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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
NASA NTRS2005

Genesis Field Recovery

The Genesis mission returned to Earth on September 8, 2004 after a nearly flawless three-year mission to collect solar matter. The intent was to deploy a drogue chute and parafoil high over the Utah desert and to catch the fragile payload capsule in mid-air by helicopter. The capsule would then be opened in a clean-room constructed for that purpose at UTTR, and a nitrogen purge was to be installed before transporting the science canister to JSC. Unfortunately, both chutes failed to deploy, causing the capsule to fall to the desert floor at a speed of nearly 200 MPH. Still, Genesis represents a milestone in the US space program, comprising the first sample return since the Apollo Missions as well as the first return of materials exposed to the space environment outside of low Earth orbit and beyond the Earth s magnetosphere for an extended period. We have no other comparable materials in all of our collections on Earth. The goal of the Genesis Mission was to collect a representative sample of the composition of the solar wind and thus, the solar nebula from which our solar system originated. This was done by allowing the naturally accelerated species to implant shallowly in the surfaces of ultra-pure, ultra-clean collector materials. These collectors included single crystal silicon (FZ and CZ), sapphire, silicon carbide; those materials coated with aluminum, silicon, diamond like carbon, and gold; and isotopically enriched polycrystalline diamond and amorphous carbon. The majority of these materials were distributed on five collector arrays. Three of the materials were housed in an electrostatic concentrator designed to increase the flux of low-mass ions. There was also a two-inch diameter bulk metallic glass collector and a gold foil, polished aluminum, and molybdenum coated platinum foil collector. An excellent review of the Genesis collector materials is offered in reference [1].

McNamara, K. M.↗
DOE OSTI2023

Baseline CGMF Calculations for Pu Isotopes

Updates were made to CGMF to run neutron-induced fission calculations on Pu isotopes of A =238, 239, 240, 241, and 242 (the release version of CGMF can calculate 239,241 Pu(n,f) and 238,240,242 Pu(sf) only). Here, we show the baseline calculations from CGMF for these neutron-induced fission reactions, including updates made for total kinetic energy systematics and fission barriers fit to be more consistent with the neutron-induced fission cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Standoff Detection of Chemical Plumes Using Swept-ECQCL Remote Detection Technology: Results from EMRTC Complex Terrain Dual Tracer Experiment

As part of the Complex Terrain Dual Tracer Experiments performed in support of the LYNM program, a swept-wavelength external cavity quantum cascade laser (swept-ECQCL) remote detection system was deployed to measure chemical plume propagation with high temporal resolution. The swept-ECQCL remote detection technology was initially developed at PNNL, and is now under further development at Opticslah, LLC under SBIR funding from DOE (DOE SBIR Phase I Project DE-SC0019855, Remote Detection Technologies, Program Manager Chris Ramos). The swept-ECQCL technology provides a single-frequency infrared laser source which is rapidly scanned over a large wavelength range to measure the infrared absorption of gases, liquids, or solids. Analysis of the measured infrared absorption spectrum is used to detect, identify, and quantify the chemical species which are present. While most previous laser-based techniques used small wavelength tuning ranges to measure only a few isolated spectral lines of a few species, the swept-ECQCL provides the large tuning range needed to measure multiple species simultaneously, and to detect broad absorption features from large molecules or condensed phase materials. In addition, the swept-ECQCL systems have high enough spectral resolution and sensitivity to measure narrow lines of small molecules and determine isotope ratios. The remote detection system for gas plumes is based on the swept-ECQCL technology, and consists of a swept-ECQCL source, infrared detector, control electronics, and beam transmitting and alignment optics. The remote detection system directs the infrared output of the swept-ECQCL to a target retro reflector typically located at ~10-10,000 m distance. The swept-ECQCL wavelength is varied continuously over a large wavelength range (>1000 nm) at rates up to 1000 Hz. The swept ECQCL light reflected from the remote target is collected and focused onto an infrared photodetector. By measuring the detector signal continuously as the swept-ECQCL wavelength is varied, a time-series of absorption spectra are measured for the gases along the beam line of-sight. Based on analysis and fitting of the measured absorption spectra, the time-dependent chemical concentrations of species are determined. Multiple species and mixtures can be detected simultaneously, for all species with absorption features above the sensor noise floor. The sensor output is thus a continuous, high-speed record of identified chemicals along with their concentrations and isotope ratios of interest. The high-speed operation enables low-noise operation in conditions of atmospheric turbulence and allows tracking of fluctuating concentrations in transient chemical plumes as needed for reliable remote detection applications. All components operate with minimal cooling requirements and no cryogens are needed. In addition, the output infrared light is invisible with an intensity below the maximum permissible exposure (MPE) threshold of 100 mW/cm2 for these infrared wavelengths and thus eye safe. The measurement concept for plume detection using the swept-ECQCL remote detection system is as follows. The ECQCL beam is directed toward a remote retro-reflector, which reflects the beam back to the ECQCL system where it is focused onto an infrared photodetector. The line-of-sight from the ECQCL system to the retro-reflector defines the measurement path. The detected infrared signal is recorded continuously and detects the absorption of molecular species along the measurement path. Before a plume release, the system measures absorption from atmospheric constituents, which in this LWIR band consist primarily of lines from H 2 O and CO 2 , with additional weaker absorption from N 2 O, NH 3 , and O 3 possibly contributing. The system then measures any changes in absorption due to the tracer compounds when they enter the measurement path. The shape of absorption features is used to distinguish and identify the different tracer compounds, and the strength of the absorption is used to determine the path-integrated concentration – also called the column density.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Proposed Generalized Header File for TSLs [Slides]

This presentation discusses several recent issues have been raised over the past few years regarding inconsistent definitions for the number of scattering atoms between incoherent elastic and inelastic (MT2 and MT4 files, respectively) in the same MF7 file and isotopic distribution of primary scattering atom. It also discusses how processing codes handle these potential issues in different ways. Additionally, it discusses how the preferred solution would be automation in order to reduce possibility of user error and a proposed solution in the form of a MF7/MT1 file is examined with preliminary examples. The presentation concludes with a question-and-answer discussion.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Evaluation of B and W UO{sub 2}/ThO{sub 2} VIII Experimental Core: Criticality and Thermal Disadvantage Factor Analysis

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock and Wilcox (B and W) Spectral Shift Control Reactor (SSCR), critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO{sub 2}-ThO{sub 2} fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated. Benchmark criticality results using different cross-section libraries are reported. ENDF/B-VI.6 calculation was run using the S(α, β) model based on ENDF/B-VI.3 data while ENDF/B-VII.0 and ENDF/BVII.1 used an S(α, β) model based on ENDF/B-VII.0 data. JEFF calculations were run using the S(α, β) model based on JEFF 3.1.1 data except for the aluminum isotope, for which the S(α, β) model was missing. The maximum absolute deviation from the benchmark model was obtained using the JEFF-3.1.1 cross sections while closer results were obtained using ENDF/ B-VII.0 and ENDF/B-VII.1. Only ENDF/B-VII.1 and ENDF/B-VII.0 results are inside the 3σ uncertainty band (+2.8σ and 2.4σ, respectively), having deviations on the order of 300 pcm. The maximum deviation is -827 pcm (-6.1σ), using JEFF.3.1.1. Sample calculations of the TDF benchmark were also performed, studying the effects of the main nuclear data libraries. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Results of Speciation Determination Measurements of Insulating Sulfur Hexafluoride Gas

Under the direction of the NNSA NA-231 Mo-99 Program, SRNL has provided technical assistance to U.S. companies under cooperative agreements. One such area of technical assistance has been for the recommendation for contamination management methods in the eventuality that tritium contaminates the insulating gas of a high-voltage ion source used in the accelerator-based process developed by SHINE Medical Technologies to produce Mo-99. The SHINE accelerator process uses tritium as the target of a deuterium ion beam, where the ion beam is produced within a high-voltage ion source insulated with sulfur hexafluoride (SF 6 ), and is separated from the tritium target by an extensive pump train. Normal operating conditions preclude the conditions necessary for contamination of the insulating gas, however certain atypical conditions could result in a tritium contamination event. Accelerator based processes for medical isotope production represent an anomalous tritium contamination challenge due to the presence of SF 6 . SF 6 decomposes through normal use as an insulator gas, and the exact speciation of the decomposition byproducts depends on a variety of factors. This creates a scenario where the anticipation of species present in the insulating gas mixture is difficult. In turn, the fate of tritium within such a chemical system is unclear. This document discusses the results of measurements made in-situ of the SF 6 insulating gas in use within a SHINE Medical Technology’s high-voltage ion source. This was done in an effort to determine the chemical speciation of the gas, allowing for the recommendation of management strategies in the event that tritium contaminates the insulating gas. The results of the measurements reported in this work suggest that tritium contamination in the insulating gas may result in two chemical forms of tritium in the system, tritiated water and molecular tritium. These two forms of tritium would be captured by two methods. Tritiated water may be captured and removed from the SF 6 by an adsorbent canister already present within the pressure vessel that houses the insulating gas while in use. It may also be separated from the SF 6 during routine purification of the SF 6 , which is performed by SHINE personnel routinely. The purification cart will perform all non-embedded systems cleanup necessary to remediate tritiated species in the system. The purification cart passes the insulating gas through several filters, followed by the liquefication of the SF 6 . The filters will separate tritiated water from the SF 6 at this stage. It is expected that molecular tritium will accumulate in the gas-phase during the liquification step, which is separated and stored. Tritium may be recovered from this mixture through more conventional hydrogen purification methods such as a palladium diffuser.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1990

Gamma-Ray Observatory - The next great observatory in space

The Gamma-Ray Observatory (GRO) which is part of NASA's Great Observatories space program is presented. The GRO is equipped with the Burst and Transient Source Experiment (which detects low-energy gamma-ray photons from 20 keV to 600 keV and locates sources of gamma-ray bursts), the Oriented Scintillation Spectrometer Experiment (which detects celestial gamma rays from 100 keV to 10 MeV and identifies the elements producing these rays by measuring the ray's spectra and time variability), the Imaging Compton Telescope (which images gamma rays with energies from 1 to 30 MeV created when cosmic rays interact with interstellar matter), and the Energetic Gamma-Ray Experiment Telescope (which detects high-energy photons associated with the most energetic processes occurring in nature). After the energies of photons from each source are classified, the gamma-ray mechanisms can be modelled. Nuclei, radioactive isotopes, and nuclear reactions can be identified, and the physical conditions at the radiation's source can also be modelled. From these models, theories can be developed about the creation of elements in the explosion and collapse of giant stars, the acceleration of charged particles to velocities approaching the speed of light, and the destruction of matter and antimatter.

Neal, Valerie↗
DOE OSTI2022

MPACT Software Management Plan (V.4.3)

The MPACT code solves a discretized form of the Boltzmann transport equation on a wide variety of geometries and is distributed with a multigroup neutron cross section library. MPACT provides an advanced geometrically resolved neutral-particle transport capability to solve the flux distribution throughout the entire problem geometry, and it can model the isotopic depletion, decay, and activation of materials. The flux solution in MPACT is provided using a 2D/1D synthesis method within the framework of the 3D coarse mesh finite difference (CMFD) method for which axial and radial correction factors are obtained from 2D method of characteristics (MOC) and 1D nodal expansion method (NEM), PN, or SN. Other key characteristics of the MPACT code include the subgroup method and the embedded self-shielding method (ESSM) for resonance treatment, depletion capability based on the ORIGEN exponential matrix method, and a simplified thermal-hydraulics method for temperature/fluid feedback. The sole purpose of the simplified feedback model is to provide a mechanism for testing during code development and to provide a limited capability for educational applications. Work performed at the code level supports the VERA-QA-001, quality assurance program plan (QAPP) and VERA-QA-002, VERA Software Quality Assurance Plan.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Fast-Spectrum Critical Assemblies with a Pb-HEU Core Surrounded by a Copper Reflector

The Department of Energy invests tens of millions of dollars each year to develop the next generation of nuclear engineering modeling & simulation (M&S) tools. These M&S tools are used to analyze advanced reactor designs and the safety of current nuclear operations. As computers become more powerful, we are able to enhance resolution in our calculations. This improved resolution is taking us to a point where the limitations of simulation capability are in the quality of data, including our ability to quantify the uncertainty and sensitivity of the data. In order to accurately model systems of interest, the industry must improve key nuclear data measurements and our confidence of how well we understand the data. Thus, M&S tools need evaluated and quality-assured experimental data for validation purposes. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) compiles benchmark experiment data in a handbook that can be used by criticality safety engineers to validate computer codes and cross-section libraries at nuclear facilities. Both critical and subcritical experiments are included in the handbook. Figure 1 organizes all the benchmark evaluations that have been performed by the isotope of interest, in this case Pb, and the neutron energy within the system. Compared to other isotopes of interest for nuclear applications, there are few benchmark evaluations for Pb systems. This has caused the latest nuclear cross-section libraries to over/underestimate changes in the neutron population compared to experimental results. Therefore, this evaluation fills an important knowledge gap in benchmark evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

"Source Term Modeling for Advanced Gas Micro-Reactors"

Maintaining the safety of the public, environment, and operating personnel is the most important factor in designing, operating, maintaining, and decommissioning nuclear reactors. In recent years, there has been a growing interest in the development of micro-reactors employing TRi-structural ISOtropic (TRISO)-coated particle fuel. In gas reactors, TRISO fuel plays an important role in the safety case for high temperature reactors because of the fission product retention properties of the fuel. This ability enables the use of a functional containment strategy for the reactor where multiple barriers are used to prevent fission product release to the environment. Part of the safety analysis of these advanced reactors is the assessment of radionuclide releases under normal and accident conditions through the multiple credited safety barriers. Using conservative assumptions, a mechanistic analysis can be performed to quantify these releases that combines the probabilistic assessment of failure with analytic solutions to radionuclide transport equations. Source term modeling for TRISO fuel has been performed for previous reactor designs; however, these models are outdated, in many cases proprietary, and need updates to be applied to the current state of TRISO fuel technology and alternative gas reactor core configurations [1]. Currently, the only publicly available source term assessment for gas reactors is an expert-based Monte Carlo simulation based on the effectiveness of the fuel kernel, coating layers, and graphite block in a modular high temperature gas reactor [2]. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that could be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. The release is calculated by the diffusion of the key safety important fission products through the kernel, silicon carbide (SiC), graphite for both intact and defective TRISO particles based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system to remove fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. The model then can calculate the fission product release for any transient temperature profile and fission product releases can then be used to assess radiological dose to the workers and the public using conventional dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy (DOE) Advanced Gas Reactor (AGR) TRISO fuel development program. The model is coded in python with inputs and outputs in excel spreadsheets, as well as python plotting utilities to aid in the interpretation of the results. References: [1] INL, NGNP Mechanistic Source Term White Paper, INL-10-17997, July 2010. [2] David A. Petti, Richard R. Hobbins, Peter Lowry, Hans Gougar, “Representative Source Terms and The Influence of Reactor Attributes on Functional Containment in Modular High Temperature Gas-cooled Reactors,” Nuclear Technology, Vol. 184, p. 181-197, Nov. 2013.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS1999

Exotic Molecules in Space: A Coordinated Astronomical Laboratory and Theoretical Study

The past three years have been a period of great progress in our laboratory investigation of molecules of astrophysical interest-the most productive by far in the 20-year history of a research program which has led to the discovery of over 20% of the 123 known interstellar and circumstellar molecules. Most of the discoveries made during this period have been the result of the construction in late 1995 and early 1996 of a Fourier transform microwave spectrometer working in the centimeter-wave band. The sensitivity of this instrument from the moment that it was turned on has exceeded our expectations by an order of magnitude. The Table below shows the 46 new molecules which have been discovered. Most are carbon chains, the dominant type of molecule which has been found in space. Several comments with respect to these molecules should be made: 1. There are probably no mistakes in any of the identifications, since these have been confirmed by the standard, powerful assays and tests used to check spectroscopic identifications: isotopic substitution, quantum calculations of the expected molecular structures, detection of hyperfine structure, Zeeman effect, etc. 2. The radio laboratory astrophysics of the entire set is complete for the time being, in the sense that essentially all the astronomically interesting radio transitions (including hfs when present) are either directly measured or can now be calculated from the derived spectroscopic constants to better than 1 part per million (or 0.3 km s-1 in radial velocity, and often much better than that). 3. Six of the forty six new molecules have already been identified in space, in every case but one on the basis of our laboratory measurements. 4. Sensitive as they are, our laboratory techniques are far from fundamental limits on sensitivity, and 5. One of the principal motivations of our research is to close the fairly small mass and size gap, now only a factor of a few, between the smallest postulated interstellar grains and the largest identified interstellar molecules.

Thaddeus, Patrick↗
DOE OSTI2025

Elucidating the corrosion mechanism of Ni-based superalloys in the presence of uranium-containing chloride molten salt

The United States Department of Energy (DOE) is committed to the advancement of nuclear reactor technology through initiatives such as the Advanced Reactor Development Program (ARDP), in an effort to diversify the United States energy portfolio towards more sustainable energy options. The ARDP includes demonstration by industry partners of molten chloride fast reactors (MCFRs). Construction of MCFRs requires qualified nuclear structural materials. Unfortunately, there are no current materials that are fully qualified by the Nuclear Regulatory Commission for the construction of molten salt reactors, including MCFRs. Adapting current structural material qualifications requires expansion of our current knowledgebase on the property-performance relationships regarding corrosion performance. In this investigation, we assess microstructural changes in a Ni-based superalloy after exposure to a UCl3¬-containing chloride salt eutectic mixture through a correlated multi-modal approach combining several advanced characterization techniques, including scanning electron microscopy/focused ion beam (SEM/FIB) and transmission electron microscopy (TEM). SEM/FIB analysis will illustrate changes in elemental composition, microstructure, and isotopic information acquired from energy x-ray dispersive spectroscopy (EDS), electron backscatter diffraction (EBSD), and secondary ion mass spectroscopy (SIMS), respectively. This information will then aid in identifying localized regions to elucidate the corrosion mechanism with TEM through a combination of electron diffraction, electron energy loss spectroscopy (EELS), and additional EDS. The findings from this investigation will further expand our assessment of the corrosion performance of structural materials in molten salt chloride systems, aiding to developing fully qualified materials for construction of MCFRs.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Integral Experiment Request 554: CED-2 Summary Report

This report summarizes the critical experiment design 2 (CED-2) or final design study for Integral Experiment Request 554 (IER-554). This experiment describes the effects and importance of adding a commercially available neutron absorber material to a known light-water low-enriched uranium assembly. The assembly in question is the Seven Percent Critical Experiments (7uPCX) at Sandia National Laboratories. The neutron absorber plates considered for the experiment are known as Boralcan and are made of boron carbide (B 4 C) particles embedded in 1100 aluminum alloy. The concentrations of B 4 C and 1100 aluminum were changed as well as the thickness and the positions of the plates, and the fuel rod configuration was adapted to ensure that the assembly would be critical in each case studied. A total of 6 critical configurations, each with a neutron absorber plate inserted, are described in this report. No results of high-quality integral experiments involving neutron absorbers made with B 4 C and 1100 aluminum plates are currently publicly available. Sensitivity to the neutron absorber plate material definition, isotopes, and cross sections of specific regions and configurations are also analyzed. As part of the final design of the experiment, a safe plate holder made of 3003 aluminum was designed. The study results indicate that these experiments are achievable with sufficiently low uncertainties and minimal modification to the assembly, and the CED-3 phase can start when funding allows.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1994

Main characteristics of the COMET/COMRADE experiments

Both the COMET (Collection en Orbite de Matiere Extra Terrestre) and the COMRADE (Collection of Micrometeorites, Residue and Debris Ejecta) programs are developed to the collection and analysis of the particles of various origins orbiting around the Earth at low altitudes (between approx. 300 and approx. 500 km). The COMET experiment is more specifically designed to be flown for a short period of time (a few days), in concordance with a meteor stream crossing the Earth. Thus, it results in a considerable enrichment in the collection of grains related to a given comet. The COMRADE experiment has been selected as a proposal for long-duration flights (a few months), in order to gain information on all sizes of particles present on low Earth orbits, including submicrometer grains. It has been accepted by ESA authorities for use on the EURECA 2 platform. The objectives of these studies are multiple. The use of passive detectors gives access to the chemical and isotopical properties of the grains in the micrometer size range, by analyzing either the particle remnant mixed with the target material, or the intact particle captured in a specific low-density material. The particle remnants of the micrometer-sized extraterrestrial grains, having impacted on purposely designed metallic collectors, are identified for complete and detailed chemical, isotopic, and organic analysis, thereby determining grain composition as well as the existence of organic and inorganic molecules, to be related with the possible cometary origin of the grains. Micrometer/submicrometer dust grains are also captured in a manner that ensures minimal particle degradation. The captured intact particles are returned to Earth for complete and detailed chemical, isotopic, spectral, mineralogical, and organic analysis.

Borg, J.↗
DOE OSTI2026

Fuel Fabrication Specification Impact Analysis for NBSR LEU Conversion

As part of a national initiative to enhance nuclear security and reduce proliferation risks, significant efforts have been undertaken by the National Nuclear Security Administration Material Management and Minimization Office of Reactor Conversion Program to convert U.S. high performance research reactors (USHPRRs) from the use of highly enriched uranium (HEU) to low-enriched uranium (LEU), including the National Bureau of Standards Reactor (NBSR). The current plan is to procure LEU fuel assemblies from commercial fabricators according to fuel specifications tailored for each USHPRR. The analysis conducted at Brookhaven National Laboratory was part of an effort to identify the sources of uncertainty in the fuel specifications that may impact the performance of the NBSR core after its conversion and, in particular, to assess the range of acceptable tolerance limits from the perspective of core safety and reactor performance. Using the stochastic neutronics code MCNP 6.2, the variations in important NBSR neutronics characteristics were analyzed as a function of the specification parameters independently and in combination. The important NBSR specification parameters analyzed were the fuel isotopic composition, the amount of impurity content in cladding, the fuel plate thickness, and the fuel element 235U mass loading. The range of variation of each specification parameter was based on the technical specification limit or available as-fabricated assay data and uncertainties. The NBSR neutronics characteristics selected for analysis were the reactor reactivity characteristics at equilibrium and the equilibrium fuel cycle length. Results show that with variations in the fabrication parameters of the as-fabricated U-10Mo fuel within the specification limitations, the excess reactivity of the NBSR LEU core remains well below the 15% Δk/k technical specification limit, and the shutdown margin is always significantly greater than the required 0.68% Δk/k. This ensures that the NBSR can be operated safely and reliably shut down for all analyzed cases within the specified fabrication limits after the LEU conversion. In the prototypic case, the fuel cycle length was 1.5 days longer than the targeted 38.5 days. In a credible worst-case scenario, where all low-reactivity parameters were combined, the fuel cycle length was reduced to 35.5 days, which is still considered manageable for reactor operations. Variations in cycle length are primarily driven by changes in 235U loading, with other parameters having secondary effects.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Extreme Temperature Sample Environment for Materials Research Using Neutron Scattering

Extreme temperatures and non-equilibrium materials are both central to Department of Energy’s (DOE) mission of addressing grand challenges in emerging energy and advanced materials technologies. The goal of this project is to develop and perfect a system for studying materials in extreme and non-equilibrium conditions using neutron scattering. Neutron scattering has many advantages in the study of materials structure including the ability to use isotopic substitution to determine partial structure factors. When used at a pulsed neutron source, the potential for event-based detection can enable a stroboscopic technique to interrogate very short lived metastable states in supercooled liquids. The work performed in this Phase I project included: (i) literature reviews and discussions with beamline scientists to establish key requirements for the instrument, (ii) laboratory-based research at Materials Development, Inc. (MDI) to test design concepts, modify control software and model designs, (iii) feasibility experiments and measurements on a prototype sample at the Spallation Neutron Source, and (iv) analysis of the results and design of the proposed Phase II instrument. The results of the work show that the proposed approach is viable and provides a strong basis for the Phase II instrument design and Phase III commercialization. MDI participated in the DOE Commercialization Assistance Program with the Larta Institute (Larta) to accelerate commercial sales of the instruments. The market research identified a served available market of ~13M including sales to beamlines, university research laboratories, and industrial glass and materials businesses. Over the ca. 10 year product life cycle, MDI expects to sell instruments with a cumulative net present value profit of approximately 5M.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Gaussian Process Optimization of Sensitivity-Based Similarity Metrics between New Nuclear Applications and New/Existing Benchmarks [Slides]

This presentation discusses Nuclear Criticality Safety (NCS) and how designing safe, new nuclear criticality experiments requires expert judgement, which could take years of experience. Sensitivity/uncertainty (S/U) analysis can be utilized by less experienced individuals to conservatively estimate uncertainties in important parameters, such as k eff , in newly proposed nuclear experiments. The presentation poses the question of how this analysis can be performed and states that the answer lies in matching new nuclear experiments with existing benchmark experiments using similarity metrics. By increasing the criticality safety of the application in this work, higher mass limits could be used in PF-4 operations. Additionally, the presentation discusses MCNP6.2®, Whisper-1.1, the software that can be used in this analysis. Also discussed is the fact that International Criticality Safety Benchmark Evaluation Project (ICSBEP) benchmarks rarely match new nuclear applications and that there are significant differences in given set of materials and/or geometry. If there are no benchmarks that match the application, the presentation discusses the possibility of creating new benchmarks. In conclusion, this work presents a Gaussian process (GP) optimization scheme that was used to generate new benchmarks with the highest sensitivity-based similarity metrics to user-defined nuclear applications. The Gaussian process optimization successfully designed 3 new experimental benchmarks that were highly correlated to the application of interest and had k eff values near critical. Optimization over c k,i-r has shown that investigating specific isotope reactions for different applications is crucial to designing benchmark experiments. Partial contribution from Pu dominates c k similarity metric. Future work includes testing new stand-alone similarity metrics or new combinations of similarity metrics as the design criterion of this optimization – design criterion is application dependent.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗