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Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai
Nuclear Engineering and Design, 305, p.270 - 276, 2016/08
Times Cited Count：3 Percentile：28.28(Nuclear Science & Technology)In our previous study, we proposed a new process for determining the in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. It was shown that both components had sufficient reliability even assuming unrealistic severe conditions. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components. The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.
Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
In our previous study, a new process for determination of in-service inspection (ISI) requirements was proposed on the basis of the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, ISI requirements for a reactor guard vessel and a core support structure of the prototype sodium-cooled fast breeder reactor in Japan, Monju, were investigated according to the proposed process. The proposed process is expected to contribute to realize effective and rational ISI by properly taking into account plant-specific features.
Takaya, Shigeru; Asayama, Tai; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Nakai, Satoru; Morishita, Masaki
Journal of Nuclear Engineering and Radiation Science, 1(1), p.011004_1 - 011004_9, 2015/01
A new process for determination of inservice inspection (ISI) requirements was proposed based on the System Based Code concept to realize effective and rational ISI by properly taking into account plant specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.
Watanabe, Daigo*; Chuman, Yasuharu*; Asayama, Tai; Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 7 Pages, 2013/07
Limit state design was newly developed in order to apply the margin exchange which is one of the innovative concepts of the System Based Code (SBC). It was shown that limit state design method is applicable to plant design instead of current design criteria. In this report, working example of a reactor vessel of a Fast Reactor subject to thermal load is conducted to demonstrate this concept. As the result allowable stress was increased by changing the acceptance criteria from current design criteria to limit state design criteria.
Takaya, Shigeru; Okajima, Satoshi; Kurisaka, Kenichi; Asayama, Tai; Machida, Hideo*; Kamishima, Yoshio*
Journal of Power and Energy Systems (Internet), 5(1), p.60 - 68, 2011/01
Takaya, Shigeru; Okajima, Satoshi; Kurisaka, Kenichi; Asayama, Tai; Machida, Hideo*; Kamishima, Yoshio*
no journal, ,
The System Based Code (SBC) concept has been proposed to achieve compatibility in matters of reliability, safety, and cost of Fast Breeder Reactors (FBR). Therefore, a quantitative index which can connect different areas is required. In addition, the determination of its target value is also one of the key points to substantiate the SBC concept. We have proposed a new method to determine the reliability targets for the structures and components in FBR plants from the safety point of view by utilizing analysis models of a probabilistic safety assessment. In this study, the effectiveness of the proposed index and the compatibility of the reliability targets derived by the new method were shown through a trial setting of In-Service Inspection request on the reactor vessel near the sodium surface based on the SBC concept.
Takaya, Shigeru; Okajima, Satoshi; Asayama, Tai; Chitose, Hiromasa*; Machida, Hideo*; Yokoi, Shinobu*; Kamishima, Yoshio*
no journal, ,
An evaluation method of the occurrence probability of a through-wall crack in a reactor vessel of a fast breeder reactor due to fatigue-creep interaction has been proposed. Input data were prepared for a trial evaluation and the proposed evaluation method was applied. The result was compared with the allowable occurrence probability derived from the safety requirements for FBR.
Takaya, Shigeru; Asayama, Tai; Machida, Hideo*; Kamishima, Yoshio*
no journal, ,
This paper describes a procedure of structural reliability evaluation of passive components of fast breeder reactors for a failure due to fatigue-creep interaction damage, and also reports results of benchmark evaluation conducted by using two independently developed codes. The structural integrity of a reactor vessel was evaluated deterministically and probabilistically. The results estimated by two codes agree well for both of deterministic and probabilistic evaluations, which shows that these codes are programmed properly according to the evaluation procedure.
Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*; Asayama, Tai
no journal, ,
This paper describes benchmark evaluation of independently programmed structural reliability evaluation codes, MSS-REAL and GENPRO/PEPPER. First, a procedure of structural reliability evaluation of passive components of fast reactors for a failure due to fatigue-creep interaction damage is explained. The procedure consists of evaluations of crack initiation, crack propagation, and crack penetration or unstable fracture. Next, input data for evaluation of structural reliability of a reactor vessel is prepared. Then, the structural reliability of a reactor vessel is evaluated deterministically and probabilistically. The results estimated by two codes agreed well for both deterministic and probabilistic evaluations, which shows that these codes are programmed properly according to the evaluation procedure.
