Volume 35, Issue 3 - Special issue: Materials for accident tolerant fuels / Issue editor: Raúl B. Rebak

Light-water reactor (LWR) fuel cladding shall retain the performance as the barrier for nuclear fuel materials and fission products in high-pressure and high-temperature coolant under irradiation conditions for long periods. The cladding also has to withstand temperature increase and severe loading under accidental conditions. As lessons learned from the accident at the Fukushima Daiichi nuclear power station, advanced cladding materials are being developed to enhance accident tolerance compared to conventional zirconium alloys. The present paper reviews the progress of the development and summarizes the subjects to be solved for enhanced accident-tolerant fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.

Surface-modified zirconium (Zr)-based alloys, mainly by fabricating protective coatings, are being developed and evaluated as accident-tolerant fuel (ATF) claddings, aiming to improve fuel reliability and safety during normal operations, anticipated operational occurrences, and accident scenarios in water-cooled reactors. In this overview, the performance of Zr alloy claddings under normal and accident conditions is first briefly summarized. In evaluating previous studies, various coating concepts are highlighted based on coating materials, focusing on their performance in autoclave hydrothermal corrosion tests and high-temperature steam oxidation tests. The challenges for the utilization of coatings, including materials selection, deposition technology, and stability under various situations, are discussed to provide some valuable guidance to future research activities.

The very high temperature (≥1400°C) steam oxidation resistance of thin-walled FeCrAl tubes is being investigated as an alternative to Zr-based alloys for enhanced accident-tolerant light water reactor fuel cladding. Initial work examined commercial FeCrAl with ~20% Cr in ramp testing with 1 min hold times at temperatures up to 1700°C. At 1400–1500°C, excellent oxidation resistance was observed with thin external alumina scales formed in contrast to the thick oxides formed on Zircaloy-4 under similar conditions. For an optimized FeCrAl with 13% Cr, one batch of tubing performed poorly at 1400°C, while the second batch formed a protective scale at 1400°C but was fully oxidized at 1500°C. Differences in performance between two test rigs suggest a role of gas velocity, and initial work on bulk alumina specimens has quantified an evaporation rate. However, some results suggest that melting is occurring well below 1500°C. This behavior is still being investigated.

To avoid accidents like that in Fukushima, the US Department of Energy is engaged with a nuclear fuel vendor to evaluate the performance of iron-chromium-aluminum (FeCrAl) alloys such as advanced powder metallurgy tubing (APMT) as accident-tolerant material for uranium dioxide fuel cladding in light water reactors (LWR). It was important to characterize the oxides formed on APMT under both boiling water reactor (BWR) and pressurized water reactor (PWR) environments for a better understanding of its environmental sustainability in LWRs. Coupons of APMT were exposed for 1 year to both hydrogenated and oxygenated high-purity water at 288°C (e.g. simulated BWR water chemistry without Pt injection) and hydrogenated high-purity water at 330°C (e.g. simulated primary PWR water chemistry without Li/B addition). Results show that after 1-year immersion, APMT always developed a chromium-rich protective oxide film on its surface. In oxygen-containing environments, the oxide consisted of a dual layer, an external thicker layer containing mostly iron oxides and a thinner internal layer rich in chromium oxide. In hydrogen environments, only a single oxide layer formed, consisting of chromium oxide. This is a similar finding as for type 304 and 316 stainless steels and for nickel-based alloy 600, which is extensively reported in the literature. General corrosion of APMT alloys under LWR operating conditions would not be a limiting factor for its performance as cladding material.