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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.

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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Comparison of the MCNP6 and FLUKA Codes in Shielding Calculation at ProNova Proton Therapy Facility

The Monte Carlo method is a powerful tool to simulate particle transport in materials, and can be effectively used to design radiation shielding. The purpose of this study is to compare the results from two Monte Carlo codes (MCNP6 and FLUKA ) in shielding design. The calculation was also compared with the measured results from the ProNova Superconducting 360 degrees (SC360) installation at the Knoxville Provision Center for Proton Therapy (PCPT) facility. The SC360 230 MeV continue wave (CW) proton beam is produced by an isochronous cyclotron, the Sumitomo P235. The maximum beam that is extracted into the beamline is 300 nA, which is of 60% of the full current generated in cyclotron. The SC360 facility has an energy selection system for adjusting the proton range, one fixed beam and one gantry room for rotating the beam around the stationary patient. At the entrance to each treatment room beamline, a thickness adjustable Beryllium degrader inserted in the beam path is used to achieve the desirable beam energy. When protons interact with beamline components, the secondary neutrons become the dominant source to determine the radiation shielding. In this study, the dose rate outside the lateral wall of the beamline vault was discussed. The calculations from two Monte Carlo codes were compared and they agreed very well. The dose rates measured by a Wendi-II Neutron Detector agreed within a factor of less than seven compared with the calculation results. The difference may be explained by a variety of factors such as the uncertainty in knowing the exact composition of shielding materials, geometry difference, and the measurement uncertainty due to instrumentation. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Reassessing the MCNP Random Number Generator

Random number generators are integral components to Monte Carlo codes. They provide the pseudorandom number sequence used to actually sample the distributions of interest. As a result, they are one of the most important components to the software. The current recommended MCNP random number generator is a 63-bit linear congruential generator (LCG). This generator is quite fast, but it has some drawbacks. First, it only has a period of 2 63 . Due to the necessarily non-optimal usage of random numbers to ensure parallel reproducibility, this amount is too few to guarantee random number sequences are not reused in all configurations the code runs under. As simulation size increases, users will need to be aware of the limitations of the generator and tune configuration variables to best suit their simulations, or they will need to assume that reuse is not negatively affecting their answers. Neither of these are optimal. Second, small LCGs are fairly weak in bit generation quality, and this can have an unknown impact on the quality of the simulation. This paper is an investigation into whether or not more modern random number generators can supersede the current ones. The goal is to find a generator that is similar or superior in speed to the LCGs, has a state space large enough to make strong guarantees about random number reuse, and passes all modern random number test suites. If such a generator is found, it would eliminate the need for the user to even be aware of the limitations of the random number generator and would simplify the use of the code. This paper will be broken into several parts. Sec. 2 will discuss the evolution of the random number generator within the MCNP code. Sec. 3 will go over what a Monte Carlo code needs from a generator to be reproducible and portable and how the current generator behaves in that light. Sec. 4 goes through how each generator was tested. Finally, Sec. 5 will discuss improvements that could be made to the code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

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]

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 4 He 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) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28 Mg) 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.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

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

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. Although the MCNP Monte Carlo code is used in this paper, the reciprocity technique can be used equally well with other Monte Carlo codes.

61 RADIATION PROTECTION AND DOSIMETRY↗
NASA NTRS2004

Infrared Spectroscopy of Star Formation in Galactic and Extragalactic Regions

Last year we submitted and had accepted a paper entitled "The Far-Infrared Emission Line and Continuum Spectrum of the Seyfert Galaxy NGC 1068," by Spinoglio, L., Malkan, M., Smith. HA, Gonzalez-Alfonso, E., and Fischer, J. This analysis was based on the SWAS Monte Carlo code modeling of the OH lines in galaxies observed by ISO. Since that meeting last spring considerable effort has been put into improving the Monte Carlo code. A group of European astronomers, including Prof. Eduardo Gonzalez-Alfonso, had been performing Monte Carlo modeling of other molecules seen in ISO galaxies. We used portions of this grant to bring Prof. Gonzalez-Alfonso to Cambridge for an intensive working visit. A second major paper on the ISO IR spectroscopy of galaxies, "The Far Infrared Spectrum of Arp 220," Gonzalez-Alfonso, E., Smith. H., Fischer, J., and Cernicharo, J., is in press. Spitzer science development was the major component of this past year;s research. This program supported the development of five Early Release Objects for Spitzer observations on which Dr. Smith was Principal Investigator or Co-Investigator, and another five proposals for GO time. The early release program is designed to rapidly present to the public and the scientific community some exciting results from Spitzer in the first months of its operation. The Spitzer instrument and science teams submitted proposals for ERO objects, and a competitive selection process narrowed these down to a small group with exciting science and realistic observational parameters. This grant supported Dr. Smith's participation in the ERO process, including developing science goals, identifying key objects for observation, and developing the detailed AOR (observing formulae) to be use by the instruments for mapping, integrating, etc.). During this year Dr. Smith worked on writing up and publishing these early results. The attached bibliography includes six of Dr. Smith's articles. During this past year Dr. Smith also led or helped to develop proposals for ten Spitzer GO Programs, and three others. Appendix B lists the programs involved.

Smith, Howard A.↗
DOE OSTI2012

Application of GRS method to evaluation of uncertainties of calculation parameters of perspective sodium-cooled fast reactor

A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used for estimation of errors of calculation parameters (K{sub eff}, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Release of Continuous Representation for S(α,β) ACE Data

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 continuous 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.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Comparisons of dosimetric accuracy and calculation time of ARCHER and MCNP5 codes for the Ir-192 brachytherapy case - Paper 1

In the last decade, the graphics processing unit (GPU) technology has emerged as a viable player in high-performance computing. A GPU card is designed to carry out thousands of single instructions simultaneously, showing unprecedented computing power when compared with the traditional CPU technology. It has been proved that some computationally intensive programs can be accelerated drastically on the GPUs including Monte Carlo simulations. Unfortunately, existing production Monte Carlo codes cannot be easily ported to the GPU environment. To address the needs, a new Monte Carlo code ARCHER (Accelerated Radiation-transport Computations in Heterogeneous Environment) is being developed at Rensselaer Polytechnic Institute, employing the latest GPU and the Intel Xeon Phi coprocessor technologies. We can execute the code which was written using C and C++ on both CPU and GPU platforms. The ARCHER GPU code was written in CUDA C 5.0 and was tested on the Tesla M2090. There have been a few studies testing the ARCHER GPU code on several different platforms. In particular, Su et al has tested the ARCHER for external-beam radiation treatment 20 MeV monoenergetic monodirectional parallel electron beam. The comparison of the calculation results between EGSnrc and ARCHERGPU was impressive. A remarkable 224- fold speed-up factor was obtained using a NVIDIA Tesla M2090 GPU compared to a CPU. In this study, we applied the ARCHER code to dose calculation for a relatively new procedure, called accelerated partial breast irradiation (APBI) brachytherapy. The APBI brachytherapy dose calculation was based on RPI-AF phantom which is a 3D heterogeneous patient model. We then compared the dosimetric accuracy and calculation times of ARCHER and MCNP5 to gain insight into the accuracy and speedup of the GPU-based ARCHER code. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

TRIPOLI-4{sup R} version 9 Shielding for the Shielding Community: Overview, Relevant Benchmarking for Shielding and Licensing Policy - Paper 49

This is a review paper of the Shielding only version of TRIPOLI-4. TRIPOLI-4 and TRIPOLI-4 S: TRIPOLI{sup R} represents a family of radiation transport codes using the Monte Carlo method (1,2). It has been continuously developed at CEA since the mid 60's. TRIPOLI-4{sup R}, being the fourth generation, has been developed starting from the mid 90's in C++. It tracks neutrons from 20 MeV down to 10{sup -5} eV, photons, electrons and positrons from 100 MeV to 1 keV, in 3D arbitrary geometries either in TRIPOLI{sup R} proprietary format or ROOT(4) format. It reads its continuous energy nuclear data from any evaluation in ENDF format, including but not limited to JEFF-3.1.1 and ENDF/BVII.0. It has various simulation modes, fixed-source (with or without fission), criticality, or both of them chained. A variety of tallies are available: averaged flux in a volume (track length and collision estimators), surface flux and current, point flux, reaction rates, deposited energy, dpa, pka, gamma spectroscopy, mesh tallies, tallies for criticality (k{sub eff}, beta eff, kinetic parameters). It provides the user with embedded response functions from IRDF and EAF as well. Several variance reduction techniques (3), including automated ones, make it particularly suitable to deep penetration problems (implicit capture, Russian roulette, exponential biasing with automatic estimation of the importance map). It may run in parallel, on single multi-core machines as well as heterogeneous networks of stations or massively parallel machines. It comes with productivity tools designed to speed-up the elaboration and verification of the input deck as well as the output analysis: 2D interactive display of geometry; 2D display of mesh tallies and collision sites; 3D input deck modeler; compatibility with ROOT tools to post-process TRIPOLI{sup R} outputs. Its V and V test base is made of more than 1000 internationally accepted benchmarks from ICSBEP and SINBAD (REPLICA, ASPIS, NESDIP) and covers all its application domains, including several shielding cases which will be detailed in the talk. TRIPOLI-4{sup R} is the reference Monte Carlo code for CEA, EDF, and also for AREVA for a number of applications. A 'Shielding-only' version has been designed to specifically answer the needs of the shielding community. It is distributed through the NEA Databank with a license for R and D, covering the fusion applications, whose terms are detailed in the paper.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Experimental study for subcriticality measurement of fuel debris in the Fukushima Daiichi reactor - 14582

Fuel debris defueling in the Fukushima Daiichi reactors is planned. Furthermore, retrieved fuel debris will be packed and stored in casks for a long time. In these processes, from the viewpoint of criticality safety, the subcriticality of the fuel debris should be confirmed. Although there are some subcriticality measurement methods, the applicability of these methods to the fuel debris must be investigated because the elemental composition and shape of the fuel debris are unknown. In this study, we focus on the exponential experiment method which is able to measure a wide range of subcriticality. Basic experiments to validate the methods were carried out using Toshiba Nuclear Critical Assembly (NCA) and {sup 3}He Position Sensitive Detector (PSD). The decay curves of neutron flux distributions were measured by the PSD detector and fitted to a theoretical formula. The elemental compositions and neutron spectrum of the fuel debris have been estimated with a severe accident code and a Monte Carlo code. In the experiment, the neutron spectrum of the fuel debris was simulated in the NCA core. The porosity of the fuel debris was assumed to be 60%, which corresponds to be H/U=7.5 and Volume Ratio=1.5. In order to evaluate the effect of the uncertainty of the compositions on the measured subcriticality, a homogeneous region and a heterogeneous region were configured in an experimental core. These regions consisted of fuel rods with different UO{sub 2} enrichment zone or different gadolinia concentration zone in the axial direction. The effective multiplication factors deduced by the exponential experiment method were a range of 0.94 to 0.98. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Reactor cell neutron dose for the molten salt breeder reactor conceptual design

The private sector’s interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Radiation Protection Around High-intensity Laser Interactions with Solid Targets

Interaction of a high-intensity optical laser with a solid target can generate an ionizing radiation hazard in the form of high-energy "hot" electrons and bremsstrahlung, resulting from hot electrons interacting with the target itself and the surrounding target chamber. Previous studies have characterized the bremsstrahlung dose yields generated by such interactions for lasers in the range of 10 17 to 10 22 W cm -2 using particle-in-cell code EPOCH and Monte Carlo code FLUKA. Electron measurements based on a depth-dose approach are presented for two laser intensities, which indicate a Maxwellian distribution is more suitable for estimating the hot electrons' energy distribution. Also, transmission factors for the resulting bremsstrahlung for common shielding materials are calculated with FLUKA, and shielding tenth-value-layer thicknesses are also derived. In combination with the bremsstrahlung dose yield, the tenth-value layers provide radiation protection programs the means to evaluate radiation hazards and design shielding for high-intensity laser facilities.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

TOPAS: An innovative proton Monte Carlo platform for research and clinical applications

Purpose: While Monte Carlo particle transport has proven useful in many areas (treatment head design, dose calculation, shielding design, and imaging studies) and has been particularly important for proton therapy (due to the conformal dose distributions and a finite beam range in the patient), the available general purpose Monte Carlo codes in proton therapy have been overly complex for most clinical medical physicists. The learning process has large costs not only in time but also in reliability. To address this issue, we developed an innovative proton Monte Carlo platform and tested the tool in a variety of proton therapy applications. Methods: Our approach was to take one of the already-established general purpose Monte Carlo codes and wrap and extend it to create a specialized user-friendly tool for proton therapy. The resulting tool, TOol for PArticle Simulation (TOPAS), should make Monte Carlo simulation more readily available for research and clinical physicists. TOPAS can model a passive scattering or scanning beam treatment head, model a patient geometry based on computed tomography (CT) images, score dose, fluence, etc., save and restart a phase space, provides advanced graphics, and is fully four-dimensional (4D) to handle variations in beam delivery and patient geometry during treatment. A custom-designed TOPAS parameter control system was placed at the heart of the code to meet requirements for ease of use, reliability, and repeatability without sacrificing flexibility. Results: We built and tested the TOPAS code. We have shown that the TOPAS parameter system provides easy yet flexible control over all key simulation areas such as geometry setup, particle source setup, scoring setup, etc. Through design consistency, we have insured that user experience gained in configuring one component, scorer or filter applies equally well to configuring any other component, scorer or filter. We have incorporated key lessons from safety management, proactively removing possible sources of user error such as line-ordering mistakes. We have modeled proton therapy treatment examples including the UCSF eye treatment head, the MGH stereotactic alignment in radiosurgery treatment head and the MGH gantry treatment heads in passive scattering and scanning modes, and we have demonstrated dose calculation based on patient-specific CT data. Initial validation results show agreement with measured data and demonstrate the capabilities of TOPAS in simulating beam delivery in 3D and 4D. Conclusions: We have demonstrated TOPAS accuracy and usability in a variety of proton therapy setups. As we are preparing to make this tool freely available for researchers in medical physics, we anticipate widespread use of this tool in the growing proton therapy community.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

Design of water-moderated heterogeneous cores in new STACY facility through JAEA/IRSN collaboration - 14363

In the post-Fukushima prospect, Japan Atomic Energy Agency, JAEA, intends to re-open its Static Experiment Critical Facility STACY, in order to contribute to the validation of fuel debris calculations in conditions close to those affected by the Fukushima Daiichi Nuclear Power Station accident. The reactor core of STACY is being modified by JAEA from homogeneous solution fuel core to heterogeneous pin- type fuel core. The first step of critical experiments is to establish reference experiments involving lattices of UO{sub 2} rods without fuel debris. A design of such experiments is in progress for the widest range of moderation ratios in the frame of JAEA/IRSN collaboration. To anticipate the specification values of materials to be manufactured (grids, rods), an evaluation of uncertainties pertaining to the different parameters driving k{sub eff} is paramount. The following uncertainties of experimental cores are evaluated to estimate the reactivity effect of engineering tolerance/or measurements of the modified STACY: (1) Density of fuel pellets, (2) Enrichment of uranium, (3) Diameter of fuel pellets, (4) Thickness of fuel claddings, (5) Impurities of fuel pellets, (6) Impurities of fuel claddings and (7) Positioning of fuel rods in the lattice. For these calculations, three types of moderation conditions were considered within the experimental parameters of STACY: the lowest moderation (lattice pitch = 1.15 cm; V{sub m}/V{sub f}=1.2), optimum moderation (lattice pitch = 1.5 cm, V{sub m}/V{sub f} = 2.9) and an over moderated case (lattice pitch = 2.55 cm; V{sub m}/V{sub f} = 11). The calculations to derive the overall uncertainty of modified STACY were performed propagating the 1σ uncertainties of parameters of existing STACY in terms of k{sub eff}. For that purpose, a perturbation calculation (differential operator) technique and a stochastic geometry modeling technique were used with the MCNP5 continuous energy Monte Carlo code and a point-wise cross section library based on the JENDL-4.0 evaluated nuclear data library. Moreover, the results of calculations were compared with those calculated by two continuous energy Monte Carlo codes using the correlated sampling technique: - An experimental version of the Japanese MVP2 code combined to the JENDL-4.0 nuclear data library, - The MORET 5 code, developed by IRSN, combined to the JEFF-3.1 nuclear data library. The proposed paper discusses all the aforementioned issues, and confirms that the uncertainty pertaining to the critical experiments of the STACY is sufficiently small. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as virtual experiments that give deeper and long lasting understanding of core principles. The student can then make sound judgements in novel situations encountered beyond routine clinical activities.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code

High-fidelity deterministic transport codes require highly accurate multigroup cross sections (MGXS). Monte Carlo is increasingly cited as a reactor-agnostic approach to MGXS generation since it is unconstrained by the engineering-based approximations that limit the applicability of deterministic MGXS generation tools. This paper introduces a new framework that uses the OpenMC Monte Carlo code to generate MGXS for use in multigroup transport codes. The openmc.mgxs module is built atop OpenMC's Python application programming interface to process tally data output by the OpenMC executable. This paper validates the module to generate MGXS that enable the multigroup OpenMOC transport code to compute eigenvalues to within 50 pcm and fission rates to within 1% of reference solutions for two heterogeneous pressurized water reactor benchmarks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

SU-E-T-761: TOMOMC, A Monte Carlo-Based Planning VerificationTool for Helical Tomotherapy

Purpose: Present a new Monte Carlo code (TOMOMC) to calculate 3D dose distributions for patients undergoing helical tomotherapy treatments. TOMOMC performs CT-based dose calculations using the actual dynamic variables of the machine (couch motion, gantry rotation, and MLC sequences). Methods: TOMOMC is based on the GEPTS (Gama Electron and Positron Transport System) general-purpose Monte Carlo system (Chibani and Li, Med. Phys. 29, 2002, 835). First, beam models for the Hi-Art Tomotherpy machine were developed for the different beam widths (1, 2.5 and 5 cm). The beam model accounts for the exact geometry and composition of the different components of the linac head (target, primary collimator, jaws and MLCs). The beams models were benchmarked by comparing calculated Pdds and lateral/transversal dose profiles with ionization chamber measurements in water. See figures 1–3. The MLC model was tuned in such a way that tongue and groove effect, inter-leaf and intra-leaf transmission are modeled correctly. See figure 4. Results: By simulating the exact patient anatomy and the actual treatment delivery conditions (couch motion, gantry rotation and MLC sinogram), TOMOMC is able to calculate the 3D patient dose distribution which is in principal more accurate than the one from the treatment planning system (TPS) since it relies on the Monte Carlo method (gold standard). Dose volume parameters based on the Monte Carlo dose distribution can also be compared to those produced by the TPS. Attached figures show isodose lines for a H&N patient calculated by TOMOMC (transverse and sagittal views). Analysis of differences between TOMOMC and TPS is an ongoing work for different anatomic sites. Conclusion: A new Monte Carlo code (TOMOMC) was developed for Tomotherapy patient-specific QA. The next step in this project is implementing GPU computing to speed up Monte Carlo simulation and make Monte Carlo-based treatment verification a practical solution.

60 APPLIED LIFE SCIENCES↗