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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Synthesis and Crystal Structures of Volatile Neptunium(IV) β-Diketonates

There is a large need for neptunium certified reference materials to support domestic nuclear forensics programs. These reference materials are required for quantification of Np forensic related samples. Isotopic separation is required to produce this material; traditional isotopic separation methods use electromagnetic separation techniques. Such methods require a gaseous introduction of Np into the separation source. This work sought to produce a volatile Np compound that would be suitable for the purpose of certified reference material production. ß-diketonates were identified as compounds likely able to serve this purpose. However, previously described synthesis requires NpCl4, which is not commercially available and prohibitively difficult to produce. A novel and facile procedure was developed here to successfully synthesize several tetravalent ß-diketonates of Np. They were characterized using FTIR and single crystal x-ray diffraction. The volatility of these compounds at elevated temperatures was then analyzed, and discussed. The synthesis procedure developed here presents a facile platform for further Np organometallic studies.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Validation of Continuous-Energy ENDF/B-VIII.0 16 O, 56 Fe, and 63,65 Cu Cross Sections for Nuclear Criticality Safety Applications

Recently completed cross-section evaluations sponsored in part by the Nuclear Criticality Safety Program were incorporated into the 2018 release of the ENDF/B-VIII.0 cross-section library. Evaluated isotopes of interest to the nuclear data and criticality safety community include 16 O, 56 Fe, and 63,65 Cu. For performance validation, benchmark models defined in the International Criticality Safety Benchmark Evaluation Project Handbook were selected based on energy-integrated k eff sensitivities to total cross sections of interest and compared with experimental values. Of the 102 benchmark configurations that were utilized, 63 are sensitive to 16 O, 32 sensitive to 63,65 Cu, and 25 sensitive to 56 Fe. Selected benchmarks were modeled in SCALE 6.2.3 Criticality Safety Analysis Sequence (CSAS) continuous-energy Monte Carlo k eff calculations with ENDF/B-VII.1, with a hybrid ENDF/B-VII.1 with ENDF/B-VIII.0 data substituted for individual isotopes of interest, and with ENDF/B-VIII.0. ENDF/B-VIII.0 showed improved agreement with experimental k eff for 56 Fe, 63 Cu, elemental copper, and full library substitution while producing lessened agreement for 16 O and 65 Cu. Furthermore, with full library and isotope-specific ENDF/B-VIII.0 performance, a best-case ENDF library was formed by excluding underperforming isotopes’ ENDF/B-VIII.0 data, reverting 16 O and 65 Cu cross sections to ENDF/B-VII.1. This resulted in the average relative deviation between calculated and experimental data improving from 1.45σ for the ENDF/B-VIII.0 library to 1.32σ for the best-case library, relative to benchmark uncertainty.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1992

Toward the last frontier - A strategy for the evolutionary development of space nuclear power systems

A number of exciting mission opportunities are being considered for the 21st century, including advanced robotic science missions to the outer planets and beyond, human exploration of the Moon and Mars, and advanced space transportation systems. All of these missions will require some form of nuclear power; however, it is clear that current budgetary constraints preclude developing many different types of space nuclear power systems. This paper reviews the specific civil space missions which have been identified, the power levels and lifetimes required, and the technologies available. From this an evolutionary space nuclear power program is developed which builds upon the experience of radioisotope thermoelectric generators, dynamic isotope power systems, and space nuclear reactors. It is strongly suggested that not only does this approach make technical and budgetary sense but that it is consistent with the normal development of new technologies.

Gary L Bennett↗
DOE OSTI2017

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED URANIUM (LEU) FUEL

The United States Department of Energy (DOE), National Nuclear Security Administration (NNSA), Office of Material Management and Minimization (MMM) is working to convert research reactors globally from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its predecessor programs have converted or verified the shutdown of 88 HEU research reactors and isotope production facilities, using LEU fuel developed in the 1980s. However, there is a small set of high performance research reactors which require a new high density LEU fuel for conversion, including six U.S. high performance research reactors (USHPRRs), which include the Advanced Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor (MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor (MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA. These reactors will share the same uranium-molybdenum (U-Mo) fuel plate design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with aluminum cladding. Three different application methods for the Zr diffusion barrier layer are being pursued – co-rolled, electroplating and plasma spraying. Fabrication of the higher density LEU U-Mo fuel has presented several challenges. Because the U-Mo fuel fabrication process uses the existing stockpile of 93% highly enriched uranium (HEU) to be blended with molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of enriching to 19.75% will not be realized. Homogeneity of the U-235 and molybdenum in the metal foil is a challenge due to U-Mo casting process results in variations between adjacent cast U-Mo plates. LEU will not require less U-235 as previously believed due to self-shielding properties of U-238 requiring additional U-235 to be loaded in each fuel element. Although the overall dimensions of the ATR fuel element remain unchanged, the weight of the LEU ATR fuel element is 72% greater than a HEU ATR fuel element. During waste recovery and recycling efforts, separation of aluminum cladding from the HEU fuel meat is a relatively simple chemical dissolution, however, separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly more challenging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

R-matrix Resolved Resonance Region Evaluation of 140,142 Ce

Oak Ridge National Laboratory completed the resolved resonance region (RRR) evaluation of the two most abundant cerium isotopes, 140 Ce (88.45%) and 142 Ce (11.11%), as requested by the US Nuclear Criticality Safety Program. These evaluations are based on recent high-resolution transmission and capture measurements performed on nat Ce and highly enriched 142 Ce samples at the JRC-Geel Linear Accelerator facility, as well as measured thermal constants available from the EXFOR database. Starting from the resonance parameters of the ENDF/B-VIII.0 library followed by a preliminary R-matrix analysis, an updated set of resonance parameters and corresponding covariance information were derived by fitting these measured data using the Reich–Moore approximation of the R -matrix theory, as implemented in the SAMMY code system. The 140 Ce RRR upper energy limit was kept at 200 keV, whereas the 142 Ce resonance region was extended from 13 to 26 keV. Updated statistical properties were obtained for the new evaluations and compared to those derived from the ENDF/B-VIII.0 nuclear data library. The new evaluation work improved some of the discrepancies found in previous work, such as the capture resonance integral and stellar Maxwellian-averaged cross sections. These integral quantities were mainly derived from the fit of the latest measured data, especially the neutron capture yield data for 142 Ce isotope.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Analysis and Design of Ex-Service Garter Spring Irradiation Capsules in the High Flux Isotope Reactor

Nickel-based super alloys such as Inconel X-750 are used to fabricate internal components for various reactor systems. Recent efforts to research and understand the effects of long-term neutron radiation damage on this material are crucial for quantifying the performance (radiation-induced creep and embrittlement) of such materials under extended use. The CANDU® ex-service garter spring irradiation program was instituted to investigate such effects on various types of Inconel X-750 garter springs, tensile specimens, and fatigue specimens. Conventional irradiation programs use rabbits, which are capsules that are generally loaded with non radioactive irradiation specimens. These rabbits are designed to realize various specimen temperatures while irradiated to high neutron fluences in the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR). To accelerate the irradiation schedule of this program, two types of material — unirradiated and previously irradiated ex-service garter spring material — were both included in the irradiation program test matrix. These two material conditions are expected to provide intermediate and end-of-life neutron dose experimental results concurrently to accelerate the overall irradiation program schedule. This document describes the B8J18 garter spring design and the supporting work required to successfully fabricate, assemble, and qualify subsequent irradiation capsules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Engineering the Big Chill: The story of JLab’s Central Helium Liquefier

This article tells the story of the Central Helium Liquefier (CHL) at the Thomas Jefferson National Accelerator Facility (JLab), one of the US National Laboratories. JLab’s successful superconducting radio frequency accelerator was only possible because a group of JLab engineers successfully tackled a complex of difficulties to build a cryogenic system that included the CHL, a task that required advancing the frontier of cryogenic technology. Ultimately, these cryogenic advances were applied far beyond JLab to the benefit of cutting-edge programs at other US national laboratories (Oak Ridge, Brookhaven, and the Facility for Rare Isotope Beams at MSU) as well as NASA. In addition, this innovation story dramatizes the sort of engineer-driven technological problem solving that allows the successful launch and operation of experimental projects. Along the way, the CHL story also provides an important addition to our understanding of the role played by engineers and industry in creating knowledge at physics laboratories.

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Activation Mechanism and Surface Intermediates during Olefin Metathesis by Supported MoO x /Al 2 O 3 Catalysts

The activation of supported MoO x /Al 2 O 3 catalysts and the resulting surface intermediates were examined with in situ DRIFTS during olefin metathesis reaction conditions, aided by C 3 D 6 -C 3 H 6 isotopically labeled switching, C 2 H 4 -C 4 H 8 titration, and temperature programmed reaction experiments. The studies shows that activation by propylene of the surface MoO x sites on Al 2 O 3 initiates by forming surface isopropoxide species that subsequently dehydrogenate to acetone and surface Mo-OH. Desorption of acetone from the catalyst surface reduces the Mo +6 sites to Mo +4 sites, creating a vacancy for coordination of the next CH 2 =CHCH 3 molecule. Oxidative addition from subsequent adsorption of propylene on the surface Mo +4 sites results in formation of surface Mo=CH 2 and Mo=CHCH3 reactive intermediates, oxidizing the reduced surface molybdena sites back to Mo +6 . This study demonstrates the activation mechanism and surface intermediates during olefin metathesis by supported MoO x /Al 2 O 3 catalysts under reaction conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

The DarkSide project

DarkSide is a graded experimental project based on radiopure argon, and is now, and will be, used in direct dark matter searches. The present DarkSide-50 detector, operating at the Gran Sasso National Laboratory, is a dual-phase, 50 kg, liquid argon time-projection-chamber surrounded by an active liquid scintillator veto. It is designed to be background free in 3 years of operation. DS-50 performances, when filled with atmospheric argon, are reported. However DS-50 filled with underground argon, shows impressive reduction of the 39Ar isotope. The application of this powerful technology in a future generation of the DarkSide program is discussed.

Agnes, P.↗
DOE OSTI2018

Fission Induced Electromagnetic Response

The FIER (Fission Induced Electromagnetic Response) code is a C++ program designed to predict delayed gamma-ray spectra following fission. Given a fissioning isotope and irradiation scheme, FIER will determine the decay chains for each fission fragment and their daughters. Each product’s time-dependent population is calculated using an analytical solution to the Bateman equations that guide decay transmutation. These populations are then used to calculate the number of gamma-ray emissions for a given counting scheme. A Monte Carlo method is also available to determine the uncertainty of the fission product populations and delayed gamma-ray intensities based on the uncertainties on the input nuclear data.

Matthews, EricFrancis [University of California - ↗
DOE OSTI2015

Recommendations for the generation, quantification, storage and handling of peptides used for mass spectrometry-based assays

The Clinical Proteomic Tumor Analysis Consortium (1) (CPTAC) of the National Cancer Institute (NCI) is a comprehensive and coordinated effort to accelerate the understanding of the molecular basis of cancer through the application of robust technologies and workflows for the quantitative measurements of proteins. The Assay Development Working Group of the CPTAC Program aims to foster broad uptake of targeted mass spectrometry-based assays employing isotopically labeled peptides for confident assignment and quantification, including multiple reaction monitoring (MRM; also referred to as Selected Reaction Monitoring), parallel reaction monitoring (PRM), and other targeted methods.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2018

Performance Evaluation of the CHKED Multi-Elemental Analysis and Software

The Hybrid K-edge Densitometer (HKED) is a high precision safeguards assay system developed by Ottmar and Eberle [1] for quantification of the U and Pu concentrations of dissolver solutions in spent nuclear fuel reprocessing. The HKED measurement supplements traditional destructive assay methods such as Isotope Dilution Mass Spectrometry (IDMS) and is an essential component of the international safeguards program implementation at these facilities. The Canberra Hybrid K-Edge Densitometry (CHKED) software was recently developed by Canberra Industries to better address the changing composition of the nuclear fuel cycle by addition of a multi-elemental analysis capability to the K-Edge transmission measurement. Our evaluation of the CHKED software was performed at the request of the International Atomic Energy Agency (IAEA) to examine the functionality and performance of the CHKED software and algorithms and identify potential areas for improvement of the software. The reader is assumed to be familiar with the HKED technique and its current implementation in safeguards assay of U and Pu in the nuclear fuel cycle. The Canberra Hybrid K-Edge Densitometry CHKED software replicates functionality of the older VMSbased Hybrid K-Edge Densitometer (HKED) software and incorporates additional analysis algorithms and tools. When the traditional HKED analyses are selected, the CHKED software appears to faithfully reproduce the measurement results as reported by the older software. The CHKED software includes a Multi-elemental K-Edge (MEKED) analysis based in part on algorithms devised at Los Alamos National Laboratory. In principle, the MEKED analysis represents a significant advancement over the traditional HKED analysis, providing a capability to analyze the actinide concentrations from complex solutions, such as low U:Pu ratios (e.g., U:Pu = 1:1) including additional minor actinide components such as Np and Am. However, this MEKED analysis is somewhat limited by reliance on a region of interest (ROI) analysis of the spectrum, the need for extensive calibration measurements and a simplistic interpretation of the KED transmission spectrum. Finally, the MEKED algorithm set associated with the CHKED software (V1.0) acquired by Oak Ridge National Laboratory (ORNL) from Canberra Industries does not seem to be “finished” (e.g., the MEKED analysis seems to only determine measurement results for two actinides with while results for the minor actinides appear to be highly dependent on declaration values).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Geochemical Changes in Response to CO 2 Injection in a CO 2 -EOR Complex in Northern Michigan

The overall purpose of the geochemical monitoring program under Midwestern Regional Carbon Sequestration Partnership (MRCSP) is to use stable and radiogenic isotope geochemistry in concert with analysis of general geochemical parameters of fluids and gases and analysis of core samples to determine geochemical processes occurring in the reef structure because of CO 2 injection..

54 ENVIRONMENTAL SCIENCES↗