Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Yada, Hiroki; Takaya, Shigeru; Morohoshi, Kyoichi*; Yokoi, Shinobu*; Miyagawa, Takayuki*
Mechanical Engineering Journal (Internet), 10(4), p.23-00044_1 - 23-00044_13, 2023/08
To develop rationalized maintenance plans for nuclear power plants, the characteristics of each plant must be considered. For sodium-cooled fast reactor (SFR) plants, constraints on inspections exist due to the specialty that equipment retaining sodium must be handled, which is one of the important points that must be considered in maintenance rationalization. In this study, we propose a maintenance optimization scheme, which is a design support tool, using risk information to develop a maintenance strategy based on the system based code (SBC) concept. The SBC concept intends to provide a theoretical procedure to optimize the reliability of structure, system and components (SSCs) by administrating every related engineering requirements throughout the life of the SSCs from design to decommissioning. ASME Boiler and Pressure Vessel Code, Code Case, N-875 was developed based on the SBC concept. The purpose of this study is to establish detailed procedures for the maintenance optimization scheme based on the procedure in Code Case N-875. Furthermore, a quantitative trial evaluation of the core support structure of the next SFR under development in Japan is also performed using the maintenance optimization scheme.
Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04
The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.
Masaki, Nobuo*; Kato, Koji*; Yamamoto, Tomohiko; Miyagawa, Takayuki*; Fujita, Satoshi*; Okamura, Shigeki*
Nihon Kenchiku Gakkai Gijutsu Hokokushu, 28(68), p.81 - 84, 2022/02
no abstracts in English
Yada, Hiroki; Wakai, Takashi; Miyagawa, Takayuki*; Machida, Hideo*
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07
Fukasawa, Tsuyoshi*; Miyagawa, Takayuki*; Uchita, Masato*; Yamamoto, Tomohiko; Miyazaki, Masashi; Okamura, Shigeki*; Fujita, Satoshi*
Nihon Kikai Gakkai Rombunshu (Internet), 87(898), p.21-00007_1 - 21-00007_17, 2021/06
This paper describes a fundamental study on the seismic safety margin for the isolated structure using laminated rubber bearings. The variation of the seismic response assumed in the isolated structure will occur under the superposition of "Variations in seismic response due to input ground motions" and "Error with design value accompanying manufacture of the isolation devices ". The seismic response analysis which allows to their conditions is important to assess the seismic safety margin for the isolated structure. This paper clarifies that the seismic safety margin of the isolated structure, which consists of rubber bearings, for Sodium-cooled Fast Reactor (SFR) is ensured against the basis ground motions of Japan Electric Association Guide 4601 (JEAG4601) and SFR through the seismic response analysis considering the variation factors of seismic response. In addition, a relationship between the seismic safety margin and the excess probability of linearity limits is discussed using the results of seismic response analysis.
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.
Mechanical Engineering Journal (Internet), 7(3), p.19-00489_1 - 19-00489_16, 2020/06
The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.
Tanaka, Masaru*; Kawara, Osami*; Ishizaka, Kaoru*; Ohata, Yuki*; Fukuike, Iori*; Kawase, Keiichi; Tokizawa, Takayuki; Miyagawa, Hiroshi*; Ishimori, Yuu
JAEA-Research 2018-001, 98 Pages, 2018/06
In the 2016 fiscal year, communication cases on general waste disposal facility construction plans in recent years were surveyed. Results suggested as follows: (1) Existing long-term relationships or agreements in local area promote local accepting. (2) An operator needs to consider alternative plans and explain reasons for the decision making to local stakeholders. (3) Even after first announcement of a new plan, an operator needs to review the plan depending on local concerns. (4) Announcement of a new plan will activate communications on local development including the site redevelopment.
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.
Tanaka, Masaru*; Aoyama, Isao*; Ishizaka, Kaoru*; Ohata, Yuki*; Fukuike, Iori*; Kawase, Keiichi; Watanabe, Masanori; Tokizawa, Takayuki; Miyagawa, Hiroshi*; Ishimori, Yuu
JAEA-Research 2017-003, 65 Pages, 2017/06
JAEA Ningyo-toge Environmental Engineering Center and Fukushima Environmental Safety Center have same challenges in risk communication. As reference, similar domestic cases were investigated by our two Centers, and requirements for building long-term relationship were clarified. As follows; (1) Develop new relationship with various stakeholders in the region. (2) Make better use of existing resources (personnel, land and facilities, etc.). (3) Make a concerted effort to create new values with local stakeholders. (4) Make an opportunity which local stakeholders confirm safety and build confidence to the project. These efforts will enhance the opportunities for operators and residents to learn about environment management and environmental protection.
Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi*; Eto, Masao*; Miyagawa, Takayuki*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In the frame work of generation IV international forum, safety design criteria and safety design guideline for the generation IV sodium-cooled fast reactors have been developing. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC. In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX was modified for the primary heat exchanger, which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator, protective wall tube type design is under investigation as an option with less R&D risks.
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.
Miyagawa, Takayuki*; Kitano, Akihiro; Okawachi, Yasushi
JAEA-Technology 2014-008, 60 Pages, 2014/05
The prototype fast breeder reactor Monju resumed the system startup test (SST) on May 6, 2010 after fourteen year and five month shutdown since the sodium leakage of the secondary heat transport system in December 1995 and reached criticality on May 8, 2010. Core Confirmation Test (CCT) is the first step of SST which consists of three steps, and finished on July 22 after 78 days test. In the evaluation of the feedback reactivity at the part of the CCT, the "self-stability" of Monju was observed when the positive reactivity was inserted with the control rod withdrawal, due to the negative feedback property of the reactor, and due to the control properties of the auxiliary cooling system. Parameters represented with reactor power, sodium temperature of the primary loops became to be stable after transient without any operations. Additionally, the quantitative feedback reactivity was evaluated using the results of this test tentatively.
Kitano, Akihiro; Miyagawa, Takayuki*; Okawachi, Yasushi; Hazama, Taira
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03
The feedback reactivity was measured in Monju start-up test conducted in 2010. The two reactivity components related either to power or to the core inlet coolant temperature were evaluated by fitting to a reactivity balance equation as a function of neutron count rate and coolant temperature. The measured feedback reactivity and the two components were compared with calculation taking account of the temperature distribution in the core. The calculated and the measured values of the feedback reactivity showed a reasonable agreement.
Jo, Takahisa; Goto, Takehiro; Yabuki, Kentaro; Ikegami, Kazunori; Miyagawa, Takayuki; Mori, Tetsuya; Kubo, Atsuhiko; Kitano, Akihiro; Nakagawa, Hiroki; Kawamura, Yoshiaki; et al.
JAEA-Technology 2010-052, 84 Pages, 2011/03
The prototype fast breeder reactor MONJU resumed the System Startup Test (SST) on May 6th 2010 after five months and fourteen years shutdown since the sodium leakage of the secondary heat transport system on December 1995. Core Confirmation Test (CCT) is the first step of SST, which consists of three steps. CCT was finished on July 22nd after 78 days tests. CCT is composed 20 test items including control rods' worth evaluation, radiation dose measurement etc..
Yabuki, Kentaro; Kitano, Akihiro; Fukushima, Masahiro; Miyagawa, Takayuki; Okawachi, Yasushi
no journal, ,
no abstracts in English
Kato, Yuko; Miyagawa, Takayuki; Kageyama, Takeshi; Okimoto, Yutaka
no journal, ,
no abstracts in English
Takano, Kazuya; Miyagawa, Takayuki; Ikegami, Kazunori; Kitano, Akihiro
no journal, ,
no abstracts in English
Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi
no journal, ,
no abstracts in English
Yabuki, Kentaro; Kitano, Akihiro; Fukushima, Masahiro; Miyagawa, Takayuki; Okawachi, Yasushi
no journal, ,
no abstracts in English