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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Neutronic impact of Autonomous Reactivity Control system inclusion in a breed-and-burn core

This work examines the impacts induced in a large breed-and-burn core through the inclusion of Autonomous Reactivity Control (ARC) systems from the neutronics perspective. A combination of numerical and analytical techniques are used to conservatively quantify the reactivity penalty, amount of additional tritium production, and depletion of the system absorber fluid. The reactivity penalty is found to be negligibly small, with the largest impact resulting from the removal of fuel material to make room for the ARC system tubes. Similarly, the depletion of the absorber fluid is found to be negligible compared to the amount of fluid present. It is found that the production of tritium within the active core is also small in comparison to that which is nominally produced, while tritium production within the lower ARC system reservoir may be of concern if not adequately considered during the assembly design phase. This issue is explored and a mitigation technique is proposed. Ultimately this study concludes that there are no issues from the steady-state neutronics perspective preventing incorporation of ARC systems into typical breed-and-burn cores.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Modeling The DLOFC Accident Scenario of HTR-PM Equilibrium Core Using NEAMS Tools

High-Temperature Gas-cooled Reactors (HTGRs) have excellent characteristics in terms of safety and high thermal efficiency, and they are gaining a large interest from the industry as a candidate of Gen-IV reactors for a wide range of applications. The High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM) is one of these designs and where helium gas is used to cool the pebble-bed region that consists of spherical fuel elements moderated with graphite. The HTR-PM design is based on the combined experience from the German pebble-bed reactor program from the 1960s through the 1990s and the HTR-10 experience in China during the 2000s. Idaho National Laboratory has a long experience in modeling of HTGRs working in developing neutronics and thermal hydraulics tools for the proper modeling of these reactors. The neutronics code Griffin has the capability to model pebble depletion . While the thermal hydraulics code Pronghorn was developed mainly to model the pebble bed reactors with the porous media assumption. In this work, an equilibrium core Multiphysics model was developed for the HTR-PM reactor to analyze the depressurized loss of forced cooling accident scenario (DLOFC). This paper is organized as follows: First, a brief description of the reactor and model specifications are provided. Then, the developed Multiphysics model is discussed. Finally, verification results of the steady-state equilibrium core and DLOFC accident are presented followed by a summary of the conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Linear source method of characteristics in HELIOS2

The Studsvik Scandpower code HELIOS2 is a fast lattice transport code that has gained wide use for reactor fuel cycle analysis in commercial, regulatory, and research environments. It contains two transport solvers - Collision Probabilities (CP) and Method of Characteristics (MoC) - with a general geometry ability and generalized depletion for a variety of fuel types, experimental facilities, and isotope production applications. The HELIOS2 lattice code has been extended with the development of a linear source MoC option. The linear source equations are derived within the scope of the HELIOS2 general geometry and non-cyclic ray tracing. The implementation was tested for fine and coarse spatial mesh models of the 2D C5G7 MOX computational benchmark, which provides a pure and extreme test of a transport solver. The linear source yields at least comparable accuracy with reduced computational resource requirements relative to the flat-source approximation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRSDate not supplied

Refractory Carbides for Hydrogen Erosion Resistance in Carbon Tubes for Nuclear Thermal Propulsion

As crewed travel becomes more common to the Moon and beyond, space nuclear propulsion-based craft offer reduced travel times and double the specific impulse compared to conventional chemical rockets. The nuclear core and its integral hydrogen propellant channels will require advanced refractory carbide coatings over a carbon/carbon substrate for hydrogen resistance. Using ab initio and other computational techniques, we studied select (Zr/Ta/Nb)C mixed carbide systems for erosion resistance and hydrogen diffusion using realistic levels of carbon vacancies and experimentally obtainable stoichiometries. Various contributions to the formation and hydrogen erosion free energies were calculated, including configurational, electronic, and vibrational terms. Our calculations indicate that ZrC is less likely to lose carbon than NbC or TaC, providing an explanation for heritage NERVA data on ZrC and NbC mass loss performance. Additionally, binary mixtures show compositions that are more stable than their end-member counterparts. Hydrogen diffusion in stoichiometric ZrC was found to be slightly higher than previously theorized, with the caveat that hydrogen saturation of carbon vacancies will hinder further hydrogen diffusion through the bulk, while hydrogen diffusion in ZrNbC is slightly enhanced over ZrC. And finally, we show that as carbon is depleted, the deformation mechanism of ZrC changes from brittle to ductile. Though many questions still remain as to the bulk performance of these materials, these trends and results are important in terms of materials selection efforts for hydrogen-resistant coatings on carbon/carbon substrates.

William C Tucker↗
NASA NTRSDate not supplied

Refractory Carbides for Hydrogen Erosion Resistance in Carbon Tubes for Nuclear Thermal Propulsion

As crewed travel becomes more common to the Moon and beyond, space nuclear propulsion-based craft offer reduced travel times and double the specific impulse compared to conventional chemical rockets. The nuclear core and its integral hydrogen propellant channels will require advanced refractory carbide coatings over a carbon/carbon substrate for hydrogen resistance. Using ab initio and other computational techniques, we studied select (Zr/Ta/Nb)C mixed carbide systems for erosion resistance and hydrogen diffusion using realistic levels of carbon vacancies and experimentally obtainable stoichiometries. Various contributions to the formation and hydrogen erosion free energies were calculated, including configurational, electronic, and vibrational terms. Our calculations indicate that ZrC is less likely to lose carbon than NbC or TaC, providing an explanation for heritage NERVA data on ZrC and NbC mass loss performance. Additionally, binary mixtures show compositions that are more stable than their end-member counterparts. Hydrogen diffusion in stoichiometric ZrC was found to be slightly higher than previously theorized, with the caveat that hydrogen saturation of carbon vacancies will hinder further hydrogen diffusion through the bulk, while hydrogen diffusion in ZrNbC is slightly enhanced over ZrC. And finally, we show that as carbon is depleted, the deformation mechanism of ZrC changes from brittle to ductile. Though many questions still remain as to the bulk performance of these materials, these trends and results are important in terms of materials selection efforts for hydrogen-resistant coatings on carbon/carbon substrates.

William C Tucker↗
DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11 (Rev. 1)

The Accelerated Basin De-inventory (ABD) program involves discarding spent nuclear fuel that is currently stored in L-Basin to the Defense Waste Processing Facility (DWPF) for vitrification. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. Savannah River Mission Completion has requested that the Savannah River National Laboratory assess the technical gaps related to the increased gadolinium poisoning requirement and the impacts of performing the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51 with H-Canyon discards present. The following summarizes the evaluation of the impacts of increasing the quantity of gadolinium (and related topics) from what was previously evaluated in the SRNL studies of gadolinium-poisoned ABD material solubility, the overall ABD flowsheet review, and increasing the fissile mass loading in glass: 1) Based on literature surveys, there is no indication that organic interactions with gadolinium will be significant at the high pH (typically >13) conditions of the Concentration, Storage, and Transfer Facilities. Any interactions of gadolinium with organics in DWPF are not expected to adversely impact DWPF or downstream facilities. Thus, there is little-to-no residual risk from organic interactions with gadolinium [Gap closed]; 2) Adding depleted uranium to ABD material, targeting 235 U enrichment of 4.90% within each transfer window, will mitigate potential impacts from an increase in soluble 235 U enrichment during sludge washing and LTAD. The plan to take advantage of previous transfers and allow 235 U enrichment of >5% during the final transfer window carries a risk that Tank 51 supernate will have a 235 U enrichment of >5%, which should be evaluated for acceptance; 3) Increasing the gadolinium mass ratio to 3.0:1 Gd: 235 U(eq SLU ) should lead to the same or higher partitioning of gadolinium into the solid phase within the DWPF Chemical Process Cell, resulting in both liquid and solid phases with expected partitioning of Gd consistent with the prior solubility study [Gap closed for SB11]; 4) There are no expected impacts on DWPF melt temperature and melter operations due to the minimal ~0.2 weight percent (wt%) increase in Gd concentration relative to previous sludge batches [Gap closed for SB11]; 5) As observed previously, Gd is expected to enter the off-gas system via physical entrainment, but at a slightly higher concentration than what was observed for SB9 melter off-gas pluggage deposits (0.07 wt%) [Gap closed for SB11] ; 6) There are no expected impacts on DWPF recycle or the Recycle Collection Tank glycolate destruction process. [Gap closed for SB11]; 7) Gd is projected to be a trace component in the SB11 glass (<0.5 wt%) and can be ignored for process control. Trace components do not significantly impact glass durability, thus the conclusions of the previous Product Consistency Test evaluation at a fissile mass loading of 2,500 g fissile/m3 glass still applies to SB11. The ~0.1 wt% increase in Gd2O3 concentration relative to the previous study will not impact the predictability of SB11 glass with the DWPF Product Composition Control System (PCCS) models for durability or the acceptability of glass according to the Waste Acceptance Product Specifications (WAPS) criterion for product consistency [Gap closed for SB11]; 8) No additional Toxicity Characteristic Leaching Procedure testing is necessary for SB11 and the hazardous waste specification of the SB11 DWPF waste form is unchanged after the addition of the ABD stream [Gap closed for SB11]. The following summarizes the evaluation of the impacts of adding two-thirds of the ABD material to Tank 51 prior to LTAD: 1) The addition of two-thirds of the ABD increases overall aluminum mass from 1.39×10 4 kg to 1.64×10 4 kg (15.5% ABD Al). The form of the insoluble portion of the Al resulting from ABD addition should be the more readily dissolved Al(OH) 3 and amorphous forms. The portion of the ABD aluminum that is processed by LTAD is expected to be completely soluble, thus requiring that less of the boehmite in the sludge be dissolved to reach the same Al target in the SB. [Gap closed for SB11]; The expected LTAD impact on other components, as related primarily to the components in ABD, are discussed. Gd is expected to remain insoluble during LTAD and not impact the solubility of other components. [Gap closed for SB11]; The addition of two-thirds of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 13,100 kg (63% ABD U) and the projected plutonium mass from 86.0 kg to 89.5 kg (3.9% ABD Pu). The addition of all of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 16,100 kg (70% ABD U) and the projected plutonium mass from 86.0 kg to 90.4 kg (5.3% ABD Pu). The 235 U enrichment will be ≤5%. The fissile uranium will be adequately poisoned by Gd and the fissile Pu will be adequately poisoned by Fe from the sludge. [Gap closed for SB11]; There is a low risk that ABD addition will impact the rheology or pumpability of the slurry. There is a low but higher risk of ABD addition prior to LTAD impacting the settling rate; Based on the evaluation of adding two-thirds of the ABD material and all of the ABD material prior to the LTAD process, there is no volume or mass limit that would need to be imposed on ABD additions prior to LTAD. [Gap closed for SB11]. Revision 1 of this report addresses a variation on the ABD additions and LTAD strategy where sodium hydroxide additions for LTAD may be performed intermittently or concurrently with an ABD addition window. The proposed change does not alter the conclusions of this evaluation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRSDate not supplied

Refractory Carbides for Hydrogen Erosion Resistance in Carbon Tubes for Nuclear Thermal Propulsion

As crewed travel becomes more common to the Moon and beyond, space nuclear propulsion-based craft offer reduced travel times and double the specific impulse compared to conventional chemical rockets. The nuclear core and its integral hydrogen propellant channels will require advanced refractory carbide coatings over a carbon/carbon substrate for hydrogen resistance. Using \textit{ab initio} and other computational techniques, we studied select (Zr/Ta/Nb)C mixed carbide systems for erosion resistance and hydrogen diffusion using realistic levels of carbon vacancies and experimentally obtainable stoichiometries. Various contributions to the formation and hydrogen erosion free energies were calculated, including configurational, electronic, and vibrational terms. Our calculations indicate that ZrC is less likely to lose carbon than NbC or TaC, providing an explanation for heritage NERVA data on ZrC and NbC mass loss performance. Additionally, binary mixtures show compositions that are more stable than their end-member counterparts. Hydrogen diffusion in stoichiometric ZrC was found to be slightly higher than previously theorized, with the caveat that hydrogen saturation of carbon vacancies will hinder further hydrogen diffusion through the bulk, while hydrogen diffusion in ZrNbC is slightly enhanced over ZrC. And finally, we show that as carbon is depleted, the deformation mechanism of ZrC changes from brittle to ductile. Though many questions still remain as to the bulk performance of these materials, these trends and results are important in terms of materials selection efforts for hydrogen-resistant coatings on carbon/carbon substrates.

William C. Tucker↗
DOE OSTI2020

Enhanced detection limits for radiokrypton analysis

In this paper, we present a method for improving detection limits of Atom Trap Trace Analysis for the krypton radioisotopes 85 Kr and 81 Kr. For the case of 85 Kr this work demonstrates that systematic use of isotopically depleted gas for calibration and extended conditioning of the instrument results in a detection limit of 900 85 Kr atoms per 11 μl of Kr gas, equivalent to a 85 Kr/Kr isotopic abundance of 3 x 10 -15 . This improvement of roughly two orders of magnitude over previously reported limits will help to expand the reach of radiokrypton dating towards longer age ranges where most of the radioisotopes have decayed. Additionally, the method offers an opportunity to investigate radiokrypton production via spontaneous fission within naturally occurring minerals to understand potential underground production of these isotopes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Performance improvements of the windowed multipole formalism using a rational fraction approximation of the Faddeeva function

The windowed multipole (WMP) formalism was introduced as a way to calculate Doppler broadened cross sections on the fly during Monte Carlo simulations. While more arithmetic is needed compared to point-wise cross section look-ups, performance remained competitive from the large memory reductions and sequential data access. The single most expensive function call in a depleted fuel assembly problem using WMP comes from the evaluation of the Faddeeva function, which previously relied on a highly accurate, highly-branching algorithm. This paper explores the use of rational fraction approximations tailored to the domain interest of reactor physics applications and the development of lower accuracy approximations sufficient for our application. The rational approximations were implemented and tested in OpenMC on an infinite medium problem to stress the cross section calculation routine and a PWR assembly problem. In both cases, the rational approximation nearly eliminated the ∼ 20% penalty previously observed when comparing to point-wise libraries. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Lattice Physics Calculations Using the Embedded Self-Shielding Method in Polaris, Part I: Methods and Implementation

Polaris is a 2-dimensional multigroup lattice physics capability in the SCALE code system for the analysis of light water reactor fuel designs. The goal of light water reactor lattice physics codes is to generate few-group homogenized cross sections for downstream full-core nodal diffusion calculations. Additionally, lattice physics calculations contain three primary components: the cross section processing calculation, the 2D transport calculation, and the depletion calculation. This paper summarizes the calculational methods and their implementation into Polaris, with an emphasis on implementation of the embedded self-shielding method. The accuracy of the embedded self-shielding method depends on the procedure used to generate self-shielding factors on the multigroup library. Numerical benchmarks calculations reveal that the accuracy of Polaris eigenvalue predictions is enhanced by (1) using heterogeneous unit cell models to generate the self-shielding factors on the library and (2) using self-shielding factors for within-group scattering cross section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Energy deposition model in STREAM with photon transport

Recent efforts have been taken to implement a photon transport module for the neutron transport analysis code STREAM, considering the explicit photon energy distribution in the problem instead of using the locally deposition assumption as employed in previous STREAM versions. The current energy release model in STREAM based on the On-the-fly energy released per fission model is modified accordingly to account for the photon energy deposition computed explicitly by this module. The impacts of the modified energy release model with photon transport in STREAM on the power distribution, fuel temperature and other core parameters during depletion with feedback calculations are studied on several problems from the VERA benchmark suit. Overall, the photon transport provides a better power map for depletion and feedback calculations, especially when strong gamma emitters are present. The fuel temperature, in general, is decreased when the photon transport is employed because fission gammas have been transported out of their birthplace. The new energy deposition model in STREAM with photon transport indicates that gamma energy accounts for about 9.7% of the total energy released in a fresh VERA5 2D core and rises to 10.5% at the End of Cycle. Around 2.5% of the total energy released will be deposited in the coolant in this 2D core. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

FY22 Progress on Computational Modeling of the Water-Based NSTF

This report summarizes the system-level modeling effort by Argonne National Laboratory (Argonne) of the Natural convection Shutdown heat removal Test Facility (NSTF) in FY22. As an extension of the effort from FY21, this year’s work focuses primarily on the two-phase modeling of the NSTF using RELAP5-3D, particularly with the inclusion of the cavity model. The results from simulations were used to compare against experimental data for benchmarking purposes of the RELAP5 deck. Additionally, RELAP5 was used as a predictive tool to guide planned test operations and identify expected system behaviors. In the first part of this report, details are provided of the cavity omitted model where heat flux is applied directly as a boundary condition to the risers. The general trend predicted by the RELAP5 model matches that from the experimental data when a single-phase natural circulation flow is first established, followed by an oscillatory two-phase period and finally a stable two-phase flow. However, the onset of oscillations is predicted early by the model due to the smaller thermal mixing region in the tank. However, by expanding the simulated thermal mixing region in the tank, the onset of oscillations predicted by the model is able to match that from the experiment. These oscillations where studied in depth and are deduced to be flashing-induced instability. The model was then modified to simulate an accident scenario case where a representative heat load based on the full-scale Framatome’s 625 MW t SC-HTGR was applied directly to the riser channels. The simulated initial and boundary conditions were identical to those performed experimentally, facilitating direct comparisons between the predicted and experimental results. It was determined that the results showed some discrepancies remain, likely due to the overprediction of vapor generation rate by the computer model. In the second part of this report, the cavity model is re-introduced where it is observed that the RELAP5 prediction is now able to capture the major trends of the observed flow commonly observed during two-phase conditions. However, the onset of oscillations is once again predicted early by the model, possibly caused by the underprediction of heat loss from the heater and cavity. This is likely due to the omission of support structures in the cavity that can act as additional pathways for heat to escape to the environment. To overcome the underprediction of heat loss, part of the insulation surrounding the cavity side panels and the back of the heaters are removed to allow heat to escape directly to the environment, which then improves the RELAP5 prediction. Parametric studies are also performed to investigate the effects of heater power, tank inventory level, and tank gas space pressure on flow behaviors, also in direct comparison to conditions tested experimentally. User option-61 in the RELAP5-3D input deck, which changes the heat transfer coefficient correlations used for calculating the vapor generation, is also investigated where it is found that by enabling the option, the overall duration of oscillations is increased and matches that from the experiment better. The RELAP5 model is further benchmarked with a header inlet- throttling case where it is observed that the prediction from the model fails to capture some major features observed in the experiment. By using a modified loss coefficient curve for the valve, the accuracy of the prediction is improved where most of the major features observed in the experiment are predicted by the model. Lastly, the model is benchmarked with an inventory depletion scenario where it is observed that despite the modeling limitation of RELAP5, the prediction shows good agreement with the experimental data where major trends and features are captured by the model. Future work will see continued development of the current RELAP5-3D input deck of the NSTF to both improve the accuracy of the model’s predictive capability and continuing serving the experimental program. The mutually beneficial relationship between analysis and experimental efforts has become integral to the parent NSTF program, and the greater objective to fully understand and accurately predict the heat removal performance of a full scale RCCS concept.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

A Custom Low-Noise Silicon Photodiode Detector Designed for Use With X-Ray Capillary Optics

We present the first results of a custom low-noise silicon photodiode detector designed for use with X-ray capillary optics. This system is intended to perform as a smaller, lighter, and low-powered alternative to currently existing X-ray telescope technologies. The detector has an active area with a diameter of 100 μm, and its small size leads to a low capacitance of well under 1 pF and a leakage current of 1 pA or less at the depletion voltage. We tested the detector with both a discrete preamp and an ultralow-noise front-end application-specific integrated circuit (ASIC). The front-end ASIC, developed at Brookhaven National Laboratory, was designed for low-capacitance detectors but had not been previously tested with a detector. Here, we discuss the detector design, the experimental setup, and the resulting detector performance. Using radioactive laboratory sources, we measure a full-width at half-maximum (FWHM) of 211 ± 8 eV at 5.9 keV with the ASIC and are able to set the threshold as low as 0.6 keV.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Update on Radiochemical Assessment of High Burnup Commercially Irradiated Fuel

This work documents an effort to collect burnup measurements on a high burnup rod, designated 6XV, and first cycle accident tolerant fuel (ATF) rod, designated 47I, to enable benchmarking of fuel performance codes and neutronics codes. In addition to measurements, Virtual Environment for Reactor Applications (VERA) full-core-depletion analysis was also performed for the rods that were experimentally analyzed to provide an opportunity for code validation. This effort focuses on collecting data from rods irradiated at Byron Generating Station and shipped to the Oak Ridge National Laboratory (ORNL) hot-cells. This data will also anchor non-destructive examination evaluations of burnup of the various fuel rods undergoing postirradiation examination (PIE) at ORNL. Previous PIE of these fuel rods provides some guidance on the burnup trend across the fuel. Axial gamma spectroscopy scans provide a measure of relative changes in burnup across a fuel pin. Mass spectrometry based burnup measurements performed for this work at specific axial locations in the fuel are fully quantitative. By combining the mass spectrometry data with the gamma scans it is possible to more quantitatively evaluate axial variations in burnup across the entire fuel pin [1]. The combined set of burnup evaluations will be made available to other organizations that have an interest in high burnup radiochemistry data for validation of neutronic simulations and source term evaluation such as the Nuclear Regulatory Commission (NRC).

Harp, Jason [Oak Ridge National Laboratory (ORNL),↗
DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using openmc

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Corrosion Response of Ni-19Cr-5Fe in Molten NaCl-MgCl 2 Salt

Ni-based alloys are strong candidates for use in high-temperature molten salt reactors due to their superior corrosion resistance and mechanical stability compared to stainless steels. In this study, the static corrosion behavior of a Ni-19Cr-5Fe model alloy was systematically evaluated in a purified molten NaCl-MgCl 2 salt at 700°C for 30, 240, and 500 h. Post-exposure analyses were conducted to assess microstructural evolution, corrosion depth, and elemental depletion profiles. Corrosion rates, quantified by chromium depletion depth, followed an inverse power-law trend with increasing exposure time, indicating diffusion-limited kinetics. This trend is attributed to limitations in mass transport in the static salt and to the progressive local depletion of reactive chromium species. Coupled electron backscatter diffraction and energy dispersive X-ray spectroscopy analysis further revealed that the grain boundary character significantly influences corrosion susceptibility: high-angle grain boundaries exhibited pronounced Cr depletion and pitting, while low-angle and Σ3 boundaries remained comparatively resistant. These results offer valuable insight into the role of microstructure in corrosion processes and reinforce the importance of time-dependent material evaluation in molten salt environments relevant to advanced reactor designs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Discriminating Uranium Isotopes Based on Fission Signatures Induced by Delayed Neutrons

The use of active interrogation to induce delayed neutron emission is a well-established technique for the characterization of special nuclear materials. Delayed neutrons have isotope-characteristic spectral and temporal signatures, which provide the basis for isotope identification. However, in bulk materials that contain an appreciable fissile (e.g., 235 U or 233 U) fraction, such as highly enriched uranium (HEU), delayed neutrons have a high probability of inducing additional fissions. As a result, the overall delayed neutron signature consists of two distinct components: the “primary” delayed neutrons (emitted directly by fission fragments) and the “secondary prompt” fission neutrons produced in fission induced by primary delayed neutrons. These prompt products differ from “primary” delayed neutrons both in their energy spectra and in the presence of coincident radiation released by the parent fission event. The presence and relative quantity of prompt products from delayed fission depend on the cross section of the material in the energy range of delayed neutrons, which may differ significantly between isotopes, thus providing an exploitable means for isotope differentiation. Here, we demonstrate two experimental approaches for discriminating between 235 U and 238 U isotopes based on the measurement of delayed neutron-induced fission products. First, HEU and depleted uranium objects are differentiated through the detection of high-energy prompt neutrons from delayed fission using both recoil-based organic liquid scintillators and thermalization spectra from a custom-built capture-gated composite detector. Secondly, coincident radiation measurements are used as the basis for discrimination by comparing the overall rates and time evolution of fission events when delayed neutrons are present.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2004

Lunar and Planetary Science XXXV: Asteroids, Meteors, Comets

Reports included:Long Term Stability of Mars Trojans; Horseshoe Asteroids and Quasi-satellites in Earth-like Orbits; Effect of Roughness on Visible Reflectance Spectra of Planetary Surface; SUBARU Spectroscopy of Asteroid (832) Karin; Determining Time Scale of Space Weathering; Change of Asteroid Reflectance Spectra by Space Weathering: Pulse Laser Irradiation on Meteorite Samples; Reflectance Spectra of CM2 Chondrite Mighei Irradiated with Pulsed Laser and Implications for Low-Albedo Asteroids and Martian Moons; Meteorite Porosities and Densities: A Review of Trends in the Data; Small Craters in the Inner Solar System: Primaries or Secondaries or Both?; Generation of an Ordinary-Chondrite Regolith by Repetitive Impact; Asteroid Modal Mineralogy Using Hapke Mixing Models: Validation with HED Meteorites; Particle Size Effect in X-Ray Fluorescence at a Large Phase Angle: Importance on Elemental Analysis of Asteroid Eros (433); An Investigation into Solar Wind Depletion of Sulfur in Troilite; Photometric Behaviour Dependent on Solar Phase Angle and Physical Characteristics of Binary Near-Earth-Asteroid (65803) 1996 GT; Spectroscopic Observations of Asteroid 4 Vesta from 1.9 to 3.5 micron: Evidence of Hydrated and/or Hydroxylated Minerals; Multi-Wavelength Observations of Asteroid 2100 Ra-Shalom: Visible, Infrared, and Thermal Spectroscopy Results; New Peculiarities of Cometary Outburst Activity; Preliminary Shape Modeling for the Asteroid (25143) Itokawa, AMICA of Hayabusa Mission; Scientific Capability of MINERVA Rover in Hayabusa Asteroid Mission; Characteristics and Current Status of Near Infrared Spectrometer for Hayabusa Mission; Sampling Strategy and Curation Plan of Hayabusa Asteroid Sample Return Mission; Visible/Near-Infrared Spectral Properties of MUSES C Target Asteroid 25143 Itokawa; Calibration of the NEAR XRS Solar Monitor; Modeling Mosaic Degradation of X-Ray Measurements of 433 Eros by NEAR-Shoemaker; Scattered Light Remediation and Recalibration of near Sheomaker s NIS Global Dataaset at 433 Eros; Evaluation of Preparation and Measuring Techniques for Interplanetary Dust Particles for the MIDAS Experiment on Rosetta; Chiron: a Proposed Remote Sensing Prompt Gamma Ray Activation Analysis Instrument for a Nuclear Powered Prometheus Mission;From Present Surveying to Future Prospecting of the Asteroid Belt; Asteroid Physical Properties Probe Microgravity Testing of a Surface Sampling System for Sample Return from Small Solar System Bodies;and Penetrator Coring Apparatus for Cometary Surfaces.

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