Abstract
Spent nuclear fuels (SNFs) are stored in stainless steel canisters at Independent Spent Fuel Storage Installations (ISFSIs) typically near the seashore. During long-term storage of these canisters in the dry cask storage system (DCSS), chloride-induced stress corrosion cracking (CISCC) could occur due to the deliquescence of concentrated salt deposits on the canister surface. To evaluate such flaws on the accessible exterior metallic portions of containment systems while in service, the ASME Section XI Code Case N-860 provides inservice inspection requirements for aging management of canisters manufactured with welded austenitic stainless steels. It is noteworthy that CISCC crack growth rate (CGR) model in Code Case N-860 consists of the constitutive equations with temperatures (canister surface temperatures, storage site yearly mean temperature, and ambient temperature measured at overpack inlet) and is independent of stress intensity factor or other environmental factors.
In this work, the mean temperature effect of the local storage site on the CGR is analyzed based on the CISCC CGR model in Code Case N-860. The specified mean temperature in the Code Case is calculated yearly however, the crack growth by CISCC can be evaluated differently if the mean temperature of storage site with a large annual range of temperature such as South Korea is applied. In that case, the monthly mean temperature is adjusted as the yearly mean temperature so the effect of averaging range for calculating the mean temperature is analyzed. Firstly, climate data of some candidate sites for the storage in South Korea are measured from Korea Meteorological Administration (KMA). The climate data of the Diablo Canyon Power Plant located in California (United States) is obtained from National Weather Service (NWS) for comparison. Yearly data from 2012 to 2020 are applied and the crack growth is estimated for sites of different annual ranges.
