Abstract
Reactor pressure vessels are components that are practically not replaceable and thus they determine lifetime of the whole nuclear power plant (NPP). Lifetime of reactor pressure vessels are determined by level of material degradation as well as by existence of defects found during in-service inspection and their size and location. According to some codes, it is necessary to have a procedure for repair of potential defects found during in-service inspection that can be non-acceptable from point of view of assurance of required RPV lifetime. A wide experimental program was realized to propose, check and approve a procedure for repair of potential defects in underclad region of WWER-440 MWe (PWR type reactors of Russian design) reactor pressure vessels during operation. Experimental program was realized on as-received materials as well as on materials subjected to simulated ageing to level representing expected material embrittlement after design reactor lifetime. Main requirements for this procedure were: RPV will not be pre-heated during repair welding, and no post-weld heat treatment will be applied. Maximum repair depth will be 40 mm including austenitic cladding thickness (8 to 10 mm). Such procedure was developed and checked by series of mechanical testing (hardness, fracture toughness, impact notch toughness as well as by detailed metallography studies).
