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Machida, Hideo*; Koizumi, Yu*; Wakai, Takashi; Takahashi, Koji*
Nihon Kikai Gakkai M&M 2019 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.OS1307_1 - OS1307_5, 2019/11
This paper describes the fracture test and fracture analysis of a pipe under displacement control load. In order to grasp the fracture behavior of the circumferential through-wall cracked pipe, which is important in evaluating the feasibility of leak before break (LBB) in sodium cooled reactor piping, a fracture test in case of a circumferential throughwall crack in the weld line between an elbow and a straight pipe was carried out. From this test, it was found that no pipe fracture occurs in the displacement control loading condition even if a large circumferential through-wall crack (180
) was assumed. The fracture analysis of the pipe was carried out using Gurson's parameters set based on the tensile test results of the tested pipe material. The analytic results agree well with the test results, and it was found that it will be possible to predict the fracture behavior of sodium cooled reactor piping.
Nakahira, Masataka
JAERI-Research 2005-030, 182 Pages, 2005/09
JAERI-Research-2005-030.pdf:12.57MB
It is difficult for Vacuum Vessel (VV) of ITER to apply a non-destructive in-service inspection (ISI) and then new safety concept is needed. Present fabrication standards are not applicable to the VV, because the access is limited to the backside of closure weld of double wall. Fabrication tolerance of VV is 
5mm even the structure is huge as high as 10m. This accuracy requires a rational method on the estimation of welding deformation. In this report, an inherent safety feature of the tokamak is proved closing up a special characteristic of termination of fusion reaction due to tiny water leak. A rational concept not to require ISI without sacrificing safety is shown based on this result. A partial penetration T-welded joint is proposed to establish a rational fabrication method of double wall. Strength and susceptibility to crevice corrosion is evaluated for this joint and feasibility is confirmed. A rational method of estimation of welding deformation for large and complex structure is proposed and the efficiency is shown by comparing analysis experimental results of full-scale test.
Nakahira, Masataka; Shibui, Masanao*
Nihon Kikai Gakkai Dai-9-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, No.04-2, p.267 - 272, 2004/06
A small water leak can cause a plasma disruption in a tokamak-type fusion machine. This plasma disruption will induce electromagnetic (EM) force acting in the vacuum vessel that is a physical barrier of tritium and activated dust. If the VV can sustain an unstable fracture by the EM force, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma in ITER is evaluated as about 2 mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated.
Nakahira, Masataka
Journal of Nuclear Science and Technology, 41(2), p.226 - 234, 2004/02
Times Cited Count：1 Percentile：10.03(Nuclear Science & Technology)A tokamak-type fusion machine is said to have inherent safety associated with plasma shutdown. A small leak of water can terminate the plasma safely and can cause a plasma disruption which will induce electromagnetic(EM) forces in the vacuum vessel (VV). From a radiological safety view point, the VV forms the physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak is proposed, with verification by experiment. Based on the analyses, the critical crack length to terminate plasma is evaluated as about 2 mm, and the critical crack length for unstable fracture is obtained as about 400 mm. It is therefore concluded that EM forces induced by small leak to terminate plasma will not cause the unstable fracture of VV, and then the inherent safety is demonstrated.
Nakahira, Masataka
JAERI-Tech 2003-087, 28 Pages, 2003/12
JAERI-Tech-2003-087.pdf:1.74MB
A tokamak-type fusion machine has been characterized as having inherent plasma shutdown safety. An extremely small leakage of cooling water will cause a plasma disruption. This plasma disruption will induce electromagnetic forces (EM forces) acting in the vacuum vessel (VV) which forms the physical barrier enclosing tritium and activated dust. If the VV has the possibility of sustaining an unstable fracture from a penetrating crack caused by EM forces, the structural safety will be assured and the inherent safety will be demonstrated. This paper analytically assures the Leak-Before-Break (LBB) concept as applied to the VV and is based on experimental leak rate data of a through crack having a very small opening. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2 mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is therefore concluded that EM forces induced by small leak to terminate plasma will not cause the unstable fracture of VV, and then the inherent safety is demonstrated.
JNC TN1400 2000-012, 250 Pages, 2000/11
JNC-TN1400-2000-012.pdf:10.18MB
no abstracts in English
Kurihara, Ryoichi
JAERI-Research 98-043, 106 Pages, 1998/08
JAERI-Research-98-043.pdf:6.65MB
no abstracts in English
Isozaki, Toshikuni; Shibata, Katsuyuki
LBB95: Specialist Meeting on Leak Before Break in Reactor Piping and Vessels, 0, 10 Pages, 1995/00
no abstracts in English
*; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki
JAERI-M 94-074, 16 Pages, 1994/05
no abstracts in English
Shibata, Katsuyuki; Isozaki, Toshikuni; Ueda, Shuzo; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo
Nucl. Eng. Des., 153, p.71 - 86, 1994/00
Times Cited Count：11 Percentile：68.99(Nuclear Science & Technology)no abstracts in English
Shibata, Katsuyuki; Isozaki, Toshikuni; Ueda, Shuzo; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo
Nihon Genshiryoku Gakkai-Shi, 35(10), p.923 - 939, 1993/10
Times Cited Count：1 Percentile：18.76(Nuclear Science & Technology)no abstracts in English
JAERI-M 93-076, 438 Pages, 1993/05
no abstracts in English
Shibata, Katsuyuki; Isozaki, Toshikuni; *; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo
Proc. of 6th German-Japanese Seminar on Structural Strength and NDE Problems in Nuclear Engineering, 19 Pages, 1993/00
no abstracts in English
Shibata, Katsuyuki
Nihon Kikai Gakkai Rombunshu, A, 58(552), p.1347 - 1352, 1992/08
no abstracts in English
Isozaki, Toshikuni; Shibata, Katsuyuki; *; Ueda, Shuzo; Kurihara, Ryoichi
Transactions of the 11th Int. Conf. on Structural Mechanics in Reactor Technology, Vol. SDO, p.401 - 412, 1991/08
no abstracts in English
Shibata, Katsuyuki
Zairyo Rikigaku Koenkai Koen Rombunshu, Vol.B, p.149 - 151, 1991/00
no abstracts in English
; Shibata, Katsuyuki; Isozaki, Toshikuni
JAERI-M 90-050, 106 Pages, 1990/03
no abstracts in English
Isozaki, Toshikuni; Shibata, Katsuyuki; ; *
Int. J. Press. Vessels Piping, 43, p.399 - 411, 1990/00
Times Cited Count：8 Percentile：75.98(Engineering, Multidisciplinary)no abstracts in English
Shibata, Katsuyuki; *; Onizawa, Kunio; Miyazono, S.
Proc. on the 4th Japanese-German Joint Seminar on Structural Strength and NDE Programs in Nucl. Eng., p.347 - 364, 1988/00
no abstracts in English
