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Uchita, Masato*; Miyagawa, Takayuki*; Dozaki, Koji*; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hayafune, Hiroki; Suzuno, Tetsuji*; Fukasawa, Tsuyoshi*; Kamishima, Yoshio*; Fujita, Satoshi*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.380 - 386, 2018/04
It is well-known that pool-type SFRs are the main streams recently in a field of Generation IV reactors. The pool-type encloses primary pumps and IHXs located around the core barrel in a main vessel. Consequently, the main vessel diameter trends to be larger than that of loop-types. From the viewpoint of commercialization in the future, a target of the vessel diameter and its weight including Sodium coolant will increase further. In this paper, the prospects are described in terms of seismic design and structural integrity for the thermal loadings to prevent buckling of the reactor vessel based on parameter studies with diameters of the vessel. In addition, the seismic isolation device which will be effective as a countermeasure is proposed in order to secure a margin against buckling of a large reactor vessel.
Takaya, Shigeru; Sasaki, Naoto*; Asayama, Tai; Kamishima, Yoshio*
Mechanical Engineering Journal (Internet), 4(3), p.16-00558_1 - 16-00558_12, 2017/06
In this study, we developed a new design rule for the prevention of seismic buckling of vessels using the load and resistance factor design method to enable more rational vessel designs. The effectiveness of the new design rule was illustrated in comparison with the current provision.
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.
Yokoi, Shinobu*; Kamishima, Yoshio*; Sadahiro, Daisuke*; Takaya, Shigeru
JAEA-Data/Code 2016-002, 38 Pages, 2016/07
JAEA-Data-Code-2016-002.pdf:1.51MB
Many efforts have been made to implement the System Based Code concept aiming at optimizing margins dispersed in existing codes and standards. Failure probability calculated based on statistical information such as a type of probability distribution, mean (or median) and variance (or standard deviation) for random variables is expected to be a promising quantitative index for margin optimization. Statistical information on material strength, which is also required to calculate the failure probability, has been already reported in JAEA-Data/Code 2015-002 "Structural Properties of Material Strength for Reliability Evaluation of Components of Fast Reactors -Austenitic Stainless Steels-" whereas others have not been identified yet. This report provides methodologies and basic ideas to determine statistical parameters of other random variables, especially variable loads, necessary for reliability evaluation.
Chikazawa, Yoshitaka; Kato, Atsushi; Hayafune, Hiroki; Shimakawa, Yoshio*; Kamishima, Yoshio*
Nuclear Technology, 192(2), p.111 - 124, 2015/11
Times Cited Count：1 Percentile：9.74(Nuclear Science & Technology)Evaluation of severe external hazards on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. For tsunam, hypothetical station blackout has been evaluated.
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 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.
Kawasaki, Nobuchika; Sakamoto, Yoshihiko; Eto, Masao*; Taniguchi, Yoshihiro*; Kamishima, Yoshio*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.760 - 769, 2015/05
The Japan Sodium-cooled Fast Reactor, JSFR, is currently under conceptual study. The concept of JSFR's reactor system is a compact reactor system to avoid excessive increase of reactor vessel diameter with structural and fluid integrities. To realize this concept, single rotating plug with advanced refueling system is adopted. Advanced refueling system consists of column type Upper Internal Structure and pantograph type Fuel Handling Machine. To realize structural and fluid integrities, top entry piping, sodium dam and flow block/guide structures are adopted. Structural integrities against seismic displacement or thermal stress and fluid integrities against vortex cavitations or cover gas entrainment can be ensured with these designs.
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.
Asayama, Tai; Miyagawa, Takayuki*; Dozaki, Koji*; Kamishima, Yoshio*; Hayashi, Masaaki*; Machida, Hideo*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept. A brief introduction to the SBC concept is followed by the technical features of structural evaluation methodologies that are based on the SBC concept. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.
Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 4 Pages, 2014/07
This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are "General rules", "Reliability evaluation", "Failure scenario setting", "Modeling", and "Failure probability calculation", respectively. Details of each chapter are explained.
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.
Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.
Okamura, Shigeki*; Eto, Masao*; Kamishima, Yoshio*; Negishi, Kazuo; Sakamoto, Yoshihiko; Kitamura, Seiji; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
This paper describes the seismic design of JSFR, which includes the seismic condition, the seismic isolation system and the seismic evaluation of primary component. JSFR employs a seismic isolation system to mitigate the earthquake force. The design seismic loading is made more severe than ever since Niigata-ken Chuetsu-oki Earthquake in 2007. The earthquake force loaded on the primary components has to be mitigated more than that of the previous seismic isolation system. We examined the advanced seismic isolation system by optimizing the performance of the previous seismic isolation system considering the natural frequency of the primary components. The advanced seismic isolation system for SFR was adopted laminated rubber bearings which are thicker than that of the previous, as well as oil dampers. The seismic evaluation of nuclear reactor components under applying the advanced seismic isolation system was performed; the performance of the system was confirmed.
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
Sakai, Takaaki; Kotake, Shoji; Aoto, Kazumi; Ito, Takaya*; Kamishima, Yoshio*; Oshima, Jun*
Proceedings of 2010 International Congress on Advances in Nuclear Power Plants (ICAPP '10) (CD-ROM), p.521 - 530, 2010/06
JAEA is now conducting "Fast Reactor Cycle Technology Development (FaCT)" project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. In this paper, current status of the conceptual design study will be summarized with related research and developments on plant technologies.
Sakai, Takaaki; Kotake, Shoji; Aoto, Kazumi; Ito, Takaya*; Kamishima, Yoshio*; Oshima, Jun*
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
JAEA is now conducting "Fast Reactor Cycle Technology Development (FaCT)" project for commercialization before 2050s. A demonstration reactor for Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since FY2007 to determine referential reactor specifications for the next stage of design work of licensing and construction study. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. In this paper, the current status of the conceptual design study for the demonstration reactor plant is summarized.
Kotake, Shoji; Sakamoto, Yoshihiko; Mihara, Takatsugu; Kubo, Shigenobu*; Uto, Nariaki; Kamishima, Yoshio*; Aoto, Kazumi; Toda, Mikio*
Nuclear Technology, 170(1), p.133 - 147, 2010/04
Times Cited Count：36 Percentile：91.14(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) is now executing "Fast Reactor Cycle Technology Development (FaCT)" project in cooperation with the Japanese electric utilities. In the FaCT project, both the conceptual design study for Japan Sodium-cooled Fast Reactor (JSFR) and the developments of innovative technologies to be adopted to JSFR are now implemented with paying attention to the consistency between the design and the innovative technologies. The current target is that the development will be accomplished around 2015, after that a licensing procedure for the demonstration JSFR will be launched. This paper describes design requirements, design characteristics of JSFR and evaluation on the performances for economic competitiveness. Furthermore, the current status of the key technology development for JSFR is briefly introduced.
Sakamoto, Yoshihiko; Kubo, Shigenobu*; Kotake, Shoji; Kamishima, Yoshio*
Dai-13-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.505 - 506, 2008/06
This paper describes the enhancement of reliability for components of the reactor system in JSFR design. As for manufacturability, compact design of the RV enables its manufacture in a factory. This results in high quality welding and in precise machining of the RV. The adoption of ring-shaped forgings contributes for securing the reliability against thermal stress as well as securing the dimension precision. Regarding maintenability, the in-vessel structures have simple configurations, so it is comparatively easy for inspection equipments to reach inspection targets. In the JSFR design, sodium boundary area is reduced significantly, which makes double-walled design of the piping easier, and reduces welding lines. So, the reactor system of JSFR is desirable to inspect the in-vessel structures efficiently, and there is a prospect of reliable plant operation. Advanced inspection technologies are also under development for the inspection of the in-vessel structures under sodium.
