COHORT-II - A Monte Carlo general purpose shielding computer code.
COHORT-II Monte Carlo computer program written in Fortran IV to model radiation shielding problems
Clemons, L. C., Jr.↗
SEARCH · Search NASA
Results for “Monte Carlo Code”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
The effects of mapping CT images to Monte Carlo materials on GEANT4 proton simulation accuracy
Purpose: Monte Carlo simulations of radiation therapy require conversion from Hounsfield units (HU) in CT images to an exact tissue composition and density. The number of discrete densities (or density bins) used in this mapping affects the simulation accuracy, execution time, and memory usage in GEANT4 and other Monte Carlo code. The relationship between the number of density bins and CT noise was examined in general for all simulations that use HU conversion to density. Additionally, the effect of this on simulation accuracy was examined for proton radiation. Methods: Relative uncertainty from CT noise was compared with uncertainty from density binning to determine an upper limit on the number of density bins required in the presence of CT noise. Error propagation analysis was also performed on continuously slowing down approximation range calculations to determine the proton range uncertainty caused by density binning. These results were verified with Monte Carlo simulations. Results: In the presence of even modest CT noise (5 HU or 0.5%) 450 density bins were found to only cause a 5% increase in the density uncertainty (i.e., 95% of density uncertainty from CT noise, 5% from binning). Larger numbers of density bins are not required as CT noise will prevent increased density accuracy; this applies across all types of Monte Carlo simulations. Examining uncertainty in proton range, only 127 density bins are required for a proton range error of <0.1 mm in most tissue and <0.5 mm in low density tissue (e.g., lung). Conclusions: By considering CT noise and actual range uncertainty, the number of required density bins can be restricted to a very modest 127 depending on the application. Reducing the number of density bins provides large memory and execution time savings in GEANT4 and other Monte Carlo packages.
60 APPLIED LIFE SCIENCES↗
Gravitational microlensing of gamma-ray bursts
A Monte Carlo code is developed to calculate gravitational microlensing in three dimensions when the lensing optical depth is low or moderate (not greater than 0.25). The code calculates positions of microimages and time delays between the microimages. The majority of lensed gamma-ray bursts should show a simple double-burst structure, as predicted by a single point mass lens model. A small fraction should show complicated multiple events due to the collective effects of several point masses (black holes). Cosmological models with a significant fraction of mass density in massive compact objects can be tested by searching for microlensing events in the current BATSE data. Our catalog generated by 10,000 Monte Carlo models is accessible through the computer network. The catalog can be used to take realistic selection effects into account.
Mao, Shude↗
Comparison of air-kerma strength determinations for HDR {sup 192}Ir sources
Purpose: To perform a comparison of the interim air-kerma strength standard for high dose rate (HDR) {sup 192}Ir brachytherapy sources maintained by University of Wisconsin Accredited Dosimetry Calibration Laboratory (UWADCL) with measurements of the various source models using modified techniques from the literature. The current interim standard was established by Goetsch et al. in 1991 and has remained unchanged to date. Methods: The improved, laser-aligned seven-distance apparatus of University of Wisconsin Medical Radiation Research Center (UWMRRC) was used to perform air-kerma strength measurements of five different HDR {sup 192}Ir source models. The results of these measurements were compared with those from well chambers traceable to the original standard. Alternative methodologies for interpolating the {sup 192}Ir air-kerma calibration coefficient from the NIST air-kerma standards at {sup 137}Cs and 250 kVp x rays (M250) were investigated and intercompared. As part of the interpolation method comparison, the Monte Carlo code EGSnrc was used to calculate updated values of A{sub wall} for the Exradin A3 chamber used for air-kerma strength measurements. The effects of air attenuation and scatter, room scatter, as well as the solution method were investigated in detail. Results: The average measurements when using the inverse N{sub K} interpolation method for the Classic Nucletron, Nucletron microSelectron, VariSource VS2000, GammaMed Plus, and Flexisource were found to be 0.47%, -0.10%, -1.13%, -0.20%, and 0.89% different than the existing standard, respectively. A further investigation of the differences observed between the sources was performed using MCNP5 Monte Carlo simulations of each source model inside a full model of an HDR 1000 Plus well chamber. Conclusions: Although the differences between the source models were found to be statistically significant, the equally weighted average difference between the seven-distance measurements and the well chambers was 0.01%, confirming that it is not necessary to update the current standard maintained at the UWADCL.
61 RADIATION PROTECTION AND DOSIMETRY↗
Preparation of nuclear libraries with deterministic and stochastic methods for LWR reflectors
The explicit reflector methodology is used in the system of codes CASMO-5 / SIMULATE-3 to include the reflector around the active core into the computational region and avoid adopting any ad-hoc or experimental albedo coefficients as boundary conditions. However, to complete the core calculation, a set of cross sections and discontinuity factors is needed for the reflector nodes and the accuracy of these nuclear parameters influences the final results, in particular along the peripheral regions of the core. In this paper the explicit reflector methodology of CASMO-5 is adopted to evaluate the few-group cross sections and discontinuity factors of the different reflector cases, based on the design of Generation II and III LWR reactors. In addition, in the perspective of using Monte-Carlo codes as a complementary option for lattice calculations of reflector configurations, the stochastic SERPENT code is also included as part of this benchmark. With the latter, the impact of applying 2-D reflector models with homogenized materials instead of explicit representation of the real geometrical structures is moreover evaluated and shown to be limited. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Watts Bar I ex-core analyses using VERA - 25430
The integration and direct coupling of the Oak Ridge National Laboratory (ORNL) Shift Monte Carlo code within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities that combine high-fidelity in-core and ex-core radiation transport. In this recent development activity, the deterministic neutronics code MPACT performs the in-core radiation transport with temperature feedback using COBRA-TF (CTF), and provides the fission source to Shift for a follow-on fixed source radiation transport calculation to determine ex-core quantities of interest. The coupling of MPACT to Shift allows the neutron source for the ex-core transport to be calculated from the time-dependent, fully coupled MPACT solution. This paper presents the first use of VERA to perform coupon irradiation for code-to-code comparison of its ex-core capabilities. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Dynamic Simulation with on the fly Doppler Broadening treatment in RMC
In conventional transient analysis, thermal hydraulics calculations are generally coupled to deterministic neutronics solvers. Due to the limited accuracy of deterministic methods and the rapid development of computer power, Monte Carlo neutronics calculations have received increasing attention nowadays. One of challenges for Monte Carlo methods coupled to thermal hydraulics calculations is the temperature dependent cross sections. Taking the detailed temperature distribution feedback into account, each nuclide may be in a large range of temperatures, pre-generated libraries at every temperature point are no longer applicable owing to unaffordable memory footprint. Different solutions have been proposed to deal with this problem. TMS, accounting for Doppler broadening effect by the explicit treatment of thermal motion, was used to model fast transients in Serpent 2. Stochastic mixing, using two sets of libraries for each nuclide and mixing them to get the correct temperature, was used in transient analysis by Sjenitzer and Hoogenboom. Besides, there are other on the fly Doppler broadening methods being developed, including series expansion of cross sections, multipole representation, different integration algorithms based on Doppler broadening equation. In this paper, on the fly Doppler broadening treatment based on integration algorithms was applied to transient calculation in Reactor Monte Carlo code RMC . In the following section, the integration algorithms implemented and dynamic simulation method in RMC is briefly introduced. Finally, a pure physical transient fuel assembly model was chosen for numerical validation. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Air-kerma strength determination of a new directional {sup 103}Pd source
Purpose: A new directional {sup 103}Pd planar source array called a CivaSheet™ has been developed by CivaTech Oncology, Inc., for potential use in low-dose-rate (LDR) brachytherapy treatments. The array consists of multiple individual polymer capsules called CivaDots, containing {sup 103}Pd and a gold shield that attenuates the radiation on one side, thus defining a hot and cold side. This novel source requires new methods to establish a source strength metric. The presence of gold material in such close proximity to the active {sup 103}Pd region causes the source spectrum to be significantly different than the energy spectra of seeds normally used in LDR brachytherapy treatments. In this investigation, the authors perform air-kerma strength (S{sub K}) measurements, develop new correction factors for these measurements based on an experimentally verified energy spectrum, and test the robustness of transferring S{sub K} to a well-type ionization chamber. Methods: S{sub K} measurements were performed with the variable-aperture free-air chamber (VAFAC) at the University of Wisconsin Medical Radiation Research Center. Subsequent measurements were then performed in a well-type ionization chamber. To realize the quantity S{sub K} from a directional source with gold material present, new methods and correction factors were considered. Updated correction factors were calculated using the MCNP 6 Monte Carlo code in order to determine S{sub K} with the presence of gold fluorescent energy lines. In addition to S{sub K} measurements, a low-energy high-purity germanium (HPGe) detector was used to experimentally verify the calculated spectrum, a sodium iodide (NaI) scintillating counter was used to verify the azimuthal and polar anisotropy, and a well-type ionization chamber was used to test the feasibility of disseminating S{sub K} values for a directional source within a cylindrically symmetric measurement volume. Results: The UW VAFAC was successfully used to measure the S{sub K} of four CivaDots with reproducibilities within 0.3%. Monte Carlo methods were used to calculate the UW VAFAC correction factors and the calculated spectrum emitted from a CivaDot was experimentally verified with HPGe detector measurements. The well-type ionization chamber showed minimal variation in response (<1.5%) as a function of source positioning angle, indicating that an American Association of Physicists in Medicine (AAPM) Accredited Dosimetry Calibration Laboratory calibrated well chamber would be a suitable device to transfer an S{sub K}-based calibration to a clinical user. S{sub K} per well-chamber ionization current ratios were consistent among the four dots measured. Additionally, the measurements and predictions of anisotropy show uniform emission within the solid angle of the VAFAC, which demonstrates the robustness of the S{sub K} measurement approach. Conclusions: This characterization of a new {sup 103}Pd directional brachytherapy source helps to establish calibration methods that could ultimately be used in the well-established AAPM Task Group 43 formalism. Monte Carlo methods accurately predict the changes in the energy spectrum caused by the fluorescent x-rays produced in the gold shield.
07 ISOTOPE AND RADIATION SOURCES↗
Multi-cycle reload analysis of a long cycle gas-cooled fast modular reactor
There is currently significant interest in deploying HALEU-fueled fast reactors, including the General Atomics (GA) Fast Modular Reactor (FMR). Such reactors can achieve very long fuel cycles, but with multi-batch loading will take decades to reach equilibrium. This motivates design and analysis of both the initial core and multi-cycle reload, which is typically performed using fast-running, deterministic fast reactor codes such as the Argonne Reactor Computation (ARC) codes. In this paper, multicycle reload of the GA FMR is analyzed using the ARC codes. The GA FMR utilizes 19.75 % enriched fuel in a 16 year cycle with a three-batch strategy, with twice-burned fuel placed on the core periphery. The GA FMR has a softened neutron spectrum due to reflecting elements in the core, so the neutronic solution is first benchmarked against the OpenMC Monte Carlo code. Discrepancy on k eff is 400–600 pcm, likely due to the softened neutron spectrum, heterogeneous fuel assembly design and central reflector. However, the rms discrepancy on the assembly power distribution is only 0.6 %, despite the presence of the central reflector. A reload strategy is devised for the first three cycles of such a reactor, ultimately spanning the first 45–48 years of its operation. The fresh core uses 19.75 %, 19.25 % and 16.75 % enriched fuel in place of fresh, once-burned and twice-burned and is then subsequently refueled with only 19.75 % enriched fuel. The cycle length is varied over 3 cycles of operation to balance fuel utilization and reactor availability, specifically with use of an extended 18-year Cycle 1, followed by a shortened 11-year Cycle 2. Cycle 3 is close to the target 16-year length. Finally, placing twice burned assemblies next to the GA FMR central reflector can reduce power peaking by 3 %, at the expense of slightly reducing the cycle length.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
SU-E-T-297: Dosimetric Assessment of An Air-Filled Balloon Applicator in HDR Vaginal Cuff Brachytherapy Using the Monte Carlo Method
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) 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.
60 APPLIED LIFE SCIENCES↗
Evaluation of the adequacy of using few-group lattice-homogenized properties for the diffusion analysis of the super critical water reactor
Two issues may affect the accuracy of computed core reactivities and flux/power distributions for the Super Critical Water Reactor (SCWR) core with traditional core-analysis code RFSP: one is the two-energy-group neutron-diffusion theory; the other is the generation of lattice-homogenized properties with the lattice code based on the single-lattice-cell model without considering the effects of the environment. These two issues are not SCWR specific; however their effect may be more significant for SCWR. It has been illustrated that the lattice-homogenized properties calculated with the single-lattice-cell model is not sufficiently accurate for heterogeneous core configurations such as ACR-1000 checkerboard-voiding and core-reflector interface when adjacent channels experience significant spectrum interaction. To evaluate the adequacy of using two-group neutron-diffusion theory with single-lattice-based lattice properties for the analysis of the SCWR core, a 2-D SCWR benchmark problem was setup with the reference solution provided by the continuous-energy Monte-Carlo code SERPENT. The assessment shows that the traditional two-group neutron-diffusion theory with the single-lattice-cell- based lattice properties is not sufficient to capture either the spectral change or the environment effect for the SCWR core. The solution of the eight-group neutron-diffusion equation with the multicell- based lattice properties is considered appropriate for the analysis of the d SCWR core. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
SU-E-T-31: A Fast Finite Size Pencil Beam (FSPB) Convolution Algorithm for a New Co-60 Arc Therapy Machine
Purpose: Present a fast Finite Size Pencil Beam (FSPB) convolution algorithm for a new Co-60 arc therapy machine. The FSPB algorithm accounts for (i) strong angular divergence (short SAD), (ii) heterogeneity effect for primary attenuation, and (iii) source energy spectrum. Methods: The FSPB algorithm is based on a 0.5×0.5-cm2 dose kernel calculated using the GEPTS (Gamma Electron and Positron Transport System) Monte Carlo code. The dose kernel is tabulated using a thin XYZ mesh (0.1 mm steps in lateral directions) for radius less than 1 cm and using an RZ mesh (with varying steps) for larger radial distance. To account for SSD effect, 11 dose kernels with SSDs varying between 30 cm to 80 cm are calculated. Maynord factor and “lateral stretching” are applied to account for differences between closest and actual SSD. Appropriate rotations and second order interpolation are used to calculate the dose from a given beamlet to a point. Results: Accuracy: Dose distributions in water with 80 cm SSD are calculated using the new FSPB convolution algorithm and full Monte Carlo simulation (gold standard). Figs 1–4 show excellent agreements between FSPB and Monte Carlo calculations for different field sizes and at different depths. The dose distribution for a prostate case is calculated using FSPB (Fig.5). Sixty conformal beams with rectum blocking are assumed. Figs 6–8 show the comparison with Monte Carlo simulation based on the same beam apertures. The excellent agreement demonstrates the accuracy of the new algorithm in handling SSD variation, oblique incidence, and scatter contribution.Speed: The FSPB convolution algorithm calculates 28 million dose points per second using a single 2.2-GHz CPU. The present algorithm is seven times faster than a similar algorithm from Gu et al. (Phys. Med. Biol. 54, 2009, 6287–6297). Conclusion: A fast and accurate FSPB convolution algorithm was developed and benchmarked.
60 APPLIED LIFE SCIENCES↗
General purpose dynamic Monte Carlo with continuous energy for transient analysis
For safety assessments transient analysis is an important tool. It can predict maximum temperatures during regular reactor operation or during an accident scenario. Despite the fact that this kind of analysis is very important, the state of the art still uses rather crude methods, like diffusion theory and point-kinetics. For reference calculations it is preferable to use the Monte Carlo method. In this paper the dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli4. Also, the method is extended for use with continuous energy. The first results of Dynamic Tripoli demonstrate that this kind of calculation is indeed accurate and the results are achieved in a reasonable amount of time. With the method implemented in Tripoli it is now possible to do an exact transient calculation in arbitrary geometry. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Increasing the energy dynamic range of solid-state nuclear track detectors using multiple surfaces
Solid-state nuclear track detectors, such as CR-39, are widely used in physics and in many inertial confinement fusion (ICF) experiments. In the ICF experiments, the particles of interest, such as D{sup 3}He-protons, have ranges of order of the detector thickness. In this case, the dynamic range of the detector can be extended by recording data on both the front and back sides of the detector. Higher energy particles which are undetectable on the front surface can then be measured on the back of the detector. Studies of track formation under the conditions on the front and back of the detector reveal significant differences. Distinct front and back energy calibrations of CR-39 are therefore necessary and are presented for protons. Utilizing multiple surfaces with additional calibrations can extend the range of detectable energies on a single piece of CR-39 by up to 7-8 MeV. The track formation process is explored with a Monte Carlo code, which shows that the track formation difference between front and back is due to the non-uniform ion energy deposition in matter.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Multigroup Constant Calculation with Static α-Eigenvalue Monte Carlo for Time-Dependent Neutron Transport Simulations
Herein, multigroup constants for deterministic methods that preserve the time-dependent physics of the neutron transport equations are derived. Alternative multigroup constant weighting spectra are discussed: (1) the fundamental k-eigenfunction, (2) the fundamental α-eigenfunction, and (3) a composite of several α-modes. To generate the fundamental α-eigenfunction for calculating the multigroup constants, a static fundamental α-eigenvalue method is implemented into the open source Monte Carlo code OpenMC. Several static and kinetic problems are devised to verify the implementations and to investigate the relative performance of the alternative multigroup constant weighting spectra. Results emphasize that as a multigroup constant weighting spectrum, the fundamental α-eigenfunction offers physical characteristics that make it advantageous (in producing accurate solutions) over the typically used fundamental k-eigenfunction.
42 ENGINEERING↗
Shielding Analysis of a Preclinical Bremsstrahlung X-ray FLASH Radiotherapy System within a Clinical Radiation Therapy Vault
A preclinical radiotherapy system producing FLASH dose rates with 12 MV bremsstrahlung x rays is being developed at Stanford University and SLAC National Accelerator Laboratory. Because of the high expected workload of 6,800 Gy w –1 at the isocenter, an efficient shielding methodology is needed to protect operators and the public while the preclinical system is operated in a radiation therapy vault designed for 6 MV x rays. Here, in this study, an analysis is performed to assess the shielding of the local treatment head and radiation vault using the Monte Carlo code FLUKA and the empirical methodology given in the National Council on Radiation Protection and Measurements Report 151. Two different treatment head shielding designs were created to compare single-layer and multilayer shielding methodologies using high-Z and low-Z materials. The multilayered shielding methodology produced designs with a 17% reduction in neutron fluence leaking from the treatment head compared to the single layered design of the same size, resulting in a decreased effective dose to operators and the public. The conservative assumptions used in the empirical methods can lead to over-shielding when treatment heads use polyethylene or multilayered shielding. High-Z/Low-Z multilayered shielding optimized via Monte Carlo is shown to be effective in the case of treatment head shielding and provide more effective shielding design for external beam radiotherapy systems that use 12 MV bremsstrahlung photons. Modifications to empirical methods used in the assessment of MV radiotherapy systems may be warranted to capture the effects of polyethylene in treatment head shielding.
61 RADIATION PROTECTION AND DOSIMETRY↗
Development of the Advanced D1S for Shutdown Dose Rate Calculations in Fusion Reactors
The assessment of the shutdown dose rate due to neutron activation is fundamental for shielding design, materials selection, licensing and planning of maintenance operations in high performances fusion devices. The shutdown dose rate calculations require the combined use of radiation transports and inventory codes. The typical approach consists in a two steps procedure. A pure neutron transport simulation is performed in a first step to calculate the spatial distribution of neutron spectra. The results are used as input for an activation calculation that provides the decay gamma source used in a second step for a pure gamma transport simulation (rigorous-two step, R2S). The Direct 1-Step (D1S) approach has been developed more than ten years ago by ENEA and the ITER team: this technique relies on a coupled neutron-decay gamma transport in a single simulation. It is based on the use of a modified version of the MCNP5 Monte Carlo code with specially tailored nuclear cross-section data for transport and FISPACT inventory code for temporal behaviour. The 'Advanced-D1S' is an improved version of D1S in which additional computational features have been introduced, such as the mesh tally capability and continuous time evolution assessment of the dose rate in single run dose. It has been validated through measurements at JET tokamak and comparison with R2S codes and a future benchmark experiment during next high performance deuterium-tritium campaign is in preparation at JET. It has also been extensively used for ITER calculations and recently applied to DEMO assessment. Present limitations and in progress code improvement strategy are described and discussed in this paper.
61 RADIATION PROTECTION AND DOSIMETRY↗
Uncertainty Analysis for a Criticality Benchmark [Slides]
This presentation discusses how KRUSTY was analyzed with LANL’s MCNP neutron transport code, Monte Carlo N-Particle. In the Monte Carlo method, neutrons are born in fission events and fly around the geometry having events (e.g., scatter, capture, fission, escape) according to the natural probabilities. “Natural probabilities” are given as neutron cross sections , which are measured and evaluated nuclear data. One of the goals of the benchmark project is to improve the nuclear cross sections (and the codes that use them). In this work, the response of interest is called k eff , which describes the neutron multiplication in a near critical system. Like most quantities, k eff is not measured directly; it is only inferred. In summation, MCNP results for k eff matched the measurements extremely well. The evaluated uncertainties are asymmetric and smaller than 0.00100 (e.g., + 0.00080 / – 0.00062). The largest uncertainties are due to alignment.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗