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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 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Collaborative or competitive phase transformation processes in an additively manufactured-maraging steel M300

Although retained austenite is present in minor amounts in as-built additively manufactured-maraging steels, its evolution during ageing is key to controlling the final microstructure. This study investigates the relationship between austenite reversion and precipitation evolution during ageing, particularly whether these processes compete or cooperate. Using Scanning/Transmission Electron Microscopy, Atom Probe Tomography and Thermo-Calc® PRISMA simulations, the presence and evolution of retained austenite, Ni 3 Ti and Fe 7 Mo 6 intermetallic phases was characterised across different ageing conditions. Experimental results revealed the presence of non-enriched austenite in the as-built condition. Nevertheless, with increasing ageing temperature, Ni and Mo segregation became evident. High-Resolution Transmission Electron Microscopy revealed that Ni 3 Ti precipitation primarily occurred within the martensitic matrix, while Fe 7 Mo 6 nucleated preferentially at the austenite-martensite interface. PRISMA simulations indicated early and rapid precipitation in solute-enriched areas (i.e., intercellular regions), with Ni 3 Ti forming prior to Fe 7 Mo 6 . Both experimental and simulation results suggested Ni diffusion controls retained austenite growth, while Ti and Mo drive Ni 3 Ti and Fe 7 Mo 6 precipitation, respectively. The evidence reported in this work supports a both collaborative and independent phase transformation mechanism, where austenite reversion and precipitation co-evolve.

42 ENGINEERING↗
DOE OSTI2020

Epsilon Metal: A Waste Form for Noble Metals from Used Nuclear Fuel

Epsilon metal (e-metal) is the metallic phase that forms as inclusions at the grain boundaries in the UO2 fuel during reactor operation. This metal is composed of Pd, Mo, Rh, Ru, and Tc. These metallic inclusions are insoluble in strong acid and remnants of these metallic inclusions have been found in the UO2 matrix that remains from the natural reactors in Gabon that were active 1.8 billion years ago, therefore e-metal should be an excellent waste form for the immobilization of the long-lived isotopes 107Pd (6.5 ? 106 a) and 99Tc (2.13 ? 105 a), with 99Tc being the isotope of interest for repository performance. Therefore, the chemical durability of this potential waste form is assessed in this study. Typically, corrosion rates for metallic materials are measured electrochemically because they are quick, inexpensive, and can reveal the mechanism by which a metal corrodes, at least initially. However, in a repository the waste form would be subjected to slowly flowing water without an applied electrical potential over long time periods. Therefore, the corrosion rates of e-metal specimens were measured with both electrochemical tests and the single-pass flow-through test (SPFT). Potentiodynamic and potentiostatic polarization results suggest that a thin passive film exists on the alloy surface, which seems to be responsible for its high corrosion resistance. Additionally, X-Ray photoelectron spectroscopic results suggest that Pd oxides are significantly enriched in the passive film Results from the SPFT show that the dissolution rates were weakly dependent on pH. Only Mo and Re were found in solution and were used for the calculation of the dissolution rates. In general, the electrochemically determined corrosion rates agree reasonably well with the initial dissolution rate measured with the SPFT test, but they are about one or two orders of magnitude higher than the steady state rates. The causes for this discrepancy are discussed.

epsilon-metal, nuclear waste form, corrosion, sing↗
DOE OSTI2020

Impact of uranium oxide (UO 2 ) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety

This work investigates nuclear reactor performance and safety characteristics of UO 2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. Additionally, the purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO 2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO 2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO 2 -Mo were shown to be relatively similar to UO 2 , but higher Mo content and 235 U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Furthermore, thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO 2 -Mo fuel. Compared with UO 2 in a pressurized water reactor (PWR), the optimized UO 2 -Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Segregation behavior and phase instability of Eurofer97 after neutron irradiation to 72 dpa

Here, a reduced activation ferritic/martensitic steel, Eurofer97, was neutron irradiated in the vicinity of 300 °C in the High Flux Isotope Reactor (HFIR) up to 72 dpa. Advanced analytical scanning transmission electron microscopy and conventional transmission electron microscopy were applied to investigate the radiation-induced segregation and phase instability behavior after neutron irradiation. Amorphization was observed in M 23 C 6 carbides. Cr-rich clusters were seen within the matrix, near the lath boundaries and close to the M 23 C 6 carbides. Cr enrichment and Fe depletion were detected at both prior austenite grain boundaries and lath boundaries, despite different segregation magnitude. In addition, the enrichment of Ni, the depletion of V, and tiny cavities (presumably helium bubbles) are also found at lath boundaries. This work interrogates the evolution of microstructures after neutron irradiation, which provides detailed understanding on the microstructural aspects controlling the mechanical integrity of Eurofer97 under high-dose neutron damage.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Computational determination of a primary diffusion mode in γ U-10Mo under irradiation

Low enriched uranium (< 20 % 235 U)-molybdenum (U-Mo) monolithic fuel is the primary candidate for high performance research and test reactors, and is in the process of being qualified to replace highly enriched uranium (≥ 20 % 23 5U) fuel. As part of the qualification process, it is critical to understand and predict the behavior of fission gas bubbles under irradiation, which affects fuel swelling and fuel failure. Mechanistic fuel models are being developed that can both reproduce the existing experimental data for fuel swelling, and be further applied to irradiation conditions beyond the experimental scope. Diffusion of species under irradiation conditions is an important parameter in the mechanistic fuel models; however, no temperature-relevant experimental diffusion data exists. In the present work, radiation-enhanced diffusion coefficients of U, Mo, and Xe in λU-10wt.%Mo were calculated in the temperature range between 300 K and 1400 K via rate-theory models and molecular dynamics simulations with an embedded-atom method interatomic potential for the U-Mo-Xe system. Accordingly, total diffusion coefficients under relevant irradiation conditions are determined using previously obtained intrinsic thermal diffusion and radiation-driven diffusion coefficients, as well as the newly calculated radiation-enhanced diffusion coefficients presented herein. Radiation-enhanced diffusion of U and Mo was dominant in the intermediate temperature range, whereas radiation-enhanced diffusion of Xe did not significantly contribute to total diffusion of Xe at the relevant fission rate densities. Radiation-enhanced diffusion of Xe became faster than both intrinsic thermal diffusion and radiation-driven diffusion at a fission rate density of 5 x 10 22 fissions/m 3 /s, which is higher than the typical fission rate density range in research reactors. The temperature regime that radiation-enhanced diffusion of each element dominated was dependent on the fission rate density. Finally, the total diffusion coefficients of U, Mo, and Xe, updated in this work, will be utilized as parameters in the mechanistic fuel models to help predict the behavior of fission gas bubbles under irradiation more accurately.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Enabling BWR fuel rod analysis in the BISON fuel performance code

Nuclear fuel vendors around the world are pursuing approaches to sustain the existing nuclear reactor fleet consisting primarily of pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). To support the industry's efforts, advanced modeling and simulation tools need to be capable of analyzing both legacy reactor concepts. BWR fuel rods are significantly different than those used in PWRs, which can affect fuel performance analysis. BWR fuel rods include: (1) an extensive use of gadolinia dopant as a burnable absorber, (2) an axial variation in fuel enrichment and gadolinia content, (3) the inclusion of a liner on the inner cladding surface to mitigate the impact of pellet-clad mechanical interaction (which impacts hydrogen and hydride distribution), (4) a lower initial fill gas pressure, (5) bottom-entry control rods, and (6) a lower coolant pressure that results in the two-phase flow boiling phenomenon. Although the primary focus of BISON has been in the area of PWR and advanced reactor fuel analyses, this paper presents the developments in BISON to support BWR fuel performance analysis. An overview of the models that account for the effects of gadolinia is highlighted. Internal mesh generation capabilities to include a liner is presented. Normal operating and transient (reactivity insertion accident) demonstration problems are presented to illustrate the impact of gadolinia, the hydrogen and hydride evolution due to the presence of the liner, and BISON's ability to simulate axially varying enrichments and dopant concentration. Bottom-entry control effects are captured by the axial power peaking factors present in the demonstration cases. Comparisons to integral experiments from the Halden IFA-681 experiments are discussed as initial validation. Reasonable comparisons are obtained for fuel centerline temperature and rod internal pressure as a function of time. In conclusion, simulations of additional experiments containing Gd 2 O 3 -bearing fuel are necessary to completely validate the code for BWR applications.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Tensile properties of co-rolled Zr/U-10Mo fuel at ambient and elevated temperatures

The United States High Performance Research Reactor (USHPRR) initiative is currently interested in replacing high-enriched uranium fuel in research reactors with high-assay low-enriched uranium fuel. Here, this work aimed to assist in that mission by providing tensile and bend test data on bare U-Mo and tensile test data on co-rolled Zr/U-10Mo plates at differing thicknesses (0.285 mm, 0.300 mm, and 0.465 mm) and temperatures (293 K, 473 K, and 623 K). Yield strength and ultimate tensile strength values are reported along with an in-depth comparison of how temperature, rolling direction, thickness, and the presence of the Zr-interlayer affect mechanical properties. Results showed that yield strength and ultimate tensile strength in Zr/U-10Mo foils lower as temperature increases by approximately 300 MPa from 293 to 473 K, and another 100 MPa from 473 to 623 K. Additionally it was found that with increased temperature, there is less differentiation in fuel necking and delamination behavior, as well as a greater perceived elongation to fracture. The elastic modulus of co-rolled mini-plates does not follow the volumetric rule of mixtures in terms of averaged strength and is actually lower than the reported moduli of both Zr and U-10Mo.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of a rapid viability RT-PCR (RV-RT-PCR) method to detect infectious SARS-CoV-2 from swabs

Since the rapid onset of the COVID-19 pandemic, its causative virus, Severe Acute Respiratory Syndrome Coronavirus-2 (SARS-CoV-2), continues to spread and increase the number of fatalities. To expedite studies on understanding potential surface transmission of the virus and to aid environmental epidemiological investigations, here we developed a rapid viability reverse transcriptase PCR (RV-RT-PCR) method that detects viable (infectious) SARS-CoV-2 from swab samples in <1 day compared to several days required by current gold-standard cell-culture-based methods. The method integrates cell-culture-based viral enrichment in a 96-well plate format with gene-specific RT-PCR-based analysis before and after sample incubation to determine the cycle threshold (C T ) difference (ΔC T ). An algorithm based on ΔC T ≥ 6 representing ~ 2-log or more increase in SARS-CoV-2 RNA following enrichment determines the presence of infectious virus. The RV-RT-PCR method with 2-hr viral infection and 9-hr post-infection incubation periods includes ultrafiltration to concentrate virions, resulting in detection of <50 SARS-CoV-2 virions in swab samples in 17 h (for a batch of 12 swabs), compared to days typically required by the cell-culture-based method. The SARS-CoV-2 RV-RT-PCR method may also be useful in clinical sample analysis and antiviral drug testing, and could serve as a model for developing rapid methods for other viruses of concern.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2025

Apatite geochemistry as a tool for understanding the petrogenesis of layered mafic-ultramafic rocks in the Bushveld Complex, South Africa

The sources of the magmas that formed the Rustenburg Layered Suite of the Bushveld Complex in South Africa remain debated, despite decades of research. Vertical and lateral variation in bulk rock and mineral separate Sr-Nd isotopic compositions, which generally indicate enriched sources, demonstrate that the layered sequence was formed by the emplacement of multiple batches of magma, crucially resulting in episodes of PGE-Cr-V mineralisation. The Lu-Hf isotope compositions of zircon are, however, at odds with the bulk rock Sr-Nd isotopic heterogeneity as they show near homogeneous compositions throughout the layered sequence (εHf (2.06 Ga) =−8). This lack of variation in Hf isotope composition has been attributed to deep, continental lithospheric mantle-related and/or crustal contamination of plume-derived Bushveld magmas. In this study, we analysed the major, trace element and Sr-Nd isotope geochemistry of apatite in the Rustenburg Layered Suite. Apatite occurs as an intercumulus mineral in the lowermost regions and a cumulus mineral in the uppermost regions of the layered sequence and can therefore be used to test existing models for the isotopic disequilibrium between bulk rock Sr-Nd and zircon Hf isotopic compositions. Apatite is largely chlorapatite in the lowermost regions and fluorapatite in the uppermost regions of the layered sequence. The Merensky Reef is unusual in that it comprises both chlorapatite and fluorapatite. Apatite throughout the layered sequence is generally unzoned and shows no evidence of late-stage alteration. Trace element data show that apatite is enriched in L/HREE, with common negative Eu-Sr anomalies. These trace element signatures are consistent with a magmatic origin for the apatite grains, with prior, or concurrent, plagioclase crystallization from the same melt. Variability in in situ Sr and Nd isotope compositions of apatite is recorded throughout the layered sequence with εNd (2.06 Ga) compositions varying between −2.5 and − 10.2 and initial 87 Sr/ 86 Sr compositions varying between 0.7079 and 0.7103 (for the Marikana dikes only). The variability in Sr-Nd isotope compositions of apatite is consistent with the bulk rock (and mineral separate) variation in Sr-Nd isotope compositions, suggesting apatite preserves primary magmatic compositions in the Rustenburg Layered Suite.

Apatite↗
DOE OSTI2022

The role of Cr concentration and temperature on cavity swelling with co-injected helium in dual-ion irradiated Fe and Fe-Cr alloys

The level of chromium plays an essential role in irradiation tolerance of Fe-Cr ferritic alloys. However, conflicting results have been reported regarding the dependence of cavity swelling under irradiation on Cr level and temperature. Here, we have performed a comprehensive set of simultaneous dual-ion (Ni + He) irradiations to high dose (~30 displacements per atom, dpa) at 400–550 °C on a series of ultra-high purity Fe and Fe-Cr binary alloys (3–14 wt.%Cr). Helium co-implantation rates of 0.1 and 10 appm He/dpa were selected to examine He synergistic effects relevant for fission and fusion reactor conditions, respectively. Cavities were observed in all irradiated samples by transmission electron microscopy. The results show that higher He implantation rate causes a shift in the swelling peak to higher temperatures in both Fe and Fe-Cr alloys. When assuming smaller cavities as biased sinks, the non-monotonic nature of the cavity swelling behavior is related to the ratio of biased to unbiased point defect sink strengths. Cr-enriched precipitates were observed in Fe-14Cr irradiated at 400 °C by atom probe tomography. Our analysis suggests the formation of Cr-enriched precipitates could suppress cavity swelling for Fe-Cr alloys with Cr content above 10 wt%.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Minimizing thickness variation in monolithic U-10Mo fuel foil and Zr interlayer during hot rolling: A microstructure-based finite element method analysis

Low-enriched uranium alloyed with 10 wt. % molybdenum (U-10Mo) has been identified as a promising alternative to highly enriched uranium fuel for the United States’ high performance research reactors. The monolithic U-10Mo fuel plate consists of a metallic U-10Mo fuel foil with a 25 µm Zr interlayer and a relatively thick cladding of aluminum alloy 6061. The Zr interlayer is typically applied during the hot co-rolling process, and this process dictates the uniformity of the Zr interlayer. Thickness variation observed in the U-10Mo and Zr interlayer has been attributed to several sources: the initial grain size of the U-10Mo castings, can materials, rolling temperature, inhomogeneous molybdenum content, and porosity in the cast U-10Mo. This thickness variation limits the ability to meet the dimensional specification; thus, a better understanding of the factors causing the nonuniform thickness is needed. In this work, we used a novel, microstructure-based finite element method to model the hot rolling process to address these concerns. Grain microstructures in U-10Mo were tessellated and explicitly considered in the finite element model. Each grain was assigned a random material property to mimic the grain strength variations induced by different grain orientations. Simulations were performed using six steel can thicknesses, four grain sizes, and with or without a Zr interlayer to investigate the influences of those variables on the thickness nonuniformity. The simulation results showed that a thinner steel can and finer U-10Mo grain size reduce thickness variations in both the U-10Mo fuel foil and Zr interlayer. The direct findings from the simulations and analysis can be used to optimize the hot rolling schedule, reduce fabrication defects, and meet the dimensional specifications. The proposed microstructure-based finite element model can be also coupled with experimental microstructure characterization data, images, and models to simulate multi-pass hot rolling.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Micro X-ray computed tomography examination of mini plate fuel with hot isostatic pressed aluminum cladding

In order to minimize proliferation risks and improve security of nuclear material, the United States high performance research and test reactors (USHPRR) program is tasked with converting nuclear reactors that are fueled with highly enriched uranium (HEU) fuels to operate with low-enriched uranium (LEU) nuclear fuels. One favorable LEU fuel configuration is plate fuel with a metallic uranium-molybdenum foil clad within an aluminum alloy (AA 6061). In this fuel, the aluminum cladding is bonded with a hot isostatic pressing (HIP) method to seal the cladding around the fuel meat. However, the HIP process parameters influence the cladding performance, as a defective or incomplete bond can cause a pathway for corrosion. Micro X-ray computed tomography (XCT), a nondestructive technique that provides volumetric imaging, can be applied to inspect fuel plate cladding at the engineering scale. In this work, XCT methodology was developed and successfully utilized to not only observe the bond line of unirradiated mini fuel plates, but to also identify subsurface abnormalities in the plates’ cladding. Importantly, in future work, this technique could be applied to fuel plates, pre- and post-irradiation, to quantify irradiation effects on cladding defects and bond line integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Gamma-ray spectrum variations for surface measurements of uranium hexafluoride cylinders

Here, this work aims to characterize features that contribute to the spectral variations that appear in gamma-23 ray spectra of filled uranium hexafluoride (UF 6 ) cylinders. While the enrichment of the UF 6 in a 24 cylinder can greatly affect gamma-ray spectra at peaks such as the 185.7 keV peak and be used to 25 verify enrichment, many other characteristics of a filled UF 6 cylinder can affect the spectra. These 26 characteristics vary on a per-cylinder basis and can include: the time between when a cylinder was 27 filled and when it was measured; the previous fill history of a cylinder; how the cylinder was most 28 recently filled; the UF 6 geometric distribution within the cylinder; and the measurement location 29 along the cylinder. This work explores spectral variations seen in measurement data and uses 30 validated simulation models to find trends in these uncertainties. These trends are analyzed to observe 31 how each parameter can contribute to the overall uncertainty of UF 6 spectral measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma

Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Here, integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Characterization of a boron-loaded deuterated liquid scintillator for fast and thermal neutron detection

The preparation and characterization of novel boron-loaded deuterated liquid scintillators (BLDLS) are presented for the first time to investigate the performance of a detector sensitive to both fast and thermal neutrons intended for future neutron spectroscopy measurements. Three deuterated toluene-based scintillation cocktails were produced, one without boron, one loaded with ortho-carborane (natural boron isotopic abundance), and one with 96 wt% 10B enriched ortho-carborane. General optical and material properties were analyzed to include the composition, density, index of refraction, photoluminescence emission spectra, and absorbance spectra in addition to the relative light yield, resolution, relative efficiency, and pulse-shape-discrimination performance for all three scintillators and a hydrogen-based EJ-309 scintillator to evaluate the effects of boron loading in comparison with a common commercially-available scintillator and feasibility for future neutron spectroscopy measurements. Finally, the detection of shielded neutron sources was explored to capitalize on the 10B thermal neutron capture reaction. The results showed loading the scintillators with boron deteriorated the PSD performance slightly compared to the unloaded deuterated scintillator and resulted in a relative light yield loss of 26% and 41% for the 10B enriched and natural boron BLDLS, respectively. Lastly, the BLDLS have a higher relative detection efficiency than the unloaded deuterated scintillator for shielded neutron sources, based on total count rate, a desirable characteristic for missions related to finding and characterizing illicitly-trafficked special nuclear material.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies

The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (~hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. In this work, we present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Preparation and characterization of isotopically pure Mo targets for nuclear science measurements

Isotopically pure molybdenum targets are essential in various nuclear science measurements. For studies of cross sections of capture reactions on molybdenum, uniform targets with areal densities of hundreds of g/cm 2 are desired. Mechanical rolling of enriched metal pieces or vacuum sputtering methods used in previous studies to prepare molybdenum targets are challenging and involve labor- and time-intensive processes. Therefore, reliable and straightforward methods to produce molybdenum targets are crucial. We have developed a new double-step method to produce targets meeting these needs. The first step consists of evaporating isotopically enriched MoO 3 onto gold backings. The second step involves the reduction of the oxide layer in a 5% H 2 -Ar environment to obtain molybdenum targets. X-ray fluorescence (XRF) spectroscopy, X-ray Diffraction (XRD), a scanning electron microscope/focused ion beam milling, and energy-dispersive X-ray spectroscopy (EDS) were used to characterize the targets. These methods show that the targets exhibit a uniform coverage of Mo over the gold backings. Furthermore, significant volume shrinkage occurring during the reduction of MoO 3 creates a porous Mo layer where the gold diffuses due to the high processing temperatures, resulting in mixed Mo–Au layers.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Electrospraying deposition and characterization of potassium chloride targets for nuclear science measurements

Natural and isotopically enriched KCl targets were prepared using specially formulated KCl solutions, deposited by electrospraying on thin (1.5 μm) gold backings, followed by a short annealing in a preheated (350-450 °C) furnace. Various techniques, such as X-ray fluorescence (XRF), scanning electron microscopy (SEM) imaging, and energy-dispersive X-ray spectroscopy (EDS), were used to characterize the produced targets. A white neutron beam at the Los Alamos Neutron Science Center (LANSCE) facility was also used to verify the estimated atomic percentages of the enriched 40 KCl targets. KCl was deposited relatively uniformly as large particles rather than films on the gold surface. As a result, this method can be broadly extended to make stable targets that are difficult to produce with other methods and more exotic radioactive targets required for research in fundamental sciences and applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗