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.
Stress corrosion cracking is a major problem in light water nuclear reactors, whether pressurised water reactors (PWRs) or boiling water reactors (BWRs). The nuclear industry needs to be able to predict the service life of these power plants and develop appropriate maintenance and repair practices to ensure safe long term operation. This important book sums up key recent research on corrosion in light water reactors and its practical applications.
The book is divided into four parts. It begins with an overview of materials degradation due to stress corrosion, corrosion potential monitoring and passivation. Part two summarises research on susceptibility of materials to stress corrosion cracking and the ways it can be initiated. The third part of the book considers stress corrosion crack propagation processes whilst the final part includes practical case studies of corrosion in particular plants. The book reviews corrosion in a range of materials such as low alloy steels, stainless steels and nickel-based alloys.
With its distinguished editor and team of contributors, Corrosion issues in light water reactors is a standard work for the nuclear industry.
- Summarises key recent research on corrosion in light water reactors
- Includes practical case studies
The nuclear industry and all those concerned with corrosion research
- No. of pages:
- © Woodhead Publishing 2007
- 22nd June 2007
- Woodhead Publishing
- eBook ISBN:
- Hardcover ISBN:
Dr Damien Féron is Deputy Head of the Service de la Corrosion et du Comportement des Matériaux dans leur Environment at CEA-Saclay and is Chair of the EFC Working Party 4 (Nuclear Corrosion).
Dr Jean-Marc Oliveis a researcher at researcher at C.N.R.S. and is Chair of the EFC Working Party 5 (Environment Sensitive Fracture).
HYDROGENIUS-AIST-Kyushu University, Japan