Report of Examination of the Sample from Core Shroud (2F2-H3) at Fukushima Dai-ni Power Station Unit-2 (Contract research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi; Nakajima, Hajime*; Shibata, Katsuyuki; Tsukada, Takashi; Suzuki, Masahide; Kiuchi, Kiyoshi; Kaji, Yoshiyuki; Kikuchi, Masahiko; Ueno, Fumiyoshi; Nakano, Junichi; et al.

JAERI-Tech 2004-015, 114 Pages, 2004/03

JAERI-Tech-2004-015.pdf:38.06MB

The Tokyo Electric Power Company (TEPCO) visually inspected the weld joint of core shroud at Fukushima Dai-ni Nuclear Power Station Unit-2 by a direction of the Nuclear and Industrial Agency, cracks were observed at outer side of the ring weld joint (H3) between a core shroud middle trunk and a middle ring. TEPCO has conducted a material examination with Nippon Nuclear Fuel Development Co. Ltd. (NFD) on the specimen including cracks sampled from the core shroud. The present examination has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage. Based on results of the present examination, the probable presence of tensile residual stress by welding process and dissolved oxygen contents in the cooling water, it was shown that the cracks were considered to be stress corrosion cracking (SCC). However, the cause of the cracks needs more consideration on the way of shroud construction.

Evaluation of in-pile and out-of-pile stress relaxation in 316L stainless steel under uniaxial loading

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

https://doi.org/10.1016/S0022-3115(02)01204-7

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 288C 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.

New in-pile water loop facility for IASCC studies at JMTR

Tsukada, Takashi; Komori, Yoshihiro; Tsuji, Hirokazu; Nakajima, Hajime; Ito, Haruhiko

Proceedings of International Conference on Water Chemistry in Nuclear Reactor Systems 2002 (CD-ROM), 5 Pages, 2002/00

Irradiation assisted stress corrosion cracking (IASCC) is caused by the synergistic effects of neutron and gamma radiation, residual and applied stresses and high temperature water environment on the structural materials of vessel internals. IASCC has been studied since the beginning of the 1980s and the phenomenological knowledge on IASCC is accrued extensively. However, mainly due to the experimental difficulties, data for the mechanistic understanding and prediction of failures of the specific in-vessel components are still insufficient and further well-controlled experiments are needed [1]. In recent years, efforts to perform the in-pile materials test for IASCC study have been made at some research reactors [2-4]. At JAERI, a high temperature water loop facility was designed to install at the Japan Materials Testing Reactor (JMTR) to carry out the in-core IASCC testing. This report describes an overview of design and specification of the loop facility.

Status of JAERI material performance database (JMPD) and analysis of irradiation assistd stress corrosion cracking (IASCC) data

Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Science and Technology, 37(11), p.949 - 958, 2000/11

https://doi.org/10.3327/jnst.37.949

no abstracts in English

Evaluation of ferromagnetic materials for induction linacs

Hashimoto, Dai; Morimoto, Iwao; Zheng, X.; Maebara, Sunao; Nakajima, Mitsuo*; Horioka, Kazuhiko*; Kono, Toshiyuki*; Shiho, Makoto

JAERI-Research 2000-018, p.66 - 0, 2000/03

JAERI-Research-2000-018.pdf:2.19MB

no abstracts in English

Study on development of filler metal for Ni-Cr-W superalloy (Joint research)

Saito, T.*; ; Takatsu, T.*; Tsuji, Hirokazu; Shindo, Masami; Nakajima, Hajime

JAERI-Research 99-036, 99 Pages, 1999/05

JAERI-Research-99-036.pdf:29.53MB

no abstracts in English

Microstructures of type 316 model alloys neutron-irradiated at 513 K to 1 dpa

Miwa, Yukio; Tsukada, Takashi; Tsuji, Hirokazu; Nakajima, Hajime

Journal of Nuclear Materials, 271-272, p.316 - 320, 1999/00

https://doi.org/10.1016/S0022-3115(98)00717-X

no abstracts in English

Material research in Tokai Research Establishment, Japan Atomic Energy Research Institute

Katsuta, Hiroji; Nakajima, Hajime

Materia, 37(11), P. 960, 1998/11

https://ci.nii.ac.jp/naid/10002546637

no abstracts in English

Creep rupture properties of trial welded joints of a Ni-Cr-W superalloy in air environment

; Saito, T.*; Tsuji, Hirokazu; Takatsu, T.*; Shindo, Masami; Nakajima, Hajime

JAERI-Research 97-032, 20 Pages, 1997/05

JAERI-Research-97-032.pdf:1.78MB

no abstracts in English

Creep rupture properties of a Ni-Cr-W superalloy in air environment

; Tsuji, Hirokazu; Shindo, Masami; Nakajima, Hajime

Journal of Nuclear Materials, 246(2-3), p.196 - 205, 1997/00

https://doi.org/10.1016/S0022-3115(97)00106-2

no abstracts in English