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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; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Journal of Pressure Vessel Technology, 137(5), p.051802_1 - 051802_7, 2015/10
Times Cited Count：2 Percentile：11.54(Engineering, Mechanical)The minimum wall thickness required to prevent seismic buckling of a reactor vessel in a fast reactor is derived using the System Based Code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.
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.
Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 6 Pages, 2013/07
In this paper, minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling is discussed on the basis of the System Based Code (SBC) concept. One of key concepts of SBC is the margin optimization. To implement this concept, reliability design method is employed, and the target reliability for seismic buckling of reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation such as distribution type, mean value and standard deviation of random variable are prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Wall thickness needed to achieve the target reliability is evaluated, and as a result, it is shown that the minimum wall thickness can be reduced from that required by a deterministic design method.
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.
Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi*; Inoue, Kazuhiko*; Kasahara, Naoto
Nuclear Engineering and Design, 238(2), p.389 - 398, 2008/02
Times Cited Count：5 Percentile：35.07(Nuclear Science & Technology)In this paper, the inelastic analysis procedures for the improved design of future fast breeder reactors were validated through the structural model tests and the evaluation of the experimental results by the inelastic analyses. First, a thermal fatigue test of a 316FR hollow cylinder with two longitudinal weldments was conducted under the condition of combined constant axial load and cyclic movement of axial temperature distribution, which simulated the loading condition near the free surface of coolant sodium in the main vessel of fast breeder reactors (FBRs). Second, the inelastic analyses were carried out in accordance with the recommended procedure by using the measured results of oscillating temperature distribution. Finally, the results of inelastic analyses were compared with the experimental results and it was validated that the recommended practice gave a conservative result for the deformation and a good estimation of strain range for the fatigue life evaluation.
Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto
Proceedings of 2006 ASME Pressure Vessels and Piping Division Conference (PVP 2006)/International Council on Pressure Vessel Technology (ICPVT-11) (CD-ROM), 7 Pages, 2006/00
Prevention of thermal ratcheting is an important problem for high temperature components of fast breeder reactors that are subjected to cyclic thermal loads. To clarify ratcheting behaviors, structural model tests were planned. Strain measurement is important for understanding the thermal ratcheting phenomenon, however the conventional measurement by strain gauge is difficult at high temperature. Then, Laser speckle strain measurement system using the dual-beam set-up was developed to apply to high temperature structural model tests. This system was applied to the thermal ratcheting tests, which demonstrated the actual operative conditions of reactor vessels. Through comparison with uniaxial test results obtained by extensometers, the laser speckle method was verified. Measured data of structural model tests were utilized to certify the guidelines of inelastic analysis for design, which provide prediction method of strain in components of fast reactor.
Watanabe, Daigo*; Chuman, Yasuharu*; Otani, Tomomi*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto
no journal, ,
no abstracts in English
Ikeda, Yoshitaka; Okano, Fuminori; Hanada, Masaya; Sakasai, Akira; Miya, Naoyuki; Watanabe, Takashi*; Daigo, Yasuhiko*; Hosogane, Nobuyuki*; Aoto, Mitsuo*
no journal, ,
JT-60, to which the Radiation Hazard Prevention Act is applied, stopped its operation in October 2008 after 18 years deuterium operation since 1991. JT-60 will be upgraded to JT-60SA with superconducting magnet coils, which is the Satellite Tokamak Program under the EU-Japan collaboration, so as to demonstrate the high-beta long-pulse plasma operation. To establish this new device, the existing JT-60 facilities such as magnetic coils, vacuum vessel, basement, diagnostics and heating system. The disassembly of JT-60 is featured by the radiactivation of all components due to neutron from the D-D reaction, and thus, one of the main issues is to manage the physical control of the radiated material for the application of clearance regulation. The disassembly has started since 2009 and will complete by the autumn of 2012. Then a new JT-60SA basement, which will be shipped from EU, will be installed within March 2013.
Okano, Fuminori; Ikeda, Yoshitaka; Sakasai, Akira; Hanada, Masaya; Watanabe, Takashi*; Daigo, Yasuhiko*; Hosogane, Nobuyuki*; Aoto, Mitsuo*
no journal, ,
no abstracts in English
Ikeda, Yoshitaka; Okano, Fuminori; Hanada, Masaya; Sakasai, Akira; JT-60 Team; Watanabe, Takashi*; Daigo, Yasuhiko*; Hosogane, Nobuyuki*; Aoto, Mitsuo*
no journal, ,
Break-even Plasma Test Facilities (JT-60) is the only tokamak device applied by the Radiation Hazard Prevention Act, and is under upgrading towards the super-conducting magnetic device "JT-60SA" aiming long pulse and high beta plasmas. The JT-60SA project is in progress as the satellite Tokamak project under the Japan-EU international program "ITER Broader Approach". At the first step towards JT-60SA, the disassembly of JT-60 tokamak and its affiliated facilities was done for three years and completed in October 2012. All disassembly and cutting works were in control of radiation management. Total number and weight of disassembly components were about 13000 and about 5400 tons, respectively. The JT-60 was composed of a lot of high mechanical strength materials such as high manganese steel which is difficult-to-machine material. New technologies such as a diamond wire-saw which cut the complicated structure simultaneously enabled the effective disassembly.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Araki, Takao*; Watanabe, Kazuhito*; Kittaka, Daigo*; Ishii, Kyoko*; Matsumiya, Hisato*
no journal, ,
Recent findings on safety characteristics of a tokamak DEMO reactor are reported in the case where all the coolant water is lost completely and instantaneously. Assuming that there are neither off-site power nor active emergency cooling, we have analyzed temporal histories of the temperatures of the reactor components using the fusion reactor thermo-hydraulic analysis code MELCOR-fus. We have found that even in such an extremely severe case, the temperatures of the vacuum vessel and in-vessel components do not reach their melting points.
