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190 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Production of Radiohalogens: Bromine and Astatine for Imaging and Therapy

Objective: This collaborative proposal between the University of Alabama at Birmingham (UAB), University of Wisconsin, Madison (UWisc) and the University of Pennsylvania (Penn) aims to research the efficient production of high purity, clinical grade radiohalogens for preclinical and translational applications by optimizing the cyclotron targetry and separation chemistry of 76,77Br and 211At for use in imaging and therapy. Proposed Research and Methods: During the course of this project, faculty and trainees at our three institutions will use the complementary infrastructure available to examine the 76Se(p,n)76Br, 77Se(p,n)77Br, 78Se(p,2n)77Br, natAs(α,2n)77Br ,80Kr(p,α)77Br and 209Bi(α, 2n)211At reactions as routes to the production of these valuable radiohalogens for imaging and therapy. Working together we aim to optimize both solid and gas targetry, including reclamation of expensive enriched material. We will develop high yielding recovery processes based on dry distillation and ion exchange techniques. We will also create a comprehensive set of quality control procedures to ensure we are producing our radioisotopes in the highest radiochemical and chemical purity possible. By bringing three institutions together to perform research and development we provide the unique opportunity to build an everlasting cooperative culture in the field of nuclear medicine research by enabling academic centers to delve into the fields of radiohalogen chemistry. In addition, each institution involved holds different skills and expertise and a collaborative grant will facilitate the cross-pollination of ideas, techniques, and build strong bonds that unite under the DOE mission. Additionally, our work does not stop at the completion of this proposal, rather this proposal allows us to gain the momentum needed to get our research and developmental efforts off the ground thereby creating a network of three institutions dedicated to the advancement of isotope production for medical research.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS1996

Sulfur and Hydrogen Isotope Anomalies in Organic Compounds from the Murchison Meteorite

Isotopic measurements have been made on organic sulfur and phosphorus compounds recently discovered in the Murchison meteorite. Carbon, hydrogen and sulfur measurements were performed on individual members of the organic sulfur compounds, alkyl sulfonates; and carbon and hydrogen measurements were made on bulk alkyl phosphonates. Cooper and Chang reported the first carbon isotopic measurements of Murchison organic sulfonates, providing insight into the potential synthetic mechanisms of these and, possibly, other organic species. Hydrogen isotopic measurements of the sulforiates now reveal deuterium excesses ranging from +660 to +2730 %. The deuterium enrichments indicate formation of the hydrocarbon portion of these compounds in a low temperature astrophysical environment consistent with that of dense molecular clouds. Measurements of the sulfur isotopes provide further constraints on the origin and mechanism of formation of these organic molecules. Recently, there has been growing documentation of sulfur isotopic anomalies in meteoritic material. Thiemens and Jackson have shown that some bulk ureilites possess excess S-33 and Thiemens et al. have reported excess S-33 in an oldhamite separate from the Norton County meteorite. Rees and Thode reported a large S-33 excess in an Allende acid residue, however, attempts to verify this measurements have been unsuccessful, possibly due to the heterogeneous nature of the carrier phase. With the recognition that sulfur isotopes may reflect chemistry in the protosolar nebula or the precursor molecular cloud, identification of potential carriers is of considerable interest. In the present study, the stable isotopes of sulfur were measured in methane sulfonic acid extracted from the Murchison meteorite. The isotopic composition was found to be: (delta)S-33 = 2.48 %, (delta)S-34 = 2.49 % and (delta)S-36 = 6.76 %. Based upon analysis of more than 60 meteoritic and numerous terrestrial samples, the mass fractionation lines are defined by Delta-33 = (delta)S-33 -0.50(delta)S-34 and Delta-36 = (delta)s-36 - 1.97 (delta)S-34. From these relationships Delta-33 = 1.24 % and Delta-36 = 0.89 % are observed. These anomalies, particularly the Delta-33, lie well outside the range of analytical uncertainty. They are the largest observed in any meteoritic component and the first found in an organosulfur compound. As discussed by Thiemens and Jackson, due to it's position on the periodic chart, sulfur undergoes chemically induced mass independent isotopic fractionations as does oxygen. Experiments by Mauersberger et. al. show that in such processes, the magnitude of fractionation for the different isotopically substituted species varies with mass and angular momentum; thus, anomalies are expected for both S-33 and S-36, but not necessarily of the same magnitude. Laboratory experiments have also confirmed that chemically produced, mass independent fractionations are mediated by molecular symmetry factors. A chemical source of fractionation requires that the sulfur isotopic anomaly was established in the gas phase, probably in reactions involving symmetric CS2. The discovery of an anomalous sulfur isotopic composition in a specific molecule containing excess deuterium is an important advance in the understanding of the cosmochemistry of sulfur. This evidence suggests that methanesulfonic acid was synthesized by interstellar processes. Further measurements and details of possible synthesis and fractionation mechanisms will be presented.

Cooper, G. W.↗
NASA NTRS2000

Chondrules in the LEW85332 Ungrouped Carbonaceous Chondrite: Fractionation Processes in the Solar Nebula

We studied 14 chondrules separated from LEW85332, an ungrouped type-3 carbonaceous chondrite related to CR chondrites; 23 elements were determined by neutron activation and the chondrules were characterized petrographically. Oxygen isotopic compositions were determined by R. N. Clayton and T. K. Mayeda for seven chondrules. Chondrule abundance ratios tend to form one of two distinct patterns. In low FeO chondrules, refractory lithophile patterns are flat (i.e., unfractionated); siderophile abundances are high and show a small decrease with increasing volatility. Although high FeO chondrules also have flat refractory lithophile abundance patterns, siderophile abundances are highly fractionated; refractory Ir is very low and Fe is very high relative to other siderophiles. We suggest that the low FeO chondrules in LEW85332 formed early in nebular history when metal was intimately mixed with silicates in the chondrule precursors, and that the viscosity of the liquid-solid mix was too high to permit expulsion of the metal by centrifugal action; their porphyritic structures are consistent with incomplete melting, which would result in relatively high viscosities. When the high FeO chondrules formed somewhat later, much of the Fe was oxidized and the melting of precursors was more extensive, FeO and other oxidized siderophiles were retained in the silicate liquid, and metal was lost, possibly expelled from low viscosity chondrule melts. The O isotopic compositions of the chondrules form a linear array of slope 0.93 +/- 0.05 on a three-isotope diagram, parallel to the carbonaceous chondrite anhydrous minerals (CCAM) line and a CR chondrule array, but offset from the latter by -1% in (delta)O-18. Some or all of this offset may reflect incorporation of O from Antarctic water during weathering. Chondrule (Delta)O-17 values correlate positively with FeO, possibly indicating that the (Delta)O-17 of the nebular gas composition increased with time. The chemical and O isotopic data suggest a temporal sequence extending from early, low FeO to late high FeO chondrules.

Wasson, John T.↗
DOE OSTI2022

Selective Detection of Diglycolamide Diastereomers By Metal-Ligand Cluster Formation

Introduction (120 words): Differentiation of isomers using mass spectrometry is a challenge, as these species tend to produce ions of identical mass-to-charge ratio. Therefore, separation with mass spectrometry must rely on chemical differences between isomeric species. Separation of stereoisomers is especially challenging, as these species often have very similar chemistry. However, recent investigations of the solvent extraction behavior of diastereomers of diglycolamide molecules have shown that different stereoisomers can have significantly different lanthanide extraction efficiencies. These studies showed that the R,S stereoisomer of 2,2'-oxybis(N,N-dioctylpropanamide efficiently extracts lanthanides, while the S,S stereoisomer does not. Thus, gas-phase metal complexation may provide a pathway for differentiating these species using mass spectrometry. Methods (120 words): Europium, samarium, and holmium clusters with the R,S, and the S,S diastereomers of 2,2'-oxybis(N,N-didecylpropanamide (trivially tetradecyl diglycolamide, or mTDDGA) were generated in gas-phase using the electrospray ionization source of a Bruker (Billerica, MA, USA) micrOTOF-Q II quadrupole time-of-flight mass spectrometer. Solutions of aqueous lanthanide nitrate and diglycolamide ligands in 2-propanol were added to 50%:50% by volume acetonitrile:2-propanol to generate the spray solutions. Ligand concentration was maintained at approximately 6 µM, while lanthanide concentration was maintained at either 6 µM or 30 µM . High resolution, high mass accuracy spectra were recorded using the time-of-flight. Mass accuracy was ensured using external calibration with Agilent (Santa Clara, CA, USA) ESI-L Low Concentration tuning mix. Preliminary data (300 words): Electrospray of solutions containing only R,S-mTDDGA or S,S-m-TDDGA yield abundant peaks at m/z= 721.7, corresponding to [mTDDGA+H]+1, m/z = 743.7, corresponding to [mTDDGA+Na]+1, and m/z = 1464.4, corresponding to the sodium-bound dimer, [Na(mTDDGA)2]+1. Spray of solutions containing 6 µM europium (Eu) with 3 µM R,S-mTDDGA produced an abundant cluster ion with an isotopic envelope centered at m/z=772.0, corresponding to [Eu(R,S-mTDDGA)3]3+. Europium-containing cluster ions with isotopic envelopes apexing at m/z = 1012.3 and 1189.0 correspond to the [Eu(R,S-mTDDGA)4]3+ and [Eu(R,S-mTDDGA)5]3+ cluster ions, respectively. Protonated and sodiated R,S-mTDDGA are present as well, along with the sodium-bound dimer. In contrast, spray of solutions containing 6 µM europium (Eu) with 3 µM S,S-mTDDGA diastereomer yielded only protonated and sodiated ligand ions; no Eu- S,S-mTDDGA ions were present. This mirrors the solution-phase behavior, where the S,S diestereomer extracts lanthanides with significantly less efficiency than the R,S diestereomer. Experiments with holmium and samarium yielded the same results. Increasing the lanthanide concentration to 30 µM resulted in limited formation of lanthanide-S,S-mTDDGA clusters. However, the abundance of the [Ln(S,S-mTDDGA)3]3+ clusters remained low compared to the abundance of the protonated and sodiated ligand. Additionally, the abundance of the [Ln(R,S-mTDDGA)3]3+ cluster ion was significantly higher than that of the protonated and sodiated ligand in experiments with spray solutions containing 30 µM lanthanide with R,S-mTDDA, supporting the hypothesis that R,S-mTDDGA has a higher intrinsic affinity for the lanthanides than S,S-mTDDGA. Thus, the intrinsic complexation with lanthanides appears to offer a route to differentiate diglycolamide diestereomers in the gas-phase. Novelty (20): Stereoisomers can be differentiated in the gas phase by exploitation of differences in metal complexation chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Directed Gas-Phase Formation of the 1-Cyanovinyl Radical (H 2 CCCN, X 2 A′) in the Interstellar Medium

The formation pathways to nitrogen-containing molecules and radicals are crucial to the understanding of the carbon–nitrogen chemistry in interstellar and atmospheric environments. While over 65 nitrogen-containing neutral species have been observed in deep space to date, their formation mechanisms─in particular, those of radical species─remain largely speculative. The crossed molecular beam technique in conjunction with electronic structure and statistical calculations was utilized to offer a detailed overview of the fundamental pathways in the gas-phase bimolecular reaction of ground-state atomic carbon (C, 3 P) with acetonitrile-d 3 (CD 3 CN, X 1 A 1 ) under single-collision conditions leading to the formation of the 1-cyanovinyl radical (D 2 CCCN, X 2 A′) coupled with deuterium atom loss. Here, the indirect reaction was initiated by barrierless carbon-atom addition, with the most probable route involving carbon addition across the carbon–nitrogen nitrile triple bond of acetonitrile, forming a three-membered ring intermediate followed by ring-opening and unimolecular decomposition via atomic deuterium loss from the C3 carbon atom. The reaction was overall exoergic, and intermediates and transition states lie lower in energy than the separated reactants, unlocking the reaction of carbon with acetonitrile in low-temperature environments such as cold molecular clouds, e.g., Taurus Molecular Cloud (TMC-1), and planetary atmospheres, e.g., Saturn’s moon Titan. In these environments, the 1-cyanovinyl radical may act as a building block for cyano-substituted polycyclic aromatic hydrocarbons and N-heterocycles, thus furthering our understanding of the complex carbon–nitrogen chemistry in deep space.

Chemical reactions↗
NASA NTRS2011

Compositional Impact of Io Volcanic Emissions on Jupiter's Magnetosphere and the Icy Galilean Moons

The magnetospheric ion population of Jupiter is dominated by the 1000 kg/s of iogenic material constantly ejected by IO volcanism as neutral gas (approx. 1 kg/s goes out as high speed dust grains), subsequent atmospheric losses to the IO torus, and radial transport of torus ions throughout the magnetosphere. As that magnetosphere is greatly distended in radial size by the iogenic plasma loading, so are surfaces of the other Galilean moons also significantly, and perhaps even dominantly, affected by iogenic plasma bombardment, e.g. at the level up to 0.2 kg/s heavy ions (mostly O and S) onto Europa as per local plasma ion measurements. In comparison, cometary impacts onto IO deliver about 0.02 kg/s of impact ejecta to Europa via ballistic transfer through the Jupiter system. The magnetosphere of this system operates as a powerful engine to produce and transport ions from the IO source to the surfaces of these other moons, and any future orbiter missions to these moons must account for surface distributions of the iogenic material and its chemical effects before real assessments can be made of sensible chemical materials otherwise arising from primordial formation and subsequent evolution of these moons. This is a fundamental problem of space weathering that must be addressed for all planetary bodies with thin atmospheres and direct surface exposure to their space plasma environments. Long-standing debates from Galileo Orbiter measurements about the origins of hydrate sulfates at Europa present examples of this problem, as to whether the sulfates arise from oceanic minerals or from iogenic sulfur chemistry. Any orbiter or landed mission to Europa for astrobiological investigations would further need to separate the potential chemical biosignatures of life or its precursors from the highly abundant background of iogenic material. Although no single ion carries a tag identifying it as of iogenic or other origin, the elemental abundance distributions of ions to be measured throughout the jovian magnetosphere and in the local moon environments can act as tracers if we know from direct measurements and models the distributions at the mostly likely sources, i.e. at IO. However, our knowledge of these abundances are very limited from earlier in-situ and remote measurements, mainly confined to major (S, O) and some minor (Na, K, Cl) species with abundances at or above a few percent relative to O. Future in-situ plasma measurements by the planned Jupiter Europa Orbiter and Jupiter Ganymede Orbiter missions should extend the abundance coverage to minor and even trace elemental species. For Europa astrobiological investigations it is also important to specify iogenic inputs and surface processing of isotopic species. We discuss the range of abundance distributions arising from models for IO hot volcanic emissions, and from the subsequent dynamics of ion injection, magnetospheric transport, and icy moon surface bombardment.

Cooper, John↗
DOE OSTI2021

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR) 5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of the development and qualification of tri-structural isotropic (TRISO) low enriched fuel for use in high-temperature gas-cooled reactors (HTGR). The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR 5/6/7 experiment were to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel, beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective is to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast (NE) flux trap of the ATR with a planned duration of 500 effective full-power days (EFPDs). The NE flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B?168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 EFPDs. Final burnup values, on a per-compact basis, ranged from 5.66% to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples (TC) had failed (the bottom three capsules lost all TCs). During the first five cycles (162B ? 165A), fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8?10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to the maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particle failures in Capsule 3, which was considered possible at the start of the irradiation, since the experiment was designed to operate beyond the high-temperature gas-cooled reactor (HTGR) normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR)-5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory in support of the development and qualification of tri-structural isotropic low-enriched fuel for use in high-temperature gas-cooled reactors. The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR-5/6/7 experiment were to: • Irradiate reference-design fuel particles to support fuel qualification. • Establish operating margins for the fuel, beyond normal operating conditions. • Provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective was to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR-5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast flux trap of the ATR with a planned duration of 500 effective full-power days. The northeast flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B–168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full-power days. Final burnup values, on a per-compact basis, ranged from 5.66 to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples had failed (the bottom three capsules lost all thermocouples). During the first five cycles (162B – 165A), the fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8–10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to a maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented a fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particles failed in Capsule 3, which was considered possible because the experiment was designed to operate beyond the high-temperature gas-cooled reactor normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to the lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, Cycle 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

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 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↗