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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Data for "Depth of nutrient uptake by deep-rooted plants is regulated by water availability"

The data set consists of strontium (Sr) isotope ratios (87Sr/86Sr), water isotopes, soil cation concentrations, soil water potential sensor data, and results of 87Sr/86Sr mixing model. The plant canopy size files include the dataset of canopy dimension of sagebrush, lupine, and sunflower. The soil and plant ICPMS (Inductively Coupled Plasma Mass Spectrometry) data file includes both of 87Sr/86Sr, and cation concentration dataset from soil exchangeable pool, apatite pool, silicate extract, atmospheric rain deposition, and plant leaf and stem tissues. The plant dendrochronology file includes the dendrochronogical ring width of several sagebrush, and dendrochemical sample data includes the 87Sr/86Sr for each separated growth ring. The modeling result gives the proportion of nutrient sources of each plants (based on their 87Sr/86Sr in leaf tissues and growth rings) from atmospheric deposition and mineral weathering. Soil water potential data includes continuous collection of soil water potential dataset at 2 depths (30 cm and 60 cm, from Nov 24 - Jun 25) of the sampling site. All the samples were collected from 2 sampling campaign June and July 2023, and rain water is a separate sampling from Aug - Sept 2023, at north-facing hillslope near pumphouse site. The data showed that the depth of cation nutrient acquisition is thus tightly coupled with, and likely determined by, water availability in soil, saprolite and bedrock. The enhanced uptake of cations and water from regions of mineral weathering could confer plant and ecosystem resilience during low water years and may impact the rate of bedrock weathering and watershed chemistry during drought. This dataset includes a file-level metadata (flmd.csv) file that lists each file contained in the dataset with associated metadata; a data dictionary (dd.csv) file that contains column/row headers used throughout the files along with a definition, units, and data type; a location metadata file (locations.csv); and a samples metadata file (samples.csv). All files are provided as comma-separated values (CSV) files (.csv). This work was supported by the Watershed Function Science Focus Area at Lawrence Berkeley National Laboratory funded by the US Department of Energy, Office of Science, Biological and Environmental Research under Contract No. DE-AC02-05CH11231.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Core-flood Effluent and Shale Surface Chemistries in Predicting Interaction between Shale, Brine, and Reactive Fluid

We report field and laboratory observations to date indicate that the efficiency of hydraulic fracturing, as it relates to hydrocarbon recovery, depends significantly on geochemical alterations to rock surfaces that diminish accessibility by partial or total plugging of the pore and fracture networks. This is caused by mineral scale deposition such as coating of fracture surfaces with precipitates, particle migration, and/or crack closure due to dissolution under stress. In reactive flow-through experiments, mineral reactions in response to acidic fluid injection significantly reduced system porosity and core permeability. The present study focuses on changes to fluid chemistry and shale surfaces (inlet and fracture walls) resulting from shale-fluid interactions and integrating these findings for an improved estimate of transport-related consequences. The reacted shale surfaces were examined by spatially-resolved scanning electron microscopy - energy dispersive spectroscopy (SEM-EDS) analysis. Importantly, inductively coupled plasma - mass spectrometry/optical emission spectroscopy (ICP-MS/OES) was utilized to probe the chemical evolution of the core-flood effluents. The three study cores selected from the Marcellus formation represent different mineralogies and structural features. In flow-through experiments, lab-generated brine and HCl-based fracture fluid (pH=2) were injected sequentially under effective stress (up to 500 psi) at reservoir temperature (80°C). SEM-EDS results confirmed by the ICP concentration trends showed significant Fe hydroxide precipitates in clay- and pyrite-rich outcrop samples due to partial oxidation of Fe-bearing phases in the case of intrusion of low salinity water-based fluids. Porosity reduction in the MSEEL (Marcellus Shale Energy and Environmental Laboratory) carbonate-rich sample is related to compaction of cores under stress due to matrix softening with substantial dissolution, and pore-filling by hydroxides, as well as barite and salts. Despite the same fluid compositions and experimental conditions used for both MSEEL samples, barite precipitation was much more intense in the MSEEL clay-rich sample due to its greater sorption capacity and additional sulfate source as well as fissile nature with multiple lengthwise cracks. ICP tests revealed time-resolved concentration trends in produced brine and reactive fluids that in turn complemented the pre-/post-reaction SEM-EDS observations.

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

Characterization of Tank 11H samples from tank closure cesium removal (TCCR) Batch 3 - intermediate and final samples

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. The TCCR unit processes dissolved salt from Tank 10H through a series of ion exchange columns containing crystalline silicotitanate (CST) and the effluent is then discharged to Tank 11H. Four interim samples pulled from Tank 11H during and just after the completion of processing of Batch 3 through the TCCR process have been analyzed for 137 Cs activity and density. The 137 Cs activity was found to decrease with each subsequent sample, which is consistent with the addition of decontaminated solution to Tank 11H. When compared to the expected composition from mixing the Tank 10H Batch 3 feed with the material already present in Tank 11H, the bulk chemical composition was as expected. A corrosion control sample collected from Tank 11H in June 2020 showed changes in the chemical composition and 137 Cs activity when compared to the composition measured at the end of Batch 2 processing. As there were no additions made to the tank during this period, these changes were attributed to leaching of the solids present in Tank 11H. Revision 1 of this report contains additional analyses of the fourth interim sample including inductively coupled plasma – mass spectrometry (ICP-MS) results and activities of other radionuclides. These results are consistent with the previously reported results, showing leaching from the Tank 11H heel. Notably, the 90 Sr activity was about 2.7x higher than calculated from a mixture of the original Tank 11H supernate and the Tank 10H supernate as treated by TCCR, suggesting additional leaching of Sr from the Tank 11H heel solids. In addition, Revision 1 includes analysis results for the Batch 3 Post-Production surface and variable depth samples collected in October of 2020. The Post-Production surface sample was similar in composition to the fourth interim surface sample, except for the increased Cs-137 and oxalate concentrations that were presumably associated with leaching of the Tank 11H heel solids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Vitrification of High-Cr Glass in Research-Scale Melter

This test report describes the experimental results from a small-scale test using the research-scale melter (RSM) at Pacific Northwest National Laboratory to demonstrate processing of a high-Cr simulated feed stream, designated HLW-HCr-16. The RSM is a small, joule-heated melter capable of processing melter feed continuously. The melter is equipped with Inconel ® 693 electrodes, Monofrax ® K-3 refractory, and an Inconel 690 pour spout. An electric kiln surrounds the melter body and minimizes heat loss from the melter body during operation. The RSM is equipped with an offgas treatment system that employs quenching, wet scrubbing, and high-efficiency mist elimination. The glass-discharge section is heated to facilitate pouring of the glass. The melter is fitted with a melt cavity that is ~25 cm (10 in.) in diameter with a nominal glass depth of 8.9 cm (3.5 in.). The melter was operated with a target glass temperature of 1150°C and target plenum temperature between 550°C and 700°C for this test. The air bubbling rate was 4.2 L/min. Overall, during the continuous operation of the melter for ~ 103 hours, ~ 141 kg of glass was produced. At the conclusion of the test, the melter and exhaust lines were visually inspected for particulate deposition and corrosion. Entrained material had adhered to the underside of the melter lid and to the exhaust piping. Enrichments in elements such as Cl, F, B, K, Li, P, Na, and S were measured in these deposits through inductively coupled plasma–optical emission spectroscopy and X-ray fluorescence analysis. When the melter electrodes and air bubbler tube were removed from the glass in the RSM, the electrodes appeared discolored, but no significant loss of metal was observed. The processing of a high-Cr simulant, HLW-HCr-16, in the RSM produced glass at an average rate of 1.36 kg/h, equaling a melter-surface-area normalized glass generation rate of 654 kg/day/m 2 . The resulting glass met the toxicity characteristic leaching procedure requirement. Test results of crystallinity, electrical conductivity, and viscosity showed good processing properties of this high-Cr high-level waste glass. RSM offgas was also sampled and analyzed at periodic intervals during steady-state operating conditions. The total decontamination factor averaged by four sampling periods was 134. The concentrations of CO and NO x in emissions were 237 to 422 and 69 to 94 parts per million by volume, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Low-Activity Waste Glass Standards Preparation and Characterization for Calibration of Analytical Instruments

A matrix of seven low-activity waste (LAW) glasses was fabricated and characterized to be used by the Hanford Waste Treatment and Immobilization Plant (WTP) as analytical instrument calibration and matrix interference standards when measuring LAW glass compositions. The matrix of glasses was designed to represent the range of LAW glass compositions that will be generated by the WTP. Samples from each of the seven glasses were measured with electron probe microanalysis (EPMA), ion chromatography (IC), and bulk inductively coupled plasma mass spectroscopy (ICP-MS) to determine the variability and accuracy across batches and within individually poured glass bar samples. High relative percent differences and standard deviations were calculated for components with low concentrations due to instrument detection limits. IC and ICP-MS percent differences and standard deviations were higher at low component concentrations. Otherwise, variability in measured compositions from IC and ICP-MS was generally lower than measurements from EPMA, with the exclusion of Cl, F, Si, and B. Overall, the two methods proved comparable and complementary. From EPMA, IC, and ICP-MS data, volatile elements were underrepresented compared to batched compositions, which suggested loss of these elements during melting. EPMA data was statistically analyzed to evaluate homogeneity within single bars and whole compositions for each of the seven glasses. Most of the variability within glass compositions occurred at the bar-to-bar level; however, no clear systematic trends were observed. Based on these analyses, when using these glasses as reference material, it is recommended that the overall mean and variance of each composition be used for performing instrument calibrations and analytical corrections.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Production and Performance of High Sulfur Concentration Waste Glasses (FY21)

In FY21 Savannah River National Laboratory (SRNL) was tasked to investigate the practical aspects of low activity waste (LAW) glass production with high sulfur loading for the Office of River Protection (ORP). Relationships between sulfur solubility, retention, and melter resonance need to be evaluated to refine waste loading limits and the super saturated melting (triple melt, grind, and wash) methodology. The impact of high sulfur concentration on durability and phase stability (crystallization and liquid-liquid immiscibility) during slow cooling also needs to be further evaluated. To accomplish this, SRNL proposed scaled melter testing using simulants with matched physicochemical properties to melter feed, as opposed to laboratory reagents. This type of testing is expected to validate applicability to design parameters and functional criteria expected during unit operations. This work was expected to commence in FY21, with the first melter tests being performed near the end of the FY. However, reduced access to Savannah River Site and personnel availability during the COVID-19 pandemic significantly reduced the amount of time and resources available to perform the task. This report documents the experimental plan and progress towards this task. The following is a summarized list of scheduled testing that will be performed on feed, condensate, cold cap and glass: 1) Melter runs; 2) Product consistency test (PCT); 3) Inductively coupled plasma optical emission spectroscopy (ICP-OES); and 4) Optical (OM) and scanning electron microscopy (SEM).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Vitrification of Hanford Tank Wastes for Condensate Recycle and Feed Composition Changeover Testing

During the vitrification of Hanford Site nuclear waste at the Waste Treatment and Immobilization Plant (WTP), which is a part of the safe and efficient retrieval, treatment, and disposal mission of the U.S. Department of Energy Office of River Protection, the offgas condensate generated from the waste-to-glass conversion is currently planned to be concentrated by evaporation in the Effluent Management Facility (EMF). This concentrated condensate can then be recycled back to the incoming waste and vitrified. To test the recycle process, an apparatus was designed and built to mimic the EMF evaporator and was then used to concentrate a volume of condensate produced during the vitrification of a sample of Hanford tank 241-AP-107 (referred to herein as AP-107) waste in a continuous laboratory-scale melter (CLSM). The concentrated condensate was added to an additional sample of AP-107 waste, to mimic one round of the recycle process, and the combined solution was vitrified, producing a second round of recycle condensate. In the current study, the EMF test apparatus was used to concentrate the second-round recycle condensate under evaporation conditions (at 45 °C and 1.4 psia) designed to emulate EMF operation. The condensate was successfully concentrated by a factor of ~10 while retaining over 95 % of the technetium-99 ( 99 Tc), Cs, and I inventories in the concentrate. Another portion of AP-107 waste was retrieved by Washington River Protection Solutions, LLC (WRPS) and transferred to Pacific Northwest National Laboratory (PNNL), where it was pretreated and then combined with the second-round recycle AP-107 condensate concentrate and glass-forming chemicals (GFCs) to form the two-time recycle AP-107 melter feed, approximating a second round to the recycling action to be performed at the WTP. A portion of AP-105 waste was also retrieved by WRPS and provided to PNNL for pretreatment and combining with GFCs to form AP-105 melter feed. The two-time recycle AP-107 and AP-105 melter feeds were processed consecutively in the CLSM. The CLSM run proceeded for 13.63 hours, producing 9.70 kg of glass for an average glass production rate of 1464 kg m 2 d -1 during the two-time recycle AP-107 feed charging and 1568 kg m 2 d -1 during the AP-105 feed charging. The rate during AP-107 charging was essentially equivalent to the rate when processing no-recycle AP-107 feed and lower than that achieved when processing one-time recycle AP-107 feed. However, all rates were within the potential range of variability when processing melter feeds with similar composition in the CLSM. Likewise, the rate during AP-105 charging was higher than the previous rate processing AP-105, but within the potential CLSM range. The cold-cap characteristics changed from the typically thin AP-107 cold cap to a foamy-edged cold cap as previously seen with AP-105 shortly after transitioning to the AP-105 melter feed. The glass produced during the CLSM run was within 10 % of its target composition for the primary glass components. The CaO and Li 2 O targets varied by more than 1 wt% between the two-time recycle AP-107 and AP-105 glass targets and it took about 2 turnovers of the CLSM glass inventory to reach a relative chemical steady state in the glass for CaO and Li 2 O after the melter feed inputs were switched. The 99 Tc and total cesium content in the melter feeds were maintained at concentrations expected to be experienced at the WTP. During the CLSM run, while processing the two-time recycle AP-107 melter feed at a relative chemical steady state, the 99 Tc/Cs ratio was 10, and 34% of 99 Tc and 74% of Cs were retained in the glass. These values were higher than those measured in the CLSM run with one-time recycle AP-107 melter feed. After the transition to processing the AP-105 melter feed, when the production reached a relative chemical steady state, the 99 Tc/Cs ratio was 77 while 44% of 99 Tc was retained in the glass. The C's retention during this time frame reached 200% due to the excess C's in the glass after the target content decreased to 15% of its initial level in the two-time recycle AP-107 melter feed to the lower target in the Ap-105 melter feed. While iodine was below inductively coupled plasma mass spectrometry analytical reporting limits in the melter feed and glass samples, it was detected in quantities above the analytical reporting limits in the liquid and filter samples collected from the CLSM offgas treatment system. The behavior of iodine in the CLSM offgas treatment system followed a similar pattern to those of 99 Tc and Cs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Characterization of actinide abundances and isotopic compositions by HR-ICP-MS

In this report, we present actinide isotopic and elemental data from fallout melt glass using high resolution inductively coupled plasma mass spectrometry (HR-ICP-MS). This is an analytical technique that requires minimal sample preparation. These data are directly compared to the results of high precision isotopic analyses and elemental assay by multi-collector (MC-) ICP-MS, which requires lengthy sample processing involving spiking and purifying aliquots of individual elements. The comparison shows broad overlap between the results of the two techniques. Thus, although the data obtained by HR-ICP-MS is of significantly lower precision, it is a viable method to obtain actinide elemental and isotopic data on a more rapid timeframe than traditional high-precision techniques.

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

High Voltage Regrown GaN P-N Diodes Enabled by Defect and Doping Control

This project studied and implemented methods to form GaN p-n diodes using selective area regrowth to achieve selective area doping. Successful selective area doping of GaN p-n diodes is an enabling factor to realize more advanced devices such as vertical transistors. The general challenge to selective area regrowth of GaN is that the primary etch method, inductively coupled plasma (ICP), damages the crystal and causes high leakage when didoes are formed by regrowth on the etched surface. Our approach used low damage etch methods following ICP etch to remove crystal damage and reduced leakage in the regrown diode. This project demonstrated 1.6 kV etched-and-regrown GaN p-n diodes using planar (non-selective) regrowth and 840 V etched-and-regrown p-n diodes using selective area regrown. Enabling factors were use of a low-damage reactive ion etch (RIE) to remove damage caused by the primary ICP etch combined with a multi-step junction terminal extension (JTE) process. Deep level defect investigation quantitatively correlated a deep level near the middle of the GaN band gap with ICP etch-induced leakage that was greatly mitigated by using a slow, low damage RIE process. This research is economically feasible for commercialization because the processes used in this project, including substrate type and source, epitaxial crystal growth and fabrication techniques are all standard to the GaN semiconductor industry. The fundamental understanding and foundational ability to produce kV-class GaN p-n diodes through etch-and-regrowth provides a path to realize high power, high efficiency GaN power switches that can significantly outperform commercial devices for next-generation electrical power conversion and transmission systems.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Sister Rod Destructive Examinations (FY2021) Appendix J: Leaching of High Burnup Used Nuclear Fuel in De-ionized Water

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The leaching experiment was a part of a larger study by Yadukrishnan Sasikumar in defending his doctoral thesis at The Open University, School of Engineering and Innovation [13]. It utilized post-CIRFT Sister Rod test specimens that were slated for disposal. The leaching experiment aims to understand the trends in the radiolytic dissolution of High Burnup Spent Nuclear Fuel (SNF) when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL de-ionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed gamma spectroscopy and inductively coupled plasma mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of fuel exposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

High-Energy X-ray Diffraction Microscopy for Nuclear Forensics FY2022 Project Report

Morphological information on nuclear material has been identified using visible light and scanning electron microscopy. These identify qualitative differences in particle morphology. Three-dimensional imaging of materials through alternating scanning electron microscopy imaging and focused ion beam milling has also been used. Unfortunately, these techniques are time- and labor-intensive, with significant sample preparation required and lengthy analysis times. Further, the resulting 3D images are qualitative, require manual identification, and do not capture statistically-representative populations. High energy X-ray 3D imaging using a direct-beam or diffracted-beam (High-Energy Diffraction Microscopy) have been developed at the Advanced Photon Source and can produce quantitative information on grains (phase, location, etc.) and pores (size distribution, sphericity) in a material. These techniques require only minutes to characterize a sample volume and are non-destructive, thus suitable for a wide range of existing samples and for confirmatory analyses to be carried out using conventional microscopy techniques. In this first year of the project, all uranium oxide samples were synthesized and characterized using conventional analyses by the analytical chemistry laboratory. Conventional analysis methods included powder x-ray diffraction, scanning electron microscopy, impurity analysis via inductively coupled plasma mass spectrometry, and infrared spectroscopy. Impurity analysis shows a drop in boron content from UO 3 to the lowest U 3 O 8 calcination temperature, but otherwise no appreciable difference in any sample. Analysis of diffraction data shows a flip of peaks from UO 3 dominated for the 600 °C calcined sample to U 3 O 8 dominated at 700 °C and 800 °C. Analysis of scanning electron microscopy images shows that with increased calcination temperature the size distribution of particles seems to increase and broaden. Both of these last findings are in line with previously published data, though this work used significantly fewer particles to simply show similar trends instead of getting truly quantitative particle analysis. Infrared analysis similarly shows ingrowth of U 3 O 8 as calcination temperature is increased, along with depression of peaks associated with UO 3 and water. Samples were prepared for analysis at the Advanced Photon Source at beamline 1-ID. It is anticipated that analysis will occur in November of 2022. AI/ML techniques to de-noise data coming out of 1-ID during the analyses was also developed during this time using previously gathered data. Preliminary results using a self-supervision technique called Noise2Selfshow good de-noising of data. Once the uranium oxide samples are analyzed, real data will be used to test the de-noising and other AI/ML techniques that may be developed in the second year of the project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

DISTRIBUTION OF ION-EXCHANGEABLE RARE EARTH ELEMENTS AND CRITICAL MINERALS IN COAL-HOSTED CLAYS OF THE POWDER RIVER BASIN, WY

Increasing demand for a more robust domestic supply of rare earth elements (REE) and critical minerals (CM) has led to significant investigation into unconventional sources. Coal and coal byproducts from the Powder River Basin (PRB) of Wyoming are potential sources of REE but lack complete characterization of mineral hosts including ion-adsorption clays (IACs). IACs, which provide a substantial portion of the global REE supply, form when hydrated REE complexes are physiosorbed to negatively charged sites on clay surfaces. Typical IAC minerals include kaolinite, illite, and smectite. REE complexes are easily leached from IACs via ion exchange, making them economically attractive. However, research regarding the partitioning behaviors of strategic metals into PRB coal-hosted clays is limited. This project aims to quantify the modes of occurrence and distribution of REE and CM in clay minerals derived from PRB coals. Strategic metals investigated in this project have historic production or promising deposits within the PRB and/or high anticipated supply risk, and include: REE (lanthanides and Y), V, Ti, platinum group elements (PGE; Pt, Pd, Rh, Ru, Ir, Os), Al, Co, Ga, Zr, and W. Semi-quantitative compositional data from portable x-ray fluorescence spectroscopy (pXRF) on coal core from the PRB is consistent with the known correlation between REE and Al content, suggesting increased REE abundance will be found in clay-rich units. pXRF data also identifies elevated V, Ti, Co, and Zr correlated with Al. Preliminary x-ray diffraction (XRD) data indicates kaolinite, illite, and dickite are abundant minerals in overburden and clay-rich partings. Quantitative analysis via inductively coupled plasma mass spectrometry (ICP-MS) provides insights into REE and CM fractionation in coals as we compare trace metal occurrence to variations in mineralogy. Additionally, we apply scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS) data to correlate clay size and morphology to trace metal occurrence. Our findings supplement a broader effort to couple REE and CM extraction with existing domestic coal production and contribute to the fundamental understanding of adsorption processes that occur in low-temperature basin environments.

Stuart, Sophia↗
DOE OSTI2023

Density Measurement of NaCl-MgCl 2 -PuCl 3 and NaCl-UCl 3 -PuCl 3 Molten Salt Systems by Neutron Radiographic Dilatometry

A novel neutron radiographic technique was developed and implemented to measure the change in volume as a function of temperature of select plutonium-containing liquid chloride mixtures from melt to 1250 K. These measurements were performed at Los Alamos National Laboratory by members of the molten salt research group under support from Gateway for Accelerated Innovation in Nuclear (GAIN) voucher NE-21-25117: “Density Measurements of Plutonium Bearing Salts via Neutron Beam Dilatometry”. This technical report contains a description of the experimental method and a summary of the results. Initial characterization of the {NaCl, MgCl 2 , UCl 3 , PuCl 3 } salts used in this experiment includes a combination of x-ray diffraction, differential scanning calorimetry, gamma ray spectroscopy, and inductively coupled plasma mass spectrometry. Four samples of NaCl + MgCl 2 + x mol% PuCl 3 and four samples of NaCl + UCl 3 + x mol% PuCl 3 were prepared at the Plutonium Facility, shipped to the Los Alamos Neutron Science Center (LANSCE), and radiographed with the density via neutron radiography (DvNR) apparatus. Density data in the form of meniscus heights were recorded as a function of temperature. The results show a linear dependence of density with temperature and a third order polynomial dependence of density with composition as a function of PuCl 3 addition in the corresponding halide salt mixture. Additions of PuCl 3 result in a significant increase in density of the salt mixture and a slight reduction in the coefficient of volumetric expansion. These results represent a first measurement of the density of plutonium chloride salt mixtures and, in combination with additional measurements, may lead to a generalized description of the density of such salt systems for arbitrary compositions and temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Partitioning of Radionuclides in Various Streams of the Mo-99 Purification Process

The Universal Neutron Irradiator (UNI) was designed to irradiate a small volume of uranium solution and to achieve high fission power without creating a large radionuclide (RN) inventory. The UNI system is equipped with a fully enclosed off-gas system with online hydrogen gas monitoring to prevent the production of an explosive atmosphere, as well as storage cylinders for capturing fission gases. It also includes a catalytic recombiner in the target capsule (to immediately recombine radiolytic hydrogen and oxygen gas in the system) and access ports for sampling the target solution and capsule headspace. In UNI Phase I experiments, 50 mL of uranyl sulfate solution was irradiated at 0.3–0.5 W/mL fission power using a tantalum target assembly to produce photoneutrons via irradiation with an electron beam. The purposes of the Phase I experiments were to test the efficacy of adding 200 ppm Fe 2+ to the target solution to prevent uranyl peroxide precipitation under representative irradiation conditions, and to monitor iodine speciation in the irradiated uranyl sulfate solution. The focus of the UNI Phase II experiments was to use an irradiated uranyl sulfate solution containing a mixture of fission products and monitor their distribution in various streams during the primary recovery column (titania-based column) that separates Mo-99 from uranium. Irradiations were designed to create sufficient activity for tracking their distribution in various processing streams using a combination of gamma counting and inductively coupled plasma mass spectrometry (ICP-MS) analyses. Obtained data will be used to better understand the composition of waste streams, accumulation of RN on the titania column, and identify what RNs accumulate in the uranyl sulfate solution. Furthermore, this knowledge will allow us to determine waste classifications for various streams produced during the Mo-99 purification process and develop an appropriate waste clean-up strategy.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water

The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Oak Ridge National Laboratory Official TM Cover: Development and Qualification of Tungsten Reference Materials for LA-ICP-MS

Reference materials have been developed to provide areal density measurements of tungsten that is deposited on graphite when using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS). Several reference materials of varying areal densities have been produced using a magnetron sputtering, thin film deposition technique. With multiple reference materials that have varying areal densities, a calibration curve has been developed for translating the detector response of an ICP-MS into a quantitative measurement. These results provide the opportunity to make calibrated comparisons across graphite samples that were exposed to tungsten transport experiments for nuclear fusion research. In addition, these methods may be applied to provide individual areal density measurements for the five unique isotopes of tungsten. It is found that areal density measurements via LA-ICP-MS agree with characterization for the same samples when using the more traditional Rutherford backscattering ion beam technique. Qualification of these reference materials was completed using ion beam laboratories at the University of Tennessee, Knoxville and Massachusetts Institute of Technology, and Sandia National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of a hybrid neural network and transfer learning model for optimized ICP-MS/MS operation

Correct function and calibration of instrumentation is a crucial assumption for any scientific experiment. One such instrument, tandem inductively coupled plasma mass spectrometer (ICP-MS/MS), has in-depth calibration settings that range across 30+ different parameters, making it difficult to determine optimal conditions without expertise and some degree of trial and error. Often, these settings are hand-tuned, a time-intensive process prone to local maxima and human error. While some automation is available, the automation also may favor local optimizations over a global optimum. In addition to these difficulties, day to day instrument variability can further complicate the calibration process. We propose a solution to this problem as a machine learning (ML) algorithm that learns how each parameter helps determine the calibration sensitivity across several elements, and re-weights parameters over time as instrument variability changes (e.g., a global neural network (NN) with a time-dependent transfer learning (TL) component). This model would be able to generate a surface of predicted calibration sensitivities and their respective parameters, and a simple multivariate algorithm would be able to pull out the optimum results with the settings associated with them. Here-in, we describe our initial findings in working towards this goal, including data extraction from historical files, exploratory data analysis, and some initial model building to better describe the data and the feasibility of our goal.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Report on the Analysis of Plutonium-238 Oxide, Campaign #P5PO

Oak Ridge National Laboratory (ORNL) is currently producing Plutonium-238 oxide for NASA space programs. Through the irradiation of NpO 2 pellets in the High Flux Isotope Reactor (HFIR), the production of Pu-238 oxide will be used for electrical power and heat on NASA spacecraft such as the Curiosity Mars Rover and the Voyager 1 and 2. The Transuranium Analytical Laboratory (TAL) leads the analytical effort for this project, which includes the dissolution of plutonium product oxide (PUP) and the analysis of the dissolved product solution using both radiological and inductively coupled plasma mass spectrometry (ICP-MS) techniques that are required to meet product specifications. The Transuranic Analytical Laboratory received three samples of Pu-238 oxide for analysis in January of 2022. This report documents the analytical protocols performed by the TAL along with the results produced from said analytical protocols. This report also includes a comparison with the general-purpose heat source (GPHS) specifications.

07 ISOTOPE AND RADIATION SOURCES↗