This review provides a comprehensive evaluation of the state-of-knowledge of radiation effects in crystalline ceramics that may be used for the immobilization of high-level nuclear waste and plutonium. The current understanding of radiation damage processes, defect generation, microstructure development, theoretical methods, and experimental methods are reviewed. Fundamental scientific and technological issues that offer opportunities for research are identified. The most important issue is the need for an understanding of the radiation-induced structural changes at the atomic, microscopic, and macroscopic levels, and the effect of these changes on the release rates of radionuclides during corrosion.

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During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In zirconium alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup increases when electron transport decreases, requiring hydrogen ingress to close the reaction.

The experimental study of grain growth in nanocrystalline metallic foils under ion irradiation showed the existence of a low-temperature regime ͑below about 0.15-0.22T m ͒, where grain growth is independent of the irradiation temperature, and a thermally assisted regime where grain growth is enhanced with increasing irradiation temperature. A model is proposed to describe grain growth under irradiation in the temperature-independent regime, based on the direct impact of the thermal spikes on grain boundaries. In the model, grain-boundary migration occurs by atomic jumps, within the thermal spikes, biased by the local grain-boundary curvature driving. The jumps in the spike are calculated based on Vineyard's analysis of thermal spikes and activated processes using a spherical geometry for the spike. The model incorporates cascade structure features such as subcascade formation, and the probability of subcascades occurring at grain boundaries. This results in a power law expression relating the average grain size with the ion dose with an exponent equal to 3, in agreement with the experimental observations. The model is applied to grain growth observed in situ in a transmission electron microscope in a wide range of doses, temperature, and irradiation conditions for four different pure metals, and shown to predict well the results in all applicable cases. Some discussions are also presented on the expansion of the model to the thermally assisted regime. The paper is organized in six sections. Section I gives background and literature review, while Secs. II and III review experimental methods and results for in situ grain growth under irradiation. Section IV derives the model proposed to find the grain-growth equation in the nonthermal regime, and in Sec. V the model is applied to the results. In Sec. VI grain growth in the thermally assisted regime is discussed and Sec. VII presents the conclusions.