SCC growth behavior of BWR core shroud materials

International Journal of Pressure Vessels and Piping

Volumn 85, Issue 9, 2008, Pages 582-592

  Yamashita, H ^a   Ooki, S ^b   Tanaka, Y ^b   Takamori, K ^b   Asano, K ^b   Suzuki, S ^b  

^a Utsue Valve Service Co Shinmachi   (Japan)

^b TOKYO ELECTRIC POWER COMPANY   (Japan)

Author keywords

Core shroud; Crack growth rate; IASCC; SCC

Indexed keywords

BOILING WATER REACTORS; CORROSION; CRACK DETECTION; CRACKING (CHEMICAL); GROWTH RATE; PIGMENTS;

CORE SHROUD; GROWTH BEHAVIOR; MATERIAL HARDENING; SCC GROWTH RATES; STRESS CORROSION CRACKING (SCC);

STRESS CORROSION CRACKING;

EID: 45849143262     PISSN: 03080161     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.ijpvp.2008.01.004     Document Type: Article

References (14)

1. Suzuki S., Kumagai K., Okamura Y., Fukuda T., Yamashita H., and Yamashita N. Stress Corrosion Damage Estimation of the core shroud based on JSME NA1-2002 standard. Maintenology 3 2 (2004) (in Japanese)
2. Suzuki S., Takamori K., Kumagai K., Ooki S., Fukuda T., Yamashita H., et al. Evaluation of SCC morphology on L-grade stainless steel in BWRs. J High Pressure Inst Jpn 42 4 (2004) 197 (in Japanese)
3. Kumagai K, Suzuki S, Mizutani J, Shitara C, Yonekura K, Mastuda M, Futami T. Evaluation of IGSCC growth behavior of 316NG PLR piping in BWR. In: Proceedings of the 2004 ASME/JSME pressure vessel and piping conference.
4. Andresen PL. K/size effects on SCC in irradiated, cold worked and unirradiated stainless. In: Proceedings of the 11th international conference on environmental degradation of materials in nuclear power systems-water reactors, Stevenson, Washington, 2003. p. 870-86.
5. Anders Jenssen, Pal Efsing, Karen Gott, Per-Olof Andersson. Crack growth behavior of irradiated type 304L stainless steel in simulated BWR environment. In: Proceedings of the 11th international conference on environmental degradation of materials in nuclear power systems-water reactors, Stevenson, Washington, 2003. p. 1015-28.
6. Ulla Ehrnsten, Pertti Aaltonen, Pertti Nenonen, Hannu Hanninen, Christer Jansson, Thomas Angeliu. Intergranular cracking of AISI 316NG stainless steel in BWR environment. In: Proceedings of the tenth international conference on environmental degradation of materials in nuclear power systems-water reactors.
7. Kodama M, Suzuki S, Nakata K, Nishimura S, Fukuya K, Kato T, Tanaka Y, Shima S. Mechanical properties of various kind of irradiated austenitic stainless steel. In: Proceedings of the eighth international conference on environmental degradation of materials in nuclear power systems-water reactors, American Nuclear Society, La Grange Park, IL, 1997.
8. Castano Martin ML, Garcia Redondo MS, de Diego Velasco G, Gomez Briceno D. Crack growth rate of hardened austenitic stainless steels in BWR and PWR environments. In: Proceedings of the 11th international conference on environmental degradation of materials in nuclear power systems-water reactors, Stevenson, Washington, 2003. p. 845-54.
9. Tetsuo Shoji, Guangfu Li, Junhyun Kwon, Shinobu Matsushima, Zhanpeng Lu. Quantification of yield strength effects on IGSCC of austenitic stainless steel in high temperature water. In: Proceedings of the 11th international conference on environmental degradation of materials in nuclear power systems-water reactors, Stevenson, Washington, 2003. p. 834-44.
10. Torimaru T, Kodama M, Tanaka S, Nakamura T, Asano K, Kumagai K. Neutron flux effects on the irradiation hardening of type 304 stainless steel. In: Proceedings at the 22nd symposium on effects of radiation on materials, Boston, 8-10 June 2004.
11. Chopra OK, Gruber EE, Shack WJ. Crack growth behavior or irradiated austenitic stainless steels in high-purity water at 289C. In: Proceedings of the 11th international conference on environmental degradation of materials in nuclear power systems-water reactors, Stevenson, Washington, 2003. p. 1027-36.
12. Documents presented at the structural integrity evaluation committee, NISA.
13. Obata M., Root J.H., Ishiyama Y., Nataka K., Sakamoto H., Anzai H., et al. Radiation-induced stress relaxation of welded type 304 stainless steel evaluated by neutron diffraction. J ASTM Int 3 1 (2006)
14. Ooki S, tanaka Y, Takamori K, Tanaka S, Suzuki S, Saito Y, Nakamura T, Kato T, Chatani K, Kodama M. Study on SCC growth behavior of BWR core shroud. In: Proceedings of the 12th international conference on environmental degradation of materials in nuclear power systems-water reactors, Salt Lake City, Utah, 2005.