Stress corrosion cracking of Ni base alloys in Pb-contaminated caustic solutions
- 1. Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Description
Steam generator tubes in pressurized water reactors(PWRs) form a pressure boundary between the primary and secondary sides. Austenitic stainless steel was used initially for the tubing, but it was changed to Alloy 600 as the results of corrosion problems. Experiences with Alloy 600 in the late 1960s and early 1970s, in the early high temperature large PWRs, identified numerous corrosion problems. Efforts for the development of a new made and the result was Alloy 690. Since the mid-1980s new and replacement steam generators have used Alloy 690. However, PWRs with Alloy 600 steam generators are still being operated. A lot of problems related to corrosion have been reported in Alloy 600 steam generator tubes of operating nuclear power plants(NPPs), and the outer diameter stress corrosion cracking (ODSCC) and intergranular attack(IGA) which have been occurring in Alloy 600 tubes are known to be the leading causes of PWR steam generator tube plugging in the USA and worldwide. According to Smith and Stratton, Alloy 690 is also reported to be susceptible to stress corrosion cracking and intergranular attack in deaerated (with argon) caustic solutions. It has been reported that contaminated lead in the secondary side cooling water is accumulated in the sludge piled on top of the tube sheet thus accelerating stress corrosion cracking in the SG tubes of NPPs. The detailed mechanism of accelerated stress corrosion cracking of Alloy 600 and 690 by Pb, however, has yet to be completely understood. It was observed that Pb dissolved in water can produce PbSCC at a Pb concentration as low as 0.1 ppm in these alloys. Lead is known to be one of the most aggressive environmental species that can accumulate in the crevice between the tubes and sludge piled on TTS in steam generators. Many laboratory experiments indicate that the stress corrosion cracking of steam generator tubing materials is accelerated in the presence of lead species in a caustic environment. In order to observe the effect of Pb on stress corrosion cracking, it is needed to investigate stress corrosion cracking susceptibility of Alloy 600 and 690 in solutions with Pb. In this study, stress corrosion cracking tests were carried out with Alloy 600 and 690 in caustic solutions with Pb
Additional details
Publishing Information
- Publisher
- KNS
- Imprint Place
- Daejeon (Korea, Republic of)
- Imprint Title
- Proceedings of the KNS spring meeting
- Imprint Pagination
- [1 CD-ROM]
- Journal Page Range
- [2 p.]
Conference
- Title
- 2009 spring meeting of the KNS
- Dates
- 18-23 May 2009
- Place
- Jeju (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 40101267
- Subject category
- S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- CRACKING; INCONEL 600; INCONEL 690; NICKEL BASE ALLOYS; PWR TYPE REACTORS; STRESS CORROSION; TUBES
- Descriptors DEC
- ALLOY-NI59CR30FE9; ALLOY-NI76CR15FE8; ALLOYS; ALUMINIUM ADDITIONS; ALUMINIUM ALLOYS; CHEMICAL REACTIONS; CHROMIUM ALLOYS; CORROSION; CORROSION RESISTANT ALLOYS; DECOMPOSITION; ENRICHED URANIUM REACTORS; HEAT RESISTANT MATERIALS; HEAT RESISTING ALLOYS; INCONEL ALLOYS; IRON ALLOYS; MATERIALS; NICKEL ALLOYS; NICKEL BASE ALLOYS; NIMONIC; POWER REACTORS; PYROLYSIS; REACTORS; THERMAL REACTORS; THERMOCHEMICAL PROCESSES; TITANIUM ADDITIONS; TITANIUM ALLOYS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 5 refs, 3 figs, 2 tabs