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At least 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

Carbon and Oxygen Isotope Measurements of Ordinary Chondrite (OC) Meteorites from Antarctica Indicate Distinct Carbonate Species Using a Stepped Acid Extraction Procedure

The purpose of this study is to characterize the stable isotope values of terrestrial, secondary carbonate minerals from five Ordinary Chondrite (OC) meteorites collected in Antarctica. These samples were identified and requested from NASA based upon their size, alteration history, and collection proximity to known Martian meteorites. They are also assumed to be carbonate-free before falling to Earth. This research addresses two questions involving Mars carbonates: 1) characterize terrestrial, secondary carbonate isotope values to apply to Martian meteorites for isolating in-situ carbonates, and 2) increase understanding of carbonates formed in cold and arid environments with Antarctica as an analog for Mars. Two samples from each meteorite, each approximately 0.5 grams, were crushed and dissolved in pure phosphoric acid for 3 sequential reactions: a) R times 0 for 1 hour at 30 degrees Centigrade (fine calcite extraction), b) R times 1 for 18 hours at 30 degrees Centigrade (course calcite extraction), and c) R times 2 for 3 hours at 150 degrees Centigrade (siderite and/or magnesite extraction). CO (sub 2) was distilled by freezing with liquid nitrogen from each sample tube, then separated from organics and sulfides with a TRACE GC using a Restek HayeSep Q 80/100 6 foot 2 millimeter stainless column, and then analyzed on a Thermo MAT 253 Isotope Ratio Mass Spectrometer (IRMS) in Dual Inlet mode. This system was built at NASA/JSC over the past 3 years and proof-tested with known carbonate standards to develop procedures, assess yield, and quantify expected error bands. Two distinct species of carbonates are found: 1) calcite, and 2) non-calcite carbonate (future testing will attempt to differentiate siderite from magnesite). Preliminary results indicate the terrestrial carbonates are formed at approximately sigma (sup 13) C equal to plus 5 per mille, which is consistent with atmospheric CO (sub 2) sigma (sup 13) C equal to minus 7 per mille and fractionation of plus12 per mille based upon polar temperature of -20 degrees Centigrade. The oxygen values fractionate sigma (sup 18) O equal to minus 10-20 per mille lighter between the R times 0 and R times 1 reactions at 30 degrees Centigrade. The carbonate oxygen isotope measurements are consistently heavier than expected with meteoric water and temperatures from Antarctica, perhaps due to secondary carbonate formation during curation in Houston, TX.

Evans, Michael E.↗
NASA NTRS1981

Noble-gas-rich separates from the Allende meteorite

Predominantly carbonaceous HF/HCl-resistant residues from the Allende meteorite are studied. Samples are characterized by SEM/EDXA, X-ray diffraction, INAA, C, S, H, N, and noble gas analyses. Isotopic data for carbon show variations no greater than 5%, while isotopic data from noble gases confirm previously established systematics. Noble gas abundances correlate with those of C and N, and concomitant partial loss of C and normal trapped gas occur during treatments with oxidizing acids. HF/HCl demineralization of bulk meteorite results in similar fractional losses of C and trapped noble gases, which leads to the conclusion that various macromolecular carbonaceous substances serve as the main host phase for normal trapped noble gases and anomalous gases in acid-resistant residues, and as the carrier of the major part of trapped noble gases lost during HF/HCl demineralization. Limits on the possible abundances of dense mineralic host phases in the residues are obtained, and considerations of the nucleogenetic origin for CCF-XE indicate that carbonaceous host phases and various forms of organic matter in carbonaceous meteorites may have a presolar origin.

Ott, U.↗
DOE OSTI2021

Evidence linking calcium to increased organo-mineral association in soils

Geochemical indicators are emerging as important predictors of soil organic carbon (SOC) dynamics, but evidence concerning the role of calcium (Ca) is scarce. This study investigates the role of Ca prevalence in SOC accumulation by comparing otherwise similar sites with (CaCO 3 -bearing) or without carbonates (CaCO 3 -free). We measured the SOC content and indicators of organic matter quality (C stable isotope composition, expressed as δ 13 C values, and thermal stability) in bulk soil samples. We then used sequential sonication and density fractionation (DF) to separate two occluded pools from free and mineral-associated SOC. The SOC content, mass, and δ 13 C values were determined in all the fractions. X-ray photoelectron spectroscopy was used to investigate the surface chemistry of selected fractions. Our hypothesis was that occlusion would be more prevalent at the CaCO 3 -bearing site due to the influence of Ca on aggregation, inhibiting oxidative transformation, and preserving lower δ 13 C values. Bulk SOC content was twice as high in the CaCO 3 -bearing profiles, which also had lower bulk δ 13 C values, and more occluded SOC. Yet, contrary to our hypothesis, occlusion only accounted for a small proportion of total SOC (< 10%). Instead, it was the heavy fraction (HF), containing mineral-associated organic C, which accounted for the majority of total SOC and for the lower bulk δ 13 C values. Overall, an increased Ca prevalence was associated with a near-doubling of mineral-associated SOC content. Future investigations should now aim to isolate Ca-mediated complexation processes that increase organo-mineral association and preserve organic matter with lower δ 13 C values.

58 GEOSCIENCES↗
NASA NTRS1989

Age structure of refractory interstellar dust and isotopic consequences

A sputtering and recycling Monte Carlo model, developed by Liffman and Clayton (1988) is used to calculate the distribution of existence times of the matter in interstellar dust composed of refractory metals. The mean age of each dust particle is defined not as the time it has existed but rather as the mass-weighted existence times of its parts at t = 6 Gyr of the modeled solar system formation. It is shown that Galactic evolution generates a mean correlation, applying to large numbers of particles binned according to size rather than according to individual particles, whose mean ages fluctuate statistically. The cosmochemical consequence is that if interstellar particles can be dynamically sorted into separate size populations during the aggregation history of solar system bodies, the collections of larger grains will constitute matter that is chemically older than collections of smaller grains. The macroscopic age difference generates isotopic anomalies by virtue of the time dependence of the secondary/primary nucleosynthesis yields. Results are compared with three different prescriptions for the sputtering of interstellar dust.

Clayton, Donald D.↗
NASA NTRS2002

SNC Meteorites and Martian Reservoirs

Jones first suggested that the inverse covariation of initial epsilon (Nd-143) and Sr-87/Sr-86 of the shergottites could be explained by interaction between mantle-derived magmas with another isotopic reservoir(s) (i.e., assimilation or contamination). In that model, magmas were generated in a source region that was isotopically similar to the Nakhla source and the second reservoir(s) was presumed to be crust. The text also permitted the second reservoir to be another type of mantle, but I can confirm that a second mantle reservoir was never seriously considered by that author. Other features of this model were that (i) it occurred at a particular time, 180 m.y. ago, and (ii) the interacting reservoirs had been separated at approximately 4.5 b.y. In a later paper Jones explored this mixing model more quantitatively and concluded that magmas from a Nakhla-like source region at 180 m.y. would fall on or near an isotopic Nd-Sr-Pb hyperplane defined by the shergottites. This criterion was a necessary prerequisite for the parent magma(s) of the shergottites to have initially been Nakhla-like isotopically. At this juncture, it is perhaps worthwhile to note that this mixing model was not presented to explain geochemical variations but as a justification for a 180 m.y. crystallization age for the shergottites and a 1.3 b.y. crystallization age for the nakhlites. In the mid-1980's crystallization ages estimated for Nakhla ranged from approximately 1.3 b.y to 4.5 b.y. Similarly, preferred crystallization ages for the shergottites ranged from 360 m.y., to 1.3 b.y., to 4.5 b.y. In all these models, the 180 m.y. event seen in the shergottites was deemed to be metamorphic. The fit between the Nakhla-like source region and the shergottite hyperplane was a validation both of the 1.3 b.y. igneous age of Nakhla and the 180 m.y. igneous age of the shergottites.

Jones, J. H.↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

14 MeV Irradiation and Analysis of a 93% 239 Pu Target in Preparation for a F2019 FY22 Pu Campaign

In this work, we present the irradiation of a 93% 239 Pu with 14 MeV neutrons and subsequent analysis of the fission and activation products. The fully assembled target, a Pu metal bead encapsulated in Al, further encapsulated in welded stainless steel, was analyzed 22 times over more than 100 days using gamma emission analysis (GEA). Using the results from these analyses, R-values and fission yields for fission products were determined. To prepare for the FY22 Pu irradiation using the GODIVA critical assembly at NCERC, the irradiated Pu target was disassembled, dissolved, and separated using chemistry provided by LANL collaborators. This chemical separation was intended to remove the Pu from solution with little to no effect on the remaining elements. The chemistry was assessed to try to determine possible routes of fractionation of the sample and the fission products. The separation process used was successful for the bulk of the analyzed fission and activation products, as well as the added radiotracers. The final dissolved solution and separated fractions were analyzed by GEA looking at the remaining fission, activation and radiotracers and thermal ionization mass spectrometry analysis looking at the Pu isotopics.

14MeV↗
NASA NTRS1991

U-Th-Pb isotopic systematics of lunar norite 78235

A pristine high-Mg noritic cumulate thought to be relict deep-seated lunar crust is studied with an eye to obtaining evidence of initial Pb isotopic composition and U/Pb ratios of early lunar magma sources and possibly of a primary magma ocean. A leaching procedure was conducted on polymineralic separates to assure the removal of secondary Pb components. The Pb from leached separates do not form a linear trend on the Pb-Pb diagram, indicating open-system behavior either from mixtures of Pb or postcrystallization disturbances. Calculated initial Pb compositions and corresponding U-238/Pb-204 (mu) values are presented, with the assumption of reasonably precise radiometric ages from the literature for norite 78236. The results obtained support the contention that high-Mg suite rocks are coeval with the ferroan anorthosites, both being produced during the earliest stages of lunar evolution.

Premo, W. R.↗
NASA NTRS1994

Noble gases in presolar diamonds II: Component abundances reflect thermal processing

Using the isotopic compositions derived in Huss and Lewis, 1994a, abundances of the P3, HL, and P6 noble-gas components were determined for 15 diamonds separates from primitive chondrites of 8 chondrite classes. Within a meteorite class, the relative abundances of these components correlate with the petrologic subtype of the host meteorite, indicating that metamorphism is primarily responsible for the variations. Relative abundances of P3, HL, and P6 among diamond samples can be understood in terms of thermal processing of a single mixture of diamonds like those now found in CI and CM2 chondrites. With relatively gentle heating, primitive diamonds first lose their low-temperature P3 gases and a 'labile' fraction of the HL component. Mass loss associated with release of these components produces an increase in the HL and P6 content of the remaining diamond relative to unprocessed diamond. Higher temperatures initiate destruction of the main HL carrier, while the HL content of the surviving diamonds remains essentially constant. At the same time, the P6 carrier begins to preferentially lose light noble gases. Meteorites that have experienced metamorphic temperatures greater than or = 650 C have lost essentially all of their presolar diamond through chemical reactions with surrounding minerals. The P3 abundance seems to be a function only of the maximum temperature experienced by the diamonds and thus is independent of the nature of the surrounding environment. If all classes inherited the same mixture of primitive diamonds, then P3 abundances would tie together the metamorphic scales in different meteorite classes. However, if the P3 abundance indicates a higher temperature than do other thermometers applicable to the host meteorite, then the P3 abundance may contain information about heating prior to accretion. Diamonds in the least metamorphosed EH, CV, and CO chondrites seem to carry a record of pre-accretionary thermal processing.

Huss, Gary R.↗
NASA NTRS2016

Assessing the Formation of Ungrouped Achondrite Northwest Africa 8186: Residue, Crystallization Product, or Recrystallized Chondrite?

The recent discoveries of primitive achondrites, metachondrites, and type 7 chondrites challenge the long held idea that all chondrites and achondrites form on separate parent bodies. These meteorites have experienced metamorphic temperatures above petrologic type 6 and have partially melted to various degrees. However, because of their isotopic and compositional similarities to both undifferentiated and differentiated groups, the provenance of these 'type 6+' meteorites remains largely unknown. CK and CV chondrites have recently been linked to a few achondrites due to their strong compositional, mineralogical, and isotopic similarities], suggesting a common origin between these meteorites. Although CVs have generally undergone low degrees of alteration near petrologic type 3, CKs have experienced a wide range of thermal alteration from petrologic type 3 to 6. Thermal evolution models on early accreting bodies predict that an early forming body can partially differentiate due to radiogenic heating, and, as a result, form radial layers of material increasing in thermal grade (types 3 to 6+) from the unmelted chondritic surface towards the differentiated core.Northwest Africa (NWA) 8186 is an ungrouped achondrite that provides compelling evidence for higher degrees of thermal processing and/or melting and differentiation on some CK/CV parent bodies. NWA 8186 plots on the CCAM line on a 3-oxygen isotope diagram directly with CK and CV chondrites and also plots with the CKs in regards to Cr isotopes. This meteorite is dominated by Nickel(II)Oxygen-rich olivine (less than 80%), lacks iron metal, and contains four oxide phases, indicating a high fOxygen (above FMQ) similar to the CKs. Additionally, NWA 8186 does not contain chondrules. We have further investigated the origins of NWA 8186 by examining and comparing the bulk composition of this CK-like achondrite with CK and CV chondrites, allowing us to assess the various scenarios in which NWA 8186 may have formed from CK/CV precursor material.

Srinivasan, P.↗
DOE OSTI2021

Isotope Production at the Jefferson Lab LERF Facility (Final Report)

The project initially proceeded well, meeting all of the Milestones. However, the need to create a new irradiation area in the LERF (the initial proposal was to use the CEBAF Injector) used up a larger fraction of the budget than foreseen. A further problem arose when the hot-cell became unavailable at VCU, followed by VCU dropping out of the collaboration. This meant that we were unable to separate the 67 Cu from the gallium. We replaced this by carrying out detailed gamma spectroscopy to determine the quantity of 67 Cu produced.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Inorganic resins enable the increased purification efficiency of 82 Sr from rubidium targets for use in PET imaging isotope production

The 82 Sr/ 82 Rb generator system has found use in positron emission tomography (PET) imaging of myocardial perfusion. 82 Sr is produced by proton irradiation of Rb metal/RbCl salt followed by chemical separation. Inorganic ion-exchangers were synthesized by hydrothermal methods and evaluated for the purification of accelerator-produced 82Sr from a RbCl target and co-produced impurities like 75 Se, Fe, Cr, and Ni. A process consisting of removal of bulk RbCl using sodium nonatitanate (SNT), then removal of remaining tracers of Rb and Se using poorly crystalline silicotitanate (pCST) is suitable for 82 Sr separation and achieves higher purity of 82 Sr vs. commonly used organic resins.

43 PARTICLE ACCELERATORS↗
DOE OSTI2026

RuKY Catalyst‐Packed Permeation Membrane for Quantitative Ammonia and d3‐Ammonia Dehydrogenation to Ultrapure Hydrogen

Ammonia is a promising carbon-free hydrogen carrier, but incomplete ammonia dehydrogenation (cracking) generates atmospheric emissions of NO x , a potent greenhouse gas. Additionally, incomplete cracking of ammonia leads to regulatory challenges in nuclear and fusion power, where tritiated ammonia (NT 3 ) emissions are strictly controlled. Therefore, we report the use of low-temperature ammonia dehydrogenation catalysts (3%Ru/1%Y/12%K/Al 2 O 3 ) in a palladium alloy H 2 permeation membrane for quantitative conversion of ammonia into hydrogen and nitrogen at industry-relevant conditions. This catalytic membrane reactor system achieved an astonishing effluent concentration of <1 ppm at 450°C under a 100% NH 3 stream, which is far beyond the 99.6% conversion target required for the adoption of ammonia as a vehicle fuel. The low-temperature ammonia dehydrogenation catalyst was tested in a packed bed reactor with NH 3 and ND 3 to both elucidate the reaction mechanism and to quantify the kinetic isotope effect of the membrane reactor. The rate-limiting step at temperatures relevant to the palladium membrane are isotope independent, indicating that the isotopologue content will not modify the desired reaction kinetics. By reducing emissions to below-trace levels with no additional separation, this work provides a path to greatly simplified and miniaturized ammonia cracking processes.

ammonia decomposition↗
DOE OSTI2025

Design Report on the MiniFuel Instrumented Test Apparatus for Understanding Radiation Effects

Most nuclear fuels irradiations at Oak Ridge National Laboratory (ORNL) over the past decade have been conducted using MiniFuel—a static capsule design employing subscale fuel specimens to collect separate-effects irradiated fuel performance data. Irradiation conditions for MiniFuel experiments are predicted pre-test using reactor physics, and thermal models are verified post-irradiation via SiC dilatometry and various spectrometry methods. Relevant fuel performance parameters are also observed post-irradiation in a hot cell, thereby providing a single data point for each parameter representing the cumulative effects of the irradiation conditions. Substantially more data can be harvested from a single test and within a shorter duration by instrumenting irradiation vehicles and measuring desired quantities in situ. This report presents the design and analysis of the MiniFuel INstrumented Irradiation Test Apparatus for Understanding Radiation Effects (MINITAURE)—an instrumented test rig based on the separate-effects MiniFuel concept that aims to capture fission gas release (FGR) and thermal conductivity degradation of fuel specimens during irradiation in the High Flux Isotope Reactor (HFIR). MINITAURE will be integrated with the Materials Irradiation Facility (MIF) located in the HFIR building outside the reactor containment. The MIF will act as the instrumentation and control center for the experiment, enabling real-time feedback from in situ sensors and control of irradiation temperatures via a gas delivery system. Two unique capsule designs were developed to capture each phenomenon: the thermal conductivity capsule, which uses a thermopile method to estimate fuel specimen thermal conductivity, and the fission gas release capsule, which will have continuous flowing gas communication to high-purity germanium detectors that are housed in the MIF for monitoring FGR. This report details the reactor physics and heat transfer modeling activities that were used to inform the experiment design and predict capsule performance. It also describes out-of-pile activities conducted to stand up this new capability and verify the measurement techniques. Modeling efforts to date have demonstrated the feasibility of the in situ measurement techniques and supported the development of the MINITAURE assembly configuration. Out-of-pile testing of the thermal conductivity measurement shows promise in capturing relative changes in thermal conductivity. However, significant errors exist in the measured absolute value, posing a need for further refinement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

DEIMoS: An Open-Source Tool for Processing High-Dimensional Mass Spectrometry Data

We present DEIMoS: Data Extraction for Integrated Multidimensional Spectrometry, a Python application programming interface and command-line tool for high-dimensional mass spectrometry data analysis workflows, offering ease of development and access to efficient algorithmic implementations. Functionality includes feature detection, feature alignment, collision cross section calibration, isotope detection, and MS/MS spectral deconvolution, with the output comprising detected features aligned across study samples and characterized by mass, CCS, tandem mass spectra, and isotopic signature. Notably, DEIMoS operates on N-dimensional data, largely agnostic to acquisition instrumentation: algorithm implementations utilize all dimensions simultaneously to (i) offer greater separation between features, improving detection sensitivity, (ii) increase alignment/feature matching confidence among datasets, and (iii) mitigate convolution artifacts in tandem mass spectra. We demonstrate DEIMoS with LC-IMS-MS/MS data, demonstrating the advantages of a multidimensional approach in each data processing step.

Colby, Sean M↗
NASA NTRS1972

Geochemistry of Apollo 15 basalt 15555 and soil 15531.

Data are presented on major and trace element concentrations determined by atomic absorption spectrophotometry, colorimetry, and isotope dilution in Apollo 15 mare basalt 15555 from the Hadley Rille area, as well as on trace element concentrations determined in plagioclase and pyroxene separates from basalt 15555 and in soil 15531 from the same area. Most of the chemical differences between basalt 15555 and soil 15531 could be accounted for if the soil were a mixture of 88% basalt, 6% KREEP (a component, identified in other Apollo soils, rich in potassium, rare-earth elements, and phosphorus), and 6% plagioclase.

Schnetzler, C. C.↗
NASA NTRS1973

Krypton and xenon in lunar fines

Data from grain-size separates, stepwise-heated fractions, and bulk analyses of 20 samples of fines and breccias from five lunar sites are used to define three-isotope and ordinate intercept correlations in an attempt to resolve the lunar heavy rare gas system in a statistically valid approach. Tables of concentrations and isotope compositions are given.

Basford, J. R.↗