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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

THE EFFECT OF TRANSIENT ACCRETION ON THE SPIN-UP OF MILLISECOND PULSARS

A millisecond pulsar is a neutron star that has been substantially spun up by accretion from a binary companion. A previously unrecognized factor governing the spin evolution of such pulsars is the crucial effect of nonsteady or transient accretion. We numerically compute the evolution of accreting neutron stars through a series of outburst and quiescent phases, considering the drastic variation of the accretion rate and the standard disk–magnetosphere interaction. We find that, for the same long-term average accretion rate, X-ray transients can spin up pulsars to rates several times higher than can persistent accretors, even when the spin-down due to electromagnetic radiation during quiescence is included. We also compute an analytical expression for the equilibrium spin frequency in transients, by taking spin equilibrium to mean that no net angular momentum is transferred to the neutron star in each outburst cycle. We find that the equilibrium spin rate for transients, which depends on the peak accretion rate during outbursts, can be much higher than that for persistent sources. This explains our numerical finding. This finding implies that any meaningful study of neutron star spin and magnetic field distributions requires the inclusion of the transient accretion effect, since most accreting neutron star sources are transients. Our finding also implies the existence of a submillisecond pulsar population, which is not observed. This may point to the need for a competing spin-down mechanism for the fastest-rotating accreting pulsars, such as gravitational radiation.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H 2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H 2 O with the alumina reaction tube at 1700°C resulted in Al(OH) 3 formation, which reacted with the specimen to form a liquid reaction product. This behavior was not observed at lower temperatures where gas velocity and H 2 O content showed typical effects on the reaction rate. For LANL, the capabilities for oxidation testing as well as exploration of a methodology for measurement of hydrogen production of samples during oxidation under water vapor atmospheres is discussed. Results obtained for available commercial alloys are summarized, and data highlighting the performance of molybdenum, a recently proposed cladding material, are presented. Finally, leveraging of these techniques in conjunction with current and companion FCRD programs is discussed with respect to work in FY14.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of Cadmium in Undissolved Anode Materials of Mark-IV Electrorefiner

The Mark-IV electrorefiner (Mk-IV ER) contains an electrolyte/molten cadmium system for refining uranium electrochemically. Typically, the anode of the Mk-IV ER consists of the chopped sodium-bonded metallic driver fuels, which have been primarily U-10Zr binary fuels. Chemical analysis of the residual anode materials after electrorefining indicates that a small amount of cadmium is removed from the Mk-IV ER along with the undissolved anode materials. Investigation of chemical analysis data indicates that the amount of cadmium in the undissolved anode materials is strongly correlated with the anode rotation speeds and the residence time of the anode in the Mk-IV ER. Discussions are given to explain the prescribed correlation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of cadmium in undissolved anode materials of Mark-IV electro-refiner

The Mark-IV electro-refiner (Mk-IV ER) is a unit process in the FCF (Fuel Conditioning Facility), which is primarily assigned to treating the used driver fuels. Mk-IV ER contains an electrolyte/molten cadmium system for refining uranium electrochemically. Typically, the anode of the Mk-IV ER consists of the chopped sodium-bonded metallic driver fuels, which have been primarily U-10Zr binary fuels. Chemical analysis of the residual anode materials after electrorefining indicates that a small amount of cadmium is removed from the Mk-IV ER along with the undissolved anode materials. Investigation of chemical analysis data indicates that the amount of cadmium in the undissolved anode materials is strongly correlated with the anode rotation speeds and the residence time of the anode in the Mk-IV ER. Discussions are given to explain the prescribed correlation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Influence of Equation-of-States on Supercritical CO2 Combustion Mixtures

Abstract Fossil fuel based direct-fired supercritical CO2 (sCO2) cycles are gaining the attention of industry, academia, and government due to their remarkable efficiency and carbon capture at high-source temperatures. Modeling plays an important role in the development of sCO2 combustors because experiments are very expensive at the designed operating conditions of these direct-fired cycles. Inaccurate density estimates are detrimental to the simulation output. Hence, this work focuses on comprehensive evaluation of the influence and applicability of various equation-of-states (EOS) which are being used in the supercritical combustion modeling literature. A state-of-the-art supercritical combustion modeling methodology is used to simulate counter-flow supercritical CO2 flames by using various equation-of-states. The results show that using the corresponding state principle to evaluate compressibility factor is not accurate. Also, van der Waal type EOSs predictions can be as accurate as complex Benedict–Webb–Rubin EOSs; hence van der Waal EOSs are more suitable to simulate sCO2 combustor simulations. Non-ideal effects are significant under the operating conditions considered in this work. The choice of EOS significantly influences the flame structure and heat release rate. Also, assuming the binary interaction parameter as zero is reasonable in sCO2 combustion simulations.

Energy & Fuels↗
DOE OSTI2014

Super-eddington accretion in the ultraluminous x-ray source NGC 1313 X-2: An ephemeral feast

We investigate the X-ray spectrum, variability, and the surrounding ionized bubble of NGC 1313 X-2 to explore the physics of super-Eddington accretion. Beyond the Eddington luminosity, the accretion disk of NGC 1313 X-2 is truncated at a large radius (∼50 times the innermost stable circular orbit), and displays the similar evolution track with both luminous Galactic black-hole and neutron star X-ray binaries (XRBs). In super-critical accretion, the speed of radiatively driven outflows from the inner disk is mildly relativistic. Such ultra-fast outflows would be overionized and might produce weak Fe K absorption lines, which may be detected by the coming X-ray mission Astro-H. If NGC 1313 X-2 is a massive stellar XRB, the high luminosity indicates that an ephemeral feast is held in the source. That is, the source must be accreting at a hyper-Eddington mass rate to give the super-Eddington emission over ∼10{sup 4}-10{sup 5} yr. The expansion of the surrounding bubble nebula with a velocity of ∼100 km s{sup –1} might indicate that it has existed over ∼10{sup 6} yr and is inflated by the radiatively driven outflows from the transient with a duty cycle of activity of ∼ a few percent. Alternatively, if the surrounding bubble nebula is produced by line-driven winds, less energy is required than the radiatively driven outflow scenario, and the radius of the Strömgren radius agrees with the nebula size. Our results are in favor of the line-driven winds scenario, which can avoid the conflict between the short accretion age and the apparently much longer bubble age inferred from the expansion velocity in the nebula.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Radiation induced athermal diffusivity in uranium mononitride

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuel considered for light water reactors and advanced reactor designs. Properties like self- and fission gas diffusivity need to be better understood, given that they influence key fuel performance phenomena such as fission gas swelling and release. In particular, the radiation induced athermal (D 3 ) diffusivity remains challenging to accurately predict and has only been sparsely characterized in UN, despite its importance as it likely governs diffusion at the low temperatures this high-thermal-conductivity fuel form may operate. Molecular Dynamics simulations are used to estimate the mean square displacement induced by a primary knock-on atom (PKA) with a given kinetic energy. These results are combined with the PKA energy distributions obtained from binary collision approximation calculations to obtain the displacement due to a particular fission fragment. Finally, this is combined with experimental fission fragment yields to determine the displacement due to an average fission event and, thus, express the athermal diffusivity as a function of the fission rate density. These results are in excellent agreement with available experimental data. In conclusion, a particular importance is given to the understanding and the quantification of the variability of these results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Demonstration of the On-the-Fly Shielding Analysis Method: Spent Fuel and Waste Disposition

This report documents work performed supporting the US Department of Energy (DOE) Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) Integrated Waste Management activities under work breakdown structure element 1.08.02.04.01, “Data and Tools Development, Validation, and Maintenance.” In particular, this report fulfills milestone M3SF-21OR020401016, “Implement on-the-fly dose analysis methodology in UNF-ST&DARDS” within work package SF-21OR02040101, “Commercial SNF Characterization - ORNL.” The Used Nuclear Fuel - Storage, Transportation & Disposal Analysis Resource and Data System (UNFST& DARDS) enables automated dose rate calculations for spent nuclear fuel (SNF) transportation packages and storage casks using a Monte Carlo radiation transport code. The explicit method uses a detailed model of the SNF system and its contents. Therefore, a dose rate calculation is required for each as-loaded transportation package or storage cask because the SNF assemblies within a canister typically have unique irradiation characteristics. An alternate method, referred to as the “on-the-fly” shielding analysis method, has been proposed that requires only a set of Monte Carlo dose rate calculations for each transportation packaging/storage cask design. The results of the Monte Carlo dose rate calculations are independent of the SNF assembly irradiation and decay characteristics. The dose rate values may then be combined with the radiation source strength of the SNF assemblies associated with a particular transportation packaging/storage cask design to determine actual dose rates. This report presents on-the-fly dose rate calculations for a representative SNF storage cask and verification of the on-the-fly dose rate calculation results by comparison with reference dose rate calculations using the explicit Monte Carlo dose rate calculation. The on-the-fly shielding analysis method was implemented in UNF-ST&DARDS. A Python program was developed to process the MAVRIC dose rate results obtained by source particle type, energy group, and fuel geometry region. A Python processor created binary files, which were saved as a special UNF-ST&DARDS library for on-the- fly shielding analyses. UNF-ST&DARDS uses the precalculated on-the-fly binary libraries generated by the Python data processor and directly executes the Python code for on-the-fly dose analysis. This Python code unzips the pre-generated binary files mentioned above, reads the data, and combines them with user-specified sources for dose and uncertainty calculations. The Python programs were verified using Excel calculations and by comparison with the values obtained with the MAVRIC post-processing utilities applied to the 3dmap files. This method can currently be used to determine dose rates for as-loaded HI-STORM FW storage casks. The UNF-ST&DARDS analysis wizard for on-the-fly shielding analysis is described in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Machine Learning for Predictive Performance Analysis in Charged Particle Beam Tools

Imaging methods driven by probes, electrons, and ions have played a dominant role in modern science and engineering. Opportunities for machine vision and AI that focus on consumer problems like driving and feature recognition, are now presenting themselves for automating aspects of the scientific processes. This proposal aims to enable and drive discovery in ultra-low energy implantation by taking advantage of faster processing, flexible control and detection methods, and architecture-agnostic workflows that will result in higher efficiency and shorter scientific development cycles. Custom microscope control, collection and analysis hardware will provide a framework for conducting novel in situ experiments revealing unprecedented insight into surface dynamics at the nanoscale. Ion implantation is a key capability for the semiconductor industry. As devices shrink, novel materials enter the manufacturing line, and quantum technologies transition to being more mainstream. Traditional implantation methods fall short in terms of energy, ion species, and positional precision. Here we demonstrate 1 keV focused ion beam Au implantation into Si and validate the results via atom probe tomography. We show the Au implant depth at 1 keV is 0.8 nm and that identical results for low energy ion implants can be achieved by either lowering the column voltage, or decelerating ions using bias – while maintaining a sub-micron beam focus. We compare our experimental results to static calculations using SRIM and dynamic calculations using binary collision approximation codes TRIDYN and IMSIL. A large discrepancy between the static and dynamic simulation is found that is due to lattice enrichment with high stopping power Au and surface sputtering. Additionally, we demonstrate how model details are particularly important to the simulation of these low-energy heavy-ion implantations. Finally, we discuss how our results pave a way to much lower implantation energies, while maintaining high spatial resolution.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

DN/DG Screening of Environmental Swipe Samples: FY2016 Report

The Delayed Neutron Delayed Gamma (DNDG) technique provides a new analytical capability to the International Atomic Energy Agency (IAEA) for detecting undeclared nuclear activities. IAEA’s Long Term R&D (LTRD) plan has a stated high urgency need to develop elemental and isotopic signatures of nuclear fuel cycle activities and processes (LTRD 2.2). The new DNDG capability is used to co-detect both uranium and plutonium as an extension of a DN only method that is already being utilized by the IAEA for the analysis of swipes to inform on undeclared nuclear activities. Analytical method involving irradiation of swipe samples potentially containing trace quantities of fissile material in a thermal neutron field, followed by the counting of delayed neutrons, is a well-known technique in the field of safeguards and nonproliferation. It is used for detecting the presence of microscopic amounts of fissile material, (typically a linear combination of 233 U, 235 U, 239 Pu, and 241 Pu)and quantifying it in terms of the equivalent mass of 235U. The delayed neutron (DN) technique is very sensitive and is been routinely employed at the High Flux Isotope Reactor (HFIR) facility at Oak Ridge National Laboratory (ORNL). Both uranium and plutonium are of high safeguards value. However, the DN technique is not well suited for distinguishing between U and Pu isotopes since the decay curves overlap closely. The delayed gamma (DG) technique will help detect the presence of 239Pu in a mixture of U and Pu. Thus the DNDG approach combines the best of both worlds; the sensitivity of DN counting and the isotopic specificity of DG counting. The present work seeks to build on the delayed neutron and delayed gamma methods that have been developed at ORNL. It is recognized that the distribution profile of heavy fission products remains fairly invariant for the fissile nuclides whereas the distribution of light fission products varies from one isotope to another. That is, the ratio of the yield of a light fission fragment to a heavy fission fragments is isotope specific. Measurement of the ratio of the net full energy peak (FEP) from low/high mass fission products is an elegant way to characterize the fraction of fissile materials present in a mixture. By empirically calibrating the ratio of the net FEP as a function of known concentration of the binary mixture, one can determine the fraction of fissile isotopes in an unknown sample. In the work done in fiscal year (FY) 2016, samples of single fissile material isotopes as well as binary mixtures were irradiated in a well thermalized irradiation field in the HFIR. Delayed neutron counting was performed using the neutron counter at the HFIR Neutron Activation Analysis (NAA) laboratory. Delayed gamma counting was performed using a shielded high purity germanium (HPGe) detector. Delayed neutron decay curve results highlighted the difficulty of distinguishing between U and Pu isotopes, and the need for including the delayed gamma component. Based on delayed gamma spectrometry, twelve ratios of low mass/high fission product gamma ray FEP have been identified as valid candidates. Linearity of the ratios, as a function of 239 Pu fraction in 235 U+ 239 Pu mixtures, was confirmed for the low mass/high mass candidates that were selected. The DNDG method we are spearheading allows not only the presence of total fissile content to be detected, but whether the material is predominantly U or predominantly Pu, or a mixture. This provides additional SG relevant information.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Fundamental Electrochemical Properties of Liquid Metals in LiCl-KCl for Separation of Alkali/Alkaline-Earths (Cs, Sr, and Ba)

In electrorefiner, uranium is recovered from used nuclear fuel using an electrorefining process in which a metallic used nuclear fuel anode is oxidized into molten LiCl-KCl-UCl 3 electrolyte and pure U is preferentially reduced onto an inert cathode. While electrorefiner systems facilitate the recycling of substantial amounts of uranium from used nuclear fuel, they also contribute to the production of nuclear waste due to the build-up of fission products such as 90 Sr and 137 Cs in the molten salt electrolyte as they are electrochemically more active than U. The accumulation of alkali/alkaline-earth elements (Ba, Sr, Cs) in the electrolyte presents a problem as Sr and Cs isotopes have high heat densities and produce large amounts of highly ionizing radiation; these hazards combined with difficulty in removing the highly stable alkali/alkaline-earth elements from the electrolyte necessitates frequent replacement and disposal of the electrolyte, which then increases the overall volume of nuclear waste. This research project focuses on evaluating the viability of using liquid metal electrodes for electrochemical separation of alkali/alkaline-earths from molten salts utilizing the strong atomic interactions between candidate liquid metals and alkali/alkaline-earths. The strength of chemical interactions was quantified by determining the thermodynamic properties (e.g., activity) in liquid metals of Bi, Sb, and Pb. Thermodynamic properties, including activities, partial molar entropies, and partial molar enthalpies, were determined using electromotive force measurements for the Sr-Bi, Sr-Sb, Sr-Pb, Ba-Bi, and Ba-Sb binary systems in order to elucidate the strength of interactions between the alkaline-earth elements and each liquid metal, and to develop a comprehensive understanding of their behavior. Activities as low as a Sr = 10 -13 at x Sr = 0.04 at T = 888 K were measured as well as liquid state solubility as high as 40 mol% at 988 K in Sr-liquid metal systems; activities as low as aBa = 10 -15 at xBa = 0.05 at T = 888 K with liquid state solubility as high as 30 mol% at 988 K in Ba-liquid metal systems. Experimental data was used as input data towards computational efforts involving first-principles calculations as well as the CALPHAD (CALculation of PHAse Diagram) technique in the case of the Sr-Sb and Ba-Bi systems to develop improved phase diagrams and provide further basis for the use of computational models in elucidating strongly interacting binary systems. Attempts to remove Sr and Ba from molten salt electrolyte using an electrochemical cell with liquid metal cathodes were successful, with post-mortem elemental analysis of the electrodes confirming significant quantities of Sr (6.5 mol%) and Ba (12.8 mol%) deposited into Bi. Furthermore, deposition results correlated well with the deposition behavior predicted from the aforementioned electromotive force measurements, inviting the possibility of using liquid metal electrodes for selectively removing alkali/alkaline-earth elements from molten LiCl-KCl electrolyte to recycle the process salt in electrorefiner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

PdAu/YSZ composite hydrogen separation membranes with enhanced stability in the presence of CO

The effects of gold composition on the tolerance of palladium-based composite membranes to CO exposure was studied under the presence of high CO concentrations. Membranes with PdAu compositions ranging up to 41% by mass were exposed to binary 50/50 H 2 /CO gas atmospheres and both hydrogen flux and purity were monitored over time. It was observed that the highest gold composition (41%) suffered no degradation in membrane performance in terms of permeate hydrogen purity and hydrogen permeation flux during the 48 h exposure. Furthermore, for lower gold compositions, X-ray diffraction patterns suggested that carbon entered the film and could be removed by exposure to a non-CO containing gas mixture over time. However, removal of the carbon adversely affected purity; interestingly, some of the purity decline could be reversed upon re-exposure to the CO gas mixture. It is hypothesized that carbon loading in the palladium films causes lattice expansion and may even fill pore defects, thus lowering leaks through the Pd films. Furthermore, this study suggests that a membrane exposed to a CO mixture should not be cycled between CO and non-CO environments, but rather run continuously for best performance. Additionally, the choice of Pd-alloy and metal composition should be considered in regard to its carbon solubility to best predict lifetime performance.

08 HYDROGEN↗
DOE OSTI2017

Sensitivity and Limitations of Structures from X-ray and Neutron-Based Diffraction Analyses of Transition Metal Oxide Lithium-Battery Electrodes

Lithium transition metal oxides are an important class of electrode materials for lithium-ion batteries. Binary or ternary (transition) metal doping brings about new opportunities to improve the electrode’s performance and often leads to more complex stoichiometries and atomic structures than the archetypal LiCoO 2 . Rietveld structural analyses of X-ray and neutron diffraction data is a widely-used approach for structural characterization of crystalline materials. But, different structural models and refinement approaches can lead to differing results, and some parameters can be difficult to quantify due to the inherent limitations of the data. Here, through the example of LiNi 0.8 Co 0.15 Al 0.05 O 2 (NCA), we demonstrated the sensitivity of various structural parameters in Rietveld structural analysis to different refinement approaches and structural models, and proposed an approach to reduce refinement uncertainties due to the inexact X-ray scattering factors of the constituent atoms within the lattice. Furthermore, this refinement approach was implemented for electrochemically-cycled NCA samples and yielded accurate structural parameters using only X-ray diffraction data. The present work provides the best practices for performing structural refinement of lithium transition metal oxides.

25 ENERGY STORAGE↗
DOE OSTI2023

Benchmark Specifications for TREAT Tests M5, M6, and M7

Detailed information describing three TREAT tests performed on metallic fuels has been collected and organized for use as a benchmark in evaluating the performance of metallic fuel models and codes. The tests, designated M5, M6, and M7, subjected EBR-II-irradiated fuel pins to a single type of overpower transient (at prototypical conditions with full coolant flow and an exponential power rise on an 8 second period) in flowing sodium loops. Six fuel pins were tested; five were ternary (U-19Pu-10Zr) alloy fuel clad in D9 with burnup ranging from 0.8 to 9.8 at. %, and one was binary alloy fuel (U-10Zr) clad in HT9 with 2.9 at. % burnup. The information gathered from the test records is expected to be useful for pre-transient characterization of the irradiated fuel pins as well as the transient analysis of the metallic fuel when subjected to severe accident conditions. This report presents benchmark specifications for the M5, M6, and M7 TREAT tests, identifies where primary sources of benchmark-related information can be found, and includes background information to help a user of the data understand their applicability and limitations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

25-Pin metallic fuel performance benchmark case based on the EBR-II X430 experiments series

A metallic fuel benchmark case was developed for the fuel performance code BISON based on 25 uranium-zirconium and uranium-plutonium-zirconium pins of the Experimental Breeder Reactor II X430 experiment series. Results of the benchmarks were compared with measurements and calculations made at the time of the experiment as well as subsequent measurements reported in 2019. The comparisons were used to quantify the accuracy of the BISON predictions and to identify patterns in the BISON differences.BISON predicted burnup, plenum pressure, and fission gas release fractions accurately. BISON temperature predictions were somewhat cooler than the temperatures determined at the time of the experiment but appeared to be reasonably accurate considering uncertainties in the legacy temperature calculations, uncertainties in the legacy linear heat rate calculations, and the high sensitivities of the BISON-predicted temperatures to BISON inputs. Fuel axial elongation predictions had errors correlated to fuel composition; BISON tended to underpredict the elongation of binary fuels and overpredict the elongation of ternary fuels. BISON cladding radial dilation predictions were also significantly lower than legacy PIE measurements.Recommendations were made to improve the BISON fuel gaseous swelling model to account for fuel composition, to add additional capabilities to the coolant channel temperature model to ease benchmark development, and to continue developing benchmark cases based on a wide range of experiments in several reactors. Once a wide array of benchmarks is developed, an attempt can be made to enhance or calibrate BISON models to improve the cladding dilation predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Nuclear Reactions in the Crusts of Accreting Neutron Stars

X-ray observations of transiently accreting neutron stars during quiescence provide information about the structure of neutron star crusts and the properties of dense matter. Interpretation of the observational data requires an understanding of the nuclear reactions that heat and cool the crust during accretion and define its nonequilibrium composition. We identify here in detail the typical nuclear reaction sequences down to a depth in the inner crust where the mass density is $\rho =2\times {10}^{12}\,{\rm{g}}\,{\mathrm{cm}}^{-3}$ using a full nuclear reaction network for a range of initial compositions. The reaction sequences differ substantially from previous work. We find a robust reduction of crust impurity at the transition to the inner crust regardless of initial composition, though shell effects can delay the formation of a pure crust somewhat to densities beyond $\rho =2\times {10}^{12}\,{\rm{g}}\,{\mathrm{cm}}^{-3}$. This naturally explains the small inner crust impurity inferred from observations of a broad range of systems. The exception are initial compositions with A ≥ 102 nuclei, where the inner crust remains impure with an impurity parameter of Q imp ≈ 20 owing to the N = 82 shell closure. In agreement with previous work, we find that nuclear heating is relatively robust and independent of initial composition, while cooling via nuclear Urca cycles in the outer crust depends strongly on initial composition. As a result, this work forms a basis for future studies of the sensitivity of crust models to nuclear physics and provides profiles of composition for realistic crust models.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗