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At least 397 records · Page 22

Multigroup Cross-section Generation in MCNP6.3 [Slides]

This presentation states that in comparison to the NJOY-produced multigroup cross sections, the MCNP-produced multigroup cross sections are generally consistent. Statistical uncertainties, however, are challenging and the unresolved resonance region may be looked at in the future. It also discusses how the SPM and LCS options were compared to each other for internal consistency. Additionally, some reactor pin-cell-like problems were used to compare to multigroup capabilities in other Monte Carlo codes (e.g., Serpent, OpenMC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

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 or 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)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Genetic algorithm-based optimisation of the few-group structure for lead fast reactors analysis

The optimal choice of the few-group structure for full-core transient analyses is still an open issue in reactor physics, especially for fast system like the lead fast reactor. One possible approach to select the group boundaries is represented by heuristic search algorithms, such as evolutionary ones. In this paper, a genetic algorithm coupled with the SIMMER code is employed to determine optimized six-group boundaries for the analysis of the ALFRED reactor. The Serpent Monte Carlo code is adopted to produce both the fine-group cross section library and the fine-group flux, used as a figure of merit to drive the genetic optimisation. The results show that the algorithm is indeed able to find satisfactory solutions that comply with the set objectives and can be reasonably interpreted in light of the underlying physics of the considered core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A BrachyPhantom for verification of dose calculation of HDR brachytherapy planning system

Purpose: To develop a calibration phantom for {sup 192}Ir high dose rate (HDR) brachytherapy units that renders possible the direct measurement of absorbed dose to water and verification of treatment planning system.Methods: A phantom, herein designated BrachyPhantom, consists of a Solid Water™ 8-cm high cylinder with a diameter of 14 cm cavity in its axis that allows the positioning of an A1SL ionization chamber with its reference measuring point at the midheight of the cylinder's axis. Inside the BrachyPhantom, at a 3-cm radial distance from the chamber's reference measuring point, there is a circular channel connected to a cylindrical-guide cavity that allows the insertion of a 6-French flexible plastic catheter from the BrachyPhantom surface. The PENELOPE Monte Carlo code was used to calculate a factor, P{sub sw}{sup lw}, to correct the reading of the ionization chamber to a full scatter condition in liquid water. The verification of dose calculation of a HDR brachytherapy treatment planning system was performed by inserting a catheter with a dummy source in the phantom channel and scanning it with a CT. The CT scan was then transferred to the HDR computer program in which a multiple treatment plan was programmed to deliver a total dose of 150 cGy to the ionization chamber. The instrument reading was then converted to absorbed dose to water using the N{sub gas} formalism and the P{sub sw}{sup lw} factor. Likewise, the absorbed dose to water was calculated using the source strength, S{sub k}, values provided by 15 institutions visited in this work.Results: A value of 1.020 (0.09%, k= 2) was found for P{sub sw}{sup lw}. The expanded uncertainty in the absorbed dose assessed with the BrachyPhantom was found to be 2.12% (k= 1). To an associated S{sub k} of 27.8 cGy m{sup 2} h{sup −1}, the total irradiation time to deliver 150 cGy to the ionization chamber point of reference was 161.0 s. The deviation between the absorbed doses to water assessed with the BrachyPhantom and those calculated by the treatment plans and using the S{sub k} values did not exceed ±3% and ±1.6%, respectively.Conclusions: The BrachyPhantom may be conveniently used for quality assurance and/or verification of HDR planning system with a priori threshold level to spot problems of 2% and ±3%, respectively, and in the long run save time for the medical physicist.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

SU-F-18C-09: Assessment of OSL Dosimeter Technology in the Validation of a Monte Carlo Radiation Transport Code for CT Dosimetry

Purpose: To assess the utility of optically stimulated luminescent (OSL) dosimeter technology in calibrating and validating a Monte Carlo radiation transport code for computed tomography (CT). Methods: Exposure data were taken using both a standard CT 100-mm pencil ionization chamber and a series of 150-mm OSL CT dosimeters. Measurements were made at system isocenter in air as well as in standard 16-cm (head) and 32-cm (body) CTDI phantoms at isocenter and at the 12 o'clock positions. Scans were performed on a Philips Brilliance 64 CT scanner for 100 and 120 kVp at 300 mAs with a nominal beam width of 40 mm. A radiation transport code to simulate the CT scanner conditions was developed using the GEANT4 physics toolkit. The imaging geometry and associated parameters were simulated for each ionization chamber and phantom combination. Simulated absorbed doses were compared to both CTDI{sub 100} values determined from the ion chamber and to CTDI{sub 100} values reported from the OSLs. The dose profiles from each simulation were also compared to the physical OSL dose profiles. Results: CTDI{sub 100} values reported by the ion chamber and OSLs are generally in good agreement (average percent difference of 9%), and provide a suitable way to calibrate doses obtained from simulation to real absorbed doses. Simulated and real CTDI{sub 100} values agree to within 10% or less, and the simulated dose profiles also predict the physical profiles reported by the OSLs. Conclusion: Ionization chambers are generally considered the standard for absolute dose measurements. However, OSL dosimeters may also serve as a useful tool with the significant benefit of also assessing the radiation dose profile. This may offer an advantage to those developing simulations for assessing radiation dosimetry such as verification of spatial dose distribution and beam width.

61 RADIATION PROTECTION AND DOSIMETRY↗

Generation of Bondarenko F factors with MCNP5 for use in scale transport codes

Generally there are three methods of cross-section processing available when using the Scale computer code. They are NITAWL, BONAMI, and CENTRM, with CENTRM being the most common and accurate, but computationally expensive. In order to improve the accuracy of BONAMI (which uses the Bondarenko Method), new Bondarenko/F Factors were to be generated that will smooth out the current F Factors that are being generated using CENTRM. The case discussed here involves using a dedicated Monte Carlo code (MCNP5) to calculate the transport solution from which shielded cross-sections can be produced directly for the transport geometry/mesh. A simple program was created to parse and collect the tallied cross-sections from the MCNP output, which was fed into a modified CLAROL input (a module that replaces or adds data in an AMPX master library), which used the cross-sections obtained from MCNP to calculate new F-factors and update the SCALE library. This approach allows the use of other methods to generate the shielded cross-sections and for easy comparison to existing results. Initial proof-of-principle calculations were carried out for an various cases using various transport solvers, such as NEWT, KENO, and XSDRN, with BONAMI in SCALE. Poor results were obtained using cross-sections generated using infinite homogeneous cases, but good results were obtained by using pin cells in an infinite lattice. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The MCNP6 Analytic Criticality Benchmark Suite

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) 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.

97 MATHEMATICS AND COMPUTING↗

Direct NeTS sampling of nuclear graphite $S(α, β, T)$ in Serpent

For advanced reactor applications, Neural Thermal Scattering (NeTS) modules were developed to predict the thermal scattering law (TSL or $S(α, β, T)$) of a nuclear graphite neutron moderator. NeTS are multi-layer, feedforward artificial neural networks, which act as universal function approximators designed for TSL datasets. In this case, a 4-layer neural network with 164 neurons per layer is trained using FLASSH evaluated data in PyTorch and serialized as a torchscript dictionary to predict $S(α, β, T)$ on-the-fly. Relative, absolute and maximum percent deviations of NeTS from File 7 data generated using the FLASSH code are on the order of 0.01%, 0.1% and 1%, respectively, with low inference latencies of 0.000172 s per $S(α, β, T)$ at a given temperature. Capturing the full dimensionality of possible inelastic neutron-lattice interactions, NeTS functionality is embedded in the Serpent Monte Carlo code, where $S(α, β, T)_{NeTS}$ sampling is conducted on-the-fly and compared to ACE look-up-tables for predicting TREAT criticality. k-eff differences between sampling algorithms of 6 pcm are observed and are within the order of Monte Carlo uncertainty. Compared to discrete and continuous-energy ACE files (30 MB and 131 MB per temperature), the NeTS format is on the order of 200–300 kB for a continuous-temperature, interpolation-free representation of $S(α, β, T)$ and cross sections. NeTS-in-Serpent runtimes comparable with ACE look-up tables are achieved by scaling NeTS for high performance computing architectures with hybrid OpenMP + MPI parallelization. This work validates a novel, self-contained reactor physics framework for predictive cross sections, and demonstrates a general methodology for embedding modern machine learning libraries within existing neutronic analysis frameworks.

Nuclear Criticality Safety Program (NCSP)↗

A Monte Carlo neutron transport code for eigenvalue calculations on a dual-GPU system and CUDA environment

Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SU-E-T-304: Study of Secondary Neutrons From Uniform Scanning Proton Beams

Purpose: Secondary neutrons are unwanted byproducts from proton therapy and exposure from secondary radiation during treatment could increase risk of developing a secondary cancer later in a patient's lifetime. The purpose of this study is to investigate secondary neutrons from uniform scanning proton beams under various beam conditions using both measurements and Monte Carlo simulations. Methods: CR-39 Plastic Track Nuclear Detectors (PNTD) were used for the measurement. CR-39 PNTD has tissue like sensitivity to the secondary neutrons but insensitive to the therapeutic protons. In this study, we devised two experimental conditions: a) hollow-phantom; phantom is bored with a hollow cylinder along the direction of the beam so that the primary proton passes through the phantom without interacting with the phantom material, b) cylindrical-phantom; a solid cylinder of diameter close to the beam diameter is placed along the beam path. CR-39 PNTDs were placed laterally inside a 60X20X35 cm3 phantom (hollow-phantom) and in air (cylindrical-phantom) at various angles with respect to the primary beam axis. We studied for three different proton energies (78 MeV, 162 MeV and 226 MeV), using a 4 cm modulation width and 5cm diameter brass aperture for the entire experiment and simulation. A comparison of the experiment was performed using the Monte Carlo code FLUKA. Results: The measured secondary neutron dose equivalent per therapeutic primary proton dose (H/D) ranges from 2.1 ± 0.2 to 25.42 ± 2.3 mSv/Gy for the hollow phantom study, and 2.7 ± 0.3 to 46.4 ± 3.4 mSv/Gy for the cylindrical phantom study. Monte Carlo simulations predicated neutron dose equivalent from measurements within a factor of 5. Conclusion: The study suggests that the production of external neutrons is significantly higher than the production of internal neutrons.

07 ISOTOPE AND RADIATION SOURCES↗

TIME-DEPENDENT MULTI-GROUP MULTI-DIMENSIONAL RELATIVISTIC RADIATIVE TRANSFER CODE BASED ON SPHERICAL HARMONIC DISCRETE ORDINATE METHOD

We develop a time-dependent, multi-group, multi-dimensional relativistic radiative transfer code, which is required to numerically investigate radiation from relativistic fluids that are involved in, e.g., gamma-ray bursts and active galactic nuclei. The code is based on the spherical harmonic discrete ordinate method (SHDOM) which evaluates a source function including anisotropic scattering in spherical harmonics and implicitly solves the static radiative transfer equation with ray tracing in discrete ordinates. We implement treatments of time dependence, multi-frequency bins, Lorentz transformation, and elastic Thomson and inelastic Compton scattering to the publicly available SHDOM code. Our code adopts a mixed-frame approach; the source function is evaluated in the comoving frame, whereas the radiative transfer equation is solved in the laboratory frame. This implementation is validated using various test problems and comparisons with the results from a relativistic Monte Carlo code. These validations confirm that the code correctly calculates the intensity and its evolution in the computational domain. The code enables us to obtain an Eddington tensor that relates the first and third moments of intensity (energy density and radiation pressure) and is frequently used as a closure relation in radiation hydrodynamics calculations.

79 ASTRONOMY AND ASTROPHYSICS↗

Preliminary performance evaluation of P++ single event proton scattering algorithms on GPUS - 25688

A need has been realized for fast, accurate charged particle transport in a wide variety of domains. Utilizing hardware acceleration through GPUs now common in user desktops, speedups in simulation comparable to those found with HPC systems can be realized for a fraction of the price. Preliminary results on the P++ proton transport code developed by the Rensselaer Radiation Measurement and Dosimetry Group (RRMDG) are presented. The code has been validated for the 1 GeV to 1 MeV energy regime and is targeted to benefit application areas such as proton radiography, space radiation research, and clinical accelerator therapy. The code is notable for employing a single event, 'on the fly' cross section based transport algorithm rather than the traditional condensed history or step based method seen in many production Monte Carlo codes. Performance improvements for the P++ code have been introduced through MPI parallelization as well as a CUDA version of the transport algorithm capable of running on NVidia GPUs. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Initiation of persistent fission chains in the fast burst reactor Caliban

We provide in this article, experimental data of initiation of persistent fission chains obtained at different supercritical states, using the Fast Burst Reactor CALIBAN. In many past papers, theory has been compared mostly to initiation experiments at various super-prompt critical states, whereas very few experimental data has been published in delayed supercritical states. To fill the lack of data, we have conducted three campaigns on the reactor at reactivities far below 0.7$ which was one of the rare lowest state ever published on a similar assembly [2][1]. We give a justification of the use of the gamma function to fit experimental results of the temporal distributions of waiting times and compare experiments with numerical simulations obtained with a zero-D punctual Monte Carlo code. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SU-E-T-222: Computational Optimization of Monte Carlo Simulation On 4D Treatment Planning Using the Cloud Computing Technology

Purpose: This study evaluated the efficiency of 4D lung radiation treatment planning using Monte Carlo simulation on the cloud. The EGSnrc Monte Carlo code was used in dose calculation on the 4D-CT image set. Methods: 4D lung radiation treatment plan was created by the DOSCTP linked to the cloud, based on the Amazon elastic compute cloud platform. Dose calculation was carried out by Monte Carlo simulation on the 4D-CT image set on the cloud, and results were sent to the FFD4D image deformation program for dose reconstruction. The dependence of computing time for treatment plan on the number of compute node was optimized with variations of the number of CT image set in the breathing cycle and dose reconstruction time of the FFD4D. Results: It is found that the dependence of computing time on the number of compute node was affected by the diminishing return of the number of node used in Monte Carlo simulation. Moreover, the performance of the 4D treatment planning could be optimized by using smaller than 10 compute nodes on the cloud. The effects of the number of image set and dose reconstruction time on the dependence of computing time on the number of node were not significant, as more than 15 compute nodes were used in Monte Carlo simulations. Conclusion: The issue of long computing time in 4D treatment plan, requiring Monte Carlo dose calculations in all CT image sets in the breathing cycle, can be solved using the cloud computing technology. It is concluded that the optimized number of compute node selected in simulation should be between 5 and 15, as the dependence of computing time on the number of node is significant.

60 APPLIED LIFE SCIENCES↗

SU-F-T-281: Monte Carlo Investigation of Sources of Dosimetric Discrepancies with 2D Arrays

Purpose: Intensity modulated radiation therapy (IMRT) poses a number of challenges for properly measuring commissioning data and quality assurance (QA). Understanding the limitations and use of dosimeters to measure these dose distributions is critical to safe IMRT implementation. In this work, we used Monte Carlo simulations to investigate the possible sources of discrepancy between our measurement with 2D array system and our dose calculation using our treatment planning system (TPS). Material and Methods: MCBEAM and MCSIM Monte Carlo codes were used for treatment head simulation and phantom dose calculation. Accurate modeling of a 6MV beam from Varian trilogy machine was verified by comparing simulated and measured percentage depth doses and profiles. Dose distribution inside the 2D array was calculated using Monte Carlo simulations and our TPS. Then Cross profiles for different field sizes were compared with actual measurements for zero and 90° gantry angle setup. Through the analysis and comparison, we tried to determine the differences and quantify a possible angular calibration factor. Results: Minimum discrepancies was seen in the comparison between the simulated and the measured profiles for the zero gantry angles at all studied field sizes (4×4cm{sup 2}, 10×10cm{sup 2}, 15×15cm{sup 2}, and 20×20cm{sup 2}). Discrepancies between our measurements and calculations increased dramatically for the cross beam profiles at the 90° gantry angle. This could ascribe mainly to the different attenuation caused by the layer of electronics at the base behind the ion chambers in the 2D array. The degree of attenuation will vary depending on the angle of beam incidence. Correction factors were implemented to correct the errors. Conclusion: Monte Carlo modeling of the 2D arrays and the derivation of angular dependence correction factors will allow for improved accuracy of the device for IMRT QA.

60 APPLIED LIFE SCIENCES↗

Modeling Proton- and Light Ion-Induced Reactions at Low Energies in the MARS15 Code

An implementation of both ALICE code and TENDL evaluated nuclear data library in order to describe nuclear reactions induced by low-energy projectiles in the Monte Carlo code MARS15 is presented. Comparisons between results of modeling and experimental data on reaction cross sections and secondary particle distributions are shown.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Fission matrix capability for MCNP, Part I - Theory

The theory underlying the fission matrix method is derived using a rigorous Green's function approach. The method is then used to investigate fundamental properties of the transport equation for a continuous-energy physics treatment. We provide evidence that an infinite set of discrete, real eigenvalues and eigenfunctions exist for the continuous-energy problem, and that the eigenvalue spectrum converges smoothly as the spatial mesh for the fission matrix is refined. We also derive equations for the adjoint solution. We show that if the mesh is sufficiently refined so that both forward and adjoint solutions are valid, then the adjoint fission matrix is identical to the transpose of the forward matrix. While the energy-dependent transport equation is strictly bi-orthogonal, we provide surprising results that the forward modes are very nearly self-adjoint for a variety of continuous-energy problems. A companion paper (Part II - Applications) describes the initial experience and results from implementing this fission matrix capability into the MCNP Monte Carlo code. (authors)

97 MATHEMATICS AND COMPUTING↗

Monte Carlo Simulation On X-Ray Produced At High Intensity Laser Facilities - Paper 20

With the rapid development of laser technology, a number of high intensity laser devices are emerging in the world. At the same time, the ionizing radiation hazard produced from the interaction between high intensity laser and solid target has gradually been observed. Because Laser-plasma interactions can accelerate electrons to high energies (10's to 1000's of keV). These hot electrons subsequently interact with ions in the target and generate Bremsstrahlung X-rays, which become an ionizing radiation source. Measurements show that significant amounts of radiation can be generated from this ionizing radiation source at high intensity laser facilities. However, very limited studies have been made on the laser-induced ionizing radiation hazard and the protection issues. In this paper, a X-ray dose calculation model was developed and X-ray dose produced in the interaction between high intensity laser and solid target was studied by simulation using Monte Carlo code. (authors)

42 ENGINEERING↗