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Wakui, Takashi; Ishii, Hideaki*; Naoe, Takashi; Kogawa, Hiroyuki; Haga, Katsuhiro; Wakai, Eiichi; Takada, Hiroshi; Futakawa, Masatoshi
Materials Transactions, 60(6), p.1026 - 1033, 2019/06
Times Cited Count：3 Percentile：17.96(Materials Science, Multidisciplinary)The mercury target has large size as 1.3
1.32.5 m
. In view of reducing the amount of wastes, we studied the structure so that the fore part could be separated. The flange is required to have high seal performance less than 110
Pa m/s. Invar with low thermal expansion is a candidate. Due to its low stiffness, however, the flange may deform when it is fastened by bolts. Practically invar is reinforced with stainless steel where all interface between them has to be bonded completely with the HIP bonding. In this study, we made specimens at four temperatures and conducted tensile tests. The specimen bonded at 973 K had little diffusion layer, and so fractured at the interface. The tensile strength reduced with increasing the temperature, and the reduced amount was about 10% at 1473 K. The analyzed residual stresses near the interface increased by 50% at maximum. Then, we concluded that the optimum temperature was 1173 K.
Onizawa, Takashi; Hashidate, Ryuta
Mechanical Engineering Journal (Internet), 6(1), p.18-00477_1 - 18-00477_15, 2019/02
Aiming at enhancing its economic competitiveness and reducing radioactive waste, JAEA has proposed an attractive plant concept and made great efforts to demonstrate the applicability of some innovative technologies to the plant. One of the most practical means is to extend the design life to 60 years. Accordingly, the material strength standards set by JSME have to be extended from 300,000 to 500,000 hours but this extension requires more precise estimation of creep rupture strength and creep strain of the materials in the long term. This paper describes the development of creep property equations of 316FR stainless steel and Mod.9Cr-1Mo steel considering changes in creep mechanisms at high temperatures in the long term based on evaluations of long-term creep properties of the materials. The creep property equations developed in this study will provide more precise estimation of the creep properties in the long term than the present creep property equations of JSME.
Ishio, Kotaro*; Nakajima, Hideo
Tetsu To Hagane, 92(2), p.90 - 96, 2006/02
no abstracts in English
Ishio, Kotaro*; Hamada, Kazuya; Nakajima, Hideo
Tetsu To Hagane, 92(1), p.30 - 35, 2006/01
no abstracts in English
Nakamura, Hiroo; Takemura, Morio*; Yamauchi, Michinori*; Fischer, U.*; Ida, Mizuho*; Mori, Seiji*; Nishitani, Takeo; Simakov, S.*; Sugimoto, Masayoshi
Fusion Engineering and Design, 75-79, p.1169 - 1172, 2005/11
Times Cited Count：6 Percentile：40.47(Nuclear Science & Technology)In the IFMIF, activated erosion/corrosion materials of Li target back wall deposits on a surface of the Li loop. Therefore, accessibility during maintenance of the Li loop pipings will depend on the activation level of the deposition materials. This paper evaluates effect of target activation on the accessibility of the Li loop pipings. Activation level is calculated by the ACT-4 code. High energy cross section above 15 MeV is introduced using IEAF-2001 data. In this calculation, target material is stainless steel 316. Area of the erosion/corrosion in the back wall is 100 cm
. The erosion/corrosion rate is 1 micron/y. Dose rate around the Li loop after one year IFMIF operation is evaluated assuming 1% deposition of the erosion/corrosion materials and uniform deposition on surface area of 33 m. Permissible level for hands-on maintenance is 10 microSv/hr. As the results, after 1 week from shutdown, close maintenance work 8 cm to the Li loop is possible. Also, after 1 month, hands-on maintenance becomes possible.
Nishi, Hiroshi
Journal of Nuclear Materials, 329-333(Part2), p.1567 - 1570, 2004/08
Times Cited Count：9 Percentile：52.68(Materials Science, Multidisciplinary)Elastic-plastic finite element analysis was performed for low cycle fatigue behavior of stainless steel/alumina-dispersion-strengthened copper (DS Cu) joint in order to investigate the fatigue life and the fracture behavior of the joint. As the results, a strain concentration was occurred near the interface of DS Cu for small strain range, however, in the DS Cu for large strain range. The fatigue life and fracture point were evaluated taking account for the strain concentration. The fatigue life and fracture point were consistent with those of the low cycle fatigue test.
Nakano, Junichi; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Kita, Satoshi; Nemoto, Yoshiyuki; Tsuji, Hirokazu; Jitsukawa, Shiro
Journal of Nuclear Materials, 307-311(Part2), p.1568 - 1572, 2002/12
Times Cited Count：12 Percentile：60.82(Materials Science, Multidisciplinary)Type 316LN stainless steel of the international thermonuclear experimental reactor (ITER) Grade (316LN-IG SS) is being considered for the first wall/ blanket component. Hot isostatic pressing (HIP) technique is expected for the fabrication of module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316LN-IG SS, tensile tests in vacuum and slow strain rate tests (SSRT) in high temperature water were performed. Specimen with the HIPed joint shows no deterioration of the tensile strength and susceptibility to SCC in oxygenated water. Thermally sensitized specimen with the HIPed joint was low susceptible to SCC in creviced environment. It is concluded that the strength at joint location is as high as that at the base alloy and the joint interface appears integrity.
Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Kita, Satoshi; Yonekawa, Minoru; Nakano, Junichi; Tsuji, Hirokazu; Nakajima, Hajime
Journal of Nuclear Materials, 307-311(Part1), p.331 - 334, 2002/12
Times Cited Count：5 Percentile：34.58(Materials Science, Multidisciplinary)Irradiation assisted stress corrosion cracking (IASCC) caused by simultaneous effects of neutron irradiation and high temperature water environments has been pointed out as one of the major concerns of in-core structural materials not only for the light water reactors (LWRs) but also for the water-cooled fusion reactor. It is necessary to evaluate precisely stress condition under irradiation environment, because stress is one of key factors on IASCC. Stress relaxation of tensile type specimens under fast neutron irradiation at 288
C has been studied for type 316L stainless steel in Japan Materials Testing Reactor (JMTR). This paper describes the in-pile and out-of-pile stress-relaxation test results of tensile type specimens for type 316L stainless steel as compared with the literature data by Foster, which were mainly obtained by bent beam specimens. Moreover these experimental results were compared with the analytical ones by using Nakagawa's model.
Hirabayashi, Takakuni; Saeki, Masakatsu; Tachikawa, Enzo
Journal of Nuclear Materials, 127, p.187 - 192, 1985/00
Times Cited Count：15 Percentile：84.25(Materials Science, Multidisciplinary)no abstracts in English
Shiraishi, K.; ;
Journal of Nuclear Materials, 119(2-3), p.268 - 277, 1985/00
Times Cited Count：8 Percentile：68.41(Materials Science, Multidisciplinary)no abstracts in English
C; ; ;
Journal of Nuclear Materials, 116(2-3), p.178 - 183, 1983/00
Times Cited Count：7 Percentile：64.85(Materials Science, Multidisciplinary)no abstracts in English
Shiraishi, K.; Aruga, T.; Katano, Y.
Journal of Nuclear Materials, 103-104, p.1053 - 1058, 1981/00
no abstracts in English
;
Nucl.Eng.Des., 55(3), p.375 - 387, 1979/00
Times Cited Count：2no abstracts in English
Chimi, Yasuhiro; Kasahara, Shigeki; Hata, Kuniki; Nishiyama, Yutaka; Seto, Hitoshi*; Chatani, Kazuhiro*; Kitsunai, Yuji*; Koshiishi, Masato*
no journal, ,
In order to investigate effects of environmental mitigation and water radiolysis caused by 
-rays from radioactive material on irradiation-assisted stress corrosion cracking (IASCC) growth behavior for highly irradiated material, crack growth tests in simulated BWR water conditions (at 563 K) are performed. The specimens made of 316L stainless steels are irradiated with neutrons up to 
12 dpa in the Japan Materials Testing Reactor (JMTR). One of the specimens is annealed at 973 K for 1 hour to show almost recovered mechanical and micro-chemical properties corresponding to the unirradiated material. For low electrochemical corrosion potential (ECP) condition, the crack growth rate (CGR) is suppressed by about one order of magnitude in high stress intensity factor (K) condition. This result indicates that environmental mitigation for crack growth can be found even under severe conditions on material and stress factors. The effects of water radiolysis on the CGRs are discussed.
Chimi, Yasuhiro; Kasahara, Shigeki*; Nishiyama, Yutaka; Seto, Hitoshi*; Chatani, Kazuhiro*; Kitsunai, Yuji*; Koshiishi, Masato*
no journal, ,
In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests in simulated BWR water conditions (at 563 K) were performed using neutron-irradiated specimens made of 316L stainless steels, and the oxide film properties and locally deformed structures near the crack tip have been investigated by transmission electron microscopy (TEM). When electrochemical corrosion potential (ECP) of the materials was lowered by deaeration and hydrogen injection into feed water, apparent suppression of oxidation inside the cracks was observed as well as suppression of the crack growth rate (CGR). In the presentation, the TEM results of the locally deformed structures along the cracks are also reported, and the relation among the CGR, oxide film properties, and locally deformed structures is discussed.
Chimi, Yasuhiro; Kasahara, Shigeki; Nishiyama, Yutaka; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*
no journal, ,
In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests using compact tension (CT) specimens made of neutron-irradiated 316L stainless steels (SSs) were performed in simulated BWR environments (at 288C). Moreover, microstructures of deformed areas were observed by transmission electron microscope (TEM) after straining tensile specimens made of neutron-irradiated 316L SSs. As a result, for lower neutron dose than 
1.9 dpa, the crack growth rates (CGRs) show effective environmental mitigation and the deformed structures show tangling of dislocations. On the other hand, for higher neutron dose than 
2.7 dpa, the CGRs show small environmental mitigation and the deformed structures consist mainly of dislocation channels. From the relationship between CGRs and deformed structures, mechanisms on IASCC growth will be discussed.
Irisawa, Eriko; Kato, Chiaki
no journal, ,
As a corrosion evaluation of stainless steel in liquid lead-bismuth alloys used as coolants for next-generation nuclear power systems, the relationship between the corrosion behaviors and the oxygen concentrations in LBE was investigated for the extra-pure stainless steel SUS316EHP, which exhibited high intergranular corrosion resistance in aqueous solution environments. Although intergranular oxidation was suppressed at high oxygen concentration, significantly faster dissolution corrosion was observed at low oxygen concentration compared to the corrosion behavior of SUS316L. The influence of chemical composition on the corrosion behavior at grain boundaries was also discussed based on microstructural observations.
