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Journal Articles

In-core SCC growth behavior of type 304 stainless steel in BWR simulated high-temperature water at JMTR

Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi; Nakano, Junichi; Matsui, Yoshinori; Kawamata, Kazuo; Shibata, Akira; Omi, Masao; Nagata, Nobuaki*; Dozaki, Koji*; et al.

Journal of Nuclear Science and Technology, 45(8), p.725 - 734, 2008/08

 Times Cited Count:6 Percentile:55.76(Nuclear Science & Technology)

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1$$times$$10$$^{25}$$n/m$$^{2}$$ in pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/$$gamma$$ radiation and stress/water environment on IASCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents environments under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests will be discussed and compared with the result obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC.

Journal Articles

Development of in-pile SCC test technique and crack initiation behavior using pre-irradiated austenitic stainless steel at JMTR

Ugachi, Hirokazu; Kaji, Yoshiyuki; Matsui, Yoshinori; Endo, Shinya; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04

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). It is considered that the reproduced IASCC by PIEs must be carefully distinguished from the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. Hence, we have embarked on a development of the test technique for the in-pile IASCC testing. We adopted the uniaxial constant load (UCL) tensile test method with small tensile specimens for in-pile SCC initiation test, and tried to evaluate the crack initiation behavior as the detection of specimen rupture or detailed observation of surface of loaded specimens. As a result of this study, it was inferred that an acceleration effect of in-pile environment for SCC initiation behavior was not observed under the test condition of this study.

Journal Articles

Comparison of SCC growth rate between in-core and EX-core tests in BWR simulated high temperature water

Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi; Matsui, Yoshinori; Omi, Masao; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 12 Pages, 2007/00

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1$$times$$10$$^{25}$$n/m$$^{2}$$ in pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/$$gamma$$ radiation and stress/water environment on SCC growth rate, we performed post irradiation examinations (PIEs) in the several dissolved oxygen contents or hydrogen peroxide added environments under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC.

Journal Articles

Grain boundary character of cracks observed in IASCC and IGSCC

Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kato, Yoshiaki; Tomita, Takeshi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 14 Pages, 2007/00

Grain boundary (GB) character of cracks observed in irradiation assisted stress corrosion cracking (IASCC) and in intergranular stress corrosion cracking (IGSCC) was examined using the orientation imaging microscope (OIM). IASCC were produced by constant load tests with 1/4T-CT specimens for pre-irradiated (1.8 dpa at 546 K) type 304 stainless steel. The tests for pre-irradiated specimens were performed by the post irradiation SCC test or the in-reactor SCC test at the Japan Materials Testing Reactor. In all specimens, cracks propagated mainly along random grain boundaries (GBs), and small amount of cracks propagated along low angle GBs ($$Sigma$$ 1), twin GBs ($$Sigma$$ 3) and coincidence site lattice (CSL) GBs ($$Sigma$$ 5-27). Fraction of the GB character was compared with the author's previous studies in which the fraction of IGSCC in thermally-sensitized type 304 stainless steel and unirradiated type 316L stainless steel were measured on CT specimens and a BWR shroud sample. The relationship between SCC behavior and the GB character was discussed. It was considered that the difference of the fraction of GB character between IASCC and IGSCC related to the deformation mode of irradiated stainless steel such as dislocation channelling.

Journal Articles

Results of in-pile SCC growth tests in high temperature water at JMTR

Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of International Conference on Water Chemistry of Nuclear Reactor Systems 2006 (CD-ROM), 5 Pages, 2006/10

Irradiation assisted stress corrosion cracking (IASCC) has been recognized as the aging issue of core-internal materials of the light water reactors (LWRs). The synergistic effect of neutron/$$gamma$$ radiation, stress and high temperature water on the materials in the reactor core is significant to understand IASCC behavior. Therefore, the in-pile IASCC testing is one of the key experiments to investigate IASCC mechanism, and also to assess the reliability of the PIE data. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing. Using the loop facility, in-pile IASCC growth tests have been successfully carried out in the irradiation capsule under simulated BWR condition. The results showed that the effect of synergy of neutron/$$gamma$$ radiation and stress/water environment on SCC growth rate was considered to be small within the present test conditions.

Journal Articles

In-pile SCC growth behavior of type 304 stainless steel in high temperature water at JMTR

Kaji, Yoshiyuki; Ugachi, Hirokazu; Tsukada, Takashi; Matsui, Yoshinori; Omi, Masao; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 14th International Conference on Nuclear Engineering (ICONE-14) (CD-ROM), 7 Pages, 2006/07

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 general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). 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 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-pile IASCC growth tests have been successfully carried out using pre-irradiated type 304 stainless steel at JMTR. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC.

Journal Articles

Present status of in-pile IASCC growth tests at JMTR

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.

Journal Articles

In-pile SCC initiation and growth testing at JMTR

Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of Symposium on Water Chemistry and Corrosion of Nuclear Power Plants in Asia, 2005 (CD-ROM), 6 Pages, 2005/10

Irradiation assisted stress corrosion cracking (IASCC) is one of the significant concerns for the in-vessel stainless steel components of the aged light water reactors (LWRs). In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. Therefore, to confirm the effect of synergy, we have started to develop the test technique to carry out the in-pile IASCC tests at JMTR, Japan Materials Testing Reactor. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack initiation/growth and a result of mock-up in-pile SCC tests using thermally sensitized specimens.

Journal Articles

Development of test techniques for in-pile SCC initiation and growth tests and the current status of in-pile testing at JMTR

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.

Journal Articles

Program of in-pile IASCC testing under the simulated actual plant condition; Thermohydraulic design study of saturated temperature capsule for IASCC irradiation test

Ide, Hiroshi; Matsui, Yoshinori; Nagao, Yoshiharu; Komori, Yoshihiro; Itabashi, Yukio; Tsuji, Hirokazu; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04

The advanced water chemistry controlled irradiation research device has been developed in JAERI to perform irradiation tests for research on IASCC. The irradiation device consists of the SATCAP (Saturated Temperature Capsule) inserted into the JMTR core and the water control unit installed out-of-core. Regarding the SATCAP, thermohydraulic design of the SATCAP was performed aiming at controlling the specimen temperature with high accuracy and increasing water flow velocity on the specimen surface to improve the controllability of water chemistry. As a result of irradiation test using the new type SATCAP, each specimen temperature and water chemistry were able to be controlled as designed.

Oral presentation

In-pile IASCC study in Japan Materials Testing Reactor, 2

Kaji, Yoshiyuki; Ugachi, Hirokazu; Matsui, Yoshinori; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

no journal, , 

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns of in-core structural materials for light water reactors. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In this study, we conducted in-pile SCC growth tests in pure water simulated boiling water reactor (BWR) coolant condition and discussed comparing with the results obtained by PIEs from a viewpoint of the synergistic effects on IASCC.

Oral presentation

In-pile IASCC study in Japan Materials Testing Reactor, 4; Water chemistry measurement by ECP under high neutron flux

Miwa, Yukio; Hanawa, Satoshi; Matsui, Yoshinori; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

no journal, , 

no abstracts in English

Oral presentation

In-pile IASCC study in Japan Materials Testing Reactor, 3; In-pile SCC initiation test

Ugachi, Hirokazu; Kaji, Yoshiyuki; Matsui, Yoshinori; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

no journal, , 

no abstracts in English

Oral presentation

Electrochemical potential measurement of stainless steel in BWR water environment under high radiation field at JMTR

Miwa, Yukio; Tsukada, Takashi; Sato, Tomonori; Hanawa, Satoshi; Matsui, Yoshinori; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*

no journal, , 

no abstracts in English

Oral presentation

Investigation of crack initiation behavior using pre-irradiated austenitic stainless steel at JMTR

Tsukada, Takashi; Ugachi, Hirokazu; Kaji, Yoshiyuki; Miwa, Yukio; Nakano, Junichi; Matsui, Yoshinori; Endo, Shinya; Kato, Yoshiaki; Nagata, Nobuaki*; Dozaki, Koji*; et al.

no journal, , 

Irradiation Assisted Stress Corrosion Cracking (IASCC) has been regarded in recent years as the matter that is related to the reliability for core components of LWRs. The in-pile IASCC initiation tests were performed using of uniaxial constant load specimens pre-irradiated to about 5$$times$$10$$^{24}$$n/m$$^{2}$$ and 1$$times$$10$$^{25}$$n/m$$^{2}$$ at JMTR under the simulated BWR core condition. From the results of in-pile and out-of-pile tests of the irradiated specimens, we confirmed that any remarkable acceleration effect of synergy of neutron/$$gamma$$ radiation and stress/water environment on SCC initiation results was not observed under the test condition of this study. On the surface of in-pile test specimen, a micro crack was observed on the surface of specimen, and it was considered to be an initiation of SCC besed on the results of out-of-pile experiments.

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