In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

NFDC (Nuclear Fuel and materials Data Center) developed standard reference data for oxidation of HANA-6 cladding material. Thermo-gravimetric analyzer (TGA) was used to measure oxidation, and the measuring device was self-calibrated using standard materials. The oxidation amount of the HANA6 cladding was measured in an oxidizing atmosphere in the temperature range of 400 to 700°C. Through this, oxidation data, oxidation rate model equation, and graph were developed. The uncertainty factors were analyzed from the oxidation model. The expanded uncertainty of oxidation data was calculated by evaluating the uncertainty for each uncertainty factor. The oxidation data produced in this study was self-rated through deliberation by a specialized committee of NFDC and third experts. It was finally registered as a reference standard through the technical committee of the National Reference Standards Center. It is believed that the standard reference data developed in this study will be helpful for increasing reliability and stability evaluation of nuclear fuel and spent fuel.