Оцінка можливості виникнення сповільненого гідридного розтріскування в оболонках паливних стрижнів зі сплавів Е110 та ZIRLO за термічного впливу під час реалізації технології зберігання відпрацьованих тепловидільних збірок у ВБК ССВЯП ВП ЗАЕС

• Date: 2025
• Main Authors: Zuyok, V., Kostenevych, O., Milenin, O., Makhnenko, O., Tretyakov, M., Hrudnytskyi, V., Lytovchenko, S.
• Format: Article
• Language: English
• Published: State Scientific and Technical Center for Nuclear and Radiation Safety 2025
• Online Access: https://nuclear-journal.com/index.php/journal/article/view/1280
• Journal Title: Nuclear and Radiation Safety

An analysis of available data on the conditions for the initiation of delayed hydride cracking (DHC) as one of the mechanisms that can cause integrity loss of fuel zirconium cladding under dry storage conditions was performed. Based on the available experimental values for different zirconium alloy grades, a recommendation was made to use 5.0±2,5 MPa·m1/2 as a conservative critical value for the stress intensity factor in the vicinity of a crack, whose exceeding may result in DHC. Stresses arising in the fuel cladding can be caused by mechanical interaction between the fuel pellet and the cladding, pressure created by gaseous fission products (GFPs) under the cladding, and residual stresses in the weld joint zone. In the implementation of the dry storage technology, the maximum reduction in the radial gap between the fuel pellet and the cladding does not exceed 2.2 μm (for E110 cladding at a GFP pressure of 2.85 MPa), which is less than the minimum initial gap typical for irradiated fuel rods (15 μm). Therefore, it was concluded that the formation/growth of stresses in the fuel cladding, which could cause DHC initiation due to mechanical contact between the fuel pellet and the cladding, can be disregarded. Using the VERLIFE methodology, the stress intensity factor (KI) was calculated for the maximum permissible size (depth) of a postulated internal crack in the nuclear power industry, which is 0.25 of its thickness. This size is greater than the maximum fretting wear (10%) of the fuel cladding and is therefore more conservative. The set of calculations enabled to conclude that under thermal impact in the implementation of the dry storage technology, which includes vacuum drying followed by filling the Multi-Purpose Canister (MPC) with helium until a steady state is reached at a temperature of 350 °C, and during long-term storage of spent nuclear fuel (SNF) in the Ventilated Concrete Casks (VCCs) at the Dry Spent Fuel Storage Facility (DSFSF) site under normal conditions, due to the internal pressure under the cladding (GFPs, pellet), DHC will not be initiated, since the maximum calculated KI value for rods with E110 and ZIRLO alloy claddings is 1.81 MPa·m1/2, which is typical for ZIRLO alloy cladding at the vacuum drying stage (436 °C) and does not exceed the conservative critical value KIC = 5.0±2,5 MPa·m1/2.