In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). To investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

In this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors. The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at similar to 380 degrees C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430 degrees C-500 degrees C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460 degrees C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at similar to 380 degrees C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460 degrees C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation. (c) 2022 Elsevier B.V. All rights reserved.

The microstructural evolution of ternary Fe-21Cr-32Ni (21Cr32Ni) model alloy and of alloy 800H was investigated with a series of in-situ ion irradiation experiments performed using the Intermediate Voltage Electron Microscope (IVEM)-Tandem Facility at Argonne National Laboratory (ANL). Samples were irradiated in-situ with 1 MeV Kr++ ions in the temperature range of 50K to 713K to doses up to 2 dpa. The size distribution of defect clusters, average defect cluster diameter, and defect cluster density were measured and compared. Results showed that the evolution of defects (i.e. the average defect cluster size and number density) in 21Cr32Ni model alloy and alloy 800H with dose were similar at irradiation temperatures up to 300K where they initially increased with dose up to 0.1 dpa after which no significant changes in defect size and density were observed with further irradiation. In addition, both alloys exhibited ordered defect structures along the <100> direction at relatively low temperatures, up to 300K, which remained stable throughout post-irradiation in-situ thermal annealing up to a temperature of 773K. During irradiation at 713K, small defect clusters were observed at low doses (<0.1 dpa) in both alloys. However, at this irradiation temperature, the clusters in 21Cr32Ni grew with a faster rate than those formed in alloy 800H, causing the microstructure in the former to be dominated by numerous large dislocation loops having both {111}- and {110}-type habit planes, and in the latter to be dominated by small defect clusters and small {111}-type dislocation loops. This may indicate that defect trapping by the solute atoms in alloy 800H at 713K can slow point defect migration to defect clusters and limit their growth.

Austenitic 21Cr32Ni model alloy thin foils, previously irradiated with 5 MeV Fe++ ions in bulk to create voids, were re-irradiated in-situ in the Intermediate Voltage Electron Microscope Facility (IVEM). The voids which had been formed under bulk-ion irradiation shrank and disappeared after in-situ Kr ion irradiation in the temperature range 50 K-713 K to an additional dose of 1 dpa. The voids were unaffected by eithersuccessive thermal annealing to 673 K and by prolonged exposure to the 200 keV electron beam at the irradiation temperature. The high void shrinkage rate observed did not change significantly for irradiation temperatures between 50 K and 713 K, suggesting that the void shrinkage process in thin foils during in-situ heavy-ion irradiation results from the interactions of displacement cascades with the voids. Possible void shrinkage mechanisms under thin foil irradiation are discussed in this study. (C) 2020 Elsevier B.V. All rights reserved.

Ion irradiation has been widely used to deliver high doses of irradiation since neutron irradiation requires very long exposure times due to its low damage rates, and the usage of hot cells to handle activated samples. One challenge to correlating the two is the absence of helium during the ion irradiation because no transmutation reaction takes place during ion irradiation. In nuclear reactors, swelling behavior can be significantly affected by the presence of helium generated from (n, ) reactions. This is because helium is insoluble at low temperatures (unlike other residual gases) and therefore, it can stabilize cavities in the matrix. To address this question helium can be implanted simultaneously to emulate reactor irradiation.

The microstructural evolution of the 21Cr32Ni model alloy under ion irradiation is investigated. A set of bulk materials were irradiated at the Michigan Ion Beam Laboratory using single beam (5 MeV Fe++) to 1, 10 and 20 dpa at 440 degrees C and dual beam (5 MeV Fe++ plus energy degraded 1.95 MeV He++ ions) to 16.6 dpa at 446 degrees C. The average diameter and number density of the faulted loops and cavities formed under irradiation were characterized using Transmission Electron Microscopy (TEM). The behavior of faulted loop in the model alloy was also investigated in-situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory (ANL). Results show that the average faulted loop diameter decreases, but the faulted loop number density increases with increasing dose. In-situ experiments showed that the faulted loops become unfaulted during ion irradiation by interacting with network dislocations. Although the average faulted loop diameter after 16.6 dpa dual beam irradiation at 446 degrees C was found to be similar to those seen in samples irradiated with single beams to 10 and 20 dpa, the faulted loop number density was significantly higher in the dual beam irradiated sample. Moreover, the dual beam irradiated model alloy exhibits a significantly higher density of smaller cavities. It is also found that the size and density of the faulted loops and voids calculated for the dual beam irradiation of 21Cr32Ni model alloy at 446 degrees C are in better agreement with those measured in a sample neutron irradiated at 375 degrees C. Further discussion is presented in this study. (C) 2018 Elsevier B.V. All rights reserved.

The microstructure of the materials in the reactor core continuously evolves during reactor operation due to the atomic displacements resulting from interactions with high-energy neutrons. For some proposed fast reactors, the amount of displacement damage is expected to reach up to ~200 dpa (displacements per atom) at operating temperatures above 400C[1]. Therefore, understanding of how materials behave under irradiation is essential. The major challenge of neutron irradiation experiments is the requirement of very long exposure times, as well as the handling of highly radioactive samples needing the usage of hot cells which increases costs.