Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR

Joyce Conermann; Regis Shogan; Koji Fujimoto; Toshio Yonezawa; Yoichiro Yamaguchi

Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR

Joyce Conermann, Regis Shogan, Koji Fujimoto, Toshio Yonezawa, Yoichiro Yamaguchi

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

Abstract

Mechanical and corrosion properties were measured on a cold worked, Type 316 stainless steel tube removed from the core of a Pressurized Water Reactor (PWR) after 23 years of service. Neutron exposure levels of the material ranged from near 0 to 65 dpa (∼4.5×10²² n/cm², E>1 MeV) and irradiation temperatures from 290 to 320°C. As expected, the strength of the material increased and the ductility decreased with irradiation. Slow strain rate and stressed O-ring crack initiation tests in PWR water were used to determine the susceptibility of the material to irradiation assisted stress corrosion cracking (IASCC). The data indicate that IASCC susceptibility may accelerate above 20 dpa and saturate after 60 dpa. At the highest fluence, limited intergranular failure was observed on the fracture surface of a specimen slow strain rate tested in inert gas.

Original language	English
Title of host publication	Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
Pages	277-287
Number of pages	11
Publication status	Published - 2005 Dec 1
Event	12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors - Salt Lake City, UT, United States Duration: 2005 Aug 14 → 2005 Aug 18

Publication series

Name	Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors

Other

Other	12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
Country	United States
City	Salt Lake City, UT
Period	05/8/14 → 05/8/18

Keywords

IASCC
PWR
Stainless Steel
Swelling

ASJC Scopus subject areas

Engineering(all)

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Cite this

Conermann, J., Shogan, R., Fujimoto, K., Yonezawa, T., & Yamaguchi, Y. (2005). Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR. In Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (pp. 277-287). (Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors).

Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR. / Conermann, Joyce; Shogan, Regis; Fujimoto, Koji; Yonezawa, Toshio; Yamaguchi, Yoichiro.

Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. 2005. p. 277-287 (Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors).

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution

Conermann, J, Shogan, R, Fujimoto, K, Yonezawa, T & Yamaguchi, Y 2005, Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR. in Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, pp. 277-287, 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Salt Lake City, UT, United States, 05/8/14.

Conermann J, Shogan R, Fujimoto K, Yonezawa T, Yamaguchi Y. Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR. In Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. 2005. p. 277-287. (Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors).

Conermann, Joyce ; Shogan, Regis ; Fujimoto, Koji ; Yonezawa, Toshio ; Yamaguchi, Yoichiro. / Irradiation effects in a highly irradiated cold worked stainless steel removed from a commercial PWR. Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. 2005. pp. 277-287 (Proceedings of the Twelfth International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors).

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abstract = "Mechanical and corrosion properties were measured on a cold worked, Type 316 stainless steel tube removed from the core of a Pressurized Water Reactor (PWR) after 23 years of service. Neutron exposure levels of the material ranged from near 0 to 65 dpa (∼4.5×1022 n/cm2, E>1 MeV) and irradiation temperatures from 290 to 320°C. As expected, the strength of the material increased and the ductility decreased with irradiation. Slow strain rate and stressed O-ring crack initiation tests in PWR water were used to determine the susceptibility of the material to irradiation assisted stress corrosion cracking (IASCC). The data indicate that IASCC susceptibility may accelerate above 20 dpa and saturate after 60 dpa. At the highest fluence, limited intergranular failure was observed on the fracture surface of a specimen slow strain rate tested in inert gas.",

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N2 - Mechanical and corrosion properties were measured on a cold worked, Type 316 stainless steel tube removed from the core of a Pressurized Water Reactor (PWR) after 23 years of service. Neutron exposure levels of the material ranged from near 0 to 65 dpa (∼4.5×1022 n/cm2, E>1 MeV) and irradiation temperatures from 290 to 320°C. As expected, the strength of the material increased and the ductility decreased with irradiation. Slow strain rate and stressed O-ring crack initiation tests in PWR water were used to determine the susceptibility of the material to irradiation assisted stress corrosion cracking (IASCC). The data indicate that IASCC susceptibility may accelerate above 20 dpa and saturate after 60 dpa. At the highest fluence, limited intergranular failure was observed on the fracture surface of a specimen slow strain rate tested in inert gas.

AB - Mechanical and corrosion properties were measured on a cold worked, Type 316 stainless steel tube removed from the core of a Pressurized Water Reactor (PWR) after 23 years of service. Neutron exposure levels of the material ranged from near 0 to 65 dpa (∼4.5×1022 n/cm2, E>1 MeV) and irradiation temperatures from 290 to 320°C. As expected, the strength of the material increased and the ductility decreased with irradiation. Slow strain rate and stressed O-ring crack initiation tests in PWR water were used to determine the susceptibility of the material to irradiation assisted stress corrosion cracking (IASCC). The data indicate that IASCC susceptibility may accelerate above 20 dpa and saturate after 60 dpa. At the highest fluence, limited intergranular failure was observed on the fracture surface of a specimen slow strain rate tested in inert gas.

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