Irradiation assisted stress corrosion cracking

Pokor, C.*; Herbelin, A.*; Couvant, T.*; Kaji, Yoshiyuki

NEA/NSC/R(2016)5 (Internet), p.317 - 360, 2017/05

In aged BWR plants, certain locations in the mid-plane of the core shroud experience fluence levels at which the materials become susceptible to irradiation assisted stress corrosion cracking (IASCC). BWRVIP (Boiling Water Reactor Vessel Internals Program) has developed crack growth disposition methodologies for evaluating intergranular stress corrosion cracking (IGSCC) in the internal components of BWRs and the Japan Nuclear Energy Safety organization (JNES) has been conducting a project related to IASCC crack growth rate data as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Although many investigators proposed prediction models for SCC and IASCC growth rates for austenitic stainless steels and Ni alloys, even more improvements of models are necessary as compared with the detailed experimental results, because these models are still preliminary models.

Proceedings of the Workshop on Reactor Safety Research, Focusing on the Integrity of Aged Components; March 17, 2003, Tokai Research Establishment, Tokai-mura

Hidaka, Akihide; Suzuki, Masahide

JAERI-Conf 2003-014, 178 Pages, 2003/09

JAERI-Conf-2003-014.pdf:19.17MB

The Workshop on Reactor Safety Research focusing on the integrity of aged components was held at the Tokai Research Establishment on March 17, 2003. The purpose of the Workshop was to obtain useful information to proceed with the reactor safety research in future and to resolve the issues on the integrity evaluation of aged components through the discussions followed by the presentations on the results of the research at JAERI on all the research subjects assigned to JAERI in the Five-Year Program of Safety Research for Nuclear Installations established by the Nuclear Safety Commission, and on those of the studies at JAERI on the integrity of core shrouds of BWR plants. Thirty-eight people from outside JAERI including the press such as Nihon Television Network Corporation and Shin-Ibaraki Shinbun and fifty-seven people from JAERI attended the Workshop. This proceeding compiles all the viewgraphs presented in the workshop, the opinions of participants for forum and the answers, and summary of questionnaire on workshop.

Design of water feeding system for IASCC irradiation tests at JMTR

Kanno, Masaru; Nabeya, Hideaki; Mori, Yuichiro*; Matsui, Yoshinori; Tobita, Masahiro*; Ide, Hiroshi; Itabashi, Yukio; Komori, Yoshihiro; Tsukada, Takashi; Tsuji, Hirokazu

JAERI-Tech 2001-080, 57 Pages, 2001/12

JAERI-Tech-2001-080.pdf:2.34MB

no abstracts in English

Status of JAERI material performance database (JMPD) and analysis of irradiation assistd stress corrosion cracking (IASCC) data

Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Science and Technology, 37(11), p.949 - 958, 2000/11

https://doi.org/10.3327/jnst.37.949

no abstracts in English

Post-irradiation mechanical properties of austenitic alloys at temperatures below 703K

Jitsukawa, Shiro; Ioka, Ikuo; Hishinuma, Akimichi

Journal of Nuclear Materials, 271-272, p.167 - 172, 1999/00

https://doi.org/10.1016/S0022-3115(98)00775-2

no abstracts in English

Effect of irradiation temperature on irradiation assisted stress corrosion cracking of model austenitic stainless steels

Tsukada, Takashi; Miwa, Yukio; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Materials, 258-263, p.1669 - 1674, 1998/00

https://doi.org/10.1016/S0022-3115(98)00317-1

no abstracts in English

Post irradiation test facilities for irradiation assisted stress corrosion cracking research

Tsukada, Takashi; Shiba, Kiyoyuki; Omi, Masao; Kizaki, Minoru; ; Nakajima, Hajime

Proc. of the 3rd Asian Symp. on Research Reactor, 8 Pages, 1991/00

no abstracts in English

Influence of temperature histories during reactor startup periods on IASCC susceptibility of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Chimi, Yasuhiro; Kitsunai, Yuji*; Koshiishi, Masato*

no journal, , 

According to existing data of slow strain rate tensile (SSRT) test under high temperature water conditions, which simulated BWR primary coolant environment, low carbon stainless steel, which was irradiated with neutrons up to about 3 dpa in BWR core, shows susceptibility of Irradiation associated stress corrosion cracking (IASCC). On the other hand, the stainless steel irradiated by using the JMTR did not show IASCC susceptibility, regardless of neutron fluence. To investigate this different result about IASCC susceptibility, the JMTR operated to simulate temperature history at start-up of BWR, and a tensile specimen of SUS316L was irradiated up to about 3 dpa under the condition. After that, the specimen was examined by SSRT test to evaluate IASCC susceptibility. The result of fracture surface observation after the SSRT test indicated that the specimen fractured by Inter-granular mode and was evaluated to be susceptible to IASCC. In the comparison of the data of IASCC sensitivity by the JMTR irradiated materials, which did not show IASCC susceptibility, the difference of them was suggested to attribute to different temperature histories at the start of irradiation. The relationship between IASCC susceptibility and the parameters obtained from tensile tests was discussed, in consideration of the difference of the tensile parameters which are suffered from the irradiation condition under the different temperature history during the start period of the irradiation.