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.

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

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.

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

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.

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.

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.

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.

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.

Full core analysis of IRIS reactor by using MCNPX.

PubMed

Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

2016-07-01

This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code. Copyright © 2016 Elsevier Ltd. All rights reserved.

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.

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

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.

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

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

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

Zehtabian, M; Zaker, N; Sina, S

2015-06-15

Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 whichmore » is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.« less

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.

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.

GEANT4 benchmark with MCNPX and PHITS for activation of concrete

NASA Astrophysics Data System (ADS)

Tesse, Robin; Stichelbaut, Frédéric; Pauly, Nicolas; Dubus, Alain; Derrien, Jonathan

2018-02-01

The activation of concrete is a real problem from the point of view of waste management. Because of the complexity of the issue, Monte Carlo (MC) codes have become an essential tool to its study. But various codes or even nuclear models exist in MC. MCNPX and PHITS have already been validated for shielding studies but GEANT4 is also a suitable solution. In these codes, different models can be considered for a concrete activation study. The Bertini model is not the best model for spallation while BIC and INCL model agrees well with previous results in literature.

FLUKA simulation studies on in-phantom dosimetric parameters of a LINAC-based BNCT

NASA Astrophysics Data System (ADS)

Ghal-Eh, N.; Goudarzi, H.; Rahmani, F.

2017-12-01

The Monte Carlo simulation code, FLUKA version 2011.2c.5, has been used to estimate the in-phantom dosimetric parameters for use in BNCT studies. The in-phantom parameters of a typical Snyder head, which are necessary information prior to any clinical treatment, have been calculated with both FLUKA and MCNPX codes, which exhibit a promising agreement. The results confirm that FLUKA can be regarded as a good alternative for the MCNPX in BNCT dosimetry simulations.

Dosimetric comparison of Monte Carlo codes (EGS4, MCNP, MCNPX) considering external and internal exposures of the Zubal phantom to electron and photon sources.

PubMed

Chiavassa, S; Lemosquet, A; Aubineau-Lanièce, I; de Carlan, L; Clairand, I; Ferrer, L; Bardiès, M; Franck, D; Zankl, M

2005-01-01

This paper aims at comparing dosimetric assessments performed with three Monte Carlo codes: EGS4, MCNP4c2 and MCNPX2.5e, using a realistic voxel phantom, namely the Zubal phantom, in two configurations of exposure. The first one deals with an external irradiation corresponding to the example of a radiological accident. The results are obtained using the EGS4 and the MCNP4c2 codes and expressed in terms of the mean absorbed dose (in Gy per source particle) for brain, lungs, liver and spleen. The second one deals with an internal exposure corresponding to the treatment of a medullary thyroid cancer by 131I-labelled radiopharmaceutical. The results are obtained by EGS4 and MCNPX2.5e and compared in terms of S-values (expressed in mGy per kBq and per hour) for liver, kidney, whole body and thyroid. The results of these two studies are presented and differences between the codes are analysed and discussed.

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.

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.

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.

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.

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

MCNPX simulation of proton dose distribution in homogeneous and CT phantoms

NASA Astrophysics Data System (ADS)

Lee, C. C.; Lee, Y. J.; Tung, C. J.; Cheng, H. W.; Chao, T. C.

2014-02-01

A dose simulation system was constructed based on the MCNPX Monte Carlo package to simulate proton dose distribution in homogeneous and CT phantoms. Conversion from Hounsfield unit of a patient CT image set to material information necessary for Monte Carlo simulation is based on Schneider's approach. In order to validate this simulation system, inter-comparison of depth dose distributions among those obtained from the MCNPX, GEANT4 and FLUKA codes for a 160 MeV monoenergetic proton beam incident normally on the surface of a homogeneous water phantom was performed. For dose validation within the CT phantom, direct comparison with measurement is infeasible. Instead, this study took the approach to indirectly compare the 50% ranges (R50%) along the central axis by our system to the NIST CSDA ranges for beams with 160 and 115 MeV energies. Comparison result within the homogeneous phantom shows good agreement. Differences of simulated R50% among the three codes are less than 1 mm. For results within the CT phantom, the MCNPX simulated water equivalent Req,50% are compatible with the CSDA water equivalent ranges from the NIST database with differences of 0.7 and 4.1 mm for 160 and 115 MeV beams, respectively.

Validation of a personalized dosimetric evaluation tool (Oedipe) for targeted radiotherapy based on the Monte Carlo MCNPX code

NASA Astrophysics Data System (ADS)

Chiavassa, S.; Aubineau-Lanièce, I.; Bitar, A.; Lisbona, A.; Barbet, J.; Franck, D.; Jourdain, J. R.; Bardiès, M.

2006-02-01

Dosimetric studies are necessary for all patients treated with targeted radiotherapy. In order to attain the precision required, we have developed Oedipe, a dosimetric tool based on the MCNPX Monte Carlo code. The anatomy of each patient is considered in the form of a voxel-based geometry created using computed tomography (CT) images or magnetic resonance imaging (MRI). Oedipe enables dosimetry studies to be carried out at the voxel scale. Validation of the results obtained by comparison with existing methods is complex because there are multiple sources of variation: calculation methods (different Monte Carlo codes, point kernel), patient representations (model or specific) and geometry definitions (mathematical or voxel-based). In this paper, we validate Oedipe by taking each of these parameters into account independently. Monte Carlo methodology requires long calculation times, particularly in the case of voxel-based geometries, and this is one of the limits of personalized dosimetric methods. However, our results show that the use of voxel-based geometry as opposed to a mathematically defined geometry decreases the calculation time two-fold, due to an optimization of the MCNPX2.5e code. It is therefore possible to envisage the use of Oedipe for personalized dosimetry in the clinical context of targeted radiotherapy.

The use of the SRIM code for calculation of radiation damage induced by neutrons

NASA Astrophysics Data System (ADS)

Mohammadi, A.; Hamidi, S.; Asadabad, Mohsen Asadi

2017-12-01

Materials subjected to neutron irradiation will being evolve to structural changes by the displacement cascades initiated by nuclear reaction. This study discusses a methodology to compute primary knock-on atoms or PKAs information that lead to radiation damage. A program AMTRACK has been developed for assessing of the PKAs information. This software determines the specifications of recoil atoms (using PTRAC card of MCNPX code) and also the kinematics of interactions. The deterministic method was used for verification of the results of (MCNPX+AMTRACK). The SRIM (formely TRIM) code is capable to compute neutron radiation damage. The PKAs information was extracted by AMTRACK program, which can be used as an input of SRIM codes for systematic analysis of primary radiation damage. Then the Bushehr Nuclear Power Plant (BNPP) radiation damage on reactor pressure vessel is calculated.

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.

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

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.

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.

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.

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

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.

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.

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.

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.

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.

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

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.

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

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

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.

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.

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.

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.

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

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.

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

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.

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

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.

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

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.

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

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.

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

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

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

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

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

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.

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.

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.

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.

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

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

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.

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.

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.

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.

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

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.

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.

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

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.

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.

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

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.

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.

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.

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.

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.

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.

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.

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.

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

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.

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.

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.

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.

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.

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.

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

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

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.

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.

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.

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.

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.

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.

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.

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.

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

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.

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.

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

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.

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

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

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.

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.

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.

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.

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.

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.

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

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.

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.

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

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.

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.

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

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

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

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%.

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.

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

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

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

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.

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

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

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.

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

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

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.

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.

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.

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.

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.

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.

Description of Transport Codes for Space Radiation Shielding

NASA Technical Reports Server (NTRS)

Kim, Myung-Hee Y.; Wilson, John W.; Cucinotta, Francis A.

2011-01-01

This slide presentation describes transport codes and their use for studying and designing space radiation shielding. When combined with risk projection models radiation transport codes serve as the main tool for study radiation and designing shielding. There are three criteria for assessing the accuracy of transport codes: (1) Ground-based studies with defined beams and material layouts, (2) Inter-comparison of transport code results for matched boundary conditions and (3) Comparisons to flight measurements. These three criteria have a very high degree with NASA's HZETRN/QMSFRG.

Los Alamos radiation transport code system on desktop computing platforms

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

Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. Themore » current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines.« less

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.

A velocity-dependent anomalous radial transport model for (2-D, 2-V) kinetic transport codes

NASA Astrophysics Data System (ADS)

Bodi, Kowsik; Krasheninnikov, Sergei; Cohen, Ron; Rognlien, Tom

2008-11-01

Plasma turbulence constitutes a significant part of radial plasma transport in magnetically confined plasmas. This turbulent transport is modeled in the form of anomalous convection and diffusion coefficients in fluid transport codes. There is a need to model the same in continuum kinetic edge codes [such as the (2-D, 2-V) transport version of TEMPEST, NEO, and the code being developed by the Edge Simulation Laboratory] with non-Maxwellian distributions. We present an anomalous transport model with velocity-dependent convection and diffusion coefficients leading to a diagonal transport matrix similar to that used in contemporary fluid transport models (e.g., UEDGE). Also presented are results of simulations corresponding to radial transport due to long-wavelength ExB turbulence using a velocity-independent diffusion coefficient. A BGK collision model is used to enable comparison with fluid transport codes.

SYMTRAN - A Time-dependent Symmetric Tandem Mirror Transport Code

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

Hua, D; Fowler, T

2004-06-15

A time-dependent version of the steady-state radial transport model in symmetric tandem mirrors in Ref. [1] has been coded up and first tests performed. Our code, named SYMTRAN, is an adaptation of the earlier SPHERE code for spheromaks, now modified for tandem mirror physics. Motivated by Post's new concept of kinetic stabilization of symmetric mirrors, it is an extension of the earlier TAMRAC rate-equation code omitting radial transport [2], which successfully accounted for experimental results in TMX. The SYMTRAN code differs from the earlier tandem mirror radial transport code TMT in that our code is focused on axisymmetric tandem mirrorsmore » and classical diffusion, whereas TMT emphasized non-ambipolar transport in TMX and MFTF-B due to yin-yang plugs and non-symmetric transitions between the plugs and axisymmetric center cell. Both codes exhibit interesting but different non-linear behavior.« less

Comparison of space radiation calculations for deterministic and Monte Carlo transport codes

NASA Astrophysics Data System (ADS)

Lin, Zi-Wei; Adams, James; Barghouty, Abdulnasser; Randeniya, Sharmalee; Tripathi, Ram; Watts, John; Yepes, Pablo

For space radiation protection of astronauts or electronic equipments, it is necessary to develop and use accurate radiation transport codes. Radiation transport codes include deterministic codes, such as HZETRN from NASA and UPROP from the Naval Research Laboratory, and Monte Carlo codes such as FLUKA, the Geant4 toolkit and HETC-HEDS. The deterministic codes and Monte Carlo codes complement each other in that deterministic codes are very fast while Monte Carlo codes are more elaborate. Therefore it is important to investigate how well the results of deterministic codes compare with those of Monte Carlo transport codes and where they differ. In this study we evaluate these different codes in their space radiation applications by comparing their output results in the same given space radiation environments, shielding geometry and material. Typical space radiation environments such as the 1977 solar minimum galactic cosmic ray environment are used as the well-defined input, and simple geometries made of aluminum, water and/or polyethylene are used to represent the shielding material. We then compare various outputs of these codes, such as the dose-depth curves and the flux spectra of different fragments and other secondary particles. These comparisons enable us to learn more about the main differences between these space radiation transport codes. At the same time, they help us to learn the qualitative and quantitative features that these transport codes have in common.

Modeling anomalous radial transport in kinetic transport codes

NASA Astrophysics Data System (ADS)

Bodi, K.; Krasheninnikov, S. I.; Cohen, R. H.; Rognlien, T. D.

2009-11-01

Anomalous transport is typically the dominant component of the radial transport in magnetically confined plasmas, where the physical origin of this transport is believed to be plasma turbulence. A model is presented for anomalous transport that can be used in continuum kinetic edge codes like TEMPEST, NEO and the next-generation code being developed by the Edge Simulation Laboratory. The model can also be adapted to particle-based codes. It is demonstrated that the model with a velocity-dependent diffusion and convection terms can match a diagonal gradient-driven transport matrix as found in contemporary fluid codes, but can also include off-diagonal effects. The anomalous transport model is also combined with particle drifts and a particle/energy-conserving Krook collision operator to study possible synergistic effects with neoclassical transport. For the latter study, a velocity-independent anomalous diffusion coefficient is used to mimic the effect of long-wavelength ExB turbulence.

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.

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

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

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

CRUNCH_PARALLEL

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

Shumaker, Dana E.; Steefel, Carl I.

The code CRUNCH_PARALLEL is a parallel version of the CRUNCH code. CRUNCH code version 2.0 was previously released by LLNL, (UCRL-CODE-200063). Crunch is a general purpose reactive transport code developed by Carl Steefel and Yabusake (Steefel Yabsaki 1996). The code handles non-isothermal transport and reaction in one, two, and three dimensions. The reaction algorithm is generic in form, handling an arbitrary number of aqueous and surface complexation as well as mineral dissolution/precipitation. A standardized database is used containing thermodynamic and kinetic data. The code includes advective, dispersive, and diffusive transport.

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

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.

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.

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.

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

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.

A Deterministic Transport Code for Space Environment Electrons

NASA Technical Reports Server (NTRS)

Nealy, John E.; Chang, C. K.; Norman, Ryan B.; Blattnig, Steve R.; Badavi, Francis F.; Adamczyk, Anne M.

2010-01-01

A deterministic computational procedure has been developed to describe transport of space environment electrons in various shield media. This code is an upgrade and extension of an earlier electron code. Whereas the former code was formulated on the basis of parametric functions derived from limited laboratory data, the present code utilizes well established theoretical representations to describe the relevant interactions and transport processes. The shield material specification has been made more general, as have the pertinent cross sections. A combined mean free path and average trajectory approach has been used in the transport formalism. Comparisons with Monte Carlo calculations are presented.

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.

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.

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.

Comparison of Space Radiation Calculations from Deterministic and Monte Carlo Transport Codes

NASA Technical Reports Server (NTRS)

Adams, J. H.; Lin, Z. W.; Nasser, A. F.; Randeniya, S.; Tripathi, r. K.; Watts, J. W.; Yepes, P.

2010-01-01

The presentation outline includes motivation, radiation transport codes being considered, space radiation cases being considered, results for slab geometry, results from spherical geometry, and summary. ///////// main physics in radiation transport codes hzetrn uprop fluka geant4, slab geometry, spe, gcr,

Implementation of an anomalous radial transport model for continuum kinetic edge codes

NASA Astrophysics Data System (ADS)

Bodi, K.; Krasheninnikov, S. I.; Cohen, R. H.; Rognlien, T. D.

2007-11-01

Radial plasma transport in magnetic fusion devices is often dominated by plasma turbulence compared to neoclassical collisional transport. Continuum kinetic edge codes [such as the (2d,2v) transport version of TEMPEST and also EGK] compute the collisional transport directly, but there is a need to model the anomalous transport from turbulence for long-time transport simulations. Such a model is presented and results are shown for its implementation in the TEMPEST gyrokinetic edge code. The model includes velocity-dependent convection and diffusion coefficients expressed as a Hermite polynominals in velocity. The specification of the Hermite coefficients can be set, e.g., by specifying the ratio of particle and energy transport as in fluid transport codes. The anomalous transport terms preserve the property of no particle flux into unphysical regions of velocity space. TEMPEST simulations are presented showing the separate control of particle and energy anomalous transport, and comparisons are made with neoclassical transport also included.

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

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

Benchmarking NNWSI flow and transport codes: COVE 1 results

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

Hayden, N.K.

1985-06-01

The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of themore » codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs.« less

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.

Modification and benchmarking of MCNP for low-energy tungsten spectra.

PubMed

Mercier, J R; Kopp, D T; McDavid, W D; Dove, S B; Lancaster, J L; Tucker, D M

2000-12-01

The MCNP Monte Carlo radiation transport code was modified for diagnostic medical physics applications. In particular, the modified code was thoroughly benchmarked for the production of polychromatic tungsten x-ray spectra in the 30-150 kV range. Validating the modified code for coupled electron-photon transport with benchmark spectra was supplemented with independent electron-only and photon-only transport benchmarks. Major revisions to the code included the proper treatment of characteristic K x-ray production and scoring, new impact ionization cross sections, and new bremsstrahlung cross sections. Minor revisions included updated photon cross sections, electron-electron bremsstrahlung production, and K x-ray yield. The modified MCNP code is benchmarked to electron backscatter factors, x-ray spectra production, and primary and scatter photon transport.

76 FR 2744 - Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-14

... DEPARTMENT OF TRANSPORTATION Office of the Secretary Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation AGENCY: Office of the Secretary, Department of Transportation..., their agents, and third party sellers of air transportation in view of recent amendments to 49 U.S.C...

Computer Code for Transportation Network Design and Analysis

DOT National Transportation Integrated Search

1977-01-01

This document describes the results of research into the application of the mathematical programming technique of decomposition to practical transportation network problems. A computer code called Catnap (for Control Analysis Transportation Network A...

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

User's manual for a material transport code on the Octopus Computer Network

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

Naymik, T.G.; Mendez, G.D.

1978-09-15

A code to simulate material transport through porous media was developed at Oak Ridge National Laboratory. This code has been modified and adapted for use at Lawrence Livermore Laboratory. This manual, in conjunction with report ORNL-4928, explains the input, output, and execution of the code on the Octopus Computer Network.

Comparison of heavy-ion transport simulations: Collision integral in a box

NASA Astrophysics Data System (ADS)

Zhang, Ying-Xun; Wang, Yong-Jia; Colonna, Maria; Danielewicz, Pawel; Ono, Akira; Tsang, Manyee Betty; Wolter, Hermann; Xu, Jun; Chen, Lie-Wen; Cozma, Dan; Feng, Zhao-Qing; Das Gupta, Subal; Ikeno, Natsumi; Ko, Che-Ming; Li, Bao-An; Li, Qing-Feng; Li, Zhu-Xia; Mallik, Swagata; Nara, Yasushi; Ogawa, Tatsuhiko; Ohnishi, Akira; Oliinychenko, Dmytro; Papa, Massimo; Petersen, Hannah; Su, Jun; Song, Taesoo; Weil, Janus; Wang, Ning; Zhang, Feng-Shou; Zhang, Zhen

2018-03-01

Simulations by transport codes are indispensable to extract valuable physical information from heavy-ion collisions. In order to understand the origins of discrepancies among different widely used transport codes, we compare 15 such codes under controlled conditions of a system confined to a box with periodic boundary, initialized with Fermi-Dirac distributions at saturation density and temperatures of either 0 or 5 MeV. In such calculations, one is able to check separately the different ingredients of a transport code. In this second publication of the code evaluation project, we only consider the two-body collision term; i.e., we perform cascade calculations. When the Pauli blocking is artificially suppressed, the collision rates are found to be consistent for most codes (to within 1 % or better) with analytical results, or completely controlled results of a basic cascade code. In orderto reach that goal, it was necessary to eliminate correlations within the same pair of colliding particles that can be present depending on the adopted collision prescription. In calculations with active Pauli blocking, the blocking probability was found to deviate from the expected reference values. The reason is found in substantial phase-space fluctuations and smearing tied to numerical algorithms and model assumptions in the representation of phase space. This results in the reduction of the blocking probability in most transport codes, so that the simulated system gradually evolves away from the Fermi-Dirac toward a Boltzmann distribution. Since the numerical fluctuations are weaker in the Boltzmann-Uehling-Uhlenbeck codes, the Fermi-Dirac statistics is maintained there for a longer time than in the quantum molecular dynamics codes. As a result of this investigation, we are able to make judgements about the most effective strategies in transport simulations for determining the collision probabilities and the Pauli blocking. Investigation in a similar vein of other ingredients in transport calculations, like the mean-field propagation or the production of nucleon resonances and mesons, will be discussed in the future publications.

Verification and benchmark testing of the NUFT computer code

NASA Astrophysics Data System (ADS)

Lee, K. H.; Nitao, J. J.; Kulshrestha, A.

1993-10-01

This interim report presents results of work completed in the ongoing verification and benchmark testing of the NUFT (Nonisothermal Unsaturated-saturated Flow and Transport) computer code. NUFT is a suite of multiphase, multicomponent models for numerical solution of thermal and isothermal flow and transport in porous media, with application to subsurface contaminant transport problems. The code simulates the coupled transport of heat, fluids, and chemical components, including volatile organic compounds. Grid systems may be cartesian or cylindrical, with one-, two-, or fully three-dimensional configurations possible. In this initial phase of testing, the NUFT code was used to solve seven one-dimensional unsaturated flow and heat transfer problems. Three verification and four benchmarking problems were solved. In the verification testing, excellent agreement was observed between NUFT results and the analytical or quasianalytical solutions. In the benchmark testing, results of code intercomparison were very satisfactory. From these testing results, it is concluded that the NUFT code is ready for application to field and laboratory problems similar to those addressed here. Multidimensional problems, including those dealing with chemical transport, will be addressed in a subsequent report.

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 4 2013-10-01 2013-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 4 2014-10-01 2014-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

49 CFR Appendix C to Part 229 - FRA Locomotive Standards-Code of Defects

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false FRA Locomotive Standards-Code of Defects C Appendix C to Part 229 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Pt. 229, App. C...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 4 2013-10-01 2013-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 4 2014-10-01 2014-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 4 2012-10-01 2012-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 4 2011-10-01 2011-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to Part 215—FRA Freight Car Standards Defect Code The following defect code has been established for use...

Methods of treating complex space vehicle geometry for charged particle radiation transport

NASA Technical Reports Server (NTRS)

Hill, C. W.

1973-01-01

Current methods of treating complex geometry models for space radiation transport calculations are reviewed. The geometric techniques used in three computer codes are outlined. Evaluations of geometric capability and speed are provided for these codes. Although no code development work is included several suggestions for significantly improving complex geometry codes are offered.

Comparisons of anomalous and collisional radial transport with a continuum kinetic edge code

NASA Astrophysics Data System (ADS)

Bodi, K.; Krasheninnikov, S.; Cohen, R.; Rognlien, T.

2009-05-01

Modeling of anomalous (turbulence-driven) radial transport in controlled-fusion plasmas is necessary for long-time transport simulations. Here the focus is continuum kinetic edge codes such as the (2-D, 2-V) transport version of TEMPEST, NEO, and the code being developed by the Edge Simulation Laboratory, but the model also has wider application. Our previously developed anomalous diagonal transport matrix model with velocity-dependent convection and diffusion coefficients allows contact with typical fluid transport models (e.g., UEDGE). Results are presented that combine the anomalous transport model and collisional transport owing to ion drift orbits utilizing a Krook collision operator that conserves density and energy. Comparison is made of the relative magnitudes and possible synergistic effects of the two processes for typical tokamak device parameters.

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.

Light transport feature for SCINFUL.

PubMed

Etaati, G R; Ghal-Eh, N

2008-03-01

An extended version of the scintillator response function prediction code SCINFUL has been developed by incorporating PHOTRACK, a Monte Carlo light transport code. Comparisons of calculated and experimental results for organic scintillators exposed to neutrons show that the extended code improves the predictive capability of SCINFUL.

First ERO2.0 modeling of Be erosion and non-local transport in JET ITER-like wall

NASA Astrophysics Data System (ADS)

Romazanov, J.; Borodin, D.; Kirschner, A.; Brezinsek, S.; Silburn, S.; Huber, A.; Huber, V.; Bufferand, H.; Firdaouss, M.; Brömmel, D.; Steinbusch, B.; Gibbon, P.; Lasa, A.; Borodkina, I.; Eksaeva, A.; Linsmeier, Ch; Contributors, JET

2017-12-01

ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport for applications in fusion research. The code has undergone a significant upgrade (ERO2.0) which allows increasing the simulation volume in order to cover the entire plasma edge of a fusion device, allowing a more self-consistent treatment of impurity transport and comparison with a larger number and variety of experimental diagnostics. In this contribution, the physics-relevant technical innovations of the new code version are described and discussed. The new capabilities of the code are demonstrated by modeling of beryllium (Be) erosion of the main wall during JET limiter discharges. Results for erosion patterns along the limiter surfaces and global Be transport including incident particle distributions are presented. A novel synthetic diagnostic, which mimics experimental wide-angle 2D camera images, is presented and used for validating various aspects of the code, including erosion, magnetic shadowing, non-local impurity transport, and light emission simulation.

HZETRN: A heavy ion/nucleon transport code for space radiations

NASA Technical Reports Server (NTRS)

Wilson, John W.; Chun, Sang Y.; Badavi, Forooz F.; Townsend, Lawrence W.; Lamkin, Stanley L.

1991-01-01

The galactic heavy ion transport code (GCRTRN) and the nucleon transport code (BRYNTRN) are integrated into a code package (HZETRN). The code package is computer efficient and capable of operating in an engineering design environment for manned deep space mission studies. The nuclear data set used by the code is discussed including current limitations. Although the heavy ion nuclear cross sections are assumed constant, the nucleon-nuclear cross sections of BRYNTRN with full energy dependence are used. The relation of the final code to the Boltzmann equation is discussed in the context of simplifying assumptions. Error generation and propagation is discussed, and comparison is made with simplified analytic solutions to test numerical accuracy of the final results. A brief discussion of biological issues and their impact on fundamental developments in shielding technology is given.

ITS version 5.0 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

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

Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theoristsmore » alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.« less

IPOLE - semi-analytic scheme for relativistic polarized radiative transport

NASA Astrophysics Data System (ADS)

Mościbrodzka, M.; Gammie, C. F.

2018-03-01

We describe IPOLE, a new public ray-tracing code for covariant, polarized radiative transport. The code extends the IBOTHROS scheme for covariant, unpolarized transport using two representations of the polarized radiation field: In the coordinate frame, it parallel transports the coherency tensor; in the frame of the plasma it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is implemented to be as spacetime- and coordinate- independent as possible. The emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, IPOLE is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth. We show that the code matches analytic results in flat space, and that it produces results that converge to those produced by Dexter's GRTRANS polarized transport code on a complicated model problem. We expect IPOLE will mainly find applications in modelling Event Horizon Telescope sources, but it may also be useful in other relativistic transport problems such as modelling for the IXPE mission.

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.

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).

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.

SOPHAEROS code development and its application to falcon tests

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

Lajtha, G.; Missirlian, M.; Kissane, M.

1996-12-31

One of the key issues in source-term evaluation in nuclear reactor severe accidents is determination of the transport behavior of fission products released from the degrading core. The SOPHAEROS computer code is being developed to predict fission product transport in a mechanistic way in light water reactor circuits. These applications of the SOPHAEROS code to the Falcon experiments, among others not presented here, indicate that the numerical scheme of the code is robust, and no convergence problems are encountered. The calculation is also very fast being three times longer on a Sun SPARC 5 workstation than real time and typicallymore » {approx} 10 times faster than an identical calculation with the VICTORIA code. The study demonstrates that the SOPHAEROS 1.3 code is a suitable tool for prediction of the vapor chemistry and fission product transport with a reasonable level of accuracy. Furthermore, the fexibility of the code material data bank allows improvement of understanding of fission product transport and deposition in the circuit. Performing sensitivity studies with different chemical species or with different properties (saturation pressure, chemical equilibrium constants) is very straightforward.« less

Intact coding region of the serotonin transporter gene in obsessive-compulsive disorder

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

Altemus, M.; Murphy, D.L.; Greenberg, B.

1996-07-26

Epidemiologic studies indicate that obsessive-compulsive disorder is genetically transmitted in some families, although no genetic abnormalities have been identified in individuals with this disorder. The selective response of obsessive-compulsive disorder to treatment with agents which block serotonin reuptake suggests the gene coding for the serotonin transporter as a candidate gene. The primary structure of the serotonin-transporter coding region was sequenced in 22 patients with obsessive-compulsive disorder, using direct PCR sequencing of cDNA synthesized from platelet serotonin-transporter mRNA. No variations in amino acid sequence were found among the obsessive-compulsive disorder patients or healthy controls. These results do not support a rolemore » for alteration in the primary structure of the coding region of the serotonin-transporter gene in the pathogenesis of obsessive-compulsive disorder. 27 refs.« less

Evolution of Structure and Composition in Saturn's Rings Due to Ballistic Transport of Micrometeoroid Impact Ejecta

NASA Astrophysics Data System (ADS)

Estrada, P. R.; Durisen, R. H.; Cuzzi, J. N.

2014-04-01

We introduce improved numerical techniques for simulating the structural and compositional evolution of planetary rings due to micrometeoroid bombardment and subsequent ballistic transport of impact ejecta. Our current, robust code, which is based on the original structural code of [1] and on the pollution transport code of [3], is capable of modeling structural changes and pollution transport simultaneously over long times on both local and global scales. We provide demonstrative simulations to compare with, and extend upon previous work, as well as examples of how ballistic transport can maintain the observed structure in Saturn's rings using available Cassini occultation optical depth data.

Comparing Turbulence Simulation with Experiment in DIII-D

NASA Astrophysics Data System (ADS)

Ross, D. W.; Bravenec, R. V.; Dorland, W.; Beer, M. A.; Hammett, G. W.; McKee, G. R.; Murakami, M.; Jackson, G. L.

2000-10-01

Gyrofluid simulations of DIII-D discharges with the GRYFFIN code(D. W. Ross et al.), Transport Task Force Workshop, Burlington, VT, (2000). are compared with transport and fluctuation measurements. The evolution of confinement-improved discharges(G. R. McKee et al.), Phys. Plasmas 7, 1870 (200) is studied at early times following impurity injection, when EXB rotational shear plays a small role. The ion thermal transport predicted by the code is consistent with the experimental values. Experimentally, changes in density profiles resulting from the injection of neon, lead to reduction in fluctuation levels and transport following the injection. This triggers subsequent changes in the shearing rate that further reduce the turbulence.(M. Murakami et al.), European Physical Society, Budapest (2000); M. Murakami et al., this meeting. Estimated uncertainties in the plasma profiles, however, make it difficult to simulate these reductions with the code. These cases will also be studied with the GS2 gyrokinetic code.

Gyrokinetic Particle Simulation of Turbulent Transport in Burning Plasmas (GPS - TTBP) Final Report

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

Chame, Jacqueline

2011-05-27

The goal of this project is the development of the Gyrokinetic Toroidal Code (GTC) Framework and its applications to problems related to the physics of turbulence and turbulent transport in tokamaks,. The project involves physics studies, code development, noise effect mitigation, supporting computer science efforts, diagnostics and advanced visualizations, verification and validation. Its main scientific themes are mesoscale dynamics and non-locality effects on transport, the physics of secondary structures such as zonal flows, and strongly coherent wave-particle interaction phenomena at magnetic precession resonances. Special emphasis is placed on the implications of these themes for rho-star and current scalings and formore » the turbulent transport of momentum. GTC-TTBP also explores applications to electron thermal transport, particle transport; ITB formation and cross-cuts such as edge-core coupling, interaction of energetic particles with turbulence and neoclassical tearing mode trigger dynamics. Code development focuses on major initiatives in the development of full-f formulations and the capacity to simulate flux-driven transport. In addition to the full-f -formulation, the project includes the development of numerical collision models and methods for coarse graining in phase space. Verification is pursued by linear stability study comparisons with the FULL and HD7 codes and by benchmarking with the GKV, GYSELA and other gyrokinetic simulation codes. Validation of gyrokinetic models of ion and electron thermal transport is pursed by systematic stressing comparisons with fluctuation and transport data from the DIII-D and NSTX tokamaks. The physics and code development research programs are supported by complementary efforts in computer sciences, high performance computing, and data management.« less

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.

A comparative study of space radiation organ doses and associated cancer risks using PHITS and HZETRN.

PubMed

Bahadori, Amir A; Sato, Tatsuhiko; Slaba, Tony C; Shavers, Mark R; Semones, Edward J; Van Baalen, Mary; Bolch, Wesley E

2013-10-21

NASA currently uses one-dimensional deterministic transport to generate values of the organ dose equivalent needed to calculate stochastic radiation risk following crew space exposures. In this study, organ absorbed doses and dose equivalents are calculated for 50th percentile male and female astronaut phantoms using both the NASA High Charge and Energy Transport Code to perform one-dimensional deterministic transport and the Particle and Heavy Ion Transport Code System to perform three-dimensional Monte Carlo transport. Two measures of radiation risk, effective dose and risk of exposure-induced death (REID) are calculated using the organ dose equivalents resulting from the two methods of radiation transport. For the space radiation environments and simplified shielding configurations considered, small differences (<8%) in the effective dose and REID are found. However, for the galactic cosmic ray (GCR) boundary condition, compensating errors are observed, indicating that comparisons between the integral measurements of complex radiation environments and code calculations can be misleading. Code-to-code benchmarks allow for the comparison of differential quantities, such as secondary particle differential fluence, to provide insight into differences observed in integral quantities for particular components of the GCR spectrum.

A comparative study of space radiation organ doses and associated cancer risks using PHITS and HZETRN

NASA Astrophysics Data System (ADS)

Bahadori, Amir A.; Sato, Tatsuhiko; Slaba, Tony C.; Shavers, Mark R.; Semones, Edward J.; Van Baalen, Mary; Bolch, Wesley E.

2013-10-01

NASA currently uses one-dimensional deterministic transport to generate values of the organ dose equivalent needed to calculate stochastic radiation risk following crew space exposures. In this study, organ absorbed doses and dose equivalents are calculated for 50th percentile male and female astronaut phantoms using both the NASA High Charge and Energy Transport Code to perform one-dimensional deterministic transport and the Particle and Heavy Ion Transport Code System to perform three-dimensional Monte Carlo transport. Two measures of radiation risk, effective dose and risk of exposure-induced death (REID) are calculated using the organ dose equivalents resulting from the two methods of radiation transport. For the space radiation environments and simplified shielding configurations considered, small differences (<8%) in the effective dose and REID are found. However, for the galactic cosmic ray (GCR) boundary condition, compensating errors are observed, indicating that comparisons between the integral measurements of complex radiation environments and code calculations can be misleading. Code-to-code benchmarks allow for the comparison of differential quantities, such as secondary particle differential fluence, to provide insight into differences observed in integral quantities for particular components of the GCR spectrum.

MILSTAMP TACs: Military Standard Transportation and Movement Procedures Transportation Account Codes. Volume 2

DTIC Science & Technology

1987-02-15

this chapter. NO - If shipment is not second des - tination transportation , obtain fund cite per yes response for question 2 above. 4. For Direct Support...return . . . . . . . . .0 . . . . . . . a. . .. A820 (8) LOGAIR/QUICKTRANS. Transportation Account Codes de - signed herein are applicable to the...oo~• na~- Transportation Tis Document Contains Tasotto Missing Page/s That Are Unavailable In The And Original Document Movement sdocument has boon

Transport calculations and accelerator experiments needed for radiation risk assessment in space.

PubMed

Sihver, Lembit

2008-01-01

The major uncertainties on space radiation risk estimates in humans are associated to the poor knowledge of the biological effects of low and high LET radiation, with a smaller contribution coming from the characterization of space radiation field and its primary interactions with the shielding and the human body. However, to decrease the uncertainties on the biological effects and increase the accuracy of the risk coefficients for charged particles radiation, the initial charged-particle spectra from the Galactic Cosmic Rays (GCRs) and the Solar Particle Events (SPEs), and the radiation transport through the shielding material of the space vehicle and the human body, must be better estimated Since it is practically impossible to measure all primary and secondary particles from all possible position-projectile-target-energy combinations needed for a correct risk assessment in space, accurate particle and heavy ion transport codes must be used. These codes are also needed when estimating the risk for radiation induced failures in advanced microelectronics, such as single-event effects, etc., and the efficiency of different shielding materials. It is therefore important that the models and transport codes will be carefully benchmarked and validated to make sure they fulfill preset accuracy criteria, e.g. to be able to predict particle fluence, dose and energy distributions within a certain accuracy. When validating the accuracy of the transport codes, both space and ground based accelerator experiments are needed The efficiency of passive shielding and protection of electronic devices should also be tested in accelerator experiments and compared to simulations using different transport codes. In this paper different multipurpose particle and heavy ion transport codes will be presented, different concepts of shielding and protection discussed, as well as future accelerator experiments needed for testing and validating codes and shielding materials.

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE PAGES

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson; ...

2018-06-14

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

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

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

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.

Full 3D visualization tool-kit for Monte Carlo and deterministic transport codes

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

Frambati, S.; Frignani, M.

2012-07-01

We propose a package of tools capable of translating the geometric inputs and outputs of many Monte Carlo and deterministic radiation transport codes into open source file formats. These tools are aimed at bridging the gap between trusted, widely-used radiation analysis codes and very powerful, more recent and commonly used visualization software, thus supporting the design process and helping with shielding optimization. Three main lines of development were followed: mesh-based analysis of Monte Carlo codes, mesh-based analysis of deterministic codes and Monte Carlo surface meshing. The developed kit is considered a powerful and cost-effective tool in the computer-aided design formore » radiation transport code users of the nuclear world, and in particular in the fields of core design and radiation analysis. (authors)« less

Radiation Transport Tools for Space Applications: A Review

NASA Technical Reports Server (NTRS)

Jun, Insoo; Evans, Robin; Cherng, Michael; Kang, Shawn

2008-01-01

This slide presentation contains a brief discussion of nuclear transport codes widely used in the space radiation community for shielding and scientific analyses. Seven radiation transport codes that are addressed. The two general methods (i.e., Monte Carlo Method, and the Deterministic Method) are briefly reviewed.

Comparison of Transport Codes, HZETRN, HETC and FLUKA, Using 1977 GCR Solar Minimum Spectra

NASA Technical Reports Server (NTRS)

Heinbockel, John H.; Slaba, Tony C.; Tripathi, Ram K.; Blattnig, Steve R.; Norbury, John W.; Badavi, Francis F.; Townsend, Lawrence W.; Handler, Thomas; Gabriel, Tony A.; Pinsky, Lawrence S.;

Monitoring Cosmic Radiation Risk: Comparisons between Observations and Predictive Codes for Naval Aviation

DTIC Science & Technology

2009-01-01

proton PARMA PHITS -based Analytical Radiation Model in the Atmosphere PCAIRE Predictive Code for Aircrew Radiation Exposure PHITS Particle and...radiation transport code utilized is called PARMA ( PHITS based Analytical Radiation Model in the Atmosphere) [36]. The particle fluxes calculated from the...same dose equivalent coefficient regulations from the ICRP-60 regulations. As a result, the transport codes utilized by EXPACS ( PHITS ) and CARI-6

Monitoring Cosmic Radiation Risk: Comparisons Between Observations and Predictive Codes for Naval Aviation

DTIC Science & Technology

2009-07-05

proton PARMA PHITS -based Analytical Radiation Model in the Atmosphere PCAIRE Predictive Code for Aircrew Radiation Exposure PHITS Particle and Heavy...transport code utilized is called PARMA ( PHITS based Analytical Radiation Model in the Atmosphere) [36]. The particle fluxes calculated from the input...dose equivalent coefficient regulations from the ICRP-60 regulations. As a result, the transport codes utilized by EXPACS ( PHITS ) and CARI-6 (PARMA

Moving from Batch to Field Using the RT3D Reactive Transport Modeling System

NASA Astrophysics Data System (ADS)

Clement, T. P.; Gautam, T. R.

2002-12-01

The public domain reactive transport code RT3D (Clement, 1997) is a general-purpose numerical code for solving coupled, multi-species reactive transport in saturated groundwater systems. The code uses MODFLOW to simulate flow and several modules of MT3DMS to simulate the advection and dispersion processes. RT3D employs the operator-split strategy which allows the code solve the coupled reactive transport problem in a modular fashion. The coupling between reaction and transport is defined through a separate module where the reaction equations are specified. The code supports a versatile user-defined reaction option that allows users to define their own reaction system through a Fortran-90 subroutine, known as the RT3D-reaction package. Further a utility code, known as BATCHRXN, allows the users to independently test and debug their reaction package. To analyze a new reaction system at a batch scale, users should first run BATCHRXN to test the ability of their reaction package to model the batch data. After testing, the reaction package can simply be ported to the RT3D environment to study the model response under 1-, 2-, or 3-dimensional transport conditions. This paper presents example problems that demonstrate the methods for moving from batch to field-scale simulations using BATCHRXN and RT3D codes. The first example describes a simple first-order reaction system for simulating the sequential degradation of Tetrachloroethene (PCE) and its daughter products. The second example uses a relatively complex reaction system for describing the multiple degradation pathways of Tetrachloroethane (PCA) and its daughter products. References 1) Clement, T.P, RT3D - A modular computer code for simulating reactive multi-species transport in 3-Dimensional groundwater aquifers, Battelle Pacific Northwest National Laboratory Research Report, PNNL-SA-28967, September, 1997. Available at: http://bioprocess.pnl.gov/rt3d.htm.

Generic reactive transport codes as flexible tools to integrate soil organic matter degradation models with water, transport and geochemistry in soils

NASA Astrophysics Data System (ADS)

Jacques, Diederik; Gérard, Fréderic; Mayer, Uli; Simunek, Jirka; Leterme, Bertrand

2016-04-01

A large number of organic matter degradation, CO2 transport and dissolved organic matter models have been developed during the last decades. However, organic matter degradation models are in many cases strictly hard-coded in terms of organic pools, degradation kinetics and dependency on environmental variables. The scientific input of the model user is typically limited to the adjustment of input parameters. In addition, the coupling with geochemical soil processes including aqueous speciation, pH-dependent sorption and colloid-facilitated transport are not incorporated in many of these models, strongly limiting the scope of their application. Furthermore, the most comprehensive organic matter degradation models are combined with simplified representations of flow and transport processes in the soil system. We illustrate the capability of generic reactive transport codes to overcome these shortcomings. The formulations of reactive transport codes include a physics-based continuum representation of flow and transport processes, while biogeochemical reactions can be described as equilibrium processes constrained by thermodynamic principles and/or kinetic reaction networks. The flexibility of these type of codes allows for straight-forward extension of reaction networks, permits the inclusion of new model components (e.g.: organic matter pools, rate equations, parameter dependency on environmental conditions) and in such a way facilitates an application-tailored implementation of organic matter degradation models and related processes. A numerical benchmark involving two reactive transport codes (HPx and MIN3P) demonstrates how the process-based simulation of transient variably saturated water flow (Richards equation), solute transport (advection-dispersion equation), heat transfer and diffusion in the gas phase can be combined with a flexible implementation of a soil organic matter degradation model. The benchmark includes the production of leachable organic matter and inorganic carbon in the aqueous and gaseous phases, as well as different decomposition functions with first-order, linear dependence or nonlinear dependence on a biomass pool. In addition, we show how processes such as local bioturbation (bio-diffusion) can be included implicitly through a Fickian formulation of transport of soil organic matter. Coupling soil organic matter models with generic and flexible reactive transport codes offers a valuable tool to enhance insights into coupled physico-chemical processes at different scales within the scope of C-biogeochemical cycles, possibly linked with other chemical elements such as plant nutrients and pollutants.

49 CFR 171.25 - Additional requirements for the use of the IMDG Code.

Code of Federal Regulations, 2010 CFR

2010-10-01

...) Packages containing primary lithium batteries and cells that are transported in accordance with Special Provision 188 of the IMDG Code must be marked “PRIMARY LITHIUM BATTERIES—FORBIDDEN FOR TRANSPORT ABOARD PASSENGER AIRCRAFT” or “LITHIUM METAL BATTERIES—FORBIDDEN FOR TRANSPORT ABOARD PASSENGER AIRCRAFT.” This...

77 FR 18716 - Transportation Security Administration Postal Zip Code Change; Technical Amendment

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-28

... organizational changes and it has no substantive effect on the public. DATES: Effective March 28, 2012. FOR... No. 1572-9] Transportation Security Administration Postal Zip Code Change; Technical Amendment AGENCY: Transportation Security Administration, DHS. ACTION: Final rule. SUMMARY: This rule is a technical change to...

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.

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.

Thermodynamic and transport properties of gaseous tetrafluoromethane in chemical equilibrium

NASA Technical Reports Server (NTRS)

Hunt, J. L.; Boney, L. R.

1973-01-01

Equations and in computer code are presented for the thermodynamic and transport properties of gaseous, undissociated tetrafluoromethane (CF4) in chemical equilibrium. The computer code calculates the thermodynamic and transport properties of CF4 when given any two of five thermodynamic variables (entropy, temperature, volume, pressure, and enthalpy). Equilibrium thermodynamic and transport property data are tabulated and pressure-enthalpy diagrams are presented.

High-fidelity plasma codes for burn physics

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

Cooley, James; Graziani, Frank; Marinak, Marty

Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental datamore » and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.« less

Development of a 1.5D plasma transport code for coupling to full orbit runaway electron simulations

NASA Astrophysics Data System (ADS)

Lore, J. D.; Del Castillo-Negrete, D.; Baylor, L.; Carbajal, L.

2017-10-01

A 1.5D (1D radial transport + 2D equilibrium geometry) plasma transport code is being developed to simulate runaway electron generation, mitigation, and avoidance by coupling to the full-orbit kinetic electron transport code KORC. The 1.5D code solves the time-dependent 1D flux surface averaged transport equations with sources for plasma density, pressure, and poloidal magnetic flux, along with the Grad-Shafranov equilibrium equation for the 2D flux surface geometry. Disruption mitigation is simulated by introducing an impurity neutral gas `pellet', with impurity densities and electron cooling calculated from ionization, recombination, and line emission rate coefficients. Rapid cooling of the electrons increases the resistivity, inducing an electric field which can be used as an input to KORC. The runaway electron current is then included in the parallel Ohm's law in the transport equations. The 1.5D solver will act as a driver for coupled simulations to model effects such as timescales for thermal quench, runaway electron generation, and pellet impurity mixtures for runaway avoidance. Current progress on the code and details of the numerical algorithms will be presented. Work supported by the US DOE under DE-AC05-00OR22725.

Benchmarking a Visual-Basic based multi-component one-dimensional reactive transport modeling tool

NASA Astrophysics Data System (ADS)

Torlapati, Jagadish; Prabhakar Clement, T.

2013-01-01

We present the details of a comprehensive numerical modeling tool, RT1D, which can be used for simulating biochemical and geochemical reactive transport problems. The code can be run within the standard Microsoft EXCEL Visual Basic platform, and it does not require any additional software tools. The code can be easily adapted by others for simulating different types of laboratory-scale reactive transport experiments. We illustrate the capabilities of the tool by solving five benchmark problems with varying levels of reaction complexity. These literature-derived benchmarks are used to highlight the versatility of the code for solving a variety of practical reactive transport problems. The benchmarks are described in detail to provide a comprehensive database, which can be used by model developers to test other numerical codes. The VBA code presented in the study is a practical tool that can be used by laboratory researchers for analyzing both batch and column datasets within an EXCEL platform.

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.

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

Development of a new version of the Vehicle Protection Factor Code (VPF3)

NASA Astrophysics Data System (ADS)

Jamieson, Terrance J.

1990-10-01

The Vehicle Protection Factor (VPF) Code is an engineering tool for estimating radiation protection afforded by armoured vehicles and other structures exposed to neutron and gamma ray radiation from fission, thermonuclear, and fusion sources. A number of suggestions for modifications have been offered by users of early versions of the code. These include: implementing some of the more advanced features of the air transport rating code, ATR5, used to perform the air over ground radiation transport analyses; allowing the ability to study specific vehicle orientations within the free field; implementing an adjoint transport scheme to reduce the number of transport runs required; investigating the possibility of accelerating the transport scheme; and upgrading the computer automated design (CAD) package used by VPF. The generation of radiation free field fluences for infinite air geometries as required for aircraft analysis can be accomplished by using ATR with the air over ground correction factors disabled. Analysis of the effects of fallout bearing debris clouds on aircraft will require additional modelling of VPF.

A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes

NASA Technical Reports Server (NTRS)

Schnittman, Jeremy David; Krolik, Julian H.

2013-01-01

We present a new code for radiation transport around Kerr black holes, including arbitrary emission and absorption mechanisms, as well as electron scattering and polarization. The code is particularly useful for analyzing accretion flows made up of optically thick disks and optically thin coronae. We give a detailed description of the methods employed in the code and also present results from a number of numerical tests to assess its accuracy and convergence.

Development of Parallel Computing Framework to Enhance Radiation Transport Code Capabilities for Rare Isotope Beam Facility Design

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

Kostin, Mikhail; Mokhov, Nikolai; Niita, Koji

A parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. It is intended to be used with older radiation transport codes implemented in Fortran77, Fortran 90 or C. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was developed and tested in conjunction with the MARS15 code. It is possible to use it with other codes such as PHITS, FLUKA andmore » MCNP after certain adjustments. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. The framework corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.« less

METHES: A Monte Carlo collision code for the simulation of electron transport in low temperature plasmas

NASA Astrophysics Data System (ADS)

Rabie, M.; Franck, C. M.

2016-06-01

We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.

STELLTRANS: A Transport Analysis Suite for Stellarators

NASA Astrophysics Data System (ADS)

Mittelstaedt, Joseph; Lazerson, Samuel; Pablant, Novimir; Weir, Gavin; W7-X Team

2016-10-01

The stellarator transport code STELLTRANS allows us to better analyze the power balance in W7-X. Although profiles of temperature and density are measured experimentally, geometrical factors are needed in conjunction with these measurements to properly analyze heat flux densities in stellarators. The STELLTRANS code interfaces with VMEC to find an equilibrium flux surface configuration and with TRAVIS to determine the RF heating and current drive in the plasma. Stationary transport equations are then considered which are solved using a boundary value differential equation solver. The equations and quantities considered are averaged over flux surfaces to reduce the system to an essentially one dimensional problem. We have applied this code to data from W-7X and were able to calculate the heat flux coefficients. We will also present extensions of the code to a predictive capacity which would utilize DKES to find neoclassical transport coefficients to update the temperature and density profiles.

Electron transport model of dielectric charging

NASA Technical Reports Server (NTRS)

Beers, B. L.; Hwang, H. C.; Lin, D. L.; Pine, V. W.

1979-01-01

A computer code (SCCPOEM) was assembled to describe the charging of dielectrics due to irradiation by electrons. The primary purpose for developing the code was to make available a convenient tool for studying the internal fields and charge densities in electron-irradiated dielectrics. The code, which is based on the primary electron transport code POEM, is applicable to arbitrary dielectrics, source spectra, and current time histories. The code calculations are illustrated by a series of semianalytical solutions. Calculations to date suggest that the front face electric field is insufficient to cause breakdown, but that bulk breakdown fields can easily be exceeded.

Capabilities overview of the MORET 5 Monte Carlo code

NASA Astrophysics Data System (ADS)

Cochet, B.; Jinaphanh, A.; Heulers, L.; Jacquet, O.

2014-06-01

The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations, describe the materials, define their own tallies in order to analyse the results. The MORET code has been initially designed to perform calculations for criticality safety assessments. New features has been introduced in the MORET 5 code to expand its use for reactor applications. This paper presents an overview of the MORET 5 code capabilities, going through the description of materials, the geometry modelling, the transport simulation and the definition of the outputs.

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

Aleman, S.E.

This report documents a finite element code designed to model subsurface flow and contaminant transport, named FACT. FACT is a transient three-dimensional, finite element code designed to simulate isothermal groundwater flow, moisture movement, and solute transport in variably saturated and fully saturated subsurface porous media.

Adaptive Nodal Transport Methods for Reactor Transient Analysis

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

Thomas Downar; E. Lewis

2005-08-31

Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

The Athena Astrophysical MHD Code in Cylindrical Geometry

NASA Astrophysics Data System (ADS)

Skinner, M. A.; Ostriker, E. C.

2011-10-01

We have developed a method for implementing cylindrical coordinates in the Athena MHD code (Skinner & Ostriker 2010). The extension has been designed to alter the existing Cartesian-coordinates code (Stone et al. 2008) as minimally and transparently as possible. The numerical equations in cylindrical coordinates are formulated to maintain consistency with constrained transport, a central feature of the Athena algorithm, while making use of previously implemented code modules such as the eigensystems and Riemann solvers. Angular-momentum transport, which is critical in astrophysical disk systems dominated by rotation, is treated carefully. We describe modifications for cylindrical coordinates of the higher-order spatial reconstruction and characteristic evolution steps as well as the finite-volume and constrained transport updates. Finally, we have developed a test suite of standard and novel problems in one-, two-, and three-dimensions designed to validate our algorithms and implementation and to be of use to other code developers. The code is suitable for use in a wide variety of astrophysical applications and is freely available for download on the web.

Transport and equilibrium in field-reversed mirrors

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

Boyd, J.K.

Two plasma models relevant to compact torus research have been developed to study transport and equilibrium in field reversed mirrors. In the first model for small Larmor radius and large collision frequency, the plasma is described as an adiabatic hydromagnetic fluid. In the second model for large Larmor radius and small collision frequency, a kinetic theory description has been developed. Various aspects of the two models have been studied in five computer codes ADB, AV, NEO, OHK, RES. The ADB code computes two dimensional equilibrium and one dimensional transport in a flux coordinate. The AV code calculates orbit average integralsmore » in a harmonic oscillator potential. The NEO code follows particle trajectories in a Hill's vortex magnetic field to study stochasticity, invariants of the motion, and orbit average formulas. The OHK code displays analytic psi(r), B/sub Z/(r), phi(r), E/sub r/(r) formulas developed for the kinetic theory description. The RES code calculates resonance curves to consider overlap regions relevant to stochastic orbit behavior.« less

Method for calculating internal radiation and ventilation with the ADINAT heat-flow code

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

Butkovich, T.R.; Montan, D.N.

1980-04-01

One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation andmore » ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation.« less

Monte Carlo Modeling of the Initial Radiation Emitted by a Nuclear Device in the National Capital Region

DTIC Science & Technology

2013-07-01

also simulated in the models. Data was derived from calculations using the three-dimensional Monte Carlo radiation transport code MCNP (Monte Carlo N...32  B.  MCNP PHYSICS OPTIONS ......................................................................................... 33  C.  HAZUS...input deck’) for the MCNP , Monte Carlo N-Particle, radiation transport code. MCNP is a general-purpose code designed to simulate neutron, photon

Considerations of MCNP Monte Carlo code to be used as a radiotherapy treatment planning tool.

PubMed

Juste, B; Miro, R; Gallardo, S; Verdu, G; Santos, A

2005-01-01

The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)⁶⁰Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.

CTViz: A tool for the visualization of transport in nanocomposites.

PubMed

Beach, Benjamin; Brown, Joshua; Tarlton, Taylor; Derosa, Pedro A

2016-05-01

A visualization tool (CTViz) for charge transport processes in 3-D hybrid materials (nanocomposites) was developed, inspired by the need for a graphical application to assist in code debugging and data presentation of an existing in-house code. As the simulation code grew, troubleshooting problems grew increasingly difficult without an effective way to visualize 3-D samples and charge transport in those samples. CTViz is able to produce publication and presentation quality visuals of the simulation box, as well as static and animated visuals of the paths of individual carriers through the sample. CTViz was designed to provide a high degree of flexibility in the visualization of the data. A feature that characterizes this tool is the use of shade and transparency levels to highlight important details in the morphology or in the transport paths by hiding or dimming elements of little relevance to the current view. This is fundamental for the visualization of 3-D systems with complex structures. The code presented here provides these required capabilities, but has gone beyond the original design and could be used as is or easily adapted for the visualization of other particulate transport where transport occurs on discrete paths. Copyright © 2016 Elsevier Inc. All rights reserved.

Identification of Trends into Dose Calculations for Astronauts through Performing Sensitivity Analysis on Calculational Models Used by the Radiation Health Office

NASA Technical Reports Server (NTRS)

Adams, Thomas; VanBaalen, Mary

2009-01-01

The Radiation Health Office (RHO) determines each astronaut s cancer risk by using models to associate the amount of radiation dose that astronauts receive from spaceflight missions. The baryon transport codes (BRYNTRN), high charge (Z) and energy transport codes (HZETRN), and computer risk models are used to determine the effective dose received by astronauts in Low Earth orbit (LEO). This code uses an approximation of the Boltzman transport formula. The purpose of the project is to run this code for various International Space Station (ISS) flight parameters in order to gain a better understanding of how this code responds to different scenarios. The project will determine how variations in one set of parameters such as, the point of the solar cycle and altitude can affect the radiation exposure of astronauts during ISS missions. This project will benefit NASA by improving mission dosimetry.

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

PubMed

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

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 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

SHIELD and HZETRN comparisons of pion production cross sections

NASA Astrophysics Data System (ADS)

Norbury, John W.; Sobolevsky, Nikolai; Werneth, Charles M.

2018-03-01

A program of comparing American (NASA) and Russian (ROSCOSMOS) space radiation transport codes has recently begun, and the first paper directly comparing the NASA and ROSCOSMOS space radiation transport codes, HZETRN and SHIELD respectively has recently appeared. The present work represents the second time that NASA and ROSCOSMOS calculations have been directly compared, and the focus here is on models of pion production cross sections used in the two transport codes mentioned above. It was found that these models are in overall moderate agreement with each other and with experimental data. Disagreements that were found are discussed.

Evaluation and utilization of beam simulation codes for the SNS ion source and low energy beam transport developmenta)

NASA Astrophysics Data System (ADS)

Han, B. X.; Welton, R. F.; Stockli, M. P.; Luciano, N. P.; Carmichael, J. R.

2008-02-01

Beam simulation codes PBGUNS, SIMION, and LORENTZ-3D were evaluated by modeling the well-diagnosed SNS base line ion source and low energy beam transport (LEBT) system. Then, an investigation was conducted using these codes to assist our ion source and LEBT development effort which is directed at meeting the SNS operational and also the power-upgrade project goals. A high-efficiency H- extraction system as well as magnetic and electrostatic LEBT configurations capable of transporting up to 100mA is studied using these simulation tools.

On the Development of a Deterministic Three-Dimensional Radiation Transport Code

NASA Technical Reports Server (NTRS)

Rockell, Candice; Tweed, John

2011-01-01

Since astronauts on future deep space missions will be exposed to dangerous radiations, there is a need to accurately model the transport of radiation through shielding materials and to estimate the received radiation dose. In response to this need a three dimensional deterministic code for space radiation transport is now under development. The new code GRNTRN is based on a Green's function solution of the Boltzmann transport equation that is constructed in the form of a Neumann series. Analytical approximations will be obtained for the first three terms of the Neumann series and the remainder will be estimated by a non-perturbative technique . This work discusses progress made to date and exhibits some computations based on the first two Neumann series terms.

Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

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

Smith, L.M.; Hochstedler, R.D.

1997-02-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of themore » accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).« less

AN OPEN-SOURCE NEUTRINO RADIATION HYDRODYNAMICS CODE FOR CORE-COLLAPSE SUPERNOVAE

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

O’Connor, Evan, E-mail: evanoconnor@ncsu.edu; CITA, Canadian Institute for Theoretical Astrophysics, Toronto, M5S 3H8

2015-08-15

We present an open-source update to the spherically symmetric, general-relativistic hydrodynamics, core-collapse supernova (CCSN) code GR1D. The source code is available at http://www.GR1Dcode.org. We extend its capabilities to include a general-relativistic treatment of neutrino transport based on the moment formalisms of Shibata et al. and Cardall et al. We pay special attention to implementing and testing numerical methods and approximations that lessen the computational demand of the transport scheme by removing the need to invert large matrices. This is especially important for the implementation and development of moment-like transport methods in two and three dimensions. A critical component of neutrinomore » transport calculations is the neutrino–matter interaction coefficients that describe the production, absorption, scattering, and annihilation of neutrinos. In this article we also describe our open-source neutrino interaction library NuLib (available at http://www.nulib.org). We believe that an open-source approach to describing these interactions is one of the major steps needed to progress toward robust models of CCSNe and robust predictions of the neutrino signal. We show, via comparisons to full Boltzmann neutrino-transport simulations of CCSNe, that our neutrino transport code performs remarkably well. Furthermore, we show that the methods and approximations we employ to increase efficiency do not decrease the fidelity of our results. We also test the ability of our general-relativistic transport code to model failed CCSNe by evolving a 40-solar-mass progenitor to the onset of collapse to a black hole.« less

Reactive transport modeling in fractured rock: A state-of-the-science review

NASA Astrophysics Data System (ADS)

MacQuarrie, Kerry T. B.; Mayer, K. Ulrich

2005-10-01

The field of reactive transport modeling has expanded significantly in the past two decades and has assisted in resolving many issues in Earth Sciences. Numerical models allow for detailed examination of coupled transport and reactions, or more general investigation of controlling processes over geologic time scales. Reactive transport models serve to provide guidance in field data collection and, in particular, enable researchers to link modeling and hydrogeochemical studies. In this state-of-science review, the key objectives were to examine the applicability of reactive transport codes for exploring issues of redox stability to depths of several hundreds of meters in sparsely fractured crystalline rock, with a focus on the Canadian Shield setting. A conceptual model of oxygen ingress and redox buffering, within a Shield environment at time and space scales relevant to nuclear waste repository performance, is developed through a review of previous research. This conceptual model describes geochemical and biological processes and mechanisms materially important to understanding redox buffering capacity and radionuclide mobility in the far-field. Consistent with this model, reactive transport codes should ideally be capable of simulating the effects of changing recharge water compositions as a result of long-term climate change, and fracture-matrix interactions that may govern water-rock interaction. Other aspects influencing the suitability of reactive transport codes include the treatment of various reaction and transport time scales, the ability to apply equilibrium or kinetic formulations simultaneously, the need to capture feedback between water-rock interactions and porosity-permeability changes, and the representation of fractured crystalline rock environments as discrete fracture or dual continuum media. A review of modern multicomponent reactive transport codes indicates a relatively high-level of maturity. Within the Yucca Mountain nuclear waste disposal program, reactive transport codes of varying complexity have been applied to investigate the migration of radionuclides and the geochemical evolution of host rock around the planned disposal facility. Through appropriate near- and far-field application of dual continuum codes, this example demonstrates how reactive transport models have been applied to assist in constraining historic water infiltration rates, interpreting the sealing of flow paths due to mineral precipitation, and investigating post-closure geochemical monitoring strategies. Natural analogue modeling studies, although few in number, are also of key importance as they allow the comparison of model results with hydrogeochemical and paleohydrogeological data over geologic time scales.

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.

Reactive transport codes for subsurface environmental simulation

DOE PAGES

Steefel, C. I.; Appelo, C. A. J.; Arora, B.; ...

2014-09-26

A general description of the mathematical and numerical formulations used in modern numerical reactive transport codes relevant for subsurface environmental simulations is presented. The formulations are followed by short descriptions of commonly used and available subsurface simulators that consider continuum representations of flow, transport, and reactions in porous media. These formulations are applicable to most of the subsurface environmental benchmark problems included in this special issue. The list of codes described briefly here includes PHREEQC, HPx, PHT3D, OpenGeoSys (OGS), HYTEC, ORCHESTRA, TOUGHREACT, eSTOMP, HYDROGEOCHEM, CrunchFlow, MIN3P, and PFLOTRAN. The descriptions include a high-level list of capabilities for each of themore » codes, along with a selective list of applications that highlight their capabilities and historical development.« less

Dust-Particle Transport in Tokamak Edge Plasmas

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

Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K

2005-09-12

Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensivemore » dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.« less

Study of no-man's land physics in the total-f gyrokinetic code XGC1

NASA Astrophysics Data System (ADS)

Ku, Seung Hoe; Chang, C. S.; Lang, J.

2014-10-01

While the ``transport shortfall'' in the ``no-man's land'' has been observed often in delta-f codes, it has not yet been observed in the global total-f gyrokinetic particle code XGC1. Since understanding the interaction between the edge and core transport appears to be a critical element in the prediction for ITER performance, understanding the no-man's land issue is an important physics research topic. Simulation results using the Holland case will be presented and the physics causing the shortfall phenomenon will be discussed. Nonlinear nonlocal interaction of turbulence, secondary flows, and transport appears to be the key.

Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System

NASA Astrophysics Data System (ADS)

Aizawa, Naoto; Iwasaki, Tomohiko

2014-06-01

Safety analysis code system of beam transport and core for accelerator driven system (ADS) is developed for the analyses of beam transients such as the change of the shape and position of incident beam. The code system consists of the beam transport analysis part and the core analysis part. TRACE 3-D is employed in the beam transport analysis part, and the shape and incident position of beam at the target are calculated. In the core analysis part, the neutronics, thermo-hydraulics and cladding failure analyses are performed by the use of ADS dynamic calculation code ADSE on the basis of the external source database calculated by PHITS and the cross section database calculated by SRAC, and the programs of the cladding failure analysis for thermoelastic and creep. By the use of the code system, beam transient analyses are performed for the ADS proposed by Japan Atomic Energy Agency. As a result, the rapid increase of the cladding temperature happens and the plastic deformation is caused in several seconds. In addition, the cladding is evaluated to be failed by creep within a hundred seconds. These results have shown that the beam transients have caused a cladding failure.

Use of FEC coding to improve statistical multiplexing performance for video transport over ATM networks

NASA Astrophysics Data System (ADS)

Kurceren, Ragip; Modestino, James W.

1998-12-01

The use of forward error-control (FEC) coding, possibly in conjunction with ARQ techniques, has emerged as a promising approach for video transport over ATM networks for cell-loss recovery and/or bit error correction, such as might be required for wireless links. Although FEC provides cell-loss recovery capabilities it also introduces transmission overhead which can possibly cause additional cell losses. A methodology is described to maximize the number of video sources multiplexed at a given quality of service (QoS), measured in terms of decoded cell loss probability, using interlaced FEC codes. The transport channel is modelled as a block interference channel (BIC) and the multiplexer as single server, deterministic service, finite buffer supporting N users. Based upon an information-theoretic characterization of the BIC and large deviation bounds on the buffer overflow probability, the described methodology provides theoretically achievable upper limits on the number of sources multiplexed. Performance of specific coding techniques using interlaced nonbinary Reed-Solomon (RS) codes and binary rate-compatible punctured convolutional (RCPC) codes is illustrated.

Clean Energy in City Codes: A Baseline Analysis of Municipal Codification across the United States

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

Cook, Jeffrey J.; Aznar, Alexandra; Dane, Alexander

Municipal governments in the United States are well positioned to influence clean energy (energy efficiency and alternative energy) and transportation technology and strategy implementation within their jurisdictions through planning, programs, and codification. Municipal governments are leveraging planning processes and programs to shape their energy futures. There is limited understanding in the literature related to codification, the primary way that municipal governments enact enforceable policies. The authors fill the gap in the literature by documenting the status of municipal codification of clean energy and transportation across the United States. More directly, we leverage online databases of municipal codes to develop nationalmore » and state-specific representative samples of municipal governments by population size. Our analysis finds that municipal governments with the authority to set residential building energy codes within their jurisdictions frequently do so. In some cases, communities set codes higher than their respective state governments. Examination of codes across the nation indicates that municipal governments are employing their code as a policy mechanism to address clean energy and transportation.« 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.

ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

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

Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

2008-04-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a methodmore » for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.« less

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.

Boltzmann Transport Code Update: Parallelization and Integrated Design Updates

NASA Technical Reports Server (NTRS)

Heinbockel, J. H.; Nealy, J. E.; DeAngelis, G.; Feldman, G. A.; Chokshi, S.

2003-01-01

The on going efforts at developing a web site for radiation analysis is expected to result in an increased usage of the High Charge and Energy Transport Code HZETRN. It would be nice to be able to do the requested calculations quickly and efficiently. Therefore the question arose, "Could the implementation of parallel processing speed up the calculations required?" To answer this question two modifications of the HZETRN computer code were created. The first modification selected the shield material of Al(2219) , then polyethylene and then Al(2219). The modified Fortran code was labeled 1SSTRN.F. The second modification considered the shield material of CO2 and Martian regolith. This modified Fortran code was labeled MARSTRN.F.

TRANSP: status and planning

NASA Astrophysics Data System (ADS)

Andre, R.; Carlsson, J.; Gorelenkova, M.; Jardin, S.; Kaye, S.; Poli, F.; Yuan, X.

2016-10-01

TRANSP is an integrated interpretive and predictive transport analysis tool that incorporates state of the art heating/current drive sources and transport models. The treatments and transport solvers are becoming increasingly sophisticated and comprehensive. For instance, the ISOLVER component provides a free boundary equilibrium solution, while the PT- SOLVER transport solver is especially suited for stiff transport models such as TGLF. TRANSP incorporates high fidelity heating and current drive source models, such as NUBEAM for neutral beam injection, the beam tracing code TORBEAM for EC, TORIC for ICRF, the ray tracing TORAY and GENRAY for EC. The implementation of selected components makes efficient use of MPI for speed up of code calculations. Recently the GENRAY-CQL3D solver for modeling of LH heating and current drive has been implemented and currently being extended to multiple antennas, to allow modeling of EAST discharges. Also, GENRAY+CQL3D is being extended to the use of EC/EBW and of HHFW for NSTX-U. This poster will describe present uses of the code worldwide, as well as plans for upgrading the physics modules and code framework. Work supported by the US Department of Energy under DE-AC02-CH0911466.

BRYNTRN: A baryon transport model

NASA Technical Reports Server (NTRS)

Wilson, John W.; Townsend, Lawrence W.; Nealy, John E.; Chun, Sang Y.; Hong, B. S.; Buck, Warren W.; Lamkin, S. L.; Ganapol, Barry D.; Khan, Ferdous; Cucinotta, Francis A.

1989-01-01

The development of an interaction data base and a numerical solution to the transport of baryons through an arbitrary shield material based on a straight ahead approximation of the Boltzmann equation are described. The code is most accurate for continuous energy boundary values, but gives reasonable results for discrete spectra at the boundary using even a relatively coarse energy grid (30 points) and large spatial increments (1 cm in H2O). The resulting computer code is self-contained, efficient and ready to use. The code requires only a very small fraction of the computer resources required for Monte Carlo codes.

ecode - Electron Transport Algorithm Testing v. 1.0

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

Franke, Brian C.; Olson, Aaron J.; Bruss, Donald Eugene

2016-10-05

ecode is a Monte Carlo code used for testing algorithms related to electron transport. The code can read basic physics parameters, such as energy-dependent stopping powers and screening parameters. The code permits simple planar geometries of slabs or cubes. Parallelization consists of domain replication, with work distributed at the start of the calculation and statistical results gathered at the end of the calculation. Some basic routines (such as input parsing, random number generation, and statistics processing) are shared with the Integrated Tiger Series codes. A variety of algorithms for uncertainty propagation are incorporated based on the stochastic collocation and stochasticmore » Galerkin methods. These permit uncertainty only in the total and angular scattering cross sections. The code contains algorithms for simulating stochastic mixtures of two materials. The physics is approximate, ranging from mono-energetic and isotropic scattering to screened Rutherford angular scattering and Rutherford energy-loss scattering (simple electron transport models). No production of secondary particles is implemented, and no photon physics is implemented.« less

MCNP capabilities for nuclear well logging calculations

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

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

Differential Cross Section Kinematics for 3-dimensional Transport Codes

NASA Technical Reports Server (NTRS)

Norbury, John W.; Dick, Frank

2008-01-01

In support of the development of 3-dimensional transport codes, this paper derives the relevant relativistic particle kinematic theory. Formulas are given for invariant, spectral and angular distributions in both the lab (spacecraft) and center of momentum frames, for collisions involving 2, 3 and n - body final states.

49 CFR 171.25 - Additional requirements for the use of the IMDG Code.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 176 of this subchapter. (3) Packages containing primary lithium batteries and cells that are transported in accordance with Special Provision 188 of the IMDG Code must be marked “PRIMARY LITHIUM BATTERIES—FORBIDDEN FOR TRANSPORT ABOARD PASSENGER AIRCRAFT” or “LITHIUM METAL BATTERIES—FORBIDDEN FOR...

49 CFR 171.25 - Additional requirements for the use of the IMDG Code.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 176 of this subchapter. (3) Packages containing primary lithium batteries and cells that are transported in accordance with Special Provision 188 of the IMDG Code must be marked “PRIMARY LITHIUM BATTERIES—FORBIDDEN FOR TRANSPORT ABOARD PASSENGER AIRCRAFT” or “LITHIUM METAL BATTERIES—FORBIDDEN FOR...

49 CFR 171.25 - Additional requirements for the use of the IMDG Code.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 176 of this subchapter. (3) Packages containing primary lithium batteries and cells that are transported in accordance with Special Provision 188 of the IMDG Code must be marked “PRIMARY LITHIUM BATTERIES—FORBIDDEN FOR TRANSPORT ABOARD PASSENGER AIRCRAFT” or “LITHIUM METAL BATTERIES—FORBIDDEN FOR...

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.

Nonperturbative methods in HZE ion transport

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Costen, Robert C.; Shinn, Judy L.

1993-01-01

A nonperturbative analytic solution of the high charge and energy (HZE) Green's function is used to implement a computer code for laboratory ion beam transport. The code is established to operate on the Langley Research Center nuclear fragmentation model used in engineering applications. Computational procedures are established to generate linear energy transfer (LET) distributions for a specified ion beam and target for comparison with experimental measurements. The code is highly efficient and compares well with the perturbation approximations.

Suite of Benchmark Tests to Conduct Mesh-Convergence Analysis of Nonlinear and Non-constant Coefficient Transport Codes

NASA Astrophysics Data System (ADS)

Zamani, K.; Bombardelli, F. A.

2014-12-01

Verification of geophysics codes is imperative to avoid serious academic as well as practical consequences. In case that access to any given source code is not possible, the Method of Manufactured Solution (MMS) cannot be employed in code verification. In contrast, employing the Method of Exact Solution (MES) has several practical advantages. In this research, we first provide four new one-dimensional analytical solutions designed for code verification; these solutions are able to uncover the particular imperfections of the Advection-diffusion-reaction equation, such as nonlinear advection, diffusion or source terms, as well as non-constant coefficient equations. After that, we provide a solution of Burgers' equation in a novel setup. Proposed solutions satisfy the continuity of mass for the ambient flow, which is a crucial factor for coupled hydrodynamics-transport solvers. Then, we use the derived analytical solutions for code verification. To clarify gray-literature issues in the verification of transport codes, we designed a comprehensive test suite to uncover any imperfection in transport solvers via a hierarchical increase in the level of tests' complexity. The test suite includes hundreds of unit tests and system tests to check vis-a-vis the portions of the code. Examples for checking the suite start by testing a simple case of unidirectional advection; then, bidirectional advection and tidal flow and build up to nonlinear cases. We design tests to check nonlinearity in velocity, dispersivity and reactions. The concealing effect of scales (Peclet and Damkohler numbers) on the mesh-convergence study and appropriate remedies are also discussed. For the cases in which the appropriate benchmarks for mesh convergence study are not available, we utilize symmetry. Auxiliary subroutines for automation of the test suite and report generation are designed. All in all, the test package is not only a robust tool for code verification but it also provides comprehensive insight on the ADR solvers capabilities. Such information is essential for any rigorous computational modeling of ADR equation for surface/subsurface pollution transport. We also convey our experiences in finding several errors which were not detectable with routine verification techniques.

Integrated modelling framework for short pulse high energy density physics experiments

NASA Astrophysics Data System (ADS)

Sircombe, N. J.; Hughes, S. J.; Ramsay, M. G.

2016-03-01

Modelling experimental campaigns on the Orion laser at AWE, and developing a viable point-design for fast ignition (FI), calls for a multi-scale approach; a complete description of the problem would require an extensive range of physics which cannot realistically be included in a single code. For modelling the laser-plasma interaction (LPI) we need a fine mesh which can capture the dispersion of electromagnetic waves, and a kinetic model for each plasma species. In the dense material of the bulk target, away from the LPI region, collisional physics dominates. The transport of hot particles generated by the action of the laser is dependent on their slowing and stopping in the dense material and their need to draw a return current. These effects will heat the target, which in turn influences transport. On longer timescales, the hydrodynamic response of the target will begin to play a role as the pressure generated from isochoric heating begins to take effect. Recent effort at AWE [1] has focussed on the development of an integrated code suite based on: the particle in cell code EPOCH, to model LPI; the Monte-Carlo electron transport code THOR, to model the onward transport of hot electrons; and the radiation hydrodynamics code CORVUS, to model the hydrodynamic response of the target. We outline the methodology adopted, elucidate on the advantages of a robustly integrated code suite compared to a single code approach, demonstrate the integrated code suite's application to modelling the heating of buried layers on Orion, and assess the potential of such experiments for the validation of modelling capability in advance of more ambitious HEDP experiments, as a step towards a predictive modelling capability for FI.

Muon simulation codes MUSIC and MUSUN for underground physics

NASA Astrophysics Data System (ADS)

Kudryavtsev, V. A.

2009-03-01

The paper describes two Monte Carlo codes dedicated to muon simulations: MUSIC (MUon SImulation Code) and MUSUN (MUon Simulations UNderground). MUSIC is a package for muon transport through matter. It is particularly useful for propagating muons through large thickness of rock or water, for instance from the surface down to underground/underwater laboratory. MUSUN is designed to use the results of muon transport through rock/water to generate muons in or around underground laboratory taking into account their energy spectrum and angular distribution.

Approximate Green's function methods for HZE transport in multilayered materials

NASA Technical Reports Server (NTRS)

Wilson, John W.; Badavi, Francis F.; Shinn, Judy L.; Costen, Robert C.

1993-01-01

A nonperturbative analytic solution of the high charge and energy (HZE) Green's function is used to implement a computer code for laboratory ion beam transport in multilayered materials. The code is established to operate on the Langley nuclear fragmentation model used in engineering applications. Computational procedures are established to generate linear energy transfer (LET) distributions for a specified ion beam and target for comparison with experimental measurements. The code was found to be highly efficient and compared well with the perturbation approximation.

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

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

Trent, D.S.; Eyler, L.L.

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

14 CFR 257.6 - Effective and compliance dates.

Code of Federal Regulations, 2011 CFR

2011-01-01

... the selling carrier is not the transporting carrier and (ii) Of the transporting carrier's identity... transportation involving a code-share arrangement of the transporting carrier's corporate name and any other name...

14 CFR 257.6 - Effective and compliance dates.

Code of Federal Regulations, 2014 CFR

2014-01-01

... the selling carrier is not the transporting carrier and (ii) Of the transporting carrier's identity... transportation involving a code-share arrangement of the transporting carrier's corporate name and any other name...

14 CFR 257.6 - Effective and compliance dates.

Code of Federal Regulations, 2013 CFR

2013-01-01

... the selling carrier is not the transporting carrier and (ii) Of the transporting carrier's identity... transportation involving a code-share arrangement of the transporting carrier's corporate name and any other name...

14 CFR 257.6 - Effective and compliance dates.

Code of Federal Regulations, 2012 CFR

2012-01-01

... the selling carrier is not the transporting carrier and (ii) Of the transporting carrier's identity... transportation involving a code-share arrangement of the transporting carrier's corporate name and any other name...

PelePhysics

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

2017-05-17

PelePhysics is a suite of physics packages that provides functionality of use to reacting hydrodynamics CFD codes. The initial release includes an interface to reaction rate mechanism evaluation, transport coefficient evaluation, and a generalized equation of state (EOS) facility. Both generic evaluators and interfaces to code from externally available tools (Fuego for chemical rates, EGLib for transport coefficients) are provided.

Analysis of JT-60SA operational scenarios

NASA Astrophysics Data System (ADS)

Garzotti, L.; Barbato, E.; Garcia, J.; Hayashi, N.; Voitsekhovitch, I.; Giruzzi, G.; Maget, P.; Romanelli, M.; Saarelma, S.; Stankiewitz, R.; Yoshida, M.; Zagórski, R.

2018-02-01

Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated.

Improvements on non-equilibrium and transport Green function techniques: The next-generation TRANSIESTA

NASA Astrophysics Data System (ADS)

Papior, Nick; Lorente, Nicolás; Frederiksen, Thomas; García, Alberto; Brandbyge, Mads

2017-03-01

We present novel methods implemented within the non-equilibrium Green function code (NEGF) TRANSIESTA based on density functional theory (DFT). Our flexible, next-generation DFT-NEGF code handles devices with one or multiple electrodes (Ne ≥ 1) with individual chemical potentials and electronic temperatures. We describe its novel methods for electrostatic gating, contour optimizations, and assertion of charge conservation, as well as the newly implemented algorithms for optimized and scalable matrix inversion, performance-critical pivoting, and hybrid parallelization. Additionally, a generic NEGF "post-processing" code (TBTRANS/PHTRANS) for electron and phonon transport is presented with several novelties such as Hamiltonian interpolations, Ne ≥ 1 electrode capability, bond-currents, generalized interface for user-defined tight-binding transport, transmission projection using eigenstates of a projected Hamiltonian, and fast inversion algorithms for large-scale simulations easily exceeding 106 atoms on workstation computers. The new features of both codes are demonstrated and bench-marked for relevant test systems.

Morse Monte Carlo Radiation Transport Code System

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

Emmett, M.B.

1975-02-01

The report contains sections containing descriptions of the MORSE and PICTURE codes, input descriptions, sample problems, deviations of the physical equations and explanations of the various error messages. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. The PICTURE code provide aid in preparing correct input data for the combinatorial geometry package CG. It provides a printed view of arbitrary two-dimensional slices through the geometry. By inspecting these pictures one maymore » determine if the geometry specified by the input cards is indeed the desired geometry. 23 refs. (WRF)« less

TReacLab: An object-oriented implementation of non-intrusive splitting methods to couple independent transport and geochemical software

NASA Astrophysics Data System (ADS)

Jara, Daniel; de Dreuzy, Jean-Raynald; Cochepin, Benoit

2017-12-01

Reactive transport modeling contributes to understand geophysical and geochemical processes in subsurface environments. Operator splitting methods have been proposed as non-intrusive coupling techniques that optimize the use of existing chemistry and transport codes. In this spirit, we propose a coupler relying on external geochemical and transport codes with appropriate operator segmentation that enables possible developments of additional splitting methods. We provide an object-oriented implementation in TReacLab developed in the MATLAB environment in a free open source frame with an accessible repository. TReacLab contains classical coupling methods, template interfaces and calling functions for two classical transport and reactive software (PHREEQC and COMSOL). It is tested on four classical benchmarks with homogeneous and heterogeneous reactions at equilibrium or kinetically-controlled. We show that full decoupling to the implementation level has a cost in terms of accuracy compared to more integrated and optimized codes. Use of non-intrusive implementations like TReacLab are still justified for coupling independent transport and chemical software at a minimal development effort but should be systematically and carefully assessed.

Applications of the microdosimetric function implemented in the macroscopic particle transport simulation code PHITS.

PubMed

Sato, Tatsuhiko; Watanabe, Ritsuko; Sihver, Lembit; Niita, Koji

2012-01-01

Microdosimetric quantities such as lineal energy are generally considered to be better indices than linear energy transfer (LET) for expressing the relative biological effectiveness (RBE) of high charge and energy particles. To calculate their probability densities (PD) in macroscopic matter, it is necessary to integrate microdosimetric tools such as track-structure simulation codes with macroscopic particle transport simulation codes. As an integration approach, the mathematical model for calculating the PD of microdosimetric quantities developed based on track-structure simulations was incorporated into the macroscopic particle transport simulation code PHITS (Particle and Heavy Ion Transport code System). The improved PHITS enables the PD in macroscopic matter to be calculated within a reasonable computation time, while taking their stochastic nature into account. The microdosimetric function of PHITS was applied to biological dose estimation for charged-particle therapy and risk estimation for astronauts. The former application was performed in combination with the microdosimetric kinetic model, while the latter employed the radiation quality factor expressed as a function of lineal energy. Owing to the unique features of the microdosimetric function, the improved PHITS has the potential to establish more sophisticated systems for radiological protection in space as well as for the treatment planning of charged-particle therapy.

ipole: Semianalytic scheme for relativistic polarized radiative transport

NASA Astrophysics Data System (ADS)

Moscibrodzka, Monika; Gammie, Charles F.

2018-04-01

ipole is a ray-tracing code for covariant, polarized radiative transport particularly useful for modeling Event Horizon Telescope sources, though may also be used for other relativistic transport problems. The code extends the ibothros scheme for covariant, unpolarized transport using two representations of the polarized radiation field: in the coordinate frame, it parallel transports the coherency tensor, and in the frame of the plasma, it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is as spacetime- and coordinate- independent as possible; the emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, ipole is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth.

3D unstructured-mesh radiation transport codes

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

Morel, J.

1997-12-31

Three unstructured-mesh radiation transport codes are currently being developed at Los Alamos National Laboratory. The first code is ATTILA, which uses an unstructured tetrahedral mesh in conjunction with standard Sn (discrete-ordinates) angular discretization, standard multigroup energy discretization, and linear-discontinuous spatial differencing. ATTILA solves the standard first-order form of the transport equation using source iteration in conjunction with diffusion-synthetic acceleration of the within-group source iterations. DANTE is designed to run primarily on workstations. The second code is DANTE, which uses a hybrid finite-element mesh consisting of arbitrary combinations of hexahedra, wedges, pyramids, and tetrahedra. DANTE solves several second-order self-adjoint forms of the transport equation including the even-parity equation, the odd-parity equation, and a new equation called the self-adjoint angular flux equation. DANTE also offers three angular discretization options:more » $$S{_}n$$ (discrete-ordinates), $$P{_}n$$ (spherical harmonics), and $$SP{_}n$$ (simplified spherical harmonics). DANTE is designed to run primarily on massively parallel message-passing machines, such as the ASCI-Blue machines at LANL and LLNL. The third code is PERICLES, which uses the same hybrid finite-element mesh as DANTE, but solves the standard first-order form of the transport equation rather than a second-order self-adjoint form. DANTE uses a standard $$S{_}n$$ discretization in angle in conjunction with trilinear-discontinuous spatial differencing, and diffusion-synthetic acceleration of the within-group source iterations. PERICLES was initially designed to run on workstations, but a version for massively parallel message-passing machines will be built. The three codes will be described in detail and computational results will be presented.« less

AN EXTENSION OF THE ATHENA++ CODE FRAMEWORK FOR GRMHD BASED ON ADVANCED RIEMANN SOLVERS AND STAGGERED-MESH CONSTRAINED TRANSPORT

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

White, Christopher J.; Stone, James M.; Gammie, Charles F.

2016-08-01

We present a new general relativistic magnetohydrodynamics (GRMHD) code integrated into the Athena++ framework. Improving upon the techniques used in most GRMHD codes, ours allows the use of advanced, less diffusive Riemann solvers, in particular HLLC and HLLD. We also employ a staggered-mesh constrained transport algorithm suited for curvilinear coordinate systems in order to maintain the divergence-free constraint of the magnetic field. Our code is designed to work with arbitrary stationary spacetimes in one, two, or three dimensions, and we demonstrate its reliability through a number of tests. We also report on its promising performance and scalability.

Common Errors in the Calculation of Aircrew Doses from Cosmic Rays

NASA Astrophysics Data System (ADS)

O'Brien, Keran; Felsberger, Ernst; Kindl, Peter

2010-05-01

Radiation doses to air crew are calculated using flight codes. Flight codes integrate dose rates over the aircraft flight path, which were calculated by transport codes or obtained by measurements from take off at a specific airport to landing at another. The dose rates are stored in various ways, such as by latitude and longitude, or in terms of the geomagnetic vertical cutoff. The transport codes are generally quite satisfactory, but the treatment of the boundary conditions is frequently incorrect. Both the treatment of solar modulation and of the effect of the geomagnetic field are often defective, leading to the systematic overestimate of the crew doses.

49 CFR 178.905 - Large Packaging identification codes.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Large Packaging identification codes. 178.905... FOR PACKAGINGS Large Packagings Standards § 178.905 Large Packaging identification codes. Large packaging code designations consist of: two numerals specified in paragraph (a) of this section; followed by...

PFLOTRAN: Reactive Flow & Transport Code for Use on Laptops to Leadership-Class Supercomputers

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

Hammond, Glenn E.; Lichtner, Peter C.; Lu, Chuan

PFLOTRAN, a next-generation reactive flow and transport code for modeling subsurface processes, has been designed from the ground up to run efficiently on machines ranging from leadership-class supercomputers to laptops. Based on an object-oriented design, the code is easily extensible to incorporate additional processes. It can interface seamlessly with Fortran 9X, C and C++ codes. Domain decomposition parallelism is employed, with the PETSc parallel framework used to manage parallel solvers, data structures and communication. Features of the code include a modular input file, implementation of high-performance I/O using parallel HDF5, ability to perform multiple realization simulations with multiple processors permore » realization in a seamless manner, and multiple modes for multiphase flow and multicomponent geochemical transport. Chemical reactions currently implemented in the code include homogeneous aqueous complexing reactions and heterogeneous mineral precipitation/dissolution, ion exchange, surface complexation and a multirate kinetic sorption model. PFLOTRAN has demonstrated petascale performance using 2{sup 17} processor cores with over 2 billion degrees of freedom. Accomplishments achieved to date include applications to the Hanford 300 Area and modeling CO{sub 2} sequestration in deep geologic formations.« less

Use of Existing CAD Models for Radiation Shielding Analysis

NASA Technical Reports Server (NTRS)

Lee, K. T.; Barzilla, J. E.; Wilson, P.; Davis, A.; Zachman, J.

2015-01-01

The utility of a radiation exposure analysis depends not only on the accuracy of the underlying particle transport code, but also on the accuracy of the geometric representations of both the vehicle used as radiation shielding mass and the phantom representation of the human form. The current NASA/Space Radiation Analysis Group (SRAG) process to determine crew radiation exposure in a vehicle design incorporates both output from an analytic High Z and Energy Particle Transport (HZETRN) code and the properties (i.e., material thicknesses) of a previously processed drawing. This geometry pre-process can be time-consuming, and the results are less accurate than those determined using a Monte Carlo-based particle transport code. The current work aims to improve this process. Although several Monte Carlo programs (FLUKA, Geant4) are readily available, most use an internal geometry engine. The lack of an interface with the standard CAD formats used by the vehicle designers limits the ability of the user to communicate complex geometries. Translation of native CAD drawings into a format readable by these transport programs is time consuming and prone to error. The Direct Accelerated Geometry -United (DAGU) project is intended to provide an interface between the native vehicle or phantom CAD geometry and multiple particle transport codes to minimize problem setup, computing time and analysis error.

Status and Plans for the TRANSP Interpretive and Predictive Simulation Code

NASA Astrophysics Data System (ADS)

Kaye, Stanley; Andre, Robert; Marina, Gorelenkova; Yuan, Xingqui; Hawryluk, Richard; Jardin, Steven; Poli, Francesca

2015-11-01

TRANSP is an integrated interpretive and predictive transport analysis tool that incorporates state of the art heating/current drive sources and transport models. The treatments and transport solvers are becoming increasingly sophisticated and comprehensive. For instance, the ISOLVER component provides a free boundary equilibrium solution, while the PT_SOLVER transport solver is especially suited for stiff transport models such as TGLF. TRANSP also incorporates such source models as NUBEAM for neutral beam injection, GENRAY, TORAY, TORBEAM, TORIC and CQL3D for ICRH, LHCD, ECH and HHFW. The implementation of selected components makes efficient use of MPI for speed up of code calculations. TRANSP has a wide international user-base, and it is run on the FusionGrid to allow for timely support and quick turnaround by the PPPL Computational Plasma Physics Group. It is being used as a basis for both analysis and development of control algorithms and discharge operational scenarios, including simulation of ITER plasmas. This poster will describe present uses of the code worldwide, as well as plans for upgrading the physics modules and code framework. Progress on implementing TRANSP as a component in the ITER IMAS will also be described. This research was supported by the U.S. Department of Energy under contracts DE-AC02-09CH11466.

Cosmic-ray propagation with DRAGON2: I. numerical solver and astrophysical ingredients

NASA Astrophysics Data System (ADS)

Evoli, Carmelo; Gaggero, Daniele; Vittino, Andrea; Di Bernardo, Giuseppe; Di Mauro, Mattia; Ligorini, Arianna; Ullio, Piero; Grasso, Dario

2017-02-01

We present version 2 of the DRAGON code designed for computing realistic predictions of the CR densities in the Galaxy. The code numerically solves the interstellar CR transport equation (including inhomogeneous and anisotropic diffusion, either in space and momentum, advective transport and energy losses), under realistic conditions. The new version includes an updated numerical solver and several models for the astrophysical ingredients involved in the transport equation. Improvements in the accuracy of the numerical solution are proved against analytical solutions and in reference diffusion scenarios. The novel features implemented in the code allow to simulate the diverse scenarios proposed to reproduce the most recent measurements of local and diffuse CR fluxes, going beyond the limitations of the homogeneous galactic transport paradigm. To this end, several applications using DRAGON2 are presented as well. This new version facilitates the users to include their own physical models by means of a modular C++ structure.

Laser Blow-Off Impurity Injection Experiments at the HSX Stellarator

NASA Astrophysics Data System (ADS)

Castillo, J. F.; Bader, A.; Likin, K. M.; Anderson, D. T.; Anderson, F. S. B.; Kumar, S. T. A.; Talmadge, J. N.

2017-10-01

Results from the HSX laser blow-off experiment are presented and compared to a synthetic diagnostic implemented in the STRAHL impurity transport modeling code in order to measure the impurity transport diffusivity and convective velocity. A laser blow-off impurity injection system is used to rapidly deposit a small, controlled quantity of aluminum into the confinement volume. Five AXUV photodiode arrays are used to take time-resolved measurements of the impurity radiation. The spatially one-dimensional impurity transport code STRAHL is used to calculate a time-dependent plasma emissivity profile. Modeled intensity signals calculated from a synthetic diagnostic code provide direct comparison between plasma simulation and experimental results. An optimization algorithm with impurity transport coefficients acting as free parameters is used to fit the model to experimental data. This work is supported by US DOE Grant DE-FG02-93ER54222.

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.

Validation of the WIMSD4M cross-section generation code with benchmark results

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

Leal, L.C.; Deen, J.R.; Woodruff, W.L.

1995-02-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment for Research and Test (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the procedure to generatemore » cross-section libraries for reactor analyses and calculations utilizing the WIMSD4M code. To do so, the results of calculations performed with group cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory(ORNL) unreflected critical spheres, the TRX critical experiments, and calculations of a modified Los Alamos highly-enriched heavy-water moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

New Parallel computing framework for radiation transport codes

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

Kostin, M.A.; /Michigan State U., NSCL; Mokhov, N.V.

A new parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was integrated with the MARS15 code, and an effort is under way to deploy it in PHITS. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility canmore » be used in single process calculations as well as in the parallel regime. Several checkpoint files can be merged into one thus combining results of several calculations. The framework also corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.« less

FLUKA: A Multi-Particle Transport Code

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

Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan

2005-12-14

This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.

Adapting HYDRUS-1D to simulate overland flow and reactive transport during sheet flow deviations

USDA-ARS?s Scientific Manuscript database

The HYDRUS-1D code is a popular numerical model for solving the Richards equation for variably-saturated water flow and solute transport in porous media. This code was adapted to solve rather than the Richards equation for subsurface flow the diffusion wave equation for overland flow at the soil sur...

Modeling of boron species in the Falcon 17 and ISP-34 integral tests

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

Lazaridis, M.; Capitao, J.A.; Drossinos, Y.

1996-09-01

The RAFT computer code for aerosol formation and transport was modified to include boron species in its chemical database. The modification was necessary to calculate fission product transport and deposition in the FAL-17 and ISP-34 Falcon tests, where boric acid was injected. The experimental results suggest that the transport of cesium is modified in the presence of boron. The results obtained with the modified RAFT code are presented; they show good agreement with experimental results for cesium and partial agreement for boron deposition in the Falcon silica tube. The new version of the RAFT code predicts the same behavior formore » iodine deposition as the previous version, where boron species were not included.« less

Comparison of Stopping Power and Range Databases for Radiation Transport Study

NASA Technical Reports Server (NTRS)

Tai, H.; Bichsel, Hans; Wilson, John W.; Shinn, Judy L.; Cucinotta, Francis A.; Badavi, Francis F.

1997-01-01

The codes used to calculate stopping power and range for the space radiation shielding program at the Langley Research Center are based on the work of Ziegler but with modifications. As more experience is gained from experiments at heavy ion accelerators, prudence dictates a reevaluation of the current databases. Numerical values of stopping power and range calculated from four different codes currently in use are presented for selected ions and materials in the energy domain suitable for space radiation transport. This study of radiation transport has found that for most collision systems and for intermediate particle energies, agreement is less than 1 percent, in general, among all the codes. However, greater discrepancies are seen for heavy systems, especially at low particle energies.

Statistical Analysis of CFD Solutions from the Third AIAA Drag Prediction Workshop

NASA Technical Reports Server (NTRS)

Morrison, Joseph H.; Hemsch, Michael J.

2007-01-01

The first AIAA Drag Prediction Workshop, held in June 2001, evaluated the results from an extensive N-version test of a collection of Reynolds-Averaged Navier-Stokes CFD codes. The code-to-code scatter was more than an order of magnitude larger than desired for design and experimental validation of cruise conditions for a subsonic transport configuration. The second AIAA Drag Prediction Workshop, held in June 2003, emphasized the determination of installed pylon-nacelle drag increments and grid refinement studies. The code-to-code scatter was significantly reduced compared to the first DPW, but still larger than desired. However, grid refinement studies showed no significant improvement in code-to-code scatter with increasing grid refinement. The third Drag Prediction Workshop focused on the determination of installed side-of-body fairing drag increments and grid refinement studies for clean attached flow on wing alone configurations and for separated flow on the DLR-F6 subsonic transport model. This work evaluated the effect of grid refinement on the code-to-code scatter for the clean attached flow test cases and the separated flow test cases.

Reactive transport modeling of stable carbon isotope fractionation in a multi-phase multi-component system during carbon sequestration

DOE PAGES

Zhang, Shuo; DePaolo, Donald J.; Zheng, Liange; ...

2014-12-31

Carbon stable isotopes can be used in characterization and monitoring of CO 2 sequestration sites to track the migration of the CO 2 plume and identify leakage sources, and to evaluate the chemical reactions that take place in the CO 2-water-rock system. However, there are few tools available to incorporate stable isotope information into flow and transport codes used for CO 2 sequestration problems. We present a numerical tool for modeling the transport of stable carbon isotopes in multiphase reactive systems relevant to geologic carbon sequestration. The code is an extension of the reactive transport code TOUGHREACT. The transport modulemore » of TOUGHREACT was modified to include separate isotopic species of CO 2 gas and dissolved inorganic carbon (CO 2, CO 3 2-, HCO 3 -,…). Any process of transport or reaction influencing a given carbon species also influences its isotopic ratio. Isotopic fractionation is thus fully integrated within the dynamic system. The chemical module and database have been expanded to include isotopic exchange and fractionation between the carbon species in both gas and aqueous phases. The performance of the code is verified by modeling ideal systems and comparing with theoretical results. Efforts are also made to fit field data from the Pembina CO 2 injection project in Canada. We show that the exchange of carbon isotopes between dissolved and gaseous carbon species combined with fluid flow and transport, produce isotopic effects that are significantly different from simple two-component mixing. These effects are important for understanding the isotopic variations observed in field demonstrations.« less

Polarization-multiplexed rate-adaptive non-binary-quasi-cyclic-LDPC-coded multilevel modulation with coherent detection for optical transport networks.

PubMed

Arabaci, Murat; Djordjevic, Ivan B; Saunders, Ross; Marcoccia, Roberto M

2010-02-01

In order to achieve high-speed transmission over optical transport networks (OTNs) and maximize its throughput, we propose using a rate-adaptive polarization-multiplexed coded multilevel modulation with coherent detection based on component non-binary quasi-cyclic (QC) LDPC codes. Compared to prior-art bit-interleaved LDPC-coded modulation (BI-LDPC-CM) scheme, the proposed non-binary LDPC-coded modulation (NB-LDPC-CM) scheme not only reduces latency due to symbol- instead of bit-level processing but also provides either impressive reduction in computational complexity or striking improvements in coding gain depending on the constellation size. As the paper presents, compared to its prior-art binary counterpart, the proposed NB-LDPC-CM scheme addresses the needs of future OTNs, which are achieving the target BER performance and providing maximum possible throughput both over the entire lifetime of the OTN, better.

TDA : Transportation Development Act : statutes and California codes of regulations.

DOT National Transportation Integrated Search

2009-03-01

The Mills-Alquist-Deddeh Act (SB 325) was enacted by the California Legislature to improve : existing public transportation services and encourage regional transportation coordination. : Known as the Transportation Development Act (TDA) of 1971, this...

A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

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

Burns, T.D. Jr.

1996-05-01

The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run withmore » little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.« less

hybrid\\scriptsize{{MANTIS}}: a CPU-GPU Monte Carlo method for modeling indirect x-ray detectors with columnar scintillators

NASA Astrophysics Data System (ADS)

Sharma, Diksha; Badal, Andreu; Badano, Aldo

2012-04-01

The computational modeling of medical imaging systems often requires obtaining a large number of simulated images with low statistical uncertainty which translates into prohibitive computing times. We describe a novel hybrid approach for Monte Carlo simulations that maximizes utilization of CPUs and GPUs in modern workstations. We apply the method to the modeling of indirect x-ray detectors using a new and improved version of the code \\scriptsize{{MANTIS}}, an open source software tool used for the Monte Carlo simulations of indirect x-ray imagers. We first describe a GPU implementation of the physics and geometry models in fast\\scriptsize{{DETECT}}2 (the optical transport model) and a serial CPU version of the same code. We discuss its new features like on-the-fly column geometry and columnar crosstalk in relation to the \\scriptsize{{MANTIS}} code, and point out areas where our model provides more flexibility for the modeling of realistic columnar structures in large area detectors. Second, we modify \\scriptsize{{PENELOPE}} (the open source software package that handles the x-ray and electron transport in \\scriptsize{{MANTIS}}) to allow direct output of location and energy deposited during x-ray and electron interactions occurring within the scintillator. This information is then handled by optical transport routines in fast\\scriptsize{{DETECT}}2. A load balancer dynamically allocates optical transport showers to the GPU and CPU computing cores. Our hybrid\\scriptsize{{MANTIS}} approach achieves a significant speed-up factor of 627 when compared to \\scriptsize{{MANTIS}} and of 35 when compared to the same code running only in a CPU instead of a GPU. Using hybrid\\scriptsize{{MANTIS}}, we successfully hide hours of optical transport time by running it in parallel with the x-ray and electron transport, thus shifting the computational bottleneck from optical to x-ray transport. The new code requires much less memory than \\scriptsize{{MANTIS}} and, as a result, allows us to efficiently simulate large area detectors.

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.

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.

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

Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). Hanford Environmental Dose Reconstruction Project

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

Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.

1994-02-01

The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters,more » and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user`s guide, and a programmer`s guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user`s guide to the model with emphasis on running the code. The user`s guide contains information about the model input and output. The third section is a programmer`s guide to the code. It discusses the hardware and software required to run the code. The programmer`s guide also discusses program structure and each of the program elements.« less

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.

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.

Minerva: Cylindrical coordinate extension for Athena

NASA Astrophysics Data System (ADS)

Skinner, M. Aaron; Ostriker, Eve C.

2013-02-01

Minerva is a cylindrical coordinate extension of the Athena astrophysical MHD code of Stone, Gardiner, Teuben, and Hawley. The extension follows the approach of Athena's original developers and has been designed to alter the existing Cartesian-coordinates code as minimally and transparently as possible. The numerical equations in cylindrical coordinates are formulated to maintain consistency with constrained transport (CT), a central feature of the Athena algorithm, while making use of previously implemented code modules such as the Riemann solvers. Angular momentum transport, which is critical in astrophysical disk systems dominated by rotation, is treated carefully.

Solar Proton Transport within an ICRU Sphere Surrounded by a Complex Shield: Combinatorial Geometry

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

The 3DHZETRN code, with improved neutron and light ion (Z (is) less than 2) transport procedures, was recently developed and compared to Monte Carlo (MC) simulations using simplified spherical geometries. It was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in general combinatorial geometry. A more complex shielding structure with internal parts surrounding a tissue sphere is considered and compared against MC simulations. It is shown that even in the more complex geometry, 3DHZETRN agrees well with the MC codes and maintains a high degree of computational efficiency.

Verification of Three Dimensional Triangular Prismatic Discrete Ordinates Transport Code ENSEMBLE-TRIZ by Comparison with Monte Carlo Code GMVP

NASA Astrophysics Data System (ADS)

Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi

2014-06-01

This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.

NPSS Space Team

NASA Technical Reports Server (NTRS)

Lavelle, Tom

2003-01-01

The objective is to increase the usability of the current NPSS code/architecture by incorporating an advanced space transportation propulsion system capability into the existing NPSS code and begin defining advanced capabilities for NPSS and provide an enhancement for the NPSS code/architecture.

Validation of the WIMSD4M cross-section generation code with benchmark results

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

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section librariesmore » for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

30 CFR 206.56 - Transportation allowances-general.

Code of Federal Regulations, 2010 CFR

2010-07-01

... oil has been determined under § 206.52 or § 206.53 of this subpart at a point (e.g., sales point or... sales type code may not exceed 50 percent of the value of the oil at the point of sale as determined under § 206.52 of this subpart. Transportation costs cannot be transferred between sales type codes or...

LLNL Mercury Project Trinity Open Science Final Report

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

Dawson, Shawn A.

The Mercury Monte Carlo particle transport code is used to simulate the transport of radiation through urban environments. These challenging calculations include complicated geometries and require significant computational resources to complete. In the proposed Trinity Open Science calculations, I will investigate computer science aspects of the code which are relevant to convergence of the simulation quantities with increasing Monte Carlo particle counts.

THC-MP: High performance numerical simulation of reactive transport and multiphase flow in porous media

NASA Astrophysics Data System (ADS)

Wei, Xiaohui; Li, Weishan; Tian, Hailong; Li, Hongliang; Xu, Haixiao; Xu, Tianfu

2015-07-01

The numerical simulation of multiphase flow and reactive transport in the porous media on complex subsurface problem is a computationally intensive application. To meet the increasingly computational requirements, this paper presents a parallel computing method and architecture. Derived from TOUGHREACT that is a well-established code for simulating subsurface multi-phase flow and reactive transport problems, we developed a high performance computing THC-MP based on massive parallel computer, which extends greatly on the computational capability for the original code. The domain decomposition method was applied to the coupled numerical computing procedure in the THC-MP. We designed the distributed data structure, implemented the data initialization and exchange between the computing nodes and the core solving module using the hybrid parallel iterative and direct solver. Numerical accuracy of the THC-MP was verified through a CO2 injection-induced reactive transport problem by comparing the results obtained from the parallel computing and sequential computing (original code). Execution efficiency and code scalability were examined through field scale carbon sequestration applications on the multicore cluster. The results demonstrate successfully the enhanced performance using the THC-MP on parallel computing facilities.

Investigation of Fluctuation-Induced Electron Transport in Hall Thrusters with a 2D Hybrid Code in the Azimuthal and Axial Coordinates

NASA Astrophysics Data System (ADS)

Fernandez, Eduardo; Borelli, Noah; Cappelli, Mark; Gascon, Nicolas

2003-10-01

Most current Hall thruster simulation efforts employ either 1D (axial), or 2D (axial and radial) codes. These descriptions crucially depend on the use of an ad-hoc perpendicular electron mobility. Several models for the mobility are typically invoked: classical, Bohm, empirically based, wall-induced, as well as combinations of the above. Experimentally, it is observed that fluctuations and electron transport depend on axial distance and operating parameters. Theoretically, linear stability analyses have predicted a number of unstable modes; yet the nonlinear character of the fluctuations and/or their contribution to electron transport remains poorly understood. Motivated by these observations, a 2D code in the azimuthal and axial coordinates has been written. In particular, the simulation self-consistently calculates the azimuthal disturbances resulting in fluctuating drifts, which in turn (if properly correlated with plasma density disturbances) result in fluctuation-driven electron transport. The characterization of the turbulence at various operating parameters and across the channel length is also the object of this study. A description of the hybrid code used in the simulation as well as the initial results will be presented.

Nonambipolar Transport and Torque in Perturbed Equilibria

NASA Astrophysics Data System (ADS)

Logan, N. C.; Park, J.-K.; Wang, Z. R.; Berkery, J. W.; Kim, K.; Menard, J. E.

2013-10-01

A new Perturbed Equilibrium Nonambipolar Transport (PENT) code has been developed to calculate the neoclassical toroidal torque from radial current composed of both passing and trapped particles in perturbed equilibria. This presentation outlines the physics approach used in the development of the PENT code, with emphasis on the effects of retaining general aspect-ratio geometric effects. First, nonambipolar transport coefficients and corresponding neoclassical toroidal viscous (NTV) torque in perturbed equilibria are re-derived from the first order gyro-drift-kinetic equation in the ``combined-NTV'' PENT formalism. The equivalence of NTV torque and change in potential energy due to kinetic effects [J-K. Park, Phys. Plas., 2011] is then used to showcase computational challenges shared between PENT and stability codes MISK and MARS-K. Extensive comparisons to a reduced model, which makes numerous large aspect ratio approximations, are used throughout to emphasize geometry dependent physics such as pitch angle resonances. These applications make extensive use of the PENT code's native interfacing with the Ideal Perturbed Equilibrium Code (IPEC), and the combination of these codes is a key step towards an iterative solver for self-consistent perturbed equilibrium torque. Supported by US DOE contract #DE-AC02-09CH11466 and the DOE Office of Science Graduate Fellowship administered by the Oak Ridge Institute for Science & Education under contract #DE-AC05-06OR23100.

The Initial Atmospheric Transport (IAT) Code: Description and Validation

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

Morrow, Charles W.; Bartel, Timothy James

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34Dmore » accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.« less

Transport studies in high-performance field reversed configuration plasmas

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

Gupta, S., E-mail: sgupta@trialphaenergy.com; Barnes, D. C.; Dettrick, S. A.

2016-05-15

A significant improvement of field reversed configuration (FRC) lifetime and plasma confinement times in the C-2 plasma, called High Performance FRC regime, has been observed with neutral beam injection (NBI), improved edge stability, and better wall conditioning [Binderbauer et al., Phys. Plasmas 22, 056110 (2015)]. A Quasi-1D (Q1D) fluid transport code has been developed and employed to carry out transport analysis of such C-2 plasma conditions. The Q1D code is coupled to a Monte-Carlo code to incorporate the effect of fast ions, due to NBI, on the background FRC plasma. Numerically, the Q1D transport behavior with enhanced transport coefficients (butmore » with otherwise classical parametric dependencies) such as 5 times classical resistive diffusion, classical thermal ion conductivity, 20 times classical electron thermal conductivity, and classical fast ion behavior fit with the experimentally measured time evolution of the excluded flux radius, line-integrated density, and electron/ion temperature. The numerical study shows near sustainment of poloidal flux for nearly 1 ms in the presence of NBI.« less

Groundwater flow with energy transport and water-ice phase change: Numerical simulations, benchmarks, and application to freezing in peat bogs

USGS Publications Warehouse

McKenzie, J.M.; Voss, C.I.; Siegel, D.I.

2007-01-01

In northern peatlands, subsurface ice formation is an important process that can control heat transport, groundwater flow, and biological activity. Temperature was measured over one and a half years in a vertical profile in the Red Lake Bog, Minnesota. To successfully simulate the transport of heat within the peat profile, the U.S. Geological Survey's SUTRA computer code was modified. The modified code simulates fully saturated, coupled porewater-energy transport, with freezing and melting porewater, and includes proportional heat capacity and thermal conductivity of water and ice, decreasing matrix permeability due to ice formation, and latent heat. The model is verified by correctly simulating the Lunardini analytical solution for ice formation in a porous medium with a mixed ice-water zone. The modified SUTRA model correctly simulates the temperature and ice distributions in the peat bog. Two possible benchmark problems for groundwater and energy transport with ice formation and melting are proposed that may be used by other researchers for code comparison. ?? 2006 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.

2006 Interregional Transportation Improvement Program.

DOT National Transportation Integrated Search

2006-01-01

The Department of Transportations (Department) five-year Interregional Transportation : Improvement Program (ITIP) is prepared pursuant to Government Code 14526 and : consists of projects funded from the interregional share, which is 25 percent of...

Comparison of EGS4 and MCNP Monte Carlo codes when calculating radiotherapy depth doses.

PubMed

Love, P A; Lewis, D G; Al-Affan, I A; Smith, C W

1998-05-01

The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

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

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

Radiation Transport and Shielding for Space Exploration and High Speed Flight Transportation

NASA Technical Reports Server (NTRS)

Maung, Khin Maung; Trapathi, R. K.

1997-01-01

Transportation of ions and neutrons in matter is of direct interest in several technologically important and scientific areas, including space radiation, cosmic ray propagation studies in galactic medium, nuclear power plants and radiological effects that impact industrial and public health. For the proper assessment of radiation exposure, both reliable transport codes and accurate data are needed. Nuclear cross section data is one of the essential inputs into the transport codes. In order to obtain an accurate parametrization of cross section data, theoretical input is indispensable especially for processes where there is little or no experimental data available. In this grant period work has been done on the studies of the use of relativistic equations and their one-body limits. The results will be useful in choosing appropriate effective one-body equation for reaction calculations. Work has also been done to improve upon the data base needed for the transport codes used in the studies of radiation transport and shielding for space exploration and high speed flight transportation. A phenomenological model was developed for the total absorption cross sections valid for any system of charged and/or uncharged collision pairs for the entire energy range. The success of the model is gratifying. It is being used by other federal agencies, national labs and universities. A list of publications based on the work during the grant period is given below and copies are enclosed with this report.

DARTAB: a program to combine airborne radionuclide environmental exposure data with dosimetric and health effects data to generate tabulations of predicted health impacts

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

Begovich, C.L.; Eckerman, K.F.; Schlatter, E.C.

1981-08-01

The DARTAB computer code combines radionuclide environmental exposure data with dosimetric and health effects data to generate tabulations of the predicted impact of radioactive airborne effluents. DARTAB is independent of the environmental transport code used to generate the environmental exposure data and the codes used to produce the dosimetric and health effects data. Therefore human dose and risk calculations need not be added to every environmental transport code. Options are included in DARTAB to permit the user to request tabulations by various topics (e.g., cancer site, exposure pathway, etc.) to facilitate characterization of the human health impacts of the effluents.more » The DARTAB code was written at ORNL for the US Environmental Protection Agency, Office of Radiation Programs.« less

Transoptr — A second order beam transport design code with optimization and constraints

NASA Astrophysics Data System (ADS)

Heighway, E. A.; Hutcheon, R. M.

1981-08-01

This code was written initially to design an achromatic and isochronous reflecting magnet and has been extended to compete in capability (for constrained problems) with TRANSPORT. Its advantage is its flexibility in that the user writes a routine to describe his transport system. The routine allows the definition of general variables from which the system parameters can be derived. Further, the user can write any constraints he requires as algebraic equations relating the parameters. All variables may be used in either a first or second order optimization.

MPACT Standard Input User s Manual, Version 2.2.0

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

Collins, Benjamin S.; Downar, Thomas; Fitzgerald, Andrew

The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

Programmers manual for a one-dimensional Lagrangian transport model

USGS Publications Warehouse

Schoellhamer, D.H.; Jobson, H.E.

1986-01-01

A one-dimensional Lagrangian transport model for simulating water-quality constituents such as temperature, dissolved oxygen , and suspended sediment in rivers is presented in this Programmers Manual. Lagrangian transport modeling techniques, the model 's subroutines, and the user-written decay-coefficient subroutine are discussed in detail. Appendices list the program codes. The Programmers Manual is intended for the model user who needs to modify code either to adapt the model to a particular need or to use reaction kinetics not provided with the model. (Author 's abstract)

Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

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

Chapman, Bryan Scott; MacQuigg, Michael Robert; Wysong, Andrew Russell

In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k eff.

Radiation transport codes for potential applications related to radiobiology and radiotherapy using protons, neutrons, and negatively charged pions

NASA Technical Reports Server (NTRS)

Armstrong, T. W.

1972-01-01

Several Monte Carlo radiation transport computer codes are used to predict quantities of interest in the fields of radiotherapy and radiobiology. The calculational methods are described and comparisions of calculated and experimental results are presented for dose distributions produced by protons, neutrons, and negatively charged pions. Comparisons of calculated and experimental cell survival probabilities are also presented.

SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages

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

Russel, E.

1997-11-01

This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.

49 CFR Appendix C to Part 215 - FRA Freight Car Standards Defect Code

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false FRA Freight Car Standards Defect Code C Appendix C... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION RAILROAD FREIGHT CAR SAFETY STANDARDS Pt. 215, App. C Appendix C to... 13/4″. (C)(1) Rim thickness is 11/16″ or less; (2) Rim thickness is 5/8″ or less; (3) Rim thickness...

Study of negative ion transport phenomena in a plasma source

NASA Astrophysics Data System (ADS)

Riz, D.; Paméla, J.

1996-07-01

NIETZSCHE (Negative Ions Extraction and Transport ZSimulation Code for HydrogEn species) is a negative ion (NI) transport code developed at Cadarache. This code calculates NI trajectories using a 3D Monte-Carlo technique, taking into account the main destruction processes, as well as elastic collisions (H-/H+) and charge exchanges (H-/H0). It determines the extraction probability of a NI created at a given position. According to the simulations, we have seen that in the case of volume production, only NI produced close to the plasma grid (PG) can be extracted. Concerning the surface production, we have studied how NI produced on the PG and accelerated by the plasma sheath backward into the source could be extracted. We demonstrate that elastic collisions and charge exchanges play an important role, which in some conditions dominates the magnetic filter effect, which acts as a magnetic mirror. NI transport in various conditions will be discussed: volume/surface production, high/low plasmas density, tent filter/transverse filter.

Cosmic-ray propagation with DRAGON2: I. numerical solver and astrophysical ingredients

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

Evoli, Carmelo; Gaggero, Daniele; Vittino, Andrea

2017-02-01

We present version 2 of the DRAGON code designed for computing realistic predictions of the CR densities in the Galaxy. The code numerically solves the interstellar CR transport equation (including inhomogeneous and anisotropic diffusion, either in space and momentum, advective transport and energy losses), under realistic conditions. The new version includes an updated numerical solver and several models for the astrophysical ingredients involved in the transport equation. Improvements in the accuracy of the numerical solution are proved against analytical solutions and in reference diffusion scenarios. The novel features implemented in the code allow to simulate the diverse scenarios proposed tomore » reproduce the most recent measurements of local and diffuse CR fluxes, going beyond the limitations of the homogeneous galactic transport paradigm. To this end, several applications using DRAGON2 are presented as well. This new version facilitates the users to include their own physical models by means of a modular C++ structure.« less

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

PubMed

Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

Un-collided-flux preconditioning for the first order transport equation

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

Rigley, M.; Koebbe, J.; Drumm, C.

2013-07-01

Two codes were tested for the first order neutron transport equation using finite element methods. The un-collided-flux solution is used as a preconditioner for each of these methods. These codes include a least squares finite element method and a discontinuous finite element method. The performance of each code is shown on problems in one and two dimensions. The un-collided-flux preconditioner shows good speedup on each of the given methods. The un-collided-flux preconditioner has been used on the second-order equation, and here we extend those results to the first order equation. (authors)

Nuclide Depletion Capabilities in the Shift Monte Carlo Code

DOE PAGES

Davidson, Gregory G.; Pandya, Tara M.; Johnson, Seth R.; ...

2017-12-21

A new depletion capability has been developed in the Exnihilo radiation transport code suite. This capability enables massively parallel domain-decomposed coupling between the Shift continuous-energy Monte Carlo solver and the nuclide depletion solvers in ORIGEN to perform high-performance Monte Carlo depletion calculations. This paper describes this new depletion capability and discusses its various features, including a multi-level parallel decomposition, high-order transport-depletion coupling, and energy-integrated power renormalization. Several test problems are presented to validate the new capability against other Monte Carlo depletion codes, and the parallel performance of the new capability is analyzed.

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.

MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations

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

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capabilitymore » of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.« less

Track-structure simulations for charged particles.

PubMed

Dingfelder, Michael

2012-11-01

Monte Carlo track-structure simulations provide a detailed and accurate picture of radiation transport of charged particles through condensed matter of biological interest. Liquid water serves as a surrogate for soft tissue and is used in most Monte Carlo track-structure codes. Basic theories of radiation transport and track-structure simulations are discussed and differences compared to condensed history codes highlighted. Interaction cross sections for electrons, protons, alpha particles, and light and heavy ions are required input data for track-structure simulations. Different calculation methods, including the plane-wave Born approximation, the dielectric theory, and semi-empirical approaches are presented using liquid water as a target. Low-energy electron transport and light ion transport are discussed as areas of special interest.

TOUGHREACT: a new code of the TOUGH Family for Non-Isothermal multiphase reactive geochemical transport in variably saturated geologic media

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

Xu, Tianfu; Sonnenthal, Eric; Spycher, Nicolas

Coupled modeling of subsurface multiphase fluid and heat flow, solute transport and chemical reactions can be used for the assessment of acid mine drainage remediation, waste disposal sites, hydrothermal convection, contaminant transport, and groundwater quality. We have developed a comprehensive numerical simulator, TOUGHREACT, which considers non-isothermal multi-component chemical transport in both liquid and gas phases. A wide range of subsurface thermo-physical-chemical processes is considered under various thermohydrological and geochemical conditions of pressure, temperature, water saturation, and ionic strength. The code can be applied to one-, two- or three-dimensional porous and fractured media with physical and chemical heterogeneity.

Aerothermo-Structural Analysis of Low Cost Composite Nozzle/Inlet Components

NASA Technical Reports Server (NTRS)

Shivakumar, Kuwigai; Challa, Preeli; Sree, Dave; Reddy, D.

1999-01-01

This research is a cooperative effort among the Turbomachinery and Propulsion Division of NASA Glenn, CCMR of NC A&T State University, and the Tuskegee University. The NC A&T is the lead center and Tuskegee University is the participating institution. Objectives of the research were to develop an integrated aerodynamic, thermal and structural analysis code for design of aircraft engine components, such as, nozzles and inlets made of textile composites; conduct design studies on typical inlets for hypersonic transportation vehicles and setup standards test examples and finally manufacture a scaled down composite inlet. These objectives are accomplished through the following seven tasks: (1) identify the relevant public domain codes for all three types of analysis; (2) evaluate the codes for the accuracy of results and computational efficiency; (3) develop aero-thermal and thermal structural mapping algorithms; (4) integrate all the codes into one single code; (5) write a graphical user interface to improve the user friendliness of the code; (6) conduct test studies for rocket based combined-cycle engine inlet; and finally (7) fabricate a demonstration inlet model using textile preform composites. Tasks one, two and six are being pursued. Selected and evaluated NPARC for flow field analysis, CSTEM for in-depth thermal analysis of inlets and nozzles and FRAC3D for stress analysis. These codes have been independently verified for accuracy and performance. In addition, graphical user interface based on micromechanics analysis for laminated as well as textile composites was developed. Demonstration of this code will be made at the conference. A rocket based combined cycle engine was selected for test studies. Flow field analysis of various inlet geometries were studied. Integration of codes is being continued. The codes developed are being applied to a candidate example of trailblazer engine proposed for space transportation. A successful development of the code will provide a simpler, faster and user-friendly tool for conducting design studies of aircraft and spacecraft engines, applicable in high speed civil transport and space missions.

Recent Progress in the Development of a Multi-Layer Green's Function Code for Ion Beam Transport

NASA Technical Reports Server (NTRS)

Tweed, John; Walker, Steven A.; Wilson, John W.; Tripathi, Ram K.

2008-01-01

To meet the challenge of future deep space programs, an accurate and efficient engineering code for analyzing the shielding requirements against high-energy galactic heavy radiation is needed. To address this need, a new Green's function code capable of simulating high charge and energy ions with either laboratory or space boundary conditions is currently under development. The computational model consists of combinations of physical perturbation expansions based on the scales of atomic interaction, multiple scattering, and nuclear reactive processes with use of the Neumann-asymptotic expansions with non-perturbative corrections. The code contains energy loss due to straggling, nuclear attenuation, nuclear fragmentation with energy dispersion and downshifts. Previous reports show that the new code accurately models the transport of ion beams through a single slab of material. Current research efforts are focused on enabling the code to handle multiple layers of material and the present paper reports on progress made towards that end.

Fully-kinetic Ion Simulation of Global Electrostatic Turbulent Transport in C-2U

NASA Astrophysics Data System (ADS)

Fulton, Daniel; Lau, Calvin; Bao, Jian; Lin, Zhihong; Tajima, Toshiki; TAE Team

2017-10-01

Understanding the nature of particle and energy transport in field-reversed configuration (FRC) plasmas is a crucial step towards an FRC-based fusion reactor. The C-2U device at Tri Alpha Energy (TAE) achieved macroscopically stable plasmas and electron energy confinement time which scaled favorably with electron temperature. This success led to experimental and theoretical investigation of turbulence in C-2U, including gyrokinetic ion simulations with the Gyrokinetic Toroidal Code (GTC). A primary objective of TAE's new C-2W device is to explore transport scaling in an extended parameter regime. In concert with the C-2W experimental campaign, numerical efforts have also been extended in A New Code (ANC) to use fully-kinetic (FK) ions and a Vlasov-Poisson field solver. Global FK ion simulations are presented. Future code development is also discussed.

Simulations of 4D edge transport and dynamics using the TEMPEST gyro-kinetic code

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, B. I.; Cohen, R. H.; Dorr, M. R.; Hittinger, J. A. F.; Kerbel, G. D.; Nevins, W. M.; Xiong, Z.; Xu, X. Q.

2006-10-01

Simulation results are presented for tokamak edge plasmas with a focus on the 4D (2r,2v) option of the TEMPEST continuum gyro-kinetic code. A detailed description of a variety of kinetic simulations is reported, including neoclassical radial transport from Coulomb collisions, electric field generation, dynamic response to perturbations by geodesic acoustic modes, and parallel transport on open magnetic-field lines. Comparison is made between the characteristics of the plasma solutions on closed and open magnetic-field line regions separated by a magnetic separatrix, and simple physical models are used to qualitatively explain the differences observed in mean flow and electric-field generation. The status of extending the simulations to 5D turbulence will be summarized. The code structure used in this ongoing project is also briefly described, together with future plans.

Statistical Analysis of the AIAA Drag Prediction Workshop CFD Solutions

NASA Technical Reports Server (NTRS)

Morrison, Joseph H.; Hemsch, Michael J.

2007-01-01

The first AIAA Drag Prediction Workshop (DPW), held in June 2001, evaluated the results from an extensive N-version test of a collection of Reynolds-Averaged Navier-Stokes CFD codes. The code-to-code scatter was more than an order of magnitude larger than desired for design and experimental validation of cruise conditions for a subsonic transport configuration. The second AIAA Drag Prediction Workshop, held in June 2003, emphasized the determination of installed pylon-nacelle drag increments and grid refinement studies. The code-to-code scatter was significantly reduced compared to the first DPW, but still larger than desired. However, grid refinement studies showed no significant improvement in code-to-code scatter with increasing grid refinement. The third AIAA Drag Prediction Workshop, held in June 2006, focused on the determination of installed side-of-body fairing drag increments and grid refinement studies for clean attached flow on wing alone configurations and for separated flow on the DLR-F6 subsonic transport model. This report compares the transonic cruise prediction results of the second and third workshops using statistical analysis.

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.

Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

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

Wagner, John C; Peplow, Douglas E.; Mosher, Scott W

2011-01-01

This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or moremore » localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.« less

Prompt Radiation Protection Factors

DTIC Science & Technology

2018-02-01

dimensional Monte-Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection factors (ratio of dose in the open to...radiation was performed using the three dimensional Monte- Carlo radiation transport code MCNP (Monte Carlo N-Particle) and the evaluation of the protection...by detonation of a nuclear device have placed renewed emphasis on evaluation of the consequences in case of such an event. The Defense Threat

Space Radiation Transport Codes: A Comparative Study for Galactic Cosmic Rays Environment

NASA Astrophysics Data System (ADS)

Tripathi, Ram; Wilson, John W.; Townsend, Lawrence W.; Gabriel, Tony; Pinsky, Lawrence S.; Slaba, Tony

For long duration and/or deep space human missions, protection from severe space radiation exposure is a challenging design constraint and may be a potential limiting factor. The space radiation environment consists of galactic cosmic rays (GCR), solar particle events (SPE), trapped radiation, and includes ions of all the known elements over a very broad energy range. These ions penetrate spacecraft materials producing nuclear fragments and secondary particles that damage biological tissues, microelectronic devices, and materials. In deep space missions, where the Earth's magnetic field does not provide protection from space radiation, the GCR environment is significantly enhanced due to the absence of geomagnetic cut-off and is a major component of radiation exposure. Accurate risk assessments critically depend on the accuracy of the input information as well as radiation transport codes used, and so systematic verification of codes is necessary. In this study, comparisons are made between the deterministic code HZETRN2006 and the Monte Carlo codes HETC-HEDS and FLUKA for an aluminum shield followed by a water target exposed to the 1977 solar minimum GCR spectrum. Interaction and transport of high charge ions present in GCR radiation environment provide a more stringent constraint in the comparison of the codes. Dose, dose equivalent and flux spectra are compared; details of the comparisons will be discussed, and conclusions will be drawn for future directions.

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.

Transportation fuel research and development : statistically validated codes and standards

DOT National Transportation Integrated Search

2007-08-28

The recent establishment of the National University Transportation Center at MST under the "Safe, Accountable, Flexible, Efficient Transportation Equity Act: A Legacy for Users," expands the research and education activities to include alternative tr...

14 CFR 257.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... this part: (a) Air transportation means foreign air transportation or interstate air transportation as defined in 49 U.S.C. 40102 (a)(23) and (25) respectively. (b) Carrier means any air carrier or foreign air... scheduled passenger air transportation, including by wet lease. (c) Code-sharing arrangement means an...

Extension of the BRYNTRN code to monoenergetic light ion beams

NASA Technical Reports Server (NTRS)

Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.

1994-01-01

A monoenergetic version of the BRYNTRN transport code is extended to beam transport of light ions (H-2, H-3, He-3, and He-4) in shielding materials (thick targets). The redistribution of energy in nuclear reactions is included in transport solutions that use nuclear fragmentation models. We also consider an equilibrium target-fragment spectrum for nuclei with mass number greater than four to include target fragmentation effects in the linear energy transfer (LET) spectrum. Illustrative results for water and aluminum shielding, including energy and LET spectra, are discussed for high-energy beams of H-2 and He-4.

Green's function methods in heavy ion shielding

NASA Technical Reports Server (NTRS)

Wilson, John W.; Costen, Robert C.; Shinn, Judy L.; Badavi, Francis F.

1993-01-01

An analytic solution to the heavy ion transport in terms of Green's function is used to generate a highly efficient computer code for space applications. The efficiency of the computer code is accomplished by a nonperturbative technique extending Green's function over the solution domain. The computer code can also be applied to accelerator boundary conditions to allow code validation in laboratory experiments.

A Green's function method for heavy ion beam transport

NASA Technical Reports Server (NTRS)

Shinn, J. L.; Wilson, J. W.; Schimmerling, W.; Shavers, M. R.; Miller, J.; Benton, E. V.; Frank, A. L.; Badavi, F. F.

1995-01-01

The use of Green's function has played a fundamental role in transport calculations for high-charge high-energy (HZE) ions. Two recent developments have greatly advanced the practical aspects of implementation of these methods. The first was the formulation of a closed-form solution as a multiple fragmentation perturbation series. The second was the effective summation of the closed-form solution through nonperturbative techniques. The nonperturbative methods have been recently extended to an inhomogeneous, two-layer transport media to simulate the lead scattering foil present in the Lawrence Berkeley Laboratories (LBL) biomedical beam line used for cancer therapy. Such inhomogeneous codes are necessary for astronaut shielding in space. The transport codes utilize the Langley Research Center atomic and nuclear database. Transport code and database evaluation are performed by comparison with experiments performed at the LBL Bevalac facility using 670 A MeV 20Ne and 600 A MeV 56Fe ion beams. The comparison with a time-of-flight and delta E detector measurement for the 20Ne beam and the plastic nuclear track detectors for 56Fe show agreement up to 35%-40% in water and aluminium targets, respectively.

A Conference on Spacecraft Charging Technology - 1978, held at U.S. Air Force Academy, Colorado Springs, Colorado, October 31 - November 2, 1978.

DTIC Science & Technology

1978-01-01

complex, applications of the code . NASCAP CODE DESCRIPTION The NASCAP code is a finite-element spacecraft-charging simulation that is written in FORTRAN ...transport code POEM (ref. 1), is applicable to arbitrary dielectrics, source spectra, and current time histories. The code calculations are illustrated by...iaxk ’. Vlbouced _DstributionL- 9TNA Availability Codes %ELECTEf Nationa Aeronautics and Dist. Spec al TAvalland/or. MAY 2 21980 Space Administration

Linear energy transfer in water phantom within SHIELD-HIT transport code

NASA Astrophysics Data System (ADS)

Ergun, A.; Sobolevsky, N.; Botvina, A. S.; Buyukcizmeci, N.; Latysheva, L.; Ogul, R.

2017-02-01

The effect of irradiation in tissue is important in hadron therapy for the dose measurement and treatment planning. This biological effect is defined by an equivalent dose H which depends on the Linear Energy Transfer (LET). Usually, H can be expressed in terms of the absorbed dose D and the quality factor K of the radiation under consideration. In literature, various types of transport codes have been used for modeling and simulation of the interaction of the beams of protons and heavier ions with tissue-equivalent materials. In this presentation we used SHIELD-HIT code to simulate decomposition of the absorbed dose by LET in water for 16O beams. A more detailed description of capabilities of the SHIELD-HIT code can be found in the literature.

Overview of Edge Simulation Laboratory (ESL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.; Dorr, M.; Hittinger, J.; Rognlien, T.; Umansky, M.; Xiong, A.; Xu, X.; Belli, E.; Candy, J.; Snyder, P.; Colella, P.; Martin, D.; Sternberg, T.; van Straalen, B.; Bodi, K.; Krasheninnikov, S.

2006-10-01

The ESL is a new collaboration to build a full-f electromagnetic gyrokinetic code for tokamak edge plasmas using continuum methods. Target applications are edge turbulence and transport (neoclassical and anomalous), and edge-localized modes. Initially the project has three major threads: (i) verification and validation of TEMPEST, the project's initial (electrostatic) edge code which can be run in 4D (neoclassical and transport-timescale applications) or 5D (turbulence); (ii) design of the next generation code, which will include more complete physics (electromagnetics, fluid equation option, improved collisions) and advanced numerics (fully conservative, high-order discretization, mapped multiblock grids, adaptivity), and (iii) rapid-prototype codes to explore the issues attached to solving fully nonlinear gyrokinetics with steep radial gradiens. We present a brief summary of the status of each of these activities.

Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

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

Charles A. Wemple; Joshua J. Cogliati

2005-04-01

A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random numbermore » generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.« less

Using Modern C++ Idiom for the Discretisation of Sets of Coupled Transport Equations in Numerical Plasma Physics

NASA Astrophysics Data System (ADS)

van Dijk, Jan; Hartgers, Bart; van der Mullen, Joost

2006-10-01

Self-consistent modelling of plasma sources requires a simultaneous treatment of multiple physical phenomena. As a result plasma codes have a high degree of complexity. And with the growing interest in time-dependent modelling of non-equilibrium plasma in three dimensions, codes tend to become increasingly hard to explain-and-maintain. As a result of these trends there has been an increased interest in the software-engineering and implementation aspects of plasma modelling in our group at Eindhoven University of Technology. In this contribution we will present modern object-oriented techniques in C++ to solve an old problem: that of the discretisation of coupled linear(ized) equations involving multiple field variables on ortho-curvilinear meshes. The `LinSys' code has been tailored to the transport equations that occur in transport physics. The implementation has been made both efficient and user-friendly by using modern idiom like expression templates and template meta-programming. Live demonstrations will be given. The code is available to interested parties; please visit www.dischargemodelling.org.

22 CFR 228.22 - Air transportation.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 22 Foreign Relations 1 2013-04-01 2013-04-01 false Air transportation. 228.22 Section 228.22... § 228.22 Air transportation. The Fly America Act, Title 49 of the United States Code, Subtitle VII, part A, subpart I, Chapter 401, 40118—Government-Financed Air Transportation, is applicable to all...

22 CFR 228.22 - Air transportation.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 22 Foreign Relations 1 2014-04-01 2014-04-01 false Air transportation. 228.22 Section 228.22... § 228.22 Air transportation. The Fly America Act, Title 49 of the United States Code, Subtitle VII, part A, subpart I, Chapter 401, 40118—Government-Financed Air Transportation, is applicable to all...

22 CFR 228.22 - Air transportation.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 22 Foreign Relations 1 2012-04-01 2012-04-01 false Air transportation. 228.22 Section 228.22... § 228.22 Air transportation. The Fly America Act, Title 49 of the United States Code, Subtitle VII, part A, subpart I, Chapter 401, 40118—Government-Financed Air Transportation, is applicable to all...

Modeling of ion orbit loss and intrinsic toroidal rotation with the COGENT code

NASA Astrophysics Data System (ADS)

Dorf, M.; Dorr, M.; Cohen, R.; Rognlien, T.; Hittinger, J.

2014-10-01

We discuss recent advances in cross-separatrix neoclassical transport simulations with COGENT, a continuum gyro-kinetic code being developed by the Edge Simulation Laboratory (ESL) collaboration. The COGENT code models the axisymmetric transport properties of edge plasmas including the effects of nonlinear (Fokker-Planck) collisions and a self-consistent electrostatic potential. Our recent work has focused on studies of ion orbit loss and the associated toroidal rotation driven by this mechanism. The results of the COGENT simulations are discussed and analyzed for the parameters of the DIII-D experiment. Work performed for USDOE at LLNL under Contract DE-AC52-07NA27344.

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

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

Kim, Kang Seog; Clarno, Kevin T.; Gentry, Cole

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Code of Federal Regulations, 2013 CFR

2013-01-01

... hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... processing and hermetically sealed containers; closures; code marking; heat processing; incubation. (a... storage and transportation as evidenced by the incubation test. (h) Lots of canned products shall be...

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Code of Federal Regulations, 2012 CFR

2012-01-01

... hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... processing and hermetically sealed containers; closures; code marking; heat processing; incubation. (a... storage and transportation as evidenced by the incubation test. (h) Lots of canned products shall be...

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Code of Federal Regulations, 2011 CFR

2011-01-01

... hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... processing and hermetically sealed containers; closures; code marking; heat processing; incubation. (a... storage and transportation as evidenced by the incubation test. (h) Lots of canned products shall be...

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Code of Federal Regulations, 2014 CFR

2014-01-01

... hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... processing and hermetically sealed containers; closures; code marking; heat processing; incubation. (a... storage and transportation as evidenced by the incubation test. (h) Lots of canned products shall be...

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Code of Federal Regulations, 2010 CFR

2010-01-01

... hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... processing and hermetically sealed containers; closures; code marking; heat processing; incubation. (a... storage and transportation as evidenced by the incubation test. (h) Lots of canned products shall be...