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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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This report provides a brief look at XCP Computational Physics Student Summer Workshop with research conclusions and ideas for future work.
This work extends the application of functional expansion tallies to 3D spherical geometries. The 3D Zernike polynomials are set as an orthonormal polynomials basis for the functional reconstruction. The study describes the construction of the complete set of polynomials, a natural expansion of the spherical harmonics polynomials where 3D Zernike moments can be evaluated as a linear combination of the geometrical moments. The 3D Zernike polynomials formulation and the computational approach implemented in Serpent 2 are presented and tested through the Godiva model from the ICSBEP criticality benchmark test cases. The implementation results are in agreement with a reference solution described in a fine-resolution mesh, enhancing also the performance and memory demand. (authors)
Not Available
Abstract Binary Collision Approximation (BCA) codes for ion-material interactions, such as SRIM, Tridyn, F-TRIDYN, and SDtrimSP, have historically been limited to screened Coulomb potentials even at low energies due to the difficulty in numerically solving the Distance of Closest Approach (DOCA) problem for attractive-repulsive potentials. Techniques such as direct n-body simulation or modifications to Newton’s method are either prohibitively costly or not guaranteed to work for all potentials. Advanced rootfinding techniques, such as companion matrix solvers, offer a solution. For many attractive-repulsive potentials, however, a companion matrix cannot be used directly, because there is no way to put the associated functions into a monomial basis form. A complementary technique is proxy rootfinding—by finding the best-fit polynomial approximant of a function, the zeros of the approximant can be guaranteed to be close to the zeros of the function. Using the Chebyshev basis and grid offers additional guarantees with regards to the quality of the approximation, the speed of convergence, and the avoidance of Runge’s phenomenon. By finding Chebyshev interpolants and using the Chebyshev-Frobenius companion matrix, the zeros of any real function on a bounded domain can be found. Here we show that using an Adaptive Chebyshev Proxy Rootfinder with Automatic Subdivision (ACPRAS) with appropriate scaling functions, numerical issues presented by attractive-repulsive potentials, including those of scale, can be handled. Using these techniques, we show that it is possible to include any physically reasonable interatomic potential in a BCA code, and to guarantee correctness of the resulting scattering angle calculations.
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The development of accident-tolerant fuel and advanced reactors fueled with high-assay low-enriched uranium has created an increased focus on systems with uranium enrichments between 5 and 20 wt % 235 U. SCALE criticality safety validation reports generated at Oak Ridge National Laboratory contain very few benchmark experiments in this range. The work presented herein is an initial investigation of expanded validation of the entire range of enrichments for uranium-fueled systems, including consideration of over 300 newly developed KENO models. The vast majority of the new models are between 4.4 wt % and 37.5 wt % enrichment and include a variety of thermal, intermediate, and mixed neutron spectra. This work finds no strong evidence of an enrichment-dependent bias, but there are indications of a spectrum-dependent bias. There are, however, only a limited number of independent experiments included at this time to distinguish intermediate enrichment effects from intermediate spectrum effects. Further work in this area is ongoing for intermediate enrichment systems. In the long run, the addition of these models to the Verified, Archived Library and Inputs and Data is planned, as funding and staff availability allow.
An overview of the MCATK algorithm is provided, along with other areas of interest to Los Alamos National Laboratory.
There is likely a trend in bias for uranium-fueled systems as a function of spectrum, but probably not as a function of enrichment. Some enrichment trends appear, but are potentially the result of outlier experimental results; further investigation is needed where possible. There are good examples in this data of markedly different biases from different systems. Future work includes review and vetting for inclusion in the VALID library and analysis with TSUNAMI.
Abstract not provided.
The thermal radiative transfer equations (TRT) model many important phenomena of interest at LANL. The TRT equations model time-dependent thermal radiation through a medium, i.e., how thermal radiation loses energy by absorption, gains energy via emission processes, and redistributes energy via scattering.
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The TVA WB1 benchmark problem was precisely modeled in accordance with the geometry and material specifications, and benchmark exercise 1 was solved with the MVP code and the nuclear data libraries based on JENDL-4.0u (JENDL-4.0 updated file) and ENDF/B-VII.0. The criticality, the control rod bank worth, the isothermal temperature coefficient (ITC), the integral rod worth of bank D, and differential soluble boron worth (DBW) were obtained. The criticality and the control rod bank worth were compared with the measured data and published calculation results with other codes, Serpent and KENO. The criticality and the control rod bank worth calculated by MVP and the libraries agree with the measured and the published calculation results within the criteria specified in the ANSI/ANS standard. From these results, the calculation model for the MVP and the libraries was verified. The other parameters such as ITC, the integral rod worth, and DBW were also compared between JENDL-4.0u and ENDF/B-VII.0. The comparison of the results shows JENDL-4.0u and ENDF/B-VII.0 give comparable results within the standard deviation, the difference between libraries is small. (author)
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Monte Carlo criticality transport codes, which rely on the power iteration procedure, are a fundamental tool for nuclear criticality safety practitioners in assessing the neutron multiplication factor (k eff ) for problems involving fissile material. In these calculations, ensuring the convergence of both the fission source distributions and the k eff estimate for accurate results is crucial. However, a converged k eff estimate does not necessarily mean the fission source distribution is also converged because the fission source and flux distribution may continue to evolve even after k eff convergence. Therefore, most Monte Carlo transport criticality codes now offer various diagnostic tests to assess fission source convergence in addition to the k eff convergence by analyzing the trends of these quantities over multiple generations.
Equivalent point source energy emission distributions were computed for various threat sources for use in simulating the detector responses for urban search scenarios. The sources include standard isotopic sources used in detector testing, medical and industrial sources occasionally encountered in urban searches, and several types of special nuclear materials. Most of the equivalent point source distributions represent small sources inside some amount of shielding, but the special nuclear material sources represent volumetrically distributed sources in spheres of metal. Text-based inputs for emission distributions are available for the Monte Carlo transport codes Monte Carlo N-Particle, SCALE/MAVRIC, and Omnibus/Shift, any of which can easily be converted to other formats. These sources were developed for use in the Radiological Anomaly Detection and Identification (RADAI) project and the follow-on project, the RADAI-Extended project, sponsored by the National Nuclear Security Administration Office of Defense Nuclear Nonproliferation Research and Development.
COG is a high-resolution code for the Monte Carlo simulation of coupled particle transport in arbitrary 3-D geometry. COG will transport neutrons, protons, deuterons, alpha particles with energies up to hundreds of GeV, and photons with energy ranges limited by the available cross section sets and physics models. Electrons can be transported via the EGS5 electron transport kernel, electrons can also be transported. The COG code is a significant upgrade from earlier Monte Carlo transport codes and has been written specifically to make it more versatile, accurate, and easy to use. COG has provisions for calculating deep penetration (shielding) problems, criticality problems, and neutron activation problems while retains all of the standard capabilities found in other Monte Carlo transport codes. COG uses high-resolution pointwise cross-section databases and makes no compromises in the transport physics, so that the results of a COG run are limited only by the accuracy of the databases used. COG runs primarily on Linux Operating System workstations with MPICH software installed – currently, Red Hat 7 & 8, Windows 10 (Windows Subsystem for Linux –WSL), Ubuntu 16, 18 & 20, OpenSUSE Leap 15.2, Fedora 32, Apple Power Mac with Intel CPU (with MacPorts installed) workstations, and LLNL LC supercomputer CTS-1 cluster with TOSS 3 are supported.
The Monte Carlo Transport Project at Lawrence Livermore National Laboratory develops two Monte Carlo transport codes used in production by a sizable internal user community. Mercury is a Monte Carlo particle transport code used to model the interaction of neutrons, gammas, and light ions with a material. Imp is an implicit Monte Carlo thermal x-ray photon transport code used to model the interaction of x-ray photons with a material. This paper describes the two codes and highlights recent developments.
The OpenMC Monte Carlo particle transport code has been continuously developed for 13 years by a large community of contributors. In that time span, the codebase has undergone significant changes that have redefined what OpenMC is and made it an enduring presence in the nuclear science and engineering community. In this paper, we discuss the evolution of programming language use in OpenMC, trends in the overall design of the programming interfaces, and implications for the future of the code.