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All sourcesNASA NTRSDOE OSTI
NASA NTRS2003

Development of A Monte Carlo Radiation Transport Code System For HEDS: Status Update

Modifications of the Monte Carlo radiation transport code HETC are underway to extend the code to include transport of energetic heavy ions, such as are found in the galactic cosmic ray spectrum in space. The new HETC code will be available for use in radiation shielding applications associated with missions, such as the proposed manned mission to Mars. In this work the current status of code modification is described. Methods used to develop the required nuclear reaction models, including total, elastic and nuclear breakup processes, and their associated databases are also presented. Finally, plans for future work on the extended HETC code system and for its validation are described.

Townsend, Lawrence W.↗
DOE OSTI2018

Detailed modeling of fission with F R E Y A

For many years, the state of the art for treating fission in radiation transport codes has involved sampling from average distributions. However, such “average” fission models have limited interaction-by-interaction capabilities. Energy is not explicitly conserved and no correlations are available because all particles are emitted isotropically and independently. However, in a true fission event, the energies, momenta and multiplicities of emitted particles are correlated. Such correlations are interesting for many modern applications, including detector development and detection of small amounts of special material. Recently, several Monte Carlo codes have become available that calculate complete fission events. Event-by-event techniques are particularly useful because it is possible to obtain the fission products as well as the prompt neutrons and photons emitted during an individual fission process, all with complete kinematic information. It is therefore possible to extract any desired correlation observables. Such codes, when included in broader Monte Carlo transport codes, such as MCNP, can be made broadly available to the community. One such code, FREYA (Fission Reaction Event Yield Algorithm), is particularly fast and can readily generate large samples of complete fission events. We briefly describe here the physics in FREYA and compare our results with relevant available data on prompt neutron and photon emission. We discuss correlated measurements in particular for validation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

SU-E-T-102: Determination of Dose Distributions and Water-Equivalence of MAGIC-F Polymer Gel for 60Co and 192Ir Brachytherapy Sources

Purpose: Analyse the water-equivalence of MAGIC-f polymer gel for {sup 60}Co and {sup 192}Ir clinical brachytherapy sources, through dose distributions simulated with PENELOPE Monte Carlo code. Methods: The real geometry of {sup 60} (BEBIG, modelo Co0.A86) and {sup 192}192Ir (Varian, model GammaMed Plus) clinical brachytherapy sources were modelled on PENELOPE Monte Carlo simulation code. The most probable emission lines of photons were used for both sources: 17 emission lines for {sup 192}Ir and 12 lines for {sup 60}. The dose distributions were obtained in a cubic water or gel homogeneous phantom (30 × 30 × 30 cm{sup 3}), with the source positioned in the middle of the phantom. In all cases the number of simulation showers remained constant at 10{sup 9} particles. A specific material for gel was constructed in PENELOPE using weight fraction components of MAGIC-f: wH = 0,1062, wC = 0,0751, wN = 0,0139, wO = 0,8021, wS = 2,58×10{sup −6} e wCu = 5,08 × 10{sup −6}. The voxel size in the dose distributions was 0.6 mm. Dose distribution maps on the longitudinal and radial direction through the centre of the source were used to analyse the water-equivalence of MAGIC-f. Results: For the {sup 60} source, the maximum diferences in relative doses obtained in the gel and water were 0,65% and 1,90%, for radial and longitudinal direction, respectively. For {sup 192}Ir, the maximum difereces in relative doses were 0,30% and 1,05%, for radial and longitudinal direction, respectively. The materials equivalence can also be verified through the effective atomic number and density of each material: Zef-MAGIC-f = 7,07 e .MAGIC-f = 1,060 g/cm{sup 3} and Zef-water = 7,22. Conclusion: The results showed that MAGIC-f is water equivalent, consequently being suitable to simulate soft tissue, for Cobalt and Iridium energies. Hence, gel can be used as a dosimeter in clinical applications. Further investigation to its use in a clinical protocol is needed.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2019

Demonstration of Comprehensive Ex-Vessel Fluence Capability

Several recent developments in the Consortium for Advanced Simulation of Light Water Reactors’ (CASL) Virtual Environment for Reactor Applications (VERA) were tested to highlight its capability to perform ex-vessel fluence and reaction rate calculations. VERA offers unique capabilities for integrating the deterministic neutronics code, MPACT, with Shift, a Monte Carlo code, to perform high-fidelity in-core and ex-core radiation transport. Applications such as pressure vessel fluence, ex-core detector response, and coupon irradiation analyses take advantage of this coupling. For these applications, MPACT performs the in-core radiation transport with temperature feedback and isotopic depletion through the direct coupling with the CTF subchannel thermal-hydraulics code. Then, MPACT provides the fission source to Shift for a follow-on fixed source radiation transport calculation that tallies all the ex-core responses of interest for each time-dependent statepoint. The variance reduction method, CADIS, which is implemented in Shift, allows for efficient performance of ex-core transport for calculation of ex-core quantities of interest. This milestone report identifies limitations to the current methodology in place and showcases VERA’s current capability to model ex-core quantities of interest.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Transport studies in high-performance field reversed configuration plasmas

A significant improvement of field reversed configuration (FRC) lifetime and plasma confinement times in the C-2 plasma, called High Performance FRC regime, has been observed with neutral beam injection (NBI), improved edge stability, and better wall conditioning [Binderbauer et al., Phys. Plasmas 22, 056110 (2015)]. A Quasi-1D (Q1D) fluid transport code has been developed and employed to carry out transport analysis of such C-2 plasma conditions. The Q1D code is coupled to a Monte-Carlo code to incorporate the effect of fast ions, due to NBI, on the background FRC plasma. Numerically, the Q1D transport behavior with enhanced transport coefficients (but with otherwise classical parametric dependencies) such as 5 times classical resistive diffusion, classical thermal ion conductivity, 20 times classical electron thermal conductivity, and classical fast ion behavior fit with the experimentally measured time evolution of the excluded flux radius, line-integrated density, and electron/ion temperature. The numerical study shows near sustainment of poloidal flux for nearly 1 ms in the presence of NBI.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Substantiation of parameters of the geometric model of the research reactor core for the calculation using the Monte Carlo method

The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Aleatoric and epistemic uncertainties in sampling based nuclear data uncertainty and sensitivity analyses

Sampling based uncertainty and sensitivity analyses due to epistemic input uncertainties, i.e. to an incomplete knowledge of uncertain input parameters, can be performed with arbitrary application programs to solve the physical problem under consideration. For the description of steady-state particle transport, direct simulations of the microscopic processes with Monte Carlo codes are often used. This introduces an additional source of uncertainty, the aleatoric sampling uncertainty, which is due to the randomness of the simulation process performed by sampling, and which adds to the total combined output sampling uncertainty. So far, this aleatoric part of uncertainty is minimized by running a sufficiently large number of Monte Carlo histories for each sample calculation, thus making its impact negligible as compared to the impact from sampling the epistemic uncertainties. Obviously, this process may cause high computational costs. The present paper shows that in many applications reliable epistemic uncertainty results can also be obtained with substantially lower computational effort by performing and analyzing two appropriately generated series of samples with much smaller number of Monte Carlo histories each. The method is applied along with the nuclear data uncertainty and sensitivity code package XSUSA in combination with the Monte Carlo transport code KENO-Va to various critical assemblies and a full scale reactor calculation. It is shown that the proposed method yields output uncertainties and sensitivities equivalent to the traditional approach, with a high reduction of computing time by factors of the magnitude of 100. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Energetic Light Fragment Production Capability in MCNP6

Monte Carlo Methods, Codes, and Applications group within the Computational Physics Division at Los Alamos National Laboratory has led the development of the transport code MCNP6 (Monte Carlo N-Particle transport code, version 6). MCNP6 is a general-purpose, continuous energy, generalized-geometry, time-dependent, Monte-Carlo radiation-transport code designed to track many particle types over broad ranges of energies. It is used around the world in applications ranging from radiation protection and dosimetry, nuclear-reactor design, nuclear criticality safety, detector design and analysis, decontamination and decommissioning, accelerator applications, medical physics, space research, and beyond. At lower energies, the code uses tables of evaluated nuclear data, while for higher energies (> 150 MeV), MCNP6 uses the cascade-exciton model, version 03.03 (CEM03.03), and the Los Alamos quark-gluon string model, version 03.03 (LAQGSM03.03) to model nuclear reactions. Emission of energetic heavy clusters heavier than {sup 4}He from nuclear reactions play a critical role in several applications, including electronics performance in space, human radiation dosages in space or other extreme radiation environments, proton- and hadron-therapy in medical physics, accelerator and shielding applications, and more. None of the available models are able to accurately predict emission of light fragments (LF) from arbitrary reactions. The CEM and LAQGSM event generators in MCNP6 describe quite well the spectra of fragments with sizes up to {sup 4}He across a broad range of target masses and incident energies (up to ∼ 5 GeV for CEM and up to ∼ 1 TeV/A for LAQGSM). However, they do not predict the high-energy tails of LF spectra heavier than {sup 4}He well. Most LF with energies above several tens of MeV are emitted during the precompound stage of a reaction. The 03.03 versions of CEM and LAQGSM do not account for precompound emission of LF larger than {sup 4}He. The goal of this research is to enable MCNP6 to produce high-energy light fragments. These energetic LF may be emitted by our models through three processes: Fermi breakup, preequilibrium, and coalescence. We explore the emission of light fragments through each of these mechanisms and demonstrate an improved agreement with experimental data achieved by extending precompound models to include emission of fragments heavier than {sup 4}He. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Reactor physics benchmark of Westinghouse PWR core design suite for high burnup/high enrichment fuel - Part 2: 2D and 3D core simulations

Westinghouse has performed a comprehensive set of code-to-code benchmark comparisons to corroborate application of its state-of-the-art NEXUS-based code suite to high burnup and high enrichment (HB/HE) fueled PWR core design and reload analysis. The benchmark consisted of a progression in complexity from modeling single 2D pin cells, through 2D lattices and finally to whole core 2D and then 3D scenarios. The accompanying paper (Part 1) presented the results for a variety of pin cell and lattices representative of high energy cores and established the ability of the MPACT code to accurately match the predictions of the Monte-Carlo code SERPENT2 such that the former may be used as a surrogate reference code in spatially complex whole core modeling applications where using Monte-Carlo methods is somewhat impractical. This paper focuses on comparisons at the whole core level using a representative high power density 3-loop Westinghouse NSSS core operating for 24 month cycles employing high enriched fuel with use of combined burnable absorbers to hold down excess reactivity. The evolution of the core reactivity and power distribution characteristics throughout the cycle are calculated using the Westinghouse NEXUS-based code suite and compared to the results using MPACT. The results of this whole core benchmark confirm the conclusions that the Westinghouse code suite accurately models HB/HE fuel designs, with very good prediction of core reactivity and power distribution throughout the cycle. The industry can be confident that the NEXUS code system can be deployed to support PWR reload safety and operational analyses for cores featuring HB/HE fuel and 24-month cycle operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Enhancing Monte Carlo Workflows for Nuclear Reactor Analysis with Metamodel-Driven Modeling

Monte Carlo codes are essential components of many reactor physics simulation workflows as high-fidelity continuous-energy neutron transport solvers. Among Monte Carlo radiation transport codes, MCNP is particularly notable due to its diverse simulation capabilities, large user base, and long validation history. Despite being a powerful simulation tool, MCNP provides limited capabilities to allow automated execution, model transformation, or support for user-defined logic and abstractions that limit its compatibility with modern workflows. Here, to better integrate MCNP into a modern scientific workflow, we have developed an intuitive yet full-featured MCNP Application Program Interface (API) in Python, named MCNPy, which provides a specialized set of classes for MCNP input development. Moreover, to guarantee that our reading, writing, and modeling capabilities remain self-consistent (and to render the huge scope of the MCNP API manageable), we have adopted a strategy of model-driven software development in which a generalized model of the MCNP input format has been created. From this generalized model, or “metamodel,” problem-specific implementations such as an engine for input validation or a codebase for programmatic operations may be automatically generated. Since MCNPy primarily acts as a Python front-end to the underlying Java API that directly interfaces with the metamodel, it is intrinsically linked to the metamodel and thus remains maintainable. With MCNPy, users can programmatically read, write, and modify any syntactically valid MCNP input file regardless of its origin. These capabilities allow users to automate complicated tasks like design optimization and model translation for nuclear systems. As examples, this work demonstrates the use of MCNPy to find the critical radius of a plutonium sphere and to translate a 9000+ line MCNP input file into a corresponding OpenMC model.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2006

The FLUKA Code: An Overview

FLUKA is a multipurpose Monte Carlo code which can transport a variety of particles over a wide energy range in complex geometries. The code is a joint project of INFN and CERN: part of its development is also supported by the University of Houston and NASA. FLUKA is successfully applied in several fields, including but not only, particle physics, cosmic ray physics, dosimetry, radioprotection, hadron therapy, space radiation, accelerator design and neutronics. The code is the standard tool used at CERN for dosimetry, radioprotection and beam-machine interaction studies. Here we give a glimpse into the code physics models with a particular emphasis to the hadronic and nuclear sector.

Ballarini, F.↗
DOE OSTI2015

GENIE event generator

GENIE is an international collaboration of physicists and programmers who develop Monte Carlo codes for neutrino-nucleus simulation. It is applicable to any neutrino experiment using beam energies between 10 MeV and 10 TeV. This short document describes recent changes in process and release schedule. The new model developments for v2.9.0 and v2.9.2 are briefly described.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

TU-H-CAMPUS-IeP1-02: Validation of a CT Monte Carlo Software

Purpose: To validate the in-house developed CT Monte Carlo calculation tool GMctdospp against reference simulation sets provided by the AAPM in the new report 195. Methods: Deposited energy was calculated in four segments (test 1) and two 10 cm long cylinders (test 2) inside a CTDI phantom (following case #4 of the AAPM report 195). The x-ray point source of a given 120 kVp spectrum was collimated to a fan beam with two thicknesses (10 mm, 80 mm) for a static and a rotational setup. In addition, a given chest geometry was used to calculate deposited energy in several organs for a 0° static and a rotational beam (following case #5 of the AAPM report 195). The results of GMctdospp were compared against the particular mean value of the four quoted Monte Carlo codes (EGSnrc, Geant 4, MCNP and Penelope). Results: Calculated values showed no outliers in any of the cases. Differences between GMctdospp and the particular mean Results: Calculated values showed no outliers in any of the cases. Differences between GMctdospp and the particular mean value were always at similar magnitude compared to the quoted codes. For case #4 (CTDI phantom) the relative differences were within 1.5 %, on average 0.4 % and for case #5 (chest phantom) within 2.5 % and on average 0.85 %. Conclusion: The results confirmed an overall uncertainty of the Monte-Carlo calculation chain in GMctdospp being <2.5 %, for most cases even better. This can be considered small compared to other sources of uncertainties, e.g. virtual source and patient models. The photon transport implemented in GMctdospp inside a voxel-based patient geometry was successfully verified.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Signal pulse emulation for scintillation detectors using Geant4 Monte Carlo with light tracking simulation

The anode pulse of a photomultiplier tube (PMT) coupled with a scintillator is used for pulse shape discrimination (PSD) analysis. We have developed a novel emulation technique for the PMT anode pulse based on optical photon transport and a PMT response function. The photon transport was calculated using Geant4 Monte Carlo code and the response function with a BC408 organic scintillator. The obtained percentage RMS value of the difference between the measured and simulated pulse with suitable scintillation properties using GSO:Ce (0.4, 1.0, 1.5 mol%), LaBr{sub 3}:Ce and BGO scintillators were 2.41%, 2.58%, 2.16%, 2.01%, and 3.32%, respectively. The proposed technique demonstrates high reproducibility of the measured pulse and can be applied to simulation studies of various radiation measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Monte Carlo modeling and simulations of the High Definition (HD120) micro MLC and validation against measurements for a 6 MV beam

Purpose: The most recent Varian micro multileaf collimator (MLC), the High Definition (HD120) MLC, was modeled using the BEAMNRC Monte Carlo code. This model was incorporated into a Varian medical linear accelerator, for a 6 MV beam, in static and dynamic mode. The model was validated by comparing simulated profiles with measurements. Methods: The Varian Trilogy (2300C/D) accelerator model was accurately implemented using the state-of-the-art Monte Carlo simulation program BEAMNRC and validated against off-axis and depth dose profiles measured using ionization chambers, by adjusting the energy and the full width at half maximum (FWHM) of the initial electron beam. The HD120 MLC was modeled by developing a new BEAMNRC component module (CM), designated HDMLC, adapting the available DYNVMLC CM and incorporating the specific characteristics of this new micro MLC. The leaf dimensions were provided by the manufacturer. The geometry was visualized by tracing particles through the CM and recording their position when a leaf boundary is crossed. The leaf material density and abutting air gap between leaves were adjusted in order to obtain a good agreement between the simulated leakage profiles and EBT2 film measurements performed in a solid water phantom. To validate the HDMLC implementation, additional MLC static patterns were also simulated and compared to additional measurements. Furthermore, the ability to simulate dynamic MLC fields was implemented in the HDMLC CM. The simulation results of these fields were compared with EBT2 film measurements performed in a solid water phantom. Results: Overall, the discrepancies, with and without MLC, between the opened field simulations and the measurements using ionization chambers in a water phantom, for the off-axis profiles are below 2% and in depth-dose profiles are below 2% after the maximum dose depth and below 4% in the build-up region. On the conditions of these simulations, this tungsten-based MLC has a density of 18.7 g cm{sup -3} and an overall leakage of about 1.1 {+-} 0.03%. The discrepancies between the film measured and simulated closed and blocked fields are below 2% and 8%, respectively. Other measurements were performed for alternated leaf patterns and the agreement is satisfactory (to within 4%). The dynamic mode for this MLC was implemented and the discrepancies between film measurements and simulations are within 4%. Conclusions: The Varian Trilogy (2300 C/D) linear accelerator including the HD120 MLC was successfully modeled and simulated using the Monte Carlo BEAMNRC code by developing an independent CM, the HDMLC CM, either in static and dynamic modes.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2011

MCNP Developments [Slides]

This presentation covers recent progress in development and support of the MCNP Monte Carlo code during FY 2011 in support of the US DOE/NNSA Nuclear Criticality Safety Program.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Development of random geometry capability in RMC code for explicit analysis of polytype particle/pebble - 18493

Monte Carlo method plays an important role in accurate simulation of High Temperature Gas-Cooled Reactors (HTR) and molten salt reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the innovative design of pebble bed, prism HTRs and molten salt reactors, there may be more than one kind of particles or pebbles in the system, such as the burnable poison particles, graphite pebbles or fuel pebbles with different enrichments. Different treatments have been used to deal with the double-heterogeneous geometry of these reactors. In some researches, the particle distributions in stochastic media are assumed the regular lattice, while in the others, the explicit packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) are used, by generating the random coordinates of the particles/pebbles. However, traditional geometry with repeated structure has difficulties in explicit modeling polytype particle/pebble, due to the arbitrary packing fractions of different kinds of particles or pebbles. Although, packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) could solve the polytype particle/pebble problems to some extent, but these methods are troublesome in geometry modeling and time consuming. In this paper, a new Random Universe Geometry method was developed and implemented in Monte Carlo code RMC, for explicit analysis of polytype particle/pebble in double heterogeneous geometry systems, such as HTRs and molten salt reactors. This method was applied to constructed mixed particles/pebbles problems and also simplified HTR-10 full core problem to test the feasibility, and the results show that the random geometry modeling can provide explicit packing fraction of different kinds of particles or pebbles. The results are also compared with that calculated by RSA method and regular structure, and agree quite well. This work would provide an effective tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗