Sato, Tomonori; Uchida, Shunsuke; Tsukada, Takashi; Sato, Yoshiyuki*; Mkel, K.*
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 13 Pages, 2007/08
measurements of electric resistance and multilateral surface analyses of oxide films on stainless steel specimens exposed to high temperature pure water containing HO and O were carried out to evaluate properties of these oxide films and their effects on ECP. During measurement of electric resistance, the potential of the specimens was kept at a natural potential with regard to the reference electrode to avoid change in oxide properties. The results can be summarized as follows. (1) Major parameters of the electric resistance measurement apparatus could be optimized so as to improve the quality of obtained data. (2) The specific electric resistance of oxide film formed under 100 ppb HO exposure at 288 C was about 3 Mcm, 150 times higher than under 200 ppb O exposure. (3) It was confirmed that the effect of potential drop caused by the electric resistance of the oxide film and corrosion current density on ECP was small.
Uchida, Shunsuke; Sato, Tomonori; Tsukada, Takashi; Sato, Yoshiyuki*; Wada, Yoichi*
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 15 Pages, 2007/08
measurements of corrosion behaviors of stainless steel specimens exposed to HO and O in high temperature water have been carried out. Oxide films on the specimens have been also examined with multilateral surface analyses. The major differences between environments containing O and HO are summarized as follows. (1) The oxide films formed under HO environment consists mainly of hematite with high electric resistance, while that under O environment consists of magnetite. (2) Differences in ECP of stainless steel exposed to HO and O are caused by differences in chemical form of oxide and oxidation of HO at the surface. (3) HO caused higher ECP and lower corrosion current density than O with the same oxidant concentrations, resulting from the effects of the thin oxide films with high electric resistance. (4) Lower corrosion current density results in lower IGSCC crack growth rate.
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 110n/m 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/ 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.
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.
Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 11 Pages, 2007/00
The effect of residual stress on corrosion behavior in type 316L austenitic stainless steel was examined by ion irradiated specimens. Ion irradiation was performed on specimens both undeformed and deformed plastically by bending constrait at 330C to average displacement damage from 1 to 45dpa. It was observed in EPR testing that deformed specimens showed higher corrosion resistance than undeformed specimens. Three-dimensional atom probe analysis was conducted on irradiated specimens. It was found that the enrichment of Ni, Si and the depletion of Cr at dislocations, and the degree of segregation was greater in undeformed specimen than in deformed specimen. It could be suggested that radiation induced segregation behavior of solute atoms as a consequence of diffusion and annihilation of irradiation defects at sink is affected by residual stress, and this also might affect the corrosion resistance.
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 ( 1), twin GBs ( 3) and coincidence site lattice (CSL) GBs ( 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.
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.