Corrosion Issues in Light Water Reactors

Part 1 Overviews: An overview of materials degradation by stress corrosion in PWRs; Corrosion potential monitoring in nuclear power environments; Kinetics of passivation of a nickel-base alloy in high temperature water. Part 2 Stress corrosion cracking: Susceptibility and initiation: IASCC susceptibility under BWR conditions of welded 304 and 347 stainless steels; The effect of lead on resistance of low alloy steel to SCC in high temperature water environment; Effect of cold work hardening on stress corrosion cracking of stainless steels in primary water of pressurized water reactors; Effect of strain-path on stress corrosion cracking of AISI 304L stainless steel in PWR primary environment at 360°C; Dynamic strain ageing of deformed nitrogen-alloyed AISI 316 stainless steels; Laboratory results of stress corrosion cracking of steam generator tubes in a “Complex” environment – An update; The effect of sulphate and chloride transients on the environmentally-assisted cracking behaviour of low-alloy RPV steels under simulated BWR conditions; Transgranular stress-corrosion cracking in austenitic stainless steels at high temperatures. Part 3 Stress corrosion cracking: Propagation: Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions – CASTOC, Part I: BWR/NWC Conditions; Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions – CASTOC, Part II: VVER Conditions; Effect of yield strength on stress corrosion crack propagation under PWR and BWR environments of hardened stainless steels; Corrosion fatigue crack growth behaviour of low-alloy RPV steels at different temperatures and loading frequencies under BWR/NWC environment; Effect of cyclic loadings on the stress corrosion crack growth rate in alloy 600 in PWR primary water; Pattern recognition model to estimate intergranular stress corrosion cracking (IGSCC) at crevices and pit sites of 304 SS in BWRs environments; Fatigue crack growth in austenitic steel AISI 304L in PWR primary water at room temperature and elevated temperature. Part 4 Practical experience: Corrosion damages of 18cr-9ni-?i steel after 25 years of operation in steam-water environments of the VK-50 reactor; Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators; Stress corrosion cracking of a KORI 1 retired steam generator tube; A systematic study of the corrosion effects of the FRAMATOME CORD-UV technology.