular codes of 20 trinucleotides and the bijective genetic codes of 20 trinucleotides coding the 20 amino acids. This approach allows us to propose a genetic scale of reading frame coding which ranges from 1/3 with the random codes (RFC probability identical in the three frames) to 1 with the comma-free circular codes (RFC probability maximal in the reading frame and null in the two shifted frames). This genetic scale shows, in particular, the reading frame coding probabilities of the 12,964,440 circular codes (PrRFC=83.2% in average), the 216 C(3) self-complementary circular codes (PrRFC=84.1% in average) including the code X identified in eukaryotic and prokaryotic genes (PrRFC=81.3%) and the 339,738,624 bijective genetic codes (PrRFC=61.5% in average) including the 52 codes without permuted trinucleotides (PrRFC=66.0% in average). Otherwise, the reading frame coding probabilities of each trinucleotide code coding an amino acid with the universal genetic code are also determined. The four amino acids Gly, Lys, Phe and Pro are coded by codes (not circular) with RFC probabilities equal to 2/3, 1/2, 1/2 and 2/3, respectively. The amino acid Leu is coded by a circular code (not comma-free) with a RFC probability equal to 18/19. The 15 other amino acids are coded by comma-free circular codes, i.e. with RFC probabilities equal to 1. The identification of coding properties in some classes of trinucleotide codes studied here may bring new insights in the origin and evolution of the genetic code. Copyright © 2014 Elsevier Ltd. All rights reserved.

Error-correction coding for digital communications

NASA Astrophysics Data System (ADS)

Clark, G. C., Jr.; Cain, J. B.

This book is written for the design engineer who must build the coding and decoding equipment and for the communication system engineer who must incorporate this equipment into a system. It is also suitable as a senior-level or first-year graduate text for an introductory one-semester course in coding theory. Fundamental concepts of coding are discussed along with group codes, taking into account basic principles, practical constraints, performance computations, coding bounds, generalized parity check codes, polynomial codes, and important classes of group codes. Other topics explored are related to simple nonalgebraic decoding techniques for group codes, soft decision decoding of block codes, algebraic techniques for multiple error correction, the convolutional code structure and Viterbi decoding, syndrome decoding techniques, and sequential decoding techniques. System applications are also considered, giving attention to concatenated codes, coding for the white Gaussian noise channel, interleaver structures for coded systems, and coding for burst noise channels.

A novel concatenated code based on the improved SCG-LDPC code for optical transmission systems

NASA Astrophysics Data System (ADS)

Yuan, Jian-guo; Xie, Ya; Wang, Lin; Huang, Sheng; Wang, Yong

2013-01-01

Based on the optimization and improvement for the construction method of systematically constructed Gallager (SCG) (4, k) code, a novel SCG low density parity check (SCG-LDPC)(3969, 3720) code to be suitable for optical transmission systems is constructed. The novel SCG-LDPC (6561,6240) code with code rate of 95.1% is constructed by increasing the length of SCG-LDPC (3969,3720) code, and in a way, the code rate of LDPC codes can better meet the high requirements of optical transmission systems. And then the novel concatenated code is constructed by concatenating SCG-LDPC(6561,6240) code and BCH(127,120) code with code rate of 94.5%. The simulation results and analyses show that the net coding gain (NCG) of BCH(127,120)+SCG-LDPC(6561,6240) concatenated code is respectively 2.28 dB and 0.48 dB more than those of the classic RS(255,239) code and SCG-LDPC(6561,6240) code at the bit error rate (BER) of 10-7.

Trace-shortened Reed-Solomon codes

NASA Technical Reports Server (NTRS)

Mceliece, R. J.; Solomon, G.

1994-01-01

Reed-Solomon (RS) codes have been part of standard NASA telecommunications systems for many years. RS codes are character-oriented error-correcting codes, and their principal use in space applications has been as outer codes in concatenated coding systems. However, for a given character size, say m bits, RS codes are limited to a length of, at most, 2(exp m). It is known in theory that longer character-oriented codes would be superior to RS codes in concatenation applications, but until recently no practical class of 'long' character-oriented codes had been discovered. In 1992, however, Solomon discovered an extensive class of such codes, which are now called trace-shortened Reed-Solomon (TSRS) codes. In this article, we will continue the study of TSRS codes. Our main result is a formula for the dimension of any TSRS code, as a function of its error-correcting power. Using this formula, we will give several examples of TSRS codes, some of which look very promising as candidate outer codes in high-performance coded telecommunications systems.

Error Control Coding Techniques for Space and Satellite Communications

NASA Technical Reports Server (NTRS)

Lin, Shu

2000-01-01

This paper presents a concatenated turbo coding system in which a Reed-Solomom outer code is concatenated with a binary turbo inner code. In the proposed system, the outer code decoder and the inner turbo code decoder interact to achieve both good bit error and frame error performances. The outer code decoder helps the inner turbo code decoder to terminate its decoding iteration while the inner turbo code decoder provides soft-output information to the outer code decoder to carry out a reliability-based soft-decision decoding. In the case that the outer code decoding fails, the outer code decoder instructs the inner code decoder to continue its decoding iterations until the outer code decoding is successful or a preset maximum number of decoding iterations is reached. This interaction between outer and inner code decoders reduces decoding delay. Also presented in the paper are an effective criterion for stopping the iteration process of the inner code decoder and a new reliability-based decoding algorithm for nonbinary codes.

An Interactive Concatenated Turbo Coding System

NASA Technical Reports Server (NTRS)

Liu, Ye; Tang, Heng; Lin, Shu; Fossorier, Marc

1999-01-01

This paper presents a concatenated turbo coding system in which a Reed-Solomon outer code is concatenated with a binary turbo inner code. In the proposed system, the outer code decoder and the inner turbo code decoder interact to achieve both good bit error and frame error performances. The outer code decoder helps the inner turbo code decoder to terminate its decoding iteration while the inner turbo code decoder provides soft-output information to the outer code decoder to carry out a reliability-based soft- decision decoding. In the case that the outer code decoding fails, the outer code decoder instructs the inner code decoder to continue its decoding iterations until the outer code decoding is successful or a preset maximum number of decoding iterations is reached. This interaction between outer and inner code decoders reduces decoding delay. Also presented in the paper are an effective criterion for stopping the iteration process of the inner code decoder and a new reliability-based decoding algorithm for nonbinary codes.

Performance Analysis of New Binary User Codes for DS-CDMA Communication

NASA Astrophysics Data System (ADS)

Usha, Kamle; Jaya Sankar, Kottareddygari

2016-03-01

This paper analyzes new binary spreading codes through correlation properties and also presents their performance over additive white Gaussian noise (AWGN) channel. The proposed codes are constructed using gray and inverse gray codes. In this paper, a n-bit gray code appended by its n-bit inverse gray code to construct the 2n-length binary user codes are discussed. Like Walsh codes, these binary user codes are available in sizes of power of two and additionally code sets of length 6 and their even multiples are also available. The simple construction technique and generation of code sets of different sizes are the salient features of the proposed codes. Walsh codes and gold codes are considered for comparison in this paper as these are popularly used for synchronous and asynchronous multi user communications respectively. In the current work the auto and cross correlation properties of the proposed codes are compared with those of Walsh codes and gold codes. Performance of the proposed binary user codes for both synchronous and asynchronous direct sequence CDMA communication over AWGN channel is also discussed in this paper. The proposed binary user codes are found to be suitable for both synchronous and asynchronous DS-CDMA communication.

Multiple component codes based generalized LDPC codes for high-speed optical transport.

PubMed

Djordjevic, Ivan B; Wang, Ting

2014-07-14

A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

Organ S values and effective doses for family members exposed to adult patients following I-131 treatment: A Monte Carlo simulation study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Han, Eun Young; Lee, Choonsik; Mcguire, Lynn

Purpose: To calculate organ S values (mGy/Bq-s) and effective doses per time-integrated activity (mSv/Bq-s) for pediatric and adult family members exposed to an adult male or female patient treated with I-131 using a series of hybrid computational phantoms coupled with a Monte Carlo radiation transport technique.Methods: A series of pediatric and adult hybrid computational phantoms were employed in the study. Three different exposure scenarios were considered: (1) standing face-to-face exposures between an adult patient and pediatric or adult family phantoms at five different separation distances; (2) an adult female patient holding her newborn child, and (3) a 1-yr-old child standingmore » on the lap of an adult female patient. For the adult patient model, two different thyroid-related diseases were considered: hyperthyroidism and differentiated thyroid cancer (DTC) with corresponding internal distributions of {sup 131}I. A general purpose Monte Carlo code, MCNPX v2.7, was used to perform the Monte Carlo radiation transport.Results: The S values show a strong dependency on age and organ location within the family phantoms at short distances. The S values and effective dose per time-integrated activity from the adult female patient phantom are relatively high at shorter distances and to younger family phantoms. At a distance of 1 m, effective doses per time-integrated activity are lower than those values based on the NRC (Nuclear Regulatory Commission) by a factor of 2 for both adult male and female patient phantoms. The S values to target organs from the hyperthyroid-patient source distribution strongly depend on the height of the exposed family phantom, so that their values rapidly decrease with decreasing height of the family phantom. Active marrow of the 10-yr-old phantom shows the highest S values among family phantoms for the DTC-patient source distribution. In the exposure scenario of mother and baby, S values and effective doses per time-integrated activity to the newborn and 1-yr-old phantoms for a hyperthyroid-patient source are higher than values for a DTC-patient source.Conclusions: The authors performed realistic assessments of {sup 131}I organ S values and effective dose per time-integrated activity from adult patients treated for hyperthyroidism and DTC to family members. In addition, the authors’ studies consider Monte Carlo simulated “mother and baby/child” exposure scenarios for the first time. Based on these results, the authors reconfirm the strong conservatism underlying the point source method recommended by the US NRC. The authors recommend that various factors such as the type of the patient's disease, the age of family members, and the distance/posture between the patient and family members must be carefully considered to provide realistic dose estimates for patient-to-family exposures.« less

Multilevel Concatenated Block Modulation Codes for the Frequency Non-selective Rayleigh Fading Channel

NASA Technical Reports Server (NTRS)

Lin, Shu; Rhee, Dojun

1996-01-01

This paper is concerned with construction of multilevel concatenated block modulation codes using a multi-level concatenation scheme for the frequency non-selective Rayleigh fading channel. In the construction of multilevel concatenated modulation code, block modulation codes are used as the inner codes. Various types of codes (block or convolutional, binary or nonbinary) are being considered as the outer codes. In particular, we focus on the special case for which Reed-Solomon (RS) codes are used as the outer codes. For this special case, a systematic algebraic technique for constructing q-level concatenated block modulation codes is proposed. Codes have been constructed for certain specific values of q and compared with the single-level concatenated block modulation codes using the same inner codes. A multilevel closest coset decoding scheme for these codes is proposed.

Unaligned instruction relocation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bertolli, Carlo; O'Brien, John K.; Sallenave, Olivier H.

In one embodiment, a computer-implemented method includes receiving source code to be compiled into an executable file for an unaligned instruction set architecture (ISA). Aligned assembled code is generated, by a computer processor. The aligned assembled code complies with an aligned ISA and includes aligned processor code for a processor and aligned accelerator code for an accelerator. A first linking pass is performed on the aligned assembled code, including relocating a first relocation target in the aligned accelerator code that refers to a first object outside the aligned accelerator code. Unaligned assembled code is generated in accordance with the unalignedmore » ISA and includes unaligned accelerator code for the accelerator and unaligned processor code for the processor. A second linking pass is performed on the unaligned assembled code, including relocating a second relocation target outside the unaligned accelerator code that refers to an object in the unaligned accelerator code.« less

Unaligned instruction relocation

DOEpatents

Bertolli, Carlo; O'Brien, John K.; Sallenave, Olivier H.; Sura, Zehra N.

2018-01-23

In one embodiment, a computer-implemented method includes receiving source code to be compiled into an executable file for an unaligned instruction set architecture (ISA). Aligned assembled code is generated, by a computer processor. The aligned assembled code complies with an aligned ISA and includes aligned processor code for a processor and aligned accelerator code for an accelerator. A first linking pass is performed on the aligned assembled code, including relocating a first relocation target in the aligned accelerator code that refers to a first object outside the aligned accelerator code. Unaligned assembled code is generated in accordance with the unaligned ISA and includes unaligned accelerator code for the accelerator and unaligned processor code for the processor. A second linking pass is performed on the unaligned assembled code, including relocating a second relocation target outside the unaligned accelerator code that refers to an object in the unaligned accelerator code.

Linear chirp phase perturbing approach for finding binary phased codes

NASA Astrophysics Data System (ADS)

Li, Bing C.

2017-05-01

Binary phased codes have many applications in communication and radar systems. These applications require binary phased codes to have low sidelobes in order to reduce interferences and false detection. Barker codes are the ones that satisfy these requirements and they have lowest maximum sidelobes. However, Barker codes have very limited code lengths (equal or less than 13) while many applications including low probability of intercept radar, and spread spectrum communication, require much higher code lengths. The conventional techniques of finding binary phased codes in literatures include exhaust search, neural network, and evolutionary methods, and they all require very expensive computation for large code lengths. Therefore these techniques are limited to find binary phased codes with small code lengths (less than 100). In this paper, by analyzing Barker code, linear chirp, and P3 phases, we propose a new approach to find binary codes. Experiments show that the proposed method is able to find long low sidelobe binary phased codes (code length >500) with reasonable computational cost.

Maximum likelihood decoding analysis of Accumulate-Repeat-Accumulate Codes

NASA Technical Reports Server (NTRS)

Abbasfar, Aliazam; Divsalar, Dariush; Yao, Kung

2004-01-01

Repeat-Accumulate (RA) codes are the simplest turbo-like codes that achieve good performance. However, they cannot compete with Turbo codes or low-density parity check codes (LDPC) as far as performance is concerned. The Accumulate Repeat Accumulate (ARA) codes, as a subclass of LDPC codes, are obtained by adding a pre-coder in front of RA codes with puncturing where an accumulator is chosen as a precoder. These codes not only are very simple, but also achieve excellent performance with iterative decoding. In this paper, the performance of these codes with (ML) decoding are analyzed and compared to random codes by very tight bounds. The weight distribution of some simple ARA codes is obtained, and through existing tightest bounds we have shown the ML SNR threshold of ARA codes approaches very closely to the performance of random codes. We have shown that the use of precoder improves the SNR threshold but interleaving gain remains unchanged with respect to RA code with puncturing.

Accumulate repeat accumulate codes

NASA Technical Reports Server (NTRS)

Abbasfar, A.; Divsalar, D.; Yao, K.

2004-01-01

In this paper we propose an innovative channel coding scheme called Accumulate Repeat Accumulate codes. This class of codes can be viewed as trubo-like codes, namely a double serial concatenation of a rate-1 accumulator as an outer code, a regular or irregular repetition as a middle code, and a punctured accumulator as an inner code.

Ensemble coding of face identity is not independent of the coding of individual identity.

PubMed

Neumann, Markus F; Ng, Ryan; Rhodes, Gillian; Palermo, Romina

2018-06-01

Information about a group of similar objects can be summarized into a compressed code, known as ensemble coding. Ensemble coding of simple stimuli (e.g., groups of circles) can occur in the absence of detailed exemplar coding, suggesting dissociable processes. Here, we investigate whether a dissociation would still be apparent when coding facial identity, where individual exemplar information is much more important. We examined whether ensemble coding can occur when exemplar coding is difficult, as a result of large sets or short viewing times, or whether the two types of coding are positively associated. We found a positive association, whereby both ensemble and exemplar coding were reduced for larger groups and shorter viewing times. There was no evidence for ensemble coding in the absence of exemplar coding. At longer presentation times, there was an unexpected dissociation, where exemplar coding increased yet ensemble coding decreased, suggesting that robust information about face identity might suppress ensemble coding. Thus, for face identity, we did not find the classic dissociation-of access to ensemble information in the absence of detailed exemplar information-that has been used to support claims of distinct mechanisms for ensemble and exemplar coding.

Performance and structure of single-mode bosonic codes

NASA Astrophysics Data System (ADS)

Albert, Victor V.; Noh, Kyungjoo; Duivenvoorden, Kasper; Young, Dylan J.; Brierley, R. T.; Reinhold, Philip; Vuillot, Christophe; Li, Linshu; Shen, Chao; Girvin, S. M.; Terhal, Barbara M.; Jiang, Liang

2018-03-01

The early Gottesman, Kitaev, and Preskill (GKP) proposal for encoding a qubit in an oscillator has recently been followed by cat- and binomial-code proposals. Numerically optimized codes have also been proposed, and we introduce codes of this type here. These codes have yet to be compared using the same error model; we provide such a comparison by determining the entanglement fidelity of all codes with respect to the bosonic pure-loss channel (i.e., photon loss) after the optimal recovery operation. We then compare achievable communication rates of the combined encoding-error-recovery channel by calculating the channel's hashing bound for each code. Cat and binomial codes perform similarly, with binomial codes outperforming cat codes at small loss rates. Despite not being designed to protect against the pure-loss channel, GKP codes significantly outperform all other codes for most values of the loss rate. We show that the performance of GKP and some binomial codes increases monotonically with increasing average photon number of the codes. In order to corroborate our numerical evidence of the cat-binomial-GKP order of performance occurring at small loss rates, we analytically evaluate the quantum error-correction conditions of those codes. For GKP codes, we find an essential singularity in the entanglement fidelity in the limit of vanishing loss rate. In addition to comparing the codes, we draw parallels between binomial codes and discrete-variable systems. First, we characterize one- and two-mode binomial as well as multiqubit permutation-invariant codes in terms of spin-coherent states. Such a characterization allows us to introduce check operators and error-correction procedures for binomial codes. Second, we introduce a generalization of spin-coherent states, extending our characterization to qudit binomial codes and yielding a multiqudit code.

LDPC Codes with Minimum Distance Proportional to Block Size

NASA Technical Reports Server (NTRS)

Divsalar, Dariush; Jones, Christopher; Dolinar, Samuel; Thorpe, Jeremy

2009-01-01

Low-density parity-check (LDPC) codes characterized by minimum Hamming distances proportional to block sizes have been demonstrated. Like the codes mentioned in the immediately preceding article, the present codes are error-correcting codes suitable for use in a variety of wireless data-communication systems that include noisy channels. The previously mentioned codes have low decoding thresholds and reasonably low error floors. However, the minimum Hamming distances of those codes do not grow linearly with code-block sizes. Codes that have this minimum-distance property exhibit very low error floors. Examples of such codes include regular LDPC codes with variable degrees of at least 3. Unfortunately, the decoding thresholds of regular LDPC codes are high. Hence, there is a need for LDPC codes characterized by both low decoding thresholds and, in order to obtain acceptably low error floors, minimum Hamming distances that are proportional to code-block sizes. The present codes were developed to satisfy this need. The minimum Hamming distances of the present codes have been shown, through consideration of ensemble-average weight enumerators, to be proportional to code block sizes. As in the cases of irregular ensembles, the properties of these codes are sensitive to the proportion of degree-2 variable nodes. A code having too few such nodes tends to have an iterative decoding threshold that is far from the capacity threshold. A code having too many such nodes tends not to exhibit a minimum distance that is proportional to block size. Results of computational simulations have shown that the decoding thresholds of codes of the present type are lower than those of regular LDPC codes. Included in the simulations were a few examples from a family of codes characterized by rates ranging from low to high and by thresholds that adhere closely to their respective channel capacity thresholds; the simulation results from these examples showed that the codes in question have low error floors as well as low decoding thresholds. As an example, the illustration shows the protograph (which represents the blueprint for overall construction) of one proposed code family for code rates greater than or equal to 1.2. Any size LDPC code can be obtained by copying the protograph structure N times, then permuting the edges. The illustration also provides Field Programmable Gate Array (FPGA) hardware performance simulations for this code family. In addition, the illustration provides minimum signal-to-noise ratios (Eb/No) in decibels (decoding thresholds) to achieve zero error rates as the code block size goes to infinity for various code rates. In comparison with the codes mentioned in the preceding article, these codes have slightly higher decoding thresholds.

76 FR 77549 - Lummi Nation-Title 20-Code of Laws-Liquor Code

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-13

... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Lummi Nation--Title 20--Code of Laws--Liquor... amendment to Lummi Nation's Title 20--Code of Laws--Liquor Code. The Code regulates and controls the... this amendment to Title 20--Lummi Nation Code of Laws--Liquor Code by Resolution 2011-038 on March 1...

Error-correction coding

NASA Technical Reports Server (NTRS)

Hinds, Erold W. (Principal Investigator)

1996-01-01

This report describes the progress made towards the completion of a specific task on error-correcting coding. The proposed research consisted of investigating the use of modulation block codes as the inner code of a concatenated coding system in order to improve the overall space link communications performance. The study proposed to identify and analyze candidate codes that will complement the performance of the overall coding system which uses the interleaved RS (255,223) code as the outer code.

Beam-dynamics codes used at DARHT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ekdahl, Jr., Carl August

Several beam simulation codes are used to help gain a better understanding of beam dynamics in the DARHT LIAs. The most notable of these fall into the following categories: for beam production – Tricomp Trak orbit tracking code, LSP Particle in cell (PIC) code, for beam transport and acceleration – XTR static envelope and centroid code, LAMDA time-resolved envelope and centroid code, LSP-Slice PIC code, for coasting-beam transport to target – LAMDA time-resolved envelope code, LSP-Slice PIC code. These codes are also being used to inform the design of Scorpius.

Amino acid codes in mitochondria as possible clues to primitive codes

NASA Technical Reports Server (NTRS)

Jukes, T. H.

1981-01-01

Differences between mitochondrial codes and the universal code indicate that an evolutionary simplification has taken place, rather than a return to a more primitive code. However, these differences make it evident that the universal code is not the only code possible, and therefore earlier codes may have differed markedly from the previous code. The present universal code is probably a 'frozen accident.' The change in CUN codons from leucine to threonine (Neurospora vs. yeast mitochondria) indicates that neutral or near-neutral changes occurred in the corresponding proteins when this code change took place, caused presumably by a mutation in a tRNA gene.

Some partial-unit-memory convolutional codes

NASA Technical Reports Server (NTRS)

Abdel-Ghaffar, K.; Mceliece, R. J.; Solomon, G.

1991-01-01

The results of a study on a class of error correcting codes called partial unit memory (PUM) codes are presented. This class of codes, though not entirely new, has until now remained relatively unexplored. The possibility of using the well developed theory of block codes to construct a large family of promising PUM codes is shown. The performance of several specific PUM codes are compared with that of the Voyager standard (2, 1, 6) convolutional code. It was found that these codes can outperform the Voyager code with little or no increase in decoder complexity. This suggests that there may very well be PUM codes that can be used for deep space telemetry that offer both increased performance and decreased implementational complexity over current coding systems.

Concatenated coding systems employing a unit-memory convolutional code and a byte-oriented decoding algorithm

NASA Technical Reports Server (NTRS)

Lee, L.-N.

1977-01-01

Concatenated coding systems utilizing a convolutional code as the inner code and a Reed-Solomon code as the outer code are considered. In order to obtain very reliable communications over a very noisy channel with relatively modest coding complexity, it is proposed to concatenate a byte-oriented unit-memory convolutional code with an RS outer code whose symbol size is one byte. It is further proposed to utilize a real-time minimal-byte-error probability decoding algorithm, together with feedback from the outer decoder, in the decoder for the inner convolutional code. The performance of the proposed concatenated coding system is studied, and the improvement over conventional concatenated systems due to each additional feature is isolated.