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Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
HPR-364, Vol.1 (CD-ROM), 10 Pages, 2005/10
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack propagation and so on, and the present status of in-pile IASCC growth tests using pre-irradiated materials at JMTR.
Kaji, Yoshiyuki
Proceedings of KNS-AESJ Joint Summer School 2005 for Students and Young Researchers, 2, p.221 - 228, 2005/08
For core internals, the main research items are intergranular stress corrosion cracking (IGSCC) of low carbon stainless steel in core shrouds and primary loop recirculation pipes in boiling water reactor (BWR), and irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. This paper describes the current status and typical results of fundamental study for mechanistic understanding of IGSCC and IASCC, development of IASCC evaluation technology for BWR plants based on post-irradiation IASCC test data as a part of METI's national project, in-pile IASCC tests.
Ugachi, Hirokazu; Kaji, Yoshiyuki; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.319 - 325, 2005/00
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this conference, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack initiation, propagation and water chemistry, and the current status of in-pile SCC tests using thermally sensitized materials at JMTR.
Shibata, Taiju; Kikuchi, Takayuki; Miyamoto, Satoshi*; Ogura, Kazutomo*
Nuclear Engineering and Design, 223(2), p.133 - 143, 2003/08
Times Cited Count：1 Percentile：10.38(Nuclear Science & Technology)The High Temperature Engineering Test Reactor (HTTR) can provide very large spaces at high temperatures for irradiation tests. The I-I type irradiation equipment was developed as the first irradiation rig. It will be served for an in-pile creep test on a stainless steel with large standard size specimens. It uses the ambient high temperature of the core for the irradiation temperature control. The target irradiation temperatures are 550 and 600
C with the target temperature deviation of 
3C. In this study, the irradiation temperature changes at transient conditions were analyzed by an FEM code and the temperature controllability of the equipment was examined by a mockup test. The controllability was evaluated with the measured temperature transient data at the core graphite components in the Rise-to-Power tests of the HTTR. The result indicates that the temperature control method of the equipment is effective to keep the irradiation temperature stable in the irradiation test.
Ugachi, Hirokazu; Tsukada, Takashi; Matsui, Yoshinori; Kaji, Yoshiyuki; Tsuji, Hirokazu; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 5 Pages, 2003/04
Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron irradiation, stress and corrosion by high temperature water. It is, therefore, essential to perform in-pile SCC tests, which are material tests under the conditions simulating those of actual LWR operation, in order to clarify the precise mechanism of the phenomenon, though mainly out-of-pile SCC tests for irradiated materials have been carried out in this research field. This paper will describe the current status of the development of several techniques for in-pile SCC initiation tests in JMTR.
Department of JMTR
JAERI-Conf 97-006, 178 Pages, 1997/03
no abstracts in English
; ; ; Watanabe, Hiroyuki; ;
JAERI-M 90-196, 45 Pages, 1990/11
no abstracts in English
; ; ; ;
JAERI-M 86-068, 17 Pages, 1986/04
no abstracts in English
; T.Johnsen*
Journal of Nuclear Science and Technology, 23(8), p.752 - 755, 1986/00
Times Cited Count：1 Percentile：27.82(Nuclear Science & Technology)no abstracts in English
; ; ; ;
JAERI-M 9104, 44 Pages, 1980/10
no abstracts in English
;
JAERI-M 8767, 68 Pages, 1980/03
no abstracts in English
;
JAERI-M 8274, 53 Pages, 1979/06
no abstracts in English
Onizawa, Kunio
no journal, ,
After TEPCO Fukushima-Daiichi NPS accident, new fuels and materials have been developed in the OECD/NEA member countries. The objective of the development is to have new fuels with enhanced accident tolerance, especially for severe accident conditions. For the new fuel development, similar to the conventional ones, many kinds of information are necessary to maintain the safety functions of the fuel. The fuel safety should be confirmed not only for severe accidents, but also for various conditions such as normal operation, design basis accidents, transportation and so on. Preparation of experimental databases with high reliability and analytical tools are also needed to confirm the safety of the new fuels. Therefore, test reactors and hot laboratories play important roles in the acquisition of in-reactor data which are the key information as the technical basis of safety judgement. In this presentation, safety requirements under various plant conditions which are considered in the new fuel development is explained. The importance of experiments using test reactors and laboratories for new fuel development is also explained including examples of JAEA facilities.
