Corrosion and Hydriding Model for Zircaloy-2 Pressure Tubes of Indian Pressurised Heavy Water Reactors

摘要

Early generation of Indian pressurised heavy water reactor (PHWR) units-MAPS-1 and 2, NAPS-1 and 2, and KAPS-1 had used Zircaloy-2 pressure tubes. Corrosion of the zirconium alloy pressure tube in the high temperature (250°C-300°C) heavy water coolant flowing through it results in formation of an oxide layer on its inside surface and evolution of deuterium (for its chemical similarity with hydrogen, it will be described as hydrogen). A part of this hydrogen is absorbed by the pressure tube material. Gradual build-up of hydrogen causes degradation in the structural integrity of the pressure tube with manifestations of either one or a combination of the nucleation and growth of hydride blisters, hydride embrittlement at service induced flaw tip, and lowering of fracture toughness of the material. Safety assessment of the operating pressure tubes against these hydride induced degradation mechanisms requires a conservative estimate of hydrogen concentration in each of these pressure tubes. Although hydrogen ingress into a pressure tube during service may be estimated from the material samples taken out from the inside surface of the tube by sliver scrape sampling technique, such exercise is not feasible to be carried out on a large number of pressure tubes. Alternatively, the numerical model for corrosion and hydrogen pickup developed using the database created by the hydrogen measured in the bulk samples from the pressure tubes removed from the different reactor units for material surveillance purposes can be used for conservatively estimating the hydrogen pickup. The present paper describes the methodology adopted for developing a numerical model for in-reactor corrosion and hydriding of Zircaloy-2 material using data on oxide thickness and hydrogen pickup generated from the pressure tubes removed from the operating Indian units.