low-volatility uranate compounds. Iodine release was observed primarily as CsI, but also as I/sub 2/; in addition, at test temperatures of 900/sup 0/C and above, significant migration of the CsI to the cooler ends of the fuel-rod segments was noted. Tellurium release was markedly restricted by rapid reaction with the Zircaloy cladding. The tests in air yielded enhanced releases of cesium and iodine, and considerable swelling of the oxidized UO/sub 2/. As anticipated, measured release fractions were greater when the test rods were ruptured at temperature by internal pressure than when the cladding failures were machined in the rods prior to testing.

An alternative target design with potential improvements, including a major increase in Pu-238 production rate and annual capacity; fewer targets to be fabricated, irradiated, and processed; and a significant replacement of a large volume of caustic-nitrate, aluminum-bearing radioactive liquid waste with a smaller volume of solid metal waste, has been conceived and evaluated using reactor physics and thermal-hydraulic analyses. The alternative target design uses pressed pellets of (NpO2)-Np-237, sintered to 92% to 93% of theoretical density, and stacked inside a Zircaloy-4 cladding tube. Four test targets were fabricated, irradiated, and examined. No melting or other potential problems were indicated. Projections from measured constituents indicated annual production could be increased by a factor of similar to 2, and the number of targets required to be fabricated, irradiated, and processed could be reduced by a factor of similar to 5.

schedule, this focus on overall performance allows the program to vet the fuel (design) variants and show a connection with the best performing historical fuel. If the good fuel behavior continues, the NGNP conceptual design can be approached with a creditable fuel, supported by early PIE data.

complies with governing requirements. Portions of the LLW system are several decades old, or older, and do not comply with current environmental protection regulations. Several subsystems of the LLW system have been designated to receive a state-of-the-art replacement and refurbishment. One such subsystem serves Building 2026, the High Radiation Level Analytical Laboratory. This assessment focuses on the scope of work for the Building 2026 replacement LLW Collection and Transfer System, including the provision of a new Monitoring and Control Station (Building 2099) to receive, store, and treat (adjust pH) low level radioactive waste.

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly accumulate burnup during separate effects irradiation testing. In general, the burnup and fuel temperatures measured postirradiation were consistent with the design calculations. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU and was achieved after only 68 effective full-power days of irradiation. The dilatometric evaluation of passive silicon carbide thermometry indicated that the fuel was irradiated at temperatures ranging from 410 to 460 degrees C. Because the irradiation temperatures and burnup were low, the UN kernels showed minimal fission gas release that was within the range of the expected recoil (athermal) release. While it is possible to measure fuel swelling using x-ray computed tomography, the observed swelling was too small to quantify in this case. Extensive microstructural characterization of the irradiated fuel was performed, and no significant irradiation induced changes were observed. (C) 2020 Elsevier B.V. All rights reserved.

Release of radionuclides from intact tristructural-isotropic (TRISO) coated particle fuel at normal and accident conditions is a primary metric of fuel performance. The distribution of fission products and actinides in the TRISO layers of individual particles provides insight on radionuclide transport and release behavior and was determined using scanning electron microscopy analysis. Particles were isolated from an irradiated fuel compact (AGR-1 Compact 4-4-2) and analyzed as-irradiated or after individual particle safety-testing at 1800 degrees C for 650 h. Particles were selected for comparison based on their remaining Ag-110m fission product inventory. These comparisons corroborated the observation that the Ag-110m inventory is a marker for relative irradiation temperature based on observed radionuclide distribution in the SiC layer. The comparison also indicated that the in-pile behavior influences the fission product and actinide species interactions with the TRISO layers during high temperature exposure after irradiation. The analysis confirms both palladium and uranium diffusion, as well as other species, are active in the UCO TRISO fuel system at 1800 degrees C and that palladium transport is active at lower temperatures relative to uranium. While diffusion across the SiC layer was observed, the intact nature of the SiC layer after the 1800 degrees C, 650-h exposure indicates the SiC layer maintained its functionality as a fission product barrier by mitigating release of radionuclides at beyond accident margin temperatures. (C) 2020 Published by Elsevier B.V.

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly screen different nuclear fuel concepts. Observations on fission gas release, irradiation conditions and microstructure of the irradiated fuel show good fuel performance at the low burnup achieved in this initial irradiation. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU as was achieved after only 68 effective full power days of irradiation. Results from silicon carbide thermometry measurements further benchmarked the MiniFuel irradiation vehicle and indicated fuel was irradiated at temperatures ranging from 410-460°C. Extensive microstructural characterization on the irradiated fuel was performed and no significant irradiation induced changes were observed.

Post-irradiation examination and elevated-temperature safety testing are being performed on compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). The compacts in the AGR-2 irradiation experiment held either tristructural isotropic (TRISO)-coated particles containing uranium oxide fuel kernels (UO2) or TRISO-coated particles containing fuel kernels with both uranium carbide and uranium oxide phases (UCO). In UO2 TRISO particles, oxygen released by uranium fission can react with the surrounding carbon in the buffer layer to form carbon monoxide (CO). Excess CO can lead to various irradiation performance issues under certain operating conditions, such as pressure-induced fracture, kernel migration, and silicon carbide (SiC) corrosion. In UCO TRISO particles, CO formation is reduced because the chemical potential for oxidation of uranium carbide is lower than for oxidation of carbon.

•More fission product diversity in IPyC/SiC layers for higher temperature compacts.•Temperature is a primary driver for fission product transport across the SiC layer.•Palladium is observed to lead uranium in the SiC layer of as‑irradiated compacts.

The High Burnup Spent Fuel Data Project, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 GW days per tonne uranium) light water reactor fuel. The project includes 32 HBU spent nuclear fuel (SNF) assemblies (the project assemblies) that are stored in a typical independent spent fuel storage installation (ISFSI) and 25 "sister rods"-9 SNF rods that were removed from the fuel assemblies prior to insertion to the ISFSI and 16 SNF rods removed from similar HBU assemblies. The sister rods provide a baseline of the condition of the HBU rods before loading, drying, and long-term dry storage. The project assemblies will be inspected after 10 years, and the physical state of the stored rods will be compared with the condition of the sister rods to identify any changes in physical properties during the dry storage period. This work focuses on key results from the nondestructive postirradiation examinations of the sister rods and summarizes the results of detailed visual examinations, gamma scans, dimensional measurements, and eddy current liftoff measurements of the combined Chalk River unidentified deposits and oxide layer on the waterside surface of the rod. The data are used to calculate fuel rod and pellet stack growth rates, estimated remaining fuel rod plenum volumes, and the percentage change in fuel rod cladding diameter.

In this work, the transmissibility of gas through the pellet stack of full-length pressurized water reactor (PWR) high burnup (HBU) spent nuclear fuel (SNF) rods is measured at room temperature and correlated to a permeability. Gas transmission and depressurization testing of eight high burn-up fuel rods revealed that gas communication from one end of the pellet stack to the other is unobstructed, but slow, at room temperature. Two traditional flow models, Darcy's Law and Muskat's application of Poiseuille's Law to compressible gas flow through porous media are explored. The derived permeability values for the two methods are compared and are considered in conjunction with rod cladding type and various indicators of rod operation such as burnup, High Duty Core Index (HDCI), and average fuel temperature. The derived permeability appears to be correlated to fuel temperature and maximum HDCI, but not to rod average burnup. Three of the HBU rods were heat-treated to simulate dry storage vacuum drying conditions prior to the measurements and the heat-treatments may have resulted in increased permeability.

MeV), and the average compact temperatures during irradiation for these compacts. Capsule 2 Compacts 2-3-2 and 2-3-1 were intentionally irradiated at notably higher temperature than compacts in the other AGR-2 capsules or those previously studied in the AGR-1 irradiation test. The Capsule 2 compacts provide insight into the irradiation and post-irradiation safety test performance at higher temperatures than typical HTGR operating temperatures.

MeV), and the average compact temperatures during irradiation for these compacts. Capsule 2 Compacts 2-3-2 and 2-3-1 were intentionally irradiated at notably higher temperature than compacts in the other AGR-2 capsules or those previously studied in the AGR-1 irradiation test. The Capsule 2 compacts provide insight into the irradiation and post-irradiation safety test performance at higher temperatures than typical HTGR operating temperatures.

MeV), and the average compact temperatures during irradiation for these compacts. Capsule 2 Compacts 2-3-2 and 2-3-1 were intentionally irradiated at notably higher temperature than compacts in the other AGR-2 capsules or those previously studied in the AGR-1 irradiation test. The Capsule 2 compacts provide insight into the irradiation and post-irradiation safety test performance at higher temperatures than typical HTGR operating temperatures.

This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).