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Kaji, Yoshiyuki; Miwa, Yukio*; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*
International Journal of Nuclear Energy Science and Engineering, 2(3), p.65 - 71, 2012/09
Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The specimens were irradiated in the core region of the Japan Materials Testing Reactor (JMTR) in simulated BWR water environments at 288 
C from 0.37 to 5.55
10
n/m
(E
1 MeV) (0.62 to 9.2 dpa). The CGRs of base metals in high electrochemical corrosion potential (ECP) condition with 10 
stress intensity factor, K 30 MPam
, increased with increasing neutron fluence until 2 dpa and the CGRs were almost the same from 2 to 10 dpa. We investigated the influence of microstructure on CGR by microstructure observation and local strain measurement around the precipitate. This paper will discuss the relationship between CGR and microstructure, radiation hardening, radiation induced segregation.
Kaji, Yoshiyuki; Miwa, Yukio; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*
Proceedings of 14th International Conference on Environmental degradation of Materials in Nuclear Power Systems (CD-ROM), p.1181 - 1191, 2009/08
The CGR tests of neutron irradiated Type 304 SS were conducted in BWR conditions and the results were compared with those of Type 304L and 316L SS, and following results were obtained. (1) The CGR increase with increasing neutron fluence and the power law of K on the CGR was observed above F2 neutron fluence level (1.4 dpa). The different tendency is observed between Type 304 SS and L-grade SS (Type 304L and 316L SS) with increasing neutron fluence above F3 (4.3 dpa) level. (2) The CGR of Type 304 SS is slightly small as compared with those of Type 304L and 316L SS at the same neutron fluence and shows an increasing tendency above 4 dpa and reaches to 1.010
m/s in 9 dpa. (3) The neutron fluence dependence on uniform elongation is different with Type 304, 304L SS and Type 316L SS, that is, the neutron fluence in which the local deformation like channeling deformation is dominant, is high for Type 316L SS.
Nagashima, Nobuo*; Hayakawa, Masao*; Tsukada, Takashi; Kaji, Yoshiyuki; Miwa, Yukio*; Ando, Masami*; Nakata, Kiyotomo*
Atsuryoku Gijutsu, 47(4), p.236 - 244, 2009/07
In this study, micro-hardness tests and AFM observations were performed on SUS316L low-carbon austenitic stainless steel pre-strained by cold rolling to investigate its deformation behavior. The following results were obtained. Despite the fact that the same plastic strain was applied, post-tensile test AFM showed narrower slip-band spacing in a reduction in area of 30% cold-rolled specimen than the unrolled specimen. Concentrated slip bands were observed near grain boundaries. Micro-hardness exceeding 300 was found to occur frequently in after tensile test specimens with a reduction in area of 30% or more, particularly at grain boundaries. It is suggested that the nonuniformity of deformation at grain boundaries plays an important role of IGSCC crack propagation mechanism of low-carbon austenitic stainless steel.
Chatani, Kazuhiro*; Takakura, Kenichi*; Ando, Masami*; Nakata, Kiyotomo*; Tanaka, Shigeaki*; Ishiyama, Yoshihide*; Hishida, Mamoru*; Kaji, Yoshiyuki
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 9 Pages, 2007/00
Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The CGR tests of 316L and 304L base metals irradiated from 0.516 to 1.0710n/m (E1MeV), and of 316L and 308L weld metals irradiated from 0.523 to 0.54110n/m (E1MeV) were performed using the reversing dc potential drop (DCPD) method under constant load at a few average stress intensity factors (K) and electrochemical corrosion potential (ECP) conditions at 288C in water. CGRs of base metals were increased with increasing neutron fluence. Clear reductions in CGRs of base metals and weld metals were measured with decreasing ECP levels.
Ando, Masami*; Nakata, Kiyotomo*; Ito, Mikiro*; Tanaka, Norihiko*; Koshiishi, Masato*; Obata, Ryoji*; Miwa, Yukio; Kaji, Yoshiyuki; Hayakawa, Masao*
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 16 Pages, 2007/00
Long term SCC growth tests for nuclear grade stainless steel (SUS316(NG)) were conducted in a simulated BWR environment using specimens taken from mock-up PLR piping weld joints to obtain the crack growth rate (CGR) of the hardened heat affected zone due to weld shrinkage around weld, in order to develop the CGR curve which will be used for flaw evaluation. The piping joints were made of forged and extracted materials with several welding techniques. The obtained CGRs were higher than that of solution heat treated material. The CGRs for hardened SUS316(NG) have a correlation with hardness regardless of materials and welding techniques. The CGRs increased with hardness in the range from 210 to 250 Hv. The CGR acceleration mechanism in hardened HAZ of low carbon stainless steel was estimated based on the strain distribution and the AFM image around a SCC crack tip. It was suggested that the interaction of the plastic strain gradient at a crack tip and local strain along GBs.
Nagashima, Nobuo*; Hayakawa, Masao*; Tsukada, Takashi; Kaji, Yoshiyuki; Miwa, Yukio; Ando, Masami*; Nakata, Kiyotomo*
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 15 Pages, 2007/00
Stress corrosion cracking (SCC) was found in shroud and PLR piping made of low-carbon austenitic stainless steels in Japanese BWR plants. The intergranular type (IG) SCC propagated in hardened heat affected zones (HAZ) around welds. Strength behavior and local plastic deformation for a low-carbon austenitic stainless steel 316L, cold-rolled at the reductions in area of 10, 30% at room temperature to simulate the hardened HAZ, were measured by a micro-hardness test machine and observed by atomic force microscopy (AFM), respectively. The tensile deformation at yield point (0.2% plastic strain) had given to the work-hardened 316L to simulate the plastic zone at the crack tip. It is suggested that one of the IGSCC propagation mechanisms for 316L was related with the intergranular strength behavior and local plastic deformation around grain boundaries.
Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Nagashima, Nobuo*; Hayakawa, Masao*; Ando, Masami*; Nakata, Kiyotomo*; Koshiishi, Masato*
no journal, ,
In recent years, incidents of the stress corrosion cracking (SCC) were frequently reported that occurred in the weld part of core shroud and primary loop recirculation (PLR) piping of low carbon stainless steels, and the cause investigation and measure become the present important issue. In this study, to investigate the effect of plastic deformation behavior at the crack, analyses of plastic deformation behavior at the SCC crack tip were performed by electron back-scattering diffraction pattern (EBSP) method. SCC crack propagated along the grain boundary in the 45 degree direction of a fatigue crack, especially random grain boundary. The plastic deformation showed 10 to 20% in the one grain region of the crack. The plastic deformation behavior was different in both of a SCC crack and large plastic deformation was observed on one side of grain. From these results, it was considered that SCC crack propagation behavior was controlled by plastic deformation of very near crack tip.
Kaji, Yoshiyuki; Miwa, Yukio; Ugachi, Hirokazu; Tsukada, Takashi; Shibata, Akira; Kato, Yoshiaki; Arai, Kensaku*; Nakata, Kiyotomo*
no journal, ,
In order to investigate the effect of dose rate on SCC growth behavior, the SCC growth tests were carried out under simulated boiling water reactor (BWR) water conditions using irradiated materials at different dose rate. It was confirmed that the effect of dose rate on SCC growth rate was considered to be small.
Kaji, Yoshiyuki; Miwa, Yukio; Shibata, Akira; Kato, Yoshiaki; Taguchi, Taketoshi; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*
no journal, ,
SCC growth tests have been carried out using type 304 stainless steel that had been pre-irradiated 0.62 to 9.2dpa under BWR simulated high temperature water condition at 288C in the JMTR. This paper describes the investigated results of crack growth rate characteristics from the point of view of microstructure, radiation hardening and radiation induced segregation.
Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi; Kondo, Keietsu; Arai, Kensaku*; Nakata, Kiyotomo*
no journal, ,
In order to investigate the effect of neutron dose rate on the mechanical property and stress corrosion cracking (SCC) growth behavior of type 304 stainless steel, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted after neutron irradiation. The specimens were irradiated up to about 1dpa with different dose rates in the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the high dose rate condition, and a little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. There was no remarkable effect of dose rates on IASCC growth behavior.
