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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

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

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

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

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

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 newest 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.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

SU-E-T-171: Evaluation of the Analytical Anisotropic Algorithm in a Small Finger Joint Phantom Using Monte Carlo Simulation

Purpose: This study investigated the performance of the anisotropic analytical algorithm (AAA) in dose calculation in radiotherapy concerning a small finger joint. Monte Carlo simulation (EGSnrc code) was used in this dosimetric evaluation. Methods: Heterogeneous finger joint phantom containing a vertical water layer (bone joint or cartilage) sandwiched by two bones with dimension 2 × 2 × 2 cm{sup 3} was irradiated by the 6 MV photon beams (field size = 4 × 4 cm{sup 2}). The central beam axis was along the length of the bone joint and the isocenter was set to the center of the joint. The joint width and beam angle were varied from 0.5–2 mm and 0°–15°, respectively. Depth doses were calculated using the AAA and DOSXYZnrc. For dosimetric comparison and normalization, dose calculations were repeated in water phantom using the same beam geometry. Results: Our AAA and Monte Carlo results showed that the AAA underestimated the joint doses by 10%–20%, and could not predict joint dose variation with changes of joint width and beam angle. The calculated bone dose enhancement for the AAA was lower than Monte Carlo and the depth of maximum dose for the phantom was smaller than that for the water phantom. From Monte Carlo results, there was a decrease of joint dose as its width increased. This reflected the smaller the joint width, the more the bone scatter contributed to the depth dose. Moreover, the joint dose was found slightly decreased with an increase of beam angle. Conclusion: The AAA could not handle variations of joint dose well with changes of joint width and beam angle based on our finger joint phantom. Monte Carlo results showed that the joint dose decreased with increase of joint width and beam angle. This dosimetry comparison should be useful to radiation staff in radiotherapy related to small bone joint.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Fringe biasing: A variance reduction technique for optically thick meshes

Fringe biasing is a stratified sampling scheme applicable to Monte Carlo thermal radiation transport codes. The thermal emission source in optically thick cells is partitioned into separate contributions from the cell interiors (where the likelihood of the particles escaping the cells is virtually zero) and the 'fringe' regions close to the cell boundaries. Thermal emission in the cell interiors can now be modelled with fewer particles, the remaining particles being concentrated in the fringes so that they are more likely to contribute to the energy exchange between cells. Unlike other techniques for improving the efficiency in optically thick regions (such as random walk and discrete diffusion treatments), fringe biasing has the benefit of simplicity, as the associated changes are restricted to the sourcing routines with the particle tracking routines being unaffected. This paper presents an analysis of the potential for variance reduction achieved from employing the fringe biasing technique. The aim of this analysis is to guide the implementation of this technique in Monte Carlo thermal radiation codes, specifically in order to aid the choice of the fringe width and the proportion of particles allocated to the fringe (which are interrelated) in multi-dimensional simulations, and to confirm that the significant levels of variance reduction achieved in simulations can be understood by studying the behaviour for simple test cases. The variance reduction properties are studied for a single cell in a slab geometry purely absorbing medium, investigating the accuracy of the scalar flux and current tallies on one of the interfaces with the surrounding medium. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

Monte Carlo Hauser-Feshbach Calculations of Prompt Fission Neutrons and Gamma Rays: Application to Thermal Neutron-Induced Fission Reactions on U-235 and Pu-239

We have developed a new code CGMF , implementing a Monte Carlo version of the Hauser-Feshbach statistical theory of nuclear reactions and applied to the de-excitation of fission fragments. A parallel version of the code has been developed as well, and was used to produced the results presented in this report. Prompt fission neutron and gamma-ray characteristics have been computed for the thermal neutron-induced fission reactions on 235 U and 239 Pu. We discuss the status of these calculations and how the results compare to available experimental data.

Atomic and Nuclear Physics↗
DOE OSTI2013

Internal neutronics-temperature coupling in Serpent 2 - Reactivity differences resulting from choice of material property correlations

This paper describes the unique way of simultaneously solving the power and temperature distributions of a nuclear system with the Monte Carlo neutron transport code Serpent 2. The coupled solution is achieved through the implementation of an internal temperature solver and material property correlations in the code. The program structure is reviewed concerning the temperature solver and the internal correlations as well as the internal coupling between these two and the neutron transport part. To estimate the reactivity differences resulting from correlation choices a simple pin-cell case has been calculated. It is established, that some correlation choices may result in difference in reactivity of approximately 100 pcm. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

In-situ ion irradiation of ZIRLO alloy

As an alternative to in-core neutron irradiations which are lengthy and costly, ex-core irradiations with energetic ions are envisioned as a possible way to probe the irradiation response of zirconium alloys used in pressurized water reactors (PWRs). Such ion irradiations, being well-controlled experiments, could offer a powerful, easier, and quicker way to study irradiation induced microstructural changes without remote handling that typically lead to structural component dimensional change (such as irradiation growth). Thus these irradiation experiments could help in the development of advanced dimensionally stable and deformation resistant alloys. For this matter, an initial investigation was performed to determine the feasibility of using ex-core ions to develop irradiation damage typically observed after neutron irradiation in Westinghouse's ZIRLO alloy. In-situ ion irradiations of ZIRLO alloy were conducted at the IVEM-Tandem Facility at Argonne National Laboratory at different dose levels and at different temperatures. The initial goal is to induce radiation damage of similar levels (in terms of dpa) to what develops during in-core neutron irradiation and analyze the potential of using ion irradiation as a substitute for in core neutron irradiation of Zr alloys. To facilitate comparison and convergence to in-core neutron irradiation, samples were ion irradiated in transmission and the amount of dpa developed during the ion irradiation was calculated using monte-Carlo based SRIM code. The TEM observations showed that the ion irradiation induced microstructure consisted of - and -type dislocation loops with the -loops developing at a later stage than the - loops, similarly to irradiation damage typically induced in neutron irradiated zirconium alloys. The study offers promises that heavy ion irradiation could potentially, under well-chosen conditions, be used to simulate neutron irradiation damage in PWRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

On the feasibility of polyurethane based 3D dosimeters with optical CT for dosimetric verification of low energy photon brachytherapy seeds

Purpose: To investigate the feasibility of and challenges yet to be addressed to measure dose from low energy (effective energy <50 keV) brachytherapy sources (Pd-103, Cs-131, and I-125) using polyurethane based 3D dosimeters with optical CT. Methods: The authors' evaluation used the following sources: models 200 (Pd-103), CS-1 Rev2 (Cs-131), and 6711 (I-125). The authors used the Monte Carlo radiation transport code MCNP5, simulations with the ScanSim optical tomography simulation software, and experimental measurements with PRESAGE{sup ®} dosimeters/optical CT to investigate the following: (1) the water equivalency of conventional (density = 1.065 g/cm{sup 3}) and deformable (density = 1.02 g/cm{sup 3}) formulations of polyurethane dosimeters, (2) the scatter conditions necessary to achieve accurate dosimetry for low energy photon seeds, (3) the change in photon energy spectrum within the dosimeter as a function of distance from the source in order to determine potential energy sensitivity effects, (4) the optimal delivered dose to balance optical transmission (per projection) with signal to noise ratio in the reconstructed dose distribution, and (5) the magnitude and characteristics of artifacts due to the presence of a channel in the dosimeter. Monte Carlo simulations were performed using both conventional and deformable dosimeter formulations. For verification, 2.8 Gy at 1 cm was delivered in 92 h using an I-125 source to a PRESAGE{sup ®} dosimeter with conventional formulation and a central channel with 0.0425 cm radius for source placement. The dose distribution was reconstructed with 0.02 and 0.04 cm{sup 3} voxel size using the Duke midsized optical CT scanner (DMOS). Results: While the conventional formulation overattenuates dose from all three sources compared to water, the current deformable formulation has nearly water equivalent attenuation properties for Cs-131 and I-125, while underattenuating for Pd-103. The energy spectrum of each source is relatively stable within the first 5 cm especially for I-125. The inherent assumption of radial symmetry in the TG43 geometry leads to a linear increase in sample points within the 3D dosimeter as a function of distance away from the source, which partially offsets the decreasing signal. Simulations of dose reconstruction using optical CT showed the feasibility of reconstructing dose out to a radius of 10 cm without saturating projection images using an optimal dose and high dynamic range scanning; the simulations also predicted that reconstruction artifacts at the channel surface due to a small discrepancy in refractive index should be negligible. Agreement of the measured with calculated radial dose function for I-125 was within 5% between 0.3 and 2.5 cm from the source, and the median difference of measured from calculated anisotropy function was within 5% between 0.3 and 2.0 cm from the source. Conclusions: 3D dosimetry using polyurethane dosimeters with optical CT looks to be a promising application to verify dosimetric distributions surrounding low energy brachytherapy sources.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Passive neutron coincidence counting with plastic scintillators for the characterization of radioactive waste drums

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microseconds are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS2015

Investigation of HZETRN 2010 as a Tool for Single Event Effect Qualification of Avionics Systems - Part II

An accurate prediction of spacecraft avionics single event effect (SEE) radiation susceptibility is key to ensuring a safe and reliable vehicle. This is particularly important for long-duration deep space missions for human exploration where there is little or no chance for a quick emergency return to Earth. Monte Carlo nuclear reaction and transport codes such as FLUKA can be used to generate very accurate models of the expected in-flight radiation environment for SEE analyses. A major downside to using a Monte Carlo-based code is that the run times can be very long (on the order of days). A more popular choice for SEE calculations is the CREME96 deterministic code, which offers significantly shorter run times (on the order of seconds). However, CREME96, though fast and easy to use, has not been updated in several years and underestimates secondary particle shower effects in spacecraft structural shielding mass. Another modeling option to consider is the deterministic code HZETRN 20104, which includes updates to address secondary particle shower effects more accurately. This paper builds on previous work by Rojdev, et al. to compare the use of HZETRN 2010 against CREME96 as a tool to verify spacecraft avionics system reliability in a space flight SEE environment. This paper will discuss modifications made to HZETRN 2010 to improve its performance for calculating SEE rates and compare results with both in-flight SEE rates and other calculation methods.

Rojdev, Kristina↗
DOE OSTI2013

Benchmark for evaluation and validation of reactor simulations (BEAVRS)

Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

The Application of MCNP6.2 Unstructured Mesh Geometry Capabilities to the Study of Heavy Ion Single Event Effect on Microelectronic Devices and the Reactor Physics Analysis - 25691

Unstructured meshes allow irregular or complex geometries to be modeled precisely in Monte Carlo radiation transport codes. Through this capability, it is possible to share an unstructured mesh model between multiple codes that use unstructured mesh geometry and thus improving coupled multi-physics simulations. In this work, unstructured meshes are used to model energy deposition from heavy ion strikes for single event effects on microelectronic devices as well as neutron flux mapping of nuclear reactor configurations. In each case, MCNP6.2's unstructured mesh functionality and applicability to the problem is demonstrated and analyzed. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

The FluxCompensator: Making Radiative Transfer Models of Hydrodynamical Simulations Directly Comparable to Real Observations

When modeling astronomical objects throughout the universe, it is important to correctly treat the limitations of the data, for instance finite resolution and sensitivity. In order to simulate these effects, and to make radiative transfer models directly comparable to real observations, we have developed an open-source Python package called the FluxCompensator that enables the post-processing of the output of 3D Monte Carlo radiative transfer codes, such as Hyperion. With the FluxCompensator, realistic synthetic observations can be generated by modeling the effects of convolution with arbitrary point-spread functions, transmission curves, finite pixel resolution, noise, and reddening. Pipelines can be applied to compute synthetic observations that simulate observatories, such as the Spitzer Space Telescope or the Herschel Space Observatory . Additionally, this tool can read in existing observations (e.g., FITS format) and use the same settings for the synthetic observations. In this paper, we describe the package as well as present examples of such synthetic observations.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2014

Herschel photometry of disks around low-mass stars in the R CrA cloud

We report photometric results from a subset of a Herschel-PACS program to observe cool dust in disks around low-mass stars as a complement to our earlier program to measure far-infrared emission from brown dwarfs. In this latest study we observed five low-mass objects in the nearby R Corona Australis region and detected at least three at 70 μm. Using a Monte Carlo radiative transfer code we have investigated the disk masses and geometry based on detailed spectral energy distribution (SED) modeling, and we compare these new results to those from our earlier larger sample of brown dwarfs. In particular, our SED analysis for these five objects shows again that disk geometries of brown dwarfs or low-mass stars are generally similar to their higher mass counterparts like T Tauri disks, but the range of disk mass extends to well below the value found in T Tauri stars.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Investigation of Imaging Spent Nuclear Fuel Dry Casks using Cosmic Ray Muons

Recently, the use of cosmic ray muons for cargo scanning applications has been demonstrated and muons have been shown to have the potential to allow for nondestructive assessment of nuclear material accountancy. Relativistic muons have the ability to penetrate dense materials and monitoring the subsequent scattering of muons can provide information about the structure and composition of materials stored in cargo containers. The suitability of cosmic muons for a number of monitoring and imaging applications has been investigated over the years and includes applications to archaeology, volcano imaging, material identification and medical diagnosis. Novel applications of cosmic ray muons have been proposed for examination of comatose patients towards bone density monitoring and also for determination of the molten nuclear fuel location in nuclear reactors having suffered from the effects of a severe accident similar to the one happened in Fukushima. There is a need to develop a capability to monitor nuclear waste and investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Monitoring nuclear waste and controlling nuclear material at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of spent nuclear fuel have been generated, 25% of which are under dry storage conditions. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. During the long storage life of the cask, intermediate handling and transportation activities take place which does create concerns regarding fuel integrity. Accidents and natural disasters also contribute to this concern. Previous work on spent nuclear fuel dry cask monitoring using muons has shown that the muon scattering distributions are adequately separated and a and a decision rule derived from Bayes' Theorem rendering the classification of different dry casks feasible. Muons present significant advantages over the existing imaging techniques such as the utilization of the passive nature of muons, the lack of radiological sources and consequently the absence of any artificial radiological dose. In the present work, a methodology is developed to demonstrate the applicability of muons for imaging of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amounts of loading. Monte Carlo simulations of dry casks with different spent nuclear fuel loading are performed to simulate the passage of polyenergetic muons having energies in the range of 1-60 GeV through matter. A 'Muon Generator' is developed to generate muons that follow the actual sea level energy distribution. The 'Muon Generator' is validated against a series of actual spectrum measurements and then coupled with the Monte Carlo particle transport code GEANT4. The accumulated dataset of scattering angles were processed with the Point-of- the-Closest-Approach (PoCA) imaging algorithm and the obtained images demonstrate that an empty dry cask can be clearly distinguished from a half-loaded or a fully loaded one. The apparent separation of the images demonstrates that the muon scattering can be used as a feature for imaging spent nuclear fuel dry casks. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

SU-E-T-81: Comparison of Microdosimetric Quantities Calculated Using the Track Structure Monte Carlo Algorithms Geant4-DNA and NOREC

Purpose: Microdosimetric quantities, such as the lineal energy, have been shown to correlate with the biological response to radiation and the relative biological effect of different radiation types. Track-structure Monte Carlo simulations are an important tool for investigating these responses and for developing mechanistic models to explain them. However, some of the cross-sectional data used in these algorithms has large uncertainties; thus, it is important to investigate how the implementation of the different codes affects the quantities of interest. Methods: Two of the most widely-used publicly available track-structure Monte Carlo codes, Geant4-DNA and NOREC, were used generate electron tracks for two particle sources. One source was a mono-energetic parallel beam of electrons with energies from 5 to 500-keV, and the lineal energy for each track was calculated in 1-mm-spheres arranged in planar arrays at multiple distances from the source. The second source was mono-energetic, uniformly-distributed, and isotropic source, and the lineal energy was scored in a single 30-mm-sphere for energies between 300-eV and 5-keV. Results: The dose-mean lineal energy for the parallel-beam simulations almost all agreed within 5%. For the uniformly-distributed source, at the lowest energies there was strong agreement between the algorithms, but the Geant4-DNA simulations showed slightly more high-energy events for more energetic electrons, but the dose-mean lineal energy agreed to within 4% for all energies. Conclusion: While there were slight differences in the results between the codes, these were consistent with previous studies of the stopping power and angular scattering distributions. Importantly, the computation time for Geant4-DNA was larger than for NOREC, largely due to approximations used in the NOREC for energies below 10-eV. This study shows that these approximation does not have a major impact on the microdosimetry on the energy and length scales investigated.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Neutron production in the beryllium filter of a linac using 6 MeV treatment beam - 25425

In most conventional radiation therapy treatments, special attention is paid for neutron contamination when working with energy beams above 8 MeV. Generally, neutron generation it is only considered for shielding requirements, not for dose study in patients or employees. The present work is focused on studying the unwanted generated photoneutrons in a Medical Linear Accelerator (Linac) Varian TrueBeam using a 6 MeV radiation treatment beam. To that, Monte Carlo (MC) simulation code MCNP6.1.1 was used. This new version of the code allows the use of unstructured mesh geometries models, offering more geometric reliable results. The particularity of the studied Linac is the presence of a beryllium filter at the treatment head. Since Beryllium causes photonuclear reactions (γ, n) at energies much lower than other Linac composing materials, this work aims to analyze if this type of unit, when using low energy treatment beams (6 MeV), produce significant neutron dose and to ensure that this unwanted radiation can be considered negligible. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Development of CSG-based radiation shielding module for ARCHER: preliminary results for photons - Paper 126

ARCHER is a parallel Monte Carlo radiation transport code being developed for various hardware platforms including multi-core CPUs, the emerging Nvidia graphics processing units (GPUs) and Intel Xeon Phi coprocessors. Previous studies have demonstrated the accuracy and speed for CT imaging and radiotherapy dosimetry applications. Radiation shielding applications, however, require ARCHER to be more flexible in defining and handling geometries, in addition to the voxel phantom geometries that are commonly used in medical physics calculations. In this paper, we present the development of a general geometry module used for radiation protection and shielding problems in ARCHER. The methods of geometry design and particle transport for the GPUs are introduced. Preliminary results comparing against the MCNP code for cases involving Cs-137 0.662 MeV gammas are reported. The radiation shielding design module of ARCHER has been developed, which supports general geometry setups in the CSG format. The code was implemented using CUDA on Nvidia GPUs, and the performance was benchmarked against the MCNP code for a simple radiation shielding problem. Results showed that ARCHER worked properly and achieved impressive speedup due to highly parallel computing nature of the Nvidia GPUs compared to the MCNP code running on multi-core CPUs. Further optimization of the geometry module will be carried out using advanced geometry structures and variance reduction techniques to improve the precision and performance of ARCHER. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗