Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 23

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Rationalization and utilization of double-wall vacuum vessel for tokamak fusion facility

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 $$pm$$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.

Journal Articles

Assessments of crack length-water leak correlation on ITER vacuum vessel and inherent safety of Tokamak-type fusion machine

Nakahira, Masataka; Shibui, Masanao*

Nippon 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.

Journal Articles

Structural safety assessment of a tokamak-type fusion facility for a through crack to cause cooling water leakage and plasma disruption

Nakahira, Masataka

Journal of Nuclear Science and Technology, 41(2), p.226 - 234, 2004/02

 Times Cited Count:1 Percentile:89.17(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.

JAEA Reports

Applicability of LBB concept to tokamak-type fusion machine

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.

JAEA Reports

None

JNC-TN1400 2000-012, 250 Pages, 2000/11

JNC-TN1400-2000-012.pdf:10.18MB

no abstracts in English

JAEA Reports

None

JNC-TN1400 2000-010, 70 Pages, 2000/10

JNC-TN1400-2000-010.pdf:2.87MB

no abstracts in English

JAEA Reports

Study on unstable fracture characteristics of light water reactor piping

Kurihara, Ryoichi

JAERI-Research 98-043, 106 Pages, 1998/08

JAERI-Research-98-043.pdf:6.65MB

no abstracts in English

Journal Articles

Influence of wetting effect at the outer surface of the pipe on increase in leak rate; Experimental results and discussion

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

JAEA Reports

Fail-safe first wall for preclusion of little leakage

*; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

JAERI-M 94-074, 16 Pages, 1994/05

JAERI-M-94-074.pdf:0.56MB

no abstracts in English

Journal Articles

Results of reliability test program on light water reactor piping

Shibata, Katsuyuki; Isozaki, Toshikuni; Ueda, Shuzo; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo

Nucl. Eng. Des., 153, p.71 - 86, 1994/00

 Times Cited Count:10 Percentile:31.87

no abstracts in English

Journal Articles

Overview of reliability test program on primary coolant piping of light water reactors

Shibata, Katsuyuki; Isozaki, Toshikuni; Ueda, Shuzo; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo

Nippon Genshiryoku Gakkai-Shi, 35(10), p.923 - 939, 1993/10

 Times Cited Count:1 Percentile:80.29(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Results of piping reliability test program at JAERI

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

Journal Articles

Evaluation of LBB in piping considering multiple fatigue crack growth

Shibata, Katsuyuki

Nippon Kikai Gakkai Rombunshu, A, 58(552), p.1347 - 1352, 1992/08

no abstracts in English

Journal Articles

Overview of piping reliability test program at the Japan Atomic Energy Research Institute

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

Journal Articles

Evaluation of LBB in piping considering multiple fatigue crack growth

Shibata, Katsuyuki

Zairyo Rikigaku Koenkai Koen Rombunshu, Vol.B, p.149 - 151, 1991/00

no abstracts in English

JAEA Reports

Development of leak analysis programs from through-wall crack

; Shibata, Katsuyuki; Isozaki, Toshikuni

JAERI-M 90-050, 106 Pages, 1990/03

JAERI-M-90-050.pdf:2.17MB

no abstracts in English

Journal Articles

Measurement of leak-rate through fatigue-cracks in pipes under four-point bending and BWR conditions

Isozaki, Toshikuni; Shibata, Katsuyuki; ; *

Int. J. Press. Vessels Piping, 43, p.399 - 411, 1990/00

 Times Cited Count:8 Percentile:22.87

no abstracts in English

Journal Articles

Progress and evaluation of test results on JAERIs ductile pipe fracture test program

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

Journal Articles

Investigation and evaluation of ductile fracture behaviors in LWR piping under bending load

Shibata, Katsuyuki; Yasuda, Yuji; Onizawa, Kunio; Miyazono, S.

Structural Mechanics in Reactor Technology,Vol.G, p.461 - 466, 1987/00

no abstracts in English

23 (Records 1-20 displayed on this page)