Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Suzuki, Seiya; Arai, Yoichi; Okamura, Nobuo; Watanabe, Masayuki
Journal of Nuclear Science and Technology, 60(7), p.839 - 848, 2023/07
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)The fuel debris, consisting of nuclear fuel materials and reactor structural materials, generated in the accident of Fukushima Daiichi Nuclear Power Plant can become deteriorated like rocks under the changes of environmental temperature. Although the fuel debris have been cooled by water for 10 years, they are affected by seasonal and/or day-and-night temperature changes. Therefore, in evaluating the aging behavior of the fuel debris, it is essential to consider the changes in environmental temperature. Assuming that the fuel debris are deteriorated, radioactive substances that have recently undergone micronization could be eluted into the cooling water, and such condition may affect defueling methods. We focused on the effect of repeated changes in environmental temperature on the occurrence of cracks, and an accelerated test using simulated fuel debris was carried out. The length of the crack increases with increasing number of heat cycle; therefore, the fuel debris become brittle by stress caused by thermal expansion and contraction. In conclusion, it was confirmed that the mechanical deterioration of the fuel debris is similar to that of rocks or minerals, and it became possible to predict changes in the length of the crack in the simulated fuel debris and environmental model.
Chimi, Yasuhiro; Sato, Kenji*; Kasahara, Shigeki; Umehara, Ryuji*; Hanawa, Satoshi
Proceedings of Contribution of Materials Investigations and Operating Experience to Light Water NPPs' Safety, Performance and Reliability (FONTEVRAUD-9) (Internet), 10 Pages, 2018/09
To investigate the influence of Zinc (Zn) injection on primary water stress corrosion cracking (PWSCC) growth behavior, crack growth tests of 10% cold-worked Alloy 600 were performed in simulated primary water environment of pressurized water reactor (PWR) at 320C with a low-concentration (5-10 ppb) Zn injection under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgHO. As a result of the crack growth tests, DH-dependence of crack growth rate (CGR) showed a similar tendency to the predicted CGR based on the CGR data without Zn injection, indicating almost no effect of a low-concentration Zn injection on the crack growth behavior. Moreover, the microstructural analyses of oxide films formed inside the crack and on the specimen surface were conducted, and the intake of Zn in the oxides was detected on the specimen surface, but not detected inside the crack. This result was considered to be the cause of no Zn injection effect on the crack growth behavior.
Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00
Times Cited Count:2 Percentile:58(Materials Science, Multidisciplinary)In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at 288C on neutron-irradiated 316L stainless steels (SSs) at 12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at 2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.
Lu, K.; Li, Y.
AIMS Materials Science, 4(2), p.439 - 451, 2017/03
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (), stress intensity factor (), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.
Shibata, Katsuyuki*; Onizawa, Kunio; Suzuki, Masahide; Li, Y.*
Nihon Kikai Gakkai M&M 2005 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.299 - 300, 2005/11
no abstracts in English
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.
*; *; O.Ivano*; Nunoya, Yoshihiko; Nakajima, Hideo; Tsuji, Hiroshi
Zairyo, 45(1), p.38 - 42, 1996/01
no abstracts in English
*; *; O.Ivano*; Nunoya, Yoshihiko; Nakajima, Hideo; Tsuji, Hiroshi
Fatigue & Fracture of Engineering Materials & Structures, 18(6), p.671 - 678, 1995/00
Times Cited Count:1 Percentile:19.21(Engineering, Mechanical)no abstracts in English
Nyilas, A.*; Obst, B.*; Nakajima, Hideo
Proceedings of 3rd International Conference on High Nitrogen Steels (HNS-93), p.339 - 344, 1993/00
no abstracts in English
Hirano, Masashi; Kosaka, Atsuo
JAERI-M 92-006, 76 Pages, 1992/02
no abstracts in English
; *; ;
Nihon Kikai Gakkai Rombunshu, A, 52(477), p.1228 - 1231, 1986/00
no abstracts in English
; ; Shiba, Koreyuki
Journal of Nuclear Science and Technology, 21(7), p.528 - 537, 1984/00
Times Cited Count:12 Percentile:75.16(Nuclear Science & Technology)no abstracts in English
*; Nakajima, Hajime; ; ; Kondo, Tatsuo
Corrosion Fatigue; Mechanics, Metallurgy, Electrochemistry and Engineering, p.256 - 286, 1983/00
no abstracts in English
; ; Shindo, Masami; ; ; ; ; ; *; *; et al.
JAERI-M 82-062, 23 Pages, 1982/06
no abstracts in English
Ueda, Shuzo; ;
Int.J.Press.Vessels Piping, 10, p.465 - 480, 1982/00
Times Cited Count:1 Percentile:61.6(Engineering, Multidisciplinary)no abstracts in English
; ; ; ;
Nucl.Eng.Des., 74(2), p.199 - 213, 1982/00
Times Cited Count:1 Percentile:22.52(Nuclear Science & Technology)no abstracts in English
*; Nakajima, Hajime; Kondo, Tatsuo; *
Zairyo, 31(346), p.703 - 709, 1982/00
no abstracts in English
Ueda, Shuzo; ;
JAERI-M 9647, 22 Pages, 1981/08
no abstracts in English
; ; ; *;
JAERI-M 9246, 46 Pages, 1981/01
no abstracts in English