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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 1,153 records · Page 64
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Effects of uncertainties in simulations of extragalactic UHECR propagation, using CRPropa and SimProp

The results of simulations of extragalactic propagation of ultra-high energy cosmic rays (UHECRs) have intrinsic uncertainties due to poorly known physical quantities and approximations used in the codes. We quantify the uncertainties in the simulated UHECR spectrum and composition due to different models of extragalactic background light (EBL), different photodisintegration setups, approximations concerning photopion production and the use of different simulation codes. We discuss the results for several representative source scenarios with proton, nitrogen or iron at injection. For this purpose we used SimProp and CRPropa, two publicly available codes for Monte Carlo simulations of UHECR propagation. CRPropa is a detailed and extensive simulation code, while SimProp aims to achieve acceptable results using a simpler code. We show that especially the choices for the EBL model and the photodisintegration setup can have a considerable impact on the simulated UHECR spectrum and composition.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Axial deformed solution of the Skyrme-Hartree-Fock-Bogolyubov equations using the transformed harmonic oscillator Basis

HFBTHO is a physics computer code that is used to model the structure of the nucleus. It is an implementation of the nuclear energy Density Functional Theory (DFT), where the energy of the nucleus is obtained by integration over space of some phenomenological energy density, which is itself a functional of the neutron and proton densities. In HFBTHO, the energy density derives either from the zero-range Dkyrme or the finite-range Gogny effective two-body interaction between nucleons. Nuclear superfluidity is treated at the Hartree-Fock-Bogoliubov (HFB) approximation, and axial-symmetry of the nuclear shape is assumed. This version is the 3rd release of the program; the two previous versions were published in Computer Physics Communications [1,2]. The previous version was released at LLNL under GPL 3 Open Source License and was given release code LLNL-CODE-573953.

Perez, R. Navarro↗
DOE OSTI2019

Activation Study of Isotopically Enriched Samples During Neutron Time-of-Flight Measurements

Over the course of a time-of-flight neutron cross section measurement, the sample isotope(s) of interest are irradiated by neutrons and photons. Ideally, the enriched sample will not have any additional radioactivity after the measurement as compared to the baseline detectable radioactivity (0.6 Bq/g) established prior to acquisition of the sample. This document presents the process for estimating the activation of a stable enriched isotope exposed to a mixed photon and neutron source using tools from the SCALE code system developed at Oak Ridge National Laboratory. The report also presents the activation analysis for the isotopes of current interest: 90;91;92;94 Zr. This information is provided for two sets of Zr samples: one set of ZrO 2 samples, and one set of metal Zr samples. It was determined that, following measurements at the GELINA linear accelerator facility, all of the enriched Zr samples would contain less than 0.6 Bq/g of radioactivity within 183 days after irradiation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Introducing single-crystal scattering and optical potentials into MCNPX: Predicting neutron emission from a convoluted moderator

Neutron transport simulation codes are an indispensable tool used for the design and construction of modern neutron scattering facilities and instrumentation. It has become increasingly clear that some neutron instrumentation has started to exploit physics that is not well-modelled by the existing codes. Particularly, the transport of neutrons through single crystals and across interfaces in MCNP(X), Geant4 and other codes ignores scattering from oriented crystals and refractive effects, and yet these are essential ingredients for the performance of monochromators and ultra-cold neutron transport respectively (to mention but two examples). In light of these developments, we have extended the MCNPX code to include a single-crystal neutron scattering model and neutron reflection/refraction physics. Furthermore, we have also generated silicon scattering kernels for single crystals of definable orientation with respect to an incoming neutron beam. As a first test of these new tools, we have chosen to model the recently developed convoluted moderator concept, in which a moderating material is interleaved with layers of perfect crystals to provide an exit path for neutrons moderated to energies below the crystal s Bragg cut off at locations deep within the moderator. Studies of simple cylindrical convoluted moderator systems of 100 mm diameter and composed of polyethylene and single crystal silicon were performed with the upgraded MCNPX code and reproduced the magnitude of effects seen in experiments compared to homogeneous moderator systems. Applying different material properties for refraction and reflection, and by replacing the silicon in the models with voids, we show that the emission enhancements seen in recent experiments are primarily caused by the transparency of the silicon/void layers. Finally the convoluted moderator experiments described by Iverson et al. were simulated and we find satisfactory agreement between the measurement and the results of simulations performed using the tools we have developed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Emittance study of a 28 GHz electron cyclotron resonance ion source for the Rare Isotope Science Project superconducting linear accelerator

A 28 GHz electron cyclotron resonance (ECR) ion source is being developed for use as an injector for the superconducting linear accelerator of the Rare Isotope Science Project. Beam extraction from the ECR ion source has been simulated using the KOBRA3-INP software. The simulation software can calculate charged particle trajectories in three dimensional complex magnetic field structures, which in this case are formed by the arrangement of five superconducting magnets. In this study, the beam emittance is simulated to understand the effects of plasma potential, mass-to-charge ratio, and spatial distribution. The results of these simulations and their comparison to experimental results are presented in this paper.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Thermal Neutron Imaging Using A New Pad-Based Position Sensitive Neutron Detector

Thermal neutrons (with mean energy of 25 meV) have a scattering mean free path of about 20 m in air. Therefore it is feasible to find localized thermal neutron sources up to ~30 m standoff distance using thermal neutron imaging. Coded aperture thermal neutron imaging was developed in our laboratory in the nineties, using He-3 filled wire chambers. Recently a new generation of coded-aperture neutron imagers has been developed. In the new design the ionization chamber has anode and cathode planes, where the anode is composed of an array of individual pads. The charge is collected on each of the individual 5x5 mm2 anode pads, (48x48 in total, corresponding to 24x24 cm2 sensitive area) and read out by application specific integrated circuits (ASICs). The high sensitivity of the ASICs allows unity gain operation mode. The new design has several advantages for field deployable imaging applications, compared to the previous generation of wire-grid based neutron detectors. Among these are the rugged design, lighter weight and use of non-flammable stopping gas. For standoff localization of thermalized neutron sources a low resolution (11x11 pixel) coded aperture mask has been fabricated. Using the new larger area detector and the coarse resolution mask we performed several standoff experiments using moderated californium and plutonium sources at Idaho National Laboratory. In this paper we will report on the development and performance of the new pad-based neutron camera, and present long range coded-aperture images of various thermalized neutron sources.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Modeling the contributions to acoustic nonlinearity from complex dislocation networks using 3D dislocation dynamics

Nonlinear ultrasonic parameters are highly sensitive to microstructural features that affect macroscale material behavior, providing a nondestructive means to characterize their evolution. Although dislocations are known to be a strong source of acoustic nonlinearity, establishing quantitative links between the acoustic nonlinearity parameter (β), measured via Second Harmonic Generation, and dislocation morphology—such as dislocation length and density—remains an open challenge. This work advances the numerical modeling of dislocation–β relationships using 3D dislocation dynamics (DD) simulations in two approaches: a “static” method computing strain and stress fields from dislocation configurations in the absence of external loading, and a “quasi-static” method to estimate β from the curvature of dislocation lines under applied load. First, the static method is combined with finite element analysis to investigate a recent assertion that heterogeneous initial strain fields can induce higher harmonic generation in a linear elastic medium; the present results do not corroborate this outcome. Then, the quasi-static method is applied to multiple-dislocation scenarios through parametric studies, revealing behaviors not predicted by analytical models, such as the competing interactions of edge and screw dislocations and the significant influence of applied stress on β. Finally, the simulations are used to model SHG experimental results and validate the hypothesis that β can decrease during plastic deformation, despite increasing dislocation density. As the DD code used here is open-source, it provides a practical platform for future investigation into microstructure–β relationships important to the interpretation of SHG results.

Materials science↗
DOE OSTI2018

Development of ARCHER towards clinical use: modeling and simulation of Varian Linac for radiation therapy dose calculations - 25449

Monte Carlo particle transport methods are well recognized as the most accurate approach to dose calculations for radiation therapy treatment planning. However, the lengthy calculation time has prevented Monte Carlo methods from being applied to dose calculations in the planning of radiation therapy. Recently, the emerging parallel computing hardware accelerators, such as general-purpose Graphics Processing Unit (GPU), have significantly boosted the performance of Monte Carlo codes. The accuracy of a clinical dose engine depends greatly on the accuracy of source term modeling and the methodology of dose calculations. Hence, in this study, an efficient and accurate source model of the Varian TrueBeam{sup R} Linac is developed and integrated with a fast photon-electron transport module of the recently developed high-performance Monte Carlo code, ARCHER. The validity of the source model and the particle transport physics are verified with established dose calculation codes such as EGSnrc and validated with experimental measurements. To demonstrate the capability of the platform in clinical applications, three clinical treatment plans (the lung, breast, and prostate) are simulated and analyzed. From the perspective of performance, the entire simulation of each clinical plan, starting from the source modeling to the end of particle transport within the patient phantom, can be finished in less than one minute using a single Titan X GPU card with 1% statistical error. In comparison, the same cases take the EGSnrc code hundreds of CPU hours to finish. The results suggest that this new version of ARCHER code has been successfully developed and tested as a clinically deployable dose engine for external photon beam IMRT plans in the clinical setting. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Development of MOSCATO: A CFD-Level Electrochemistry and Corrosion Simulator for Molten Salt Systems

For both coolant and fueled variants of molten salt reactors (MSRs), the corrosion of structural materials is a significant challenge. The corrosion stems from chemical and electrochemical reactions initiated by fissile material, fission products, and impurities in the salt. Lower-fidelity models rely on empirical correlations for mass transfer, simplified lumped temperature profiles, and similar assumptions. They do not capture detailed spatial variations in complex geometries, creating the need for high-fidelity modeling to bridge this gap.As we approach the demonstration and possible deployment of MSRs in this decade, the development of a high-fidelity, high-performance simulator becomes imperative. To simulate the complex electrochemical environment and corrosion within molten salt systems, we have developed the Molten Salt Chemistry And TranspOrt (MOSCATO) code. This endeavor is comprised of three essential components. First, mass transfer equations are coupled with the Navier-Stokes equations in order to account for the transport of species in the salt. Second, the diffusion of alloy constituents, such as Cr, Fe, Ni, etc. is simulated within the structural metals. Third, the alloy and salt domains are coupled to account for the heterogeneous chemical and electrochemical reactions that occur at the salt-alloy interface.MOSCATO manages all three components within the framework of the highly scalable, open-source spectral element method computational fluid dynamics code Nek5000/NekRS. This integration enables MOSCATO to harness the immense computational power of modern high-performance computing resources, ensuring both high fidelity and computational speed.In addition to code development, we have initiated a comprehensive verification and validation campaign, utilizing data from diverse sources. First, MOSCATO's electrochemical solver was verified with reference numerical data. Then validation occurred against experiments: one of a thermal galvanic cell and the other for corrosion in flowing molten salt of FLiNaK (LiF-NaF-KF). This campaign verified and validated MOSCATO as a reliable tool for simulating electrochemical environments and corrosion in molten salt systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Drekar v.2.0

Drekar is an application code that solves partial differential equations for fluids that can be optionally coupled to electromagnetics. Drekar solves low-mach compressible and incompressible computational fluid dynamics (CFD), compressible and incompressible resistive magnetohydrodynamics (MHD), and multiple species plasmas interacting with electromagnetic fields. Drekar discretization technology includes continuous and discontinuous finite element formulations, stabilized finite element formulations, mixed integration finite element bases (nodal, edge, face, volume) and an initial arbitrary Lagrangian Eulerian (ALE) capability. Drekar contains the implementation of the discretized physics and leverages the open source Trilinos project for both parallel solver capabilities and general finite element discretization tools. The code will be released open source under a BSD license. The code is used for fundamental research for simulation of fluids and plasmas on high performance computing environments.

Seefeldt, Ben↗
DOE OSTI2013

Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely productive Waste Generator Instructions (WGIs) have been used occasionally in the past at large sites for treatment and packaging of TRU waste. The WGIs have resulted in highly efficient waste treatment, packaging and certification for disposal of TRU waste at WIPP. For example, a single WGI at LANL, combined with an increase in gram loading, resulted in a mind boggling 6,400% increase in waste loading for {sup 238}Pu heat source waste. In fact, the WGI combined with a new Contact Handled (CH) TRU Waste Content (TRUCON) Code provided a massive increase in shippable wattage per Transuranic Package Transporter-II (TRUPACT-II) over the previously used and more restrictive TRUCON Code that have been used previously for the heat source waste. In fact, the use of the WGI process at LANL's TA-55 facility reduced non-compliant drums for WIPP certification and disposal from a 13% failure rate down to a 0.5% failure rate and is expected to further reduce the failure rate to zero drums per year. The inherent value of the WGI is that it can be implemented in a site's current procedure issuance process and it provides documented proof of what actions were taken for each waste stream packaged. The WGI protocol provides a key floor-level operational component to achieve goal alignment between actual site operations, the WIPP TRU waste packaging instructions, and DOE O 435.1. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

WE-A-17A-07: Evaluation of a Grid-Based Boltzmann Solver for Nuclear Medicine Voxel-Based Dose Calculations

Purpose: Grid-based Boltzmann solvers (GBBS) have been successfully implemented in radiation oncology clinics as dose calculations for e×ternal photon beams and 192Ir sealed-source brachytherapy. We report on the evaluation of a GBBS for nuclear medicine vo×el-based absorbed doses. Methods: Vo×el-S-values were calculated for monoenergetic betas and photons (1, 0.1, 0.01 MeV), 90Y, and 131I for 3 mm vo×el sizes using Monte Carlo (DOS×YZnrc) and GBBS (Attila 8.1-beta5, Transpire). The source distribution was uniform throughout a single vo×el. The material was an infinite 1.04 g/cc soft tissue slab. To e×plore convergence properties of the GBBS 3 tetrahedral meshes, 3 energy group structures, 3 different square Chebyschev-Legendre quadrature set orders (Sn), and 4×2013;7 spherical harmonic e×pansion terms (Pn) were investigated for a total of 168 discretizations per source. The mesh, energy group, and quadrature sets are 8×, 3×, and 16×, respectively, finer than the corresponding coarse discretization. GBBS cross sections were generated with full electronphoton-coupling using the vendors e×tended CEP×S code. For accuracy, percent differences (%Δ) in source vo×el absorbed doses between MC and GBBS are reported for the coarsest and finest discretization. For convergence, ratios of the two finest discretization solutions are reported along each variable. Results: For 1 MeV, 0.1 MeV, 0.01 MeV, Y90, and I-131 beta sources the %Δ in the source vo×el for (coarsest,finest) discretization were (+2.0,−6.4), (−8.0, −7.5), (−13.8, −13.4), (+0.9,−5.5), and (− 10.1,−9.0) respectively. The corresponding %Δ for photons were (+33.7,−7.1), (−9.4, −9.8), (−17.4, −15.2), and (−1.7,−7.7), respectively. For betas, the convergence ratio of mesh, energy, Sn, and Pn ranged from 0.991–1.000. For gammas, the convergence ratio of mesh, Sn, and Pn ranged from 0.998–1.003 while the ratio for energy ranged from 0.964–1.001. Conclusions: GBBS is promising for nuclear medicine vo×el-based dose calculations. Ongoing work includes evaluating GBBS in bone, lung, and realistic clinical PET/SPECT-based activity distributions. Research reported in this publication was supported by the National Cancer Institute of the National Institutes of Health under Award Number R01CA138986. The content is solely the responsibility of the authors and does not necessarily represent the official views of the National Institutes of Health.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Multi-phenomenology Yield Characterization

This report serves as the first delivery of a four-year applied science effort to transform and advance the error bounds for the yield estimate of an explosion. Each year’s delivery will be in this form, culminating in the submission of this work for peer review to a scientific journal. Importantly, the yearly progress reports can then also be viewed as expanding drafts working towards a formal journal article submission. For the first tranche of funding, we collaborated with Air Force Technical Applications Center (AFTAC) scientists to identify unclassified real-world data that demonstrate and validate our advanced error propagation methods. Collaboration includes visits to AFTAC and telecons. For this development, we illustrate the fusion of seismic, acoustic, optical, and surface effect signatures from an explosion. The mathematics and code being adapted to this specific application (Williams et al., 2021) involves physics models of multiple sensor signatures. We have also identified related physics models and have integrated them into code. Current methods of underground explosion yield estimation for the Threshold Test Ban Treaty (TTBT) have served the US treaty monitoring mission well for decades. A research objective of the Defense Nuclear Nonproliferation Research and Development (DNN R&D) office of the National Nuclear Security Administration (NNSA) has always been to provide new technical capabilities for monitoring lower thresholds. The general model and error propagation code to be developed in this project is based on significant advances in error modeling and propagation needed to analyze data at lower detection thresholds. The second tranche of funding for this project began on May 1, 2022, and planned work for the second tranche includes: i) completing the integration of physical model code into the general error model framework; this code accommodates a wide range of linear/nonlinear source models, fixed/ random effects, and frequentist/Bayesian analyses (the purpose of which is not to dictate to users how to analyze data, but instead to allow users the maximum flexibility in their work); ii) illustrative application of code to identified data, and; iii) initial planning with AFTAC researchers on delivery of code to the Common Development Environment at AFTAC, and continued writing of the planned final journal article submission (year two of this progress report), with particular emphasis on descriptions of data identified for this effort.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2024

Development and validation of fully open-source R2S shutdown dose rate capabilities in OpenMC *

We present the first fully open-source capabilities for shutdown dose rate (SDR) calculations of fusion energy facilities based on the Rigorous 2-Step (R2S) methodology. These capabilities have been implemented in the OpenMC Monte Carlo particle transport code, building on its existing capabilities while also leveraging new features that have been added to the code to support SDR calculations, such as decay photon source generation. Each of the individual physics components in the R2S workflow—neutron transport, activation, decay photon source generation, and photon transport—have been verified through code-to-code comparisons with MCNP6.2 and FISPACT-II 4.0. These comparisons generally demonstrate excellent agreement between codes for each of the physics components. The full cell-based R2S workflow was validated by performing a simulation of the first experimental campaign from the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from the Shielding INtegral Benchmark Archive and Database (SINBAD). For short cooling times, the dose calculated by OpenMC agrees with the experimental measurements within the stated experimental uncertainties. For longer cooling times, an overprediction of the shutdown dose was observed relative to experiment, which is consistent with previous studies in the literature. Altogether, these features constitute a combination of capabilities in a single, open-source codebase to provide the fusion community with a readily-accessible option for SDR calculations and a platform for rapidly analyzing the performance of fusion technology.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

OpenSn: A massively parallel, open-source simulation environment for discrete ordinates radiation transport

OpenSn is an open-source, massively parallel deterministic radiation transport code for solving the discrete-ordinates ( S N ) form of the Boltzmann transport equation on unstructured, arbitrary polyhedral meshes. It supports high-fidelity simulations involving steady-state, eigenvalue, and adjoint problems for neutral particles (e.g., neutrons, photons, multi-particles), using the multigroup approximation in energy. OpenSn combines angular discretization via discrete ordinates with a discontinuous Galerkin finite element method (DGFEM) in space, enabling accurate resolution of transport physics on arbitrary polyhedral cells, included locally refined spatial grids. It includes multiple angular quadrature types, including locally refined angular quadratures. Written in modern C++ with a Python API, OpenSn runs efficiently on platforms ranging from laptops to supercomputers. The transport sweep algorithm is implemented using a task-based, directed-acyclic-graph (DAG) approach for each angle and supports asynchronous parallelism across thousands of MPI ranks. Group-set aggregation improves compute intensity, and synthetic acceleration techniques (e.g., diffusion synthetic acceleration, second-moment method) enhance solver convergence. OpenSn has been verified on reactor physics problems and demonstrated excellent weak and strong scaling performance on more than 32,768 processes, making it a versatile and robust platform for large-scale transport simulations in complex geometries.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Modeling of Gas Dynamic Processes in the Interelectrode Gap of a Multielement Thermionic Electricity Generating Channel by Solving the Kinetic Equation

This work is devoted to the application of a numerical method of solving Boltzmann’s equation for modeling the behavior of radionuclides (Kr, Xe, Rb, Sr, Cs, Ba) in the cavity of the interelectrode gap of a multielement electricity generating channel. Modeling methods were developed and software system implemented in the course of this work. Calculations were performed for two structural arrangements: with uni- and bilateral extraction of radionuclides into the vacuum-cesium system. Data on the pressure and flow distributions were obtained. The krypton and xenon pressure near the collector as functions of their flow and the cesium gas pressure at the gap exits were determined computationally.

07 ISOTOPE AND RADIATION SOURCES↗