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JAEA Reports

Evaluating techniques and phenomena of Stress Corrosion Cracking (SCC) in Light Water Reactors (LWRs); SCC evaluating techniques for predicting core internal and pipe aging of LWRs, technical data collection (Contract research)

Yamamoto, Masahiro; Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Tsujikawa, Shigeo*; Hattori, Shigeo*; Yoshii, Tsuguyasu*; et al.

JAEA-Review 2012-007, 404 Pages, 2012/03

JAEA-Review-2012-007.pdf:36.72MB

There are many LWRs which have been operated for more than 20 years in Japan and it is expected that technique corresponding to aging plants are necessary established for safety operation in LWRs. A lot of troubles related to SCC are reported and many investigations are concerned with SCC mechanism and technical evaluation. In this paper, those research data were collected as possible widely and reviewed systematically. Current circumstances concerned with SCC in LWRs were reviewed specifically as follows: SCC incidents, SCC evaluation methods for crack initiation and propagation, the investigations concerned with SCC mechanism and monitoring technique for corrosive environment. Influences with reactor types (BWR, PWR), materials (stainless steels, Ni alloys) and SCC evaluating methods (laboratories and actual plants) were summarized as graphs and tables easy to understand in common/difference points concerned with SCC. From these arranged results, future themes were considered and remarked SCC phenomenon was summarized in actual plants. As for SCC evaluations under the accelerate conditions in the laboratory test, it was suggested that a computational prediction and modeling including statistical technique and microscopic analysis in crack initiation were important. Furthermore it was suggested that monitoring techniques predicting SCC initiation and grasping plant circumstance in operation and feasibility in actual plants were important.

Journal Articles

Stress corrosion cracking behavior of type 304 stainless steel irradiated under different neutron dose rates at JMTR

Kaji, Yoshiyuki; Kondo, Keietsu; Aoyagi, Yoshiteru; Kato, Yoshiaki; Taguchi, Taketoshi; Takada, Fumiki; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Takakura, Kenichi*; et al.

Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1203 - 1216, 2011/08

In order to investigate the effect of neutron dose rate on tensile property and irradiation assisted stress corrosion cracking (IASCC) growth behavior, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating the temperature of boiling water reactor (BWR) up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the dose rate, but there was little effect on CGR. Increase of the yield strength of specimens irradiated with the low dose rate condition was caused by the increase of number density of frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, there was little effect on local plastic deformation behavior near crack tip in the crystal plasticity simulation.

Journal Articles

Effects of alloy composition of carbon steel on the flow accelerated corrosion and oxide film properties in neutral water condition

Sato, Tomonori; Ugachi, Hirokazu; Tsukada, Takashi; Uchida, Shunsuke

Proceedings of 14th International Conference on Environmental degradation of Materials in Nuclear Power Systems (CD-ROM), p.985 - 995, 2009/08

In order to improve a theoretical prediction of flow accelerated corrosion (FAC), the individual effects of parameters which are water chemistry, material and flow dynamics should be understood. In this study, to determine the effects of minor alloying elements of carbon steel on FAC, corrosion rate of the specimens of carbon steel with various concentrations of minor alloying elements were measured. And the oxide films were examined by surface analyses techniques, e.g. Raman spectroscopy. Obtained effects of contents of Cr, Ni and Cu in carbon steel on FAC behavior are as follows; (1) FAC was suppressed for the carbon steel with more than 0.03% of Cr content. (2) FAC rate decreased as Ni content increased. (3) The hematite rich oxide was observed for Ni added carbon steel, while only magnetite was observed for Cr added one. (4) The clear effects of Cu on FAC rate was not observed.

JAEA Reports

Development on crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation, 1 (Contract research)

Izumo, Hironobu; Chimi, Yasuhiro; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; et al.

JAEA-Technology 2009-011, 31 Pages, 2009/04

JAEA-Technology-2009-011.pdf:4.38MB

Regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) for austenitic stainless steel of the light water reactor (LWR), a lot of data that concerns the post irradiation evaluation (PIE) is acquired. However, IASCC occurs in LWR condition. Therefore, it is necessary to confirm adequacy of the PIE data comparing the experiment data under the simulated LWR condition. Bigger specimen is needed to acquire the effective data for the destruction dynamics in the study of stress corrosion cracking under neutron irradiation condition. Therefore, development of a new crack growth unit which can load to bigger is necessary to the neutron irradiation test. As a result, a prospect was provided in the unit that could load to specimen by changing load mechanism to the lever type from the linear type. And also, in the development of crack propagation unit, some technical issues were extracted from the discussion of the unit structure adopting the LVDT (Linear Variable Differential Transformer).

Journal Articles

Effects of alloy composition and flow condition on the flow accelerated corrosion in neutral water condition

Sato, Tomonori; Ugachi, Hirokazu; Tsukada, Takashi; Uchida, Shunsuke

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

The major mechanism of Flow accelerated corrosion (FAC) is the dissolution of the protective oxide on carbon steel, which is enhanced by mass transfer and erosion under high flow velocity conditions. In this study, the effects of alloy composition and flow velocity on FAC of carbon steel were evaluated by measuring FAC rate of tube type carbon steel specimens in the neutral water condition at 150$$^{circ}$$C. Obtained results are summarized in follows. (1) High FAC rate was depended upon the v$$^{1.2}$$ in the tube type specimen made of the standard alloy. (2) FAC was mitigated for the carbon steel with more than 0.03% of Cr content. (3) FAC rate decreased as Ni content increased in more than 0.1% of Ni content. (4) The difference in chemical composition of oxide film between Ni added carbon steel and Cr added one was confirmed. The hematite rich oxide was observed for Ni added carbon steel. (5) The effects of Cu on FAC rate was not observed up to 0.1% of Cu content.

Journal Articles

Experiments simulating IGSCC under irradiation in BWR

Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Miwa, Yukio; Nakano, Junichi; Sato, Tomonori; Uchida, Shunsuke

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Since the intergranular stress corrosion cracking (IGSCC) has been a major issue of degradation and failure of structural materials in the boiling water reactors (BWRs), various types of experiments were carried out to investigate IGSCC behavior under the environmental conditions simulated those in BWR. This paper describes a summary of the relevant experimental techniques and the experiences of two types of experiments performed by the authors. For the in-pile SCC experiments, IGSCC crack growth rates were obtained as a function of stress intensity factor in high temperature water. On the other hand, SCC and corrosion tests were performed on un-irradiated specimens in high temperature water by injecting hydrogen peroxide, H$$_{2}$$O$$_{2}$$ to simulate water radiolysis condition. In order to understand IGSCC behavior under irradiation in the reactor core from a mechanistic viewpoint, combinations of various types of experiments are essentially required.

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:54.81(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.

JAEA Reports

Fabrication of irradiation capsule for IASCC irradiation tests, 2; Irradiation capsule for crack propagation test (Joint research)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.

JAEA-Technology 2008-012, 36 Pages, 2008/03

JAEA-Technology-2008-012.pdf:10.09MB

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack propagation test is reported.

JAEA Reports

Fabrication of irradiation capsule for IASCC irradiation tests, 1; Irradiation capsule for crack growth test (Joint research)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.

JAEA-Technology 2008-011, 46 Pages, 2008/03

JAEA-Technology-2008-011.pdf:19.39MB

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported.

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

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

Remote-welding technique for assembling In-Pile IASCC capsule in hot cell

Kawamata, Kazuo; Ishii, Toshimitsu; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Omi, Masao; Shimizu, Michio; Matsui, Yoshinori; Ugachi, Hirokazu; Kaji, Yoshiyuki; Tsukada, Takashi; et al.

JAEA-Conf 2006-003, p.115 - 125, 2006/05

no abstracts in English

Journal Articles

PIE technologies for the study of stress corrosion cracking of reactor structural materials

Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio

JAEA-Conf 2006-003, p.253 - 265, 2006/05

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the 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) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).

Journal Articles

Development of in-pile capsule for IASCC study at JMTR

Matsui, Yoshinori; Hanawa, Satoshi; Ide, Hiroshi; Tobita, Masahiro*; Hosokawa, Jinsaku; Onuma, Yuichi; Kawamata, Kazuo; Kanazawa, Yoshiharu; Iwamatsu, Shigemi; Saito, Junichi; et al.

JAEA-Conf 2006-003, p.105 - 114, 2006/05

Irradiation assisted stress corrosion cracking (IASCC) caused by the simultaneous effects of radiation, stress and high temperature water environment is considered to be one of the critical concerns of in-core structural materials not only for light water reactors (LWRs) but also for water-cooled fusion reactors. In the research field of IASCC, post-irradiation examinations (PIEs) for irradiated materials have been mainly carried out, because there are many difficulties on SCC tests under neutron irradiation environment. Hence we have embarked on a development of the test techniques for performing the in-pile SCC tests. In this paper, we describe the developed several in-pile test techniques and the current status of in-pile SCC tests at Japan Materials Testing Reactor (JMTR).

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

In-core ECP sensor designed for the IASCC experiments at JMTR

Tsukada, Takashi; Miwa, Yukio; Ugachi, Hirokazu; Matsui, Yoshinori; Itabashi, Yukio; Nagata, Nobuaki*; Dozaki, Koji*

Proceedings of International Conference on Water Chemistry of Nuclear Reactor Systems (CD-ROM), 5 Pages, 2004/10

IASCC initiation and propagation tests will be performed on the per-irradiated specimen in the Japan Materials Testing Reactor (JMTR). Since in core, the radiolysis of water causes a generation of various kind of radical species and some oxidizing species such as hydrogen peroxide, the water chemistry in irradiation capsules must be assessed by measurements of the electrochemical corrosion potential (ECP). For the in-core measurement of ECP in JMTR, we fabricated and tested the Fe/Fe$$_{3}$$O$$_{4}$$ type ECP sensor. After the fabrication, the function of each sensor was examined in high temperature water by out-of-core thermal cycling and high temperature holding tests.

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