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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 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Source Discretization Near Curvilinear Mesh Refinements in One Dimeions (Summer Internship Final Report)

Lawrence Livermore National Lab developed the seismic wave code SW4 for earthquake simulations, and has recently added the ability to specify mesh-refinement interfaces in the curvilinear mesh. The primary purpose of this is to allow for high resolution near the surface, with relatively fewer total points in the mesh leading to shorter computation time. Presently, the code does not allow for point-source forcing terms (commonly used in seismic simulations) to be located near these curvilinear mesh refinement interfaces. The goal of this internship was to investigate the discretization of point-source forcing terms in the seismic wave equation when the sources are located near the interfaces between two curvilinear meshes. Ultimately this will be incorporated into SW4 to allow for more realistic simulations to be carried out in this faster implementation.

97 MATHEMATICS AND COMPUTING↗
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 OSTI2023

Monte Carlo Radiation Transport for Astrophysical Transients Powered by Circumstellar Interaction

In this paper, we introduce SuperLite, an open-source Monte Carlo radiation transport code designed to produce synthetic spectra for astrophysical transient phenomena affected by circumstellar interaction. SuperLite utilizes Monte Carlo methods for semi-implicit, semirelativistic radiation transport in high-velocity shocked outflows, employing multigroup structured opacity calculations. The code enables rapid post-processing of hydrodynamic profiles to generate high-quality spectra that can be compared with observations of transient events, including superluminous supernovae, pulsational pair-instability supernovae, and other peculiar transients. We present the methods employed in SuperLite and compare the code's performance to that of other radiative transport codes, such as SuperNu and CMFGEN. We show that SuperLite has successfully passed standard Monte Carlo radiation transport tests and can reproduce spectra of typical supernovae of Type Ia, Type IIP, and Type IIn.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2013

Benchmarking the New RESRAD-OFFSITE Source Term Model with DUST-MS and GoldSim - 13377

RESRAD-OFFSITE is a computer code developed by Argonne National Laboratory under the sponsorship of U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC). It is designed on the basis of RESRAD (onsite) code, a computer code designated by DOE and NRC for evaluating soil-contaminated sites for compliance with human health protection requirements pertaining to license termination or environmental remediation. RESRAD-OFFSITE has enhanced capabilities of modeling radionuclide transport to offsite locations and calculating potential radiation exposure to offsite receptors. Recently, a new source term model was incorporated into RESRAD-OFFSITE to enhance its capability further. This new source term model allows simulation of radionuclide releases from different waste forms, in addition to the soil sources originally considered in RESRAD (onsite) and RESRAD-OFFSITE codes. With this new source term model, a variety of applications can be achieved by using RESRAD-OFFSITE, including but not limited to, assessing the performance of radioactive waste disposal facilities. This paper presents the comparison of radionuclide release rates calculated by the new source term model of RESRAD-OFFSITE versus those calculated by DUST-MS and GoldSim, respectively. The focus of comparison is on the release rates of radionuclides from the bottom of the contaminated zone that was assumed to contain radioactive source materials buried in soil. The transport of released contaminants outside of the primary contaminated zone is beyond the scope of this paper. Overall, the agreement between the RESRAD-OFFSITE results and the DUST-MS and GoldSim results is fairly good, with all three codes predicting identical or similar radionuclide release profiles over time. Numerical dispersion in the DUST-MS and GoldSim results was identified as potentially contributing to the disagreement in the release rates. In general, greater discrepancy in the release rates was found for short-lived, fast-moving radionuclides than for long-lived, slow-moving radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Client-side and Server-side Programming in Sirepo Framework

Sirepo [1] is an interactive simulation environment for accelerator and beamline design. It is integrated with numbers of widely used scientific codes, including Synchrotron Radiation Workshop (SRW) [2] for accurate X–ray source and optics simulations. Sirepo is a fully open source and could be quickly extended to support the codes from different scientific domains. Sirepo framework covers both client and server–side simulation job management, user management and front–end data presentation. In this present note, we extend Sirepo by adding a new SRW source type for the beamline simulations.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Uniform and Half‐Space Models for Magnetic Dipole Sources in a Conducting Medium: I. Theory and Results

Underground nuclear events result in electromagnetic signals near the event location. Of particular interest, because its signal can be potentially enhanced by the use of large cavities to decouple the seismic signal, is the generation of an EMP signal within the nuclear detonation (NUDET) cavity, from the formation of a bubble in the local magnetic field. This mechanism is equivalent to a buried time varying magnetic dipole. In order to model such signals Los Alamos National Laboratory is developing a detailed model for the propagation of RF signals from buried sources to nearby sensors, with a focus on sources that are magnetic dipole in character. In order to validate this code, models for simple geometries involving magnetic dipole sources have been implemented. These models represent the fields from vertical and horizontal dipoles in cylindrical coordinates, either in a uniform linear isotropic electrical medium, or in a space consisting of two half spaces interfacing at the plane z=0, with the upper half space having the electrical properties of free space and the lower half space having arbitrary linear conductivity and electrical permittivity. While primarily intended for the validation of the more detailed model, the shorter turnaround time of these models facilitates the study of the general properties of geomagnetic bubble signals. This report summarizes the theory and example results for those models. The use of the model for model validation is illustrated with the near and surface fields for dipoles with Gaussian time dependence. Also examined are the surface fields expected for the geomagnetic bubble predicted by the numerical model of Kovalenko et al for a 1 kt nuclear detonation. The model of Kovalenko et al was also rescaled to allow comparison with Zablocki’s electric field observations for Bilby, Event III, and Event V. While the rescaled Kovalenko et al model had approximately the right time scale for the duration of the main peak, and a slower fall than rise also in qualitative agreement with the observations, the predicted signals had very different polarizations, were too low by more than an order of magnitude, and had less temporal structure. A simple comparison with the vertical B field data of Sweeney, where the detailed event geometries and yields are classified, but the sensor range and bounds on the yields are given, suggest that the scaled Kovalenko et al model gives too small a peak field compared to observations for that data set as well, except perhaps for the Borate event. A companion report will give a detailed discussion of the implementation and code.

58 GEOSCIENCES↗
DOE OSTI2026

Polarized Deep-Inelastic Scattering with Spin Correlations in Herwig 7

This repository is the research software and reproducibility companion for the HerwigPol polarized deep-inelastic scattering implementation developed for Herwig 7. It brings together the modified Herwig and ThePEG source snapshots, the curated POLDIS fixed-order reference code, the custom Rivet analyses, the DIS validation workflow, and the paper source in a single formal repository layout. The repository is intended to preserve the source-level ingredients needed to rebuild and re-run the validated DIS studies. It therefore tracks code, input cards, workflow drivers, and technical notes, while intentionally excluding generated artifacts such as build products, campaign outputs, merged YODA files, plots, and rendered paper outputs.

Papaefstathioou, Andreas [Kennesaw State Universit↗
DOE OSTI2020

Characterization and differentiation of aluminum powders used in improvised explosive devices – Part 1: Proof of concept of the utility of particle micromorphometry

Abstract Aluminum (Al) powders are commonly used in improvised explosive devices as metallic fuels, a component of explosive mixtures. These powders can be obtained readily from industrial‐scale and consumer products, and produced using unsophisticated “kitchen chemistry” techniques. This research demonstrates the potential of automated particle micromorphometry for comparisons between known source and questioned Al powders recovered from IEDs, as well as for insight into the method of Al powder manufacture. Al powder samples were obtained from legitimate manufacturers, and 56 samples were produced “in‐house” from Al‐containing spray paints and ball‐milled Al foils. Transmitted light microscope images of Al powder particles were acquired using an automated stage with automated z‐focus; 17 size and shape parameters were measured for all particles. Approximately 37,000–2,500,000 particles/sample were analyzed using an open‐source statistical package with customized code. Dimensionality reduction was required for processing the large datasets: eight of the 17 measured variables were selected based on inspection of the correlation matrix. Data from four subsamples from each of the 56 samples produced using “in‐house” methods were analyzed using ANOVA to assess the within‐ and between‐sample variation. High within‐sample variation was noted; however, ANOVA and post‐hoc Tukey's honestly significant difference (HSD) tests demonstrated that the between‐sample variation was substantially larger than the within‐sample variation. Each sample could be differentiated from all other samples in the test set. Future experiments will focus on ways to reduce the within‐sample variation, and additional statistical and microanalytical methods to classify sources and confidently constrain the method of Al powder manufacture.

Baldaino, JenaMarie↗
DOE OSTI2021

Nyx: A Massively Parallel AMR Code for Computational Cosmology

Nyx is a highly parallel, adaptive mesh, finite-volume N-body compressible hydrodynamics solver for cosmological simulations. It has been used to simulate different cosmological scenarios with a recent focus on the intergalactic medium and Lyman alpha forest. Together, Nyx, the compressible astrophysical simulation code, Castro, and the low Mach number code MAESTROeX, make up the AMReX-Astrophysics Suite of open-source, adaptive mesh, performance-portable astrophysical simulation codes. Other examples of cosmological simulation research codes include Enzo, Enzo-P/Cello, RAMSES, ART, FLASH, Cholla, as well as Gadget, Gasoline, Arepo, Gizmo, and SWIFT.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. Here, we present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Tokamak plasma high field side response to an n = 3 magnetic perturbation: a comparison of 3D equilibrium solutions from seven different codes

In comparing equilibrium solutions for a DIII-D shot that is amenable to analysis by both stellarator and tokamak three-dimensional (3D) equilibrium codes, a significant disagreement has been seen between solutions of the VMEC stellarator equilibrium code and solutions of tokamak perturbative 3D equilibrium codes. In this research, the source of that disagreement has been investigated, and that investigation has led to new insights into the domain of validity of the different equilibrium calculations, and to a finding that the manner in which localized screening currents at low order rational surfaces are handled can affect global properties of the equilibrium solution. The perturbative treatment has been found to break down at surprisingly small perturbation amplitudes due to overlap of the calculated perturbed flux surfaces, and that treatment is not valid in the pedestal region of the DIII-D shot studied. The perturbative treatment is valid, however, further into the interior of the plasma, and flux surface overlap does not account for the disagreement investigated here. Calculated equilibrium solutions for simple model cases and comparison of the 3D equilibrium solutions with those of other codes indicate that the disagreement arises from a difference in handling of localized currents at low order rational surfaces, with such currents being absent in VMEC and present in the perturbative codes. The significant differences in the global equilibrium solutions associated with the presence or absence of very localized screening currents at rational surfaces suggests that it may be possible to extract information about localized currents from appropriate measurements of global equilibrium plasma properties. Therefore, that would require improved diagnostic capability on the high field side of the tokamak plasma, a region difficult to access with diagnostics.

3-DIMENSIONAL MAGNETOHYDRODYNAMIC EQUILIBRIA↗
DOE OSTI2016

CDAC Student Report: Summary of LLNL Internship

Multiple objectives motivated me to apply for an internship at LLNL: I wanted to experience the work environment at a national lab, to learn about research and job opportunities at LLNL in particular, and to gain greater experience with code development, particularly within the realm of high performance computing (HPC). This summer I was selected to participate in LLNL's Computational Chemistry and Material Science Summer Institute (CCMS). CCMS is a 10 week program hosted by the Quantum Simulations group leader, Dr. Eric Schwegler. CCMS connects graduate students to mentors at LLNL involved in similar re- search and provides weekly seminars on a broad array of topics from within chemistry and materials science. Dr. Xavier Andrade and Dr. Erik Draeger served as my co-mentors over the summer, and Dr. Andrade continues to mentor me now that CCMS has concluded. Dr. Andrade is a member of the Quantum Simulations group within the Physical and Life Sciences at LLNL, and Dr. Draeger leads the HPC group within the Center for Applied Scientific Computing (CASC). The two have worked together to develop Qb@ll, an open-source first principles molecular dynamics code that was the platform for my summer research project.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Collision Tracking in OpenMC: Methods and Applications in Neutron Noise, Neutron Imaging, Time-of-Flight, and Multiplicity Counting

We present the development and application of a collision tracking feature within the OpenMC Monte Carlo particle transport code, designed for diverse applications such as neutron spectroscopy, scatter camera system, neutron noise, and multiplicity counting simulations. This feature enables the tracking of individual particle collisions, with potential applications in nuclear nonproliferation, reactor physics, and nuclear security. Additionally, the feature holds potential for the calibration of neutron detectors, specifically in converting light output into energy deposited within the detectors. The implementation consists of a set of filters—such as reaction type, energy, cell, and material—that constrain the set of collisions that are tracked, extensions to the Python API to enable simple input specification, and support for writing either OpenMC’s native HDF5-based format or the Monte Carlo particle list format. This feature was added to the official OpenMC release in version 0.15.3. In this work, the feature will be applied to showcase scenarios such as time-of-flight simulations, scatter-camera imaging for neutron source localization, neutron-noise analysis to extract integral kinetic parameters such as the prompt decay constant α, and multiplicity counting to estimate the mass of special nuclear materials. Ultimately, this feature aims to expand the application scope of open-source Monte Carlo particle transport codes such as OpenMC.

Monte Carlo code↗
DOE OSTI2024

Methodology to Generate Decay Gamma Sources for the Second Target Station Full Target Assembly

This report details the specifics of applying the Position-Averaged Method to the target wedges and staves. The methodology used to calculate the decay gamma sources for a single target wedge to the calculation of the Second Target Station (STS) target assembly activation and the resulting decay gamma source terms are detailed in this report. The methodology has been applied to the entire STS target assembly including all of the target wedges, staves, shaft, and drive. The target staves and shaft have been vertically segmented to ensure the decay gamma source gradient is sufficiently captured. MCNP ® Code Version 6.2.0 with the RNUCS patch coupled with CINDER2008 from the AARE V1.0 package are the main tools used to calculate the decay gamma source terms.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Numerical simulation of ion charge breeding in electron beam ion source

The Electron Beam Ion Source particle-in-cell code (EBIS-PIC) tracks ions in an EBIS electron beam while updating electric potential self-consistently and atomic processes by the Monte Carlo method. Recent improvements to the code are reported in this paper. The ionization module has been improved by using experimental ionization energies and shell effects. The acceptance of injected ions and the emittance of extracted ion beam are calculated by extending EBIS-PIC to the beam line transport region. An EBIS-PIC simulation is performed for a Cs charge-breeding experiment at BNL. The charge state distribution agrees well with experiments, and additional simulation results of radial profiles and velocity space distributions of the trapped ions are presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Extension of OpenMC for Fixed Source Transmutation Calculations

This report documents work performed under a Strategic Partnership Project between Argonne National Laboratory (ANL) and the United Kingdom Atomic Energy Authority (UKAEA). The overall goal of this project is to extend OpenMC [1], a community-developed Monte Carlo particle transport code, to be able to perform fixed-source transmutation calculations. In fission and fusion reactors, the high flux of energetic neutrons causes materials within the reactor to “transmute,” or undergo a nuclear reaction that results in the addition/removal of neutrons and protons from the nucleus of an atom. If subjected to these reactions for long enough, the overall composition and physical properties of the material itself begin to change as a result of transmutation. Such a feature is vital for predicting the decrease in tritium production rate within a breeder blanket during the lifetime of a fusion reactor. OpenMC is capable of simulating neutron transport in fission/fusion systems, thereby allowing it to estimate the flux that causes transmutation. It is also capable of solving the transmutation equations, which determine how the composition of a material changes over time due to neutron irradiation and radioactive decay. However, solving the transmutation equations was previously only possible when the source of neutrons came from a fission system. In a fusion system, the source of neutrons is typically determined by a separate code and then given as an input to the particle transport simulation. This is known as a fixed source calculation. Through this project, we have extended OpenMC to solve the transmutation equations for a fixed source calculation. Evaluating the change in material compositions due to transmutation and its effect on physical properties is of key importance to a range of engineering analyses for fission and fusion systems. For example, in a fusion reactor, estimating the dose rate at different physical locations resulting from irradiated materials in the reactor allows designers to ensure that workers are not exposed to doses beyond applicable regulations. In order to properly dispose of irradiated materials, designers also need to estimate the radiotoxicity, which again relies on knowledge of the material composition at some future time. The specific tasks for this project that were agreed to between ANL and UKAEA were as follows: 1. Make changes and additions in the openmc.deplete and related modules in OpenMC to support transmutation calculations following a fixed source transport simulation. 2. Make necessary changes to OpenMC to model transmutation due to an arbitrary set of reactions needed for fusion applications. Use this new capability to generate a depletion chain file based on the TENDL nuclear data library. 3. Improve the openmc.deplete module in OpenMC to keep track of gases produced as a by-product of nuclear reactions during transmutation calculations. 4. Validate the new capabilities by carrying out fixed-source transmutation calculations on a suitable benchmark problem using OpenMC and a comparable Monte Carlo neutron transport code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗