Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

SABRINA: an interactive three-dimensional geometry-mnodeling program for MCNP

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

West, J.T. III

SABRINA is a fully interactive three-dimensional geometry-modeling program for MCNP, a Los Alamos Monte Carlo code for neutron and photon transport. In SABRINA, a user constructs either body geometry or surface geometry models and debugs spatial descriptions for the resulting objects. This enhanced capability significantly reduces effort in constructing and debugging complicated three-dimensional geometry models for Monte Carlo analysis. 2 refs., 33 figs.

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

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

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

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

Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – C k's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usagemore » are discussed.« less

Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.

PubMed

Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P

2007-01-01

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

Features of MCNP6 Relevant to Medical Radiation Physics

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

Hughes, H. Grady III; Goorley, John T.

2012-08-29

MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-basedmore » isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.« less

Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

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

Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl; Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago; Molina, F.

2016-07-07

The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

Monte Carlo N Particle code - Dose distribution of clinical electron beams in inhomogeneous phantoms

PubMed Central

Nedaie, H. A.; Mosleh-Shirazi, M. A.; Allahverdi, M.

2013-01-01

Electron dose distributions calculated using the currently available analytical methods can be associated with large uncertainties. The Monte Carlo method is the most accurate method for dose calculation in electron beams. Most of the clinical electron beam simulation studies have been performed using non- MCNP [Monte Carlo N Particle] codes. Given the differences between Monte Carlo codes, this work aims to evaluate the accuracy of MCNP4C-simulated electron dose distributions in a homogenous phantom and around inhomogeneities. Different types of phantoms ranging in complexity were used; namely, a homogeneous water phantom and phantoms made of polymethyl methacrylate slabs containing different-sized, low- and high-density inserts of heterogeneous materials. Electron beams with 8 and 15 MeV nominal energy generated by an Elekta Synergy linear accelerator were investigated. Measurements were performed for a 10 cm × 10 cm applicator at a source-to-surface distance of 100 cm. Individual parts of the beam-defining system were introduced into the simulation one at a time in order to show their effect on depth doses. In contrast to the first scattering foil, the secondary scattering foil, X and Y jaws and applicator provide up to 5% of the dose. A 2%/2 mm agreement between MCNP and measurements was found in the homogenous phantom, and in the presence of heterogeneities in the range of 1-3%, being generally within 2% of the measurements for both energies in a "complex" phantom. A full-component simulation is necessary in order to obtain a realistic model of the beam. The MCNP4C results agree well with the measured electron dose distributions. PMID:23533162

Monte Carlo dose calculations of beta-emitting sources for intravascular brachytherapy: a comparison between EGS4, EGSnrc, and MCNP.

PubMed

Wang, R; Li, X A

2001-02-01

The dose parameters for the beta-particle emitting 90Sr/90Y source for intravascular brachytherapy (IVBT) have been calculated by different investigators. At a distant distance from the source, noticeable differences are seen in these parameters calculated using different Monte Carlo codes. The purpose of this work is to quantify as well as to understand these differences. We have compared a series of calculations using an EGS4, an EGSnrc, and the MCNP Monte Carlo codes. Data calculated and compared include the depth dose curve for a broad parallel beam of electrons, and radial dose distributions for point electron sources (monoenergetic or polyenergetic) and for a real 90Sr/90Y source. For the 90Sr/90Y source, the doses at the reference position (2 mm radial distance) calculated by the three code agree within 2%. However, the differences between the dose calculated by the three codes can be over 20% in the radial distance range interested in IVBT. The difference increases with radial distance from source, and reaches 30% at the tail of dose curve. These differences may be partially attributed to the different multiple scattering theories and Monte Carlo models for electron transport adopted in these three codes. Doses calculated by the EGSnrc code are more accurate than those by the EGS4. The two calculations agree within 5% for radial distance <6 mm.

MUFFSgenMC: An Open Source MUon Flexible Framework for Spectral GENeration for Monte Carlo Applications

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

Chatzidakis, Stylianos; Greulich, Christopher

A cosmic ray Muon Flexible Framework for Spectral GENeration for Monte Carlo Applications (MUFFSgenMC) has been developed to support state-of-the-art cosmic ray muon tomographic applications. The flexible framework allows for easy and fast creation of source terms for popular Monte Carlo applications like GEANT4 and MCNP. This code framework simplifies the process of simulations used for cosmic ray muon tomography.

Verification of BWR Turbine Skyshine Dose with the MCNP5 Code Based on an Experiment Made at SHIMANE Nuclear Power Station

NASA Astrophysics Data System (ADS)

Tayama, Ryuichi; Wakasugi, Kenichi; Kawanaka, Ikunori; Kadota, Yoshinobu; Murakami, Yasuhiro

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

Performance Study of Monte Carlo Codes on Xeon Phi Coprocessors — Testing MCNP 6.1 and Profiling ARCHER Geometry Module on the FS7ONNi Problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Wolfe, Noah; Lin, Hui; Zieb, Kris; Ji, Wei; Caracappa, Peter; Carothers, Christopher; Xu, X. George

2017-09-01

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

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.

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.

Benchmarking study of the MCNP code against cold critical experiments

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

Sitaraman, S.

1991-01-01

The purpose of this study was to benchmark the widely used Monte Carlo code MCNP against a set of cold critical experiments with a view to using the code as a means of independently verifying the performance of faster but less accurate Monte Carlo and deterministic codes. The experiments simulated consisted of both fast and thermal criticals as well as fuel in a variety of chemical forms. A standard set of benchmark cold critical experiments was modeled. These included the two fast experiments, GODIVA and JEZEBEL, the TRX metallic uranium thermal experiments, the Babcock and Wilcox oxide and mixed oxidemore » experiments, and the Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory (PNL) nitrate solution experiments. The principal case studied was a small critical experiment that was performed with boiling water reactor bundles.« less

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Comparison of Three Methods of Calculation, Experimental and Monte Carlo Simulation in Investigation of Organ Doses (Thyroid, Sternum, Cervical Vertebra) in Radioiodine Therapy

PubMed Central

Shahbazi-Gahrouei, Daryoush; Ayat, Saba

2012-01-01

Radioiodine therapy is an effective method for treating thyroid cancer carcinoma, but it has some affects on normal tissues, hence dosimetry of vital organs is important to weigh the risks and benefits of this method. The aim of this study is to measure the absorbed doses of important organs by Monte Carlo N Particle (MCNP) simulation and comparing the results of different methods of dosimetry by performing a t-paired test. To calculate the absorbed dose of thyroid, sternum, and cervical vertebra using the MCNP code, *F8 tally was used. Organs were simulated by using a neck phantom and Medical Internal Radiation Dosimetry (MIRD) method. Finally, the results of MCNP, MIRD, and Thermoluminescent dosimeter (TLD) measurements were compared by SPSS software. The absorbed dose obtained by Monte Carlo simulations for 100, 150, and 175 mCi administered 131I was found to be 388.0, 427.9, and 444.8 cGy for thyroid, 208.7, 230.1, and 239.3 cGy for sternum and 272.1, 299.9, and 312.1 cGy for cervical vertebra. The results of paired t-test were 0.24 for comparing TLD dosimetry and MIRD calculation, 0.80 for MCNP simulation and MIRD, and 0.19 for TLD and MCNP. The results showed no significant differences among three methods of Monte Carlo simulations, MIRD calculation and direct experimental dosimetry using TLD. PMID:23717806

New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

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

Brown, Forrest B.

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

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.

A comparison between EGS4 and MCNP computer modeling of an in vivo X-ray fluorescence system.

PubMed

Al-Ghorabie, F H; Natto, S S; Al-Lyhiani, S H

2001-03-01

The Monte Carlo computer codes EGS4 and MCNP were used to develop a theoretical model of a 180 degrees geometry in vivo X-ray fluorescence system for the measurement of platinum concentration in head and neck tumors. The model included specification of the photon source, collimators, phantoms and detector. Theoretical results were compared and evaluated against X-ray fluorescence data obtained experimentally from an existing system developed by the Swansea In Vivo Analysis and Cancer Research Group. The EGS4 results agreed well with the MCNP results. However, agreement between the measured spectral shape obtained using the experimental X-ray fluorescence system and the simulated spectral shape obtained using the two Monte Carlo codes was relatively poor. The main reason for the disagreement between the results arises from the basic assumptions which the two codes used in their calculations. Both codes assume a "free" electron model for Compton interactions. This assumption will underestimate the results and invalidates any predicted and experimental spectra when compared with each other.

Monte Carlo MCNP-4B-based absorbed dose distribution estimates for patient-specific dosimetry.

PubMed

Yoriyaz, H; Stabin, M G; dos Santos, A

2001-04-01

This study was intended to verify the capability of the Monte Carlo MCNP-4B code to evaluate spatial dose distribution based on information gathered from CT or SPECT. A new three-dimensional (3D) dose calculation approach for internal emitter use in radioimmunotherapy (RIT) was developed using the Monte Carlo MCNP-4B code as the photon and electron transport engine. It was shown that the MCNP-4B computer code can be used with voxel-based anatomic and physiologic data to provide 3D dose distributions. This study showed that the MCNP-4B code can be used to develop a treatment planning system that will provide such information in a time manner, if dose reporting is suitably optimized. If each organ is divided into small regions where the average energy deposition is calculated with a typical volume of 0.4 cm(3), regional dose distributions can be provided with reasonable central processing unit times (on the order of 12-24 h on a 200-MHz personal computer or modest workstation). Further efforts to provide semiautomated region identification (segmentation) and improvement of marrow dose calculations are needed to supply a complete system for RIT. It is envisioned that all such efforts will continue to develop and that internal dose calculations may soon be brought to a similar level of accuracy, detail, and robustness as is commonly expected in external dose treatment planning. For this study we developed a code with a user-friendly interface that works on several nuclear medicine imaging platforms and provides timely patient-specific dose information to the physician and medical physicist. Future therapy with internal emitters should use a 3D dose calculation approach, which represents a significant advance over dose information provided by the standard geometric phantoms used for more than 20 y (which permit reporting of only average organ doses for certain standardized individuals)

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

Comparative Dosimetric Estimates of a 25 keV Electron Micro-beam with three Monte Carlo Codes

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

Mainardi, Enrico; Donahue, Richard J.; Blakely, Eleanor A.

2002-09-11

The calculations presented compare the different performances of the three Monte Carlo codes PENELOPE-1999, MCNP-4C and PITS, for the evaluation of Dose profiles from a 25 keV electron micro-beam traversing individual cells. The overall model of a cell is a water cylinder equivalent for the three codes but with a different internal scoring geometry: hollow cylinders for PENELOPE and MCNP, whereas spheres are used for the PITS code. A cylindrical cell geometry with scoring volumes with the shape of hollow cylinders was initially selected for PENELOPE and MCNP because of its superior simulation of the actual shape and dimensions ofmore » a cell and for its improved computer-time efficiency if compared to spherical internal volumes. Some of the transfer points and energy transfer that constitute a radiation track may actually fall in the space between spheres, that would be outside the spherical scoring volume. This internal geometry, along with the PENELOPE algorithm, drastically reduced the computer time when using this code if comparing with event-by-event Monte Carlo codes like PITS. This preliminary work has been important to address dosimetric estimates at low electron energies. It demonstrates that codes like PENELOPE can be used for Dose evaluation, even with such small geometries and energies involved, which are far below the normal use for which the code was created. Further work (initiated in Summer 2002) is still needed however, to create a user-code for PENELOPE that allows uniform comparison of exact cell geometries, integral volumes and also microdosimetric scoring quantities, a field where track-structure codes like PITS, written for this purpose, are believed to be superior.« less

Comparison study of photon attenuation characteristics of Lead-Boron Polyethylene by MCNP code, XCOM and experimental data

NASA Astrophysics Data System (ADS)

Zhang, Lei; Jia, Mingchun; Gong, Junjun; Xia, Wenming

2017-08-01

The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.

Determining the mass attenuation coefficient, effective atomic number, and electron density of raw wood and binderless particleboards of Rhizophora spp. by using Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Marashdeh, Mohammad W.; Al-Hamarneh, Ibrahim F.; Abdel Munem, Eid M.; Tajuddin, A. A.; Ariffin, Alawiah; Al-Omari, Saleh

Rhizophora spp. wood has the potential to serve as a solid water or tissue equivalent phantom for photon and electron beam dosimetry. In this study, the effective atomic number (Zeff) and effective electron density (Neff) of raw wood and binderless Rhizophora spp. particleboards in four different particle sizes were determined in the 10-60 keV energy region. The mass attenuation coefficients used in the calculations were obtained using the Monte Carlo N-Particle (MCNP5) simulation code. The MCNP5 calculations of the attenuation parameters for the Rhizophora spp. samples were plotted graphically against photon energy and discussed in terms of their relative differences compared with those of water and breast tissue. Moreover, the validity of the MCNP5 code was examined by comparing the calculated attenuation parameters with the theoretical values obtained by the XCOM program based on the mixture rule. The results indicated that the MCNP5 process can be followed to determine the attenuation of gamma rays with several photon energies in other materials.

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

PubMed

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

Benchmarking the MCNP code for Monte Carlo modelling of an in vivo neutron activation analysis system.

PubMed

Natto, S A; Lewis, D G; Ryde, S J

1998-01-01

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

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.

Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: a case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital.

PubMed

Sheu, R J; Sheu, R D; Jiang, S H; Kao, C H

2005-01-01

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted.

Methodology of full-core Monte Carlo calculations with leakage parameter evaluations for benchmark critical experiment analysis

NASA Astrophysics Data System (ADS)

Sboev, A. G.; Ilyashenko, A. S.; Vetrova, O. A.

1997-02-01

The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of Keff value.

Recent advances and future prospects for Monte Carlo

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

Brown, Forrest B

2010-01-01

The history of Monte Carlo methods is closely linked to that of computers: The first known Monte Carlo program was written in 1947 for the ENIAC; a pre-release of the first Fortran compiler was used for Monte Carlo In 1957; Monte Carlo codes were adapted to vector computers in the 1980s, clusters and parallel computers in the 1990s, and teraflop systems in the 2000s. Recent advances include hierarchical parallelism, combining threaded calculations on multicore processors with message-passing among different nodes. With the advances In computmg, Monte Carlo codes have evolved with new capabilities and new ways of use. Production codesmore » such as MCNP, MVP, MONK, TRIPOLI and SCALE are now 20-30 years old (or more) and are very rich in advanced featUres. The former 'method of last resort' has now become the first choice for many applications. Calculations are now routinely performed on office computers, not just on supercomputers. Current research and development efforts are investigating the use of Monte Carlo methods on FPGAs. GPUs, and many-core processors. Other far-reaching research is exploring ways to adapt Monte Carlo methods to future exaflop systems that may have 1M or more concurrent computational processes.« less

Monte Carlo method for calculating the radiation skyshine produced by electron accelerators

NASA Astrophysics Data System (ADS)

Kong, Chaocheng; Li, Quanfeng; Chen, Huaibi; Du, Taibin; Cheng, Cheng; Tang, Chuanxiang; Zhu, Li; Zhang, Hui; Pei, Zhigang; Ming, Shenjin

2005-06-01

Using the MCNP4C Monte Carlo code, the X-ray skyshine produced by 9 MeV, 15 MeV and 21 MeV electron linear accelerators were calculated respectively with a new two-step method combined with the split and roulette variance reduction technique. Results of the Monte Carlo simulation, the empirical formulas used for skyshine calculation and the dose measurements were analyzed and compared. In conclusion, the skyshine dose measurements agreed reasonably with the results computed by the Monte Carlo method, but deviated from computational results given by empirical formulas. The effect on skyshine dose caused by different structures of accelerator head is also discussed in this paper.

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

Hiller, Mauritius M.; Veinot, Kenneth G.; Easterly, Clay E.

In this study, methods are addressed to reduce the computational time to compute organ-dose rate coefficients using Monte Carlo techniques. Several variance reduction techniques are compared including the reciprocity method, importance sampling, weight windows and the use of the ADVANTG software package. For low-energy photons, the runtime was reduced by a factor of 10 5 when using the reciprocity method for kerma computation for immersion of a phantom in contaminated water. This is particularly significant since impractically long simulation times are required to achieve reasonable statistical uncertainties in organ dose for low-energy photons in this source medium and geometry. Althoughmore » the MCNP Monte Carlo code is used in this paper, the reciprocity technique can be used equally well with other Monte Carlo codes.« less

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.

Comparison of scientific computing platforms for MCNP4A Monte Carlo calculations

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

Hendricks, J.S.; Brockhoff, R.C.

1994-04-01

The performance of seven computer platforms is evaluated with the widely used and internationally available MCNP4A Monte Carlo radiation transport code. All results are reproducible and are presented in such a way as to enable comparison with computer platforms not in the study. The authors observed that the HP/9000-735 workstation runs MCNP 50% faster than the Cray YMP 8/64. Compared with the Cray YMP 8/64, the IBM RS/6000-560 is 68% as fast, the Sun Sparc10 is 66% as fast, the Silicon Graphics ONYX is 90% as fast, the Gateway 2000 model 4DX2-66V personal computer is 27% as fast, and themore » Sun Sparc2 is 24% as fast. In addition to comparing the timing performance of the seven platforms, the authors observe that changes in compilers and software over the past 2 yr have resulted in only modest performance improvements, hardware improvements have enhanced performance by less than a factor of [approximately]3, timing studies are very problem dependent, MCNP4Q runs about as fast as MCNP4.« less

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

Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry.

PubMed

Sohrabpour, M; Hassanzadeh, M; Shahriari, M; Sharifzadeh, M

2002-10-01

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators.

Benchmark study for total enery electrons in thick slabs

NASA Technical Reports Server (NTRS)

Jun, I.

2002-01-01

The total energy deposition profiles when highenergy electrons impinge on a thick slab of elemental aluminum, copper, and tungsten have been computed using representative Monte Carlo codes (NOVICE, TIGER, MCNP), and compared in this paper.

Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. RAdiation Detection Scenario Analysis Toolbox (RADSAT), a code which couples deterministic and Monte Carlo transport to perform radiation detection scenario analysis in three dimensions [1], was used as the building block for the methods derived in this work. RADSAT was capable of performing coupled deterministic-Monte Carlo simulations for gamma-only and neutron-only problems. The purpose of this work was to develop the methodology necessary to perform coupled neutron-photon calculations and add this capability to RADSAT. Performing coupled neutron-photon calculations requires four main steps: the deterministic neutron transport calculation, the neutron-induced photon spectrum calculation, the deterministic photon transport calculation, and the Monte Carlo detector response calculation. The necessary requirements for each of these steps were determined. A major challenge in utilizing multigroup deterministic transport methods for neutron-photon problems was maintaining the discrete neutron-induced photon signatures throughout the simulation. Existing coupled neutron-photon cross-section libraries and the methods used to produce neutron-induced photons were unsuitable for high-resolution gamma-ray spectroscopy applications. Central to this work was the development of a method for generating multigroup neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so the neutron-induced photon signatures were preserved. The RADSAT-NG cross-section library was developed as a specialized multigroup neutron-photon cross-section set for the simulation of high-resolution gamma-ray spectroscopy applications. The methodology and cross sections were tested using code-to-code comparison with MCNP5 [2] and NJOY [3]. A simple benchmark geometry was used for all cases compared with MCNP. The geometry consists of a cubical sample with a 252Cf neutron source on one side and a HPGe gamma-ray spectrometer on the opposing side. Different materials were examined in the cubical sample: polyethylene (C2H4), P, N, O, and Fe. The cross sections for each of the materials were compared to cross sections collapsed using NJOY. Comparisons of the volume-averaged neutron flux within the sample, volume-averaged photon flux within the detector, and high-purity gamma-ray spectrometer response (only for polyethylene) were completed using RADSAT and MCNP. The code-to-code comparisons show promising results for the coupled Monte Carlo-deterministic method. The RADSAT-NG cross-section production method showed good agreement with NJOY for all materials considered although some additional work is needed in the resonance region and in the first and last energy bin. Some cross section discrepancies existed in the lowest and highest energy bin, but the overall shape and magnitude of the two methods agreed. For the volume-averaged photon flux within the detector, typically the five most intense lines agree to within approximately 5% of the MCNP calculated flux for all of materials considered. The agreement in the code-to-code comparisons cases demonstrates a proof-of-concept of the method for use in RADSAT for coupled neutron-photon problems in high-resolution gamma-ray spectroscopy applications. One of the primary motivators for using the coupled method over pure Monte Carlo method is the potential for significantly lower computational times. For the code-to-code comparison cases, the run times for RADSAT were approximately 25--500 times shorter than for MCNP, as shown in Table 1. This was assuming a 40 mCi 252Cf neutron source and 600 seconds of "real-world" measurement time. The only variance reduction technique implemented in the MCNP calculation was forward biasing of the source toward the sample target. Improved MCNP runtimes could be achieved with the addition of more advanced variance reduction techniques.

Benchmark of PENELOPE code for low-energy photon transport: dose comparisons with MCNP4 and EGS4.

PubMed

Ye, Sung-Joon; Brezovich, Ivan A; Pareek, Prem; Naqvi, Shahid A

2004-02-07

The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

Reducing statistical uncertainties in simulated organ doses of phantoms immersed in water

DOE PAGES

Hiller, Mauritius M.; Veinot, Kenneth G.; Easterly, Clay E.; ...

2016-08-13

In this study, methods are addressed to reduce the computational time to compute organ-dose rate coefficients using Monte Carlo techniques. Several variance reduction techniques are compared including the reciprocity method, importance sampling, weight windows and the use of the ADVANTG software package. For low-energy photons, the runtime was reduced by a factor of 10 5 when using the reciprocity method for kerma computation for immersion of a phantom in contaminated water. This is particularly significant since impractically long simulation times are required to achieve reasonable statistical uncertainties in organ dose for low-energy photons in this source medium and geometry. Althoughmore » the MCNP Monte Carlo code is used in this paper, the reciprocity technique can be used equally well with other Monte Carlo codes.« less

Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

PubMed

Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

2009-11-01

Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

Using the Monte Carlo method for assessing the tissue and organ doses of patients in dental radiography

NASA Astrophysics Data System (ADS)

Makarevich, K. O.; Minenko, V. F.; Verenich, K. A.; Kuten, S. A.

2016-05-01

This work is dedicated to modeling dental radiographic examinations to assess the absorbed doses of patients and effective doses. For simulating X-ray spectra, the TASMIP empirical model is used. Doses are assessed on the basis of the Monte Carlo method by using MCNP code for voxel phantoms of ICRP. The results of the assessment of doses to individual organs and effective doses for different types of dental examinations and features of X-ray tube are presented.

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

Monte Carlo simulation of x-ray spectra in diagnostic radiology and mammography using MCNP4C

NASA Astrophysics Data System (ADS)

Ay, M. R.; Shahriari, M.; Sarkar, S.; Adib, M.; Zaidi, H.

2004-11-01

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

DOE PAGES

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola; ...

2017-05-01

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

Performance and accuracy of criticality calculations performed using WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

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

Bergmann, Ryan M.; Rowland, Kelly L.; Radnović, Nikola

In this companion paper to "Algorithmic Choices in WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs" (doi:10.1016/j.anucene.2014.10.039), the WARP Monte Carlo neutron transport framework for graphics processing units (GPUs) is benchmarked against production-level central processing unit (CPU) Monte Carlo neutron transport codes for both performance and accuracy. We compare neutron flux spectra, multiplication factors, runtimes, speedup factors, and costs of various GPU and CPU platforms running either WARP, Serpent 2.1.24, or MCNP 6.1. WARP compares well with the results of the production-level codes, and it is shown that on the newestmore » hardware considered, GPU platforms running WARP are between 0.8 to 7.6 times as fast as CPU platforms running production codes. Also, the GPU platforms running WARP were between 15% and 50% as expensive to purchase and between 80% to 90% as expensive to operate as equivalent CPU platforms performing at an equal simulation rate.« less

TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking

NASA Astrophysics Data System (ADS)

Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.

2014-06-01

The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

NASA Astrophysics Data System (ADS)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Geometry creation for MCNP by Sabrina and XSM

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

Van Riper, K.A.

The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSMmore » as of January 1, 1994.« less

Implementation and testing of the on-the-fly thermal scattering Monte Carlo sampling method for graphite and light water in MCNP6

DOE PAGES

Pavlou, Andrew T.; Ji, Wei; Brown, Forrest B.

2016-01-23

Here, a proper treatment of thermal neutron scattering requires accounting for chemical binding through a scattering law S(α,β,T). Monte Carlo codes sample the secondary neutron energy and angle after a thermal scattering event from probability tables generated from S(α,β,T) tables at discrete temperatures, requiring a large amount of data for multiscale and multiphysics problems with detailed temperature gradients. We have previously developed a method to handle this temperature dependence on-the-fly during the Monte Carlo random walk using polynomial expansions in 1/T to directly sample the secondary energy and angle. In this paper, the on-the-fly method is implemented into MCNP6 andmore » tested in both graphite-moderated and light water-moderated systems. The on-the-fly method is compared with the thermal ACE libraries that come standard with MCNP6, yielding good agreement with integral reactor quantities like k-eigenvalue and differential quantities like single-scatter secondary energy and angle distributions. The simulation runtimes are comparable between the two methods (on the order of 5–15% difference for the problems tested) and the on-the-fly fit coefficients only require 5–15 MB of total data storage.« less

On the effect of updated MCNP photon cross section data on the simulated response of the HPA TLD.

PubMed

Eakins, Jonathan

2009-02-01

The relative response of the new Health Protection Agency thermoluminescence dosimeter (TLD) has been calculated for Narrow Series X-ray distribution and (137)Cs photon sources using the Monte Carlo code MCNP5, and the results compared with those obtained during its design stage using the predecessor code, MCNP4c2. The results agreed at intermediate energies (approximately 0.1 MeV to (137)Cs), but differed at low energies (<0.1 MeV) by up to approximately 10%. This disparity has been ascribed to differences in the default photon interaction data used by the two codes, and derives ultimately from the effect on absorbed dose of the recent updates to the photoelectric cross sections. The sources of these data have been reviewed.

Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6 [Production of energetic light fragments in CEM, LAQGSM, and MCNP6

DOE PAGES

Mashnik, Stepan Georgievich; Kerby, Leslie Marie; Gudima, Konstantin K.; ...

2017-03-23

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6 [Production of energetic light fragments in CEM, LAQGSM, and MCNP6

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

Mashnik, Stepan Georgievich; Kerby, Leslie Marie; Gudima, Konstantin K.

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

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

White, Morgan C.

2000-07-01

The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a selectmore » group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to calculate radiation dose due to the neutron environment around a MEA is shown. An uncertainty of a factor of three in the MEA calculations is shown to be due to uncertainties in the geometry modeling. It is believed that the methodology is sound and that good agreement between simulation and experiment has been demonstrated.« less

SU-F-T-111: Investigation of the Attila Deterministic Solver as a Supplement to Monte Carlo for Calculating Out-Of-Field Radiotherapy Dose

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

Mille, M; Lee, C; Failla, G

Purpose: To use the Attila deterministic solver as a supplement to Monte Carlo for calculating out-of-field organ dose in support of epidemiological studies looking at the risks of second cancers. Supplemental dosimetry tools are needed to speed up dose calculations for studies involving large-scale patient cohorts. Methods: Attila is a multi-group discrete ordinates code which can solve the 3D photon-electron coupled linear Boltzmann radiation transport equation on a finite-element mesh. Dose is computed by multiplying the calculated particle flux in each mesh element by a medium-specific energy deposition cross-section. The out-of-field dosimetry capability of Attila is investigated by comparing averagemore » organ dose to that which is calculated by Monte Carlo simulation. The test scenario consists of a 6 MV external beam treatment of a female patient with a tumor in the left breast. The patient is simulated by a whole-body adult reference female computational phantom. Monte Carlo simulations were performed using MCNP6 and XVMC. Attila can export a tetrahedral mesh for MCNP6, allowing for a direct comparison between the two codes. The Attila and Monte Carlo methods were also compared in terms of calculation speed and complexity of simulation setup. A key perquisite for this work was the modeling of a Varian Clinac 2100 linear accelerator. Results: The solid mesh of the torso part of the adult female phantom for the Attila calculation was prepared using the CAD software SpaceClaim. Preliminary calculations suggest that Attila is a user-friendly software which shows great promise for our intended application. Computational performance is related to the number of tetrahedral elements included in the Attila calculation. Conclusion: Attila is being explored as a supplement to the conventional Monte Carlo radiation transport approach for performing retrospective patient dosimetry. The goal is for the dosimetry to be sufficiently accurate for use in retrospective epidemiological investigations.« less

MCNP/X TRANSPORT IN THE TABULAR REGIME

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

HUGHES, H. GRADY

2007-01-08

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

MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides.

PubMed

Hendriks, P H G M; Maucec, M; de Meijer, R J

2002-09-01

gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of 40K and the series of 232Th and 238U are used to describe the source. A procedure is proposed which excludes the time-consuming electron tracking in less relevant areas of the geometry. The simulated gamma-ray spectra are benchmarked against laboratory data.

Development of a new multi-modal Monte-Carlo radiotherapy planning system.

PubMed

Kumada, H; Nakamura, T; Komeda, M; Matsumura, A

2009-07-01

A new multi-modal Monte-Carlo radiotherapy planning system (developing code: JCDS-FX) is under development at Japan Atomic Energy Agency. This system builds on fundamental technologies of JCDS applied to actual boron neutron capture therapy (BNCT) trials in JRR-4. One of features of the JCDS-FX is that PHITS has been applied to particle transport calculation. PHITS is a multi-purpose particle Monte-Carlo transport code. Hence application of PHITS enables to evaluate total doses given to a patient by a combined modality therapy. Moreover, JCDS-FX with PHITS can be used for the study of accelerator based BNCT. To verify calculation accuracy of the JCDS-FX, dose evaluations for neutron irradiation of a cylindrical water phantom and for an actual clinical trial were performed, then the results were compared with calculations by JCDS with MCNP. The verification results demonstrated that JCDS-FX is applicable to BNCT treatment planning in practical use.

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

DOE PAGES

Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; ...

2015-12-21

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Somemore » specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.« less

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

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

Pecchia, M.; D'Auria, F.; Mazzantini, O.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

Calculation of the effective dose from natural radioactivity in soil using MCNP code.

PubMed

Krstic, D; Nikezic, D

2010-01-01

Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. Copyright 2009 Elsevier Ltd. All rights reserved.

Monte Carlo simulations and benchmark measurements on the response of TE(TE) and Mg(Ar) ionization chambers in photon, electron and neutron beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Huang, Tseng-Te; Liu, Yuan-Hao; Chen, Wei-Lin; Chen, Yen-Fu; Wu, Shu-Wei; Nievaart, Sander; Jiang, Shiang-Huei

2015-06-01

The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7.8-16.5% below 120 kVp X-ray beams. In this study, we were especially interested in BNCT doses where low energy photon contribution is less to ignore, MCNP model is recognized as the most suitable to simulate wide photon-electron and neutron energy distributed responses of the paired ICs. Also, MCNP provides the best prediction of BNCT source adjustment by the detector's neutron and photon responses.

Accelerating Pseudo-Random Number Generator for MCNP on GPU

NASA Astrophysics Data System (ADS)

Gong, Chunye; Liu, Jie; Chi, Lihua; Hu, Qingfeng; Deng, Li; Gong, Zhenghu

2010-09-01

Pseudo-random number generators (PRNG) are intensively used in many stochastic algorithms in particle simulations, artificial neural networks and other scientific computation. The PRNG in Monte Carlo N-Particle Transport Code (MCNP) requires long period, high quality, flexible jump and fast enough. In this paper, we implement such a PRNG for MCNP on NVIDIA's GTX200 Graphics Processor Units (GPU) using CUDA programming model. Results shows that 3.80 to 8.10 times speedup are achieved compared with 4 to 6 cores CPUs and more than 679.18 million double precision random numbers can be generated per second on GPU.

An Electron/Photon/Relaxation Data Library for MCNP6

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

Hughes, III, H. Grady

The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

Monte Carlo source simulation technique for solution of interference reactions in INAA experiments: a preliminary report

NASA Astrophysics Data System (ADS)

Allaf, M. Athari; Shahriari, M.; Sohrabpour, M.

2004-04-01

A new method using Monte Carlo source simulation of interference reactions in neutron activation analysis experiments has been developed. The neutron spectrum at the sample location has been simulated using the Monte Carlo code MCNP and the contributions of different elements to produce a specified gamma line have been determined. The produced response matrix has been used to measure peak areas and the sample masses of the elements of interest. A number of benchmark experiments have been performed and the calculated results verified against known values. The good agreement obtained between the calculated and known values suggests that this technique may be useful for the elimination of interference reactions in neutron activation analysis.

Collision of Physics and Software in the Monte Carlo Application Toolkit (MCATK)

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

Sweezy, Jeremy Ed

2016-01-21

The topic is presented in a series of slides organized as follows: MCATK overview, development strategy, available algorithms, problem modeling (sources, geometry, data, tallies), parallelism, miscellaneous tools/features, example MCATK application, recent areas of research, and summary and future work. MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library with continuous energy neutron and photon transport. Designed to build specialized applications and to provide new functionality in existing general-purpose Monte Carlo codes like MCNP, it reads ACE formatted nuclear data generated by NJOY. The motivation behind MCATK was to reduce costs. MCATK physics involves continuous energy neutron & gammamore » transport with multi-temperature treatment, static eigenvalue (k eff and α) algorithms, time-dependent algorithm, and fission chain algorithms. MCATK geometry includes mesh geometries and solid body geometries. MCATK provides verified, unit-test Monte Carlo components, flexibility in Monte Carlo application development, and numerous tools such as geometry and cross section plotters.« less

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part II: gadolinium neutron capture therapy models and therapeutic effects.

PubMed

Wangerin, K; Culbertson, C N; 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 gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

PubMed

Daures, J; Gouriou, J; Bordy, J M

2011-03-01

This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

WE-DE-201-05: Evaluation of a Windowless Extrapolation Chamber Design and Monte Carlo Based Corrections for the Calibration of Ophthalmic Applicators

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

Hansen, J; Culberson, W; DeWerd, L

Purpose: To test the validity of a windowless extrapolation chamber used to measure surface dose rate from planar ophthalmic applicators and to compare different Monte Carlo based codes for deriving correction factors. Methods: Dose rate measurements were performed using a windowless, planar extrapolation chamber with a {sup 90}Sr/{sup 90}Y Tracerlab RA-1 ophthalmic applicator previously calibrated at the National Institute of Standards and Technology (NIST). Capacitance measurements were performed to estimate the initial air gap width between the source face and collecting electrode. Current was measured as a function of air gap, and Bragg-Gray cavity theory was used to calculate themore » absorbed dose rate to water. To determine correction factors for backscatter, divergence, and attenuation from the Mylar entrance window found in the NIST extrapolation chamber, both EGSnrc Monte Carlo user code and Monte Carlo N-Particle Transport Code (MCNP) were utilized. Simulation results were compared with experimental current readings from the windowless extrapolation chamber as a function of air gap. Additionally, measured dose rate values were compared with the expected result from the NIST source calibration to test the validity of the windowless chamber design. Results: Better agreement was seen between EGSnrc simulated dose results and experimental current readings at very small air gaps (<100 µm) for the windowless extrapolation chamber, while MCNP results demonstrated divergence at these small gap widths. Three separate dose rate measurements were performed with the RA-1 applicator. The average observed difference from the expected result based on the NIST calibration was −1.88% with a statistical standard deviation of 0.39% (k=1). Conclusion: EGSnrc user code will be used during future work to derive correction factors for extrapolation chamber measurements. Additionally, experiment results suggest that an entrance window is not needed in order for an extrapolation chamber to provide accurate dose rate measurements for a planar ophthalmic applicator.« less

SABRINA - An interactive geometry modeler for MCNP (Monte Carlo Neutron Photon)

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

West, J.T.; Murphy, J.

SABRINA is an interactive three-dimensional geometry modeler developed to produce complicated models for the Los Alamos Monte Carlo Neutron Photon program MCNP. SABRINA produces line drawings and color-shaded drawings for a wide variety of interactive graphics terminals. It is used as a geometry preprocessor in model development and as a Monte Carlo particle-track postprocessor in the visualization of complicated particle transport problem. SABRINA is written in Fortran 77 and is based on the Los Alamos Common Graphics System, CGS. 5 refs., 2 figs.

Monte Carlo dose calculations in homogeneous media and at interfaces: a comparison between GEPTS, EGSnrc, MCNP, and measurements.

PubMed

Chibani, Omar; Li, X Allen

2002-05-01

Three Monte Carlo photon/electron transport codes (GEPTS, EGSnrc, and MCNP) are bench-marked against dose measurements in homogeneous (both low- and high-Z) media as well as at interfaces. A brief overview on physical models used by each code for photon and electron (positron) transport is given. Absolute calorimetric dose measurements for 0.5 and 1 MeV electron beams incident on homogeneous and multilayer media are compared with the predictions of the three codes. Comparison with dose measurements in two-layer media exposed to a 60Co gamma source is also performed. In addition, comparisons between the codes (including the EGS4 code) are done for (a) 0.05 to 10 MeV electron beams and positron point sources in lead, (b) high-energy photons (10 and 20 MeV) irradiating a multilayer phantom (water/steel/air), and (c) simulation of a 90Sr/90Y brachytherapy source. A good agreement is observed between the calorimetric electron dose measurements and predictions of GEPTS and EGSnrc in both homogeneous and multilayer media. MCNP outputs are found to be dependent on the energy-indexing method (Default/ITS style). This dependence is significant in homogeneous media as well as at interfaces. MCNP(ITS) fits more closely the experimental data than MCNP(DEF), except for the case of Be. At low energy (0.05 and 0.1 MeV), MCNP(ITS) dose distributions in lead show higher maximums in comparison with GEPTS and EGSnrc. EGS4 produces too penetrating electron-dose distributions in high-Z media, especially at low energy (<0.1 MeV). For positrons, differences between GEPTS and EGSnrc are observed in lead because GEPTS distinguishes positrons from electrons for both elastic multiple scattering and bremsstrahlung emission models. For the 60Co source, a quite good agreement between calculations and measurements is observed with regards to the experimental uncertainty. For the other cases (10 and 20 MeV photon sources and the 90Sr/90Y beta source), a good agreement is found between the three codes. In conclusion, differences between GEPTS and EGSnrc results are found to be very small for almost all media and energies studied. MCNP results depend significantly on the electron energy-indexing method.

Monte Carlo simulations for angular and spatial distributions in therapeutic-energy proton beams

NASA Astrophysics Data System (ADS)

Lin, Yi-Chun; Pan, C. Y.; Chiang, K. J.; Yuan, M. C.; Chu, C. H.; Tsai, Y. W.; Teng, P. K.; Lin, C. H.; Chao, T. C.; Lee, C. C.; Tung, C. J.; Chen, A. E.

2017-11-01

The purpose of this study is to compare the angular and spatial distributions of therapeutic-energy proton beams obtained from the FLUKA, GEANT4 and MCNP6 Monte Carlo codes. The Monte Carlo simulations of proton beams passing through two thin targets and a water phantom were investigated to compare the primary and secondary proton fluence distributions and dosimetric differences among these codes. The angular fluence distributions, central axis depth-dose profiles, and lateral distributions of the Bragg peak cross-field were calculated to compare the proton angular and spatial distributions and energy deposition. Benchmark verifications from three different Monte Carlo simulations could be used to evaluate the residual proton fluence for the mean range and to estimate the depth and lateral dose distributions and the characteristic depths and lengths along the central axis as the physical indices corresponding to the evaluation of treatment effectiveness. The results showed a general agreement among codes, except that some deviations were found in the penumbra region. These calculated results are also particularly helpful for understanding primary and secondary proton components for stray radiation calculation and reference proton standard determination, as well as for determining lateral dose distribution performance in proton small-field dosimetry. By demonstrating these calculations, this work could serve as a guide to the recent field of Monte Carlo methods for therapeutic-energy protons.

Calibration with MCNP of NaI detector for the determination of natural radioactivity levels in the field.

PubMed

Cinelli, Giorgia; Tositti, Laura; Mostacci, Domiziano; Baré, Jonathan

2016-05-01

In view of assessing natural radioactivity with on-site quantitative gamma spectrometry, efficiency calibration of NaI(Tl) detectors is investigated. A calibration based on Monte Carlo simulation of detector response is proposed, to render reliable quantitative analysis practicable in field campaigns. The method is developed with reference to contact geometry, in which measurements are taken placing the NaI(Tl) probe directly against the solid source to be analyzed. The Monte Carlo code used for the simulations was MCNP. Experimental verification of the calibration goodness is obtained by comparison with appropriate standards, as reported. On-site measurements yield a quick quantitative assessment of natural radioactivity levels present ((40)K, (238)U and (232)Th). On-site gamma spectrometry can prove particularly useful insofar as it provides information on materials from which samples cannot be taken. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

MCNP modelling of vaginal and uterine applicators used in intracavitary brachytherapy and comparison with radiochromic film measurements

NASA Astrophysics Data System (ADS)

Ceccolini, E.; Gerardy, I.; Ródenas, J.; van Dycke, M.; Gallardo, S.; Mostacci, D.

Brachytherapy is an advanced cancer treatment that is minimally invasive, minimising radiation exposure to the surrounding healthy tissues. Microselectron© Nucletron devices with 192Ir source can be used for gynaecological brachytherapy, in patients with vaginal or uterine cancer. Measurements of isodose curves have been performed in a PMMA phantom and compared with Monte Carlo calculations and TPS (Plato software of Nucletron BPS 14.2) evaluation. The isodose measurements have been performed with radiochromic films (Gafchromic EBT©). The dose matrix has been obtained after digitalisation and use of a dose calibration curve obtained with a 6 MV photon beam provided by a medical linear accelerator. A comparison between the calculated and the measured matrix has been performed. The calculated dose matrix is obtained with a simulation using the MCNP5 Monte Carlo code (F4MESH tally).

Brachytherapy dosimetry of 125I and 103Pd sources using an updated cross section library for the MCNP Monte Carlo transport code.

PubMed

Bohm, Tim D; DeLuca, Paul M; DeWerd, Larry A

2003-04-01

Permanent implantation of low energy (20-40 keV) photon emitting radioactive seeds to treat prostate cancer is an important treatment option for patients. In order to produce accurate implant brachytherapy treatment plans, the dosimetry of a single source must be well characterized. Monte Carlo based transport calculations can be used for source characterization, but must have up to date cross section libraries to produce accurate dosimetry results. This work benchmarks the MCNP code and its photon cross section library for low energy photon brachytherapy applications. In particular, we calculate the emitted photon spectrum, air kerma, depth dose in water, and radial dose function for both 125I and 103Pd based seeds and compare to other published results. Our results show that MCNP's cross section library differs from recent data primarily in the photoelectric cross section for low energies and low atomic number materials. In water, differences as large as 10% in the photoelectric cross section and 6% in the total cross section occur at 125I and 103Pd photon energies. This leads to differences in the dose rate constant of 3% and 5%, and differences as large as 18% and 20% in the radial dose function for the 125I and 103Pd based seeds, respectively. Using a partially updated photon library, calculations of the dose rate constant and radial dose function agree with other published results. Further, the use of the updated photon library allows us to verify air kerma and depth dose in water calculations performed using MCNP's perturbation feature to simulate updated cross sections. We conclude that in order to most effectively use MCNP for low energy photon brachytherapy applications, we must update its cross section library. Following this update, the MCNP code system will be a very effective tool for low energy photon brachytherapy dosimetry applications.

An approach to design a 90Sr radioisotope thermoelectric generator using analytical and Monte Carlo methods with ANSYS, COMSOL, and MCNP.

PubMed

Khajepour, Abolhasan; Rahmani, Faezeh

2017-01-01

In this study, a 90 Sr radioisotope thermoelectric generator (RTG) with power of milliWatt was designed to operate in the determined temperature (300-312K). For this purpose, the combination of analytical and Monte Carlo methods with ANSYS and COMSOL software as well as the MCNP code was used. This designed RTG contains 90 Sr as a radioisotope heat source (RHS) and 127 coupled thermoelectric modules (TEMs) based on bismuth telluride. Kapton (2.45mm in thickness) and Cryotherm sheets (0.78mm in thickness) were selected as the thermal insulators of the RHS, as well as a stainless steel container was used as a generator chamber. The initial design of the RHS geometry was performed according to the amount of radioactive material (strontium titanate) as well as the heat transfer calculations and mechanical strength considerations. According to the Monte Carlo simulation performed by the MCNP code, approximately 0.35 kCi of 90 Sr is sufficient to generate heat power in the RHS. To determine the optimal design of the RTG, the distribution of temperature as well as the dissipated heat and input power to the module were calculated in different parts of the generator using the ANSYS software. Output voltage according to temperature distribution on TEM was calculated using COMSOL. Optimization of the dimension of the RHS and heat insulator was performed to adapt the average temperature of the hot plate of TEM to the determined hot temperature value. This designed RTG generates 8mW in power with an efficiency of 1%. This proposed approach of combination method can be used for the precise design of various types of RTGs. Copyright © 2016 Elsevier Ltd. All rights reserved.

The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

NASA Astrophysics Data System (ADS)

Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

2018-05-01

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

Shielding analysis of the Microtron MT-25 bunker using the MCNP-4C code and NCRP Report 51.

PubMed

Casanova, A O; López, N; Gelen, A; Guevara, M V Manso; Díaz, O; Cimino, L; D'Alessandro, K; Melo, J C

2004-01-01

A cyclic electron accelerator Microtron MT-25 will be installed in Havana, Cuba. Electrons, neutrons and gamma radiation up to 25 MeV can be produced in the MT-25. A detailed shielding analysis for the bunker is carried out using two ways: the NCRP-51 Report and the Monte Carlo Method (MCNP-4C Code). The walls and ceiling thicknesses are estimated with dose constraints of 0.5 and 20 mSv y(-1), respectively, and an area occupancy factor of 1/16. Both results are compared and a preliminary bunker design is shown. Copyright 2004 Oxford University Press

LANL Summer 2016 Report

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

Mendoza, Paul Michael

The Monte Carlo N-Particle (MCNP) transport code developed at Los Alamos National Laboratory (LANL) utilizes nuclear cross-section data in a compact ENDF (ACE) format. The accuracy of MCNP calculations depends on the accuracy of nuclear ACE data tables, which depends on the accuracy of the original ENDF files. There are some noticeable differences in ENDF files from one generation to the next, even among the more common fissile materials. As the next generation of ENDF files is being prepared, several software tools were developed to simulate a large number of benchmarks in MCNP (over 1000), collect data from these simulations,more » and visually represent the results.« less

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

PubMed

El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

2008-09-01

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Comparison of UWCC MOX fuel measurements to MCNP-REN calculations

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

Abhold, M.; Baker, M.; Jie, R.

1998-12-31

The development of neutron coincidence counting has greatly improved the accuracy and versatility of neutron-based techniques to assay fissile materials. Today, the shift register analyzer connected to either a passive or active neutron detector is widely used by both domestic and international safeguards organizations. The continued development of these techniques and detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model, as it is currently used, fails to accurately predict detector response in highly multiplying mediums such as mixed-oxide (MOX) lightmore » water reactor fuel assemblies. For this reason, efforts have been made to modify the currently used Monte Carlo codes and to develop new analytical methods so that this model is not required to predict detector response. The authors describe their efforts to modify a widely used Monte Carlo code for this purpose and also compare calculational results with experimental measurements.« less

Radiation shielding evaluation of the BNCT treatment room at THOR: a TORT-coupled MCNP Monte Carlo simulation study.

PubMed

Chen, A Y; Liu, Y-W H; Sheu, R J

2008-01-01

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.

Distributed multitasking ITS with PVM

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

Fan, W.C.; Halbleib, J.A. Sr.

1995-12-31

Advances in computer hardware and communication software have made it possible to perform parallel-processing computing on a collection of desktop workstations. For many applications, multitasking on a cluster of high-performance workstations has achieved performance comparable to or better than that on a traditional supercomputer. From the point of view of cost-effectiveness, it also allows users to exploit available but unused computational resources and thus achieve a higher performance-to-cost ratio. Monte Carlo calculations are inherently parallelizable because the individual particle trajectories can be generated independently with minimum need for interprocessor communication. Furthermore, the number of particle histories that can be generatedmore » in a given amount of wall-clock time is nearly proportional to the number of processors in the cluster. This is an important fact because the inherent statistical uncertainty in any Monte Carlo result decreases as the number of histories increases. For these reasons, researchers have expended considerable effort to take advantage of different parallel architectures for a variety of Monte Carlo radiation transport codes, often with excellent results. The initial interest in this work was sparked by the multitasking capability of the MCNP code on a cluster of workstations using the Parallel Virtual Machine (PVM) software. On a 16-machine IBM RS/6000 cluster, it has been demonstrated that MCNP runs ten times as fast as on a single-processor CRAY YMP. In this paper, we summarize the implementation of a similar multitasking capability for the coupled electronphoton transport code system, the Integrated TIGER Series (ITS), and the evaluation of two load-balancing schemes for homogeneous and heterogeneous networks.« less

Benchmarking the MCNP Monte Carlo code with a photon skyshine experiment

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

Olsher, R.H.; Hsu, Hsiao Hua; Harvey, W.F.

1993-07-01

The MCNP Monte Carlo transport code is used by the Los Alamos National Laboratory Health and Safety Division for a broad spectrum of radiation shielding calculations. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with the Kansas State Univ. (KSU) photon skyshine experiment of 1977. The KSU experiment for the unshielded source geometry was simulated in great detail to include the contribution of groundshine, in-silo photon scatter, and the effect of spectral degradation in the source capsule. The standard deviation of the KSUmore » experimental data was stated to be 7%, while the statistical uncertainty of the simulation was kept at or under 1%. The results of the simulation agreed closely with the experimental data, generally to within 6%. At distances of under 100 m from the silo, the modeling of the in-silo scatter was crucial to achieving close agreement with the experiment. Specifically, scatter off the top layer of the source cask accounted for [approximately]12% of the dose at 50 m. At distance >300m, using the [sup 60]Co line spectrum led to a dose overresponse as great as 19% at 700 m. It was necessary to use the actual source spectrum, which includes a Compton tail from photon collisions in the source capsule, to achieve close agreement with experimental data. These results highlight the importance of using Monte Carlo transport techniques to account for the nonideal features of even simple experiments''.« less

MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

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

Mashnik, Stepan Georgievich; Kerby, Leslie Marie

2015-05-22

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

Organ dose conversion coefficients based on a voxel mouse model and MCNP code for external photon irradiation.

PubMed

Zhang, Xiaomin; Xie, Xiangdong; Cheng, Jie; Ning, Jing; Yuan, Yong; Pan, Jie; Yang, Guoshan

2012-01-01

A set of conversion coefficients from kerma free-in-air to the organ absorbed dose for external photon beams from 10 keV to 10 MeV are presented based on a newly developed voxel mouse model, for the purpose of radiation effect evaluation. The voxel mouse model was developed from colour images of successive cryosections of a normal nude male mouse, in which 14 organs or tissues were segmented manually and filled with different colours, while each colour was tagged by a specific ID number for implementation of mouse model in Monte Carlo N-particle code (MCNP). Monte Carlo simulation with MCNP was carried out to obtain organ dose conversion coefficients for 22 external monoenergetic photon beams between 10 keV and 10 MeV under five different irradiation geometries conditions (left lateral, right lateral, dorsal-ventral, ventral-dorsal, and isotropic). Organ dose conversion coefficients were presented in tables and compared with the published data based on a rat model to investigate the effect of body size and weight on the organ dose. The calculated and comparison results show that the organ dose conversion coefficients varying the photon energy exhibits similar trend for most organs except for the bone and skin, and the organ dose is sensitive to body size and weight at a photon energy approximately <0.1 MeV.

Investigation of some possible changes in Am-Be neutron source configuration in order to increase the thermal neutron flux using Monte Carlo code

NASA Astrophysics Data System (ADS)

Basiri, H.; Tavakoli-Anbaran, H.

2018-01-01

Am-Be neutrons source is based on (α, n) reaction and generates neutrons in the energy range of 0-11 MeV. Since the thermal neutrons are widely used in different fields, in this work, we investigate how to improve the source configuration in order to increase the thermal flux. These suggested changes include a spherical moderator instead of common cylindrical geometry, a reflector layer and an appropriate materials selection in order to achieve the maximum thermal flux. All calculations were done by using MCNP1 Monte Carlo code. Our final results indicated that a spherical paraffin moderator, a layer of beryllium as a reflector can efficiently increase the thermal neutron flux of Am-Be source.

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

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

Monte Carlo simulations in Nuclear Medicine

NASA Astrophysics Data System (ADS)

Loudos, George K.

2007-11-01

Molecular imaging technologies provide unique abilities to localise signs of disease before symptoms appear, assist in drug testing, optimize and personalize therapy, and assess the efficacy of treatment regimes for different types of cancer. Monte Carlo simulation packages are used as an important tool for the optimal design of detector systems. In addition they have demonstrated potential to improve image quality and acquisition protocols. Many general purpose (MCNP, Geant4, etc) or dedicated codes (SimSET etc) have been developed aiming to provide accurate and fast results. Special emphasis will be given to GATE toolkit. The GATE code currently under development by the OpenGATE collaboration is the most accurate and promising code for performing realistic simulations. The purpose of this article is to introduce the non expert reader to the current status of MC simulations in nuclear medicine and briefly provide examples of current simulated systems, and present future challenges that include simulation of clinical studies and dosimetry applications.

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

Haghighat, A.; Sjoden, G.E.; Wagner, J.C.

In the past 10 yr, the Penn State Transport Theory Group (PSTTG) has concentrated its efforts on developing accurate and efficient particle transport codes to address increasing needs for efficient and accurate simulation of nuclear systems. The PSTTG's efforts have primarily focused on shielding applications that are generally treated using multigroup, multidimensional, discrete ordinates (S{sub n}) deterministic and/or statistical Monte Carlo methods. The difficulty with the existing public codes is that they require significant (impractical) computation time for simulation of complex three-dimensional (3-D) problems. For the S{sub n} codes, the large memory requirements are handled through the use of scratchmore » files (i.e., read-from and write-to-disk) that significantly increases the necessary execution time. Further, the lack of flexible features and/or utilities for preparing input and processing output makes these codes difficult to use. The Monte Carlo method becomes impractical because variance reduction (VR) methods have to be used, and normally determination of the necessary parameters for the VR methods is very difficult and time consuming for a complex 3-D problem. For the deterministic method, the authors have developed the 3-D parallel PENTRAN (Parallel Environment Neutral-particle TRANsport) code system that, in addition to a parallel 3-D S{sub n} solver, includes pre- and postprocessing utilities. PENTRAN provides for full phase-space decomposition, memory partitioning, and parallel input/output to provide the capability of solving large problems in a relatively short time. Besides having a modular parallel structure, PENTRAN has several unique new formulations and features that are necessary for achieving high parallel performance. For the Monte Carlo method, the major difficulty currently facing most users is the selection of an effective VR method and its associated parameters. For complex problems, generally, this process is very time consuming and may be complicated due to the possibility of biasing the results. In an attempt to eliminate this problem, the authors have developed the A{sup 3}MCNP (automated adjoint accelerated MCNP) code that automatically prepares parameters for source and transport biasing within a weight-window VR approach based on the S{sub n} adjoint function. A{sup 3}MCNP prepares the necessary input files for performing multigroup, 3-D adjoint S{sub n} calculations using TORT.« less

Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.

PubMed

Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

2017-06-01

The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.

Calculated organ doses for Mayak production association central hall using ICRP and MCNP.

PubMed

Choe, Dong-Ok; Shelkey, Brenda N; Wilde, Justin L; Walk, Heidi A; Slaughter, David M

2003-03-01

As part of an ongoing dose reconstruction project, equivalent organ dose rates from photons and neutrons were estimated using the energy spectra measured in the central hall above the graphite reactor core located in the Russian Mayak Production Association facility. Reconstruction of the work environment was necessary due to the lack of personal dosimeter data for neutrons in the time period prior to 1987. A typical worker scenario for the central hall was developed for the Monte Carlo Neutron Photon-4B (MCNP) code. The resultant equivalent dose rates for neutrons and photons were compared with the equivalent dose rates derived from calculations using the conversion coefficients in the International Commission on Radiological Protection Publications 51 and 74 in order to validate the model scenario for this Russian facility. The MCNP results were in good agreement with the results of the ICRP publications indicating the modeling scenario was consistent with actual work conditions given the spectra provided. The MCNP code will allow for additional orientations to accurately reflect source locations.

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

Assessment of background hydrogen by the Monte Carlo computer code MCNP-4A during measurements of total body nitrogen.

PubMed

Ryde, S J; al-Agel, F A; Evans, C J; Hancock, D A

2000-05-01

The use of a hydrogen internal standard to enable the estimation of absolute mass during measurement of total body nitrogen by in vivo neutron activation is an established technique. Central to the technique is a determination of the H prompt gamma ray counts arising from the subject. In practice, interference counts from other sources--e.g., neutron shielding--are included. This study reports use of the Monte Carlo computer code, MCNP-4A, to investigate the interference counts arising from shielding both with and without a phantom containing a urea solution. Over a range of phantom size (depth 5 to 30 cm, width 20 to 40 cm), the counts arising from shielding increased by between 4% and 32% compared with the counts without a phantom. For any given depth, the counts increased approximately linearly with width. For any given width, there was little increase for depths exceeding 15 centimeters. The shielding counts comprised between 15% and 26% of those arising from the urea phantom. These results, although specific to the Swansea apparatus, suggest that extraneous hydrogen counts can be considerable and depend strongly on the subject's size.

Enhancements to the MCNP6 background source

DOE PAGES

McMath, Garrett E.; McKinney, Gregg W.

2015-10-19

The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutronmore » background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.« less

Computational Model of D-Region Ion Production Caused by Energetic Electron Precipitations Based on General Monte Carlo Transport Calculations

NASA Astrophysics Data System (ADS)

Kouznetsov, A.; Cully, C. M.

2017-12-01

During enhanced magnetic activities, large ejections of energetic electrons from radiation belts are deposited in the upper polar atmosphere where they play important roles in its physical and chemical processes, including VLF signals subionospheric propagation. Electron deposition can affect D-Region ionization, which are estimated based on ionization rates derived from energy depositions. We present a model of D-region ion production caused by an arbitrary (in energy and pitch angle) distribution of fast (10 keV - 1 MeV) electrons. The model relies on a set of pre-calculated results obtained using a general Monte Carlo approach with the latest version of the MCNP6 (Monte Carlo N-Particle) code for the explicit electron tracking in magnetic fields. By expressing those results using the ionization yield functions, the pre-calculated results are extended to cover arbitrary magnetic field inclinations and atmospheric density profiles, allowing ionization rate altitude profile computations in the range of 20 and 200 km at any geographic point of interest and date/time by adopting results from an external atmospheric density model (e.g. NRLMSISE-00). The pre-calculated MCNP6 results are stored in a CDF (Common Data Format) file, and IDL routines library is written to provide an end-user interface to the model.

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.

a Proposed Benchmark Problem for Scatter Calculations in Radiographic Modelling

NASA Astrophysics Data System (ADS)

Jaenisch, G.-R.; Bellon, C.; Schumm, A.; Tabary, J.; Duvauchelle, Ph.

2009-03-01

Code Validation is a permanent concern in computer modelling, and has been addressed repeatedly in eddy current and ultrasonic modeling. A good benchmark problem is sufficiently simple to be taken into account by various codes without strong requirements on geometry representation capabilities, focuses on few or even a single aspect of the problem at hand to facilitate interpretation and to avoid that compound errors compensate themselves, yields a quantitative result and is experimentally accessible. In this paper we attempt to address code validation for one aspect of radiographic modeling, the scattered radiation prediction. Many NDT applications can not neglect scattered radiation, and the scatter calculation thus is important to faithfully simulate the inspection situation. Our benchmark problem covers the wall thickness range of 10 to 50 mm for single wall inspections, with energies ranging from 100 to 500 keV in the first stage, and up to 1 MeV with wall thicknesses up to 70 mm in the extended stage. A simple plate geometry is sufficient for this purpose, and the scatter data is compared on a photon level, without a film model, which allows for comparisons with reference codes like MCNP. We compare results of three Monte Carlo codes (McRay, Sindbad and Moderato) as well as an analytical first order scattering code (VXI), and confront them to results obtained with MCNP. The comparison with an analytical scatter model provides insights into the application domain where this kind of approach can successfully replace Monte-Carlo calculations.

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.

Absorbed fractions in a voxel-based phantom calculated with the MCNP-4B code.

PubMed

Yoriyaz, H; dos Santos, A; Stabin, M G; Cabezas, R

2000-07-01

A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. MCNP-4B absorbed fractions for photons in the mathematical phantom of Snyder et al. agreed well with reference values. Results obtained through radiation transport simulation in the voxel-based phantom, in general, agreed well with reference values. Considerable discrepancies, however, were found in some cases due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the voxel-based phantom, which is not considered in the mathematical phantom.

Multi-threading performance of Geant4, MCNP6, and PHITS Monte Carlo codes for tetrahedral-mesh geometry.

PubMed

Han, Min Cheol; Yeom, Yeon Soo; Lee, Hyun Su; Shin, Bangho; Kim, Chan Hyeong; Furuta, Takuya

2018-05-04

In this study, the multi-threading performance of the Geant4, MCNP6, and PHITS codes was evaluated as a function of the number of threads (N) and the complexity of the tetrahedral-mesh phantom. For this, three tetrahedral-mesh phantoms of varying complexity (simple, moderately complex, and highly complex) were prepared and implemented in the three different Monte Carlo codes, in photon and neutron transport simulations. Subsequently, for each case, the initialization time, calculation time, and memory usage were measured as a function of the number of threads used in the simulation. It was found that for all codes, the initialization time significantly increased with the complexity of the phantom, but not with the number of threads. Geant4 exhibited much longer initialization time than the other codes, especially for the complex phantom (MRCP). The improvement of computation speed due to the use of a multi-threaded code was calculated as the speed-up factor, the ratio of the computation speed on a multi-threaded code to the computation speed on a single-threaded code. Geant4 showed the best multi-threading performance among the codes considered in this study, with the speed-up factor almost linearly increasing with the number of threads, reaching ~30 when N  =  40. PHITS and MCNP6 showed a much smaller increase of the speed-up factor with the number of threads. For PHITS, the speed-up factors were low when N  =  40. For MCNP6, the increase of the speed-up factors was better, but they were still less than ~10 when N  =  40. As for memory usage, Geant4 was found to use more memory than the other codes. In addition, compared to that of the other codes, the memory usage of Geant4 more rapidly increased with the number of threads, reaching as high as ~74 GB when N  =  40 for the complex phantom (MRCP). It is notable that compared to that of the other codes, the memory usage of PHITS was much lower, regardless of both the complexity of the phantom and the number of threads, hardly increasing with the number of threads for the MRCP.

Total reaction cross sections in CEM and MCNP6 at intermediate energies

DOE PAGES

Kerby, Leslie M.; Mashnik, Stepan G.

2015-05-14

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

Total reaction cross sections in CEM and MCNP6 at intermediate energies

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

Kerby, Leslie M.; Mashnik, Stepan G.

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy.

PubMed

Krstic, D; Markovic, V M; Jovanovic, Z; Milenkovic, B; Nikezic, D; Atanackovic, J

2014-10-01

Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Evaluation of RAPID for a UNF cask benchmark problem

NASA Astrophysics Data System (ADS)

Mascolino, Valerio; Haghighat, Alireza; Roskoff, Nathan J.

2017-09-01

This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.

Multiple Detector Optimization for Hidden Radiation Source Detection

DTIC Science & Technology

2015-03-26

important in achieving operationally useful methods for optimizing detector emplacement, the 2-D attenuation model approach promises to speed up the...process of hidden source detection significantly. The model focused on detection of the full energy peak of a radiation source. Methods to optimize... radioisotope identification is possible without using a computationally intensive stochastic model such as the Monte Carlo n-Particle (MCNP) code

Calculations of skyshine from an intense portable electron linac

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

Estes, G.P.; Hughes, H.G.; Fry, D.A.

1994-12-31

The MCNP Monte carlo code has been used at Los Alamos to calculate skyshine and terrain albedo efects from an intense portable electron linear accelerator that is to be used by the Russian Federation to radiograph nuclear weapons that may have been damaged by accidents. Relative dose rate profiles have been calculated. The design of the accelerator, along with a diagram, is presented.

Distributed multitasking ITS with PVM

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

Fan, W.C.; Halbleib, J.A. Sr.

1995-02-01

Advances of computer hardware and communication software have made it possible to perform parallel-processing computing on a collection of desktop workstations. For many applications, multitasking on a cluster of high-performance workstations has achieved performance comparable or better than that on a traditional supercomputer. From the point of view of cost-effectiveness, it also allows users to exploit available but unused computational resources, and thus achieve a higher performance-to-cost ratio. Monte Carlo calculations are inherently parallelizable because the individual particle trajectories can be generated independently with minimum need for interprocessor communication. Furthermore, the number of particle histories that can be generated inmore » a given amount of wall-clock time is nearly proportional to the number of processors in the cluster. This is an important fact because the inherent statistical uncertainty in any Monte Carlo result decreases as the number of histories increases. For these reasons, researchers have expended considerable effort to take advantage of different parallel architectures for a variety of Monte Carlo radiation transport codes, often with excellent results. The initial interest in this work was sparked by the multitasking capability of MCNP on a cluster of workstations using the Parallel Virtual Machine (PVM) software. On a 16-machine IBM RS/6000 cluster, it has been demonstrated that MCNP runs ten times as fast as on a single-processor CRAY YMP. In this paper, we summarize the implementation of a similar multitasking capability for the coupled electron/photon transport code system, the Integrated TIGER Series (ITS), and the evaluation of two load balancing schemes for homogeneous and heterogeneous networks.« less

AN ASSESSMENT OF MCNP WEIGHT WINDOWS

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

J. S. HENDRICKS; C. N. CULBERTSON

2000-01-01

The weight window variance reduction method in the general-purpose Monte Carlo N-Particle radiation transport code MCNPTM has recently been rewritten. In particular, it is now possible to generate weight window importance functions on a superimposed mesh, eliminating the need to subdivide geometries for variance reduction purposes. Our assessment addresses the following questions: (1) Does the new MCNP4C treatment utilize weight windows as well as the former MCNP4B treatment? (2) Does the new MCNP4C weight window generator generate importance functions as well as MCNP4B? (3) How do superimposed mesh weight windows compare to cell-based weight windows? (4) What are the shortcomingsmore » of the new MCNP4C weight window generator? Our assessment was carried out with five neutron and photon shielding problems chosen for their demanding variance reduction requirements. The problems were an oil well logging problem, the Oak Ridge fusion shielding benchmark problem, a photon skyshine problem, an air-over-ground problem, and a sample problem for variance reduction.« less

Comparison of penumbra regions produced by ancient Gamma knife model C and Gamma ART 6000 using Monte Carlo MCNP6 simulation.

PubMed

Banaee, Nooshin; Asgari, Sepideh; Nedaie, Hassan Ali

2018-07-01

The accuracy of penumbral measurements in radiotherapy is pivotal because dose planning computers require accurate data to adequately modeling the beams, which in turn are used to calculate patient dose distributions. Gamma knife is a non-invasive intracranial technique based on principles of the Leksell stereotactic system for open deep brain surgeries, invented and developed by Professor Lars Leksell. The aim of this study is to compare the penumbra widths of Leksell Gamma Knife model C and Gamma ART 6000. Initially, the structure of both systems were simulated by using Monte Carlo MCNP6 code and after validating the accuracy of simulation, beam profiles of different collimators were plotted. MCNP6 beam profile calculations showed that the penumbra values of Leksell Gamma knife model C and Gamma ART 6000 for 18, 14, 8 and 4 mm collimators are 9.7, 7.9, 4.3, 2.6 and 8.2, 6.9, 3.6, 2.4, respectively. The results of this study showed that since Gamma ART 6000 has larger solid angle in comparison with Gamma Knife model C, it produces better beam profile penumbras than Gamma Knife model C in the direct plane. Copyright © 2017 Elsevier Ltd. All rights reserved.

Simulation of irradiation exposure of electronic devices due to heavy ion therapy with Monte Carlo Code MCNP6

NASA Astrophysics Data System (ADS)

Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang; Buck, Arnulf

2017-09-01

During heavy ion irradiation therapy the patient has to be located exactly at the right position to make sure that the Bragg peak occurs in the tumour. The patient has to be moved in the range of millimetres to scan the ill tissue. For that reason a special table was developed which allows exact positioning. The electronic control can be located outside the surgery. But that has some disadvantage for the construction. To keep the system compact it would be much more comfortable to put the electronic control inside the surgery. As a lot of high energetic secondary particles are produced during the therapy causing a high dose in the room it is important to find positions with low dose rates. Therefore, investigations are needed where the electronic devices should be located to obtain a minimum of radiation, help to prevent the failure of sensitive devices. The dose rate was calculated for carbon ions with different initial energy and protons over the entire therapy room with Monte Carlo particle tracking using MCNP6. The types of secondary particles were identified and the dose rate for a thin silicon layer and an electronic mixture material was determined. In addition, the shielding effect of several selected material layers was calculated using MCNP6.

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

MCNP calculations for container inspection with tagged neutrons

NASA Astrophysics Data System (ADS)

Boghen, G.; Donzella, A.; Filippini, V.; Fontana, A.; Lunardon, M.; Moretto, S.; Pesente, S.; Zenoni, A.

2005-12-01

We are developing an innovative tagged neutrons inspection system (TNIS) for cargo containers: the system will allow us to assay the chemical composition of suspect objects, previously identified by a standard X-ray radiography. The operation of the system is extensively being simulated by using the MCNP Monte Carlo code to study different inspection geometries, cargo loads and hidden threat materials. Preliminary simulations evaluating the signal and the signal over background ratio expected as a function of the system parameters are presented. The results for a selection of cases are briefly discussed and demonstrate that the system can operate successfully in different filling conditions.

Skyshine line-beam response functions for 20- to 100-MeV photons

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

Brockhoff, R.C.; Shultis, J.K.; Faw, R.E.

1996-06-01

The line-beam response function, needed for skyshine analyses based on the integral line-beam method, was evaluated with the MCNP Monte Carlo code for photon energies from 20 to 100 MeV and for source-to-detector distances out to 1,000 m. These results are compared with point-kernel results, and the effects of bremsstrahlung and positron transport in the air are found to be important in this energy range. The three-parameter empirical formula used in the integral line-beam skyshine method was fit to the MCNP results, and values of these parameters are reported for various source energies and angles.

Detection of radioactive particles offshore by γ-ray spectrometry Part I: Monte Carlo assessment of detection depth limits

NASA Astrophysics Data System (ADS)

Maučec, M.; de Meijer, R. J.; Rigollet, C.; Hendriks, P. H. G. M.; Jones, D. G.

2004-06-01

A joint research project between the British Geological Survey and Nuclear Geophysics Division of the Kernfysisch Versneller Instituut, Groningen, the Netherlands, was commissioned by the United Kingdom Atomic Energy Authority to establish the efficiency of a towed seabed γ-ray spectrometer for the detection of 137Cs-containing radioactive particles offshore Dounreay, Scotland. Using the MCNP code, a comprehensive Monte Carlo feasibility study was carried out to model various combinations of geological matrices, particle burial depth and lateral displacement, source activity and detector material. To validate the sampling and absolute normalisation procedures of MCNP for geometries including multiple (natural and induced) heterogeneous sources in environmental monitoring, a benchmark experiment was conducted. The study demonstrates the ability of seabed γ-ray spectrometry to locate radioactive particles offshore and to distinguish between γ count rate increases due to particles from those due to enhanced natural radioactivity. The information presented in this study will be beneficial for estimation of the inventory of 137Cs particles and their activity distribution and for the recovery of particles from the sea floor. In this paper, the Monte Carlo assessment of the detection limits is presented. The estimation of the required towing speed and acquisition times and their application to radioactive particle detection and discrimination offshore formed a supplementary part of this study.

Comparison of Monte Carlo simulation of gamma ray attenuation coefficients of amino acids with XCOM program and experimental data

NASA Astrophysics Data System (ADS)

Elbashir, B. O.; Dong, M. G.; Sayyed, M. I.; Issa, Shams A. M.; Matori, K. A.; Zaid, M. H. M.

2018-06-01

The mass attenuation coefficients (μ/ρ), effective atomic numbers (Zeff) and electron densities (Ne) of some amino acids obtained experimentally by the other researchers have been calculated using MCNP5 simulations in the energy range 0.122-1.330 MeV. The simulated values of μ/ρ, Zeff, and Ne were compared with the previous experimental work for the amino acids samples and a good agreement was noticed. Moreover, the values of mean free path (MFP) for the samples were calculated using MCNP5 program and compared with the theoretical results obtained by XCOM. The investigation of μ/ρ, Zeff, Ne and MFP values of amino acids using MCNP5 simulations at various photon energies when compared with the XCOM values and previous experimental data for the amino acids samples revealed that MCNP5 code provides accurate photon interaction parameters for amino acids.

Quantitative basis for component factors of gas flow proportional counting efficiencies

NASA Astrophysics Data System (ADS)

Nichols, Michael C.

This dissertation investigates the counting efficiency calibration of a gas flow proportional counter with beta-particle emitters in order to (1) determine by measurements and simulation the values of the component factors of beta-particle counting efficiency for a proportional counter, (2) compare the simulation results and measured counting efficiencies, and (3) determine the uncertainty of the simulation and measurements. Monte Carlo simulation results by the MCNP5 code were compared with measured counting efficiencies as a function of sample thickness for 14C, 89Sr, 90Sr, and 90Y. The Monte Carlo model simulated strontium carbonate with areal thicknesses from 0.1 to 35 mg cm-2. The samples were precipitated as strontium carbonate with areal thicknesses from 3 to 33 mg cm-2 , mounted on membrane filters, and counted on a low background gas flow proportional counter. The estimated fractional standard deviation was 2--4% (except 6% for 14C) for efficiency measurements of the radionuclides. The Monte Carlo simulations have uncertainties estimated to be 5 to 6 percent for carbon-14 and 2.4 percent for strontium-89, strontium-90, and yttrium-90. The curves of simulated counting efficiency vs. sample areal thickness agreed within 3% of the curves of best fit drawn through the 25--49 measured points for each of the four radionuclides. Contributions from this research include development of uncertainty budgets for the analytical processes; evaluation of alternative methods for determining chemical yield critical to the measurement process; correcting a bias found in the MCNP normalization of beta spectra histogram; clarifying the interpretation of the commonly used ICRU beta-particle spectra for use by MCNP; and evaluation of instrument parameters as applied to the simulation model to obtain estimates of the counting efficiency from simulated pulse height tallies.

Benchmark study for charge deposition by high energy electrons in thick slabs

NASA Technical Reports Server (NTRS)

Jun, I.

2002-01-01

The charge deposition profiles created when highenergy (1, 10, and 100 MeV) electrons impinge ona thick slab of elemental aluminum, copper, andtungsten are presented in this paper. The chargedeposition profiles were computed using existing representative Monte Carlo codes: TIGER3.0 (1D module of ITS3.0) and MCNP version 4B. The results showed that TIGER3.0 and MCNP4B agree very well (within 20% of each other) in the majority of the problem geometry. The TIGER results were considered to be accurate based on previous studies. Thus, it was demonstrated that MCNP, with its powerful geometry capability and flexible source and tally options, could be used in calculations of electron charging in high energy electron-rich space radiation environments.

Theory and Performance of AIMS for Active Interrogation

NASA Astrophysics Data System (ADS)

Walters, William J.; Royston, Katherine E. K.; Haghighat, Alireza

2014-06-01

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) determination of neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, γ) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water. In the first step, a response-function formulation has been developed to calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, γ) cross sections to find the resulting gamma source distribution. Finally, in the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma flux at a detector window. A code, AIMS (Active Interrogation for Monitoring Special-Nuclear-materials), has been written to output the gamma current for an source-detector assembly scanning across the cargo using the pre-calculated values and takes significantly less time than a reference MCNP5 calculation.

Application of the MCNP5 code to the Modeling of vaginal and intra-uterine applicators used in intracavitary brachytherapy: a first approach

NASA Astrophysics Data System (ADS)

Gerardy, I.; Rodenas, J.; Van Dycke, M.; Gallardo, S.; Tondeur, F.

2008-02-01

Brachytherapy is a radiotherapy treatment where encapsulated radioactive sources are introduced within a patient. Depending on the technique used, such sources can produce high, medium or low local dose rates. The Monte Carlo method is a powerful tool to simulate sources and devices in order to help physicists in treatment planning. In multiple types of gynaecological cancer, intracavitary brachytherapy (HDR Ir-192 source) is used combined with other therapy treatment to give an additional local dose to the tumour. Different types of applicators are used in order to increase the dose imparted to the tumour and to limit the effect on healthy surrounding tissues. The aim of this work is to model both applicator and HDR source in order to evaluate the dose at a reference point as well as the effect of the materials constituting the applicators on the near field dose. The MCNP5 code based on the Monte Carlo method has been used for the simulation. Dose calculations have been performed with *F8 energy deposition tally, taking into account photons and electrons. Results from simulation have been compared with experimental in-phantom dose measurements. Differences between calculations and measurements are lower than 5%.The importance of the source position has been underlined.

Shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses using WinXCom and MCNP5 code

NASA Astrophysics Data System (ADS)

Dong, M. G.; El-Mallawany, R.; Sayyed, M. I.; Tekin, H. O.

2017-12-01

Gamma ray shielding properties of 80TeO2-5TiO2-(15-x) WO3-xAnOm glasses, where AnOm is Nb2O5 = 0.01, 5, Nd2O3 = 3, 5 and Er2O3 = 5 mol% have been achieved. Shielding parameters; mass attenuation coefficients, half value layers, and macroscopic effective removal cross section for fast neutrons have been computed by using WinXCom program and MCNP5 Monte Carlo code. In addition, by using Geometric Progression method (G-P), exposure buildup factor values were also calculated. Variations of shielding parameters are discussed for the effect of REO addition into the glasses and photon energy.

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

An investigation of voxel geometries for MCNP-based radiation dose calculations.

PubMed

Zhang, Juying; Bednarz, Bryan; Xu, X George

2006-11-01

Voxelized geometry such as those obtained from medical images is increasingly used in Monte Carlo calculations of absorbed doses. One useful application of calculated absorbed dose is the determination of fluence-to-dose conversion factors for different organs. However, confusion still exists about how such a geometry is defined and how the energy deposition is best computed, especially involving a popular code, MCNP5. This study investigated two different types of geometries in the MCNP5 code, cell and lattice definitions. A 10 cm x 10 cm x 10 cm test phantom, which contained an embedded 2 cm x 2 cm x 2 cm target at its center, was considered. A planar source emitting parallel photons was also considered in the study. The results revealed that MCNP5 does not calculate total target volume for multi-voxel geometries. Therefore, tallies which involve total target volume must be divided by the user by the total number of voxels to obtain a correct dose result. Also, using planar source areas greater than the phantom size results in the same fluence-to-dose conversion factor.

An analysis of MCNP cross-sections and tally methods for low-energy photon emitters.

PubMed

Demarco, John J; Wallace, Robert E; Boedeker, Kirsten

2002-04-21

Monte Carlo calculations are frequently used to analyse a variety of radiological science applications using low-energy (10-1000 keV) photon sources. This study seeks to create a low-energy benchmark for the MCNP Monte Carlo code by simulating the absolute dose rate in water and the air-kerma rate for monoenergetic point sources with energies between 10 keV and 1 MeV. The analysis compares four cross-section datasets as well as the tally method for collision kerma versus absorbed dose. The total photon attenuation coefficient cross-section for low atomic number elements has changed significantly as cross-section data have changed between 1967 and 1989. Differences of up to 10% are observed in the photoelectric cross-section for water at 30 keV between the standard MCNP cross-section dataset (DLC-200) and the most recent XCOM/NIST tabulation. At 30 keV, the absolute dose rate in water at 1.0 cm from the source increases by 7.8% after replacing the DLC-200 photoelectric cross-sections for water with those from the XCOM/NIST tabulation. The differences in the absolute dose rate are analysed when calculated with either the MCNP absorbed dose tally or the collision kerma tally. Significant differences between the collision kerma tally and the absorbed dose tally can occur when using the DLC-200 attenuation coefficients in conjunction with a modern tabulation of mass energy-absorption coefficients.

A Monte Carlo simulation and setup optimization of output efficiency to PGNAA thermal neutron using 252Cf neutrons

NASA Astrophysics Data System (ADS)

Zhang, Jin-Zhao; Tuo, Xian-Guo

2014-07-01

We present the design and optimization of a prompt γ-ray neutron activation analysis (PGNAA) thermal neutron output setup based on Monte Carlo simulations using MCNP5 computer code. In these simulations, the moderator materials, reflective materials, and structure of the PGNAA 252Cf neutrons of thermal neutron output setup are optimized. The simulation results reveal that the thin layer paraffin and the thick layer of heavy water moderating effect work best for the 252Cf neutron spectrum. Our new design shows a significantly improved performance of the thermal neutron flux and flux rate, that are increased by 3.02 times and 3.27 times, respectively, compared with the conventional neutron source design.

Commissioning and initial acceptance tests for a commercial convolution dose calculation algorithm for radiotherapy treatment planning in comparison with Monte Carlo simulation and measurement

PubMed Central

Moradi, Farhad; Mahdavi, Seyed Rabi; Mostaar, Ahmad; Motamedi, Mohsen

2012-01-01

In this study the commissioning of a dose calculation algorithm in a currently used treatment planning system was performed and the calculation accuracy of two available methods in the treatment planning system i.e., collapsed cone convolution (CCC) and equivalent tissue air ratio (ETAR) was verified in tissue heterogeneities. For this purpose an inhomogeneous phantom (IMRT thorax phantom) was used and dose curves obtained by the TPS (treatment planning system) were compared with experimental measurements and Monte Carlo (MCNP code) simulation. Dose measurements were performed by using EDR2 radiographic films within the phantom. Dose difference (DD) between experimental results and two calculation methods was obtained. Results indicate maximum difference of 12% in the lung and 3% in the bone tissue of the phantom between two methods and the CCC algorithm shows more accurate depth dose curves in tissue heterogeneities. Simulation results show the accurate dose estimation by MCNP4C in soft tissue region of the phantom and also better results than ETAR method in bone and lung tissues. PMID:22973081

Reconstruction of Human Monte Carlo Geometry from Segmented Images

NASA Astrophysics Data System (ADS)

Zhao, Kai; Cheng, Mengyun; Fan, Yanchang; Wang, Wen; Long, Pengcheng; Wu, Yican

2014-06-01

Human computational phantoms have been used extensively for scientific experimental analysis and experimental simulation. This article presented a method for human geometry reconstruction from a series of segmented images of a Chinese visible human dataset. The phantom geometry could actually describe detailed structure of an organ and could be converted into the input file of the Monte Carlo codes for dose calculation. A whole-body computational phantom of Chinese adult female has been established by FDS Team which is named Rad-HUMAN with about 28.8 billion voxel number. For being processed conveniently, different organs on images were segmented with different RGB colors and the voxels were assigned with positions of the dataset. For refinement, the positions were first sampled. Secondly, the large sums of voxels inside the organ were three-dimensional adjacent, however, there were not thoroughly mergence methods to reduce the cell amounts for the description of the organ. In this study, the voxels on the organ surface were taken into consideration of the mergence which could produce fewer cells for the organs. At the same time, an indexed based sorting algorithm was put forward for enhancing the mergence speed. Finally, the Rad-HUMAN which included a total of 46 organs and tissues was described by the cuboids into the Monte Carlo Monte Carlo Geometry for the simulation. The Monte Carlo geometry was constructed directly from the segmented images and the voxels was merged exhaustively. Each organ geometry model was constructed without ambiguity and self-crossing, its geometry information could represent the accuracy appearance and precise interior structure of the organs. The constructed geometry largely retaining the original shape of organs could easily be described into different Monte Carlo codes input file such as MCNP. Its universal property was testified and high-performance was experimentally verified

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

COMPTEL neutron response at 17 MeV

NASA Technical Reports Server (NTRS)

Oneill, Terrence J.; Ait-Ouamer, Farid; Morris, Joann; Tumer, O. Tumay; White, R. Stephen; Zych, Allen D.

1992-01-01

The Compton imaging telescope (COMPTEL) instrument of the Gamma Ray Observatory was exposed to 17 MeV d,t neutrons prior to launch. These data were analyzed and compared with Monte Carlo calculations using the MCNP(LANL) code. Energy and angular resolutions are compared and absolute efficiencies are calculated at 0 and 30 degrees incident angle. The COMPTEL neutron responses at 17 MeV and higher energies are needed to understand solar flare neutron data.

Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination

PubMed Central

Liu, B; Xu, J; Liu, T; Ouyang, X

2012-01-01

Objective To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Methods Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a 252Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D–D neutron generator can create neutrons at up to 1013 n s−1 with current technology. All these enable an effective and low-cost method of killing anthrax spores. Results There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. Conclusion The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g 252Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D–D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D–D neutron generator output >1013 n s−1 should be attainable in the near future. This indicates that we could use a D–D neutron generator to sterilise anthrax contamination within several seconds. PMID:22573293

Development and Validation of a Monte Carlo Simulation Tool for Multi-Pinhole SPECT

PubMed Central

Mok, Greta S. P.; Du, Yong; Wang, Yuchuan; Frey, Eric C.; Tsui, Benjamin M. W.

2011-01-01

Purpose In this work, we developed and validated a Monte Carlo simulation (MCS) tool for investigation and evaluation of multi-pinhole (MPH) SPECT imaging. Procedures This tool was based on a combination of the SimSET and MCNP codes. Photon attenuation and scatter in the object, as well as penetration and scatter through the collimator detector, are modeled in this tool. It allows accurate and efficient simulation of MPH SPECT with focused pinhole apertures and user-specified photon energy, aperture material, and imaging geometry. The MCS method was validated by comparing the point response function (PRF), detection efficiency (DE), and image profiles obtained from point sources and phantom experiments. A prototype single-pinhole collimator and focused four- and five-pinhole collimators fitted on a small animal imager were used for the experimental validations. We have also compared computational speed among various simulation tools for MPH SPECT, including SimSET-MCNP, MCNP, SimSET-GATE, and GATE for simulating projections of a hot sphere phantom. Results We found good agreement between the MCS and experimental results for PRF, DE, and image profiles, indicating the validity of the simulation method. The relative computational speeds for SimSET-MCNP, MCNP, SimSET-GATE, and GATE are 1: 2.73: 3.54: 7.34, respectively, for 120-view simulations. We also demonstrated the application of this MCS tool in small animal imaging by generating a set of low-noise MPH projection data of a 3D digital mouse whole body phantom. Conclusions The new method is useful for studying MPH collimator designs, data acquisition protocols, image reconstructions, and compensation techniques. It also has great potential to be applied for modeling the collimator-detector response with penetration and scatter effects for MPH in the quantitative reconstruction method. PMID:19779896

Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.

PubMed

Liu, B; Xu, J; Liu, T; Ouyang, X

2012-10-01

To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.

Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

NASA Astrophysics Data System (ADS)

Fensin, Michael Lorne

Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established and supported radiation transport code, for further development of a Monte Carlo-linked depletion methodology which is essential to the future development of advanced reactor technologies that exceed the limitations of current deterministic based methods.

Analytic score distributions for a spatially continuous tridirectional Monte Carol transport problem

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

Booth, T.E.

1996-01-01

The interpretation of the statistical error estimates produced by Monte Carlo transport codes is still somewhat of an art. Empirically, there are variance reduction techniques whose error estimates are almost always reliable, and there are variance reduction techniques whose error estimates are often unreliable. Unreliable error estimates usually result from inadequate large-score sampling from the score distribution`s tail. Statisticians believe that more accurate confidence interval statements are possible if the general nature of the score distribution can be characterized. Here, the analytic score distribution for the exponential transform applied to a simple, spatially continuous Monte Carlo transport problem is provided.more » Anisotropic scattering and implicit capture are included in the theory. In large part, the analytic score distributions that are derived provide the basis for the ten new statistical quality checks in MCNP.« less

Monte Carol-Based Dosimetry of Beta-Emitters for Intravascular Brachytherapy

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

Choi, C.K.

2002-06-25

Monte Carlo simulations for radiation dosimetry and the experimental verifications of the simulations have been developed for the treatment geometry of intravascular brachytherapy, a form of radionuclide therapy for occluded coronary disease (restenosis). Monte Carlo code, MCNP4C, has been used to calculate the radiation dose from the encapsulated array of B-emitting seeds (Sr/Y-source train). Solid water phantoms have been fabricated to measure the dose on the radiochromic films that were exposed to the beta source train for both linear and curved coronary vessel geometries. While the dose difference for the 5-degree curved vessel at the prescription point of f+2.0 mmmore » is within the 10% guideline set by the AAPM, however, the difference increased dramatically to 16.85% for the 10-degree case which requires additional adjustment for the acceptable dosimetry planning. The experimental dose measurements agree well with the simulation results« less

MCNP simulation of a Theratron 780 radiotherapy unit.

PubMed

Miró, R; Soler, J; Gallardo, S; Campayo, J M; Díez, S; Verdú, G

2005-01-01

A Theratron 780 (MDS Nordion) 60Co radiotherapy unit has been simulated with the Monte Carlo code MCNP. The unit has been realistically modelled: the cylindrical source capsule and its housing, the rectangular collimator system, both the primary and secondary jaws and the air gaps between the components. Different collimator openings, ranging from 5 x 5 cm2 to 20 x 20 cm2 (narrow and broad beams) at a source-surface distance equal to 80 cm have been used during the study. In the present work, we have calculated spectra as a function of field size. A study of the variation of the electron contamination of the 60Co beam has also been performed.

Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.

PubMed

Hordósy, G

2005-01-01

In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002.

Source terms, shielding calculations and soil activation for a medical cyclotron.

PubMed

Konheiser, J; Naumann, B; Ferrari, A; Brachem, C; Müller, S E

2016-12-01

Calculations of the shielding and estimates of soil activation for a medical cyclotron are presented in this work. Based on the neutron source term from the 18 O(p,n) 18 F reaction produced by a 28 MeV proton beam, neutron and gamma dose rates outside the building were estimated with the Monte Carlo code MCNP6 (Goorley et al 2012 Nucl. Technol. 180 298-315). The neutron source term was calculated with the MCNP6 code and FLUKA (Ferrari et al 2005 INFN/TC_05/11, SLAC-R-773) code as well as with supplied data by the manufacturer. MCNP and FLUKA calculations yielded comparable results, while the neutron yield obtained using the manufacturer-supplied information is about a factor of 5 smaller. The difference is attributed to the missing channels in the manufacturer-supplied neutron source terms which considers only the 18 O(p,n) 18 F reaction, whereas the MCNP and FLUKA calculations include additional neutron reaction channels. Soil activation was performed using the FLUKA code. The estimated dose rate based on MCNP6 calculations in the public area is about 0.035 µSv h -1 and thus significantly below the reference value of 0.5 µSv h -1 (2011 Strahlenschutzverordnung, 9 Auflage vom 01.11.2011, Bundesanzeiger Verlag). After 5 years of continuous beam operation and a subsequent decay time of 30 d, the activity concentration of the soil is about 0.34 Bq g -1 .

LIBMAKER

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

2015-08-01

Version 00 COG LibMaker contains various utilities to convert common data formats into a format usable by the COG - Multi-particle Monte Carlo Code System package, (C00777MNYCP01). Utilities included: ACEtoCOG - ACE formatted neutron data: Currently ENDFB7R0.BNL, ENDFB7R1.BNL, JEFF3.1, JEFF3.1.1, JEFF3.1.2, MCNP.50c, MCNP.51c, MCNP.55c, MCNP.66c, and MCNP.70c. ACEUtoCOG - ACEU formatted photonuclear data: Currently PN.MCNP.30c and PN.MCNP.70u. ACTLtoCOG - Creates a COG library from ENDL formatted activation data COG library. EDDLtoCOG - Creates a COG library from ENDL formatted LLNL deuteron data. ENDLtoCOG - Creates a COG library from ENDL formatted LLNL neutron data. EPDLtoCOG - Creates a COG librarymore » from ENDL formatted LLNL photon data. LEX - Creates a COG dictionary file. SAB.ACEtoCOG - Creates a COG library from ACE formatted S(a,b) data. SABtoCOG - Creates a COG library from ENDF6 formatted S(a,b) data. URRtoCOG - Creates a COG library from ACE formatted probability table data. This package also includes library checking and bit swapping capability.« less

Assay of the Martian Regolith with Neutrons

NASA Technical Reports Server (NTRS)

Drake, Darrell M.

1997-01-01

The purpose of the research is to combine experiments and Monte Carlo transport of neutrons through volume of soil in an attempt to model neutron leakage from planetary surfaces. Emphasis is given to the change of neutron spectra as a function of water content and location. During the first stage of effort, two experiments were conducted in which leakage of neutrons from a Pu-Be source through about 30 g/cm(exp 2) of soil were measured with several counters. A Monte Carlo code, MCNP, has been used to model many of the 100 individual runs of the experiment. Hydrogen is the element that has the most dramatic effect on the neutron spectrum and its effect on the neutron spectrum is almost the same whether it is in the form of water or polyethylene. In order to simulate various water configurations, sheets of polyethylene have been used between layers of soil as well as water in several concentrations up to 18%. Comparison of experimental results to theoretical predictions made with the MCNP code were disappointing for low concentrations of water. We have made extensive calculations to see if room return could be the cause of the discrepancies. Water concentrations of the 'dry' soil were measured by two different laboratories and differed only by 0.5%. We have made calculations to optimize the next experiment and are investigating other methods of determining the water content of 'dry' soil.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

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

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

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

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

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

2015-06-15

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

Performance of MCNP4A on seven computing platforms

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

Hendricks, J.S.; Brockhoff, R.C.

1994-12-31

The performance of seven computer platforms has been evaluated with the MCNP4A Monte Carlo radiation transport code. For the first time we report timing results using MCNP4A and its new test set and libraries. Comparisons are made on platforms not available to us in previous MCNP timing studies. By using MCNP4A and its 325-problem test set, a widely-used and readily-available physics production code is used; the timing comparison is not limited to a single ``typical`` problem, demonstrating the problem dependence of timing results; the results are reproducible at the more than 100 installations around the world using MCNP; comparison ofmore » performance of other computer platforms to the ones tested in this study is possible because we present raw data rather than normalized results; and a measure of the increase in performance of computer hardware and software over the past two years is possible. The computer platforms reported are the Cray-YMP 8/64, IBM RS/6000-560, Sun Sparc10, Sun Sparc2, HP/9000-735, 4 processor 100 MHz Silicon Graphics ONYX, and Gateway 2000 model 4DX2-66V PC. In 1991 a timing study of MCNP4, the predecessor to MCNP4A, was conducted using ENDF/B-V cross-section libraries, which are export protected. The new study is based upon the new MCNP 25-problem test set which utilizes internationally available data. MCNP4A, its test problems and the test data library are available from the Radiation Shielding and Information Center in Oak Ridge, Tennessee, or from the NEA Data Bank in Saclay, France. Anyone with the same workstation and compiler can get the same test problem sets, the same library files, and the same MCNP4A code from RSIC or NEA and replicate our results. And, because we report raw data, comparison of the performance of other compute platforms and compilers can be made.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Validation of MCNP: SPERT-D and BORAX-V fuel

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

Crawford, C.; Palmer, B.

1992-11-01

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

Image enhancement using MCNP5 code and MATLAB in neutron radiography.

PubMed

Tharwat, Montaser; Mohamed, Nader; Mongy, T

2014-07-01

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work. Copyright © 2014 Elsevier Ltd. All rights reserved.

SABRINA: an interactive solid geometry modeling program for Monte Carlo

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

West, J.T.

SABRINA is a fully interactive three-dimensional geometry modeling program for MCNP. In SABRINA, a user interactively constructs either body geometry, or surface geometry models, and interactively debugs spatial descriptions for the resulting objects. This enhanced capability significantly reduces the effort in constructing and debugging complicated three-dimensional geometry models for Monte Carlo Analysis.

Monte Carlo based dosimetry for neutron capture therapy of brain tumors

NASA Astrophysics Data System (ADS)

Zaidi, Lilia; Belgaid, Mohamed; Khelifi, Rachid

2016-11-01

Boron Neutron Capture Therapy (BNCT) is a biologically targeted, radiation therapy for cancer which combines neutron irradiation with a tumor targeting agent labeled with a boron10 having a high thermal neutron capture cross section. The tumor area is subjected to the neutron irradiation. After a thermal neutron capture, the excited 11B nucleus fissions into an alpha particle and lithium recoil nucleus. The high Linear Energy Transfer (LET) emitted particles deposit their energy in a range of about 10μm, which is of the same order of cell diameter [1], at the same time other reactions due to neutron activation with body component are produced. In-phantom measurement of physical dose distribution is very important for BNCT planning validation. Determination of total absorbed dose requires complex calculations which were carried out using the Monte Carlo MCNP code [2].

The MCNP Simulation of a PGNAA System at TRR-1/M1

NASA Astrophysics Data System (ADS)

Sangaroon, S.; Ratanatongchai, W.; Picha, R.; Khaweerat, S.; Channuie, J.

2017-06-01

The prompt-gamma neutron activation analysis system (PGNAA) has been installed at Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1999. The purpose of the system is for elemental and isotopic analyses. The system mainly consists of a series of the moderator and collimator, neutron and gamma-ray shielding and the HPGe detector. In this work, the condition of the system is carried out based on the Monte Carlo method using Monte Carlo N-Particle transport code and the experiment. The flux ratios (Φthermal/Φepithermal and Φthermal/Φfast) and thermal neutron flux have been obtained. The simulated prompt gamma rays of the Portland cement sample have been carried out. The simulation provides significant contribution in upgrading the PGNAA station to be available in various applications.

The use of tetrahedral mesh geometries in Monte Carlo simulation of applicator based brachytherapy dose distributions

NASA Astrophysics Data System (ADS)

Paiva Fonseca, Gabriel; Landry, Guillaume; White, Shane; D'Amours, Michel; Yoriyaz, Hélio; Beaulieu, Luc; Reniers, Brigitte; Verhaegen, Frank

2014-10-01

Accounting for brachytherapy applicator attenuation is part of the recommendations from the recent report of AAPM Task Group 186. To do so, model based dose calculation algorithms require accurate modelling of the applicator geometry. This can be non-trivial in the case of irregularly shaped applicators such as the Fletcher Williamson gynaecological applicator or balloon applicators with possibly irregular shapes employed in accelerated partial breast irradiation (APBI) performed using electronic brachytherapy sources (EBS). While many of these applicators can be modelled using constructive solid geometry (CSG), the latter may be difficult and time-consuming. Alternatively, these complex geometries can be modelled using tessellated geometries such as tetrahedral meshes (mesh geometries (MG)). Recent versions of Monte Carlo (MC) codes Geant4 and MCNP6 allow for the use of MG. The goal of this work was to model a series of applicators relevant to brachytherapy using MG. Applicators designed for 192Ir sources and 50 kV EBS were studied; a shielded vaginal applicator, a shielded Fletcher Williamson applicator and an APBI balloon applicator. All applicators were modelled in Geant4 and MCNP6 using MG and CSG for dose calculations. CSG derived dose distributions were considered as reference and used to validate MG models by comparing dose distribution ratios. In general agreement within 1% for the dose calculations was observed for all applicators between MG and CSG and between codes when considering volumes inside the 25% isodose surface. When compared to CSG, MG required longer computation times by a factor of at least 2 for MC simulations using the same code. MCNP6 calculation times were more than ten times shorter than Geant4 in some cases. In conclusion we presented methods allowing for high fidelity modelling with results equivalent to CSG. To the best of our knowledge MG offers the most accurate representation of an irregular APBI balloon applicator.

Treating electron transport in MCNP{sup trademark}

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

Hughes, H.G.

1996-12-31

The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. Themore » theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.« less

Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

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

Brown, Forrest B.

2016-09-20

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such asmore » MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.« less

TH-AB-207A-07: Radiation Dose Simulation for a Newly Proposed Dynamic Bowtie Filters for CT Using Fast Monte Carlo Methods

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

Liu, T; Lin, H; Gao, Y

Purpose: Dynamic bowtie filter is an innovative design capable of modulating the X-ray and balancing the flux in the detectors, and it introduces a new way of patient-specific CT scan optimizations. This study demonstrates the feasibility of performing fast Monte Carlo dose calculation for a type of dynamic bowtie filter for cone-beam CT (Liu et al. 2014 9(7) PloS one) using MIC coprocessors. Methods: The dynamic bowtie filter in question consists of a highly attenuating bowtie component (HB) and a weakly attenuating bowtie (WB). The HB is filled with CeCl3 solution and its surface is defined by a transcendental equation.more » The WB is an elliptical cylinder filled with air and immersed in the HB. As the scanner rotates, the orientation of WB remains the same with the static patient. In our Monte Carlo simulation, the HB was approximated by 576 boxes. The phantom was a voxelized elliptical cylinder composed of PMMA and surrounded by air (44cm×44cm×40cm, 1000×1000×1 voxels). The dose to the PMMA phantom was tallied with 0.15% statistical uncertainty under 100 kVp source. Two Monte Carlo codes ARCHER and MCNP-6.1 were compared. Both used double-precision. Compiler flags that may trade accuracy for speed were avoided. Results: The wall time of the simulation was 25.4 seconds by ARCHER on a 5110P MIC, 40 seconds on a X5650 CPU, and 523 seconds by the multithreaded MCNP on the same CPU. The high performance of ARCHER is attributed to the parameterized geometry and vectorization of the program hotspots. Conclusion: The dynamic bowtie filter modeled in this study is able to effectively reduce the dynamic range of the detected signals for the photon-counting detectors. With appropriate software optimization methods, the accelerator-based (MIC and GPU) Monte Carlo dose engines have shown good performance and can contribute to patient-specific CT scan optimizations.« less

Evaluation of the DRAGON code for VHTR design analysis.

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

Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

2006-01-12

This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by themore » IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.« less

Copper benchmark experiment for the testing of JEFF-3.2 nuclear data for fusion applications

NASA Astrophysics Data System (ADS)

Angelone, M.; Flammini, D.; Loreti, S.; Moro, F.; Pillon, M.; Villar, R.; Klix, A.; Fischer, U.; Kodeli, I.; Perel, R. L.; Pohorecky, W.

2017-09-01

A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 70 cm3) aimed at testing and validating the recent nuclear data libraries for fusion applications was performed in the frame of the European Fusion Program at the 14 MeV ENEA Frascati Neutron Generator (FNG). Reaction rates, neutron flux spectra and doses were measured using different experimental techniques (e.g. activation foils techniques, NE213 scintillator and thermoluminescent detectors). This paper first summarizes the analyses of the experiment carried-out using the MCNP5 Monte Carlo code and the European JEFF-3.2 library. Large discrepancies between calculation (C) and experiment (E) were found for the reaction rates both in the high and low neutron energy range. The analysis was complemented by sensitivity/uncertainty analyses (S/U) using the deterministic and Monte Carlo SUSD3D and MCSEN codes, respectively. The S/U analyses enabled to identify the cross sections and energy ranges which are mostly affecting the calculated responses. The largest discrepancy among the C/E values was observed for the thermal (capture) reactions indicating severe deficiencies in the 63,65Cu capture and elastic cross sections at lower rather than at high energy. Deterministic and MC codes produced similar results. The 14 MeV copper experiment and its analysis thus calls for a revision of the JEFF-3.2 copper cross section and covariance data evaluation. A new analysis of the experiment was performed with the MCNP5 code using the revised JEFF-3.3-T2 library released by NEA and a new, not yet distributed, revised JEFF-3.2 Cu evaluation produced by KIT. A noticeable improvement of the C/E results was obtained with both new libraries.

The MCNP6 Analytic Criticality Benchmark Suite

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

Brown, Forrest B.

2016-06-16

Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling)more » and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.« less

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

A quantitative three-dimensional dose attenuation analysis around Fletcher-Suit-Delclos due to stainless steel tube for high-dose-rate brachytherapy by Monte Carlo calculations.

PubMed

Parsai, E Ishmael; Zhang, Zhengdong; Feldmeier, John J

2009-01-01

The commercially available brachytherapy treatment-planning systems today, usually neglects the attenuation effect from stainless steel (SS) tube when Fletcher-Suit-Delclos (FSD) is used in treatment of cervical and endometrial cancers. This could lead to potential inaccuracies in computing dwell times and dose distribution. A more accurate analysis quantifying the level of attenuation for high-dose-rate (HDR) iridium 192 radionuclide ((192)Ir) source is presented through Monte Carlo simulation verified by measurement. In this investigation a general Monte Carlo N-Particles (MCNP) transport code was used to construct a typical geometry of FSD through simulation and compare the doses delivered to point A in Manchester System with and without the SS tubing. A quantitative assessment of inaccuracies in delivered dose vs. the computed dose is presented. In addition, this investigation expanded to examine the attenuation-corrected radial and anisotropy dose functions in a form parallel to the updated AAPM Task Group No. 43 Report (AAPM TG-43) formalism. This will delineate quantitatively the inaccuracies in dose distributions in three-dimensional space. The changes in dose deposition and distribution caused by increased attenuation coefficient resulted from presence of SS are quantified using MCNP Monte Carlo simulations in coupled photon/electron transport. The source geometry was that of the Vari Source wire model VS2000. The FSD was that of the Varian medical system. In this model, the bending angles of tandem and colpostats are 15 degrees and 120 degrees , respectively. We assigned 10 dwell positions to the tandem and 4 dwell positions to right and left colpostats or ovoids to represent a typical treatment case. Typical dose delivered to point A was determined according to Manchester dosimetry system. Based on our computations, the reduction of dose to point A was shown to be at least 3%. So this effect presented by SS-FSD systems on patient dose is of concern.

Dosimetric study of a brachytherapy treatment of esophagus with Brazilian 192Ir sources using an anthropomorphic phantom

NASA Astrophysics Data System (ADS)

Neves, Lucio P.; Santos, William S.; Gorski, Ronan; Perini, Ana P.; Maia, Ana F.; Caldas, Linda V. E.; Orengo, Gilberto

2014-11-01

Several radioisotopes are produced at Instituto de Pesquisas Energéticas e Nucleares for the use in medical treatments, including the activation of 192Ir sources. These sources are suitable for brachytherapy treatments, due to their low or high activity, depending on the concentration of 192Ir, easiness to manufacture, small size, stable daughter products and the possibility of re-utilization. They may be used for the treatment of prostate, cervix, head and neck, skin, breast, gallbladder, uterus, vagina, lung, rectum, and eye cancer treatment. In this work, the use of some 192Ir sources was studied for the treatment of esophagus cancer, especially the dose determination of important structures, such as those on the mediastinum. This was carried out utilizing a FASH anthropomorphic phantom and the MCNP5 Monte Carlo code to transport the radiation through matter. It was possible to observe that the doses at lungs, breast, esophagus, thyroid and heart were the highest, which was expected due to their proximity to the source. Therefore, the data are useful to assess the representative dose specific to brachytherapy treatments on the esophagus for radiation protection purposes. The use of brachytherapy sources was studied for the treatment of esophagus cancer. FASH anthropomorphic phantom and MCNP5 Monte Carlo code were employed. The doses at lungs, breast, esophagus, thyroid and heart were the highest. The data is useful to assess the representative doses of treatments on the esophagus.

Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

NASA Astrophysics Data System (ADS)

Peri, Eyal; Orion, Itzhak

2017-09-01

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

Simulation the spatial resolution of an X-ray imager based on zinc oxide nanowires in anodic aluminium oxide membrane by using MCNP and OPTICS Codes

NASA Astrophysics Data System (ADS)

Samarin, S. N.; Saramad, S.

2018-05-01

The spatial resolution of a detector is a very important parameter for x-ray imaging. A bulk scintillation detector because of spreading of light inside the scintillator does't have a good spatial resolution. The nanowire scintillators because of their wave guiding behavior can prevent the spreading of light and can improve the spatial resolution of traditional scintillation detectors. The zinc oxide (ZnO) scintillator nanowire, with its simple construction by electrochemical deposition in regular hexagonal structure of Aluminum oxide membrane has many advantages. The three dimensional absorption of X-ray energy in ZnO scintillator is simulated by a Monte Carlo transport code (MCNP). The transport, attenuation and scattering of the generated photons are simulated by a general-purpose scintillator light response simulation code (OPTICS). The results are compared with a previous publication which used a simulation code of the passage of particles through matter (Geant4). The results verify that this scintillator nanowire structure has a spatial resolution less than one micrometer.

Characterization of Filters Loaded With Reactor Strontium Carbonate - 13203

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

Josephson, Walter S.; Steen, Franciska H.

A collection of three highly radioactive filters containing reactor strontium carbonate were being prepared for disposal. All three filters were approximately characterized at the time of manufacture by gravimetric methods. The first filter had been partially emptied, and the quantity of residual activity was uncertain. Dose rate to activity modeling using the Monte-Carlo N Particle (MCNP) code was selected to confirm the gravimetric characterization of the full filters, and to fully characterize the partially emptied filter. Although dose rate to activity modeling using MCNP is a common technique, it is not often used for Bremsstrahlung-dominant materials such as reactor strontium.more » As a result, different MCNP modeling options were compared to determine the optimum approach. This comparison indicated that the accuracy of the results were heavily dependent on the MCNP modeling details and the location of the dose rate measurement point. The optimum model utilized a photon spectrum generated by the Oak Ridge Isotope Generation and Depletion (ORIGEN) code and dose rates measured at 30 cm. Results from the optimum model agreed with the gravimetric estimates within 15%. It was demonstrated that dose rate to activity modeling can be successful for Bremsstrahlung-dominant radioactive materials. However, the degree of success is heavily dependent on the choice of modeling techniques. (authors)« less

Benchmark test of transport calculations of gold and nickel activation with implications for neutron kerma at Hiroshima.

PubMed

Hoshi, M; Hiraoka, M; Hayakawa, N; Sawada, S; Munaka, M; Kuramoto, A; Oka, T; Iwatani, K; Shizuma, K; Hasai, H

1992-11-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a 252Cf fission neutron source to validate the use of the code for the energy spectrum analyses of Hiroshima atomic bomb neutrons. Nuclear data libraries used in the Monte Carlo neutron and photon transport code calculation were ENDF/B-III, ENDF/B-IV, LASL-SUB, and ENDL-73. The neutron moderators used were granite (the main component of which is SiO2, with a small fraction of hydrogen), Newlight [polyethylene with 3.7% boron (natural)], ammonium chloride (NH4Cl), and water (H2O). Each moderator was 65 cm thick. The neutron detectors were gold and nickel foils, which were used to detect thermal and epithermal neutrons (4.9 eV) and fast neutrons (> 0.5 MeV), respectively. Measured activity data from neutron-irradiated gold and nickel foils in these moderators decreased to about 1/1,000th or 1/10,000th, which correspond to about 1,500 m ground distance from the hypocenter in Hiroshima. For both gold and nickel detectors, the measured activities and the calculated values agreed within 10%. The slopes of the depth-yield relations in each moderator, except granite, were similar for neutrons detected by the gold and nickel foils. From the results of these studies, the Monte Carlo neutron and photon transport code was verified to be accurate enough for use with the elements hydrogen, carbon, nitrogen, oxygen, silicon, chlorine, and cadmium, and for the incident 252Cf fission spectrum neutrons.

Determination of spatial dose distribution in UCC treatments with LDR brachytherapy using Monte Carlo methods.

PubMed

Benites-Rengifo, Jorge Luis; Vega-Carrillo, Hector Rene

2018-05-19

Using Monte Carlos methods, with the MCNP5 code, a gynecological phantom and a vaginal cylinder were modeled. The spatial distribution of absorbed dose rates in Uterine Cervical Cancer treatment through low dose rate brachytherapy was determined. A liquid water gynecology computational phantom, including a vaginal cylinder applicator made of Lucite, was designed. The applicator has a linear array of four radioactive sources of Cesium 137. Around the vaginal cylinder, 13 water spherical cells of 0.5 cm-diameter were modeled to calculate absorbed dose emulating the procedure made by the treatment planning system. The gamma-ray fluence distribution was estimated, as well as the absorbed doses resulting approximately symmetrical for cells located at upper and lower of vaginal cylinder. Obtained results allow the use of the radioactive decay law to determine dose rate for Uterine Cervical Cancer using low dose rate brachytherapy. Copyright © 2018 Elsevier Ltd. All rights reserved.

Theoretical modeling of a portable x-ray tube based KXRF system to measure lead in bone

PubMed Central

Specht, Aaron J; Weisskopf, Marc G; Nie, Linda Huiling

2017-01-01

Objective K-shell x-ray fluorescence (KXRF) techniques have been used to identify health effects resulting from exposure to metals for decades, but the equipment is bulky and requires significant maintenance and licensing procedures. A portable x-ray fluorescence (XRF) device was developed to overcome these disadvantages, but introduced a measurement dependency on soft tissue thickness. With recent advances to detector technology, an XRF device utilizing the advantages of both systems should be feasible. Approach In this study, we used Monte Carlo simulations to test the feasibility of an XRF device with a high-energy x-ray tube and detector operable at room temperature. Main Results We first validated the use of Monte Carlo N-particle transport code (MCNP) for x-ray tube simulations, and found good agreement between experimental and simulated results. Then, we optimized x-ray tube settings and found the detection limit of the high-energy x-ray tube based XRF device for bone lead measurements to be 6.91 μg g−1 bone mineral using a cadmium zinc telluride detector. Significance In conclusion, this study validated the use of MCNP in simulations of x-ray tube physics and XRF applications, and demonstrated the feasibility of a high-energy x-ray tube based XRF for metal exposure assessment. PMID:28169835

Theoretical modeling of a portable x-ray tube based KXRF system to measure lead in bone.

PubMed

Specht, Aaron J; Weisskopf, Marc G; Nie, Linda Huiling

2017-03-01

K-shell x-ray fluorescence (KXRF) techniques have been used to identify health effects resulting from exposure to metals for decades, but the equipment is bulky and requires significant maintenance and licensing procedures. A portable x-ray fluorescence (XRF) device was developed to overcome these disadvantages, but introduced a measurement dependency on soft tissue thickness. With recent advances to detector technology, an XRF device utilizing the advantages of both systems should be feasible. In this study, we used Monte Carlo simulations to test the feasibility of an XRF device with a high-energy x-ray tube and detector operable at room temperature. We first validated the use of Monte Carlo N-particle transport code (MCNP) for x-ray tube simulations, and found good agreement between experimental and simulated results. Then, we optimized x-ray tube settings and found the detection limit of the high-energy x-ray tube based XRF device for bone lead measurements to be 6.91 µg g -1 bone mineral using a cadmium zinc telluride detector. In conclusion, this study validated the use of MCNP in simulations of x-ray tube physics and XRF applications, and demonstrated the feasibility of a high-energy x-ray tube based XRF for metal exposure assessment.

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

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

Um, Dallsun

For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

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

Characterization and Performance Evaluation of an HPXe Detector for Nuclear Explosion Monitoring Applications

DTIC Science & Technology

2007-09-01

performance of the detector, and to compare the performance with sodium iodide and germanium detectors. Monte Carlo ( MCNP ) simulation was used to...aluminum ~50% more efficient), and to estimate optimum shield dimensions for an HPXe based nuclear explosion monitor. MCNP modeling was also used to...detector were calculated with MCNP by using input activity levels as measured in routine NEM runs at Pacific Northwest National Laboratory (PNNL

Visualization of nuclear particle trajectories in nuclear oil-well logging

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

Case, C.R.; Chiaramonte, J.M.

Nuclear oil-well logging measures specific properties of subsurface geological formations as a function of depth in the well. The knowledge gained is used to evaluate the hydrocarbon potential of the surrounding oil field. The measurements are made by lowering an instrument package into an oil well and slowly extracting it at a constant speed. During the extraction phase, neutrons or gamma rays are emitted from the tool, interact with the formation, and scatter back to the detectors located within the tool. Even though only a small percentage of the emitted particles ever reach the detectors, mathematical modeling has been verymore » successful in the accurate prediction of these detector responses. The two dominant methods used to model these devices have been the two-dimensional discrete ordinates method and the three-dimensional Monte Carlo method has routinely been used to investigate the response characteristics of nuclear tools. A special Los Alamos National Laboratory version of their standard MCNP Monte carlo code retains the details of each particle history of later viewing within SABRINA, a companion three-dimensional geometry modeling and debugging code.« less

A Pulsed Sphere Tutorial

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

Cullen, Dermott E.

2017-01-30

Here I attempt to explain what physically happens when we pulse an object with neutrons, specifically what we expect the time dependent behavior of the neutron population to look like. Emphasis is on the time dependent emission of both prompt and delayed neutrons. I also describe how the TART Monte Carlo transport code models this situation; see the appendix for a complete description of the model used by TART. I will also show that, as we expect, MCNP and MERCURY, produce similar results using the same delayed neutron model (again, see the appendix).

MCNP modelling of the wall effects observed in tissue-equivalent proportional counters.

PubMed

Hoff, J L; Townsend, L W

2002-01-01

Tissue-equivalent proportional counters (TEPCs) utilise tissue-equivalent materials to depict homogeneous microscopic volumes of human tissue. Although both the walls and gas simulate the same medium, they respond to radiation differently. Density differences between the two materials cause distortions, or wall effects, in measurements, with the most dominant effect caused by delta rays. This study uses a Monte Carlo transport code, MCNP, to simulate the transport of secondary electrons within a TEPC. The Rudd model, a singly differential cross section with no dependence on electron direction, is used to describe the energy spectrum obtained by the impact of two iron beams on water. Based on the models used in this study, a wall-less TEPC had a higher lineal energy (keV.micron-1) as a function of impact parameter than a solid-wall TEPC for the iron beams under consideration. An important conclusion of this study is that MCNP has the ability to model the wall effects observed in TEPCs.

An Assessment of the Detection of Highly Enriched Uranium and its Use in an Improvised Nuclear Device using the Monte Carlo Computer Code MCNP-5

NASA Astrophysics Data System (ADS)

Cochran, Thomas

2007-04-01

In 2002 and again in 2003, an investigative journalist unit at ABC News transported a 6.8 kilogram metallic slug of depleted uranium (DU) via shipping container from Istanbul, Turkey to Brooklyn, NY and from Jakarta, Indonesia to Long Beach, CA. Targeted inspection of these shipping containers by Department of Homeland Security (DHS) personnel, included the use of gamma-ray imaging, portal monitors and hand-held radiation detectors, did not uncover the hidden DU. Monte Carlo analysis of the gamma-ray intensity and spectrum of a DU slug and one consisting of highly-enriched uranium (HEU) showed that DU was a proper surrogate for testing the ability of DHS to detect the illicit transport of HEU. Our analysis using MCNP-5 illustrated the ease of fully shielding an HEU sample to avoid detection. The assembly of an Improvised Nuclear Device (IND) -- a crude atomic bomb -- from sub-critical pieces of HEU metal was then examined via Monte Carlo criticality calculations. Nuclear explosive yields of such an IND as a function of the speed of assembly of the sub-critical HEU components were derived. A comparison was made between the more rapid assembly of sub-critical pieces of HEU in the ``Little Boy'' (Hiroshima) weapon's gun barrel and gravity assembly (i.e., dropping one sub-critical piece of HEU on another from a specified height). Based on the difficulty of detection of HEU and the straightforward construction of an IND utilizing HEU, current U.S. government policy must be modified to more urgently prioritize elimination of and securing the global inventories of HEU.

Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

NASA Astrophysics Data System (ADS)

Gonzales, Matthew Alejandro

The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research version of MCNP6. Temperature feedback results from the cross sections themselves, changes in the probability density functions, as well as changes in the density of the materials. The focus of this work is specific to the Doppler temperature feedback which result from Doppler broadening of cross sections as well as changes in the probability density function within the scattering kernel. This method is compared against published results using Mosteller's numerical benchmark to show accurate evaluations of the Doppler temperature coefficient, fuel assembly calculations, and a benchmark solution based on the heavy gas model for free-gas elastic scattering. An infinite medium benchmark for neutron free gas elastic scattering for large scattering ratios and constant absorption cross section has been developed using the heavy gas model. An exact closed form solution for the neutron energy spectrum is obtained in terms of the confluent hypergeometric function and compared against spectra for the free gas scattering model in MCNP6. Results show a quick increase in convergence of the analytic energy spectrum to the MCNP6 code with increasing target size, showing absolute relative differences of less than 5% for neutrons scattering with carbon. The analytic solution has been generalized to accommodate piecewise constant in energy absorption cross section to produce temperature feedback. Results reinforce the constraints in which heavy gas theory may be applied resulting in a significant target size to accommodate increasing cross section structure. The energy dependent piecewise constant cross section heavy gas model was used to produce a benchmark calculation of the Doppler temperature coefficient to show accurate calculations when using the adjoint-weighted method. Results show the Doppler temperature coefficient using adjoint weighting and cross section derivatives accurately obtains the correct solution within statistics as well as reduce computer runtimes by a factor of 50.

Hybrid parallel code acceleration methods in full-core reactor physics calculations

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

Courau, T.; Plagne, L.; Ponicot, A.

2012-07-01

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadraturemore » required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)« less

Validation of the analytical methods in the LWR code BOXER for gadolinium-loaded fuel pins

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

Paratte, J.M.; Arkuszewski, J.J.; Kamboj, B.K.

1990-01-01

Due to the very high absorption occurring in gadolinium-loaded fuel pins, calculations of lattices with such pins present are a demanding test of the analysis methods in light water reactor (LWR) cell and assembly codes. Considerable effort has, therefore, been devoted to the validation of code methods for gadolinia fuel. The goal of the work reported in this paper is to check the analysis methods in the LWR cell/assembly code BOXER and its associated cross-section processing code ETOBOX, by comparison of BOXER results with those from a very accurate Monte Carlo calculation for a gadolinium benchmark problem. Initial results ofmore » such a comparison have been previously reported. However, the Monte Carlo calculations, done with the MCNP code, were performed at Los Alamos National Laboratory using ENDF/B-V data, while the BOXER calculations were performed at the Paul Scherrer Institute using JEF-1 nuclear data. This difference in the basic nuclear data used for the two calculations, caused by the restricted nature of these evaluated data files, led to associated uncertainties in a comparison of the results for methods validation. In the joint investigations at the Georgia Institute of Technology and PSI, such uncertainty in this comparison was eliminated by using ENDF/B-V data for BOXER calculations at Georgia Tech.« less

A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations

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

Haeck, Wim; Parsons, Donald Kent; White, Morgan Curtis

Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in themore » details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.« less

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

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

Benchmark of neutron production cross sections with Monte Carlo codes

NASA Astrophysics Data System (ADS)

Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

2018-02-01

Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first stage of a nucleus-nucleus collision also affects the low-energy neutron production. Thus, in this case, a proper combination of two physics models is desired to reproduce the measured results. In addition, code users should be aware that certain models consistently produce secondary neutrons within a constant fraction of another model in certain energy regions, which might be correlated to different physics treatments in different models.

Simulation of prompt gamma-ray emission during proton radiotherapy.

PubMed

Verburg, Joost M; Shih, Helen A; Seco, Joao

2012-09-07

The measurement of prompt gamma rays emitted from proton-induced nuclear reactions has been proposed as a method to verify in vivo the range of a clinical proton radiotherapy beam. A good understanding of the prompt gamma-ray emission during proton therapy is key to develop a clinically feasible technique, as it can facilitate accurate simulations and uncertainty analysis of gamma detector designs. Also, the gamma production cross-sections may be incorporated as prior knowledge in the reconstruction of the proton range from the measurements. In this work, we performed simulations of proton-induced nuclear reactions with the main elements of human tissue, carbon-12, oxygen-16 and nitrogen-14, using the nuclear reaction models of the GEANT4 and MCNP6 Monte Carlo codes and the dedicated nuclear reaction codes TALYS and EMPIRE. For each code, we made an effort to optimize the input parameters and model selection. The results of the models were compared to available experimental data of discrete gamma line cross-sections. Overall, the dedicated nuclear reaction codes reproduced the experimental data more consistently, while the Monte Carlo codes showed larger discrepancies for a number of gamma lines. The model differences lead to a variation of the total gamma production near the end of the proton range by a factor of about 2. These results indicate a need for additional theoretical and experimental study of proton-induced gamma emission in human tissue.

SU-F-T-657: In-Room Neutron Dose From High Energy Photon Beams

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

Christ, D; Ding, G

Purpose: To estimate neutron dose inside the treatment room from photodisintegration events in high energy photon beams using Monte Carlo simulations and experimental measurements. Methods: The Monte Carlo code MCNP6 was used for the simulations. An Eberline ESP-1 Smart Portable Neutron Detector was used to measure neutron dose. A water phantom was centered at isocenter on the treatment couch, and the detector was placed near the phantom. A Varian 2100EX linear accelerator delivered an 18MV open field photon beam to the phantom at 400MU/min, and a camera captured the detector readings. The experimental setup was modeled in the Monte Carlomore » simulation. The source was modeled for two extreme cases: a) hemispherical photon source emitting from the target and b) cone source with an angle of the primary collimator cone. The model includes the target, primary collimator, flattening filter, secondary collimators, water phantom, detector and concrete walls. Energy deposition tallies were measured for neutrons in the detector and for photons at the center of the phantom. Results: For an 18MV beam with an open 10cm by 10cm field and the gantry at 180°, the Monte Carlo simulations predict the neutron dose in the detector to be 0.11% of the photon dose in the water phantom for case a) and 0.01% for case b). The measured neutron dose is 0.04% of the photon dose. Considering the range of neutron dose predicted by Monte Carlo simulations, the calculated results are in good agreement with measurements. Conclusion: We calculated in-room neutron dose by using Monte Carlo techniques, and the predicted neutron dose is confirmed by experimental measurements. If we remodel the source as an electron beam hitting the target for a more accurate representation of the bremsstrahlung fluence, it is feasible that the Monte Carlo simulations can be used to help in shielding designs.« less

Development and application of a hybrid transport methodology for active interrogation systems

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

Royston, K.; Walters, W.; Haghighat, A.

A hybrid Monte Carlo and deterministic methodology has been developed for application to active interrogation systems. The methodology consists of four steps: i) neutron flux distribution due to neutron source transport and subcritical multiplication; ii) generation of gamma source distribution from (n, 7) interactions; iii) determination of gamma current at a detector window; iv) detection of gammas by the detector. This paper discusses the theory and results of the first three steps for the case of a cargo container with a sphere of HEU in third-density water cargo. To complete the first step, a response-function formulation has been developed tomore » calculate the subcritical multiplication and neutron flux distribution. Response coefficients are pre-calculated using the MCNP5 Monte Carlo code. The second step uses the calculated neutron flux distribution and Bugle-96 (n, 7) cross sections to find the resulting gamma source distribution. In the third step the gamma source distribution is coupled with a pre-calculated adjoint function to determine the gamma current at a detector window. The AIMS (Active Interrogation for Monitoring Special-Nuclear-Materials) software has been written to output the gamma current for a source-detector assembly scanning across a cargo container using the pre-calculated values and taking significantly less time than a reference MCNP5 calculation. (authors)« less

A Deep Penetration Problem Calculation Using AETIUS:An Easy Modeling Discrete Ordinates Transport Code UsIng Unstructured Tetrahedral Mesh, Shared Memory Parallel

NASA Astrophysics Data System (ADS)

KIM, Jong Woon; LEE, Young-Ouk

2017-09-01

As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.

Calibration of a portable HPGe detector using MCNP code for the determination of 137Cs in soils.

PubMed

Gutiérrez-Villanueva, J L; Martín-Martín, A; Peña, V; Iniguez, M P; de Celis, B; de la Fuente, R

2008-10-01

In situ gamma spectrometry provides a fast method to determine (137)Cs inventories in soils. To improve the accuracy of the estimates, one can use not only the information on the photopeak count rates but also on the peak to forward-scatter ratios. Before applying this procedure to field measurements, a calibration including several experimental simulations must be carried out in the laboratory. In this paper it is shown that Monte Carlo methods are a valuable tool to minimize the number of experimental measurements needed for the calibration.

Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

NASA Astrophysics Data System (ADS)

Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

2017-09-01

Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

Dosimetry of gamma chamber blood irradiator using PAGAT gel dosimeter and Monte Carlo simulations

PubMed Central

Mohammadyari, Parvin; Zehtabian, Mehdi; Sina, Sedigheh; Tavasoli, Ali Reza

2014-01-01

Currently, the use of blood irradiation for inactivating pathogenic microbes in infected blood products and preventing graft‐versus‐host disease (GVHD) in immune suppressed patients is greater than ever before. In these systems, dose distribution and uniformity are two important concepts that should be checked. In this study, dosimetry of the gamma chamber blood irradiator model Gammacell 3000 Elan was performed by several dosimeter methods including thermoluminescence dosimeters (TLD), PAGAT gel dosimetry, and Monte Carlo simulations using MCNP4C code. The gel dosimeter was put inside a glass phantom and the TL dosimeters were placed on its surface, and the phantom was then irradiated for 5 min and 27 sec. The dose values at each point inside the vials were obtained from the magnetic resonance imaging of the phantom. For Monte Carlo simulations, all components of the irradiator were simulated and the dose values in a fine cubical lattice were calculated using tally F6. This study shows that PAGAT gel dosimetry results are in close agreement with the results of TL dosimetry, Monte Carlo simulations, and the results given by the vendor, and the percentage difference between the different methods is less than 4% at different points inside the phantom. According to the results obtained in this study, PAGAT gel dosimetry is a reliable method for dosimetry of the blood irradiator. The major advantage of this kind of dosimetry is that it is capable of 3D dose calculation. PACS number: 87.53.Bn PMID:24423829

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

2010-01-01

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(10{sup 2-4}), 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

Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

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

Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

2007-04-15

X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement.more » Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities.« less

Investigating the response of Micromegas detector to low-energy neutrons using Monte Carlo simulation

NASA Astrophysics Data System (ADS)

Khezripour, S.; Negarestani, A.; Rezaie, M. R.

2017-08-01

Micromegas detector has recently been used for high-energy neutron (HEN) detection, but the aim of this research is to investigate the response of the Micromegas detector to low-energy neutron (LEN). For this purpose, a Micromegas detector (with air, P10, BF3, 3He and Ar/BF3 mixture) was optimized for the detection of 60 keV neutrons using the MCNP (Monte Carlo N Particle) code. The simulation results show that the optimum thickness of the cathode is 1 mm and the optimum of microgrid location is 100 μm above the anode. The output current of this detector for Ar (3%) + BF3 (97%) mixture is greater than the other ones. This mixture is considered as the appropriate gas for the Micromegas neutron detector providing the output current for 60 keV neutrons at the level of 97.8 nA per neutron. Consecuently, this detector can be introduced as LEN detector.

Monte Carlo calculation of energy deposition in ionization chambers for tritium measurements

NASA Astrophysics Data System (ADS)

Zhilin, Chen; Shuming, Peng; Dan, Meng; Yuehong, He; Heyi, Wang

2014-10-01

Energy deposition in ionization chambers for tritium measurements has been theoretically studied using Monte Carlo code MCNP 5. The influence of many factors, including carrier gas, chamber size, wall materials and gas pressure, has been evaluated in the simulations. It is found that β rays emitted by tritium deposit much more energy into chambers flowing through with argon than with deuterium in them, as much as 2.7 times higher at pressure 100 Pa. As chamber size gets smaller, energy deposition decreases sharply. For an ionization chamber of 1 mL, β rays deposit less than 1% of their energy at pressure 100 Pa and only 84% even if gas pressure is as high as 100 kPa. It also indicates that gold plated ionization chamber results in the highest deposition ratio while aluminum one leads to the lowest. In addition, simulations were validated by comparison with experimental data. Results show that simulations agree well with experimental data.

Dosimetric characterization of the M−15 high‐dose‐rate Iridium−192 brachytherapy source using the AAPM and ESTRO formalism

PubMed Central

Thanh, Minh‐Tri Ho; Munro, John J.

2015-01-01

The Source Production & Equipment Co. (SPEC) model M−15 is a new Iridium−192 brachytherapy source model intended for use as a temporary high‐dose‐rate (HDR) brachytherapy source for the Nucletron microSelectron Classic afterloading system. The purpose of this study is to characterize this HDR source for clinical application by obtaining a complete set of Monte Carlo calculated dosimetric parameters for the M‐15, as recommended by AAPM and ESTRO, for isotopes with average energies greater than 50 keV. This was accomplished by using the MCNP6 Monte Carlo code to simulate the resulting source dosimetry at various points within a pseudoinfinite water phantom. These dosimetric values next were converted into the AAPM and ESTRO dosimetry parameters and the respective statistical uncertainty in each parameter also calculated and presented. The M−15 source was modeled in an MCNP6 Monte Carlo environment using the physical source specifications provided by the manufacturer. Iridium−192 photons were uniformly generated inside the iridium core of the model M−15 with photon and secondary electron transport replicated using photoatomic cross‐sectional tables supplied with MCNP6. Simulations were performed for both water and air/vacuum computer models with a total of 4×109 sources photon history for each simulation and the in‐air photon spectrum filtered to remove low‐energy photons below δ=10%keV. Dosimetric data, including D(r,θ),gL(r),F(r,θ),Φan(r), and φ¯an, and their statistical uncertainty were calculated from the output of an MCNP model consisting of an M−15 source placed at the center of a spherical water phantom of 100 cm diameter. The air kerma strength in free space, SK, and dose rate constant, Λ, also was computed from a MCNP model with M−15 Iridium−192 source, was centered at the origin of an evacuated phantom in which a critical volume containing air at STP was added 100 cm from the source center. The reference dose rate, D˙(r0,θ0)≡D˙(1cm,π/2), is found to be 4.038±0.064 cGy mCi−1 h−1. The air kerma strength, SK, is reported to be 3.632±0.086 cGy cm2 mCi−1 g−1, and the dose rate constant, Λ, is calculated to be 1.112±0.029 cGy h−1 U−1. The normalized dose rate, radial dose function, and anisotropy function with their uncertainties were computed and are represented in both tabular and graphical format in the report. A dosimetric study was performed of the new M−15 Iridium−192 HDR brachytherapy source using the MCNP6 radiation transport code. Dosimetric parameters, including the dose‐rate constant, radial dose function, and anisotropy function, were calculated in accordance with the updated AAPM and ESTRO dosimetric parameters for brachytherapy sources of average energy greater than 50 keV. These data therefore may be applied toward the development of a treatment planning program and for clinical use of the source. PACS numbers: 87.56.bg, 87.53.Jw PMID:26103489

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Cellular dosimetry of (111)In using monte carlo N-particle computer code: comparison with analytic methods and correlation with in vitro cytotoxicity.

PubMed

Cai, Zhongli; Pignol, Jean-Philippe; Chan, Conrad; Reilly, Raymond M

2010-03-01

Our objective was to compare Monte Carlo N-particle (MCNP) self- and cross-doses from (111)In to the nucleus of breast cancer cells with doses calculated by reported analytic methods (Goddu et al. and Farragi et al.). A further objective was to determine whether the MCNP-predicted surviving fraction (SF) of breast cancer cells exposed in vitro to (111)In-labeled diethylenetriaminepentaacetic acid human epidermal growth factor ((111)In-DTPA-hEGF) could accurately predict the experimentally determined values. MCNP was used to simulate the transport of electrons emitted by (111)In from the cell surface, cytoplasm, or nucleus. The doses to the nucleus per decay (S values) were calculated for single cells, closely packed monolayer cells, or cell clusters. The cell and nucleus dimensions of 6 breast cancer cell lines were measured, and cell line-specific S values were calculated. For self-doses, MCNP S values of nucleus to nucleus agreed very well with those of Goddu et al. (ratio of S values using analytic methods vs. MCNP = 0.962-0.995) and Faraggi et al. (ratio = 1.011-1.024). MCNP S values of cytoplasm and cell surface to nucleus compared fairly well with the reported values (ratio = 0.662-1.534 for Goddu et al.; 0.944-1.129 for Faraggi et al.). For cross doses, the S values to the nucleus were independent of (111)In subcellular distribution but increased with cluster size. S values for monolayer cells were significantly different from those of single cells and cell clusters. The MCNP-predicted SF for monolayer MDA-MB-468, MDA-MB-231, and MCF-7 cells agreed with the experimental data (relative error of 3.1%, -1.0%, and 1.7%). The single-cell and cell cluster models were less accurate in predicting the SF. For MDA-MB-468 cells, relative error was 8.1% using the single-cell model and -54% to -67% using the cell cluster model. Individual cell-line dimensions had large effects on S values and were needed to estimate doses and SF accurately. MCNP simulation compared well with the reported analytic methods in the calculation of subcellular S values for single cells and cell clusters. Application of a monolayer model was most accurate in predicting the SF of breast cancer cells exposed in vitro to (111)In-DTPA-hEGF.

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

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

On the development of radiation tolerant surveillance camera from consumer-grade components

NASA Astrophysics Data System (ADS)

Klemen, Ambrožič; Luka, Snoj; Lars, Öhlin; Jan, Gunnarsson; Niklas, Barringer

2017-09-01

In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

Conversion coefficients for determination of dispersed photon dose during radiotherapy: NRUrad input code for MCNP.

PubMed

Shahmohammadi Beni, Mehrdad; Ng, C Y P; Krstic, D; Nikezic, D; Yu, K N

2017-01-01

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient's body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5.

Conversion coefficients for determination of dispersed photon dose during radiotherapy: NRUrad input code for MCNP

PubMed Central

Krstic, D.; Nikezic, D.

2017-01-01

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient’s body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5. PMID:28362837

Gamma-ray dose from an overhead plume

DOE PAGES

McNaughton, Michael W.; Gillis, Jessica McDonnel; Ruedig, Elizabeth; ...

2017-05-01

Standard plume models can underestimate the gamma-ray dose when most of the radioactive material is above the heads of the receptors. Typically, a model is used to calculate the air concentration at the height of the receptor, and the dose is calculated by multiplying the air concentration by a concentration-to-dose conversion factor. Models indicate that if the plume is emitted from a stack during stable atmospheric conditions, the lower edges of the plume may not reach the ground, in which case both the ground-level concentration and the dose are usually reported as zero. However, in such cases, the dose frommore » overhead gamma-emitting radionuclides may be substantial. Such underestimates could impact decision making in emergency situations. The Monte Carlo N-Particle code, MCNP, was used to calculate the overhead shine dose and to compare with standard plume models. At long distances and during unstable atmospheric conditions, the MCNP results agree with the standard models. As a result, at short distances, where many models calculate zero, the true dose (as modeled by MCNP) can be estimated with simple equations.« less

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

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

Du, T. F.; Chen, Z. J.; Peng, X. Y.

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometermore » at EAST are studied for future data interpretation.« less

Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

PubMed

Mohammadi, A; Hassanzadeh, M; Gharib, M

2016-02-01

In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. Copyright © 2015 Elsevier Ltd. All rights reserved.

Monte Carlo simulation of energy-dispersive x-ray fluorescence and applications

NASA Astrophysics Data System (ADS)

Li, Fusheng

Four key components with regards to Monte Carlo Library Least Squares (MCLLS) have been developed by the author. These include: a comprehensive and accurate Monte Carlo simulation code - CEARXRF5 with Differential Operators (DO) and coincidence sampling, Detector Response Function (DRF), an integrated Monte Carlo - Library Least-Squares (MCLLS) Graphical User Interface (GUI) visualization System (MCLLSPro) and a new reproducible and flexible benchmark experiment setup. All these developments or upgrades enable the MCLLS approach to be a useful and powerful tool for a tremendous variety of elemental analysis applications. CEARXRF, a comprehensive and accurate Monte Carlo code for simulating the total and individual library spectral responses of all elements, has been recently upgraded to version 5 by the author. The new version has several key improvements: input file format fully compatible with MCNP5, a new efficient general geometry tracking code, versatile source definitions, various variance reduction techniques (e.g. weight window mesh and splitting, stratifying sampling, etc.), a new cross section data storage and accessing method which improves the simulation speed by a factor of four and new cross section data, upgraded differential operators (DO) calculation capability, and also an updated coincidence sampling scheme which including K-L and L-L coincidence X-Rays, while keeping all the capabilities of the previous version. The new Differential Operators method is powerful for measurement sensitivity study and system optimization. For our Monte Carlo EDXRF elemental analysis system, it becomes an important technique for quantifying the matrix effect in near real time when combined with the MCLLS approach. An integrated visualization GUI system has been developed by the author to perform elemental analysis using iterated Library Least-Squares method for various samples when an initial guess is provided. This software was built on the Borland C++ Builder platform and has a user-friendly interface to accomplish all qualitative and quantitative tasks easily. That is to say, the software enables users to run the forward Monte Carlo simulation (if necessary) or use previously calculated Monte Carlo library spectra to obtain the sample elemental composition estimation within a minute. The GUI software is easy to use with user-friendly features and has the capability to accomplish all related tasks in a visualization environment. It can be a powerful tool for EDXRF analysts. A reproducible experiment setup has been built and experiments have been performed to benchmark the system. Two types of Standard Reference Materials (SRM), stainless steel samples from National Institute of Standards and Technology (NIST) and aluminum alloy samples from Alcoa Inc., with certified elemental compositions, are tested with this reproducible prototype system using a 109Cd radioisotope source (20mCi) and a liquid nitrogen cooled Si(Li) detector. The results show excellent agreement between the calculated sample compositions and their reference values and the approach is very fast.

Use of the FLUKA Monte Carlo code for 3D patient-specific dosimetry on PET-CT and SPECT-CT images*

PubMed Central

Botta, F; Mairani, A; Hobbs, R F; Vergara Gil, A; Pacilio, M; Parodi, K; Cremonesi, M; Coca Pérez, M A; Di Dia, A; Ferrari, M; Guerriero, F; Battistoni, G; Pedroli, G; Paganelli, G; Torres Aroche, L A; Sgouros, G

2014-01-01

Patient-specific absorbed dose calculation for nuclear medicine therapy is a topic of increasing interest. 3D dosimetry at the voxel level is one of the major improvements for the development of more accurate calculation techniques, as compared to the standard dosimetry at the organ level. This study aims to use the FLUKA Monte Carlo code to perform patient-specific 3D dosimetry through direct Monte Carlo simulation on PET-CT and SPECT-CT images. To this aim, dedicated routines were developed in the FLUKA environment. Two sets of simulations were performed on model and phantom images. Firstly, the correct handling of PET and SPECT images was tested under the assumption of homogeneous water medium by comparing FLUKA results with those obtained with the voxel kernel convolution method and with other Monte Carlo-based tools developed to the same purpose (the EGS-based 3D-RD software and the MCNP5-based MCID). Afterwards, the correct integration of the PET/SPECT and CT information was tested, performing direct simulations on PET/CT images for both homogeneous (water) and non-homogeneous (water with air, lung and bone inserts) phantoms. Comparison was performed with the other Monte Carlo tools performing direct simulation as well. The absorbed dose maps were compared at the voxel level. In the case of homogeneous water, by simulating 108 primary particles a 2% average difference with respect to the kernel convolution method was achieved; such difference was lower than the statistical uncertainty affecting the FLUKA results. The agreement with the other tools was within 3–4%, partially ascribable to the differences among the simulation algorithms. Including the CT-based density map, the average difference was always within 4% irrespective of the medium (water, air, bone), except for a maximum 6% value when comparing FLUKA and 3D-RD in air. The results confirmed that the routines were properly developed, opening the way for the use of FLUKA for patient-specific, image-based dosimetry in nuclear medicine. PMID:24200697

Analysis of activation and shutdown contact dose rate for EAST neutral beam port

NASA Astrophysics Data System (ADS)

Chen, Yuqing; Wang, Ji; Zhong, Guoqiang; Li, Jun; Wang, Jinfang; Xie, Yahong; Wu, Bin; Hu, Chundong

2017-12-01

For the safe operation and maintenance of neutral beam injector (NBI), specific activity and shutdown contact dose rate of the sample material SS316 are estimated around the experimental advanced superconducting tokamak (EAST) neutral beam port. Firstly, the neutron emission intensity is calculated by TRANSP code while the neutral beam is co-injected to EAST. Secondly, the neutron activation and shutdown contact dose rates for the neutral beam sample materials SS316 are derived by the Monte Carlo code MCNP and the inventory code FISPACT-2007. The simulations indicate that the primary radioactive nuclides of SS316 are 58Co and 54Mn. The peak contact dose rate is 8.52 × 10-6 Sv/h after EAST shutdown one second. That is under the International Thermonuclear Experimental Reactor (ITER) design values 1 × 10-5 Sv/h.

Monte Carlo simulation for Neptun 10 PC medical linear accelerator and calculations of output factor for electron beam

PubMed Central

Bahreyni Toossi, Mohammad Taghi; Momennezhad, Mehdi; Hashemi, Seyed Mohammad

2012-01-01

Aim Exact knowledge of dosimetric parameters is an essential pre-requisite of an effective treatment in radiotherapy. In order to fulfill this consideration, different techniques have been used, one of which is Monte Carlo simulation. Materials and methods This study used the MCNP-4Cb to simulate electron beams from Neptun 10 PC medical linear accelerator. Output factors for 6, 8 and 10 MeV electrons applied to eleven different conventional fields were both measured and calculated. Results The measurements were carried out by a Wellhofler-Scanditronix dose scanning system. Our findings revealed that output factors acquired by MCNP-4C simulation and the corresponding values obtained by direct measurements are in a very good agreement. Conclusion In general, very good consistency of simulated and measured results is a good proof that the goal of this work has been accomplished. PMID:24377010

Neutron field measurement at the Experimental Advanced Superconducting Tokamak using a Bonner sphere spectrometer

NASA Astrophysics Data System (ADS)

Hu, Zhimeng; Zhong, Guoqiang; Ge, Lijian; Du, Tengfei; Peng, Xingyu; Chen, Zhongjing; Xie, Xufei; Yuan, Xi; Zhang, Yimo; Sun, Jiaqi; Fan, Tieshuan; Zhou, Ruijie; Xiao, Min; Li, Kai; Hu, Liqun; Chen, Jun; Zhang, Hui; Gorini, Giuseppe; Nocente, Massimo; Tardocchi, Marco; Li, Xiangqing; Chen, Jinxiang; Zhang, Guohui

2018-07-01

The neutron field measurement was performed in the Experimental Advanced Superconducting Tokamak (EAST) experimental hall using a Bonner sphere spectrometer (BSS) based on a 3He thermal neutron counter. The measured spectra and the corresponding integrated neutron fluence and dose values deduced from the spectra at two exposed positions were compared to the calculated results obtained by a general Monte Carlo code MCNP5, and good agreements were found. The applicability of a homemade dose survey meter installed at EAST was also verified with the comparison of the ambient dose equivalent H*(10) values measured by the meter and BSS.

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

Rezaei-Ochbelagh, D.; Salman-Nezhad, S.; Asadi, A.

External photon beam radiotherapy is carried out in a way to achieve an 'as low as possible' a dose in healthy tissues surrounding the target. One of these surroundings can be heart as a vital organ of body. As it is impossible to directly determine the absorbed dose by heart, using phantoms is one way to acquire information around it. The other way is Monte Carlo method. In this work we have presented a simulation of heart geometry by introducing of different surfaces in MCNP code. We used 14 surface equations in order to determine human heart modeling. Those surfacesmore » are borders of heart walls and contents.« less

Multigroup cross section library for GFR2400

NASA Astrophysics Data System (ADS)

Čerba, Štefan; Vrban, Branislav; Lüley, Jakub; Haščík, Ján; Nečas, Vladimír

2017-09-01

In this paper the development and optimization of the SBJ_E71 multigroup cross section library for GFR2400 applications is discussed. A cross section processing scheme, merging Monte Carlo and deterministic codes, was developed. Several fine and coarse group structures and two weighting flux options were analysed through 18 benchmark experiments selected from the handbook of ICSBEP and based on performed similarity assessments. The performance of the collapsed version of the SBJ_E71 library was compared with MCNP5 CE ENDF/B VII.1 and the Korean KAFAX-E70 library. The comparison was made based on integral parameters of calculations performed on full core homogenous models.

SABRINA - an interactive geometry modeler for MCNP

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

West, J.T.; Murphy, J.

One of the most difficult tasks when analyzing a complex three-dimensional system with Monte Carlo is geometry model development. SABRINA attempts to make the modeling process more user-friendly and less of an obstacle. It accepts both combinatorial solid bodies and MCNP surfaces and produces MCNP cells. The model development process in SABRINA is highly interactive and gives the user immediate feedback on errors. Users can view their geometry from arbitrary perspectives while the model is under development and interactively find and correct modeling errors. An example of a SABRINA display is shown. It represents a complex three-dimensional shape.

A comparison of skyshine computational methods.

PubMed

Hertel, Nolan E; Sweezy, Jeremy E; Shultis, J Kenneth; Warkentin, J Karl; Rose, Zachary J

2005-01-01

A variety of methods employing radiation transport and point-kernel codes have been used to model two skyshine problems. The first problem is a 1 MeV point source of photons on the surface of the earth inside a 2 m tall and 1 m radius silo having black walls. The skyshine radiation downfield from the point source was estimated with and without a 30-cm-thick concrete lid on the silo. The second benchmark problem is to estimate the skyshine radiation downfield from 12 cylindrical canisters emplaced in a low-level radioactive waste trench. The canisters are filled with ion-exchange resin with a representative radionuclide loading, largely 60Co, 134Cs and 137Cs. The solution methods include use of the MCNP code to solve the problem by directly employing variance reduction techniques, the single-scatter point kernel code GGG-GP, the QADMOD-GP point kernel code, the COHORT Monte Carlo code, the NAC International version of the SKYSHINE-III code, the KSU hybrid method and the associated KSU skyshine codes.

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

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

Connaway, H. M.; Lee, C. H.

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To supportmore » this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Whole body counter calibration using Monte Carlo modeling with an array of phantom sizes based on national anthropometric reference data

USDA-ARS?s Scientific Manuscript database

During construction of the whole body counter (WBC) at the Children’s Nutrition Research Center (CNRC), efficiency calibration was needed to translate acquired counts of 40K to actual grams of potassium for measurement of total body potassium (TBK) in a diverse subject population. The MCNP Monte Car...

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

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Monte Carlo parametric studies of neutron interrogation with the Associated Particle Technique for cargo container inspections

NASA Astrophysics Data System (ADS)

Deyglun, Clément; Carasco, Cédric; Pérot, Bertrand

2014-06-01

The detection of Special Nuclear Materials (SNM) by neutron interrogation is extensively studied by Monte Carlo simulation at the Nuclear Measurement Laboratory of CEA Cadarache (French Alternative Energies and Atomic Energy Commission). The active inspection system is based on the Associated Particle Technique (APT). Fissions induced by tagged neutrons (i.e. correlated to an alpha particle in the DT neutron generator) in SNM produce high multiplicity coincidences which are detected with fast plastic scintillators. At least three particles are detected in a short time window following the alpha detection, whereas nonnuclear materials mainly produce single events, or pairs due to (n,2n) and (n,n'γ) reactions. To study the performances of an industrial cargo container inspection system, Monte Carlo simulations are performed with the MCNP-PoliMi transport code, which records for each neutron history the relevant information: reaction types, position and time of interactions, energy deposits, secondary particles, etc. The output files are post-processed with a specific tool developed with ROOT data analysis software. Particles not correlated with an alpha particle (random background), counting statistics, and time-energy resolutions of the data acquisition system are taken into account in the numerical model. Various matrix compositions, suspicious items, SNM shielding and positions inside the container, are simulated to assess the performances and limitations of an industrial system.

Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.

PubMed

Heuel-Fabianek, Burkhard; Hille, Ralf

2005-01-01

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Multiprocessing MCNP on an IBM RS/6000 cluster

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

McKinney, G.W.; West, J.T.

1993-01-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors (P) and the fraction of task time that multiprocesses (f), can be formulated using Amdahl's Law S ((f,P) = 1 f + f/P). However, for most applications this theoretical limit cannot be achieved, due to additional terms not included in Amdahl's Law. Monte Carlo transport is a natural candidate for multiprocessing, since the particle tracks are generally independent and the precision of the result increases as the square root of the number of particles tracked.« less

Self-optimizing Monte Carlo method for nuclear well logging simulation

NASA Astrophysics Data System (ADS)

Liu, Lianyan

1997-09-01

In order to increase the efficiency of Monte Carlo simulation for nuclear well logging problems, a new method has been developed for variance reduction. With this method, an importance map is generated in the regular Monte Carlo calculation as a by-product, and the importance map is later used to conduct the splitting and Russian roulette for particle population control. By adopting a spatial mesh system, which is independent of physical geometrical configuration, the method allows superior user-friendliness. This new method is incorporated into the general purpose Monte Carlo code MCNP4A through a patch file. Two nuclear well logging problems, a neutron porosity tool and a gamma-ray lithology density tool are used to test the performance of this new method. The calculations are sped up over analog simulation by 120 and 2600 times, for the neutron porosity tool and for the gamma-ray lithology density log, respectively. The new method enjoys better performance by a factor of 4~6 times than that of MCNP's cell-based weight window, as per the converged figure-of-merits. An indirect comparison indicates that the new method also outperforms the AVATAR process for gamma-ray density tool problems. Even though it takes quite some time to generate a reasonable importance map from an analog run, a good initial map can create significant CPU time savings. This makes the method especially suitable for nuclear well logging problems, since one or several reference importance maps are usually available for a given tool. Study shows that the spatial mesh sizes should be chosen according to the mean-free-path. The overhead of the importance map generator is 6% and 14% for neutron and gamma-ray cases. The learning ability towards a correct importance map is also demonstrated. Although false-learning may happen, physical judgement can help diagnose with contributon maps. Calibration and analysis are performed for the neutron tool and the gamma-ray tool. Due to the fact that a very good initial importance map is always available after the first point has been calculated, high computing efficiency is maintained. The availability of contributon maps provides an easy way of understanding the logging measurement and analyzing for the depth of investigation.

Monte Carlo study of a 60Co calibration field of the Dosimetry Laboratory Seibersdorf.

PubMed

Hranitzky, C; Stadtmann, H

2007-01-01

The gamma radiation fields of the reference irradiation facility of the Dosimetry Laboratory Seibersdorf with collimated beam geometry are used for calibrating radiation protection dosemeters. A close-to-reality simulation model of the facility including the complex geometry of a 60Co source was set up using the Monte Carlo code MCNP. The goal of this study is to characterise the radionuclide gamma calibration field and resulting air-kerma distributions inside the measurement hall with a total of 20 m in length. For the whole range of source-detector-distances (SDD) along the central beam axis, simulated and measured relative air-kerma values are within +/-0.6%. Influences on the accuracy of the simulation results are investigated, including e.g., source mass density effects or detector volume dependencies. A constant scatter contribution from the lead ring-collimator of approximately 1% and an increasing scatter contribution from the concrete floor for distances above 7 m are identified, resulting in a total air-kerma scatter contribution below 5%, which is in accordance to the ISO 4037-1 recommendations.

Shielding analyses of an AB-BNCT facility using Monte Carlo simulations and simplified methods

NASA Astrophysics Data System (ADS)

Lai, Bo-Lun; Sheu, Rong-Jiun

2017-09-01

Accurate Monte Carlo simulations and simplified methods were used to investigate the shielding requirements of a hypothetical accelerator-based boron neutron capture therapy (AB-BNCT) facility that included an accelerator room and a patient treatment room. The epithermal neutron beam for BNCT purpose was generated by coupling a neutron production target with a specially designed beam shaping assembly (BSA), which was embedded in the partition wall between the two rooms. Neutrons were produced from a beryllium target bombarded by 1-mA 30-MeV protons. The MCNP6-generated surface sources around all the exterior surfaces of the BSA were established to facilitate repeated Monte Carlo shielding calculations. In addition, three simplified models based on a point-source line-of-sight approximation were developed and their predictions were compared with the reference Monte Carlo results. The comparison determined which model resulted in better dose estimation, forming the basis of future design activities for the first ABBNCT facility in Taiwan.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Measurement and simulation of thermal neutron flux distribution in the RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

An MCNP-based model of a medical linear accelerator x-ray photon beam.

PubMed

Ajaj, F A; Ghassal, N M

2003-09-01

The major components in the x-ray photon beam path of the treatment head of the VARIAN Clinac 2300 EX medical linear accelerator were modeled and simulated using the Monte Carlo N-Particle radiation transport computer code (MCNP). Simulated components include x-ray target, primary conical collimator, x-ray beam flattening filter and secondary collimators. X-ray photon energy spectra and angular distributions were calculated using the model. The x-ray beam emerging from the secondary collimators were scored by considering the total x-ray spectra from the target as the source of x-rays at the target position. The depth dose distribution and dose profiles at different depths and field sizes have been calculated at a nominal operating potential of 6 MV and found to be within acceptable limits. It is concluded that accurate specification of the component dimensions, composition and nominal accelerating potential gives a good assessment of the x-ray energy spectra.

Development of low level 226Ra analysis for live fish using gamma-ray spectrometry

NASA Astrophysics Data System (ADS)

Chandani, Z.; Prestwich, W. V.; Byun, S. H.

2017-06-01

A low level 226Ra analysis method for live fish was developed using a 4π NaI(Tl) gamma-ray spectrometer. In order to find out the best algorithm for accomplishing the lowest detection limit, the gamma-ray spectrum from a 226Ra point was collected and nine different methods were attempted for spectral analysis. The lowest detection limit of 0.99 Bq for an hour counting occurred when the spectrum was integrated in the energy region of 50-2520 keV. To extend 226Ra analysis to live fish, a Monte Carlo simulation model with a cylindrical fish in a water container was built using the MCNP code. From simulation results, the spatial distribution of the efficiency and the efficiency correction factor for the live fish model were determined. The MCNP model will be able to be conveniently modified when a different fish or container geometry is employed as fish grow up in real experiments.

An evaluation of a manganese bath system having a new geometry through MCNP modelling.

PubMed

Khabaz, Rahim

2012-12-01

In this study, an approximate symmetric cylindrical manganese bath system with equal diameter and height was appraised using a Monte Carlo simulation. For nine sizes of the tank filled with MnSO(4).H(2)O solution of three different concentrations, the necessary correction factors involved in the absolute measurement of neutron emission rate were determined by a detailed modelling of the MCNP4C code with the ENDF/B-VII.0 neutron cross section data library. The results obtained were also used to determine the optimum dimensions of the bath for each concentration of solution in the calibration of (241)Am-Be and (252)Cf sources. Also, the amount of gamma radiation produced as a result of (n,γ) the reaction with the nuclei of the manganese sulphate solution that escaped from the boundary of each tank was evaluated. This gamma can be important for the background in NaI(Tl) detectors and issues concerned with radiation protection.

Lung Dosimetry for Radioiodine Treatment Planning in the Case of Diffuse Lung Metastases

PubMed Central

Song, Hong; He, Bin; Prideaux, Andrew; Du, Yong; Frey, Eric; Kasecamp, Wayne; Ladenson, Paul W.; Wahl, Richard L.; Sgouros, George

2010-01-01

The lungs are the most frequent sites of distant metastasis in differentiated thyroid carcinoma. Radioiodine treatment planning for these patients is usually performed following the Benua– Leeper method, which constrains the administered activity to 2.96 GBq (80 mCi) whole-body retention at 48 h after administration to prevent lung toxicity in the presence of iodine-avid lung metastases. This limit was derived from clinical experience, and a dosimetric analysis of lung and tumor absorbed dose would be useful to understand the implications of this limit on toxicity and tumor control. Because of highly nonuniform lung density and composition as well as the nonuniform activity distribution when the lungs contain tumor nodules, Monte Carlo dosimetry is required to estimate tumor and normal lung absorbed dose. Reassessment of this toxicity limit is also appropriate in light of the contemporary use of recombinant thyrotropin (thyroid-stimulating hormone) (rTSH) to prepare patients for radioiodine therapy. In this work we demonstrated the use of MCNP, a Monte Carlo electron and photon transport code, in a 3-dimensional (3D) imaging–based absorbed dose calculation for tumor and normal lungs. Methods A pediatric thyroid cancer patient with diffuse lung metastases was administered 37MBq of 131I after preparation with rTSH. SPECT/CT scans were performed over the chest at 27, 74, and 147 h after tracer administration. The time–activity curve for 131I in the lungs was derived from the whole-body planar imaging and compared with that obtained from the quantitative SPECT methods. Reconstructed and coregistered SPECT/CT images were converted into 3D density and activity probability maps suitable for MCNP4b input. Absorbed dose maps were calculated using electron and photon transport in MCNP4b. Administered activity was estimated on the basis of the maximum tolerated dose (MTD) of 27.25 Gy to the normal lungs. Computational efficiency of the MCNP4b code was studied with a simple segmentation approach. In addition, the Benua–Leeper method was used to estimate the recommended administered activity. The standard dosing plan was modified to account for the weight of this pediatric patient, where the 2.96-GBq (80 mCi) whole-body retention was scaled to 2.44 GBq (66 mCi) to give the same dose rate of 43.6 rad/h in the lungs at 48 h. Results Using the MCNP4b code, both the spatial dose distribution and a dose–volume histogram were obtained for the lungs. An administered activity of 1.72 GBq (46.4 mCi) delivered the putative MTD of 27.25 Gy to the lungs with a tumor absorbed dose of 63.7 Gy. Directly applying the Benua–Leeper method, an administered activity of 3.89 GBq (105.0 mCi) was obtained, resulting in tumor and lung absorbed doses of 144.2 and 61.6 Gy, respectively, when the MCNP-based dosimetry was applied. The voxel-by-voxel calculation time of 4,642.3 h for photon transport was reduced to 16.8 h when the activity maps were segmented into 20 regions. Conclusion MCNP4b–based, patient-specific 3D dosimetry is feasible and important in the dosimetry of thyroid cancer patients with avid lung metastases that exhibit prolonged retention in the lungs. PMID:17138741

Praseodymium-142 glass seeds for the brachytherapy of prostate cancer

NASA Astrophysics Data System (ADS)

Jung, Jae Won

A beta-emitting glass seed was proposed for the brachytherapy treatment of prostate cancer. Criteria for seed design were derived and several beta-emitting nuclides were examined for suitability. 142Pr was selected as the isotope of choice. Seeds 0.08 cm in diameter and 0.9 cm long were manufactured for testing. The seeds were activated in the Texas A&M University research reactor. The activity produced was as expected when considering the meta-stable state and epi-thermal neutron flux. The MCNP5 Monte Carlo code was used to calculate the quantitative dosimetric parameters suggested in the American Association of Physicists in Medicine (AAPM) TG-43/60. The Monte Carlo calculation results were compared with those from a dose point kernel code. The dose profiles agree well with each other. The gamma dose of 142Pr was evaluated. The gamma dose is 0.3 Gy at 1.0 cm with initial activity of 5.95 mCi and is insignificant to other organs. Measurements were performed to assess the 2-dimensional axial dose distributions using Gafchromic radiochromic film. The radiochromic film was calibrated using an X-ray machine calibrated against a National Institute of Standards and Technology (NIST) traceable ion chamber. A calibration curve was derived using a least squares fit of a second order polynomial. The measured dose distribution agrees well with results from the Monte Carlo simulation. The dose was 130.8 Gy at 6 mm from the seed center with initial activity of 5.95 mCi. AAPM TG-43/60 parameters were determined. The reference dose rate for 2 mm and 6 mm were 0.67 and 0.02 cGy/s/mCi, respectively. The geometry function, radial dose function and anisotropy function were generated.

Dosimetric parameters of three new solid core I‐125 brachytherapy sources

PubMed Central

Solberg, Timothy D.; DeMarco, John J.; Hugo, Geoffrey; Wallace, Robert E.

2002-01-01

Monte Carlo calculations and TLD measurements have been performed for the purpose of characterizing dosimetric properties of new commercially available brachytherapy sources. All sources tested consisted of a solid core, upon which a thin layer of I125 has been adsorbed, encased within a titanium housing. The PharmaSeed BT‐125 source manufactured by Syncor is available in silver or palladium core configurations while the ADVANTAGE source from IsoAid has silver only. Dosimetric properties, including the dose rate constant, radial dose function, and anisotropy characteristics were determined according to the TG‐43 protocol. Additionally, the geometry function was calculated exactly using Monte Carlo and compared with both the point and line source approximations. The 1999 NIST standard was followed in determining air kerma strength. Dose rate constants were calculated to be 0.955±0.005,0.967±0.005, and 0.962±0.005 cGyh−1U−1 for the PharmaSeed BT‐125‐1, BT‐125‐2, and ADVANTAGE sources, respectively. TLD measurements were in excellent agreement with Monte Carlo calculations. Radial dose function, g(r), calculated to a distance of 10 cm, and anisotropy function F(r, θ), calculated for radii from 0.5 to 7.0 cm, were similar among all source configurations. Anisotropy constants, ϕ¯an, were calculated to be 0.941, 0.944, and 0.960 for the three sources, respectively. All dosimetric parameters were found to be in close agreement with previously published data for similar source configurations. The MCNP Monte Carlo code appears to be ideally suited to low energy dosimetry applications. PACS number(s): 87.53.–j PMID:11958652

Air kerma strength characterization of a GZP6 Cobalt-60 brachytherapy source

PubMed Central

Toossi, Mohammad Taghi Bahreyni; Ghorbani, Mahdi; Mowlavi, Ali Asghar; Taheri, Mojtaba; Layegh, Mohsen; Makhdoumi, Yasha; Meigooni, Ali Soleimani

2010-01-01

Background Task group number 40 (TG-40) of the American Association of Physicists in Medicine (AAPM) has recommended calibration of any brachytherapy source before its clinical use. GZP6 afterloading brachytherapy unit is a 60Co high dose rate (HDR) system recently being used in some of the Iranian radiotherapy centers. Aim In this study air kerma strength (AKS) of 60Co source number three of this unit was estimated by Monte Carlo simulation and in air measurements. Materials and methods Simulation was performed by employing the MCNP-4C Monte Carlo code. Self-absorption of the source core and its capsule were taken into account when calculating air kerma strength. In-air measurements were performed according to the multiple distance method; where a specially designed jig and a 0.6 cm3 Farmer type ionization chamber were used for the measurements. Monte Carlo simulation, in air measurement and GZP6 treatment planning results were compared for primary air kerma strength (as for November 8th 2005). Results Monte Carlo calculated and in air measured air kerma strength were respectively equal to 17240.01 μGym2 h−1 and 16991.83 μGym2 h−1. The value provided by the GZP6 treatment planning system (TPS) was “15355 μGym2 h−1”. Conclusion The calculated and measured AKS values are in good agreement. Calculated-TPS and measured-TPS AKS values are also in agreement within the uncertainties related to our calculation, measurements and those certified by the GZP6 manufacturer. Considering the uncertainties, the TPS value for AKS is validated by our calculations and measurements, however, it is incorporated with a large uncertainty. PMID:24376948

Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

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

Karthikeyan, R.; Tellier, R. L.; Hebert, A.

2006-07-01

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« 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.

A possible approach to 14MeV neutron moderation: A preliminary study case.

PubMed

Flammini, D; Pilotti, R; Pietropaolo, A

2017-07-01

Deuterium-Tritium (D-T) interactions produce almost monochromatic neutrons with about 14MeV energy. These neutrons are used in benchmark experiments as well as for neutron cross sections assessment in fusion reactors technology. The possibility to moderate 14MeV neutrons for purposes beyond fusion is worth to be studied in relation to projects of intense D-T sources. In this preliminary study, carried out using the MCNP Monte Carlo code, the moderation of 14MeV neutrons is approached foreseeing the use of combination of metallic materials as pre-moderator and reflectors coupled to standard water moderators. Copyright © 2017 Elsevier Ltd. All rights reserved.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

NASA Astrophysics Data System (ADS)

McArthur, Matthew S.; Rees, Lawrence B.; Czirr, J. Bart

2016-08-01

Using the combination of a neutron-sensitive 6Li glass scintillator detector with a neutron-insensitive 7Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on 6Li. We used this detector with a 252Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

Monte Carlo modelling of large scale NORM sources using MCNP.

PubMed

Wallace, J D

2013-12-01

The representative Monte Carlo modelling of large scale planar sources (for comparison to external environmental radiation fields) is undertaken using substantial diameter and thin profile planar cylindrical sources. The relative impact of source extent, soil thickness and sky-shine are investigated to guide decisions relating to representative geometries. In addition, the impact of source to detector distance on the nature of the detector response, for a range of source sizes, has been investigated. These investigations, using an MCNP based model, indicate a soil cylinder of greater than 20 m diameter and of no less than 50 cm depth/height, combined with a 20 m deep sky section above the soil cylinder, are needed to representatively model the semi-infinite plane of uniformly distributed NORM sources. Initial investigation of the effect of detector placement indicate that smaller source sizes may be used to achieve a representative response at shorter source to detector distances. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

The effect of tandem-ovoid titanium applicator on points A, B, bladder, and rectum doses in gynecological brachytherapy using 192Ir.

PubMed

Sadeghi, Mohammad Hosein; Sina, Sedigheh; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani

2018-02-01

The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy.

Vectorized Monte Carlo methods for reactor lattice analysis

NASA Technical Reports Server (NTRS)

Brown, F. B.

1984-01-01

Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

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

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.« less

FY2012 summary of tasks completed on PROTEUS-thermal work.

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

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Monte Carlo simulation of random, porous (foam) structures for neutron detection

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Fronk, Ryan G.; Shultis, J. Kenneth; Roberts, Jeremy A.; Edwards, Nathaniel S.; Stevenson, Sarah R.; Tiner, Christopher N.; McGregor, Douglas S.

2017-01-01

Porous media incorporating highly neutron-sensitive materials are of interest for use in the development of neutron detectors. Previous studies have shown experimentally the feasibility of 6LiF-saturated, multi-layered detectors; however, the random geometry of porous materials has limited the effectiveness of simulation efforts. The results of scatterless neutron transport and subsequent charged reaction product ion energy deposition are reported here using a novel Monte Carlo method and compared to results obtained by MCNP6. This new Dynamic Path Generation (DPG) Monte Carlo method was developed in order to overcome the complexities of modeling a random porous geometry in MCNP6. The DPG method is then applied to determine the optimal coating thickness for 10B4C-coated reticulated vitreous-carbon (RVC) foams. The optimal coating thickness for 4.1275 cm-thick 10B4C-coated reticulated vitreous carbon foams with porosities of 5, 10, 20, 30, 45, and 80 pores per inch (PPI) were determined for ionizing gas pressures of 1.0 and 2.8 atm. A simulated, maximum, intrinsic thermal-neutron detection efficiency of 62.8±0.25% was predicted for an 80 PPI RVC foam with a 0.2 μm thick coating of 10B4C, for a lower level discriminator setting of 75 keV and an argon pressure of 2.8 atm.

Impact of the vaginal applicator and dummy pellets on the dosimetry parameters of Cs-137 brachytherapy source.

PubMed

Sina, Sedigheh; Faghihi, Reza; Meigooni, Ali S; Mehdizadeh, Simin; Mosleh Shirazi, M Amin; Zehtabian, Mehdi

2011-05-19

In this study, dose rate distribution around a spherical 137Cs pellet source, from a low-dose-rate (LDR) Selectron remote afterloading system used in gynecological brachytherapy, has been determined using experimental and Monte Carlo simulation techniques. Monte Carlo simulations were performed using MCNP4C code, for a single pellet source in water medium and Plexiglas, and measurements were performed in Plexiglas phantom material using LiF TLD chips. Absolute dose rate distribution and the dosimetric parameters, such as dose rate constant, radial dose functions, and anisotropy functions, were obtained for a single pellet source. In order to investigate the effect of the applicator and surrounding pellets on dosimetric parameters of the source, the simulations were repeated for six different arrangements with a single active source and five non-active pellets inside central metallic tubing of a vaginal cylindrical applicator. In commercial treatment planning systems (TPS), the attenuation effects of the applicator and inactive spacers on total dose are neglected. The results indicate that this effect could lead to overestimation of the calculated F(r,θ), by up to 7% along the longitudinal axis of the applicator, especially beyond the applicator tip. According to the results obtained in this study, in a real situation in treatment of patients using cylindrical vaginal applicator and using several active pellets, there will be a large discrepancy between the result of superposition and Monte Carlo simulations.

Development of the 3DHZETRN code for space radiation protection

NASA Astrophysics Data System (ADS)

Wilson, John; Badavi, Francis; Slaba, Tony; Reddell, Brandon; Bahadori, Amir; Singleterry, Robert

Space radiation protection requires computationally efficient shield assessment methods that have been verified and validated. The HZETRN code is the engineering design code used for low Earth orbit dosimetric analysis and astronaut record keeping with end-to-end validation to twenty percent in Space Shuttle and International Space Station operations. HZETRN treated diffusive leakage only at the distal surface limiting its application to systems with a large radius of curvature. A revision of HZETRN that included forward and backward diffusion allowed neutron leakage to be evaluated at both the near and distal surfaces. That revision provided a deterministic code of high computational efficiency that was in substantial agreement with Monte Carlo (MC) codes in flat plates (at least to the degree that MC codes agree among themselves). In the present paper, the 3DHZETRN formalism capable of evaluation in general geometry is described. Benchmarking will help quantify uncertainty with MC codes (Geant4, FLUKA, MCNP6, and PHITS) in simple shapes such as spheres within spherical shells and boxes. Connection of the 3DHZETRN to general geometry will be discussed.

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.

Evaluation and comparison of absorbed dose for electron beams by LiF and diamond dosimeters

NASA Astrophysics Data System (ADS)

Mosia, G. J.; Chamberlain, A. C.

2007-09-01

The absorbed dose response of LiF and diamond thermoluminescent dosimeters (TLDs), calibrated in 60Co γ-rays, has been determined using the MCNP4B Monte Carlo code system in mono-energetic megavoltage electron beams from 5 to 20 MeV. Evaluation of the dose responses was done against the dose responses of published works by other investigators. Dose responses of both dosimeters were compared to establish if any relation exists between them. The dosimeters were irradiated in a water phantom with the centre of their top surfaces (0.32×0.32 cm 2), placed at dmax perpendicular to the radiation beam on the central axis. For LiF TLD, dose responses ranged from 0.945±0.017 to 0.997±0.011. For the diamond TLD, the dose response ranged from 0.940±0.017 to 1.018±0.011. To correct for dose responses by both dosimeters, energy correction factors were generated from dose response results of both TLDs. For LiF TLD, these correction factors ranged from 1.003 up to 1.058 and for diamond TLD the factors ranged from 0.982 up to 1.064. The results show that diamond TLDs can be used in the place of the well-established LiF TLDs and that Monte Carlo code systems can be used in dose determinations for radiotherapy treatment planning.

Multiprocessing MCNP on an IBN RS/6000 cluster

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

McKinney, G.W.; West, J.T.

1993-01-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors P and the fraction f of task time that multiprocesses, can be formulated using Amdahl's law: S(f, P) =1/(1-f+f/P). However, for most applications, this theoretical limit cannot be achieved because of additional terms (e.g., multitasking overhead, memory overlap, etc.) that are not included in Amdahl's law. Monte Carlo transport is a natural candidate for multiprocessing because the particle tracks are generally independent, and the precision of the result increases as the square Foot of the number of particles tracked.« less

Multiprocessing MCNP on an IBM RS/6000 cluster

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

McKinney, G.W.; West, J.T.

1993-03-01

The advent of high-performance computer systems has brought to maturity programming concepts like vectorization, multiprocessing, and multitasking. While there are many schools of thought as to the most significant factor in obtaining order-of-magnitude increases in performance, such speedup can only be achieved by integrating the computer system and application code. Vectorization leads to faster manipulation of arrays by overlapping instruction CPU cycles. Discrete ordinates codes, which require the solving of large matrices, have proved to be major benefactors of vectorization. Monte Carlo transport, on the other hand, typically contains numerous logic statements and requires extensive redevelopment to benefit from vectorization.more » Multiprocessing and multitasking provide additional CPU cycles via multiple processors. Such systems are generally designed with either common memory access (multitasking) or distributed memory access. In both cases, theoretical speedup, as a function of the number of processors (P) and the fraction of task time that multiprocesses (f), can be formulated using Amdahl`s Law S ((f,P) = 1 f + f/P). However, for most applications this theoretical limit cannot be achieved, due to additional terms not included in Amdahl`s Law. Monte Carlo transport is a natural candidate for multiprocessing, since the particle tracks are generally independent and the precision of the result increases as the square root of the number of particles tracked.« less

Indoor Fast Neutron Generator for Biophysical and Electronic Applications

NASA Astrophysics Data System (ADS)

Cannuli, A.; Caccamo, M. T.; Marchese, N.; Tomarchio, E. A.; Pace, C.; Magazù, S.

2018-05-01

This study focuses the attention on an indoor fast neutron generator for biophysical and electronic applications. More specifically, the findings obtained by several simulations with the MCNP Monte Carlo code, necessary for the realization of a shield for indoor measurements, are presented. Furthermore, an evaluation of the neutron spectrum modification caused by the shielding is reported. Fast neutron generators are a valid and interesting available source of neutrons, increasingly employed in a wide range of research fields, such as science and engineering. The employed portable pulsed neutron source is a MP320 Thermo Scientific neutron generator, able to generate 2.5 MeV neutrons with a neutron yield of 2.0 x 106 n/s, a pulse rate of 250 Hz to 20 KHz and a duty factor varying from 5% to 100%. The neutron generator, based on Deuterium-Deuterium nuclear fusion reactions, is employed in conjunction with a solid-state photon detector, made of n-type high-purity germanium (PINS-GMX by ORTEC) and it is mainly addressed to biophysical and electronic studies. The present study showed a proposal for the realization of a shield necessary for indoor applications for MP320 neutron generator, with a particular analysis of the transport of neutrons simulated with Monte Carlo code and described the two main lines of research in which the source will be used.

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.

Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions.

PubMed

Stewart, Robert D; Streitmatter, Seth W; Argento, David C; Kirkby, Charles; Goorley, John T; Moffitt, Greg; Jevremovic, Tatjana; Sandison, George A

2015-11-07

To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, (137)Cs γ-rays, neutrons and light ions relative to γ-rays from (60)Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that (137)Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from (60)Co (RBEDSB  =  1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than (60)Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as (60)Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

Theoretical study of depth profiling with gamma- and X-ray spectrometry based on measurements of intensity ratios

NASA Astrophysics Data System (ADS)

Bártová, H.; Trojek, T.; Johnová, K.

2017-11-01

This article describes the method for the estimation of depth distribution of radionuclides in a material with gamma-ray spectrometry, and the identification of a layered structure of a material with X-ray fluorescence analysis. This method is based on the measurement of a ratio of two gamma or X-ray lines of a radionuclide or a chemical element, respectively. Its principle consists in different attenuation coefficient for these two lines in a measured material. The main aim of this investigation was to show how the detected ratio of these two lines depends on depth distribution of an analyte and mainly how this ratio depends on density and chemical composition of measured materials. Several different calculation arrangements were made and a lot of Monte Carlo simulation with the code MCNP - Monte Carlo N-Particle (Briesmeister, 2000) was performed to answer these questions. For X-ray spectrometry, the calculated Kα/Kβ diagrams were found to be almost independent upon matrix density and composition. Thanks to this phenomenon it would be possible to draw only one Kα/Kβ diagram for an element whose depth distribution is examined.

Monte Carlo simulation of a NaI(Tl) detector for in situ radioactivity measurements in the marine environment.

PubMed

Zhang, Yingying; Li, Changkai; Liu, Dongyan; Zhang, Ying; Liu, Yan

2015-04-01

To develop in situ NaI(Tl) detector for radioactivity measurement in the marine environment, the Monte Carlo N-Particle (MCNP) Transport Code was utilized to simulate the measurement of NaI(Tl) detector immersed in seawater, taking into account the material and geometry of the detector, and the interactions between the photons with the atoms of the seawater and the detector. The simulation results of the marine detection efficiency and distance were deduced and analyzed. In order to test their reliability, the field measurement was made at open sea and the experimental value of the marine detection efficiency was deduced and seems to be in good agreement with the simulated one. The minimum detectable activity for (137)Cs in the seawater of NaI(Tl) detector developed was determined mathematically at last. The simulation method and results in the paper can be used for the better design and quantitative calculation of in situ NaI(Tl) detector for radioactivity measurement in the marine environment, and also for some applications such as the installation on the marine monitoring platform and the quantitative analysis of radionuclides. Copyright © 2015 Elsevier Ltd. All rights reserved.

Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

PubMed

Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

2014-12-01

An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). 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 emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

Neutron dose estimation in a zero power nuclear reactor

NASA Astrophysics Data System (ADS)

Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

2016-10-01

This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

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

The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

PubMed

Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

2014-08-01

The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

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.

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

Van Der Marck, S. C.

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differencesmore » are probably caused by elements such as Be, C, Fe, Zr, W. (authors)« less

Shutdown Dose Rate Analysis for the long-pulse D-D Operation Phase in KSTAR

NASA Astrophysics Data System (ADS)

Park, Jin Hun; Han, Jung-Hoon; Kim, D. H.; Joo, K. S.; Hwang, Y. S.

2017-09-01

KSTAR is a medium size fully superconducting tokamak. The deuterium-deuterium (D-D) reaction in the KSTAR tokamak generates neutrons with a peak yield of 3.5x1016 per second through a pulse operation of 100 seconds. The effect of neutron generation from full D-D high power KSTAR operation mode to the machine, such as activation, shutdown dose rate, and nuclear heating, are estimated for an assurance of safety during operation, maintenance, and machine upgrade. The nuclear heating of the in-vessel components, and neutron activation of the surrounding materials have been investigated. The dose rates during operation and after shutdown of KSTAR have been calculated by a 3D CAD model of KSTAR with the Monte Carlo code MCNP5 (neutron flux and decay photon), the inventory code FISPACT (activation and decay photon) and the FENDL 2.1 nuclear data library.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

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

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Development of a patient-specific dosimetry estimation system in nuclear medicine examination

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

Lin, H. H.; Dong, S. L.; Yang, H. J.

2011-07-01

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by thismore » system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)« less

Calculation of absorbed dose and biological effectiveness from photonuclear reactions in a bremsstrahlung beam of end point 50 MeV.

PubMed

Gudowska, I; Brahme, A; Andreo, P; Gudowski, W; Kierkegaard, J

1999-09-01

The absorbed dose due to photonuclear reactions in soft tissue, lung, breast, adipose tissue and cortical bone has been evaluated for a scanned bremsstrahlung beam of end point 50 MeV from a racetrack accelerator. The Monte Carlo code MCNP4B was used to determine the photon source spectrum from the bremsstrahlung target and to simulate the transport of photons through the treatment head and the patient. Photonuclear particle production in tissue was calculated numerically using the energy distributions of photons derived from the Monte Carlo simulations. The transport of photoneutrons in the patient and the photoneutron absorbed dose to tissue were determined using MCNP4B; the absorbed dose due to charged photonuclear particles was calculated numerically assuming total energy absorption in tissue voxels of 1 cm3. The photonuclear absorbed dose to soft tissue, lung, breast and adipose tissue is about (0.11-0.12)+/-0.05% of the maximum photon dose at a depth of 5.5 cm. The absorbed dose to cortical bone is about 45% larger than that to soft tissue. If the contributions from all photoparticles (n, p, 3He and 4He particles and recoils of the residual nuclei) produced in the soft tissue and the accelerator, and from positron radiation and gammas due to induced radioactivity and excited states of the nuclei, are taken into account the total photonuclear absorbed dose delivered to soft tissue is about 0.15+/-0.08% of the maximum photon dose. It has been estimated that the RBE of the photon beam of 50 MV acceleration potential is approximately 2% higher than that of conventional 60Co radiation.

The effect of tandem-ovoid titanium applicator on points A, B, bladder, and rectum doses in gynecological brachytherapy using 192Ir

PubMed Central

Sadeghi, Mohammad Hosein; Mehdizadeh, Amir; Faghihi, Reza; Moharramzadeh, Vahed; Meigooni, Ali Soleimani

2018-01-01

Purpose The dosimetry procedure by simple superposition accounts only for the self-shielding of the source and does not take into account the attenuation of photons by the applicators. The purpose of this investigation is an estimation of the effects of the tandem and ovoid applicator on dose distribution inside the phantom by MCNP5 Monte Carlo simulations. Material and methods In this study, the superposition method is used for obtaining the dose distribution in the phantom without using the applicator for a typical gynecological brachytherapy (superposition-1). Then, the sources are simulated inside the tandem and ovoid applicator to identify the effect of applicator attenuation (superposition-2), and the dose at points A, B, bladder, and rectum were compared with the results of superposition. The exact dwell positions, times of the source, and positions of the dosimetry points were determined in images of a patient and treatment data of an adult woman patient from a cancer center. The MCNP5 Monte Carlo (MC) code was used for simulation of the phantoms, applicators, and the sources. Results The results of this study showed no significant differences between the results of superposition method and the MC simulations for different dosimetry points. The difference in all important dosimetry points was found to be less than 5%. Conclusions According to the results, applicator attenuation has no significant effect on the calculated points dose, the superposition method, adding the dose of each source obtained by the MC simulation, can estimate the dose to points A, B, bladder, and rectum with good accuracy. PMID:29619061

A voxel-based mouse for internal dose calculations using Monte Carlo simulations (MCNP).

PubMed

Bitar, A; Lisbona, A; Thedrez, P; Sai Maurel, C; Le Forestier, D; Barbet, J; Bardies, M

2007-02-21

Murine models are useful for targeted radiotherapy pre-clinical experiments. These models can help to assess the potential interest of new radiopharmaceuticals. In this study, we developed a voxel-based mouse for dosimetric estimates. A female nude mouse (30 g) was frozen and cut into slices. High-resolution digital photographs were taken directly on the frozen block after each section. Images were segmented manually. Monoenergetic photon or electron sources were simulated using the MCNP4c2 Monte Carlo code for each source organ, in order to give tables of S-factors (in Gy Bq-1 s-1) for all target organs. Results obtained from monoenergetic particles were then used to generate S-factors for several radionuclides of potential interest in targeted radiotherapy. Thirteen source and 25 target regions were considered in this study. For each source region, 16 photon and 16 electron energies were simulated. Absorbed fractions, specific absorbed fractions and S-factors were calculated for 16 radionuclides of interest for targeted radiotherapy. The results obtained generally agree well with data published previously. For electron energies ranging from 0.1 to 2.5 MeV, the self-absorbed fraction varies from 0.98 to 0.376 for the liver, and from 0.89 to 0.04 for the thyroid. Electrons cannot be considered as 'non-penetrating' radiation for energies above 0.5 MeV for mouse organs. This observation can be generalized to radionuclides: for example, the beta self-absorbed fraction for the thyroid was 0.616 for I-131; absorbed fractions for Y-90 for left kidney-to-left kidney and for left kidney-to-spleen were 0.486 and 0.058, respectively. Our voxel-based mouse allowed us to generate a dosimetric database for use in preclinical targeted radiotherapy experiments.

Validation of the MCNP6 electron-photon transport algorithm: multiple-scattering of 13- and 20-MeV electrons in thin foils

NASA Astrophysics Data System (ADS)

Dixon, David A.; Hughes, H. Grady

2017-09-01

This paper presents a validation test comparing angular distributions from an electron multiple-scattering experiment with those generated using the MCNP6 Monte Carlo code system. In this experiment, a 13- and 20-MeV electron pencil beam is deflected by thin foils with atomic numbers from 4 to 79. To determine the angular distribution, the fluence is measured down range of the scattering foil at various radii orthogonal to the beam line. The characteristic angle (the angle for which the max of the distribution is reduced by 1/e) is then determined from the angular distribution and compared with experiment. Multiple scattering foils tested herein include beryllium, carbon, aluminum, copper, and gold. For the default electron-photon transport settings, the calculated characteristic angle was statistically distinguishable from measurement and generally broader than the measured distributions. The average relative difference ranged from 5.8% to 12.2% over all of the foils, source energies, and physics settings tested. This validation illuminated a deficiency in the computation of the underlying angular distributions that is well understood. As a result, code enhancements were made to stabilize the angular distributions in the presence of very small substeps. However, the enhancement only marginally improved results indicating that additional algorithmic details should be studied.

INDIVIDUAL DOSIMETRY IN DISPOSAL REPOSITORY OF HEAT-GENERATING NUCLEAR WASTE.

PubMed

Pang, Bo; Saurí Suárez, Héctor; Becker, Frank

2016-09-01

Certain working scenarios in a disposal facility of heat-generating nuclear waste might lead to an enhanced level of radiation exposure for workers in such facilities. Hence, a realistic estimation of the personal dose during individual working scenarios is desired. In this study, the general-purpose Monte Carlo N-Particle code MCNP6 (Pelowitz, D. B. (ed). MCNP6 user manual LA-CP-13-00634, Rev. 0 (2013)) was applied to simulate a representative radiation field in a disposal facility. A tool to estimate the personal dose was then proposed by taking into account the influence of individual motion sequences during working scenarios. As basis for this approach, a movable whole-body phantom was developed to describe individual body gestures of the workers during motion sequences. In this study, the proposed method was applied to the German concept of geological disposal in rock salt. The feasibility of the proposed approach was demonstrated with an example of working scenario in an emplacement drift of a rock salt mine. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

An improved MCNP version of the NORMAN voxel phantom for dosimetry studies.

PubMed

Ferrari, P; Gualdrini, G

2005-09-21

In recent years voxel phantoms have been developed on the basis of tomographic data of real individuals allowing new sets of conversion coefficients to be calculated for effective dose. Progress in radiation studies brought ICRP to revise its recommendations and a new report, already circulated in draft form, is expected to change the actual effective dose evaluation method. In the present paper the voxel phantom NORMAN developed at HPA, formerly NRPB, was employed with MCNP Monte Carlo code. A modified version of the phantom, NORMAN-05, was developed to take into account the new set of tissues and weighting factors proposed in the cited ICRP draft. Air kerma to organ equivalent dose and effective dose conversion coefficients for antero-posterior and postero-anterior parallel photon beam irradiations, from 20 keV to 10 MeV, have been calculated and compared with data obtained in other laboratories using different numerical phantoms. Obtained results are in good agreement with published data with some differences for the effective dose calculated employing the proposed new tissue weighting factors set in comparison with previous evaluations based on the ICRP 60 report.

Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

2017-09-01

The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

Reanalysis of tritium production in a sphere of /sup 6/LiD irradiated by 14-MeV neutrons

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

Fawcett, L.R. Jr.

1985-08-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD irradiated by a central source of 14-MeV neutrons has been reanalyzed. The /sup 6/LiD sphere consisted of 10 solid hemispherical nested shells with ampules of /sup 6/LiH, /sup 7/LiH, and activation foils located 2.2, 5, 7.7, 12.6, 20, and 30 cm from the center. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport through the /sup 6/LiD, tritium production in the ampules, and foil activation. The MCNP input model was three-dimensional and employed ENDF/B-V cross sections for transport, tritiummore » production, and (where available) foil activation. The reanalyzed experimentally observed-to-calculated values of tritium production were 1.053 +- 2.1% in /sup 6/LiH and 0.999 +- 2.1% in /sup 7/LiH. The recalculated foil activation observed-to-calculated ratios were not generally improved over those reported in the original analysis.« less

(U) Introduction to Monte Carlo Methods

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

Hungerford, Aimee L.

2017-03-20

Monte Carlo methods are very valuable for representing solutions to particle transport problems. Here we describe a “cook book” approach to handling the terms in a transport equation using Monte Carlo methods. Focus is on the mechanics of a numerical Monte Carlo code, rather than the mathematical foundations of the method.

Dosimetric verification of the anisotropic analytical algorithm in lung equivalent heterogeneities with and without bone equivalent heterogeneities

PubMed Central

Ono, Kaoru; Endo, Satoru; Tanaka, Kenichi; Hoshi, Masaharu; Hirokawa, Yutaka

2010-01-01

Purpose: In this study, the authors evaluated the accuracy of dose calculations performed by the convolution∕superposition based anisotropic analytical algorithm (AAA) in lung equivalent heterogeneities with and without bone equivalent heterogeneities. Methods: Calculations of PDDs using the AAA and Monte Carlo simulations (MCNP4C) were compared to ionization chamber measurements with a heterogeneous phantom consisting of lung equivalent and bone equivalent materials. Both 6 and 10 MV photon beams of 4×4 and 10×10 cm2 field sizes were used for the simulations. Furthermore, changes of energy spectrum with depth for the heterogeneous phantom using MCNP were calculated. Results: The ionization chamber measurements and MCNP calculations in a lung equivalent phantom were in good agreement, having an average deviation of only 0.64±0.45%. For both 6 and 10 MV beams, the average deviation was less than 2% for the 4×4 and 10×10 cm2 fields in the water-lung equivalent phantom and the 4×4 cm2 field in the water-lung-bone equivalent phantom. Maximum deviations for the 10×10 cm2 field in the lung equivalent phantom before and after the bone slab were 5.0% and 4.1%, respectively. The Monte Carlo simulation demonstrated an increase of the low-energy photon component in these regions, more for the 10×10 cm2 field compared to the 4×4 cm2 field. Conclusions: The low-energy photon by Monte Carlo simulation component increases sharply in larger fields when there is a significant presence of bone equivalent heterogeneities. This leads to great changes in the build-up and build-down at the interfaces of different density materials. The AAA calculation modeling of the effect is not deemed to be sufficiently accurate. PMID:20879604

Use of Fluka to Create Dose Calculations

NASA Technical Reports Server (NTRS)

Lee, Kerry T.; Barzilla, Janet; Townsend, Lawrence; Brittingham, John

2012-01-01

Monte Carlo codes provide an effective means of modeling three dimensional radiation transport; however, their use is both time- and resource-intensive. The creation of a lookup table or parameterization from Monte Carlo simulation allows users to perform calculations with Monte Carlo results without replicating lengthy calculations. FLUKA Monte Carlo transport code was used to develop lookup tables and parameterizations for data resulting from the penetration of layers of aluminum, polyethylene, and water with areal densities ranging from 0 to 100 g/cm^2. Heavy charged ion radiation including ions from Z=1 to Z=26 and from 0.1 to 10 GeV/nucleon were simulated. Dose, dose equivalent, and fluence as a function of particle identity, energy, and scattering angle were examined at various depths. Calculations were compared against well-known results and against the results of other deterministic and Monte Carlo codes. Results will be presented.

Monte Carlo reference data sets for imaging research: Executive summary of the report of AAPM Research Committee Task Group 195.

PubMed

Sechopoulos, Ioannis; Ali, Elsayed S M; Badal, Andreu; Badano, Aldo; Boone, John M; Kyprianou, Iacovos S; Mainegra-Hing, Ernesto; McMillan, Kyle L; McNitt-Gray, Michael F; Rogers, D W O; Samei, Ehsan; Turner, Adam C

2015-10-01

The use of Monte Carlo simulations in diagnostic medical imaging research is widespread due to its flexibility and ability to estimate quantities that are challenging to measure empirically. However, any new Monte Carlo simulation code needs to be validated before it can be used reliably. The type and degree of validation required depends on the goals of the research project, but, typically, such validation involves either comparison of simulation results to physical measurements or to previously published results obtained with established Monte Carlo codes. The former is complicated due to nuances of experimental conditions and uncertainty, while the latter is challenging due to typical graphical presentation and lack of simulation details in previous publications. In addition, entering the field of Monte Carlo simulations in general involves a steep learning curve. It is not a simple task to learn how to program and interpret a Monte Carlo simulation, even when using one of the publicly available code packages. This Task Group report provides a common reference for benchmarking Monte Carlo simulations across a range of Monte Carlo codes and simulation scenarios. In the report, all simulation conditions are provided for six different Monte Carlo simulation cases that involve common x-ray based imaging research areas. The results obtained for the six cases using four publicly available Monte Carlo software packages are included in tabular form. In addition to a full description of all simulation conditions and results, a discussion and comparison of results among the Monte Carlo packages and the lessons learned during the compilation of these results are included. This abridged version of the report includes only an introductory description of the six cases and a brief example of the results of one of the cases. This work provides an investigator the necessary information to benchmark his/her Monte Carlo simulation software against the reference cases included here before performing his/her own novel research. In addition, an investigator entering the field of Monte Carlo simulations can use these descriptions and results as a self-teaching tool to ensure that he/she is able to perform a specific simulation correctly. Finally, educators can assign these cases as learning projects as part of course objectives or training programs.

Monte Carlo calculated TG-60 dosimetry parameters for the {beta}{sup -} emitter {sup 153}Sm brachytherapy source

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

Sadeghi, Mahdi; Taghdiri, Fatemeh; Hamed Hosseini, S.

Purpose: The formalism recommended by Task Group 60 (TG-60) of the American Association of Physicists in Medicine (AAPM) is applicable for {beta} sources. Radioactive biocompatible and biodegradable {sup 153}Sm glass seed without encapsulation is a {beta}{sup -} emitter radionuclide with a short half-life and delivers a high dose rate to the tumor in the millimeter range. This study presents the results of Monte Carlo calculations of the dosimetric parameters for the {sup 153}Sm brachytherapy source. Methods: Version 5 of the (MCNP) Monte Carlo radiation transport code was used to calculate two-dimensional dose distributions around the source. The dosimetric parameters ofmore » AAPM TG-60 recommendations including the reference dose rate, the radial dose function, the anisotropy function, and the one-dimensional anisotropy function were obtained. Results: The dose rate value at the reference point was estimated to be 9.21{+-}0.6 cGy h{sup -1} {mu}Ci{sup -1}. Due to the low energy beta emitted from {sup 153}Sm sources, the dose fall-off profile is sharper than the other beta emitter sources. The calculated dosimetric parameters in this study are compared to several beta and photon emitting seeds. Conclusions: The results show the advantage of the {sup 153}Sm source in comparison with the other sources because of the rapid dose fall-off of beta ray and high dose rate at the short distances of the seed. The results would be helpful in the development of the radioactive implants using {sup 153}Sm seeds for the brachytherapy treatment.« less

PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code

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

Iandola, F N; O'Brien, M J; Procassini, R J

2010-11-29

Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improvesmore » usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.« less

Methodology comparison for gamma-heating calculations in material-testing reactors

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

Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A.

2015-07-01

The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physicalmore » models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear heating is represented by the physical quantity called absorbed dose (energy deposition induced by particle-matter interactions, divided by mass). Its calculation with Monte Carlo codes is possible but computationally expensive as it requires transport simulation of charged particles, along with neutrons and photons. For that reason, the calculation of another physical quantity, called KERMA, is often preferred, as KERMA calculation with Monte Carlo codes only requires transport of neutral particles. However, KERMA is only an estimator of the absorbed dose and many conditions must be fulfilled for KERMA to be equal to absorbed dose, including so-called condition of electronic equilibrium. Also, Monte Carlo computations of absorbed dose still present some physical approximations, even though there is only a limited number of them. Some of these approximations are linked to the way how Monte Carlo codes apprehend the transport simulation of charged particles and the productive and destructive interactions between photons, electrons and positrons. There exists a huge variety of electromagnetic shower models which tackle this topic. Differences in the implementation of these models can lead to discrepancies in calculated values of absorbed dose between different Monte Carlo codes. The magnitude of order of such potential discrepancies should be quantified for JHR gamma-heating calculations. We consequently present a two-pronged plan. In a first phase, we intend to perform compared absorbed dose / KERMA Monte Carlo calculations in the JHR. This way, we will study the presence or absence of electronic equilibrium in the different JHR structures and experimental devices and we will give recommendations for the choice of KERMA or absorbed dose when calculating gamma heating in the JHR. In a second phase, we intend to perform compared TRIPOLI4 / MCNP absorbed dose calculations in a simplified JHR-representative geometry. For this comparison, we will use the same nuclear data library for both codes (the European library JEFF3.1.1 and photon library EPDL97) so as to isolate the effects from electromagnetic shower models on absorbed dose calculation. This way, we hope to get insightful feedback on these models and their implementation in Monte Carlo codes. (authors)« less

G4DARI: Geant4/GATE based Monte Carlo simulation interface for dosimetry calculation in radiotherapy.

PubMed

Slimani, Faiçal A A; Hamdi, Mahdjoub; Bentourkia, M'hamed

2018-05-01

Monte Carlo (MC) simulation is widely recognized as an important technique to study the physics of particle interactions in nuclear medicine and radiation therapy. There are different codes dedicated to dosimetry applications and widely used today in research or in clinical application, such as MCNP, EGSnrc and Geant4. However, such codes made the physics easier but the programming remains a tedious task even for physicists familiar with computer programming. In this paper we report the development of a new interface GEANT4 Dose And Radiation Interactions (G4DARI) based on GEANT4 for absorbed dose calculation and for particle tracking in humans, small animals and complex phantoms. The calculation of the absorbed dose is performed based on 3D CT human or animal images in DICOM format, from images of phantoms or from solid volumes which can be made from any pure or composite material to be specified by its molecular formula. G4DARI offers menus to the user and tabs to be filled with values or chemical formulas. The interface is described and as application, we show results obtained in a lung tumor in a digital mouse irradiated with seven energy beams, and in a patient with glioblastoma irradiated with five photon beams. In conclusion, G4DARI can be easily used by any researcher without the need to be familiar with computer programming, and it will be freely available as an application package. Copyright © 2018 Elsevier Ltd. All rights reserved.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

NASA Astrophysics Data System (ADS)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

Evaluation of the effective dose during BNCT at TRR thermal column epithermal facility.

PubMed

Jarahi, Hossein; Kasesaz, Yaser; Saleh-Koutahi, Seyed Mohsen

2016-04-01

An epithermal neutron beam has been designed for Boron neutron Capture Therapy (BNCT) at the thermal column of Tehran Research Reactor (TRR) recently. In this paper the whole body effective dose, as well as the equivalent doses of several organs have been calculated in this facility using MCNP4C Monte Carlo code. The effective dose has been calculated by using the absorbed doses determined for each individual organ, taking into account the radiation and tissue weighting factors. The ICRP 110 whole body male phantom has been used as a patient model. It was found that the effective dose during BNCT of a brain tumor is equal to 0.90Sv. This effective dose may induce a 4% secondary cancer risk. Copyright © 2016 Elsevier Ltd. All rights reserved.

Investigation on the reflector/moderator geometry and its effect on the neutron beam design in BNCT.

PubMed

Kasesaz, Y; Rahmani, F; Khalafi, H

2015-12-01

In order to provide an appropriate neutron beam for Boron Neutron Capture Therapy (BNCT), a special Beam Shaping Assembly (BSA) must be designed based on the neutron source specifications. A typical BSA includes moderator, reflector, collimator, thermal neutron filter, and gamma filter. In common BSA, the reflector is considered as a layer which covers the sides of the moderator materials. In this paper, new reflector/moderator geometries including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. It was found that the proposed configurations have a significant effect to improve the thermal to epithermal neutron flux ratio which is an important neutron beam parameter. Copyright © 2015 Elsevier Ltd. All rights reserved.

Thermal neutron radiative capture on cadmium as a counting technique at the INES beam line at ISIS: A preliminary investigation of detector cross-talk.

PubMed

Festa, G; Grazzi, F; Pietropaolo, A; Scherillo, A; Schooneveld, E M

2017-12-01

Experimental tests are presented that assess the cross-talk level among three scintillation detectors used as neutron counters exploiting the thermal neutron radiative capture on Cd. The measurements were done at the INES diffractometer operating at the ISIS spallation neutron source (Rutherford Appleton Laboratory, UK). These tests follow a preliminary set of measurements performed on the same instrument to study the effectiveness of this thermal neutron counting strategy in neutron diffraction measurements, typically performed on INES using squashed 3 He filled gas tubes. The experimental data were collected in two different geometrical configurations of the detectors and compared to results of Monte Carlo simulations, performed using the MCNP code. Copyright © 2017 Elsevier Ltd. All rights reserved.

WE-AB-204-11: Development of a Nuclear Medicine Dosimetry Module for the GPU-Based Monte Carlo Code ARCHER

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

Liu, T; Lin, H; Xu, X

Purpose: To develop a nuclear medicine dosimetry module for the GPU-based Monte Carlo code ARCHER. Methods: We have developed a nuclear medicine dosimetry module for the fast Monte Carlo code ARCHER. The coupled electron-photon Monte Carlo transport kernel included in ARCHER is built upon the Dose Planning Method code (DPM). The developed module manages the radioactive decay simulation by consecutively tracking several types of radiation on a per disintegration basis using the statistical sampling method. Optimization techniques such as persistent threads and prefetching are studied and implemented. The developed module is verified against the VIDA code, which is based onmore » Geant4 toolkit and has previously been verified against OLINDA/EXM. A voxelized geometry is used in the preliminary test: a sphere made of ICRP soft tissue is surrounded by a box filled with water. Uniform activity distribution of I-131 is assumed in the sphere. Results: The self-absorption dose factors (mGy/MBqs) of the sphere with varying diameters are calculated by ARCHER and VIDA respectively. ARCHER’s result is in agreement with VIDA’s that are obtained from a previous publication. VIDA takes hours of CPU time to finish the computation, while it takes ARCHER 4.31 seconds for the 12.4-cm uniform activity sphere case. For a fairer CPU-GPU comparison, more effort will be made to eliminate the algorithmic differences. Conclusion: The coupled electron-photon Monte Carlo code ARCHER has been extended to radioactive decay simulation for nuclear medicine dosimetry. The developed code exhibits good performance in our preliminary test. The GPU-based Monte Carlo code is developed with grant support from the National Institute of Biomedical Imaging and Bioengineering through an R01 grant (R01EB015478)« less

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

NASA Astrophysics Data System (ADS)

Gomin, E. A.; Davidenko, V. D.; Zinchenko, A. S.; Kharchenko, I. K.

2017-12-01

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

Monte Carlo tests of the ELIPGRID-PC algorithm

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

Davidson, J.R.

1995-04-01

The standard tool for calculating the probability of detecting pockets of contamination called hot spots has been the ELIPGRID computer code of Singer and Wickman. The ELIPGRID-PC program has recently made this algorithm available for an IBM{reg_sign} PC. However, no known independent validation of the ELIPGRID algorithm exists. This document describes a Monte Carlo simulation-based validation of a modified version of the ELIPGRID-PC code. The modified ELIPGRID-PC code is shown to match Monte Carlo-calculated hot-spot detection probabilities to within {plus_minus}0.5% for 319 out of 320 test cases. The one exception, a very thin elliptical hot spot located within a rectangularmore » sampling grid, differed from the Monte Carlo-calculated probability by about 1%. These results provide confidence in the ability of the modified ELIPGRID-PC code to accurately predict hot-spot detection probabilities within an acceptable range of error.« less

OEDIPE: a new graphical user interface for fast construction of numerical phantoms and MCNP calculations.

PubMed

Franck, D; de Carlan, L; Pierrat, N; Broggio, D; Lamart, S

2007-01-01

Although great efforts have been made to improve the physical phantoms used to calibrate in vivo measurement systems, these phantoms represent a single average counting geometry and usually contain a uniform distribution of the radionuclide over the tissue substitute. As a matter of fact, significant corrections must be made to phantom-based calibration factors in order to obtain absolute calibration efficiencies applicable to a given individual. The importance of these corrections is particularly crucial when considering in vivo measurements of low energy photons emitted by radionuclides deposited in the lung such as actinides. Thus, it was desirable to develop a method for calibrating in vivo measurement systems that is more sensitive to these types of variability. Previous works have demonstrated the possibility of such a calibration using the Monte Carlo technique. Our research programme extended such investigations to the reconstruction of numerical anthropomorphic phantoms based on personal physiological data obtained by computed tomography. New procedures based on a new graphical user interface (GUI) for development of computational phantoms for Monte Carlo calculations and data analysis are being developed to take advantage of recent progress in image-processing codes. This paper presents the principal features of this new GUI. Results of calculations and comparison with experimental data are also presented and discussed in this work.

ACCELERATING FUSION REACTOR NEUTRONICS MODELING BY AUTOMATIC COUPLING OF HYBRID MONTE CARLO/DETERMINISTIC TRANSPORT ON CAD GEOMETRY

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

Biondo, Elliott D; Ibrahim, Ahmad M; Mosher, Scott W

2015-01-01

Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNcemore » reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).« less

Whole body counter calibration using Monte Carlo modeling with an array of phantom sizes based on national anthropometric reference data

NASA Astrophysics Data System (ADS)

Shypailo, R. J.; Ellis, K. J.

2011-05-01

During construction of the whole body counter (WBC) at the Children's Nutrition Research Center (CNRC), efficiency calibration was needed to translate acquired counts of 40K to actual grams of potassium for measurement of total body potassium (TBK) in a diverse subject population. The MCNP Monte Carlo n-particle simulation program was used to describe the WBC (54 detectors plus shielding), test individual detector counting response, and create a series of virtual anthropomorphic phantoms based on national reference anthropometric data. Each phantom included an outer layer of adipose tissue and an inner core of lean tissue. Phantoms were designed for both genders representing ages 3.5 to 18.5 years with body sizes from the 5th to the 95th percentile based on body weight. In addition, a spherical surface source surrounding the WBC was modeled in order to measure the effects of subject mass on room background interference. Individual detector measurements showed good agreement with the MCNP model. The background source model came close to agreement with empirical measurements, but showed a trend deviating from unity with increasing subject size. Results from the MCNP simulation of the CNRC WBC agreed well with empirical measurements using BOMAB phantoms. Individual detector efficiency corrections were used to improve the accuracy of the model. Nonlinear multiple regression efficiency calibration equations were derived for each gender. Room background correction is critical in improving the accuracy of the WBC calibration.

SU-C-209-05: Monte Carlo Model of a Prototype Backscatter X-Ray (BSX) Imager for Projective and Selective Object-Plane Imaging

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

Rolison, L; Samant, S; Baciak, J

Purpose: To develop a Monte Carlo N-Particle (MCNP) model for the validation of a prototype backscatter x-ray (BSX) imager, and optimization of BSX technology for medical applications, including selective object-plane imaging. Methods: BSX is an emerging technology that represents an alternative to conventional computed tomography (CT) and projective digital radiography (DR). It employs detectors located on the same side as the incident x-ray source, making use of backscatter and avoiding ring geometry to enclose the imaging object. Current BSX imagers suffer from low spatial resolution. A MCNP model was designed to replicate a BSX prototype used for flaw detection inmore » industrial materials. This prototype consisted of a 1.5mm diameter 60kVp pencil beam surrounded by a ring of four 5.0cm diameter NaI scintillation detectors. The imaging phantom consisted of a 2.9cm thick aluminum plate with five 0.6cm diameter holes drilled halfway. The experimental image was created using a raster scanning motion (in 1.5mm increments). Results: A qualitative comparison between the physical and simulated images showed very good agreement with 1.5mm spatial resolution in plane perpendicular to incident x-ray beam. The MCNP model developed the concept of radiography by selective plane detection (RSPD) for BSX, whereby specific object planes can be imaged by varying kVp. 10keV increments in mean x-ray energy yielded 4mm thick slice resolution in the phantom. Image resolution in the MCNP model can be further increased by increasing the number of detectors, and decreasing raster step size. Conclusion: MCNP modelling was used to validate a prototype BSX imager and introduce the RSPD concept, allowing for selective object-plane imaging. There was very good visual agreement between the experimental and MCNP imaging. Beyond optimizing system parameters for the existing prototype, new geometries can be investigated for volumetric image acquisition in medical applications. This material is based upon work supported under an Integrated University Program Graduate Fellowship sponsored by the Department of Energy Office of Nuclear Energy.« less

Monte Carlo modeling of ion chamber performance using MCNP.

PubMed

Wallace, J D

2012-12-01

Ion Chambers have a generally flat energy response with some deviations at very low (<100 keV) and very high (>2 MeV) energies. Some improvements in the low energy response can be achieved through use of high atomic number gases, such as argon and xenon, and higher chamber pressures. This work looks at the energy response of high pressure xenon-filled ion chambers using the MCNP Monte Carlo package to develop geometric models of a commercially available high pressure ion chamber (HPIC). The use of the F6 tally as an estimator of the energy deposited in a region of interest per unit mass, and the underlying assumptions associated with its use are described. The effect of gas composition, chamber gas pressure, chamber wall thickness, and chamber holder wall thicknesses on energy response are investigated and reported. The predicted energy response curve for the HPIC was found to be similar to that reported by other investigators. These investigations indicate that improvements to flatten the overall energy response of the HPIC down to 70 keV could be achieved through use of 3 mm-thick stainless steel walls for the ion chamber.

Advanced Computational Methods for Monte Carlo Calculations

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

Brown, Forrest B.

This course is intended for graduate students who already have a basic understanding of Monte Carlo methods. It focuses on advanced topics that may be needed for thesis research, for developing new state-of-the-art methods, or for working with modern production Monte Carlo codes.

Comparison of GATE/GEANT4 with EGSnrc and MCNP for electron dose calculations at energies between 15 keV and 20 MeV.

PubMed

Maigne, L; Perrot, Y; Schaart, D R; Donnarieix, D; Breton, V

2011-02-07

The GATE Monte Carlo simulation platform based on the GEANT4 toolkit has come into widespread use for simulating positron emission tomography (PET) and single photon emission computed tomography (SPECT) imaging devices. Here, we explore its use for calculating electron dose distributions in water. Mono-energetic electron dose point kernels and pencil beam kernels in water are calculated for different energies between 15 keV and 20 MeV by means of GATE 6.0, which makes use of the GEANT4 version 9.2 Standard Electromagnetic Physics Package. The results are compared to the well-validated codes EGSnrc and MCNP4C. It is shown that recent improvements made to the GEANT4/GATE software result in significantly better agreement with the other codes. We furthermore illustrate several issues of general interest to GATE and GEANT4 users who wish to perform accurate simulations involving electrons. Provided that the electron step size is sufficiently restricted, GATE 6.0 and EGSnrc dose point kernels are shown to agree to within less than 3% of the maximum dose between 50 keV and 4 MeV, while pencil beam kernels are found to agree to within less than 4% of the maximum dose between 15 keV and 20 MeV.

Theory and methods for measuring the effective multiplication constant in ADS

NASA Astrophysics Data System (ADS)

Rugama Saez, Yolanda

2001-10-01

In the thesis an absolute measurements technique for the subcriticality determination is presented. The ADS is a hybrid system where a subcritical system is fed by a proton accelerator. There are different proposals to define an ADS, one is to use plutonium and minor actinides from power plants waste as fuel to be transmuted into non radioactive isotopes (transmuter/burner, ATW). Another proposal is to use a Th232-U233 cycle (Energy Amplifier), being that thorium is an interesting and abundant fertile isotope. The development of accelerator driven systems (ADS) requires the development of methods to monitor and control the subcriticality of this kind of system without interfering with its normal operation mode. With this finality, we have applied noise analysis techniques that allow us to characterise the system when it is operating. The method presented in this thesis is based on the stochastic neutron and photon transport theory that can be implemented by presently available neutron/photon transport codes. In this work, first we analyse the stochastic transport theory which has been applied to define a parameter to determine the subcritical reactivity monitoring measurements. Finally we give the main limitations and recommendations for these subcritical monitoring methodology. As a result of the theoretical methodology, done in the first part of this thesis, a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The first one, LAHET, simulates the spallation collisions and the high energy transport and the other, MCNP-DSP, is used to estimate the counting statistics from a neutron/photon ray counter in a fissile system, as well as the transport for neutron with energies less than 20 MeV. From the coupling of both codes we developed the LAHET/MCNP-DSP code which, has the capability to simulate the total process in the ADS from the proton interaction to the signal detector processing. In these simulations, we compute the cross power spectral densities between pairs of detectors located inside the system which, is defined as the measured parameter. From the comparison of the theoretical predictions with the Monte Carlo simulations, we obtain some practical and simple methods to determine the system multiplication constant. (Abstract shortened by UMI.)

NRMC - A GPU code for N-Reverse Monte Carlo modeling of fluids in confined media

NASA Astrophysics Data System (ADS)

Sánchez-Gil, Vicente; Noya, Eva G.; Lomba, Enrique

2017-08-01

NRMC is a parallel code for performing N-Reverse Monte Carlo modeling of fluids in confined media [V. Sánchez-Gil, E.G. Noya, E. Lomba, J. Chem. Phys. 140 (2014) 024504]. This method is an extension of the usual Reverse Monte Carlo method to obtain structural models of confined fluids compatible with experimental diffraction patterns, specifically designed to overcome the problem of slow diffusion that can appear under conditions of tight confinement. Most of the computational time in N-Reverse Monte Carlo modeling is spent in the evaluation of the structure factor for each trial configuration, a calculation that can be easily parallelized. Implementation of the structure factor evaluation in NVIDIA® CUDA so that the code can be run on GPUs leads to a speed up of up to two orders of magnitude.

Implementation and validation of collapsed cone superposition for radiopharmaceutical dosimetry of photon emitters

NASA Astrophysics Data System (ADS)

Sanchez-Garcia, Manuel; Gardin, Isabelle; Lebtahi, Rachida; Dieudonné, Arnaud

2015-10-01

Two collapsed cone (CC) superposition algorithms have been implemented for radiopharmaceutical dosimetry of photon emitters. The straight CC (SCC) superposition method uses a water energy deposition kernel (EDKw) for each electron, positron and photon components, while the primary and scatter CC (PSCC) superposition method uses different EDKw for primary and once-scattered photons. PSCC was implemented only for photons originating from the nucleus, precluding its application to positron emitters. EDKw are linearly scaled by radiological distance, taking into account tissue density heterogeneities. The implementation was tested on 100, 300 and 600 keV mono-energetic photons and 18F, 99mTc, 131I and 177Lu. The kernels were generated using the Monte Carlo codes MCNP and EGSnrc. The validation was performed on 6 phantoms representing interfaces between soft-tissues, lung and bone. The figures of merit were γ (3%, 3 mm) and γ (5%, 5 mm) criterions corresponding to the computation comparison on 80 absorbed doses (AD) points per phantom between Monte Carlo simulations and CC algorithms. PSCC gave better results than SCC for the lowest photon energy (100 keV). For the 3 isotopes computed with PSCC, the percentage of AD points satisfying the γ (5%, 5 mm) criterion was always over 99%. A still good but worse result was found with SCC, since at least 97% of AD-values verified the γ (5%, 5 mm) criterion, except a value of 57% for the 99mTc with the lung/bone interface. The CC superposition method for radiopharmaceutical dosimetry is a good alternative to Monte Carlo simulations while reducing computation complexity.

Implementation and validation of collapsed cone superposition for radiopharmaceutical dosimetry of photon emitters.

PubMed

Sanchez-Garcia, Manuel; Gardin, Isabelle; Lebtahi, Rachida; Dieudonné, Arnaud

2015-10-21

Two collapsed cone (CC) superposition algorithms have been implemented for radiopharmaceutical dosimetry of photon emitters. The straight CC (SCC) superposition method uses a water energy deposition kernel (EDKw) for each electron, positron and photon components, while the primary and scatter CC (PSCC) superposition method uses different EDKw for primary and once-scattered photons. PSCC was implemented only for photons originating from the nucleus, precluding its application to positron emitters. EDKw are linearly scaled by radiological distance, taking into account tissue density heterogeneities. The implementation was tested on 100, 300 and 600 keV mono-energetic photons and (18)F, (99m)Tc, (131)I and (177)Lu. The kernels were generated using the Monte Carlo codes MCNP and EGSnrc. The validation was performed on 6 phantoms representing interfaces between soft-tissues, lung and bone. The figures of merit were γ (3%, 3 mm) and γ (5%, 5 mm) criterions corresponding to the computation comparison on 80 absorbed doses (AD) points per phantom between Monte Carlo simulations and CC algorithms. PSCC gave better results than SCC for the lowest photon energy (100 keV). For the 3 isotopes computed with PSCC, the percentage of AD points satisfying the γ (5%, 5 mm) criterion was always over 99%. A still good but worse result was found with SCC, since at least 97% of AD-values verified the γ (5%, 5 mm) criterion, except a value of 57% for the (99m)Tc with the lung/bone interface. The CC superposition method for radiopharmaceutical dosimetry is a good alternative to Monte Carlo simulations while reducing computation complexity.

Monte Carlo simulation of gamma-ray interactions in an over-square high-purity germanium detector for in-vivo measurements

NASA Astrophysics Data System (ADS)

Saizu, Mirela Angela

2016-09-01

The developments of high-purity germanium detectors match very well the requirements of the in-vivo human body measurements regarding the gamma energy ranges of the radionuclides intended to be measured, the shape of the extended radioactive sources, and the measurement geometries. The Whole Body Counter (WBC) from IFIN-HH is based on an “over-square” high-purity germanium detector (HPGe) to perform accurate measurements of the incorporated radionuclides emitting X and gamma rays in the energy range of 10 keV-1500 keV, under conditions of good shielding, suitable collimation, and calibration. As an alternative to the experimental efficiency calibration method consisting of using reference calibration sources with gamma energy lines that cover all the considered energy range, it is proposed to use the Monte Carlo method for the efficiency calibration of the WBC using the radiation transport code MCNP5. The HPGe detector was modelled and the gamma energy lines of 241Am, 57Co, 133Ba, 137Cs, 60Co, and 152Eu were simulated in order to obtain the virtual efficiency calibration curve of the WBC. The Monte Carlo method was validated by comparing the simulated results with the experimental measurements using point-like sources. For their optimum matching, the impact of the variation of the front dead layer thickness and of the detector photon absorbing layers materials on the HPGe detector efficiency was studied, and the detector’s model was refined. In order to perform the WBC efficiency calibration for realistic people monitoring, more numerical calculations were generated simulating extended sources of specific shape according to the standard man characteristics.

Investigation of Radiation Protection Methodologies for Radiation Therapy Shielding Using Monte Carlo Simulation and Measurement

NASA Astrophysics Data System (ADS)

Tanny, Sean

The advent of high-energy linear accelerators for dedicated medical use in the 1950's by Henry Kaplan and the Stanford University physics department began a revolution in radiation oncology. Today, linear accelerators are the standard of care for modern radiation therapy and can generate high-energy beams that can produce tens of Gy per minute at isocenter. This creates a need for a large amount of shielding material to properly protect members of the public and hospital staff. Standardized vault designs and guidance on shielding properties of various materials are provided by the National Council on Radiation Protection (NCRP) Report 151. However, physicists are seeking ways to minimize the footprint and volume of shielding material needed which leads to the use of non-standard vault configurations and less-studied materials, such as high-density concrete. The University of Toledo Dana Cancer Center has utilized both of these methods to minimize the cost and spatial footprint of the requisite radiation shielding. To ensure a safe work environment, computer simulations were performed to verify the attenuation properties and shielding workloads produced by a variety of situations where standard recommendations and guidance documents were insufficient. This project studies two areas of concern that are not addressed by NCRP 151, the radiation shielding workload for the vault door with a non-standard design, and the attenuation properties of high-density concrete for both photon and neutron radiation. Simulations have been performed using a Monte-Carlo code produced by the Los Alamos National Lab (LANL), Monte Carlo Neutrons, Photons 5 (MCNP5). Measurements have been performed using a shielding test port designed into the maze of the Varian Edge treatment vault.

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

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

Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

2015-01-01

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members acceptmore » the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).« less

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User’s Guide

DOE PAGES

Favorite, Jeffrey A.; Perko, Zoltan; Kiedrowski, Brian C.; ...

2017-03-01

The ability to perform sensitivity analyses using adjoint-based first-order sensitivity theory has existed for decades. This paper provides guidance on how adjoint sensitivity methods can be used to predict the effect of material density and composition uncertainties in critical experiments, including when these uncertain parameters are correlated or constrained. Two widely used Monte Carlo codes, MCNP6 (Ref. 2) and SCALE 6.2 (Ref. 3), are both capable of computing isotopic density sensitivities in continuous energy and angle. Additionally, Perkó et al. have shown how individual isotope density sensitivities, easily computed using adjoint methods, can be combined to compute constrained first-order sensitivitiesmore » that may be used in the uncertainty analysis. This paper provides details on how the codes are used to compute first-order sensitivities and how the sensitivities are used in an uncertainty analysis. Constrained first-order sensitivities are computed in a simple example problem.« less

Cross section of the 197Au(n,2n)196Au reaction

NASA Astrophysics Data System (ADS)

Kalamara, A.; Vlastou, R.; Kokkoris, M.; Diakaki, M.; Serris, M.; Patronis, N.; Axiotis, M.; Lagoyannis, A.

2017-09-01

The 197Au(n,2n)196Au reaction cross section has been measured at two energies, namely at 17.1 MeV and 20.9 MeV, by means of the activation technique, relative to the 27Al(n,α)24Na reference reaction cross section. Quasi-monoenergetic neutron beams were produced at the 5.5 MV Tandem T11/25 accelerator laboratory of NCSR "Demokritos", by means of the 3H(d,n)4He reaction, implementing a new Ti-tritiated target of ˜ 400 GBq activity. The induced γ-ray activity at the targets and reference foils has been measured with HPGe detectors. The cross section for the population of the second isomeric (12-) state m2 of 196Au was independently determined. Auxiliary Monte Carlo simulations were performed using the MCNP code. The present results are in agreement with previous experimental data and with theoretical calculations of the measured reaction cross sections, which were carried out with the use of the EMPIRE code.

Intrinsic Radiation Source Generation with the ISC Package: Data Comparisons and Benchmarking

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

Solomon, Clell J. Jr.

The characterization of radioactive emissions from unstable isotopes (intrinsic radiation) is necessary for shielding and radiological-dose calculations from radioactive materials. While most radiation transport codes, e.g., MCNP [X-5 Monte Carlo Team, 2003], provide the capability to input user prescribed source definitions, such as radioactive emissions, they do not provide the capability to calculate the correct radioactive-source definition given the material compositions. Special modifications to MCNP have been developed in the past to allow the user to specify an intrinsic source, but these modification have not been implemented into the primary source base [Estes et al., 1988]. To facilitate the descriptionmore » of the intrinsic radiation source from a material with a specific composition, the Intrinsic Source Constructor library (LIBISC) and MCNP Intrinsic Source Constructor (MISC) utility have been written. The combination of LIBISC and MISC will be herein referred to as the ISC package. LIBISC is a statically linkable C++ library that provides the necessary functionality to construct the intrinsic-radiation source generated by a material. Furthermore, LIBISC provides the ability use different particle-emission databases, radioactive-decay databases, and natural-abundance databases allowing the user flexibility in the specification of the source, if one database is preferred over others. LIBISC also provides functionality for aging materials and producing a thick-target bremsstrahlung photon source approximation from the electron emissions. The MISC utility links to LIBISC and facilitates the description of intrinsic-radiation sources into a format directly usable with the MCNP transport code. Through a series of input keywords and arguments the MISC user can specify the material, age the material if desired, and produce a source description of the radioactive emissions from the material in an MCNP readable format. Further details of using the MISC utility can be obtained from the user guide [Solomon, 2012]. The remainder of this report presents a discussion of the databases available to LIBISC and MISC, a discussion of the models employed by LIBISC, a comparison of the thick-target bremsstrahlung model employed, a benchmark comparison to plutonium and depleted-uranium spheres, and a comparison of the available particle-emission databases.« less

MCNP6 unstructured mesh application to estimate the photoneutron distribution and induced activity inside a linac bunker

NASA Astrophysics Data System (ADS)

Juste, B.; Morató, S.; Miró, R.; Verdú, G.; Díez, S.

2017-08-01

Unwanted neutrons in radiation therapy treatments are typically generated by photonuclear reactions. High-energy beams emitted by medical Linear Accelerators (LinAcs) interact with high atomic number materials situated in the accelerator head and release neutrons. Since neutrons have a high relative biological effectiveness, even low neutron doses may imply significant exposure of patients. It is also important to study radioactivity induced by these photoneutrons when interacting with the different materials and components of the treatment head facility and the shielding room walls, since persons not present during irradiation (e.g. medical staff) may be exposed to them even when the accelerator is not operating. These problems are studied in this work in order to contribute to challenge the radiation protection in these treatment locations. The work has been performed by simulation using the latest state of the art of Monte-Carlo computer code MCNP6. To that, a detailed model of particles transport inside the bunker and treatment head has been carried out using a meshed geometry model. The LinAc studied is an Elekta Precise accelerator with a treatment photon energy of 15 MeV used at the Hospital Clinic Universitari de Valencia, Spain.

Testing of ENDF71x: A new ACE-formatted neutron data library based on ENDF/B-VII.1

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

Gardiner, S. J.; Conlin, J. L.; Kiedrowski, B. C.

The ENDF71x library [1] is the most thoroughly tested set of ACE-format data tables ever released by the Nuclear Data Team at Los Alamos National Laboratory (LANL). It is based on ENDF/B-VII. 1, the most recently released set of evaluated nuclear data files produced by the US Cross Section Evaluation Working Group (CSEWG). A variety of techniques were used to test and verify the ENDF7 1x library before its public release. These include the use of automated checking codes written by members of the Nuclear Data Team, visual inspections of key neutron data, MCNP6 calculations designed to test data formore » every included combination of isotope and temperature as comprehensively as possible, and direct comparisons between ENDF71x and previous ACE library releases. Visual inspection of some of the most important neutron data revealed energy balance problems and unphysical discontinuities in the cross sections for some nuclides. Doppler broadening of the total cross sections with increasing temperature was found to be qualitatively correct. Test calculations performed using MCNP prompted two modifications to the MCNP6 source code and also exposed bad secondary neutron yields for {sup 231,233}Pa that are present in both ENDF/B-VII.1 and ENDF/B-VII.0. A comparison of ENDF71x with its predecessor ACE library, ENDF70, showed that dramatic changes have been made in the neutron cross section data for a number of isotopes between ENDF/B-VII.0 and ENDF/B-VII.1. Based on the results of these verification tests and the validation tests performed by Kahler, et al. [2], the ENDF71x library is recommended for use in all Monte Carlo applications. (authors)« less

FW-CADIS Method for Global and Semi-Global Variance Reduction of Monte Carlo Radiation Transport Calculations

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

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

2014-01-01

This paper presents a new hybrid (Monte Carlo/deterministic) method for increasing the efficiency of Monte Carlo calculations of distributions, such as flux or dose rate distributions (e.g., mesh tallies), as well as responses at multiple localized detectors and spectra. This method, referred to as Forward-Weighted CADIS (FW-CADIS), is an extension of the Consistent Adjoint Driven Importance Sampling (CADIS) method, which has been used for more than a decade to very effectively improve the efficiency of Monte Carlo calculations of localized quantities, e.g., flux, dose, or reaction rate at a specific location. The basis of this method is the development ofmore » an importance function that represents the importance of particles to the objective of uniform Monte Carlo particle density in the desired tally regions. Implementation of this method utilizes the results from a forward deterministic calculation to develop a forward-weighted source for a deterministic adjoint calculation. The resulting adjoint function is then used to generate consistent space- and energy-dependent source biasing parameters and weight windows that are used in a forward Monte Carlo calculation to obtain more uniform statistical uncertainties in the desired tally regions. The FW-CADIS method has been implemented and demonstrated within the MAVRIC sequence of SCALE and the ADVANTG/MCNP framework. Application of the method to representative, real-world problems, including calculation of dose rate and energy dependent flux throughout the problem space, dose rates in specific areas, and energy spectra at multiple detectors, is presented and discussed. Results of the FW-CADIS method and other recently developed global variance reduction approaches are also compared, and the FW-CADIS method outperformed the other methods in all cases considered.« less

Analysis of neutron and gamma-ray streaming along the maze of NRCAM thallium production target room.

PubMed

Raisali, G; Hajiloo, N; Hamidi, S; Aslani, G

2006-08-01

Study of the shield performance of a thallium-203 production target room has been investigated in this work. Neutron and gamma-ray equivalent dose rates at various points of the maze are calculated by simulating the transport of streaming neutrons, and photons using Monte Carlo method. For determination of neutron and gamma-ray source intensities and their energy spectrum, we have applied SRIM 2003 and ALICE91 computer codes to Tl target and its Cu substrate for a 145 microA of 28.5 MeV protons beam. The MCNP/4C code has been applied with neutron source term in mode n p to consider both prompt neutrons and secondary gamma-rays. Then the code is applied for the prompt gamma-rays as the source term. The neutron-flux energy spectrum and equivalent dose rates for neutron and gamma-rays in various positions in the maze have been calculated. It has been found that the deviation between calculated and measured dose values along the maze is less than 20%.

Masters Thesis- Criticality Alarm System Design Guide with Accompanying Alarm System Development for the Radioisotope Production Laboratory in Richland, Washington

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

Greenfield, Bryce A.

2009-12-01

A detailed instructional manual was created to guide criticality safety engineers through the process of designing a criticality alarm system (CAS) for Department of Energy (DOE) hazard class 1 and 2 facilities. Regulatory and technical requirements were both addressed. A list of design tasks and technical subtasks are thoroughly analyzed to provide concise direction for how to complete the analysis. An example of the application of the design methodology, the Criticality Alarm System developed for the Radioisotope Production Laboratory (RPL) of Richland, Washington is also included. The analysis for RPL utilizes the Monte Carlo code MCNP5 for establishing detector coveragemore » in the facility. Significant improvements to the existing CAS were made that increase the reliability, transparency, and coverage of the system.« less

Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model

PubMed Central

Zhang, Xiaomin; Xie, Xiangdong; Qu, Decheng; Ning, Jing; Zhou, Hongmei; Pan, Jie; Yang, Guoshan

2016-01-01

A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10−9 MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal–ventral, ventral–dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 MeV. PMID:26661852

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.

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Depletion Calculations Based on Perturbations. Application to the Study of a Rep-Like Assembly at Beginning of Cycle with TRIPOLI-4®.

NASA Astrophysics Data System (ADS)

Dieudonne, Cyril; Dumonteil, Eric; Malvagi, Fausto; M'Backé Diop, Cheikh

2014-06-01

For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this paper we present a methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time the implementation of this method in the TRIPOLI-4® code will be discussed, as well as the precise calculation scheme a meme to bring important speed-up of the depletion calculation. Finally, this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes.

Some Experimental and Monte Carlo Investigations of the Plastic Scintillators for the Current Mode Measurements at Pulsed Neutron Sources

NASA Astrophysics Data System (ADS)

Rogov, A.; Pepyolyshev, Yu.; Carta, M.; d'Angelo, A.

Scintillation detector (SD) is widely used in neutron and gamma-spectrometry in a count mode. The organic scintillators for the count mode of the detector operation are investigated rather well. Usually, they are applied for measurement of amplitude and time distributions of pulses caused by single interaction events of neutrons or gamma's with scintillator material. But in a large area of scientific research scintillation detectors can alternatively be used on a current mode by recording the average current from the detector. For example,the measurements of the neutron pulse shape at the pulsed reactors or another pulsed neutron sources. So as to get a rather large volume of experimental data at pulsed neutron sources, it is necessary to use the current mode detector for registration of fast neutrons. Many parameters of the SD are changed with a transition from an accounting mode to current one. For example, the detector efficiency is different in counting and current modes. Many effects connected with time accuracy become substantial. Besides, for the registration of solely fast neutrons, as must be in many measurements, in the mixed radiation field of the pulsed neutron sources, SD efficiency has to be determined with a gamma-radiation shield present. Here is no calculations or experimental data on SD current mode operation up to now. The response functions of the detectors can be either measured in high-precision reference fields or calculated by a computer simulation. We have used the MCNP code [1] and carried out some experiments for investigation of the plastic performances in a current mode. There are numerous programs performing simulating similar to the MCNP code. For example, for neutrons there are [2-4], for photons - [5-8]. However, all known codes to use (SCINFUL, NRESP4, SANDYL, EGS49) have more stringent restrictions on the source, geometry and detector characteristics. In MCNP code a lot of these restrictions are absent and you need only to write special additions for proton and electron recoil and transfer energy to light output. These code modifications allow taking into account all processes in organic scintillator influence the light yield.

Dosimetric investigation of LDR brachytherapy ¹⁹²Ir wires by Monte Carlo and TPS calculations.

PubMed

Bozkurt, Ahmet; Acun, Hediye; Kemikler, Gonul

2013-01-01

The aim of this study was to investigate the dose rate distribution around (192)Ir wires used as radioactive sources in low-dose-rate brachytherapy applications. Monte Carlo modeling of a 0.3-mm diameter source and its surrounding water medium was performed for five different wire lengths (1-5 cm) using the MCNP software package. The computed dose rates per unit of air kerma at distances from 0.1 up to 10 cm away from the source were first verified with literature data sets. Then, the simulation results were compared with the calculations from the XiO CMS commercial treatment planning system. The study results were found to be in concordance with the treatment planning system calculations except for the shorter wires at close distances.

Monte Carlo Calculations of Polarized Microwave Radiation Emerging from Cloud Structures

NASA Technical Reports Server (NTRS)

Kummerow, Christian; Roberti, Laura

1998-01-01

The last decade has seen tremendous growth in cloud dynamical and microphysical models that are able to simulate storms and storm systems with very high spatial resolution, typically of the order of a few kilometers. The fairly realistic distributions of cloud and hydrometeor properties that these models generate has in turn led to a renewed interest in the three-dimensional microwave radiative transfer modeling needed to understand the effect of cloud and rainfall inhomogeneities upon microwave observations. Monte Carlo methods, and particularly backwards Monte Carlo methods have shown themselves to be very desirable due to the quick convergence of the solutions. Unfortunately, backwards Monte Carlo methods are not well suited to treat polarized radiation. This study reviews the existing Monte Carlo methods and presents a new polarized Monte Carlo radiative transfer code. The code is based on a forward scheme but uses aliasing techniques to keep the computational requirements equivalent to the backwards solution. Radiative transfer computations have been performed using a microphysical-dynamical cloud model and the results are presented together with the algorithm description.

MO-AB-BRA-02: A Novel Scatter Imaging Modality for Real-Time Image Guidance During Lung SBRT

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

Redler, G; Bernard, D; Templeton, A

2015-06-15

Purpose: A novel scatter imaging modality is developed and its feasibility for image-guided radiation therapy (IGRT) during stereotactic body radiation therapy (SBRT) for lung cancer patients is assessed using analytic and Monte Carlo models as well as experimental testing. Methods: During treatment, incident radiation interacts and scatters from within the patient. The presented methodology forms an image of patient anatomy from the scattered radiation for real-time localization of the treatment target. A radiographic flat panel-based pinhole camera provides spatial information regarding the origin of detected scattered radiation. An analytical model is developed, which provides a mathematical formalism for describing themore » scatter imaging system. Experimental scatter images are acquired by irradiating an object using a Varian TrueBeam accelerator. The differentiation between tissue types is investigated by imaging simple objects of known compositions (water, lung, and cortical bone equivalent). A lung tumor phantom, simulating materials and geometry encountered during lung SBRT treatments, is fabricated and imaged to investigate image quality for various quantities of delivered radiation. Monte Carlo N-Particle (MCNP) code is used for validation and testing by simulating scatter image formation using the experimental pinhole camera setup. Results: Analytical calculations, MCNP simulations, and experimental results when imaging the water, lung, and cortical bone equivalent objects show close agreement, thus validating the proposed models and demonstrating that scatter imaging differentiates these materials well. Lung tumor phantom images have sufficient contrast-to-noise ratio (CNR) to clearly distinguish tumor from surrounding lung tissue. CNR=4.1 and CNR=29.1 for 10MU and 5000MU images (equivalent to 0.5 and 250 second images), respectively. Conclusion: Lung SBRT provides favorable treatment outcomes, but depends on accurate target localization. A comprehensive approach, employing multiple simulation techniques and experiments, is taken to demonstrate the feasibility of a novel scatter imaging modality for the necessary real-time image guidance.« less

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.

Novel low-kVp beamlet system for choroidal melanoma

PubMed Central

Esquivel, Carlos; Fuller, Clifton D; Waggener, Robert G; Wong, Adrian; Meltz, Martin; Blough, Melissa; Eng, Tony Y; Thomas, Charles R

2006-01-01

Background Treatment of choroidal melanoma with radiation often involves placement of customized brachytherapy eye-plaques. However, the dosimetric properties inherent in source-based radiotherapy preclude facile dose optimization to critical ocular structures. Consequently, we have constructed a novel system for utilizing small beam low-energy radiation delivery, the Beamlet Low-kVp X-ray, or "BLOKX" system. This technique relies on an isocentric rotational approach to deliver dose to target volumes within the eye, while potentially sparing normal structures. Methods Monte Carlo N-Particle (MCNP) transport code version 5.0(14) was used to simulate photon interaction with normal and tumor tissues within modeled right eye phantoms. Five modeled dome-shaped tumors with a diameter and apical height of 8 mm and 6 mm, respectively, were simulated distinct positions with respect to the macula iteratively. A single fixed 9 × 9 mm2 beamlet, and a comparison COMS protocol plaque containing eight I-125 seeds (apparent activity of 8 mCi) placed on the scleral surface of the eye adjacent to the tumor, were utilized to determine dosimetric parameters at tumor and adjacent tissues. After MCNP simulation, comparison of dose distribution at each of the 5 tumor positions for each modality (BLOKX vs. eye-plaque) was performed. Results Tumor-base doses ranged from 87.1–102.8 Gy for the BLOKX procedure, and from 335.3–338.6 Gy for the eye-plaque procedure. A reduction of dose of at least 69% to tumor base was noted when using the BLOKX. The BLOKX technique showed a significant reduction of dose, 89.8%, to the macula compared to the episcleral plaque. A minimum 71.0 % decrease in dose to the optic nerve occurred when the BLOKX was used. Conclusion The BLOKX technique allows more favorable dose distribution in comparison to standard COMS brachytherapy, as simulated using a Monte Carlo iterative mathematical modeling. Future series to determine clinical utility of such an approach are warranted. PMID:16965624

A method of estimating conceptus doses resulting from multidetector CT examinations during all stages of gestation

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

Damilakis, John; Tzedakis, Antonis; Perisinakis, Kostas

Purpose: Current methods for the estimation of conceptus dose from multidetector CT (MDCT) examinations performed on the mother provide dose data for typical protocols with a fixed scan length. However, modified low-dose imaging protocols are frequently used during pregnancy. The purpose of the current study was to develop a method for the estimation of conceptus dose from any MDCT examination of the trunk performed during all stages of gestation. Methods: The Monte Carlo N-Particle (MCNP) radiation transport code was employed in this study to model the Siemens Sensation 16 and Sensation 64 MDCT scanners. Four mathematical phantoms were used, simulatingmore » women at 0, 3, 6, and 9 months of gestation. The contribution to the conceptus dose from single simulated scans was obtained at various positions across the phantoms. To investigate the effect of maternal body size and conceptus depth on conceptus dose, phantoms of different sizes were produced by adding layers of adipose tissue around the trunk of the mathematical phantoms. To verify MCNP results, conceptus dose measurements were carried out by means of three physical anthropomorphic phantoms, simulating pregnancy at 0, 3, and 6 months of gestation and thermoluminescence dosimetry (TLD) crystals. Results: The results consist of Monte Carlo-generated normalized conceptus dose coefficients for single scans across the four mathematical phantoms. These coefficients were defined as the conceptus dose contribution from a single scan divided by the CTDI free-in-air measured with identical scanning parameters. Data have been produced to take into account the effect of maternal body size and conceptus position variations on conceptus dose. Conceptus doses measured with TLD crystals showed a difference of up to 19% compared to those estimated by mathematical simulations. Conclusions: Estimation of conceptus doses from MDCT examinations of the trunk performed on pregnant patients during all stages of gestation can be made using the method developed in the current study.« less

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.

SKIRT: The design of a suite of input models for Monte Carlo radiative transfer simulations

NASA Astrophysics Data System (ADS)

Baes, M.; Camps, P.

2015-09-01

The Monte Carlo method is the most popular technique to perform radiative transfer simulations in a general 3D geometry. The algorithms behind and acceleration techniques for Monte Carlo radiative transfer are discussed extensively in the literature, and many different Monte Carlo codes are publicly available. On the contrary, the design of a suite of components that can be used for the distribution of sources and sinks in radiative transfer codes has received very little attention. The availability of such models, with different degrees of complexity, has many benefits. For example, they can serve as toy models to test new physical ingredients, or as parameterised models for inverse radiative transfer fitting. For 3D Monte Carlo codes, this requires algorithms to efficiently generate random positions from 3D density distributions. We describe the design of a flexible suite of components for the Monte Carlo radiative transfer code SKIRT. The design is based on a combination of basic building blocks (which can be either analytical toy models or numerical models defined on grids or a set of particles) and the extensive use of decorators that combine and alter these building blocks to more complex structures. For a number of decorators, e.g. those that add spiral structure or clumpiness, we provide a detailed description of the algorithms that can be used to generate random positions. Advantages of this decorator-based design include code transparency, the avoidance of code duplication, and an increase in code maintainability. Moreover, since decorators can be chained without problems, very complex models can easily be constructed out of simple building blocks. Finally, based on a number of test simulations, we demonstrate that our design using customised random position generators is superior to a simpler design based on a generic black-box random position generator.

Results for Phase I of the IAEA Coordinated Research Program on HTGR Uncertainties

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

Strydom, Gerhard; Bostelmann, Friederike; Yoon, Su Jong

2015-01-01

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied. High Temperature Gas-cooled Reactors (HTGR) has its own peculiarities, coated particle design, large graphite quantities, different materials and high temperatures that also require other simulationmore » requirements. The IAEA has therefore launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the HTR-PM (INET, China). This report summarizes the contributions of the HTGR Methods Simulation group at Idaho National Laboratory (INL) up to this point of the CRP. The activities at INL have been focused so far on creating the problem specifications for the prismatic design, as well as providing reference solutions for the exercises defined for Phase I. An overview is provided of the HTGR UAM objectives and scope, and the detailed specifications for Exercises I-1, I-2, I-3 and I-4 are also included here for completeness. The main focus of the report is the compilation and discussion of reference results for Phase I (i.e. for input parameters at their nominal or best-estimate values), which is defined as the first step of the uncertainty quantification process. These reference results can be used by other CRP participants for comparison with other codes or their own reference results. The status on the Monte Carlo modeling of the experimental VHTRC facility is also discussed. Reference results were obtained for the neutronics stand-alone cases (Ex. I-1 and Ex. I-2) using the (relatively new) Monte Carlo code Serpent, and comparisons were performed with the more established Monte Carlo codes MCNP and KENO-VI. For the thermal-fluids stand-alone cases (Ex. I-3 and I-4) the commercial CFD code CFX was utilized to obtain reference results that can be compared with lower fidelity tools.« less

Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part I: Benchmark comparisons of WIMS-D5 and DRAGON cell and control rod parameters with MCNP5

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

Mollerach, R.; Leszczynski, F.; Fink, J.

2006-07-01

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure-vessel design with 451 vertical coolant channels, and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update calculation methods and models (cell, supercell and reactor) was recently carried out coveringmore » cell, supercell (control rod) and core calculations. As a validation of the new models some benchmark comparisons were done with Monte Carlo calculations with MCNP5. This paper presents comparisons of cell and supercell benchmark problems based on a slightly idealized model of the Atucha-I core obtained with the WIMS-D5 and DRAGON codes with MCNP5 results. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, and more symmetric than Atucha-II Cell parameters compared include cell k-infinity, relative power levels of the different rings of fuel rods, and some two-group macroscopic cross sections. Supercell comparisons include supercell k-infinity changes due to the control rods (tubes) of steel and hafnium. (authors)« less

PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

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

Matthew Ellis; Derek Gaston; Benoit Forget

In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes.more » An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.« less

WARP

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

Bergmann, Ryan M.; Rowland, Kelly L.

2017-04-12

WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed at UC Berkeley to efficiently execute on NVIDIA graphics processing unit (GPU) platforms. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo method, namely, that very few physical and geometrical simplifications are applied. WARP is able to calculate multiplication factors, neutron flux distributions (in both space and energy), and fission source distributions for time-independent neutron transport problems. It can run in both criticality or fixed source modes, but fixed source mode is currentlymore » not robust, optimized, or maintained in the newest version. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. The goal of developing WARP is to investigate algorithms that can grow into a full-featured, continuous energy, Monte Carlo neutron transport code that is accelerated by running on GPUs. The crux of the effort is to make Monte Carlo calculations faster while producing accurate results. Modern supercomputers are commonly being built with GPU coprocessor cards in their nodes to increase their computational efficiency and performance. GPUs execute efficiently on data-parallel problems, but most CPU codes, including those for Monte Carlo neutral particle transport, are predominantly task-parallel. WARP uses a data-parallel neutron transport algorithm to take advantage of the computing power GPUs offer.« less

SU-E-T-297: Dosimetric Assessment of An Air-Filled Balloon Applicator in HDR Vaginal Cuff Brachytherapy Using the Monte Carlo Method

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

Jiang, H; Lee, Y; Pokhrel, D

2015-06-15

Purpose: As an alternative to cylindrical applicators, air inflated balloon applicators have been introduced into HDR vaginal cuff brachytherapy treatment to achieve sufficient dose to vagina mucosa as well as to spare rectum and bladder. In general, TG43 formulae based treatment planning systems do not take into account tissue inhomogeneity, and air in the balloon applicator can cause higher delivered dose to mucosa than treatment plan reported. We investigated dosimetric effect of air in balloon applicator using the Monte Carlo method. Methods: The thirteen-catheter Capri applicator with a Nucletron Ir-192 seed was modeled for various balloon diameters (2cm to 3.5cm)more » using the MCNP Monte Carlo code. Ir-192 seed was placed in both central and peripheral catheters to replicate real patient situations. Existence of charged particle equilibrium (CPE) with air balloon was evaluated by comparing kerma and dose at various distances (1mm to 70mm) from surface of air-filled applicator. Also mucosa dose by an air-filled applicator was compared with by a water-filled applicator to evaluate dosimetry accuracy of planning system without tissue inhomogeneity correction. Results: Beyond 1mm from air/tissue interface, the difference between kerma and dose was within 2%. CPE (or transient CPE) condition was deemed existent, and in this region no electron transport was necessary in Monte Carlo simulations. At 1mm or less, the deviation of dose from kerma became more apparent. Increase of dose to mucosa depended on diameter of air balloon. The increment of dose to mucosa was 2.5% and 4.3% on average for 2cm and 3.5cm applicators, respectively. Conclusion: After introduction of air balloon applicator, CPE fails only at the proximity of air/tissue interface. Although dose to mucosa is increased, there is no significant dosimetric difference (<5%) between air and water filled applicators. Tissue inhomogeneity correction is not necessary for air-filled applicators.« less

A comparative evaluation of luminescence detectors: RPL-GD-301, TLD-100 and OSL-AL2O3:C, using Monte Carlo simulations

NASA Astrophysics Data System (ADS)

Benali, A.-H.; Medkour Ishak-Boushaki, G.; Nourreddine, A.-M.; Allab, M.; Papadimitroulas, P.

2017-07-01

The luminescent dosimeters are widely used in clinical practice, for the monitoring of patient dose in external radiation therapy. Three of the most common dosimeter categories are the thermoluminescence (TLDs), the radiophotoluminescence (RPLs) and the optically stimulated luminescence (OSLs), with similar physical processes on their properties. The aim of the present study is to compare and evaluate the dosimetric properties of three specific luminescent detectors namely: a) RPL glass dosimeter, commercially known as GD-301, b) lithium fluoride TLD-100 (LiF:Mg,Ti) and c) carbon-doped aluminum oxide (Al2O3:C). For this purpose, Monte Carlo simulations were applied, using the MCNP5 code to estimate the responses of these dosimeters in terms of absorbed dose, output factor, the angular and energy dependence. In the present study, we found that the differences between the output factors were less than ± 4.2% for all detector materials RPLGD, TLD and OSLD. The variations in sensitivity for angles up to ± 80 degrees from the central axis of the beam were approximately 0.5%, 0.8% and 1.5% for the TLD-100, GD-301 and Al2O3:C, respectively. The energy dependence of the RPL and OSL dosimeters are stated as less than a 2.2%, and within 5.8% for TLD.

Developing a cosmic ray muon sampling capability for muon tomography and monitoring applications

NASA Astrophysics Data System (ADS)

Chatzidakis, S.; Chrysikopoulou, S.; Tsoukalas, L. H.

2015-12-01

In this study, a cosmic ray muon sampling capability using a phenomenological model that captures the main characteristics of the experimentally measured spectrum coupled with a set of statistical algorithms is developed. The "muon generator" produces muons with zenith angles in the range 0-90° and energies in the range 1-100 GeV and is suitable for Monte Carlo simulations with emphasis on muon tomographic and monitoring applications. The muon energy distribution is described by the Smith and Duller (1959) [35] phenomenological model. Statistical algorithms are then employed for generating random samples. The inverse transform provides a means to generate samples from the muon angular distribution, whereas the Acceptance-Rejection and Metropolis-Hastings algorithms are employed to provide the energy component. The predictions for muon energies 1-60 GeV and zenith angles 0-90° are validated with a series of actual spectrum measurements and with estimates from the software library CRY. The results confirm the validity of the phenomenological model and the applicability of the statistical algorithms to generate polyenergetic-polydirectional muons. The response of the algorithms and the impact of critical parameters on computation time and computed results were investigated. Final output from the proposed "muon generator" is a look-up table that contains the sampled muon angles and energies and can be easily integrated into Monte Carlo particle simulation codes such as Geant4 and MCNP.

Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

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

Tippayakul, C.; Ivanov, K.; Misu, S.

2006-07-01

This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross sectionmore » library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)« less

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

PubMed

Iwatani, K; Hoshi, M; Shizuma, K; Hiraoka, M; Hayakawa, N; Oka, T; Hasai, H

1994-10-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

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

Amharrak, H.; Reynard-Carette, C.; Carette, M.

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR),more » under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)« less

Quantitative criteria for assessment of gamma-ray imager performance

NASA Astrophysics Data System (ADS)

Gottesman, Steve; Keller, Kristi; Malik, Hans

2015-08-01

In recent years gamma ray imagers such as the GammaCamTM and Polaris have demonstrated good imaging performance in the field. Imager performance is often summarized as "resolution", either angular, or spatial at some distance from the imager, however the definition of resolution is not always related to the ability to image an object. It is difficult to quantitatively compare imagers without a common definition of image quality. This paper examines three categories of definition: point source; line source; and area source. It discusses the details of those definitions and which ones are more relevant for different situations. Metrics such as Full Width Half Maximum (FWHM), variations on the Rayleigh criterion, and some analogous to National Imagery Interpretability Rating Scale (NIIRS) are discussed. The performance against these metrics is evaluated for a high resolution coded aperture imager modeled using Monte Carlo N-Particle (MCNP), and for a medium resolution imager measured in the lab.

The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

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

Eccleston, G.W.; Menlove, H.O.; Abhold, M.

1998-12-31

The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 tomore » 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.« less

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

NASA Astrophysics Data System (ADS)

Zhu, Qingjun; Li, Jia; Liu, Songlin

2016-07-01

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

In-vivo assessment of total body protein in rats by prompt-γ neutron activation analysis

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Boozer, Carol N.; Ma, Ruimei; Yasumura, Seiichi

1997-02-01

A prompt-(gamma) neutron activation analysis facility for in vivo determination of total body protein (TBP) in rats has been designed. TBP is determined in vivo by assessment of total body nitrogen. The facility is based on a 252Cf radionuclide neutron source within a heavy water moderator assembly and two NaI(Tl) scintillation detectors. The in vivo precision of the technique, as estimated by three repeated measurements of 15 rats is 6 percent, for a radiation dose equivalent of 60 mSv. The radiation dose per measurement is sufficiently low to enable serial measurements on the same animal. MCNP-4A Monte Carlo transport code was utilized to calculate thermal neutron flux correction factors to account for differences in size and shape of the rats and calibration phantoms. Good agrement was observed in comparing body nitrogen assessment by prompt-(gamma) neutron activation and chemical carcass analysis.

Reusable shielding material for neutron- and gamma-radiation

NASA Astrophysics Data System (ADS)

Calzada, Elbio; Grünauer, Florian; Schillinger, Burkhard; Türck, Harald

2011-09-01

At neutron research facilities all around the world radiation shieldings are applied to reduce the background of neutron and gamma radiation as far as possible in order to perform high quality measurements and to fulfill the radiation protection requirements. The current approach with cement-based compounds has a number of shortcomings: "Heavy concrete" contains a high amount of elements, which are not desired to obtain a high attenuation of neutron and/or gamma radiation (e.g. calcium, carbon, oxygen, silicon and aluminum). A shielding material with a high density of desired nuclei such as iron, hydrogen and boron was developed for the redesign of the neutron radiography facility ANTARES at beam tube 4 (located at a cold neutron source) of FRM-II. The composition of the material was optimized by help of the Monte Carlo code MCNP5. With this shielding material a considerable higher attenuation of background radiation can be obtained compared to usual heavy concretes.

Tritium assay of Li/sub 2/O in the LBM/LOTUS experiments

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

Quanci, J.; Azam, S.; Bertone, P.

1986-11-01

The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM.more » PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results for the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system.« less

Contrast cancellation technique applied to digital x-ray imaging using silicon strip detectors

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

Avila, C.; Lopez, J.; Sanabria, J. C.

2005-12-15

Dual-energy mammographic imaging experimental tests have been performed using a compact dichromatic imaging system based on a conventional x-ray tube, a mosaic crystal, and a 384-strip silicon detector equipped with full-custom electronics with single photon counting capability. For simulating mammal tissue, a three-component phantom, made of Plexiglass, polyethylene, and water, has been used. Images have been collected with three different pairs of x-ray energies: 16-32 keV, 18-36 keV, and 20-40 keV. A Monte Carlo simulation of the experiment has also been carried out using the MCNP-4C transport code. The Alvarez-Macovski algorithm has been applied both to experimental and simulated datamore » to remove the contrast between two of the phantom materials so as to enhance the visibility of the third one.« less

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Monte Carlo simulations of medical imaging modalities

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

Estes, G.P.

Because continuous-energy Monte Carlo radiation transport calculations can be nearly exact simulations of physical reality (within data limitations, geometric approximations, transport algorithms, etc.), it follows that one should be able to closely approximate the results of many experiments from first-principles computations. This line of reasoning has led to various MCNP studies that involve simulations of medical imaging modalities and other visualization methods such as radiography, Anger camera, computerized tomography (CT) scans, and SABRINA particle track visualization. It is the intent of this paper to summarize some of these imaging simulations in the hope of stimulating further work, especially as computermore » power increases. Improved interpretation and prediction of medical images should ultimately lead to enhanced medical treatments. It is also reasonable to assume that such computations could be used to design new or more effective imaging instruments.« less

Analysis of Naval Ammunition Stock Positioning

DTIC Science & Technology

2015-12-01

model takes once the Monte -Carlo simulation determines the assigned probabilities for site-to-site locations. Column two shows how the simulation...stockpiles and positioning them at coastal Navy facilities. A Monte -Carlo simulation model was developed to simulate expected cost and delivery...TERMS supply chain management, Monte -Carlo simulation, risk, delivery performance, stock positioning 15. NUMBER OF PAGES 85 16. PRICE CODE 17

ME(SSY)**2: Monte Carlo Code for Star Cluster Simulations

NASA Astrophysics Data System (ADS)

Freitag, Marc Dewi

2013-02-01

ME(SSY)**2 stands for “Monte-carlo Experiments with Spherically SYmmetric Stellar SYstems." This code simulates the long term evolution of spherical clusters of stars; it was devised specifically to treat dense galactic nuclei. It is based on the pioneering Monte Carlo scheme proposed by Hénon in the 70's and includes all relevant physical ingredients (2-body relaxation, stellar mass spectrum, collisions, tidal disruption, ldots). It is basically a Monte Carlo resolution of the Fokker-Planck equation. It can cope with any stellar mass spectrum or velocity distribution. Being a particle-based method, it also allows one to take stellar collisions into account in a very realistic way. This unique code, featuring most important physical processes, allows million particle simulations, spanning a Hubble time, in a few CPU days on standard personal computers and provides a wealth of data only rivalized by N-body simulations. The current version of the software requires the use of routines from the "Numerical Recipes in Fortran 77" (http://www.nrbook.com/a/bookfpdf.php).

Force field development with GOMC, a fast new Monte Carlo molecular simulation code

NASA Astrophysics Data System (ADS)

Mick, Jason Richard

In this work GOMC (GPU Optimized Monte Carlo) a new fast, flexible, and free molecular Monte Carlo code for the simulation atomistic chemical systems is presented. The results of a large Lennard-Jonesium simulation in the Gibbs ensemble is presented. Force fields developed using the code are also presented. To fit the models a quantitative fitting process is outlined using a scoring function and heat maps. The presented n-6 force fields include force fields for noble gases and branched alkanes. These force fields are shown to be the most accurate LJ or n-6 force fields to date for these compounds, capable of reproducing pure fluid behavior and binary mixture behavior to a high degree of accuracy.

Monte Carlo simulation of proton track structure in biological matter

DOE PAGES

Quinto, Michele A.; Monti, Juan M.; Weck, Philippe F.; ...

2017-05-25

Here, understanding the radiation-induced effects at the cellular and subcellular levels remains crucial for predicting the evolution of irradiated biological matter. In this context, Monte Carlo track-structure simulations have rapidly emerged among the most suitable and powerful tools. However, most existing Monte Carlo track-structure codes rely heavily on the use of semi-empirical cross sections as well as water as a surrogate for biological matter. In the current work, we report on the up-to-date version of our homemade Monte Carlo code TILDA-V – devoted to the modeling of the slowing-down of 10 keV–100 MeV protons in both water and DNA –more » where the main collisional processes are described by means of an extensive set of ab initio differential and total cross sections.« less

Monte Carlo simulation of proton track structure in biological matter

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

Quinto, Michele A.; Monti, Juan M.; Weck, Philippe F.

Here, understanding the radiation-induced effects at the cellular and subcellular levels remains crucial for predicting the evolution of irradiated biological matter. In this context, Monte Carlo track-structure simulations have rapidly emerged among the most suitable and powerful tools. However, most existing Monte Carlo track-structure codes rely heavily on the use of semi-empirical cross sections as well as water as a surrogate for biological matter. In the current work, we report on the up-to-date version of our homemade Monte Carlo code TILDA-V – devoted to the modeling of the slowing-down of 10 keV–100 MeV protons in both water and DNA –more » where the main collisional processes are described by means of an extensive set of ab initio differential and total cross sections.« less

D-T Neutron Skyshine Experiments at JAERI/FNS

NASA Astrophysics Data System (ADS)

Nishitani, Takeo; Ochiai, Kentaro; Yoshida, Shigeo; Tanaka, Ryohei; Wakisaka, Masashi; Nakao, Makoto; Sato, Satoshi; Yamauchi, Michinori; Hori, Jun-Ichi; Takahashi, Akito; Kaneko, Jun-Ichi; Sawamura, Teruko

The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ˜1.7×1011n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9 × 0.9 m2 was open during the experimental period.The radiation dose rate outside the target room was measured as far as about 550 m away from the D-T target point with a spherical rem-counter. The highest neutron dose was about 0.5 μSv/hr at a distance of 30 m from the D-T target point and the dose rate was attenuated to 0.002 μSv/hr at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 250 m. The neutron spectra were evaluated with a 3He detector with different thickness of polyethylene neutron moderators. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation detectors.

Fast fission neutron detection using the Cherenkov effect

NASA Astrophysics Data System (ADS)

Millard, Matthew James

The Cherenkov effect in optically clear media of varying indices of refraction and composition was investigated for quantification of fast neutrons. The ultimate application of the proposed detection system is criticality monitoring. The optically clear medium, composed of select target nuclei, was coupled to a photomultiplier tube. Neutron reaction products of the target nuclei contained within the optical medium emit beta particles and gamma rays that produce Cherenkov photons within the medium which can be detected. Assessed media include quartz (SiO2), sapphire (Al2O3), spinel (MgAl2O4), and zinc sulfide (ZnS), which were irradiated with un-moderated 252Cf. Monte Carlo N-Particle (MCNP) code simulations were conducted to quantify the neutron flux incident on the media. High resolution gamma-ray spectroscopic measurements of the samples were conducted to verify the MCNP estimate. The threshold reactions of interest were 28Si (n, p) 28Al, 27 Al (n, p) 27Mg, 24Mg(n, p)24 Na, and 64Zn(n, p)64Cu which have neutron reaction cross sections in the 1 to 10 MeV range on the order of 0.1 barn. The detection system offers a unique way to measure a criticality event; it can count in place, making retrieval by emergency personnel unnecessary.

Data and methods to estimate fetal dose from fluoroscopically guided prophylactic hypogastric artery balloon occlusion

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

Solomou, G.; Stratakis, J.; Perisinakis, K.

Purpose: To provide data for estimation of fetal radiation dose (D{sub F}) from prophylactic hypogastric artery balloon occlusion (HABO) procedures. Methods: The Monte-Carlo-N-particle (MCNP) transport code and mathematical phantoms representing a pregnant patient at the ninth month of gestation were employed. PA, RAO 20° and LAO 20° fluoroscopy projections of left and right internal iliac arteries were simulated. Projection-specific normalized fetal dose (NFD) data were produced for various beam qualities. The effects of projection angle, x-ray field location relative to the fetus, field size, maternal body size, and fetal size on NFD were investigated. Presented NFD values were compared tomore » corresponding values derived using a physical anthropomorphic phantom simulating pregnancy at the third trimester and thermoluminescence dosimeters. Results: NFD did not considerably vary when projection angle was altered by ±5°, whereas it was found to markedly depend on tube voltage, filtration, x-ray field location and size, and maternal body size. Differences in NFD < 7.5% were observed for naturally expected variations in fetal size. A difference of less than 13.5% was observed between NFD values estimated by MCNP and direct measurements. Conclusions: Data and methods provided allow for reliable estimation of radiation burden to the fetus from HABO.« less

An MCNP-based model for the evaluation of the photoneutron dose in high energy medical electron accelerators.

PubMed

Carinou, Eleutheria; Stamatelatos, Ion Evangelos; Kamenopoulou, Vassiliki; Georgolopoulou, Paraskevi; Sandilos, Panayotis

The development of a computational model for the treatment head of a medical electron accelerator (Elekta/Philips SL-18) by the Monte Carlo code mcnp-4C2 is discussed. The model includes the major components of the accelerator head and a pmma phantom representing the patient body. Calculations were performed for a 14 MeV electron beam impinging on the accelerator target and a 10 cmx10 cm beam area at the isocentre. The model was used in order to predict the neutron ambient dose equivalent at the isocentre level and moreover the neutron absorbed dose distribution within the phantom. Calculations were validated against experimental measurements performed by gold foil activation detectors. The results of this study indicated that the equivalent dose at tissues or organs adjacent to the treatment field due to photoneutrons could be up to 10% of the total peripheral dose, for the specific accelerator characteristics examined. Therefore, photoneutrons should be taken into account when accurate dose calculations are required to sensitive tissues that are adjacent to the therapeutic X-ray beam. The method described can be extended to other accelerators and collimation configurations as well, upon specification of treatment head component dimensions, composition and nominal accelerating potential.

Effective dose in the manufacturing process of rutile covered welding electrodes.

PubMed

Herranz, M; Rozas, S; Pérez, C; Idoeta, R; Núñez-Lagos, R; Legarda, F

2013-03-01

Shielded metal arc welding using covered electrodes is the most common welding process. Sometimes the covering contains naturally occurring radioactive materials (NORMs). In Spain the most used electrodes are those covered with rutile mixed with other materials. Rutile contains some detectable natural radionuclides, so it can be considered a NORM. This paper mainly focuses on the use of MCNP (Monte Carlo N-Particle Transport Code) as a predictive tool to obtain doses in a factory which produces this type of electrode and assess the radiological impact in a specific facility after estimating the internal dose.To do this, in the facility, areas of highest radiation and positions of workers were identified, radioactive content of rutile and rutile covered electrodes was measured, and, considering a worst possible scenario, external dose at working points has been calculated using MCNP. This procedure has been validated comparing the results obtained with those from a pressurised ionisation chamber and TLD dosimeters. The internal dose has been calculated using DCAL (dose and risk calculation). The doses range between 8.8 and 394 μSv yr(-1), always lower than the effective dose limit for the public, 1 mSv yr(-1). The highest dose corresponds to the mixing area.

Release of Continuous Representation for S(α,β) ACE Data

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

Conlin, Jeremy Lloyd; Parsons, Donald Kent

2014-03-20

For low energy neutrons, the default free gas model for scattering cross sections is not always appropriate. Molecular effects or crystalline structure effects can affect the neutron scattering cross sections. These effects are included in the S(α; β) thermal neutron scattering data and are tabulated in file 7 of the ENDF6 format files. S stands for scattering. α is a momentum transfer variable and is an energy transfer variable. The S(α; β) cross sections can include coherent elastic scattering (no E change for the neutron, but specific scattering angles), incoherent elastic scattering (no E change for the neutron, but continuousmore » scattering angles), and inelastic scattering (E change for the neutron, and change in angle as well). Every S(α; β) material will have inelastic scattering and may have either coherent or incoherent elastic scattering (but not both). Coherent elastic scattering cross sections have distinctive jagged-looking Bragg edges, whereas the other cross sections are much smoother. The evaluated files from the NNDC are processed locally in the THERMR module of NJOY. Data can be produced either for continuous energy Monte Carlo codes (using ACER) or embedded in multi-group cross sections for deterministic (or even multi-group Monte Carlo) codes (using GROUPR). Currently, the S(α; β) files available for MCNP use discrete energy changes for inelastic scattering. That is, the scattered neutrons can only be emitted at specific energies— rather than across a continuous spectrum of energies. The discrete energies are chosen to preserve the average secondary neutron energy, i.e., in an integral sense, but the discrete treatment does not preserve any differential quantities in energy or angle.« less

The Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE

NASA Astrophysics Data System (ADS)

Vandenbroucke, B.; Wood, K.

2018-04-01

We present the public Monte Carlo photoionization and moving-mesh radiation hydrodynamics code CMACIONIZE, which can be used to simulate the self-consistent evolution of HII regions surrounding young O and B stars, or other sources of ionizing radiation. The code combines a Monte Carlo photoionization algorithm that uses a complex mix of hydrogen, helium and several coolants in order to self-consistently solve for the ionization and temperature balance at any given type, with a standard first order hydrodynamics scheme. The code can be run as a post-processing tool to get the line emission from an existing simulation snapshot, but can also be used to run full radiation hydrodynamical simulations. Both the radiation transfer and the hydrodynamics are implemented in a general way that is independent of the grid structure that is used to discretize the system, allowing it to be run both as a standard fixed grid code, but also as a moving-mesh code.

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.

Preliminary estimates of nucleon fluxes in a water target exposed to solar-flare protons: BRYNTRN versus Monte Carlo code

NASA Technical Reports Server (NTRS)

Shinn, Judy L.; Wilson, John W.; Lone, M. A.; Wong, P. Y.; Costen, Robert C.

1994-01-01

A baryon transport code (BRYNTRN) has previously been verified using available Monte Carlo results for a solar-flare spectrum as the reference. Excellent results were obtained, but the comparisons were limited to the available data on dose and dose equivalent for moderate penetration studies that involve minor contributions from secondary neutrons. To further verify the code, the secondary energy spectra of protons and neutrons are calculated using BRYNTRN and LAHET (Los Alamos High-Energy Transport code, which is a Monte Carlo code). These calculations are compared for three locations within a water slab exposed to the February 1956 solar-proton spectrum. Reasonable agreement was obtained when various considerations related to the calculational techniques and their limitations were taken into account. Although the Monte Carlo results are preliminary, it appears that the neutron albedo, which is not currently treated in BRYNTRN, might be a cause for the large discrepancy seen at small penetration depths. It also appears that the nonelastic neutron production cross sections in BRYNTRN may underestimate the number of neutrons produced in proton collisions with energies below 200 MeV. The notion that the poor energy resolution in BRYNTRN may cause a large truncation error in neutron elastic scattering requires further study.

Experimental benchmarking of a Monte Carlo dose simulation code for pediatric CT

NASA Astrophysics Data System (ADS)

Li, Xiang; Samei, Ehsan; Yoshizumi, Terry; Colsher, James G.; Jones, Robert P.; Frush, Donald P.

2007-03-01

In recent years, there has been a desire to reduce CT radiation dose to children because of their susceptibility and prolonged risk for cancer induction. Concerns arise, however, as to the impact of dose reduction on image quality and thus potentially on diagnostic accuracy. To study the dose and image quality relationship, we are developing a simulation code to calculate organ dose in pediatric CT patients. To benchmark this code, a cylindrical phantom was built to represent a pediatric torso, which allows measurements of dose distributions from its center to its periphery. Dose distributions for axial CT scans were measured on a 64-slice multidetector CT (MDCT) scanner (GE Healthcare, Chalfont St. Giles, UK). The same measurements were simulated using a Monte Carlo code (PENELOPE, Universitat de Barcelona) with the applicable CT geometry including bowtie filter. The deviations between simulated and measured dose values were generally within 5%. To our knowledge, this work is one of the first attempts to compare measured radial dose distributions on a cylindrical phantom with Monte Carlo simulated results. It provides a simple and effective method for benchmarking organ dose simulation codes and demonstrates the potential of Monte Carlo simulation for investigating the relationship between dose and image quality for pediatric CT patients.

Monte Carlo modelling the dosimetric effects of electrode material on diamond detectors.

PubMed

Baluti, Florentina; Deloar, Hossain M; Lansley, Stuart P; Meyer, Juergen

2015-03-01

Diamond detectors for radiation dosimetry were modelled using the EGSnrc Monte Carlo code to investigate the influence of electrode material and detector orientation on the absorbed dose. The small dimensions of the electrode/diamond/electrode detector structure required very thin voxels and the use of non-standard DOSXYZnrc Monte Carlo model parameters. The interface phenomena was investigated by simulating a 6 MV beam and detectors with different electrode materials, namely Al, Ag, Cu and Au, with thickens of 0.1 µm for the electrodes and 0.1 mm for the diamond, in both perpendicular and parallel detector orientation with regards to the incident beam. The smallest perturbations were observed for the parallel detector orientation and Al electrodes (Z = 13). In summary, EGSnrc Monte Carlo code is well suited for modelling small detector geometries. The Monte Carlo model developed is a useful tool to investigate the dosimetric effects caused by different electrode materials. To minimise perturbations cause by the detector electrodes, it is recommended that the electrodes should be made from a low-atomic number material and placed parallel to the beam direction.

Parallel Monte Carlo transport modeling in the context of a time-dependent, three-dimensional multi-physics code

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

Procassini, R.J.

1997-12-31

The fine-scale, multi-space resolution that is envisioned for accurate simulations of complex weapons systems in three spatial dimensions implies flop-rate and memory-storage requirements that will only be obtained in the near future through the use of parallel computational techniques. Since the Monte Carlo transport models in these simulations usually stress both of these computational resources, they are prime candidates for parallelization. The MONACO Monte Carlo transport package, which is currently under development at LLNL, will utilize two types of parallelism within the context of a multi-physics design code: decomposition of the spatial domain across processors (spatial parallelism) and distribution ofmore » particles in a given spatial subdomain across additional processors (particle parallelism). This implementation of the package will utilize explicit data communication between domains (message passing). Such a parallel implementation of a Monte Carlo transport model will result in non-deterministic communication patterns. The communication of particles between subdomains during a Monte Carlo time step may require a significant level of effort to achieve a high parallel efficiency.« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

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

O'Rourke, Patrick Francis

The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.

Inclusion of Scatter in HADES: Final Report

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

Aufderheide, M B

Covert nuclear attack is one of the foremost threats facing the United States and is a primary focus of the War on Terror. The Domestic Nuclear Detection Office (DNDO), within the Department of Homeland Security (DHS), is chartered to develop, and improve domestic systems to detect and interdict smuggling for the illicit use of a nuclear explosive device, fissile material or radiologica1 material. The CAARS (Cargo Advanced Automated Radiography System) program is a major part of the DHS effort to enhance US security by harnessing cutting-edge technologies to detect radiological and nuclear threats at points of entry to the Unitedmore » States. DNDO has selected vendors to develop complete radiographic systems. It is crucial that the initial design and testing concepts for the systems be validated and compared prior to the substantial efforts to build and deploy prototypes and subsequent large-scale production. An important aspect of these systems is the scatter which interferes with imaging. Monte Carlo codes, such as MCNP (X-5 Monte Carlo Team, 2005 Revision) allow scatter to be calculatied, but these calculations are very time consuming. It would be useful to have a fast scatter estimation algorithm in a fast ray tracing code. We have been extending the HADES ray-tracing radiographic simulation code to model vendor systems in a flexible and quick fashion and to use this tool to study a variety of questions involving system performance and the comparative value of surrogates. To enable this work, HADES has been linked to the BRL-CAD library (BRL-CAD Open Source Project, 2010), in order to enable the inclusion of complex CAD geometries in simulations, scanner geometries have been implemented in HADES, and the novel detector responses have been included in HADES. A major extension of HADES which has been required by this effort is the inclusion of scatter in these radiographic simulations. Ray tracing codes generally do not easily allow the inclusion of scatter, because these codes define a source and a grid of detector pixels and only compute the attenuation along rays between these points. Scatter is an extremely complex set of processes which can involve rays which change directions many times between the source and detector. Scatter from outside the field of view of the imaging system, as well as within the field of view, can have an important role in image formation. In this report, we will describe how we implemented a treatment of scatter in HADES. We begin with a discussion of how we define scatter in Section 2, followed by a description of how single Compton scatter is now included in HADES in Section 3. In Section 4 we report a set of verification tests against MCNP and tests of how the technique scales with image size, number of scatters allowed and number of processors used in the calculations. In Section 5, we describe how we plan to extend this approach to other forms of scatter and conclude in Section 6. It should be emphasized that the purpose of this report is to show that a form of scatter has been implemented in HADES and has been verified against MCNP. Validation, the process of comparing simulation and experiment, is a future task.« less

Rates for neutron-capture reactions on tungsten isotopes in iron meteorites. [Abstract only

NASA Technical Reports Server (NTRS)

Masarik, J.; Reedy, R. C.

1994-01-01

High-precision W isotopic analyses by Harper and Jacobsen indicate the W-182/W-183 ratio in the Toluca iron meteorite is shifted by -(3.0 +/- 0.9) x 10(exp -4) relative to a terrestrial standard. Possible causes of this shift are neutron-capture reactions on W during Toluca's approximately 600-Ma exposure to cosmic ray particles or radiogenic growth of W-182 from 9-Ma Hf-182 in the silicate portion of the Earth after removal of W to the Earth's core. Calculations for the rates of neutron-capture reactions on W isotopes were done to study the first possibility. The LAHET Code System (LCS) which consists of the Los Alamos High Energy Transport (LAHET) code and the Monte Carlo N-Particle(MCNP) transport code was used to numerically simulate the irradiation of the Toluca iron meteorite by galactic-cosmic-ray (GCR) particles and to calculate the rates of W(n, gamma) reactions. Toluca was modeled as a 3.9-m-radius sphere with the composition of a typical IA iron meteorite. The incident GCR protons and their interactions were modeled with LAHET, which also handled the interactions of neutrons with energies above 20 MeV. The rates for the capture of neutrons by W-182, W-183, and W-186 were calculated using the detailed library of (n, gamma) cross sections in MCNP. For this study of the possible effect of W(n, gamma) reactions on W isotope systematics, we consider the peak rates. The calculated maximum change in the normalized W-182/W-183 ratio due to neutron-capture reactions cannot account for more than 25% of the mass 182 deficit observed in Toluca W.

SUPREM-DSMC: A New Scalable, Parallel, Reacting, Multidimensional Direct Simulation Monte Carlo Flow Code

NASA Technical Reports Server (NTRS)

Campbell, David; Wysong, Ingrid; Kaplan, Carolyn; Mott, David; Wadsworth, Dean; VanGilder, Douglas

2000-01-01

An AFRL/NRL team has recently been selected to develop a scalable, parallel, reacting, multidimensional (SUPREM) Direct Simulation Monte Carlo (DSMC) code for the DoD user community under the High Performance Computing Modernization Office (HPCMO) Common High Performance Computing Software Support Initiative (CHSSI). This paper will introduce the JANNAF Exhaust Plume community to this three-year development effort and present the overall goals, schedule, and current status of this new code.

Determination of efficiency of an aged HPGe detector for gaseous sources by self absorption correction and point source methods

NASA Astrophysics Data System (ADS)

Sarangapani, R.; Jose, M. T.; Srinivasan, T. K.; Venkatraman, B.

2017-07-01

Methods for the determination of efficiency of an aged high purity germanium (HPGe) detector for gaseous sources have been presented in the paper. X-ray radiography of the detector has been performed to get detector dimensions for computational purposes. The dead layer thickness of HPGe detector has been ascertained from experiments and Monte Carlo computations. Experimental work with standard point and liquid sources in several cylindrical geometries has been undertaken for obtaining energy dependant efficiency. Monte Carlo simulations have been performed for computing efficiencies for point, liquid and gaseous sources. Self absorption correction factors have been obtained using mathematical equations for volume sources and MCNP simulations. Self-absorption correction and point source methods have been used to estimate the efficiency for gaseous sources. The efficiencies determined from the present work have been used to estimate activity of cover gas sample of a fast reactor.

Diagnosing Undersampling Biases in Monte Carlo Eigenvalue and Flux Tally Estimates

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

Perfetti, Christopher M.; Rearden, Bradley T.; Marshall, William J.

2017-02-08

Here, this study focuses on understanding the phenomena in Monte Carlo simulations known as undersampling, in which Monte Carlo tally estimates may not encounter a sufficient number of particles during each generation to obtain unbiased tally estimates. Steady-state Monte Carlo simulations were performed using the KENO Monte Carlo tools within the SCALE code system for models of several burnup credit applications with varying degrees of spatial and isotopic complexities, and the incidence and impact of undersampling on eigenvalue and flux estimates were examined. Using an inadequate number of particle histories in each generation was found to produce a maximum bias of ~100 pcm in eigenvalue estimates and biases that exceeded 10% in fuel pin flux tally estimates. Having quantified the potential magnitude of undersampling biases in eigenvalue and flux tally estimates in these systems, this study then investigated whether Markov Chain Monte Carlo convergence metrics could be integrated into Monte Carlo simulations to predict the onset and magnitude of undersampling biases. Five potential metrics for identifying undersampling biases were implemented in the SCALE code system and evaluated for their ability to predict undersampling biases by comparing the test metric scores with the observed undersampling biases. Finally, of the five convergence metrics that were investigated, three (the Heidelberger-Welch relative half-width, the Gelman-Rubin more » $$\\hat{R}_c$$ diagnostic, and tally entropy) showed the potential to accurately predict the behavior of undersampling biases in the responses examined.« less

Monte Carlo Investigation on the Effect of Heterogeneities on Strut Adjusted Volume Implant (SAVI) Dosimetry

NASA Astrophysics Data System (ADS)

Koontz, Craig

Breast cancer is the most prevalent cancer for women with more than 225,000 new cases diagnosed in the United States in 2012 (ACS, 2012). With the high prevalence, comes an increased emphasis on researching new techniques to treat this disease. Accelerated partial breast irradiation (APBI) has been used as an alternative to whole breast irradiation (WBI) in order to treat occult disease after lumpectomy. Similar recurrence rates have been found using ABPI after lumpectomy as with mastectomy alone, but with the added benefit of improved cosmetic and psychological results. Intracavitary brachytherapy devices have been used to deliver the APBI prescription. However, inability to produce asymmetric dose distributions in order to avoid overdosing skin and chest wall has been an issue with these devices. Multi-lumen devices were introduced to overcome this problem. Of these, the Strut-Adjusted Volume Implant (SAVI) has demonstrated the greatest ability to produce an asymmetric dose distribution, which would have greater ability to avoid skin and chest wall dose, and thus allow more women to receive this type of treatment. However, SAVI treatments come with inherent heterogeneities including variable backscatter due to the proximity to the tissue-air and tissue-lung interfaces and variable contents within the cavity created by the SAVI. The dose calculation protocol based on TG-43 does not account for heterogeneities and thus will not produce accurate dosimetry; however Acuros, a model-based dose calculation algorithm manufactured by Varian Medical Systems, claims to accurately account for heterogeneities. Monte Carlo simulation can calculate the dosimetry with high accuracy. In this thesis, a model of the SAVI will be created for Monte Carlo, specifically using MCNP code, in order to explore the affects of heterogeneities on the dose distribution. This data will be compared to TG-43 and Acuros calculated dosimetry to explore their accuracy.

Gold nanoparticle‐based brachytherapy enhancement in choroidal melanoma using a full Monte Carlo model of the human eye

PubMed Central

Vaez‐zadeh, Mehdi; Masoudi, S. Farhad; Rahmani, Faezeh; Knaup, Courtney; Meigooni, Ali S.

2015-01-01

The effects of gold nanoparticles (GNPs) in 125I brachytherapy dose enhancement on choroidal melanoma are examined using the Monte Carlo simulation technique. Usually, Monte Carlo ophthalmic brachytherapy dosimetry is performed in a water phantom. However, here, the compositions of human eye have been considered instead of water. Both human eye and water phantoms have been simulated with MCNP5 code. These simulations were performed for a fully loaded 16 mm COMS eye plaque containing 13 125I seeds. The dose delivered to the tumor and normal tissues have been calculated in both phantoms with and without GNPs. Normally, the radiation therapy of cancer patients is designed to deliver a required dose to the tumor while sparing the surrounding normal tissues. However, as the normal and cancerous cells absorbed dose in an almost identical fashion, the normal tissue absorbed radiation dose during the treatment time. The use of GNPs in combination with radiotherapy in the treatment of tumor decreases the absorbed dose by normal tissues. The results indicate that the dose to the tumor in an eyeball implanted with COMS plaque increases with increasing GNPs concentration inside the target. Therefore, the required irradiation time for the tumors in the eye is decreased by adding the GNPs prior to treatment. As a result, the dose to normal tissues decreases when the irradiation time is reduced. Furthermore, a comparison between the simulated data in an eye phantom made of water and eye phantom made of human eye composition, in the presence of GNPs, shows the significance of utilizing the composition of eye in ophthalmic brachytherapy dosimetry Also, defining the eye composition instead of water leads to more accurate calculations of GNPs radiation effects in ophthalmic brachytherapy dosimetry. PACS number: 87.53.Jw, 87.85.Rs, 87.10.Rt PMID:26699318

An empirical approach to estimate near-infra-red photon propagation and optically induced drug release in brain tissues

NASA Astrophysics Data System (ADS)

Prabhu Verleker, Akshay; Fang, Qianqian; Choi, Mi-Ran; Clare, Susan; Stantz, Keith M.

2015-03-01

The purpose of this study is to develop an alternate empirical approach to estimate near-infra-red (NIR) photon propagation and quantify optically induced drug release in brain metastasis, without relying on computationally expensive Monte Carlo techniques (gold standard). Targeted drug delivery with optically induced drug release is a noninvasive means to treat cancers and metastasis. This study is part of a larger project to treat brain metastasis by delivering lapatinib-drug-nanocomplexes and activating NIR-induced drug release. The empirical model was developed using a weighted approach to estimate photon scattering in tissues and calibrated using a GPU based 3D Monte Carlo. The empirical model was developed and tested against Monte Carlo in optical brain phantoms for pencil beams (width 1mm) and broad beams (width 10mm). The empirical algorithm was tested against the Monte Carlo for different albedos along with diffusion equation and in simulated brain phantoms resembling white-matter (μs'=8.25mm-1, μa=0.005mm-1) and gray-matter (μs'=2.45mm-1, μa=0.035mm-1) at wavelength 800nm. The goodness of fit between the two models was determined using coefficient of determination (R-squared analysis). Preliminary results show the Empirical algorithm matches Monte Carlo simulated fluence over a wide range of albedo (0.7 to 0.99), while the diffusion equation fails for lower albedo. The photon fluence generated by empirical code matched the Monte Carlo in homogeneous phantoms (R2=0.99). While GPU based Monte Carlo achieved 300X acceleration compared to earlier CPU based models, the empirical code is 700X faster than the Monte Carlo for a typical super-Gaussian laser beam.

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.

Specific absorbed fractions of electrons and photons for Rad-HUMAN phantom using Monte Carlo method

NASA Astrophysics Data System (ADS)

Wang, Wen; Cheng, Meng-Yun; Long, Peng-Cheng; Hu, Li-Qin

2015-07-01

The specific absorbed fractions (SAF) for self- and cross-irradiation are effective tools for the internal dose estimation of inhalation and ingestion intakes of radionuclides. A set of SAFs of photons and electrons were calculated using the Rad-HUMAN phantom, which is a computational voxel phantom of a Chinese adult female that was created using the color photographic image of the Chinese Visible Human (CVH) data set by the FDS Team. The model can represent most Chinese adult female anatomical characteristics and can be taken as an individual phantom to investigate the difference of internal dose with Caucasians. In this study, the emission of mono-energetic photons and electrons of 10 keV to 4 MeV energy were calculated using the Monte Carlo particle transport calculation code MCNP. Results were compared with the values from ICRP reference and ORNL models. The results showed that SAF from the Rad-HUMAN have similar trends but are larger than those from the other two models. The differences were due to the racial and anatomical differences in organ mass and inter-organ distance. The SAFs based on the Rad-HUMAN phantom provide an accurate and reliable data for internal radiation dose calculations for Chinese females. Supported by Strategic Priority Research Program of Chinese Academy of Sciences (XDA03040000), National Natural Science Foundation of China (910266004, 11305205, 11305203) and National Special Program for ITER (2014GB112001)

Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

NASA Astrophysics Data System (ADS)

Ródenas, José

2017-11-01

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

Calculation of dose distribution in compressible breast tissues using finite element modeling, Monte Carlo simulation and thermoluminescence dosimeters

NASA Astrophysics Data System (ADS)

Mohammadyari, Parvin; Faghihi, Reza; Mosleh-Shirazi, Mohammad Amin; Lotfi, Mehrzad; Rahim Hematiyan, Mohammad; Koontz, Craig; Meigooni, Ali S.

2015-12-01

Compression is a technique to immobilize the target or improve the dose distribution within the treatment volume during different irradiation techniques such as AccuBoost® brachytherapy. However, there is no systematic method for determination of dose distribution for uncompressed tissue after irradiation under compression. In this study, the mechanical behavior of breast tissue between compressed and uncompressed states was investigated. With that, a novel method was developed to determine the dose distribution in uncompressed tissue after irradiation of compressed breast tissue. Dosimetry was performed using two different methods, namely, Monte Carlo simulations using the MCNP5 code and measurements using thermoluminescent dosimeters (TLD). The displacement of the breast elements was simulated using a finite element model and calculated using ABAQUS software. From these results, the 3D dose distribution in uncompressed tissue was determined. The geometry of the model was constructed from magnetic resonance images of six different women volunteers. The mechanical properties were modeled by using the Mooney-Rivlin hyperelastic material model. Experimental dosimetry was performed by placing the TLD chips into the polyvinyl alcohol breast equivalent phantom. The results determined that the nodal displacements, due to the gravitational force and the 60 Newton compression forces (with 43% contraction in the loading direction and 37% expansion in the orthogonal direction) were determined. Finally, a comparison of the experimental data and the simulated data showed agreement within 11.5%  ±  5.9%.

Calculation of dose distribution in compressible breast tissues using finite element modeling, Monte Carlo simulation and thermoluminescence dosimeters.

PubMed

Mohammadyari, Parvin; Faghihi, Reza; Mosleh-Shirazi, Mohammad Amin; Lotfi, Mehrzad; Hematiyan, Mohammad Rahim; Koontz, Craig; Meigooni, Ali S

2015-12-07

Compression is a technique to immobilize the target or improve the dose distribution within the treatment volume during different irradiation techniques such as AccuBoost(®) brachytherapy. However, there is no systematic method for determination of dose distribution for uncompressed tissue after irradiation under compression. In this study, the mechanical behavior of breast tissue between compressed and uncompressed states was investigated. With that, a novel method was developed to determine the dose distribution in uncompressed tissue after irradiation of compressed breast tissue. Dosimetry was performed using two different methods, namely, Monte Carlo simulations using the MCNP5 code and measurements using thermoluminescent dosimeters (TLD). The displacement of the breast elements was simulated using a finite element model and calculated using ABAQUS software. From these results, the 3D dose distribution in uncompressed tissue was determined. The geometry of the model was constructed from magnetic resonance images of six different women volunteers. The mechanical properties were modeled by using the Mooney-Rivlin hyperelastic material model. Experimental dosimetry was performed by placing the TLD chips into the polyvinyl alcohol breast equivalent phantom. The results determined that the nodal displacements, due to the gravitational force and the 60 Newton compression forces (with 43% contraction in the loading direction and 37% expansion in the orthogonal direction) were determined. Finally, a comparison of the experimental data and the simulated data showed agreement within 11.5%  ±  5.9%.

The Development of WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs

NASA Astrophysics Data System (ADS)

Bergmann, Ryan

Graphics processing units, or GPUs, have gradually increased in computational power from the small, job-specific boards of the early 1990s to the programmable powerhouses of today. Compared to more common central processing units, or CPUs, GPUs have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU-optimized parallel algorithms are not directly portable to GPU architectures (or at least not without losing substantial performance), transport codes need to be rewritten to execute efficiently on GPUs. Unless this is done, reactor simulations cannot take full advantage of these new supercomputers. WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed in this work as to efficiently implement a continuous energy Monte Carlo neutron transport algorithm on a GPU. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo Method, namely, very few physical and geometrical simplifications. WARP is able to calculate multiplication factors, flux tallies, and fission source distributions for time-independent problems, and can run in both criticality or fixed source modes. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. WARP uses an event-based algorithm, but with some important differences. Moving data is expensive, so WARP uses a remapping vector of pointer/index pairs to direct GPU threads to the data they need to access. The remapping vector is sorted by reaction type after every transport iteration using a high-efficiency parallel radix sort, which serves to keep the reaction types as contiguous as possible and removes completed histories from the transport cycle. The sort reduces the amount of divergence in GPU ``thread blocks,'' keeps the SIMD units as full as possible, and eliminates using memory bandwidth to check if a neutron in the batch has been terminated or not. Using a remapping vector means the data access pattern is irregular, but this is mitigated by using large batch sizes where the GPU can effectively eliminate the high cost of irregular global memory access. WARP modifies the standard unionized energy grid implementation to reduce memory traffic. Instead of storing a matrix of pointers indexed by reaction type and energy, WARP stores three matrices. The first contains cross section values, the second contains pointers to angular distributions, and a third contains pointers to energy distributions. This linked list type of layout increases memory usage, but lowers the number of data loads that are needed to determine a reaction by eliminating a pointer load to find a cross section value. Optimized, high-performance GPU code libraries are also used by WARP wherever possible. The CUDA performance primitives (CUDPP) library is used to perform the parallel reductions, sorts and sums, the CURAND library is used to seed the linear congruential random number generators, and the OptiX ray tracing framework is used for geometry representation. OptiX is a highly-optimized library developed by NVIDIA that automatically builds hierarchical acceleration structures around user-input geometry so only surfaces along a ray line need to be queried in ray tracing. WARP also performs material and cell number queries with OptiX by using a point-in-polygon like algorithm. WARP has shown that GPUs are an effective platform for performing Monte Carlo neutron transport with continuous energy cross sections. Currently, WARP is the most detailed and feature-rich program in existence for performing continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs, but compared to production codes like Serpent and MCNP, WARP has limited capabilities. Despite WARP's lack of features, its novel algorithm implementations show that high performance can be achieved on a GPU despite the inherently divergent program flow and sparse data access patterns. WARP is not ready for everyday nuclear reactor calculations, but is a good platform for further development of GPU-accelerated Monte Carlo neutron transport. In it's current state, it may be a useful tool for multiplication factor searches, i.e. determining reactivity coefficients by perturbing material densities or temperatures, since these types of calculations typically do not require many flux tallies. (Abstract shortened by UMI.)

A collision history-based approach to Sensitivity/Perturbation calculations in the continuous energy Monte Carlo code SERPENT

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

Giuseppe Palmiotti

In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the eects of nuclear data perturbation on several response functions: the eective multiplication factor, reaction rate ratios and bilinear ratios (e.g., eective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.

Monte Carlo Particle Lists: MCPL

NASA Astrophysics Data System (ADS)

Kittelmann, T.; Klinkby, E.; Knudsen, E. B.; Willendrup, P.; Cai, X. X.; Kanaki, K.

2017-09-01

A binary format with lists of particle state information, for interchanging particles between various Monte Carlo simulation applications, is presented. Portable C code for file manipulation is made available to the scientific community, along with converters and plugins for several popular simulation packages.

Four pi calibration and modeling of a bare germanium detector in a cylindrical field source

NASA Astrophysics Data System (ADS)

Dewberry, R. A.; Young, J. E.

2012-05-01

In this paper we describe a 4π cylindrical field acquisition configuration surrounding a bare (unshielded, uncollimated) high purity germanium detector. We perform an efficiency calibration with a flexible planar source and model the configuration in the 4π cylindrical field. We then use exact calculus to model the flux on the cylindrical sides and end faces of the detector. We demonstrate that the model accurately represents the experimental detection efficiency compared to that of a point source and to Monte Carlo N-particle (MCNP) calculations of the flux. The model sums over the entire source surface area and the entire detector surface area including both faces and the detector's cylindrical sides. Agreement between the model and both experiment and the MCNP calculation is within 8%.

Mathematical simulations of photon interactions using Monte Carlo analysis to evaluate the uncertainty associated with in vivo K X-ray fluorescence measurements of stable lead in bone

NASA Astrophysics Data System (ADS)

Lodwick, Camille J.

This research utilized Monte Carlo N-Particle version 4C (MCNP4C) to simulate K X-ray fluorescent (K XRF) measurements of stable lead in bone. Simulations were performed to investigate the effects that overlying tissue thickness, bone-calcium content, and shape of the calibration standard have on detector response in XRF measurements at the human tibia. Additional simulations of a knee phantom considered uncertainty associated with rotation about the patella during XRF measurements. Simulations tallied the distribution of energy deposited in a high-purity germanium detector originating from collimated 88 keV 109Cd photons in backscatter geometry. Benchmark measurements were performed on simple and anthropometric XRF calibration phantoms of the human leg and knee developed at the University of Cincinnati with materials proven to exhibit radiological characteristics equivalent to human tissue and bone. Initial benchmark comparisons revealed that MCNP4C limits coherent scatter of photons to six inverse angstroms of momentum transfer and a Modified MCNP4C was developed to circumvent the limitation. Subsequent benchmark measurements demonstrated that Modified MCNP4C adequately models photon interactions associated with in vivo K XRF of lead in bone. Further simulations of a simple leg geometry possessing tissue thicknesses from 0 to 10 mm revealed increasing overlying tissue thickness from 5 to 10 mm reduced predicted lead concentrations an average 1.15% per 1 mm increase in tissue thickness (p < 0.0001). An anthropometric leg phantom was mathematically defined in MCNP to more accurately reflect the human form. A simulated one percent increase in calcium content (by mass) of the anthropometric leg phantom's cortical bone demonstrated to significantly reduce the K XRF normalized ratio by 4.5% (p < 0.0001). Comparison of the simple and anthropometric calibration phantoms also suggested that cylindrical calibration standards can underestimate lead content of a human leg up to 4%. The patellar bone structure in which the fluorescent photons originate was found to vary dramatically with measurement angle. The relative contribution of lead signal from the patella declined from 65% to 27% when rotated 30°. However, rotation of the source-detector about the patella from 0 to 45° demonstrated no significant effect on the net K XRF response at the knee.

Positron follow-up in liquid water: I. A new Monte Carlo track-structure code.

PubMed

Champion, C; Le Loirec, C

2006-04-07

When biological matter is irradiated by charged particles, a wide variety of interactions occur, which lead to a deep modification of the cellular environment. To understand the fine structure of the microscopic distribution of energy deposits, Monte Carlo event-by-event simulations are particularly suitable. However, the development of these track-structure codes needs accurate interaction cross sections for all the electronic processes: ionization, excitation, positronium formation and even elastic scattering. Under these conditions, we have recently developed a Monte Carlo code for positrons in water, the latter being commonly used to simulate the biological medium. All the processes are studied in detail via theoretical differential and total cross-section calculations performed by using partial wave methods. Comparisons with existing theoretical and experimental data in terms of stopping powers, mean energy transfers and ranges show very good agreements. Moreover, thanks to the theoretical description of positronium formation, we have access, for the first time, to the complete kinematics of the electron capture process. Then, the present Monte Carlo code is able to describe the detailed positronium history, which will provide useful information for medical imaging (like positron emission tomography) where improvements are needed to define with the best accuracy the tumoural volumes.

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.

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.

Combined experimental and Monte Carlo verification of brachytherapy plans for vaginal applicators

NASA Astrophysics Data System (ADS)

Sloboda, Ron S.; Wang, Ruqing

1998-12-01

Dose rates in a phantom around a shielded and an unshielded vaginal applicator containing Selectron low-dose-rate sources were determined by experiment and Monte Carlo simulation. Measurements were performed with thermoluminescent dosimeters in a white polystyrene phantom using an experimental protocol geared for precision. Calculations for the same set-up were done using a version of the EGS4 Monte Carlo code system modified for brachytherapy applications into which a new combinatorial geometry package developed by Bielajew was recently incorporated. Measured dose rates agree with Monte Carlo estimates to within 5% (1 SD) for the unshielded applicator, while highlighting some experimental uncertainties for the shielded applicator. Monte Carlo calculations were also done to determine a value for the effective transmission of the shield required for clinical treatment planning, and to estimate the dose rate in water at points in axial and sagittal planes transecting the shielded applicator. Comparison with dose rates generated by the planning system indicates that agreement is better than 5% (1 SD) at most positions. The precision thermoluminescent dosimetry protocol and modified Monte Carlo code are effective complementary tools for brachytherapy applicator dosimetry.

On the utility of graphics cards to perform massively parallel simulation of advanced Monte Carlo methods

PubMed Central

Lee, Anthony; Yau, Christopher; Giles, Michael B.; Doucet, Arnaud; Holmes, Christopher C.

2011-01-01

We present a case-study on the utility of graphics cards to perform massively parallel simulation of advanced Monte Carlo methods. Graphics cards, containing multiple Graphics Processing Units (GPUs), are self-contained parallel computational devices that can be housed in conventional desktop and laptop computers and can be thought of as prototypes of the next generation of many-core processors. For certain classes of population-based Monte Carlo algorithms they offer massively parallel simulation, with the added advantage over conventional distributed multi-core processors that they are cheap, easily accessible, easy to maintain, easy to code, dedicated local devices with low power consumption. On a canonical set of stochastic simulation examples including population-based Markov chain Monte Carlo methods and Sequential Monte Carlo methods, we nd speedups from 35 to 500 fold over conventional single-threaded computer code. Our findings suggest that GPUs have the potential to facilitate the growth of statistical modelling into complex data rich domains through the availability of cheap and accessible many-core computation. We believe the speedup we observe should motivate wider use of parallelizable simulation methods and greater methodological attention to their design. PMID:22003276

Energetic properties' investigation of removing flattening filter at phantom surface: Monte Carlo study using BEAMnrc code, DOSXYZnrc code and BEAMDP code

NASA Astrophysics Data System (ADS)

Bencheikh, Mohamed; Maghnouj, Abdelmajid; Tajmouati, Jaouad

2017-11-01

The Monte Carlo calculation method is considered to be the most accurate method for dose calculation in radiotherapy and beam characterization investigation, in this study, the Varian Clinac 2100 medical linear accelerator with and without flattening filter (FF) was modelled. The objective of this study was to determine flattening filter impact on particles' energy properties at phantom surface in terms of energy fluence, mean energy, and energy fluence distribution. The Monte Carlo codes used in this study were BEAMnrc code for simulating linac head, DOSXYZnrc code for simulating the absorbed dose in a water phantom, and BEAMDP for extracting energy properties. Field size was 10 × 10 cm2, simulated photon beam energy was 6 MV and SSD was 100 cm. The Monte Carlo geometry was validated by a gamma index acceptance rate of 99% in PDD and 98% in dose profiles, gamma criteria was 3% for dose difference and 3mm for distance to agreement. In without-FF, the energetic properties was as following: electron contribution was increased by more than 300% in energy fluence, almost 14% in mean energy and 1900% in energy fluence distribution, however, photon contribution was increased 50% in energy fluence, and almost 18% in mean energy and almost 35% in energy fluence distribution. The removing flattening filter promotes the increasing of electron contamination energy versus photon energy; our study can contribute in the evolution of removing flattening filter configuration in future linac.

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.

CAD-Based Shielding Analysis for ITER Port Diagnostics

NASA Astrophysics Data System (ADS)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

MO-FG-CAMPUS-TeP3-02: Benchmarks of a Proton Relative Biological Effectiveness (RBE) Model for DNA Double Strand Break (DSB) Induction in the FLUKA, MCNP, TOPAS, and RayStation™ Treatment Planning System

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

Stewart, R; Streitmatter, S; Traneus, E

2016-06-15

Purpose: Validate implementation of a published RBE model for DSB induction (RBEDSB) in several general purpose Monte Carlo (MC) code systems and the RayStation™ treatment planning system (TPS). For protons and other light ions, DSB induction is a critical initiating molecular event that correlates well with the RBE for cell survival. Methods: An efficient algorithm to incorporate information on proton and light ion RBEDSB from the independently tested Monte Carlo Damage Simulation (MCDS) has now been integrated into MCNP (Stewart et al. PMB 60, 8249–8274, 2015), FLUKA, TOPAS and a research build of the RayStation™ TPS. To cross-validate the RBEDSBmore » model implementation LET distributions, depth-dose and lateral (dose and RBEDSB) profiles for monodirectional monoenergetic (100 to 200 MeV) protons incident on a water phantom are compared. The effects of recoil and secondary ion production ({sub 2}H{sub +}, {sub 3}H{sub +}, {sub 3}He{sub 2+}, {sub 4}He{sub 2+}), spot size (3 and 10 mm), and transport physics on beam profiles and RBEDSB are examined. Results: Depth-dose and RBEDSB profiles among all of the MC models are in excellent agreement using a 1 mm distance criterion (width of a voxel). For a 100 MeV proton beam (10 mm spot), RBEDSB = 1.2 ± 0.03 (− 2–3%) at the tip of the Bragg peak and increases to 1.59 ± 0.3 two mm distal to the Bragg peak. RBEDSB tends to decrease as the kinetic energy of the incident proton increases. Conclusion: The model for proton RBEDSB has been accurately implemented into FLUKA, MCNP, TOPAS and the RayStation™TPS. The transport of secondary light ions (Z > 1) has a significant impact on RBEDSB, especially distal to the Bragg peak, although light ions have a small effect on (dosexRBEDSB) profiles. The ability to incorporate spatial variations in proton RBE within a TPS creates new opportunities to individualize treatment plans and increase the therapeutic ratio. Dr. Erik Traneus is employed full-time as a Research Scientist at RaySearch Laboratories. The research build of the RayStation used in the study was made available to the University of Washington free of charge. RaySearch Laboratories did not provide any monetary support for the reported studies.« less

Active interrogation of highly enriched uranium

NASA Astrophysics Data System (ADS)

Fairrow, Nannette Lea

Safeguarding special nuclear material (SNM) in the Department of Energy Complex is vital to the national security of the United States. Active and passive nondestructive assays are used to confirm the presence of SNM in various configurations ranging from waste to nuclear weapons. Confirmation measurements for nuclear weapons are more challenging because the design complicates the detection of a distinct signal for highly enriched uranium. The emphasis of this dissertation was to investigate a new nondestructive assay technique that provides an independent and distinct signal to confirm the presence of highly enriched uranium (HEU). Once completed and tested this assay method could be applied to confirmation measurements of nuclear weapons. The new system uses a 14-MeV neutron source for interrogation and records the arrival time of neutrons between the pulses with a high efficiency detection system. The data is then analyzed by the Feynman reduced variance method. The analysis determined the amount of correlation in the data and provided a unique signature of correlated fission neutrons. Measurements of HEU spheres were conducted at Los Alamos with the new system. Then, Monte Carlo calculations were performed to verify hypothesis made about the behavior of the neutrons in the experiment. Comparisons of calculated counting rates by the Monte Carlo N-Particle Transport Code (MCNP) were made with the experimental data to confirm that the measured response reflected the desired behavior of neutron interactions in the highly enriched uranium. In addition, MCNP calculations of the delayed neutron build-up were compared with the measured data. Based on the results obtained from this dissertation, this measurement method has the potential to be expanded to include mass determinations of highly enriched uranium. Although many safeguards techniques exist for measuring special nuclear material, the number of assays that can be used to confirm HEU in shielded systems is limited. These assays also rely on secondary characteristics of the material to be measured. A review of the nondestructive techniques with potential applications for nuclear weapons confirmatory measurements were evaluated with summaries of the pros and cons involved in implementing the methods at production type facilities.

Applying Quantum Monte Carlo to the Electronic Structure Problem

NASA Astrophysics Data System (ADS)

Powell, Andrew D.; Dawes, Richard

2016-06-01

Two distinct types of Quantum Monte Carlo (QMC) calculations are applied to electronic structure problems such as calculating potential energy curves and producing benchmark values for reaction barriers. First, Variational and Diffusion Monte Carlo (VMC and DMC) methods using a trial wavefunction subject to the fixed node approximation were tested using the CASINO code.[1] Next, Full Configuration Interaction Quantum Monte Carlo (FCIQMC), along with its initiator extension (i-FCIQMC) were tested using the NECI code.[2] FCIQMC seeks the FCI energy for a specific basis set. At a reduced cost, the efficient i-FCIQMC method can be applied to systems in which the standard FCIQMC approach proves to be too costly. Since all of these methods are statistical approaches, uncertainties (error-bars) are introduced for each calculated energy. This study tests the performance of the methods relative to traditional quantum chemistry for some benchmark systems. References: [1] R. J. Needs et al., J. Phys.: Condensed Matter 22, 023201 (2010). [2] G. H. Booth et al., J. Chem. Phys. 131, 054106 (2009).

PEPSI — a Monte Carlo generator for polarized leptoproduction

NASA Astrophysics Data System (ADS)

Mankiewicz, L.; Schäfer, A.; Veltri, M.

1992-09-01

We describe PEPSI (Polarized Electron Proton Scattering Interactions), a Monte Carlo program for polarized deep inelastic leptoproduction mediated by electromagnetic interaction, and explain how to use it. The code is a modification of the LEPTO 4.3 Lund Monte Carlo for unpolarized scattering. The hard virtual gamma-parton scattering is generated according to the polarization-dependent QCD cross-section of the first order in α S. PEPSI requires the standard polarization-independent JETSET routines to simulate the fragmentation into final hadrons.

The Monte Carlo code MCPTV--Monte Carlo dose calculation in radiation therapy with carbon ions.

PubMed

Karg, Juergen; Speer, Stefan; Schmidt, Manfred; Mueller, Reinhold

2010-07-07

The Monte Carlo code MCPTV is presented. MCPTV is designed for dose calculation in treatment planning in radiation therapy with particles and especially carbon ions. MCPTV has a voxel-based concept and can perform a fast calculation of the dose distribution on patient CT data. Material and density information from CT are taken into account. Electromagnetic and nuclear interactions are implemented. Furthermore the algorithm gives information about the particle spectra and the energy deposition in each voxel. This can be used to calculate the relative biological effectiveness (RBE) for each voxel. Depth dose distributions are compared to experimental data giving good agreement. A clinical example is shown to demonstrate the capabilities of the MCPTV dose calculation.

Massively parallelized Monte Carlo software to calculate the light propagation in arbitrarily shaped 3D turbid media

NASA Astrophysics Data System (ADS)

Zoller, Christian; Hohmann, Ansgar; Ertl, Thomas; Kienle, Alwin

2017-07-01

The Monte Carlo method is often referred as the gold standard to calculate the light propagation in turbid media [1]. Especially for complex shaped geometries where no analytical solutions are available the Monte Carlo method becomes very important [1, 2]. In this work a Monte Carlo software is presented, to simulate the light propagation in complex shaped geometries. To improve the simulation time the code is based on OpenCL such that graphics cards can be used as well as other computing devices. Within the software an illumination concept is presented to realize easily all kinds of light sources, like spatial frequency domain (SFD), optical fibers or Gaussian beam profiles. Moreover different objects, which are not connected to each other, can be considered simultaneously, without any additional preprocessing. This Monte Carlo software can be used for many applications. In this work the transmission spectrum of a tooth and the color reconstruction of a virtual object are shown, using results from the Monte Carlo software.

Acceleration of Monte Carlo simulation of photon migration in complex heterogeneous media using Intel many-integrated core architecture.

PubMed

Gorshkov, Anton V; Kirillin, Mikhail Yu

2015-08-01

Over two decades, the Monte Carlo technique has become a gold standard in simulation of light propagation in turbid media, including biotissues. Technological solutions provide further advances of this technique. The Intel Xeon Phi coprocessor is a new type of accelerator for highly parallel general purpose computing, which allows execution of a wide range of applications without substantial code modification. We present a technical approach of porting our previously developed Monte Carlo (MC) code for simulation of light transport in tissues to the Intel Xeon Phi coprocessor. We show that employing the accelerator allows reducing computational time of MC simulation and obtaining simulation speed-up comparable to GPU. We demonstrate the performance of the developed code for simulation of light transport in the human head and determination of the measurement volume in near-infrared spectroscopy brain sensing.

Optimization of the Efficiency of a Neutron Detector to Measure (α, n) Reaction Cross-Section

NASA Astrophysics Data System (ADS)

Perello, Jesus; Montes, Fernando; Ahn, Tony; Meisel, Zach; Joint InstituteNuclear Astrophysics Team

2015-04-01

Nucleosynthesis, the origin of elements, is one of the greatest mysteries in physics. A recent particular nucleosynthesis process of interest is the charge-particle process (cpp). In the cpp, elements form by nuclear fusion reactions during supernovae. This process of nuclear fusion, (α,n), will be studied by colliding beam elements produced and accelerated at the National Superconducting Cyclotron Laboratory (NSCL) to a helium-filled cell target. The elements will fuse with α (helium nuclei) and emit neutrons during the reaction. The neutrons will be detected for a count of fused-elements, thus providing us the probability of such reactions. The neutrons will be detected using the Neutron Emission Ratio Observer (NERO). Currently, NERO's efficiency varies for neutrons at the expected energy range (0-12 MeV). To study (α,n), NERO's efficiency must be near-constant at these energies. Monte-Carlo N-Particle Transport Code (MCNP6), a software package that simulates nuclear processes, was used to optimize NERO configuration for the experiment. MCNP6 was used to simulate neutron interaction with different NERO configurations at the expected neutron energies. By adding additional 3He detectors and polyethylene, a near-constant efficiency at these energies was obtained in the simulations. With the new NERO configuration, study of the (α,n) reactions can begin, which may explain how elements are formed in the cpp. SROP MSU, NSF, JINA, McNair Society.

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

MacFarlane, R. E.

An accurate representation of the scattering of neutrons by the materials used to build cold sources at neutron scattering facilities is important for the initial design and optimization of a cold source, and for the analysis of experimental results obtained using the cold source. In practice, this requires a good representation of the physics of scattering from the material, a method to convert this into observable quantities (such as scattering cross sections), and a method to use the results in a neutron transport code (such as the MCNP Monte Carlo code). At Los Alamos, the authors have been developing thesemore » capabilities over the last ten years. The final set of cold-moderator evaluations, together with evaluations for conventional moderator materials, was released in 1994. These materials have been processed into MCNP data files using the NJOY Nuclear Data Processing System. Over the course of this work, they were able to develop a new module for NJOY called LEAPR based on the LEAP + ADDELT code from the UK as modified by D.J. Picton for cold-moderator calculations. Much of the physics for methane came from Picton`s work. The liquid hydrogen work was originally based on a code using the Young-Koppel approach that went through a number of hands in Europe (including Rolf Neef and Guy Robert). It was generalized and extended for LEAPR, and depends strongly on work by Keinert and Sax of the University of Stuttgart. Thus, their collection of cold-moderator scattering kernels is truly an international effort, and they are glad to be able to return the enhanced evaluations and processing techniques to the international community. In this paper, they give sections on the major cold moderator materials (namely, solid methane, liquid methane, and liquid hydrogen) using each section to introduce the relevant physics for that material and to show typical results.« less

Skyshine study for next generation of fusion devices

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

Gohar, Y.; Yang, S.

1987-02-01

A shielding analysis for next generation of fusion devices (ETR/INTOR) was performed to study the dose equivalent outside the reactor building during operation including the contribution from neutrons and photons scattered back by collisions with air nuclei (skyshine component). Two different three-dimensional geometrical models for a tokamak fusion reactor based on INTOR design parameters were developed for this study. In the first geometrical model, the reactor geometry and the spatial distribution of the deuterium-tritium neutron source were simplified for a parametric survey. The second geometrical model employed an explicit representation of the toroidal geometry of the reactor chamber and themore » spatial distribution of the neutron source. The MCNP general Monte Carlo code for neutron and photon transport was used to perform all the calculations. The energy distribution of the neutron source was used explicitly in the calculations with ENDF/B-V data. The dose equivalent results were analyzed as a function of the concrete roof thickness of the reactor building and the location outside the reactor building.« less

An accelerator-based Boron Neutron Capture Therapy (BNCT) facility based on the 7Li(p,n)7Be

NASA Astrophysics Data System (ADS)

Musacchio González, Elizabeth; Martín Hernández, Guido

2017-09-01

BNCT (Boron Neutron Capture Therapy) is a therapeutic modality used to irradiate tumors cells previously loaded with the stable isotope 10B, with thermal or epithermal neutrons. This technique is capable of delivering a high dose to the tumor cells while the healthy surrounding tissue receive a much lower dose depending on the 10B biodistribution. In this study, therapeutic gain and tumor dose per target power, as parameters to evaluate the treatment quality, were calculated. The common neutron-producing reaction 7Li(p,n)7Be for accelerator-based BNCT, having a reaction threshold of 1880.4 keV, was considered as the primary source of neutrons. Energies near the reaction threshold for deep-seated brain tumors were employed. These calculations were performed with the Monte Carlo N-Particle (MCNP) code. A simple but effective beam shaping assembly (BSA) was calculated producing a high therapeutic gain compared to previously proposed facilities with the same nuclear reaction.

Materials for Low-Energy Neutron Radiation Shielding

NASA Technical Reports Server (NTRS)

Singleterry, Robert C., Jr.; Thibeault, Sheila A.

2000-01-01

Various candidate aircraft and spacecraft materials were analyzed and compared in a low-energy neutron environment using the Monte Carlo N-Particle (MCNP) transport code with an energy range up to 20 MeV. Some candidate materials have been tested in particle beams, and others seemed reasonable to analyze in this manner before deciding to test them. The two metal alloys analyzed are actual materials being designed into or used in aircraft and spacecraft today. This analysis shows that hydrogen-bearing materials have the best shielding characteristics over the metal alloys. It also shows that neutrons above 1 MeV are reflected out of the face of the slab better by larger quantities of carbon in the material. If a low-energy absorber is added to the material, fewer neutrons are transmitted through the material. Future analyses should focus on combinations of scatterers and absorbers to optimize these reaction channels and on the higher energy neutron component (above 50 MeV).

Quasi-heterogeneous efficient 3-D discrete ordinates CANDU calculations using Attila

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

Preeti, T.; Rulko, R.

2012-07-01

In this paper, 3-D quasi-heterogeneous large scale parallel Attila calculations of a generic CANDU test problem consisting of 42 complete fuel channels and a perpendicular to fuel reactivity device are presented. The solution method is that of discrete ordinates SN and the computational model is quasi-heterogeneous, i.e. fuel bundle is partially homogenized into five homogeneous rings consistently with the DRAGON code model used by the industry for the incremental cross-section generation. In calculations, the HELIOS-generated 45 macroscopic cross-sections library was used. This approach to CANDU calculations has the following advantages: 1) it allows detailed bundle (and eventually channel) power calculationsmore » for each fuel ring in a bundle, 2) it allows the exact reactivity device representation for its precise reactivity worth calculation, and 3) it eliminates the need for incremental cross-sections. Our results are compared to the reference Monte Carlo MCNP solution. In addition, the Attila SN method performance in CANDU calculations characterized by significant up scattering is discussed. (authors)« less

Doses and risks from the ingestion of Dounreay fuel fragments.

PubMed

Darley, P J; Charles, M W; Fell, T P; Harrison, J D

2003-01-01

The radiological implications of ingestion of nuclear fuel fragments present in the marine environment around Dounreay have been reassessed by using the Monte Carlo code MCNP to obtain improved estimates of the doses to target cells in the walls of the lower large intestine resulting from the passage of a fragment. The approach takes account of the reduction in dose due to attenuation within the intestinal wall and self-absorption of radiation in the fuel fragment itself. In addition, dose is calculated on the basis of a realistic estimate of the anatomical volume of the lumen, rather than being based on the average mass of the contents, as in the current ICRP model. Our best estimates of doses from the ingestion of the largest Dounreay particles are at least a factor of 30 lower than those predicted using the current ICRP model. The new ICRP model will address the issues raised here and provide improved estimates of dose.

The feasibility of well-logging measurements of arsenic levels using neutron-activation analysis

USGS Publications Warehouse

Oden, C.P.; Schweitzer, J.S.; McDowell, G.M.

2006-01-01

Arsenic is an extremely toxic metal, which poses a significant problem in many mining environments. Arsenic contamination is also a major problem in ground and surface waters. A feasibility study was conducted to determine if neutron-activation analysis is a practical method of measuring in situ arsenic levels. The response of hypothetical well-logging tools to arsenic was simulated using a readily available Monte Carlo simulation code (MCNP). Simulations were made for probes with both hyperpure germanium (HPGe) and bismuth germanate (BGO) detectors using accelerator and isotopic neutron sources. Both sources produce similar results; however, the BGO detector is much more susceptible to spectral interference than the HPGe detector. Spectral interference from copper can preclude low-level arsenic measurements when using the BGO detector. Results show that a borehole probe could be built that would measure arsenic concentrations of 100 ppm by weight to an uncertainty of 50 ppm in about 15 min. ?? 2006 Elsevier Ltd. All rights reserved.

Scattered Dose Calculations and Measurements in a Life-Like Mouse Phantom

PubMed Central

Welch, David; Turner, Leah; Speiser, Michael; Randers-Pehrson, Gerhard; Brenner, David J.

2017-01-01

Anatomically accurate phantoms are useful tools for radiation dosimetry studies. In this work, we demonstrate the construction of a new generation of life-like mouse phantoms in which the methods have been generalized to be applicable to the fabrication of any small animal. The mouse phantoms, with built-in density inhomogeneity, exhibit different scattering behavior dependent on where the radiation is delivered. Computer models of the mouse phantoms and a small animal irradiation platform were devised in Monte Carlo N-Particle code (MCNP). A baseline test replicating the irradiation system in a computational model shows minimal differences from experimental results from 50 Gy down to 0.1 Gy. We observe excellent agreement between scattered dose measurements and simulation results from X-ray irradiations focused at either the lung or the abdomen within our phantoms. This study demonstrates the utility of our mouse phantoms as measurement tools with the goal of using our phantoms to verify complex computational models. PMID:28140787

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.

A novel design of beam shaping assembly to use D-T neutron generator for BNCT.

PubMed

Kasesaz, Yaser; Karimi, Marjan

2016-12-01

In order to use 14.1MeV neutrons produced by d-T neutron generators, two special and novel Beam Shaping Assemblies (BSA), including multi-layer and hexagonal lattice have been suggested and the effect of them has been investigated by MCNP4C Monte Carlo code. The results show that the proposed BSA can provide the qualified epithermal neutron beam for BNCT. The final epithermal neutron flux is about 6e9 n/cm2.s. The final proposed BSA has some different advantages: 1) it consists of usual and well-known materials (Pb, Al, Fluental and Cd); 2) it has a simple geometry; 3) it does not need any additional gamma filter; 4) it can provide high flux of epithermal neutrons. As this type of neutron source is under development in the world, it seems that they can be used clinically in a hospital considering the proposed BSA. Copyright © 2016 Elsevier Ltd. All rights reserved.

Analysis of tritium production in concentric spheres of oralloy and /sup 6/LiD irradiated by 14-MeV neutrons

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

Fawcett, L.R. Jr.; Roberts, R.R. II; Hunter, R.E.

1988-03-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD with an oralloy core irradiated by a central source of 14-MeV neutrons have been calculated and compared with experimental measurements. The experimental assembly consisted of an oralloy sphere surrounded by three solid /sup 6/LiD concentric shells with ampules of /sup 6/LiH and /sup 7/LiH and activation foils located in several positions throughout the assembly. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport throughout the system, tritium production in the ampules, and foil activation. The overall experimentally observed-to-calculated ratiosmore » of tritium production were 0.996 +- 2.5% in /sup 6/Li ampules and 0.903 +- 5.2% in /sup 7/Li ampules. Observed-to-calculated ratios for foil activation are also presented. 11 refs., 4 figs., 7 tabs.« less

Determining organ dose conversion coefficients for external neutron irradiation by using a voxel mouse model.

PubMed

Zhang, Xiaomin; Xie, Xiangdong; Qu, Decheng; Ning, Jing; Zhou, Hongmei; Pan, Jie; Yang, Guoshan

2016-03-01

A set of fluence-to-dose conversion coefficients has been calculated for neutrons with energies <20 MeV using a developed voxel mouse model and Monte Carlo N-particle code (MCNP), for the purpose of neutron radiation effect evaluation. The calculation used 37 monodirectional monoenergetic neutron beams in the energy range 10(-9) MeV to 20 MeV, under five different source irradiation configurations: left lateral, right lateral, dorsal-ventral, ventral-dorsal, and isotropic. Neutron fluence-to-dose conversion coefficients for selected organs of the body were presented in the paper, and the effect of irradiation geometry conditions, neutron energy and the organ location on the organ dose was discussed. The results indicated that neutron dose conversion coefficients clearly show sensitivity to irradiation geometry at neutron energy below 1 MeV. © The Author 2015. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Ford Motor Company NDE facility shielding design.

PubMed

Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

SU-E-T-107: An Investigation Into Attenuation Effects On the Energy Spectrum for {sup 137}Cs Using Monte Carlo Techniques

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

Taneja, S; Bartol, L; Culberson, W

2015-06-15

Purpose: The calibration of radiation protection instrumentation including ionization chambers, scintillators, and Geiger Mueller (GM) counters used as survey meters are often done using {sup 137}Cs irradiators. During calibration, irradiators use a combination of attenuators with various thicknesses to modulate the beam to a known air-kerma rate. The variations in energy spectra as a result of these attenuators are not accounted for and may play a role in the energy-dependent response of survey meters. This study uses an experimentally validated irradiator geometry modeled in the MCNP5 transport code to characterize the effects of attenuation on the energy spectrum. Methods: Amore » Hopewell Designs G-10 {sup 137}Cs irradiator with lead attenuators of thicknesses of 0.635, 1.22, 2.22, and 4.32 cm, was used in this study. The irradiator geometry was modeled in MCNP5 and validated by comparing measured and simulated percent depth dose (PDD) and cross-field profiles. Variations in MCNP5 simulated spectra with increasing amounts of attenuation were characterized using the relative intensity of the 662 keV peak and the average energy. Results: Simulated and measured PDDs and profiles agreed within the associated uncertainty. The relative intensity of the 662 keV peak for simulated spectra normalized to the intensity of the unattenuated spectra ranged from 0.16% to 100% based on attenuation thickness. The average energy for simulated spectra for attenuators ranged from 582 keV with no attenuation to 653 keV with 5.54 cm of attenuation. Statistical uncertainty for MCNP5 simulations ranged from 0.11% to 3.69%. Conclusion: This study successfully used MCNP5 to validate a {sup 137}Cs irradiator geometry and characterize variations in energy spectra between different amounts of attenuation. Variations in the average energy of up to 12% were determined through simulations, and future work will aim to determine the effects of these differences on survey meter response and calibration.« less

Benchmarking and validation of a Geant4-SHADOW Monte Carlo simulation for dose calculations in microbeam radiation therapy.

PubMed

Cornelius, Iwan; Guatelli, Susanna; Fournier, Pauline; Crosbie, Jeffrey C; Sanchez Del Rio, Manuel; Bräuer-Krisch, Elke; Rosenfeld, Anatoly; Lerch, Michael

2014-05-01

Microbeam radiation therapy (MRT) is a synchrotron-based radiotherapy modality that uses high-intensity beams of spatially fractionated radiation to treat tumours. The rapid evolution of MRT towards clinical trials demands accurate treatment planning systems (TPS), as well as independent tools for the verification of TPS calculated dose distributions in order to ensure patient safety and treatment efficacy. Monte Carlo computer simulation represents the most accurate method of dose calculation in patient geometries and is best suited for the purpose of TPS verification. A Monte Carlo model of the ID17 biomedical beamline at the European Synchrotron Radiation Facility has been developed, including recent modifications, using the Geant4 Monte Carlo toolkit interfaced with the SHADOW X-ray optics and ray-tracing libraries. The code was benchmarked by simulating dose profiles in water-equivalent phantoms subject to irradiation by broad-beam (without spatial fractionation) and microbeam (with spatial fractionation) fields, and comparing against those calculated with a previous model of the beamline developed using the PENELOPE code. Validation against additional experimental dose profiles in water-equivalent phantoms subject to broad-beam irradiation was also performed. Good agreement between codes was observed, with the exception of out-of-field doses and toward the field edge for larger field sizes. Microbeam results showed good agreement between both codes and experimental results within uncertainties. Results of the experimental validation showed agreement for different beamline configurations. The asymmetry in the out-of-field dose profiles due to polarization effects was also investigated, yielding important information for the treatment planning process in MRT. This work represents an important step in the development of a Monte Carlo-based independent verification tool for treatment planning in MRT.

Control of the Low-energy X-rays by Using MCNP5 and Numerical Analysis for a New Concept Intra-oral X-ray Imaging System

NASA Astrophysics Data System (ADS)

Huh, Jangyong; Ji, Yunseo; Lee, Rena

2018-05-01

An X-ray control algorithm to modulate the X-ray intensity distribution over the FOV (field of view) has been developed by using numerical analysis and MCNP5, a particle transport simulation code on the basis of the Monte Carlo method. X-rays, which are widely used in medical diagnostic imaging, should be controlled in order to maximize the performance of the X-ray imaging system. However, transporting X-rays, like a liquid or a gas is conveyed through a physical form such as pipes, is not possible. In the present study, an X-ray control algorithm and technique to uniformize the Xray intensity projected on the image sensor were developed using a flattening filter and a collimator in order to alleviate the anisotropy of the distribution of X-rays due to intrinsic features of the X-ray generator. The proposed method, which is combined with MCNP5 modeling and numerical analysis, aimed to optimize a flattening filter and a collimator for a uniform distribution of X-rays. Their size and shape were estimated from the method. The simulation and the experimental results both showed that the method yielded an intensity distribution over an X-ray field of 6×4 cm2 at SID (source to image-receptor distance) of 5 cm with a uniformity of more than 90% when the flattening filter and the collimator were mounted on the system. The proposed algorithm and technique are not only confined to flattening filter development but can also be applied for other X-ray related research and development efforts.

Optimization of the Monte Carlo code for modeling of photon migration in tissue.

PubMed

Zołek, Norbert S; Liebert, Adam; Maniewski, Roman

2006-10-01

The Monte Carlo method is frequently used to simulate light transport in turbid media because of its simplicity and flexibility, allowing to analyze complicated geometrical structures. Monte Carlo simulations are, however, time consuming because of the necessity to track the paths of individual photons. The time consuming computation is mainly associated with the calculation of the logarithmic and trigonometric functions as well as the generation of pseudo-random numbers. In this paper, the Monte Carlo algorithm was developed and optimized, by approximation of the logarithmic and trigonometric functions. The approximations were based on polynomial and rational functions, and the errors of these approximations are less than 1% of the values of the original functions. The proposed algorithm was verified by simulations of the time-resolved reflectance at several source-detector separations. The results of the calculation using the approximated algorithm were compared with those of the Monte Carlo simulations obtained with an exact computation of the logarithm and trigonometric functions as well as with the solution of the diffusion equation. The errors of the moments of the simulated distributions of times of flight of photons (total number of photons, mean time of flight and variance) are less than 2% for a range of optical properties, typical of living tissues. The proposed approximated algorithm allows to speed up the Monte Carlo simulations by a factor of 4. The developed code can be used on parallel machines, allowing for further acceleration.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

A Monte Carlo and experimental investigation of the dosimetric behavior of low- and medium-perturbation diodes used for entrance in vivo dosimetry in megavoltage photon beams.

PubMed

Mosleh-Shirazi, Mohammad Amin; Karbasi, Sareh; Shahbazi-Gahrouei, Daryoush; Monadi, Shahram

2012-11-08

Full buildup diodes can cause significant dose perturbation if they are used on most or all of radiotherapy fractions. Given the importance of frequent in vivo measurements in complex treatments, using thin buildup (low-perturbation) diodes instead is gathering interest. However, such diodes are strictly unsuitable for high-energy photons; therefore, their use requires evaluation and careful measurement of correction factors (CFs). There is little published data on such factors for low-perturbation diodes, and none on diode characterization for 9 MV X-rays. We report on MCNP4c Monte Carlo models of low-perturbation (EDD5) and medium-perturbation (EDP10) diodes, and a comparison of source-to-surface distance, field size, temperature, and orientation CFs for cobalt-60 and 9 MV beams. Most of the simulation results were within 4% of the measurements. The results suggest against the use of the EDD5 in axial angles beyond ± 50° and exceeding the range 0° to +50° tilt angle at 9 MV. Outside these ranges, although the EDD5 can be used for accurate in vivo dosimetry at 9 MV, its CF variations were found to be 1.5-7.1 times larger than the EDP10 and, therefore, should be applied carefully. Finally, the MCNP diode models are sufficiently reliable tools for independent verification of potentially inaccurate measurements.

Building Process Improvement Business Cases Using Bayesian Belief Networks and Monte Carlo Simulation

DTIC Science & Technology

2009-07-01

simulation. The pilot described in this paper used this two-step approach within a Define, Measure, Analyze, Improve, and Control ( DMAIC ) framework to...networks, BBN, Monte Carlo simulation, DMAIC , Six Sigma, business case 15. NUMBER OF PAGES 35 16. PRICE CODE 17. SECURITY CLASSIFICATION OF

Million-body star cluster simulations: comparisons between Monte Carlo and direct N-body

NASA Astrophysics Data System (ADS)

Rodriguez, Carl L.; Morscher, Meagan; Wang, Long; Chatterjee, Sourav; Rasio, Frederic A.; Spurzem, Rainer

2016-12-01

We present the first detailed comparison between million-body globular cluster simulations computed with a Hénon-type Monte Carlo code, CMC, and a direct N-body code, NBODY6++GPU. Both simulations start from an identical cluster model with 106 particles, and include all of the relevant physics needed to treat the system in a highly realistic way. With the two codes `frozen' (no fine-tuning of any free parameters or internal algorithms of the codes) we find good agreement in the overall evolution of the two models. Furthermore, we find that in both models, large numbers of stellar-mass black holes (>1000) are retained for 12 Gyr. Thus, the very accurate direct N-body approach confirms recent predictions that black holes can be retained in present-day, old globular clusters. We find only minor disagreements between the two models and attribute these to the small-N dynamics driving the evolution of the cluster core for which the Monte Carlo assumptions are less ideal. Based on the overwhelming general agreement between the two models computed using these vastly different techniques, we conclude that our Monte Carlo approach, which is more approximate, but dramatically faster compared to the direct N-body, is capable of producing an accurate description of the long-term evolution of massive globular clusters even when the clusters contain large populations of stellar-mass black holes.

Monte Carlo simulations of {sup 3}He ion physical characteristics in a water phantom and evaluation of radiobiological effectiveness

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

Taleei, Reza; Guan, Fada; Peeler, Chris

Purpose: {sup 3}He ions may hold great potential for clinical therapy because of both their physical and biological properties. In this study, the authors investigated the physical properties, i.e., the depth-dose curves from primary and secondary particles, and the energy distributions of helium ({sup 3}He) ions. A relative biological effectiveness (RBE) model was applied to assess the biological effectiveness on survival of multiple cell lines. Methods: In light of the lack of experimental measurements and cross sections, the authors used Monte Carlo methods to study the energy deposition of {sup 3}He ions. The transport of {sup 3}He ions in watermore » was simulated by using three Monte Carlo codes—FLUKA, GEANT4, and MCNPX—for incident beams with Gaussian energy distributions with average energies of 527 and 699 MeV and a full width at half maximum of 3.3 MeV in both cases. The RBE of each was evaluated by using the repair-misrepair-fixation model. In all of the simulations with each of the three Monte Carlo codes, the same geometry and primary beam parameters were used. Results: Energy deposition as a function of depth and energy spectra with high resolution was calculated on the central axis of the beam. Secondary proton dose from the primary {sup 3}He beams was predicted quite differently by the three Monte Carlo systems. The predictions differed by as much as a factor of 2. Microdosimetric parameters such as dose mean lineal energy (y{sub D}), frequency mean lineal energy (y{sub F}), and frequency mean specific energy (z{sub F}) were used to characterize the radiation beam quality at four depths of the Bragg curve. Calculated RBE values were close to 1 at the entrance, reached on average 1.8 and 1.6 for prostate and head and neck cancer cell lines at the Bragg peak for both energies, but showed some variations between the different Monte Carlo codes. Conclusions: Although the Monte Carlo codes provided different results in energy deposition and especially in secondary particle production (most of the differences between the three codes were observed close to the Bragg peak, where the energy spectrum broadens), the results in terms of RBE were generally similar.« less

QMCPACK: an open source ab initio quantum Monte Carlo package for the electronic structure of atoms, molecules and solids

NASA Astrophysics Data System (ADS)

Kim, Jeongnim; Baczewski, Andrew D.; Beaudet, Todd D.; Benali, Anouar; Chandler Bennett, M.; Berrill, Mark A.; Blunt, Nick S.; Josué Landinez Borda, Edgar; Casula, Michele; Ceperley, David M.; Chiesa, Simone; Clark, Bryan K.; Clay, Raymond C., III; Delaney, Kris T.; Dewing, Mark; Esler, Kenneth P.; Hao, Hongxia; Heinonen, Olle; Kent, Paul R. C.; Krogel, Jaron T.; Kylänpää, Ilkka; Li, Ying Wai; Lopez, M. Graham; Luo, Ye; Malone, Fionn D.; Martin, Richard M.; Mathuriya, Amrita; McMinis, Jeremy; Melton, Cody A.; Mitas, Lubos; Morales, Miguel A.; Neuscamman, Eric; Parker, William D.; Pineda Flores, Sergio D.; Romero, Nichols A.; Rubenstein, Brenda M.; Shea, Jacqueline A. R.; Shin, Hyeondeok; Shulenburger, Luke; Tillack, Andreas F.; Townsend, Joshua P.; Tubman, Norm M.; Van Der Goetz, Brett; Vincent, Jordan E.; ChangMo Yang, D.; Yang, Yubo; Zhang, Shuai; Zhao, Luning

2018-05-01

QMCPACK is an open source quantum Monte Carlo package for ab initio electronic structure calculations. It supports calculations of metallic and insulating solids, molecules, atoms, and some model Hamiltonians. Implemented real space quantum Monte Carlo algorithms include variational, diffusion, and reptation Monte Carlo. QMCPACK uses Slater–Jastrow type trial wavefunctions in conjunction with a sophisticated optimizer capable of optimizing tens of thousands of parameters. The orbital space auxiliary-field quantum Monte Carlo method is also implemented, enabling cross validation between different highly accurate methods. The code is specifically optimized for calculations with large numbers of electrons on the latest high performance computing architectures, including multicore central processing unit and graphical processing unit systems. We detail the program’s capabilities, outline its structure, and give examples of its use in current research calculations. The package is available at http://qmcpack.org.

QMCPACK: an open source ab initio quantum Monte Carlo package for the electronic structure of atoms, molecules and solids.

PubMed

Kim, Jeongnim; Baczewski, Andrew T; Beaudet, Todd D; Benali, Anouar; Bennett, M Chandler; Berrill, Mark A; Blunt, Nick S; Borda, Edgar Josué Landinez; Casula, Michele; Ceperley, David M; Chiesa, Simone; Clark, Bryan K; Clay, Raymond C; Delaney, Kris T; Dewing, Mark; Esler, Kenneth P; Hao, Hongxia; Heinonen, Olle; Kent, Paul R C; Krogel, Jaron T; Kylänpää, Ilkka; Li, Ying Wai; Lopez, M Graham; Luo, Ye; Malone, Fionn D; Martin, Richard M; Mathuriya, Amrita; McMinis, Jeremy; Melton, Cody A; Mitas, Lubos; Morales, Miguel A; Neuscamman, Eric; Parker, William D; Pineda Flores, Sergio D; Romero, Nichols A; Rubenstein, Brenda M; Shea, Jacqueline A R; Shin, Hyeondeok; Shulenburger, Luke; Tillack, Andreas F; Townsend, Joshua P; Tubman, Norm M; Van Der Goetz, Brett; Vincent, Jordan E; Yang, D ChangMo; Yang, Yubo; Zhang, Shuai; Zhao, Luning

2018-05-16

QMCPACK is an open source quantum Monte Carlo package for ab initio electronic structure calculations. It supports calculations of metallic and insulating solids, molecules, atoms, and some model Hamiltonians. Implemented real space quantum Monte Carlo algorithms include variational, diffusion, and reptation Monte Carlo. QMCPACK uses Slater-Jastrow type trial wavefunctions in conjunction with a sophisticated optimizer capable of optimizing tens of thousands of parameters. The orbital space auxiliary-field quantum Monte Carlo method is also implemented, enabling cross validation between different highly accurate methods. The code is specifically optimized for calculations with large numbers of electrons on the latest high performance computing architectures, including multicore central processing unit and graphical processing unit systems. We detail the program's capabilities, outline its structure, and give examples of its use in current research calculations. The package is available at http://qmcpack.org.

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

Ortiz-Ramírez, Pablo, E-mail: rapeitor@ug.uchile.cl; Ruiz, Andrés

The Monte Carlo simulation of the gamma spectroscopy systems is a common practice in these days. The most popular softwares to do this are MCNP and Geant4 codes. The intrinsic spatial efficiency method is a general and absolute method to determine the absolute efficiency of a spectroscopy system for any extended sources, but this was only demonstrated experimentally for cylindrical sources. Due to the difficulty that the preparation of sources with any shape represents, the simplest way to do this is by the simulation of the spectroscopy system and the source. In this work we present the validation of themore » intrinsic spatial efficiency method for sources with different geometries and for photons with an energy of 661.65 keV. In the simulation the matrix effects (the auto-attenuation effect) are not considered, therefore these results are only preliminaries. The MC simulation is carried out using the FLUKA code and the absolute efficiency of the detector is determined using two methods: the statistical count of Full Energy Peak (FEP) area (traditional method) and the intrinsic spatial efficiency method. The obtained results show total agreement between the absolute efficiencies determined by the traditional method and the intrinsic spatial efficiency method. The relative bias is lesser than 1% in all cases.« less

TH-E-18A-01: Developments in Monte Carlo Methods for Medical Imaging

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

Badal, A; Zbijewski, W; Bolch, W

Monte Carlo simulation methods are widely used in medical physics research and are starting to be implemented in clinical applications such as radiation therapy planning systems. Monte Carlo simulations offer the capability to accurately estimate quantities of interest that are challenging to measure experimentally while taking into account the realistic anatomy of an individual patient. Traditionally, practical application of Monte Carlo simulation codes in diagnostic imaging was limited by the need for large computational resources or long execution times. However, recent advancements in high-performance computing hardware, combined with a new generation of Monte Carlo simulation algorithms and novel postprocessing methods,more » are allowing for the computation of relevant imaging parameters of interest such as patient organ doses and scatter-to-primaryratios in radiographic projections in just a few seconds using affordable computational resources. Programmable Graphics Processing Units (GPUs), for example, provide a convenient, affordable platform for parallelized Monte Carlo executions that yield simulation times on the order of 10{sup 7} xray/ s. Even with GPU acceleration, however, Monte Carlo simulation times can be prohibitive for routine clinical practice. To reduce simulation times further, variance reduction techniques can be used to alter the probabilistic models underlying the x-ray tracking process, resulting in lower variance in the results without biasing the estimates. Other complementary strategies for further reductions in computation time are denoising of the Monte Carlo estimates and estimating (scoring) the quantity of interest at a sparse set of sampling locations (e.g. at a small number of detector pixels in a scatter simulation) followed by interpolation. Beyond reduction of the computational resources required for performing Monte Carlo simulations in medical imaging, the use of accurate representations of patient anatomy is crucial to the virtual generation of medical images and accurate estimation of radiation dose and other imaging parameters. For this, detailed computational phantoms of the patient anatomy must be utilized and implemented within the radiation transport code. Computational phantoms presently come in one of three format types, and in one of four morphometric categories. Format types include stylized (mathematical equation-based), voxel (segmented CT/MR images), and hybrid (NURBS and polygon mesh surfaces). Morphometric categories include reference (small library of phantoms by age at 50th height/weight percentile), patient-dependent (larger library of phantoms at various combinations of height/weight percentiles), patient-sculpted (phantoms altered to match the patient's unique outer body contour), and finally, patient-specific (an exact representation of the patient with respect to both body contour and internal anatomy). The existence and availability of these phantoms represents a very important advance for the simulation of realistic medical imaging applications using Monte Carlo methods. New Monte Carlo simulation codes need to be thoroughly validated before they can be used to perform novel research. Ideally, the validation process would involve comparison of results with those of an experimental measurement, but accurate replication of experimental conditions can be very challenging. It is very common to validate new Monte Carlo simulations by replicating previously published simulation results of similar experiments. This process, however, is commonly problematic due to the lack of sufficient information in the published reports of previous work so as to be able to replicate the simulation in detail. To aid in this process, the AAPM Task Group 195 prepared a report in which six different imaging research experiments commonly performed using Monte Carlo simulations are described and their results provided. The simulation conditions of all six cases are provided in full detail, with all necessary data on material composition, source, geometry, scoring and other parameters provided. The results of these simulations when performed with the four most common publicly available Monte Carlo packages are also provided in tabular form. The Task Group 195 Report will be useful for researchers needing to validate their Monte Carlo work, and for trainees needing to learn Monte Carlo simulation methods. In this symposium we will review the recent advancements in highperformance computing hardware enabling the reduction in computational resources needed for Monte Carlo simulations in medical imaging. We will review variance reduction techniques commonly applied in Monte Carlo simulations of medical imaging systems and present implementation strategies for efficient combination of these techniques with GPU acceleration. Trade-offs involved in Monte Carlo acceleration by means of denoising and “sparse sampling” will be discussed. A method for rapid scatter correction in cone-beam CT (<5 min/scan) will be presented as an illustration of the simulation speeds achievable with optimized Monte Carlo simulations. We will also discuss the development, availability, and capability of the various combinations of computational phantoms for Monte Carlo simulation of medical imaging systems. Finally, we will review some examples of experimental validation of Monte Carlo simulations and will present the AAPM Task Group 195 Report. Learning Objectives: Describe the advances in hardware available for performing Monte Carlo simulations in high performance computing environments. Explain variance reduction, denoising and sparse sampling techniques available for reduction of computational time needed for Monte Carlo simulations of medical imaging. List and compare the computational anthropomorphic phantoms currently available for more accurate assessment of medical imaging parameters in Monte Carlo simulations. Describe experimental methods used for validation of Monte Carlo simulations in medical imaging. Describe the AAPM Task Group 195 Report and its use for validation and teaching of Monte Carlo simulations in medical imaging.« less

Comparison of Geant4-DNA simulation of S-values with other Monte Carlo codes

NASA Astrophysics Data System (ADS)

André, T.; Morini, F.; Karamitros, M.; Delorme, R.; Le Loirec, C.; Campos, L.; Champion, C.; Groetz, J.-E.; Fromm, M.; Bordage, M.-C.; Perrot, Y.; Barberet, Ph.; Bernal, M. A.; Brown, J. M. C.; Deleuze, M. S.; Francis, Z.; Ivanchenko, V.; Mascialino, B.; Zacharatou, C.; Bardiès, M.; Incerti, S.

2014-01-01

Monte Carlo simulations of S-values have been carried out with the Geant4-DNA extension of the Geant4 toolkit. The S-values have been simulated for monoenergetic electrons with energies ranging from 0.1 keV up to 20 keV, in liquid water spheres (for four radii, chosen between 10 nm and 1 μm), and for electrons emitted by five isotopes of iodine (131, 132, 133, 134 and 135), in liquid water spheres of varying radius (from 15 μm up to 250 μm). The results have been compared to those obtained from other Monte Carlo codes and from other published data. The use of the Kolmogorov-Smirnov test has allowed confirming the statistical compatibility of all simulation results.

Parameter dependence of the MCNP electron transport in determining dose distributions.

PubMed

Reynaert, N; Palmans, H; Thierens, H; Jeraj, R

2002-10-01

In this paper, a detailed study of the electron transport in MCNP is performed, separating the effects of the energy binning technique on the energy loss rate, the scattering angles, and the sub-step length as a function of energy. As this problem is already well known, in this paper we focus on the explanation as to why the default mode of MCNP can lead to large deviations. The resolution dependence was investigated as well. An error in the MCNP code in the energy binning technique in the default mode (DBCN 18 card = 0) was revealed, more specific in the updating of cross sections when a sub-step is performed corresponding to a high-energy loss. This updating error is not present in the ITS mode (DBCN 18 card = 1) and leads to a systematically lower dose deposition rate in the default mode. The effect is present for all energies studied (0.5-10 MeV) and depends on the geometrical resolution of the scoring regions and the energy grid resolution. The effect of the energy binning technique is of the same order of that of the updating error for energies below 2 MeV, and becomes less important for higher energies. For a 1 MeV point source surrounded by homogeneous water, the deviation of the default MCNP results at short distances attains 9% and remains approximately the same for all energies. This effect could be corrected by removing the completion of an energy step each time an electron changes from an energy bin during a sub-step. Another solution consists of performing all calculations in the ITS mode. Another problem is the resolution dependence, even in the ITS mode. The higher the resolution is chosen (the smaller the scoring regions) the faster the energy is deposited along the electron track. It is proven that this is caused by starting a new energy step when crossing a surface. The resolution effect should be investigated for every specific case when calculating dose distributions around beta sources. The resolution should not be higher than 0.85*(1-EFAC)*CSDA, where EFAC is the energy loss per energy step and CSDA a continuous slowing down approximation range. This effect could as well be removed by determining the cross sections for energy loss and multiple scattering at the average energy of an energy step and by sampling the cross sections for each sub-step. Overall, we conclude that MCNP cannot be used without a caution due to possible errors in the electron transport. When care is taken, it is possible to obtain correct results that are in agreement with other Monte Carlo codes.

Metis: A Pure Metropolis Markov Chain Monte Carlo Bayesian Inference Library

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

Bates, Cameron Russell; Mckigney, Edward Allen

The use of Bayesian inference in data analysis has become the standard for large scienti c experiments [1, 2]. The Monte Carlo Codes Group(XCP-3) at Los Alamos has developed a simple set of algorithms currently implemented in C++ and Python to easily perform at-prior Markov Chain Monte Carlo Bayesian inference with pure Metropolis sampling. These implementations are designed to be user friendly and extensible for customization based on speci c application requirements. This document describes the algorithmic choices made and presents two use cases.

EUPDF: An Eulerian-Based Monte Carlo Probability Density Function (PDF) Solver. User's Manual

NASA Technical Reports Server (NTRS)

Raju, M. S.

1998-01-01

EUPDF is an Eulerian-based Monte Carlo PDF solver developed for application with sprays, combustion, parallel computing and unstructured grids. It is designed to be massively parallel and could easily be coupled with any existing gas-phase flow and spray solvers. The solver accommodates the use of an unstructured mesh with mixed elements of either triangular, quadrilateral, and/or tetrahedral type. The manual provides the user with the coding required to couple the PDF code to any given flow code and a basic understanding of the EUPDF code structure as well as the models involved in the PDF formulation. The source code of EUPDF will be available with the release of the National Combustion Code (NCC) as a complete package.

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

NASA Astrophysics Data System (ADS)

Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

2017-04-01

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

General Monte Carlo reliability simulation code including common mode failures and HARP fault/error-handling

NASA Technical Reports Server (NTRS)

Platt, M. E.; Lewis, E. E.; Boehm, F.

1991-01-01

A Monte Carlo Fortran computer program was developed that uses two variance reduction techniques for computing system reliability applicable to solving very large highly reliable fault-tolerant systems. The program is consistent with the hybrid automated reliability predictor (HARP) code which employs behavioral decomposition and complex fault-error handling models. This new capability is called MC-HARP which efficiently solves reliability models with non-constant failures rates (Weibull). Common mode failure modeling is also a specialty.

The Serpent Monte Carlo Code: Status, Development and Applications in 2013

NASA Astrophysics Data System (ADS)

Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

2014-06-01

The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

An unbiased Hessian representation for Monte Carlo PDFs.

PubMed

Carrazza, Stefano; Forte, Stefano; Kassabov, Zahari; Latorre, José Ignacio; Rojo, Juan

We develop a methodology for the construction of a Hessian representation of Monte Carlo sets of parton distributions, based on the use of a subset of the Monte Carlo PDF replicas as an unbiased linear basis, and of a genetic algorithm for the determination of the optimal basis. We validate the methodology by first showing that it faithfully reproduces a native Monte Carlo PDF set (NNPDF3.0), and then, that if applied to Hessian PDF set (MMHT14) which was transformed into a Monte Carlo set, it gives back the starting PDFs with minimal information loss. We then show that, when applied to a large Monte Carlo PDF set obtained as combination of several underlying sets, the methodology leads to a Hessian representation in terms of a rather smaller set of parameters (MC-H PDFs), thereby providing an alternative implementation of the recently suggested Meta-PDF idea and a Hessian version of the recently suggested PDF compression algorithm (CMC-PDFs). The mc2hessian conversion code is made publicly available together with (through LHAPDF6) a Hessian representations of the NNPDF3.0 set, and the MC-H PDF set.

NOTE: Monte Carlo evaluation of kerma in an HDR brachytherapy bunker

NASA Astrophysics Data System (ADS)

Pérez-Calatayud, J.; Granero, D.; Ballester, F.; Casal, E.; Crispin, V.; Puchades, V.; León, A.; Verdú, G.

2004-12-01

In recent years, the use of high dose rate (HDR) after-loader machines has greatly increased due to the shift from traditional Cs-137/Ir-192 low dose rate (LDR) to HDR brachytherapy. The method used to calculate the required concrete and, where appropriate, lead shielding in the door is based on analytical methods provided by documents published by the ICRP, the IAEA and the NCRP. The purpose of this study is to perform a more realistic kerma evaluation at the entrance maze door of an HDR bunker using the Monte Carlo code GEANT4. The Monte Carlo results were validated experimentally. The spectrum at the maze entrance door, obtained with Monte Carlo, has an average energy of about 110 keV, maintaining a similar value along the length of the maze. The comparison of results from the aforementioned values with the Monte Carlo ones shows that results obtained using the albedo coefficient from the ICRP document more closely match those given by the Monte Carlo method, although the maximum value given by MC calculations is 30% greater.

Scaling GDL for Multi-cores to Process Planck HFI Beams Monte Carlo on HPC

NASA Astrophysics Data System (ADS)

Coulais, A.; Schellens, M.; Duvert, G.; Park, J.; Arabas, S.; Erard, S.; Roudier, G.; Hivon, E.; Mottet, S.; Laurent, B.; Pinter, M.; Kasradze, N.; Ayad, M.

2014-05-01

After reviewing the majors progress done in GDL -now in 0.9.4- on performance and plotting capabilities since ADASS XXI paper (Coulais et al. 2012), we detail how a large code for Planck HFI beams Monte Carlo was successfully transposed from IDL to GDL on HPC.

A Novel Implementation of Massively Parallel Three Dimensional Monte Carlo Radiation Transport

NASA Astrophysics Data System (ADS)

Robinson, P. B.; Peterson, J. D. L.

2005-12-01

The goal of our summer project was to implement the difference formulation for radiation transport into Cosmos++, a multidimensional, massively parallel, magneto hydrodynamics code for astrophysical applications (Peter Anninos - AX). The difference formulation is a new method for Symbolic Implicit Monte Carlo thermal transport (Brooks and Szöke - PAT). Formerly, simultaneous implementation of fully implicit Monte Carlo radiation transport in multiple dimensions on multiple processors had not been convincingly demonstrated. We found that a combination of the difference formulation and the inherent structure of Cosmos++ makes such an implementation both accurate and straightforward. We developed a "nearly nearest neighbor physics" technique to allow each processor to work independently, even with a fully implicit code. This technique coupled with the increased accuracy of an implicit Monte Carlo solution and the efficiency of parallel computing systems allows us to demonstrate the possibility of massively parallel thermal transport. This work was performed under the auspices of the U.S. Department of Energy by University of California Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48

Accelerating Monte Carlo simulations of photon transport in a voxelized geometry using a massively parallel graphics processing unit.

PubMed

Badal, Andreu; Badano, Aldo

2009-11-01

It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDATM programming model (NVIDIA Corporation, Santa Clara, CA). An outline of the new code and a sample x-ray imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.

Analysis of sewage sludge using an experimental prompt gamma neutron activation analysis (pgnaa) set-up with an am-be source

NASA Astrophysics Data System (ADS)

Idiri, Z.; Redjem, F.; Beloudah, N.

2016-09-01

An experimental PGNAA set-up using a 1 Ci Am-Be source has been developed and used for analysis of bulk sewage sludge samples issued from a wastewater treatment plant situated in an industrial area of Algiers. The sample dimensions were optimized using thermal neutron flux calculations carried out with the MCNP5 Monte Carlo Code. A methodology is then proposed to perform quantitative analysis using the absolute method. For this, average thermal neutron flux inside the sludge samples is deduced using average thermal neutron flux in reference water samples and thermal flux measurements with the aid of a 3He neutron detector. The average absolute gamma detection efficiency is determined using the prompt gammas emitted by chlorine dissolved in a water sample. The gamma detection efficiency is normalized for sludge samples using gamma attenuation factors calculated with the MCNP5 code for water and sludge. Wet and dehydrated sludge samples were analyzed. Nutritive elements (Ca, N, P, K) and heavy metals elements like Cr and Mn were determined. For some elements, the PGNAA values were compared to those obtained using Atomic Absorption Spectroscopy (AAS) and Inductively Coupled Plasma (ICP) methods. Good agreement is observed between the different values. Heavy element concentrations are very high compared to normal values; this is related to the fact that the wastewater treatment plant is treating not only domestic but also industrial wastewater that is probably rejected by industries without removal of pollutant elements. The detection limits for almost all elements of interest are sufficiently low for the method to be well suited for such analysis.

Geant4 Modifications for Accurate Fission Simulations

NASA Astrophysics Data System (ADS)

Tan, Jiawei; Bendahan, Joseph

Monte Carlo is one of the methods to simulate the generation and transport of radiation through matter. The most widely used radiation simulation codes are MCNP and Geant4. The simulation of fission production and transport by MCNP has been thoroughly benchmarked. There is an increasing number of users that prefer using Geant4 due to the flexibility of adding features. However, it has been found that Geant4 does not have the proper fission-production cross sections and does not produce the correct fission products. To achieve accurate results for studies in fissionable material applications, Geant4 was modified to correct these inaccuracies and to add new capabilities. The fission model developed by the Lawrence Livermore National Laboratory was integrated into the neutron-fission modeling package. The photofission simulation capability was enabled using the same neutron-fission library under the assumption that nuclei fission in the same way, independent of the excitation source. The modified fission code provides the correct multiplicity of prompt neutrons and gamma rays, and produces delayed gamma rays and neutrons with time and energy dependencies that are consistent with ENDF/B-VII. The delayed neutrons are now directly produced by a custom package that bypasses the fragment cascade model. The modifications were made for U-235, U-238 and Pu-239 isotopes; however, the new framework allows adding new isotopes easily. The SLAC nuclear data library is used for simulation of isotopes with an atomic number above 92 because it is not available in Geant4. Results of the modified Geant4.10.1 package of neutron-fission and photofission for prompt and delayed radiation are compared with ENDFB-VII and with results produced with the original package.

Dose conversion coefficients based on the Chinese mathematical phantom and MCNP code for external photon irradiation.

PubMed

Qiu, Rui; Li, Junli; Zhang, Zhan; Liu, Liye; Bi, Lei; Ren, Li

2009-02-01

A set of conversion coefficients from kerma free-in-air to the organ-absorbed dose are presented for external monoenergetic photon beams from 10 keV to 10 MeV based on the Chinese mathematical phantom, a whole-body mathematical phantom model. The model was developed based on the methods of the Oak Ridge National Laboratory mathematical phantom series and data from the Chinese Reference Man and the Reference Asian Man. This work is carried out to obtain the conversion coefficients based on this model, which represents the characteristics of the Chinese population, as the anatomical parameters of the Chinese are different from those of Caucasians. Monte Carlo simulation with MCNP code is carried out to calculate the organ dose conversion coefficients. Before the calculation, the effects from the physics model and tally type are investigated, considering both the calculation efficiency and precision. In the calculation irradiation conditions include anterior-posterior, posterior-anterior, right lateral, left lateral, rotational and isotropic geometries. Conversion coefficients from this study are compared with those recommended in the Publication 74 of International Commission on Radiological Protection (ICRP74) since both the sets of data are calculated with mathematical phantoms. Overall, consistency between the two sets of data is observed and the difference for more than 60% of the data is below 10%. However, significant deviations are also found, mainly for the superficial organs (up to 65.9%) and bone surface (up to 66%). The big difference of the dose conversion coefficients for the superficial organs at high photon energy could be ascribed to kerma approximation for the data in ICRP74. Both anatomical variations between races and the calculation method contribute to the difference of the data for bone surface.

Common radiation analysis model for 75,000 pound thrust NERVA engine (1137400E)

NASA Technical Reports Server (NTRS)

Warman, E. A.; Lindsey, B. A.

1972-01-01

The mathematical model and sources of radiation used for the radiation analysis and shielding activities in support of the design of the 1137400E version of the 75,000 lbs thrust NERVA engine are presented. The nuclear subsystem (NSS) and non-nuclear components are discussed. The geometrical model for the NSS is two dimensional as required for the DOT discrete ordinates computer code or for an azimuthally symetrical three dimensional Point Kernel or Monte Carlo code. The geometrical model for the non-nuclear components is three dimensional in the FASTER geometry format. This geometry routine is inherent in the ANSC versions of the QAD and GGG Point Kernal programs and the COHORT Monte Carlo program. Data are included pertaining to a pressure vessel surface radiation source data tape which has been used as the basis for starting ANSC analyses with the DASH code to bridge into the COHORT Monte Carlo code using the WANL supplied DOT angular flux leakage data. In addition to the model descriptions and sources of radiation, the methods of analyses are briefly described.

Supernova Light Curves and Spectra from Two Different Codes: Supernu and Phoenix

NASA Astrophysics Data System (ADS)

Van Rossum, Daniel R; Wollaeger, Ryan T

2014-08-01

The observed similarities between light curve shapes from Type Ia supernovae, and in particular the correlation of light curve shape and brightness, have been actively studied for more than two decades. In recent years, hydronamic simulations of white dwarf explosions have advanced greatly, and multiple mechanisms that could potentially produce Type Ia supernovae have been explored in detail. The question which of the proposed mechanisms is (or are) possibly realized in nature remains challenging to answer, but detailed synthetic light curves and spectra from explosion simulations are very helpful and important guidelines towards answering this question.We present results from a newly developed radiation transport code, Supernu. Supernu solves the supernova radiation transfer problem uses a novel technique based on a hybrid between Implicit Monte Carlo and Discrete Diffusion Monte Carlo. This technique enhances the efficiency with respect to traditional implicit monte carlo codes and thus lends itself perfectly for multi-dimensional simulations. We show direct comparisons of light curves and spectra from Type Ia simulations with Supernu versus the legacy Phoenix code.

Development of a Research Reactor Protocol for Neutron Multiplication Measurements

DOE PAGES

Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.; ...

2018-03-20

A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

Development of a Research Reactor Protocol for Neutron Multiplication Measurements

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

Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.

A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

Fast GPU-based Monte Carlo simulations for LDR prostate brachytherapy.

PubMed

Bonenfant, Éric; Magnoux, Vincent; Hissoiny, Sami; Ozell, Benoît; Beaulieu, Luc; Després, Philippe

2015-07-07

The aim of this study was to evaluate the potential of bGPUMCD, a Monte Carlo algorithm executed on Graphics Processing Units (GPUs), for fast dose calculations in permanent prostate implant dosimetry. It also aimed to validate a low dose rate brachytherapy source in terms of TG-43 metrics and to use this source to compute dose distributions for permanent prostate implant in very short times. The physics of bGPUMCD was reviewed and extended to include Rayleigh scattering and fluorescence from photoelectric interactions for all materials involved. The radial and anisotropy functions were obtained for the Nucletron SelectSeed in TG-43 conditions. These functions were compared to those found in the MD Anderson Imaging and Radiation Oncology Core brachytherapy source registry which are considered the TG-43 reference values. After appropriate calibration of the source, permanent prostate implant dose distributions were calculated for four patients and compared to an already validated Geant4 algorithm. The radial function calculated from bGPUMCD showed excellent agreement (differences within 1.3%) with TG-43 accepted values. The anisotropy functions at r = 1 cm and r = 4 cm were within 2% of TG-43 values for angles over 17.5°. For permanent prostate implants, Monte Carlo-based dose distributions with a statistical uncertainty of 1% or less for the target volume were obtained in 30 s or less for 1 × 1 × 1 mm(3) calculation grids. Dosimetric indices were very similar (within 2.7%) to those obtained with a validated, independent Monte Carlo code (Geant4) performing the calculations for the same cases in a much longer time (tens of minutes to more than a hour). bGPUMCD is a promising code that lets envision the use of Monte Carlo techniques in a clinical environment, with sub-minute execution times on a standard workstation. Future work will explore the use of this code with an inverse planning method to provide a complete Monte Carlo-based planning solution.

Fast GPU-based Monte Carlo simulations for LDR prostate brachytherapy

NASA Astrophysics Data System (ADS)

Bonenfant, Éric; Magnoux, Vincent; Hissoiny, Sami; Ozell, Benoît; Beaulieu, Luc; Després, Philippe

2015-07-01

The aim of this study was to evaluate the potential of bGPUMCD, a Monte Carlo algorithm executed on Graphics Processing Units (GPUs), for fast dose calculations in permanent prostate implant dosimetry. It also aimed to validate a low dose rate brachytherapy source in terms of TG-43 metrics and to use this source to compute dose distributions for permanent prostate implant in very short times. The physics of bGPUMCD was reviewed and extended to include Rayleigh scattering and fluorescence from photoelectric interactions for all materials involved. The radial and anisotropy functions were obtained for the Nucletron SelectSeed in TG-43 conditions. These functions were compared to those found in the MD Anderson Imaging and Radiation Oncology Core brachytherapy source registry which are considered the TG-43 reference values. After appropriate calibration of the source, permanent prostate implant dose distributions were calculated for four patients and compared to an already validated Geant4 algorithm. The radial function calculated from bGPUMCD showed excellent agreement (differences within 1.3%) with TG-43 accepted values. The anisotropy functions at r = 1 cm and r = 4 cm were within 2% of TG-43 values for angles over 17.5°. For permanent prostate implants, Monte Carlo-based dose distributions with a statistical uncertainty of 1% or less for the target volume were obtained in 30 s or less for 1 × 1 × 1 mm3 calculation grids. Dosimetric indices were very similar (within 2.7%) to those obtained with a validated, independent Monte Carlo code (Geant4) performing the calculations for the same cases in a much longer time (tens of minutes to more than a hour). bGPUMCD is a promising code that lets envision the use of Monte Carlo techniques in a clinical environment, with sub-minute execution times on a standard workstation. Future work will explore the use of this code with an inverse planning method to provide a complete Monte Carlo-based planning solution.

LLNL Mercury Project Trinity Open Science Final Report

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

Brantley, Patrick; Dawson, Shawn; McKinley, Scott

2016-04-20

The Mercury Monte Carlo particle transport code developed at Lawrence Livermore National Laboratory (LLNL) is used to simulate the transport of radiation through urban environments. These challenging calculations include complicated geometries and require significant computational resources to complete. As a result, a question arises as to the level of convergence of the calculations with Monte Carlo simulation particle count. In the Trinity Open Science calculations, one main focus was to investigate convergence of the relevant simulation quantities with Monte Carlo particle count to assess the current simulation methodology. Both for this application space but also of more general applicability, wemore » also investigated the impact of code algorithms on parallel scaling on the Trinity machine as well as the utilization of the Trinity DataWarp burst buffer technology in Mercury via the LLNL Scalable Checkpoint/Restart (SCR) library.« less

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

PubMed

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

2016-10-01

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

A proposed benchmark problem for cargo nuclear threat monitoring

NASA Astrophysics Data System (ADS)

Wesley Holmes, Thomas; Calderon, Adan; Peeples, Cody R.; Gardner, Robin P.

2011-10-01

There is currently a great deal of technical and political effort focused on reducing the risk of potential attacks on the United States involving radiological dispersal devices or nuclear weapons. This paper proposes a benchmark problem for gamma-ray and X-ray cargo monitoring with results calculated using MCNP5, v1.51. The primary goal is to provide a benchmark problem that will allow researchers in this area to evaluate Monte Carlo models for both speed and accuracy in both forward and inverse calculational codes and approaches for nuclear security applications. A previous benchmark problem was developed by one of the authors (RPG) for two similar oil well logging problems (Gardner and Verghese, 1991, [1]). One of those benchmarks has recently been used by at least two researchers in the nuclear threat area to evaluate the speed and accuracy of Monte Carlo codes combined with variance reduction techniques. This apparent need has prompted us to design this benchmark problem specifically for the nuclear threat researcher. This benchmark consists of conceptual design and preliminary calculational results using gamma-ray interactions on a system containing three thicknesses of three different shielding materials. A point source is placed inside the three materials lead, aluminum, and plywood. The first two materials are in right circular cylindrical form while the third is a cube. The entire system rests on a sufficiently thick lead base so as to reduce undesired scattering events. The configuration was arranged in such a manner that as gamma-ray moves from the source outward it first passes through the lead circular cylinder, then the aluminum circular cylinder, and finally the wooden cube before reaching the detector. A 2 in.×4 in.×16 in. box style NaI (Tl) detector was placed 1 m from the point source located in the center with the 4 in.×16 in. side facing the system. The two sources used in the benchmark are 137Cs and 235U.

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

PubMed

Gharehaghaji, Nahideh; Dadgar, Habib Alah

2018-01-01

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

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.

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

Portable LQCD Monte Carlo code using OpenACC

NASA Astrophysics Data System (ADS)

Bonati, Claudio; Calore, Enrico; Coscetti, Simone; D'Elia, Massimo; Mesiti, Michele; Negro, Francesco; Fabio Schifano, Sebastiano; Silvi, Giorgio; Tripiccione, Raffaele

2018-03-01

Varying from multi-core CPU processors to many-core GPUs, the present scenario of HPC architectures is extremely heterogeneous. In this context, code portability is increasingly important for easy maintainability of applications; this is relevant in scientific computing where code changes are numerous and frequent. In this talk we present the design and optimization of a state-of-the-art production level LQCD Monte Carlo application, using the OpenACC directives model. OpenACC aims to abstract parallel programming to a descriptive level, where programmers do not need to specify the mapping of the code on the target machine. We describe the OpenACC implementation and show that the same code is able to target different architectures, including state-of-the-art CPUs and GPUs.

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.

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

Sweezy, Jeremy Ed

A photon next-event fluence estimator at a point has been implemented in the Monte Carlo Application Toolkit (MCATK). The next-event estimator provides an expected value estimator for the flux at a point due to all source and collision events. An advantage of the next-event estimator over track-length estimators, which are normally employed in MCATK, is that flux estimates can be made in locations that have no random walk particle tracks. The next-event estimator allows users to calculate radiographs and estimate response for detectors outside of the modeled geometry. The next-event estimator is not yet accessable through the MCATK FlatAPI formore » C and Fortran. The next-event estimator in MCATK has been tested against MCNP6 using 5 suites of test problems. No issues were found in the MCATK implementation. One issue was found in the exclusion radius approximation in MCNP6. The theory, implementation, and testing are described in this document.« less

MCNP study for epithermal neutron irradiation of an isolated liver at the Finnish BNCT facility.

PubMed

Kotiluoto, P; Auterinen, I

2004-11-01

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.

Fixed forced detection for fast SPECT Monte-Carlo simulation

NASA Astrophysics Data System (ADS)

Cajgfinger, T.; Rit, S.; Létang, J. M.; Halty, A.; Sarrut, D.

2018-03-01

Monte-Carlo simulations of SPECT images are notoriously slow to converge due to the large ratio between the number of photons emitted and detected in the collimator. This work proposes a method to accelerate the simulations based on fixed forced detection (FFD) combined with an analytical response of the detector. FFD is based on a Monte-Carlo simulation but forces the detection of a photon in each detector pixel weighted by the probability of emission (or scattering) and transmission to this pixel. The method was evaluated with numerical phantoms and on patient images. We obtained differences with analog Monte Carlo lower than the statistical uncertainty. The overall computing time gain can reach up to five orders of magnitude. Source code and examples are available in the Gate V8.0 release.

Fixed forced detection for fast SPECT Monte-Carlo simulation.

PubMed

Cajgfinger, T; Rit, S; Létang, J M; Halty, A; Sarrut, D

2018-03-02

Monte-Carlo simulations of SPECT images are notoriously slow to converge due to the large ratio between the number of photons emitted and detected in the collimator. This work proposes a method to accelerate the simulations based on fixed forced detection (FFD) combined with an analytical response of the detector. FFD is based on a Monte-Carlo simulation but forces the detection of a photon in each detector pixel weighted by the probability of emission (or scattering) and transmission to this pixel. The method was evaluated with numerical phantoms and on patient images. We obtained differences with analog Monte Carlo lower than the statistical uncertainty. The overall computing time gain can reach up to five orders of magnitude. Source code and examples are available in the Gate V8.0 release.

MC3: Multi-core Markov-chain Monte Carlo code

NASA Astrophysics Data System (ADS)

Cubillos, Patricio; Harrington, Joseph; Lust, Nate; Foster, AJ; Stemm, Madison; Loredo, Tom; Stevenson, Kevin; Campo, Chris; Hardin, Matt; Hardy, Ryan

2016-10-01

MC3 (Multi-core Markov-chain Monte Carlo) is a Bayesian statistics tool that can be executed from the shell prompt or interactively through the Python interpreter with single- or multiple-CPU parallel computing. It offers Markov-chain Monte Carlo (MCMC) posterior-distribution sampling for several algorithms, Levenberg-Marquardt least-squares optimization, and uniform non-informative, Jeffreys non-informative, or Gaussian-informative priors. MC3 can share the same value among multiple parameters and fix the value of parameters to constant values, and offers Gelman-Rubin convergence testing and correlated-noise estimation with time-averaging or wavelet-based likelihood estimation methods.

Monte Carlo Simulation of Nonlinear Radiation Induced Plasmas. Ph.D. Thesis

NASA Technical Reports Server (NTRS)

Wang, B. S.

1972-01-01

A Monte Carlo simulation model for radiation induced plasmas with nonlinear properties due to recombination was, employing a piecewise linearized predict-correct iterative technique. Several important variance reduction techniques were developed and incorporated into the model, including an antithetic variates technique. This approach is especially efficient for plasma systems with inhomogeneous media, multidimensions, and irregular boundaries. The Monte Carlo code developed has been applied to the determination of the electron energy distribution function and related parameters for a noble gas plasma created by alpha-particle irradiation. The characteristics of the radiation induced plasma involved are given.

Monte Carlo charged-particle tracking and energy deposition on a Lagrangian mesh.

PubMed

Yuan, J; Moses, G A; McKenty, P W

2005-10-01

A Monte Carlo algorithm for alpha particle tracking and energy deposition on a cylindrical computational mesh in a Lagrangian hydrodynamics code used for inertial confinement fusion (ICF) simulations is presented. The straight line approximation is used to follow propagation of "Monte Carlo particles" which represent collections of alpha particles generated from thermonuclear deuterium-tritium (DT) reactions. Energy deposition in the plasma is modeled by the continuous slowing down approximation. The scheme addresses various aspects arising in the coupling of Monte Carlo tracking with Lagrangian hydrodynamics; such as non-orthogonal severely distorted mesh cells, particle relocation on the moving mesh and particle relocation after rezoning. A comparison with the flux-limited multi-group diffusion transport method is presented for a polar direct drive target design for the National Ignition Facility. Simulations show the Monte Carlo transport method predicts about earlier ignition than predicted by the diffusion method, and generates higher hot spot temperature. Nearly linear speed-up is achieved for multi-processor parallel simulations.

QMCPACK : an open source ab initio quantum Monte Carlo package for the electronic structure of atoms, molecules and solids

DOE PAGES

Kim, Jeongnim; Baczewski, Andrew T.; Beaudet, Todd D.; ...

2018-04-19

QMCPACK is an open source quantum Monte Carlo package for ab-initio electronic structure calculations. It supports calculations of metallic and insulating solids, molecules, atoms, and some model Hamiltonians. Implemented real space quantum Monte Carlo algorithms include variational, diffusion, and reptation Monte Carlo. QMCPACK uses Slater-Jastrow type trial wave functions in conjunction with a sophisticated optimizer capable of optimizing tens of thousands of parameters. The orbital space auxiliary field quantum Monte Carlo method is also implemented, enabling cross validation between different highly accurate methods. The code is specifically optimized for calculations with large numbers of electrons on the latest high performancemore » computing architectures, including multicore central processing unit (CPU) and graphical processing unit (GPU) systems. We detail the program’s capabilities, outline its structure, and give examples of its use in current research calculations. The package is available at http://www.qmcpack.org.« less

QMCPACK : an open source ab initio quantum Monte Carlo package for the electronic structure of atoms, molecules and solids

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

Kim, Jeongnim; Baczewski, Andrew T.; Beaudet, Todd D.

QMCPACK is an open source quantum Monte Carlo package for ab-initio electronic structure calculations. It supports calculations of metallic and insulating solids, molecules, atoms, and some model Hamiltonians. Implemented real space quantum Monte Carlo algorithms include variational, diffusion, and reptation Monte Carlo. QMCPACK uses Slater-Jastrow type trial wave functions in conjunction with a sophisticated optimizer capable of optimizing tens of thousands of parameters. The orbital space auxiliary field quantum Monte Carlo method is also implemented, enabling cross validation between different highly accurate methods. The code is specifically optimized for calculations with large numbers of electrons on the latest high performancemore » computing architectures, including multicore central processing unit (CPU) and graphical processing unit (GPU) systems. We detail the program’s capabilities, outline its structure, and give examples of its use in current research calculations. The package is available at http://www.qmcpack.org.« less

Monte Carlo simulation of radiation transport and dose deposition from locally released gold nanoparticles labeled with 111In, 177Lu or 90Y incorporated into tissue implantable depots

NASA Astrophysics Data System (ADS)

Lai, Priscilla; Cai, Zhongli; Pignol, Jean-Philippe; Lechtman, Eli; Mashouf, Shahram; Lu, Yijie; Winnik, Mitchell A.; Jaffray, David A.; Reilly, Raymond M.

2017-11-01

Permanent seed implantation (PSI) brachytherapy is a highly conformal form of radiation therapy but is challenged with dose inhomogeneity due to its utilization of low energy radiation sources. Gold nanoparticles (AuNP) conjugated with electron emitting radionuclides have recently been developed as a novel form of brachytherapy and can aid in homogenizing dose through physical distribution of radiolabeled AuNP when injected intratumorally (IT) in suspension. However, the distribution is unpredictable and precise placement of many injections would be difficult. Previously, we reported the design of a nanoparticle depot (NPD) that can be implanted using PSI techniques and which facilitates controlled release of AuNP. We report here the 3D dose distribution resulting from a NPD incorporating AuNP labeled with electron emitters (90Y, 177Lu, 111In) of different energies using Monte Carlo based voxel level dosimetry. The MCNP5 Monte Carlo radiation transport code was used to assess differences in dose distribution from simulated NPD and conventional brachytherapy sources, positioned in breast tissue simulating material. We further compare these dose distributions in mice bearing subcutaneous human breast cancer xenografts implanted with 177Lu-AuNP NPD, or injected IT with 177Lu-AuNP in suspension. The radioactivity distributions were derived from registered SPECT/CT images and time-dependent dose was estimated. Results demonstrated that the dose distribution from NPD reduced the maximum dose 3-fold when compared to conventional seeds. For simulated NPD, as well as NPD implanted in vivo, 90Y delivered the most homogeneous dose distribution. The tumor radioactivity in mice IT injected with 177Lu-AuNP redistributed while radioactivity in the NPD remained confined to the implant site. The dose distribution from radiolabeled AuNP NPD were predictable and concentric in contrast to IT injected radiolabeled AuNP, which provided irregular and temporally variant dose distributions. The use of NPD may serve as an intermediate between PSI and radiation delivered by radiolabeled AuNP by providing a controlled method to improve delivery of prescribed doses as well as homogenize dose from low penetrating electron sources.

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

NASA Astrophysics Data System (ADS)

Jaboulay, Jean-Charles; Brun, Emeric; Hugot, François-Xavier; Huynh, Tan-Dat; Malouch, Fadhel; Mancusi, Davide; Tsilanizara, Aime

2017-09-01

After fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant.

Monte Carlo simulation of liver cancer treatment with 166Ho-loaded glass microspheres

NASA Astrophysics Data System (ADS)

da Costa Guimarães, Carla; Moralles, Maurício; Roberto Martinelli, José

2014-02-01

Microspheres loaded with pure beta-emitter radioisotopes are used in the treatment of some types of liver cancer. The Instituto de Pesquisas Energéticas e Nucleares (IPEN) is developing 166Ho-loaded glass microspheres as an alternative to the commercially available 90Y microspheres. This work describes the implementation of a Monte Carlo code to simulate both the irradiation effects and the imaging of 166Ho and 90Y sources localized in different parts of the liver. Results obtained with the code and perspectives for the future are discussed.

Skyshine radiation from a pressurized water reactor containment dome

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

Peng, W.H.

1986-06-01

The radiation dose rates resulting from airborne activities inside a postaccident pressurized water reactor containment are calculated by a discrete ordinates/Monte Carlo combined method. The calculated total dose rates and the skyshine component are presented as a function of distance from the containment at three different elevations for various gamma-ray source energies. The one-dimensional (ANISN code) is used to approximate the skyshine dose rates from the hemisphere dome, and the results are compared favorably to more rigorous results calculated by a three-dimensional Monte Carlo code.

A Monte Carlo and experimental investigation of the dosimetric behavior of low‐ and medium‐perturbation diodes used for entrance in vivo dosimetry in megavoltage photon beams

PubMed Central

Mosleh‐Shirazi, Mohammad Amin; Shahbazi‐Gahrouei, Daryoush; Monadi, Shahram

2012-01-01

Full buildup diodes can cause significant dose perturbation if they are used on most or all of radiotherapy fractions. Given the importance of frequent in vivo measurements in complex treatments, using thin buildup (low‐perturbation) diodes instead is gathering interest. However, such diodes are strictly unsuitable for high‐energy photons; therefore, their use requires evaluation and careful measurement of correction factors (CFs). There is little published data on such factors for low‐perturbation diodes, and none on diode characterization for 9 MV X‐rays. We report on MCNP4c Monte Carlo models of low‐perturbation (EDD5) and medium‐perturbation (EDP10) diodes, and a comparison of source‐to‐surface distance, field size, temperature, and orientation CFs for cobalt‐60 and 9 MV beams. Most of the simulation results were within 4% of the measurements. The results suggest against the use of the EDD5 in axial angles beyond ±50° and exceeding the range 0° to +50° tilt angle at 9 MV. Outside these ranges, although the EDD5 can be used for accurate in vivo dosimetry at 9 MV, its CF variations were found to be 1.5–7.1 times larger than the EDP10 and, therefore, should be applied carefully. Finally, the MCNP diode models are sufficiently reliable tools for independent verification of potentially inaccurate measurements. PACS numbers: 87.10.Rt; 87.50.cm; 87.55.km; 87.56.Fc PMID:23149783

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

MCNP-based computational model for the Leksell gamma knife.

PubMed

Trnka, Jiri; Novotny, Josef; Kluson, Jaroslav

2007-01-01

We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large volumes such as for the total skull volume. The differences observed in treatment of scattered radiation between the MC method and the LGP may be important in this case. We have also studied the influence of differential direction sampling of primary photons and have found that, due to the anisotropic sampling, doses around the isocenter deviate from each other by up to 6%. With caution about the details of the calculation settings, it is possible to employ the MCNP Monte Carlo code for independent verification of the Leksell Gamma Knife radiation field properties.

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

Material Assessment for ITER's Collective Thomson Scattering first mirror

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

Santos, R.; Policarpo, H.; Goncalves, B.

2015-07-01

The International Thermonuclear Energy Reactor (ITER) Collective Thomson Scattering (CTS) system is a diagnostic instrument that measures plasma density and velocity through Thomson scattering of microwave radiation. Some of the key components of the CTS are quasi-optical mirrors that are used to produce astigmatic beam patterns, which have impact on the strength and spatial resolution of the diagnostic signal. The mirrors are exposed to neutron radiation, which may alter the quality of the signal received. In this work, three different materials (molybdenum (Mo), stainless steel 316 (SS-316) and tungsten (W)) are considered for the first mirror of the CTS. Themore » objective is to access which of the material studied are best suited for this mirror, considering different neutron radiation loads simulated scenarios defined by ITER, based on the resultant stresses and temperature distributions. For it, the neutron irradiation, and subsequent heat-load on the mirrors are simulated using the Monte Carlo N-Particle (MCNP) code. Based on the MCNP heat-load results, transient thermal-structural Finite Element Analysis (FEA) of the mirror over a 400 s discharge, with and without cooling on the rear side, are conducted using in commercial FEA software ANSYS. Results show that of the tested materials Mo and W are the most suitable material for this application. Even though, this study does not yet consider the variation of the material properties with temperature, it presents a quick initial satisfactory assessment that may be considered in subsequent and more complex analysis. (authors)« less

Rapid Acute Dose Assessment Using MCNP6

NASA Astrophysics Data System (ADS)

Owens, Andrew Steven

Acute radiation doses due to physical contact with a high-activity radioactive source have proven to be an occupational hazard. Multiple radiation injuries have been reported due to manipulating a radioactive source with bare hands or by placing a radioactive source inside a shirt or pants pocket. An effort to reconstruct the radiation dose must be performed to properly assess and medically manage the potential biological effects from such doses. Using the reference computational phantoms defined by the International Commission on Radiological Protection (ICRP) and the Monte Carlo N-Particle transport code (MCNP6), dose rate coefficients are calculated to assess doses for common acute doses due to beta and photon radiation sources. The research investigates doses due to having a radioactive source in either a breast pocket or pants back pocket. The dose rate coefficients are calculated for discrete energies and can be used to interpolate for any given energy of photon or beta emission. The dose rate coefficients allow for quick calculation of whole-body dose, organ dose, and/or skin dose if the source, activity, and time of exposure are known. Doses are calculated with the dose rate coefficients and compared to results from the International Atomic Energy Agency (IAEA) reports from accidents that occurred in Gilan, Iran and Yanango, Peru. Skin and organ doses calculated with the dose rate coefficients appear to agree, but there is a large discrepancy when comparing whole-body doses assessed using biodosimetry and whole-body doses assessed using the dose rate coefficients.

Status of the Monte Carlo library least-squares (MCLLS) approach for non-linear radiation analyzer problems

NASA Astrophysics Data System (ADS)

Gardner, Robin P.; Xu, Libai

2009-10-01

The Center for Engineering Applications of Radioisotopes (CEAR) has been working for over a decade on the Monte Carlo library least-squares (MCLLS) approach for treating non-linear radiation analyzer problems including: (1) prompt gamma-ray neutron activation analysis (PGNAA) for bulk analysis, (2) energy-dispersive X-ray fluorescence (EDXRF) analyzers, and (3) carbon/oxygen tool analysis in oil well logging. This approach essentially consists of using Monte Carlo simulation to generate the libraries of all the elements to be analyzed plus any other required background libraries. These libraries are then used in the linear library least-squares (LLS) approach with unknown sample spectra to analyze for all elements in the sample. Iterations of this are used until the LLS values agree with the composition used to generate the libraries. The current status of the methods (and topics) necessary to implement the MCLLS approach is reported. This includes: (1) the Monte Carlo codes such as CEARXRF, CEARCPG, and CEARCO for forward generation of the necessary elemental library spectra for the LLS calculation for X-ray fluorescence, neutron capture prompt gamma-ray analyzers, and carbon/oxygen tools; (2) the correction of spectral pulse pile-up (PPU) distortion by Monte Carlo simulation with the code CEARIPPU; (3) generation of detector response functions (DRF) for detectors with linear and non-linear responses for Monte Carlo simulation of pulse-height spectra; and (4) the use of the differential operator (DO) technique to make the necessary iterations for non-linear responses practical. In addition to commonly analyzed single spectra, coincidence spectra or even two-dimensional (2-D) coincidence spectra can also be used in the MCLLS approach and may provide more accurate results.

Use of single scatter electron monte carlo transport for medical radiation sciences

DOEpatents

Svatos, Michelle M.

2001-01-01

The single scatter Monte Carlo code CREEP models precise microscopic interactions of electrons with matter to enhance physical understanding of radiation sciences. It is designed to simulate electrons in any medium, including materials important for biological studies. It simulates each interaction individually by sampling from a library which contains accurate information over a broad range of energies.

Stochastic Analysis of Orbital Lifetimes of Spacecraft

NASA Technical Reports Server (NTRS)

Sasamoto, Washito; Goodliff, Kandyce; Cornelius, David

2008-01-01

A document discusses (1) a Monte-Carlo-based methodology for probabilistic prediction and analysis of orbital lifetimes of spacecraft and (2) Orbital Lifetime Monte Carlo (OLMC)--a Fortran computer program, consisting of a previously developed long-term orbit-propagator integrated with a Monte Carlo engine. OLMC enables modeling of variances of key physical parameters that affect orbital lifetimes through the use of probability distributions. These parameters include altitude, speed, and flight-path angle at insertion into orbit; solar flux; and launch delays. The products of OLMC are predicted lifetimes (durations above specified minimum altitudes) for the number of user-specified cases. Histograms generated from such predictions can be used to determine the probabilities that spacecraft will satisfy lifetime requirements. The document discusses uncertainties that affect modeling of orbital lifetimes. Issues of repeatability, smoothness of distributions, and code run time are considered for the purpose of establishing values of code-specific parameters and number of Monte Carlo runs. Results from test cases are interpreted as demonstrating that solar-flux predictions are primary sources of variations in predicted lifetimes. Therefore, it is concluded, multiple sets of predictions should be utilized to fully characterize the lifetime range of a spacecraft.

Accelerating Monte Carlo simulations of photon transport in a voxelized geometry using a massively parallel graphics processing unit

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

Badal, Andreu; Badano, Aldo

Purpose: It is a known fact that Monte Carlo simulations of radiation transport are computationally intensive and may require long computing times. The authors introduce a new paradigm for the acceleration of Monte Carlo simulations: The use of a graphics processing unit (GPU) as the main computing device instead of a central processing unit (CPU). Methods: A GPU-based Monte Carlo code that simulates photon transport in a voxelized geometry with the accurate physics models from PENELOPE has been developed using the CUDA programming model (NVIDIA Corporation, Santa Clara, CA). Results: An outline of the new code and a sample x-raymore » imaging simulation with an anthropomorphic phantom are presented. A remarkable 27-fold speed up factor was obtained using a GPU compared to a single core CPU. Conclusions: The reported results show that GPUs are currently a good alternative to CPUs for the simulation of radiation transport. Since the performance of GPUs is currently increasing at a faster pace than that of CPUs, the advantages of GPU-based software are likely to be more pronounced in the future.« less

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

A method for radiological characterization based on fluence conversion coefficients

NASA Astrophysics Data System (ADS)

Froeschl, Robert

2018-06-01

Radiological characterization of components in accelerator environments is often required to ensure adequate radiation protection during maintenance, transport and handling as well as for the selection of the proper disposal pathway. The relevant quantities are typical the weighted sums of specific activities with radionuclide-specific weighting coefficients. Traditional methods based on Monte Carlo simulations are radionuclide creation-event based or the particle fluences in the regions of interest are scored and then off-line weighted with radionuclide production cross sections. The presented method bases the radiological characterization on a set of fluence conversion coefficients. For a given irradiation profile and cool-down time, radionuclide production cross-sections, material composition and radionuclide-specific weighting coefficients, a set of particle type and energy dependent fluence conversion coefficients is computed. These fluence conversion coefficients can then be used in a Monte Carlo transport code to perform on-line weighting to directly obtain the desired radiological characterization, either by using built-in multiplier features such as in the PHITS code or by writing a dedicated user routine such as for the FLUKA code. The presented method has been validated against the standard event-based methods directly available in Monte Carlo transport codes.

Monte Carlo simulation of β γ coincidence system using plastic scintillators in 4π geometry

NASA Astrophysics Data System (ADS)

Dias, M. S.; Piuvezam-Filho, H.; Baccarelli, A. M.; Takeda, M. N.; Koskinas, M. F.

2007-09-01

A modified version of a Monte Carlo code called Esquema, developed at the Nuclear Metrology Laboratory in IPEN, São Paulo, Brazil, has been applied for simulating a 4 πβ(PS)-γ coincidence system designed for primary radionuclide standardisation. This system consists of a plastic scintillator in 4 π geometry, for alpha or electron detection, coupled to a NaI(Tl) counter for gamma-ray detection. The response curves for monoenergetic electrons and photons have been calculated previously by Penelope code and applied as input data to code Esquema. The latter code simulates all the disintegration processes, from the precursor nucleus to the ground state of the daughter radionuclide. As a result, the curve between the observed disintegration rate as a function of the beta efficiency parameter can be simulated. A least-squares fit between the experimental activity values and the Monte Carlo calculation provided the actual radioactive source activity, without need of conventional extrapolation procedures. Application of this methodology to 60Co and 133Ba radioactive sources is presented and showed results in good agreement with a conventional proportional counter 4 πβ(PC)-γ coincidence system.

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

Lin, H; Gao, Y; Liu, T

Purpose: To develop quantitative clinical guidelines between supine Deep Inspiratory Breath Hold (DIBH) and prone free breathing treatments for breast patients, we applied 3D deformable phantoms to perform Monte Carlo simulation to predict corresponding Dose to the Organs at Risk (OARs). Methods: The RPI-adult female phantom (two selected cup sizes: A and D) was used to represent the female patient, and it was simulated using the MCNP6 Monte Carlo code. Doses to OARs were investigated for supine DIBH and prone treatments, considering two breast sizes. The fluence maps of the 6-MV opposed tangential fields were exported. In the Monte Carlomore » simulation, the fluence maps allow each simulated photon particle to be weighed in the final dose calculation. The relative error of all dose calculations was kept below 5% by simulating 3*10{sup 7} photons for each projection. Results: In terms of dosimetric accuracy, the RPI Adult Female phantom with cup size D in DIBH positioning matched with a DIBH treatment plan of the patient. Based on the simulation results, for cup size D phantom, prone positioning reduced the cardiac dose and the dose to other OARs, while cup size A phantom benefits more from DIBH positioning. Comparing simulation results for cup size A and D phantom, dose to OARs was generally higher for the large breast size due to increased scattering arising from a larger portion of the body in the primary beam. The lower dose that was registered for the heart in the large breast phantom in prone positioning was due to the increase of the distance between the heart and the primary beam when the breast was pendulous. Conclusion: Our 3D deformable phantom appears an excellent tool to predict dose to the OARs for the supine DIBH and prone positions, which might help quantitative clinical decisions. Further investigation will be conducted. National Institutes of Health R01EB015478.« less

Some Notes on Neutron Up-Scattering and the Doppler-Broadening of High-Z Scattering Resonances

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

Parsons, Donald Kent

When neutrons are scattered by target nuclei at elevated temperatures, it is entirely possible that the neutron will actually gain energy (i.e., up-scatter) from the interaction. This phenomenon is in addition to the more usual case of the neutron losing energy (i.e., down-scatter). Furthermore, the motion of the target nuclei can also cause extended neutron down-scattering, i.e., the neutrons can and do scatter to energies lower than predicted by the simple asymptotic models. In recent years, more attention has been given to temperature-dependent scattering cross sections for materials in neutron multiplying systems. This has led to the inclusion of neutronmore » up-scatter in deterministic codes like Partisn and to free gas scattering models for material temperature effects in Monte Carlo codes like MCNP and cross section processing codes like NJOY. The free gas scattering models have the effect of Doppler Broadening the scattering cross section output spectra in energy and angle. The current state of Doppler-Broadening numerical techniques used at Los Alamos for scattering resonances will be reviewed, and suggestions will be made for further developments. The focus will be on the free gas scattering models currently in use and the development of new models to include high-Z resonance scattering effects. These models change the neutron up-scattering behavior.« less