Abstract
Japan Atomic Energy Research Institute (JAERI) and Power Reactor and Nuclear Fuel Development Corporation (PNC) initiated a cooperative research on the evaluation of fracture behaviors of neutron irradiated material in 1991. The research program includes a study of stress corrosion cracking (SCC) of the wrapper tube material of fuel assembly irradiated in the experimental fast reactor `JOYO`. By August 1992 SCC tests by the slow strain rate technique (SSRT) were carried out on the irradiated material. In this cooperative program the wrapper tube material of type 316 stainless steel irradiated in `JOYO` up to a neutron fluence of 8 x 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) at about 425degC was tested by SSRT at 60degC, 200degC and 300degC in the water environment at the hot laboratory of JAERI. The tests showed that the irradiated wrapper tube material fractured by fully ductile mode at 60degC in high purity water with dissolved oxygen of 32 ppm but it was fractured by intergranular cracking at 300degC in the same environment. It is concluded that the stainless steel irradiated in FBR has no susceptibility to SSC at 60degC but susceptibility appears at the higher temperatures in the water. (author).
Tsukada, Takashi;
Shiba, Kiyoyuki;
Nakajima, Hajime
[1]
- Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; and others
Citation Formats
Tsukada, Takashi, Shiba, Kiyoyuki, and Nakajima, Hajime.
Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, (1).
Japan: N. p.,
1992.
Web.
Tsukada, Takashi, Shiba, Kiyoyuki, & Nakajima, Hajime.
Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, (1).
Japan.
Tsukada, Takashi, Shiba, Kiyoyuki, and Nakajima, Hajime.
1992.
"Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, (1)."
Japan.
@misc{etde_10123242,
title = {Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, (1)}
author = {Tsukada, Takashi, Shiba, Kiyoyuki, and Nakajima, Hajime}
abstractNote = {Japan Atomic Energy Research Institute (JAERI) and Power Reactor and Nuclear Fuel Development Corporation (PNC) initiated a cooperative research on the evaluation of fracture behaviors of neutron irradiated material in 1991. The research program includes a study of stress corrosion cracking (SCC) of the wrapper tube material of fuel assembly irradiated in the experimental fast reactor `JOYO`. By August 1992 SCC tests by the slow strain rate technique (SSRT) were carried out on the irradiated material. In this cooperative program the wrapper tube material of type 316 stainless steel irradiated in `JOYO` up to a neutron fluence of 8 x 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) at about 425degC was tested by SSRT at 60degC, 200degC and 300degC in the water environment at the hot laboratory of JAERI. The tests showed that the irradiated wrapper tube material fractured by fully ductile mode at 60degC in high purity water with dissolved oxygen of 32 ppm but it was fractured by intergranular cracking at 300degC in the same environment. It is concluded that the stainless steel irradiated in FBR has no susceptibility to SSC at 60degC but susceptibility appears at the higher temperatures in the water. (author).}
place = {Japan}
year = {1992}
month = {Nov}
}
title = {Irradiation assisted stress corrosion cracking of stainless steel irradiated in FBR, (1)}
author = {Tsukada, Takashi, Shiba, Kiyoyuki, and Nakajima, Hajime}
abstractNote = {Japan Atomic Energy Research Institute (JAERI) and Power Reactor and Nuclear Fuel Development Corporation (PNC) initiated a cooperative research on the evaluation of fracture behaviors of neutron irradiated material in 1991. The research program includes a study of stress corrosion cracking (SCC) of the wrapper tube material of fuel assembly irradiated in the experimental fast reactor `JOYO`. By August 1992 SCC tests by the slow strain rate technique (SSRT) were carried out on the irradiated material. In this cooperative program the wrapper tube material of type 316 stainless steel irradiated in `JOYO` up to a neutron fluence of 8 x 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) at about 425degC was tested by SSRT at 60degC, 200degC and 300degC in the water environment at the hot laboratory of JAERI. The tests showed that the irradiated wrapper tube material fractured by fully ductile mode at 60degC in high purity water with dissolved oxygen of 32 ppm but it was fractured by intergranular cracking at 300degC in the same environment. It is concluded that the stainless steel irradiated in FBR has no susceptibility to SSC at 60degC but susceptibility appears at the higher temperatures in the water. (author).}
place = {Japan}
year = {1992}
month = {Nov}
}