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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Release of MCNP6.2 and WHISPER-1.1 - Guidance for NCS users

In 2017, a new version of the MCNP6 Monte Carlo code package was released: MCNP6.2. This is the first version of MCNP to include the WHISPER package to support sensitivity-uncertainty based nuclear criticality safety validation. The coding changes to MCNP6.2 physics are relatively insignificant. Corrections to the S(α,β) thermal scattering numerics are generally negligible relative to problem statistics (or, in rare cases, prevent aborts). Similarly, the changes to adjoint-weighting for computing kinetics parameters may result in small differences, generally negligible compared to problem statistics. The change to the MCNP6.2 geometry treatment to correctly handle coincident surfaces in problems with universe/fill features will produce different round-off in the geometry tracking. This will produce differences in results relative to previous versions, but those differences should be small relative to problem statistics, and are not a concern. An immediate benefit of using MCNP6.2 (rather than MCNP6.1) is that the new version is typically 1.5-2 times faster.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Technical Note: Improved implementation of doppler broadening in MCNP5

Purpose: Incoherent scattering has a substantial effect on spectroscopic measurements and simulations. Many general-purpose Monte Carlo codes include models that account for the effects of bound electrons on incoherent scattering, including Doppler broadening (DB). This work investigates the DB model used in the Monte Carlo N-particle transport code (MCNP5). Methods: Simulations were run with three versions of MCNP5: v1.51, v1.60, and a modified form of v1.60 (v1.60m). All simulations used the MCPLIB04 photon data library, which presents the electron subshell data for incoherent scattering in the form of a probability density function. In v1.60m, the source code was altered to sample the electron subshell from a cumulative density function instead. Each version of the code was tested using an identical set of simulations that investigated DB in a slab of silicon at scattering angles of 15 Degree-Sign , 30 Degree-Sign , and 45 Degree-Sign . For each angle, simulations were run for multiple energies between 200 keV and 800 keV. The spectrum of singly-scattered photons at the exit of the slab was scored. Spectra were analytically calculated for comparison. Results: In v1.51, DB was modeled for incident photon energies below 760 keV, 384 keV, and 260 keV at scattering angles of 15 Degree-Sign , 30 Degree-Sign , and 45 Degree-Sign , respectively. Above these energy thresholds, v1.51 did not model DB. The spectra calculated using v1.60 and v1.60m exhibited DB for all energy-angle combinations; however, v1.60m, exhibited more energy broadening than did v1.60. The spectra calculated with v1.60m agreed with the analytical calculations. Conclusions: MCNP5 v1.51 and v1.60 model partial broadening when used with the MCPLIB04 data library. MCNP5 v1.60m models DB more accurately due to the form of the electron subshell data. In response to these results, Los Alamos National Laboratory has released a new photon data library, MCPLIB84, that presents the electron subshell data in cumulative distribution form. MCNP5 v1.60 should be used with this library when incoherent scattering has a significant impact on simulation results.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Monte-Carlo fluid approaches to detached plasmas in non-axisymmetric divertor configurations

Fluid transport modeling in three-dimensional boundaries of toroidal confinement devices is reviewed with the emphasis on a Monte-Carlo approach to simulate detached plasmas. The loss of axisymmetry in such configurations presents a major challenge for numerical implementation of the standard fluid model widely applied to fusion experimental devices. A large-scale effort has been made to address this problem under complementary aspects including different magnetic topologies and numerical techniques. Here in this paper, we give a brief review of the different strategies pioneered and the challenges involved. A more detailed description is provided for the Monte-Carlo code—EMC3-Eirene, where the physics model and the basic idea behind the applied Monte-Carlo method are presented. The focus is put on its applications to detachment studies for stellarators and tokamaks. Here, major achievements and difficulties encountered are described. Model limitations and further development plans are discussed.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

A High-Order Low-Order Algorithm with Exponentially Convergent Monte Carlo for Thermal Radiative Transfer

In this paper, we have implemented a new high-order low-order (HOLO) algorithm for solving thermal radiative transfer problems. The low-order (LO) system is based on the spatial and angular moments of the transport equation and a linear-discontinuous finite-element spatial representation, producing equations similar to the standard S 2 equations. The LO solver is fully implicit in time and efficiently resolves the nonlinear temperature dependence at each time step. The high-order (HO) solver utilizes exponentially convergent Monte Carlo (ECMC) to give a globally accurate solution for the angular intensity to a fixed-source pure-absorber transport problem. This global solution is used to compute consistency terms, which require the HO and LO solutions to converge toward the same solution. The use of ECMC allows for the efficient reduction of statistical noise in the Monte Carlo solution, reducing inaccuracies introduced through the LO consistency terms. Finally, we compare results with an implicit Monte Carlo code for one-dimensional gray test problems and demonstrate the efficiency of ECMC over standard Monte Carlo in this HOLO algorithm.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Status of the OECD/NEA Watts Bar unit 1 benchmark and calculations of local reactor core power data of the Zero Power physics tests

The paper aims to provide an update on the status of the Organization for Economic Cooperation and Development (OECD) / the Nuclear Energy Agency (NEA) Tennessee Valley Authority (TVA) Watts Bar 1 (WB1) Multi-Physics Multi-Cycle depletion benchmark and the results associated with Exercise 1. The benchmark relies on the set of benchmark progression problems developed by the Department of Energy (DOE) Consortium for the Advanced Simulation of Light Water Reactors (CASL) for the Virtual Environment for Reactor (VERA), which are based on real plant design and operational data. The OECD/NEA TVA WB1 benchmark is designed for validation of both traditional and novel high-fidelity multi-physics codes to analyze Pressurized Water Reactors (PWR) depletion cycles. The activities are conducted under the Expert Group on Reactor Systems Multi-Physics (EGMUP) at NEA/OECD. In this work, we analyze Exercise 1 of the benchmark: stand-alone Three-Dimensional (3D) neutronics at Start-up Zero Power Physics Test (ZPPT) at Hot Zero Power Conditions (HZP) and the resulting power maps. The code used to model the exercise is the continuous energy Monte Carlo code Serpent 2.1.31 along with the nuclear data library ENDF/B-VII.1. Serpent results are compared with the high-fidelity deterministic code MPACT, which is part of VERA. The paper presents three selected results from power calculations required output. The results compared are the normalized axially integrated radial core power maps, the normalized axial averaged core power shapes, and the normalized core hottest and coldest assemblies radial power maps. The Root Mean Square Deviation (RMSD) between Serpent and MPACT is 0.51 % for the axially integrated radial core power map, 1.41 % for the axial averaged core power shape, 2.42 % for the hottest assembly axial power shape, 0.88 % for the coldest assembly axial integrated hottest assembly radial pin power map, and 0.17 % for the axially integrated coldest assembly radial pin power map. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Shutdown dose rate analysis with the Shift Monte Carlo radiation transport code and modular verification workflow

Calculation of the shutdown dose rate is crucial for safe fusion reactor operations. The Rigorous-two-step (R2S) method is a method that requires connected neutron transport, activation, and gamma transport. Shift has integrated variance reduction with a deterministic solver Denovo, supports multiple geometry formats, and is scalable. These features make it an attractive transport solver choice for an R2S workflow. An R2S workflow for the Shift Monte Carlo code is developed and compared to the existing Oak Ridge National Laboratory Shutdown Dose Rate Code Suite (ORCS) workflow. Also, a Python framework for integrating two R2S workflows is developed to mix and match each step in the R2S workflow for improved collaboration and verification experience. In this study, results show that the Shift-Denovo R2S workflow and the ORCS workflow calculate the shutdown dose rate of the ITER Shutdown Dose Rate benchmark problem with an average relative error of 2.285%.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Nuclear Data and Cross Section Testing Using ENDF/B-VIII.0

With the release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library, nuclear criticality safety practitioners and engineers have access to the latest cross section sets available for their analyses. However, these cross sections must be rigorously tested and validated to ensure that the nuclear data are responsive to the needs of the individuals responsible for developing, implementing, and maintaining computational tools for criticality safety applications. Thus, the ENDF/B-VIII.0 library is tested and validated with a large collection of experiments that were vetted by the International Criticality Safety Benchmark Evaluation Project and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. A selection of benchmark experiments for use within the criticality safety community were prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Nuclear Energy and Fuel Cycle Division at Oak Ridge National Laboratory. The performance of the ENDF/B-VIII.0 library is assessed by using VALID models of benchmark experiments with the beta 12 version of SCALE 6.3 KENO V.a and KENO-VI Monte Carlo codes. The performance is compared with the results obtained from with the ENDF/B-VII.1 library. This report considers multigroup (MG) and continuous energy (CE) formats of the ENDF/B-VIII.0 and -VII.1 libraries. The benchmark experiments within VALID that validate the ENDF/B-VIII.0 library cover 15 broad system categories by using a range of fissile materials, uranium enrichments, plutonium isotopic vectors, and mixed uranium/plutonium systems. These forms are represented as metals, solutions, or various arrays of rods or plates that cover a variety of neutron energy spectra: thermal, fast, mixed, and intermediate. Over 600 cases were considered for use with the KENO V.a and KENO-VI codes with the ENDF/B-VIII.0 library. The results of the Monte Carlo comparison of ENDF/B-VIII.0 to ENDF/B-VII.1 with both KENO V.a and KENO-VI indicate that there is a less than 0.53% Δk difference between the bias of calculated k eff from the expected values. The CE ENDF/B-VIII.0 library results in smaller magnitude biases than the ENDF/B-VII.1 data for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-SOL-THERM, PU-SOL-THERM, and U233-MET-FAST systems, while the MG results yielded smaller magnitude biases for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-COMP-THERM, LEU-SOL THERM, MIX-COMP-FAST, and U233-MET-FAST systems. Most notable are the adjustments to the plutonium and 233 U cross section data, which has resulted in noticeably lower biases in the ENDF/B VIII.0 results for the mixed, plutonium, and 233 U systems. Results of the sensitivity data file comparison generated from TSUNAMI-3D for selected VALID cases for the ENDF/B-VIII.0 library indicate a very high level of agreement with correlation coefficients of the effect of nuclear data uncertainty on k eff (the c k integral parameter) all above 0.99. This indicates that cases with the ENDF/B-VIII.0 library would see very similar responses to any nuclear data errors or change as those with the ENDF/B-VII.1 library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

New capabilities of the MORET 6 Monte Carlo neutron transport code

The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations in a user-friendly way. New features have been introduced to extend the application field of MORET beyond the usual criticality calculations for which it has been initially designed. The most important change is the addition of an analog fixed source mode which allows studies of systems of any reactivity combined with very flexible outputs. Other useful improvements have been added concerning the geometric part, the fission matrix, the multigroup sensitivity coefficients and the outputs. This paper presents an overview of these new features. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2006

Lunar Surface Reactor Shielding Study

Nuclear reactor system could provide power to support a long term human exploration to the moon. Such a system would require shielding to protect astronauts from its emitted radiations. Shielding studies have been performed for a Gas Cooled Reactor (GCR) system because it is considered to be the most suitable nuclear reactor system available for lunar exploration, based on its tolerance of oxidizing lunar regolith and its good conversion efficiency (Wright, 2003). The goals of the shielding studies were to provide optimal material shielding configuration that reduces the dose (rem) to the required level in order to protect astronauts, and to estimate the mass of regolith that would provide an equivalent protective effect if it were used as the shielding material. All calculations were performed using MCNPX code, a Monte Carlo transport code.

Monte Carlo N-Particle eXtended (MCNPX)↗
DOE OSTI2015

SU-E-T-556: Monte Carlo Generated Dose Distributions for Orbital Irradiation Using a Single Anterior-Posterior Electron Beam and a Hanging Lens Shield

Purpose: Various cancers of the eye are successfully treated with radiotherapy utilizing one anterior-posterior (A/P) beam that encompasses the entire content of the orbit. In such cases, a hanging lens shield can be used to spare dose to the radiosensitive lens of the eye to prevent cataracts. Methods: This research focused on Monte Carlo characterization of dose distributions resulting from a single A-P field to the orbit with a hanging shield in place. Monte Carlo codes were developed which calculated dose distributions for various electron radiation energies, hanging lens shield radii, shield heights above the eye, and beam spoiler configurations. Film dosimetry was used to benchmark the coding to ensure it was calculating relative dose accurately. Results: The Monte Carlo dose calculations indicated that lateral and depth dose profiles are insensitive to changes in shield height and electron beam energy. Dose deposition was sensitive to shield radius and beam spoiler composition and height above the eye. Conclusion: The use of a single A/P electron beam to treat cancers of the eye while maintaining adequate lens sparing is feasible. Shield radius should be customized to have the same radius as the patient’s lens. A beam spoiler should be used if it is desired to substantially dose the eye tissues lying posterior to the lens in the shadow of the lens shield. The compromise between lens sparing and dose to diseased tissues surrounding the lens can be modulated by varying the beam spoiler thickness, spoiler material composition, and spoiler height above the eye. The sparing ratio is a metric that can be used to evaluate the compromise between lens sparing and dose to surrounding tissues. The higher the ratio, the more dose received by the tissues immediately posterior to the lens relative to the dose received by the lens.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

IRSN working program status on tools for evaluation of SFR cores static neutronics safety parameters

As technical support of the French Nuclear Safety Authority, IRSN will be in charge of safety assessment of any future project of Sodium Fast Reactor (SFR) that could be built in France. One of the main safety topics will deal with reactivity control. Since the design and safety assessment of the last two SFR plants in France (Phenix and Superphenix, more than thirty years ago), methods, codes and safety objectives have evolved. That is why a working program on core neutronic simulations has been launched in order to be able to evaluate accuracy of future core characteristics computations. The first step consists in getting experienced with the ERANOS well-known deterministic code used in the past for Phenix and Superphenix. Then Monte-Carlo codes have been tested to help in the interpretation of ERANOS results and to define what place this kind of codes can have in a new SFR safety demonstration. This experience is based on open benchmark computations. Different cases are chosen to cover a wide range of configurations. The paper shows, as an example, criticality results obtained with ERANOS, SCALE and MORET, and the first conclusions based on these results. In the future, this work will be extended to other safety parameters such as sodium void and Doppler effects, kinetic parameters or flux distributions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

String model for spinning quark jets

A string model of quark hadronization, taking the quark spin degree of freedom into account, is proposed. The method for using the model in a Monte-Carlo code for jet generation is given.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2014

The FLUKA Code: Developments and Challenges for High Energy and Medical Applications

The FLUKA Monte Carlo code is used extensively at CERN for all beam-machine interactions, radioprotection calculations and facility design of forthcoming projects. Such needs require the code to be consistently reliable over the entire energy range (from MeV to TeV) for all projectiles (full suite of elementary particles and heavy ions). Outside CERN, among various applications worldwide, FLUKA serves as a core tool for the HIT and CNAO hadron-therapy facilities in Europe. Therefore, medical applications further impose stringent requirements in terms of reliability and predictive power, which demands constant refinement of sophisticated nuclear models and continuous code improvement. Some of the latest developments implemented in FLUKA are presented in this paper, with particular emphasis on issues and concerns pertaining to CERN and medical applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User's Guide

The ability to perform sensitivity analyses using adjoint-based sensitivity theory has existed for decades . This paper provides guidance on how adjoint sensitivity methods can be used to predict the effect of material density and composition uncertainties in critical experiments, including when these uncertain parameters are correlated or constrained. Two widely used Monte Carlo codes, MCNP6 and SCALE, are both capable of computing isotopic sensitivities in continuous energy and angle. Additionally, Perko et al. have shown how individual isotope sensitivities, easily computed using adjoint methods, can be combined to compute constrained sensitivities that may be used in the uncertainty analysis. This paper provides details on how the codes are used to compute sensitivities and how the sensitivities are used in an uncertainty analysis. Constrained sensitivities are computed in a simple example problem. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Critical and subcritical mass computations for {sup 247}Cm - 14378

Experimental nuclear criticality data are unavailable for many special actinides so subcritical limits are established using computations and conservative practices. Computations are done using codes in the SCALE6.1 system and the MCNP6, version 1.0, Monte Carlo code with cross sections based on Evaluated Nuclear Data File (ENDF) B-VI, B-VII.0, and B-VII.1 data. Both SCALE multigroup and MCNP continuous energy cross sections are used to estimate critical masses of unreflected, stainless steel reflected, and water reflected spherical {sup 247}Cm metal/water systems. For well moderated systems the critical masses derived using ENDF/B-VII data are about half those previously reported. The difference appears to be caused by changes in the ENDF/B-VII fission cross sections and an overall increase in the value of ν-bar. Subcritical mass limits are derived based on the critical mass estimates. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Collision of Physics and Software in the Monte Carlo Application Toolkit (MCATK)

The topic is presented in a series of slides organized as follows: MCATK overview, development strategy, available algorithms, problem modeling (sources, geometry, data, tallies), parallelism, miscellaneous tools/features, example MCATK application, recent areas of research, and summary and future work. MCATK is a C++ component-based Monte Carlo neutron-gamma transport software library with continuous energy neutron and photon transport. Designed to build specialized applications and to provide new functionality in existing general-purpose Monte Carlo codes like MCNP, it reads ACE formatted nuclear data generated by NJOY. The motivation behind MCATK was to reduce costs. MCATK physics involves continuous energy neutron & gamma transport with multi-temperature treatment, static eigenvalue (k eff and α) algorithms, time-dependent algorithm, and fission chain algorithms. MCATK geometry includes mesh geometries and solid body geometries. MCATK provides verified, unit-test Monte Carlo components, flexibility in Monte Carlo application development, and numerous tools such as geometry and cross section plotters.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗