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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Validation of numerical models for seismic fluid-structure-interaction analysis of nuclear, safety-related equipment

Seismic design, qualification, and risk assessment of nuclear safety-related vessels filled with liquid will have to consider the interaction between the vessel (tank), the contained liquid, and submerged components, if any. Seismic fluid-structure-interaction (FSI) analysis of nuclear vessels will rely on numerical models, which are required to be verified and validated. This study validates previously-verified numerical models using test data generated from earthquake-simulator experiments involving a cylindrical tank. Two solvers in LS-DYNA for FSI simulations are used for the numerical analysis: Arbitrary-Lagrangian-Eulerian (ALE) and Incompressible Computational Fluid Dynamics (ICFD). Numerical and test results are compared for responses critical to seismic design of advanced reactor vessels: hydrodynamic pressures on the tank wall, reactions at the support, and wave heights of the contained liquid. Analysis is performed for one-, two-, and three-directional seismic inputs with a range of intensity, and rocking motions. The accuracy of the numerical results and different methods for outputting wave heights are discussed. Recommendations for validation of seismic FSI numerical models of advanced reactors are provided. As a result, the validation exercise presented in this study is broadly applicable to cylindrical tanks, regardless of industry sector.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Mass ordering sum rule for the neutrino disappearance channels in T2K, NOvA, and JUNO

We revisit, reformulate, and extend a method for determining the neutrino mass ordering by using precision measurements of the atmospheric Δ m 2 s in both electron and muon neutrino disappearance channels, first proposed by the authors in 2005 [H. Nunokawa , Phys. Rev. D 72, 013009 (2005)]. The mass ordering is a very important outstanding question for our understanding of the elusive neutrino and determination of the mass ordering has consequences to particle physics, nuclear physics, and cosmology. The JUNO reactor experiment will start data taking this year, and the precision of the atmospheric Δ m 2 s from electron antineutrino measurements will improve by a factor of 3 from Daya Bay’s 2.4% to 0.8% within a year. This measurement, when combined with the atmospheric Δ m 2 ’s measurements from T2K and NOvA for muon neutrino disappearance, will contribute substantially to the Δ χ 2 between the two remaining neutrino mass orderings. In this paper we derive for the first time a mass ordering sum rule that can be used to address the possibility that JUNO’s atmospheric Δ m 2 ’s measurement, when combined with other experiments in particular T2K and NOvA, can determine the neutrino mass ordering at the 3 σ confidence level within one year of operation. For a confidence level of 5 σ in a single experiment, we will have to wait until the middle of the next decade when the DUNE experiment is operating.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR Northeast Flux Trap”). This depletion analysis provides fast fluence and fission heat rate data for all components (fuel compacts, graphite rings, stainless steel retainer, etc.), which are used as inputs for the thermal analysis codes. The graphite temperatures from thermocouples (TCs) in each capsule were used to calibrate these thermal analysis codes. However, given the high rate of TC failure under the harsh irradiation and thermal conditions in the AGR capsules, the thermal analysis results are very useful in aiding TC data qualification, increasing the confidence in delineating failures of the measuring instruments (TCs) from physical mechanisms that may have shifted the system thermal response.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Electrical Engineering in Los Alamos Neutron Science Center Accelerator

The field of electrical engineering plays a significant role in particle accelerator design and operations. Los Alamos National Laboratories LANSCE facility utilizes the electrical energy concepts of power distribution, plasma generation, radio frequency energy, electrostatic acceleration, signals and diagnostics. The culmination of these fields produces a machine of incredible potential with uses such as isotope production, neutron spallation, neutron imaging and particle analysis. The key isotope produced in LANSCE isotope production facility is Strontium-82 which is utilized for medical uses such as cancer treatment and positron emission tomography also known as PET scans. Neutron spallation is one of the very few methods used to produce neutrons for scientific research the other methods are natural decay of transuranic elements from nuclear reactors. Accelerator produce neutrons by accelerating charged particles into neutron dense elements such as tungsten imparting a neutral particle with kinetic energy, this has the benefit of producing a large number of neutrons as well as minimizing the waste generated. Utilizing the accelerator scientist can gain an understanding of how various particles behave and interact with matter to better understand the natural laws of physics and the universe around us.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

NRC Multiphysics Analysis Capability Deployment (FY2021--Part 1)

This report details progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” The tasks completed for this report are as follows: First, Task 1d: The net radiation transfer method was implemented into MOOSE for modeling reactor cavity cooling system geometries. RCCS models for two experiments were created: (1) Natural Convection Shutdown Heat Removal Test Facility (NSTF) experiment R022, and (2) HTTR VCS mockup. For validation, computed temperature distributions were compared to measured temperatures. Next, Task 4c: An algorithm for computing the pebble bed reactor equilibrium core isotopic com-position was developed and an initial version is implemented into the reactor multi-physics code Griffin. Initial results for a simplified axisymmetric pebble bed reactor are presented. Finally, Task 7: generation of a reference plant model for molten salt cooled pebble bed reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

IAEA Photonuclear Data Library 2019

Here, photo-induced reaction cross section data are of importance for a variety of current or emerging applications, such as radiation shielding design and radiation transport analyses, calculations of absorbed dose in the human body during radiotherapy, physics and technology of fission reactors (influence of photo-reactions on neutron balance) and fusion reactors (plasma diagnostics and shielding), activation analyses, safeguards and inspection technologies, nuclear waste transmutation, medical isotope production and astrophysical applications. To address these data needs the IAEA Photonuclear Data library was produced in 1999, containing evaluated photo-induced cross sections and neutron spectra for 164 nuclides which were deemed relevant for the applications. Since the release of the IAEA Photonuclear Data Library however, new experimental data as well as new methods to assess the reliability of experimental cross sections have become available. Theoretical models and input parameters used to evaluate photo-induced reactions have improved significantly over the years. In addition, new measurements of partial photoneutron cross sections using mono-energetic photon beams and advanced neutron detection systems have been performed allowing for the validation of the evaluations and assessments of the experimental data. Furthermore, technological advances have led to the construction of new and more powerful gamma-beam facilities, therefore new data needs are emerging. We report our coordinated efforts to address these data needs and present the results of the new up-to-date evaluations included in the new updated IAEA Photonuclear Data Library consisting of 219 nuclides. The new library includes 188 new evaluations produced by the CRP evaluators, and one evaluation taken from the JENDL/PD-2016 library, while 20 evaluations were retained from the previous 1999 IAEA Photonuclear Data Library. In most of the cases, the photon energy goes up to 200 MeV. A total of 55 nuclides are new in this library reflecting the progress in measurements but also the developing data needs. In this paper we discuss the new assessment method and make recommendations to the user community in cases where the experimental data are discrepant and the assessments disagree. In addition, in the absence of experimental data, we present model predictions for photo-induced reaction cross section on nuclides of potential interest to medical radioisotope production.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

fuel cycle cost analysis of Th/U MOX fuel in a block-type HTR - 18619

With a rapid development of nuclear energy in these years, the supply of nuclear fuel is becoming more and more important. Thorium regains attention because of its abundant reserves and excellent physical properties. As a next generation reactor, the high temperature reactor (HTR) is a promising choice for thorium utilization because of its deeper burnup, harder spectrum and better inherent safety. Therefore, a 600 MWth 108-column modular high temperature reactor (MHR) core model with 20%UO{sub 2}-80%ThO{sub 2}; MOX fuel is adapted. A traditional two-step calculation scheme, lattice calculation then core calculation, is used to get the neutronic performance of equilibrium cycle including uranium enrichment and mass of fuel. Using these data, the fuel cycle cost of HTR can be calculated by the levelized lifetime cost method. In order to take into account the variation of unit price, a well-established, standard statistical analysis technique is used by Monte Carlo simulation, (thus, the probability distribution of fuel cycle cost can be obtained. Furthermore, sensitivity calculations are made to analyze the impact on the fuel cycle cost of variations in the unit prices for each fuel cycle component. The result shows that the most influential parameters are unit price of natural uranium, enrichment and discount rate. Finally a comparison of fuel cycle cost between 100%UO{sub 2} and 20%UO{sub 2}-80%ThO{sub 2} MOX fuel is made. The latter is 11% cheaper than the former. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Better Method to Calculate Fuel Burnup in Pebble Bed Reactors Using Machine Learning

Burnup measurement is an important step in material control and accountancy (MC&A) at nuclear reactors, and may be done by examining gamma spectra of fuel samples. Traditional approaches rely on known correlations to specific photopeaks (e.g. 137 Cs) and operate via a standard linear regression method. However, the quality of these regression methods is limited even in the best case, and is significantly poorer at short fuel cool-down times, due to the elevated radiation background by short life-time isotopes, and self-shielding effect of the fuel. For practical operation of pebble bed reactors (PBRs), quick measurements (in minutes) and short cooling times (in hours) are required from a safety and security perspective. We investigated the efficacy and performance of machine learning (ML) methods to predict the burnup of the pebble fuel from full gamma spectra (rather than specific discrete photopeaks) and found a full-spectrum ML approach to far outperform baseline regression predictions in all measurement and cooling conditions - including in operational-like measurement conditions. We also performed model and data ablation experiments to determine the relative performance impact of our ML methods' capacity to model data nonlinearities and the inherent additional information in full spectra. Applying our ML methods, we found a number of surprising results, including improved accuracy at shorter fuel cooling times (the opposite of the norm), remarkable robustness to spectrum compression (via rebinning), and competitive burnup predictions even when using background signal only (i.e. explicitly omitting known isotope photopeaks).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Comparing Ecological Risk between D and D Facilities and Operating Facilities at the Hanford Site - 19531

Many countries are faced with a large remediation task resulting from World War II, the Cold War, and subsequent industrial development. In the U.S. the Department of Energy (DOE) faces the largest hazardous waste remediation tasks, with a commitment to protect human health and the environment during and following remediation. DoE's biggest challenge is remediation of the many remaining cleanup areas at the Hanford Site, Washington. Some Hanford sites have extensive ecological resources and ecosystems that require protection or restoration. We briefly 1) describe a rapid evaluation method for ecological resources on units slated for cleanup that allows comparison among the different areas at the Hanford Site, 2) examine some D4 (deactivation, decommissioning, decontamination and demolition) facilities and Operating facilities that require remediation, 3) describe potential ecological harm to resources on these sites, and 4) compare how risks from potential harm vary at the Hanford Site. The ecological evaluation includes three phases: 1) general steps (identify resource categories, identify cleanup areas), 2) ecological descriptions (describe resources on the site to be cleaned up), and 3) ecological ratings (rate the value of the resources on a scale that can be applied across remediation sites). After completion of these steps for the sites requiring cleanup, a comparison of potential risks can be made among types of facilities (e.g. D4 facilities, legacy waste sites, operating facilities). D4 facilities are those slated for deactivation, decommissioning, decontamination and demolition, while operating facilities at Hanford include waste treatment plants, sludge basins, waste trenches, Central Waste Complex, storage facilities, and disposal facilities, among others required to support the D4 and other remediation activities, before they too are remediated. Currently the potential harm to ecological (and eco-cultural) resources on the Hanford Site for remediation at D4 facilities is very low, but increases to very high (for reactors) during remediation when there is physical disruption, increased traffic and personnel, and possible increased contamination. Following remediation, the potential harm to ecological resources is low, and the value may increase due to restoration of native vegetation on sites that were largely industrial prior to remediation. In contrast, the potential harm to ecological resources at Operating Facilities was higher currently than for D4 facilities, but was low to medium following cleanup. The higher harm (risk) at D4 facilities partly derives from the location of the reactors near the Columbia River, a resource extremely important to Tribal Nations and to other peoples in the Northwestern United States, as well as for the iconic and threatened salmon populations. Methods described in this paper provide Congress, DOE managers, regulators, Tribes, and the general public with assurance that ecological and eco-cultural resources and the environment are being protected during and following remediation. The analysis of specific sites provides managers, regulators and the public with information to make cost-effective decisions about remediation that continues to enhance ecological value while protecting human health. (authors)

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

Upgrading Fast Pyrolysis Liquids

The liquid formed by the fast pyrolysis (FP) of biomass is commonly termed bio-oil. Biomass is converted to bio-oil through FP at 400-550°C typically at 1 atm. During FP, the biomass is held at temperature for a residence time on the order of 1-2 seconds. FP is beneficial from the standpoint that a large portion of the total mass and carbon contained in the biomass is converted and retained in the liquid bio-oil fraction. However, bio-oil is acidic, viscous, and thermally unstable. Bio-oil must be upgraded in order to be considered as a liquid transportation fuel or separated and further treated to produce value-added chemicals. Bio-oil is generally regarded as difficult to upgrade due to its thermal instability and chemical complexity which is derived in part from the incorporation of highly reactive species such as aldehydes and terminal olefins as well as the presence of sugars and anhydrosugar species which undergo secondary polymerization reactions when heated. Upgrading strategies considered in this chapter include methods to stabilize and deoxygenate bio-oil through low temperature hydroprocessing to stabilize bio-oil followed by higher temperature “deep hydrotreating” to remove the majority of oxygen present; converting bio-oil catalytically (often with zeolites such as HZSM-5) prior to condensation with catalysts inside the FP reactor or in a separate fixed or fluidized bed after the FP reactor; catalytic cracking with acid catalysts such as zeolites after condensation; or through physical or chemical separations to improve bio-oil stability, extract valuable chemicals, or produce H¬2. Bio-oil upgrading by co-processing with refinery streams such as vacuum gas oil in a fluid catalytic cracking (FCC) unit in a refinery has also been investigated. Numerous excellent reviews, perspectives and other chapters have been published in the preceding 7 years.[1-10] Reviews specific to individual sections within this chapter are at the beginning of the individual sections. This chapter is intended as an update to the original article authored by Prof. Anthony V. Bridgwater, Thermochemical Processing of Biomass: Conversion into Fuels, Chemicals and Power, First Edition, Edited by Robert C. Brown, © 2011 John Wiley & Sons, Ltd. We gratefully acknowledge the original contribution by Prof. Bridgwater and have kept much of his original discussions intact with updated references. The scope of this updated chapter includes references published since about 2010. In the interest of space many of the original references included in the original version were not included. Thus, Prof. Bridgwater’s original chapter should be viewed as a companion to this updated chapter to provide a holistic set of references to the numerous bio-oil upgrading processes discussed herein.

Albrecht, Karl O.↗
DOE OSTI2023

Fuel Salt Qualification and Associated Fuel Cycle Opportunities & Issues

NRC staff have defined fuel qualification as a process which provides high confidence that the physical and chemical behavior of fuel is sufficiently understood so that it can be adequately modeled for both normal and accident conditions, reflecting the role of the fuel design in the overall safety of the facility. Uncertainties are defined so that calculated fission product releases include the appropriate margins to ensure conservative calculation of radiological dose consequences. The Advisory Committee on Reactor Safeguards (ACRS) has accepted the methodology described in NUREG/CR-7299 as an acceptable method for fuel qualification for molten salt reactors. The fuel salt qualification process described in NUREG/CR-7299 is based upon understanding the role that the fuel salt plays in achieving the fundamental safety functions, which in-turn is based on the fuel salt thermophysical and thermochemical properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Large Cask, Small Building, AREVA's Safe MLLW Disposal Solution Despite Facility Limitations - 16295

In order to begin decommissioning of a 1950's research reactor, Class C Mixed Low Level Waste (MLLW) had to be removed from the facility and disposed of. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to the building layout, crane capacity and pool size, the standard TN RAM cask loading process could not be utilized. Unique solutions and approaches had to be developed to complete the project. AREVA designed a smaller liner insert that would house the four MLLW components in a single waste package to solve the problems related to crane capacity and physical pool dimensions. AREVA also designed and fabricated a shield bell to provide radiation shielding while moving the liner insert from the reactor pool to the cask that was staged adjacent to the reactor pool. The AREVA team worked with the disposal facility to cover requirements for disposal of Class B MLLW. This included designing spacers filled with inert material to fulfill the requirement for maximum 10% void space in the waste package. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for disposing MLLW requiring macro-encapsulation in a single waste package. This solution also provides a model for how to adapt large shipping casks and liners to smaller aging commercial, DOE or research facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Response of thermal and fast-ion transport to beam ion population, rotation and T e / T i in the DIII-D steady state hybrid scenario

The thermal and fast-ion transport properties of DIII-D steady-state hybrid discharges with normalized beta β N ≳ 3 are studied at low injected torque and an increased electron to ion temperature ratio T e / T i . Linear stability analysis performed with the TGLF turbulent code indicates that a high-k mode is usually dominant at smaller radii, whereas a low-k mode is usually dominant at larger radii in these plasmas. A reduction in the net injected torque from 8.6 to 4.3 N-m leads to reduced E × B shear and hence, an enhanced turbulence that was observed on the Doppler backscattering diagnostic and was also computed with TGLF. As T e / T i in the core was increased from 0.57 to 0.66 by adding electron cyclotron current drive (ECCD) to these plasmas, higher levels of transport are observed with increased high-k modes indicated by TGLF. The fast-ion transport level varied over an order of magnitude in these discharges depending on whether Alfvén eigenmodes, fishbones, or no instabilities were observed. Hybrid plasmas with fishbones have decreased fast-ion transport, compared to plasmas with Alfvén eigenmodes, since they are resonant with a smaller portion of phase space and their resonance is farther from the wall. This reduction in fast-ion transport with ECCD mitigates the increase in turbulent transport, resulting in higher performance than expected during strong electron heating. Similarly, the lowest fast-ion transport was observed in the low torque plasma, which also led to better than expected performance at this torque value. Finally, the thermal and fast-ion transport changes observed as the torque/rotation and T e / T i are varied indicate possible methods for transferring this scenario to a reactor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Characterization of low concentration uranium glass working materials

A series of uranium-doped silicate glasses were created at (Lawrence Livermore National Laboratory) LLNL, to be used as working reference material analogs for low uranium concentration research. Specifically, the aim of this effort was the generation of well-characterized glasses spanning a range of concentrations and compositions, and of sufficient homogeneity in uranium concentration and isotopic composition, for instrumentation research and development purposes. While the glasses produced here are not intended to replace or become standard materials for uranium concentration or uranium isotopic composition, it is hoped that they will help fill a current gap, providing low-level uranium glasses sufficient for methods development and method comparisons within the limitations of the produced glass suite. Glasses are available for research use by request.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Review of capture cross sections relevant for intentional nuclear forensics

The NA-22 Intentional Forensics Venture is developing a system for tagging nuclear fuel using various methods of information encoding. One of the main methods under development is the insertion of isotopically enriched tracers into the fuel. In order to aid in the understanding of the neutronic performance of these taggants, we assess the quality of the nuclear data underpinning simulations, which are driven by the neutron-capture cross sections. We present these cross sections of naturally occurring isotopes of the elements provided in the neutron sublibrary of the planned ENDF/B-VIII.1 Feb. 2023 library release. We make this assessment using a rubric designed for this effort, which quantifies orthogonal features related to the overall quality. The quality metric highlights 6 aspects: experimental data, resonance evaluations, integral metrics, covariances, fission products, and documentation. We focus on energy ranges relevant for reactor applications. We also discuss additional sources for new, high-quality cross-section data that may be utilized on the time scale of the venture, including existing global data, new experiments, and computational methods. Finally, overall outlook is presented with conclusions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Studies of EMP generation using a Lienard-Wiechert formulation, and progress toward a 3-D EMP code

Here, we describe studies of the generation of electro-magnetic pulse (EMP) fields under the idealized conditions of a delta-function gamma pulse, a spherical cap of Compton electrons located at a common radius from the burst, and the absence of the conduction current. Various processes (geomagnetic rotation, drag, and scattering) contribute to the time derivative of the Compton current density, which appears as the source term in the Lienard–Wiechert expression for the radiation field. These processes are considered separately and in combination, and their influences on the waveform of the EMP assessed. Using a separate computational approach, different numerical models of multiple scattering are tested, in an effort to assess the validity of existing codes and earlier results. We also describe plans for and progress toward a comprehensive 3-D EMP code, which is to employ recently developed numerical methods and is to offer a variety of models, including the Lienard–Wiechert formulation described here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗