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Facility for stress corrosion cracking test of irradiated material in high temperature water

Technical Report:

Abstract

Irradiation assisted stress corrosion cracking (hereafter, IASCC) is a synergistic effect of neutron irradiation, aqueous environment and stress on nuclear core materials. In order to investigate the IASCC a facility for stress corrosion cracking test of irradiated materials was developed. The slow strain rate technique (SSRT) in high temperature and pressure water environment was employed. The post irradiation SSRT test facility consists of a SSRT test machine and a water circulation and purification system. In the design of the facility attentions were paid to safe and reliable remote operation of heavily irradiated materials in hot cell. Main specification of the facility is as follows; Maximum load capacity : 30 kN, Strain rate range : 10{sup -7} s{sup -1} {approx} 10{sup -3} s{sup -1}, Test temperature/pressure : 300degC and 10 MPa in maximum, Water supply rate : 5 l/h, Controllable range of dissolved oxygen (DO) : 0.01 {approx} 32 ppm. Water chemistry is monitored on DO, electric conductivity, and pH. Electrochemical potential and electric current are monitored and controlled on specimen. A preliminary test was carried out on the type 316 stainless steel irradiated up to 8 dpa in Oak Ridge Research Reactor under spectrally tailored condition. As a result of  More>>
Authors:
Tsukada, Takashi; Shiba, Kiyoyuki; Nakajima, Hajime; Kizaki, Minoru; Sudo, Kenji; [1]  Bell, G E.C.; Ohmi, Masao; Goto, Ichiro
  1. Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Publication Date:
Jun 01, 1992
Product Type:
Technical Report
Report Number:
JAERI-M-92-081
Reference Number:
SCA: 420500; PA: JPN-92:011110; SN: 93000918515
Resource Relation:
Other Information: PBD: Jun 1992
Subject:
42 ENGINEERING; STRESS CORROSION; CRACKS; IRRADIATION DEVICES; IRRADIATION; STRAINS; MANIPULATORS; STEEL-CR17NI12MO3; HOT CELLS; NEUTRONS; AUTOCLAVES; COOLANT LOOPS; TEMPERATURE RANGE 0400-1000 K; 420500; MATERIALS TESTING
OSTI ID:
10111140
Research Organizations:
Japan Atomic Energy Research Inst., Tokyo (Japan)
Country of Origin:
Japan
Language:
Japanese
Other Identifying Numbers:
Other: ON: DE93753189; TRN: JP9211110
Availability:
OSTI; NTIS; INIS
Submitting Site:
JPN
Size:
35 p.
Announcement Date:
Jun 30, 2005

Technical Report:

Citation Formats

Tsukada, Takashi, Shiba, Kiyoyuki, Nakajima, Hajime, Kizaki, Minoru, Sudo, Kenji, Bell, G E.C., Ohmi, Masao, and Goto, Ichiro. Facility for stress corrosion cracking test of irradiated material in high temperature water. Japan: N. p., 1992. Web.
Tsukada, Takashi, Shiba, Kiyoyuki, Nakajima, Hajime, Kizaki, Minoru, Sudo, Kenji, Bell, G E.C., Ohmi, Masao, & Goto, Ichiro. Facility for stress corrosion cracking test of irradiated material in high temperature water. Japan.
Tsukada, Takashi, Shiba, Kiyoyuki, Nakajima, Hajime, Kizaki, Minoru, Sudo, Kenji, Bell, G E.C., Ohmi, Masao, and Goto, Ichiro. 1992. "Facility for stress corrosion cracking test of irradiated material in high temperature water." Japan.
@misc{etde_10111140,
title = {Facility for stress corrosion cracking test of irradiated material in high temperature water}
author = {Tsukada, Takashi, Shiba, Kiyoyuki, Nakajima, Hajime, Kizaki, Minoru, Sudo, Kenji, Bell, G E.C., Ohmi, Masao, and Goto, Ichiro}
abstractNote = {Irradiation assisted stress corrosion cracking (hereafter, IASCC) is a synergistic effect of neutron irradiation, aqueous environment and stress on nuclear core materials. In order to investigate the IASCC a facility for stress corrosion cracking test of irradiated materials was developed. The slow strain rate technique (SSRT) in high temperature and pressure water environment was employed. The post irradiation SSRT test facility consists of a SSRT test machine and a water circulation and purification system. In the design of the facility attentions were paid to safe and reliable remote operation of heavily irradiated materials in hot cell. Main specification of the facility is as follows; Maximum load capacity : 30 kN, Strain rate range : 10{sup -7} s{sup -1} {approx} 10{sup -3} s{sup -1}, Test temperature/pressure : 300degC and 10 MPa in maximum, Water supply rate : 5 l/h, Controllable range of dissolved oxygen (DO) : 0.01 {approx} 32 ppm. Water chemistry is monitored on DO, electric conductivity, and pH. Electrochemical potential and electric current are monitored and controlled on specimen. A preliminary test was carried out on the type 316 stainless steel irradiated up to 8 dpa in Oak Ridge Research Reactor under spectrally tailored condition. As a result of the test an occurrence of IASCC and a good performance of the facility were confirmed. (J.P.N.).}
place = {Japan}
year = {1992}
month = {Jun}
}