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Journal Articles

Estimation of creep behavior of thick rubber bearings from 47 years observation in an actual building

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

Journal Articles

Development of leak before break assessment guidelines for sodium cooled fast reactors in Japan

Yada, Hiroki; Wakai, Takashi; Miyagawa, Takayuki*; Machida, Hideo*

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 10 Pages, 2021/07

Journal Articles

Fundamental study on seismic safety margin for seismic isolated structure using the laminated rubber bearings

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.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

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.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

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.

JAEA Reports

Research of the tasks on risk communication enforcement in fiscal year 2016 (Contract research)

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

JAEA-Research-2018-001.pdf:2.49MB

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.

Journal Articles

Seismic evaluation for a large-sized reactor vessel targeting SFRs in Japan

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.

JAEA Reports

Research of the tasks on risk communication enforcement in fiscal year 2015 (Contract research)

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-Research-2017-003.pdf:2.92MB

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.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 3; Progress of component design

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.

Journal Articles

Elaboration of the system based code concept; Activities in JSME and ASME, 1; Overview

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.

JAEA Reports

Monju system start-up test report evaluation of the feedback reactivity

Miyagawa, Takayuki*; Kitano, Akihiro; Okawachi, Yasushi

JAEA-Technology 2014-008, 60 Pages, 2014/05

JAEA-Technology-2014-008.pdf:29.75MB

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.

Journal Articles

Evaluation of feedback reactivity in Monju start-up test

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.

JAEA Reports

Core confirmation test in system startup test of the fast breeder reactor MONJU

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

JAEA-Technology-2010-052.pdf:17.14MB

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..

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 2; Criticality, control rod worth measurement

Yabuki, Kentaro; Kitano, Akihiro; Fukushima, Masahiro; Miyagawa, Takayuki; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 3; Evaluation of neutron source

Kato, Yuko; Miyagawa, Takayuki; Kageyama, Takeshi; Okimoto, Yutaka

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 4; Neutron instrumentation soundness measurement

Takano, Kazuya; Miyagawa, Takayuki; Ikegami, Kazunori; Kitano, Akihiro

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "zero power reactor physics test", 8; Feedback reactivity measurement

Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "Reactor Physics Test", 2; Criticality, control rod worth measurement

Yabuki, Kentaro; Kitano, Akihiro; Fukushima, Masahiro; Miyagawa, Takayuki; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "Reactor Physics Test", 8; Feedback reactivity measurement

Miyagawa, Takayuki; Kitano, Akihiro; Muranaka, Makoto; Kato, Mitsuya*; Okawachi, Yasushi

no journal, , 

no abstracts in English

Oral presentation

Prototype FBR Monju system start up test "Reactor Physics Test", 3; Evaluation of neutron source

Kato, Yuko; Miyagawa, Takayuki; Kageyama, Takeshi; Okimoto, Yutaka

no journal, , 

no abstracts in English

41 (Records 1-20 displayed on this page)