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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

Development of prototype reactor maintenance, 3; Application to valves of sodium-cooled reactor prototype

Chikazawa, Yoshitaka; Takaya, Shigeru; Tagawa, Akihiro; Kubo, Shigenobu

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 6 Pages, 2019/05

A maintenance management required to prototype nuclear power reactors has been developed. One of important mission of a prototype reactor is to develop maintenance program for commercial reactors step by step securing safety. Since operating experience at the early stage is limited, the maintenance program for the prototype reactor should be a progressive one. It has to be modified and improved frequently taking into account R&D insight and operation experiences. Additionally, the maintenance program has to consider features of the prototype reactor even at the early stage. To select maintenance grades on particular components/systems, risk informed and graded approaches are effective. And maintenance programs have to take into account degradation mechanism originally due to reactor features. In this paper, applications for maintenance program on sodium valves of prototype fast breeder reactor Monju are studied as an example of prototype sodium-cooled reactors (SFR).

Journal Articles

Development of under sodium viewer for next generation sodium-cooled Fast reactor; Imaging test in sodium

Aizawa, Kosuke; Chikazawa, Yoshitaka; Ara, Kuniaki; Yui, Masahiro*; Jinno, Kentaro*; Hiramatsu, Takashi*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 7 Pages, 2019/05

Inspection technique in opaque liquid metal coolant is one of important issues for sodium-cooled fast reactors. Various under sodium viewers (USVs), including horizontal USVs for obstacle detection and imaging USVs, have been developed in several research institutes and countries. We aim practical realization of imaging USV which adopts an optical receiving system, which measures the vibration displacement of diaphragm by using a laser as a receiving sensor. This study mainly focuses on the sensitivity improvement of a receiving sensor. An issue for the sensitivity improvement of the receiving sensor is the sound pressure propagation inside the receiving sensor. Prototype tests in the water and sodium were conducted in order to resolve the issue. In addition, imaging experiments in the water and sodium were conducted using the improved receiving sensor. From the results of imaging experiments, the relation between obtained wave profile and the regeneration imaging was confirmed.

Journal Articles

In-vessel thermal-hydraulics analyses of the ASTRID-600MWe reactor with STAR-CCM+ code to supply boundary conditions for mechanical evaluation

Onoda, Yuichi; Chikazawa, Yoshitaka; Nakamura, Hironori*; Barbier, D.*; Dirat, J.-F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

no abstracts in English

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.

Journal Articles

Performance evaluation of eddy current flowmeter in Monju

Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohashi, Yuko

Journal of Nuclear Science and Technology, 55(12), p.1393 - 1401, 2018/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor's thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter's signal intensity and the primary sodium's flow rate under 10-100% flow rate condition. The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.

Journal Articles

Demonstration of under sodium viewer in Monju

Aizawa, Kosuke; Sasaki, Koei; Chikazawa, Yoshitaka; Fukuie, Masaru*; Jimbo, Noboru*

Nuclear Technology, 204(1), p.74 - 82, 2018/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Development of inspection technique in opaque liquid metal coolant is one of the important issues to ensure the safety of Liquid Metal Fast Breeder Reactor (LMFBR). Performance tests of an Under Sodium Viewer (USV), which was developed to detect an obstacle in the reactor vessel (RV) of LMFBR Monju, have been carried out. The ultrasonic sensors and reflectors are located across the core inside of the Monju's RV. The USV can detect an obstacle existing in between the core top and the Upper Core Structure (UCS) bottom by identifying differences of echo signals. This report describes the USV performance tests. In the tests, the reference echo signals under various conditions were accumulated and the signal to noise ratio successfully exceeded the target value. Measured signals clearly differed from with and without an obstacle. These experimental results show the performance of the USV for detecting an obstacle in the specified place.

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.

Journal Articles

Advanced sodium-cooled fast reactor development regarding GIF safety design criteria

Hayafune, Hiroki; Chikazawa, Yoshitaka; Kamide, Hideki; Iwasaki, Mikinori*; Shoji, Takashi*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 11 Pages, 2017/06

Design studies on a next generation sodium-cooled fast reactor (SFR) considering the safety design criteria (SDC) developed in the generation IV international forum (GIF) was summarized. To meet SDC including the lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, the heat removal function was enhanced to avoid loss of the function even if any internal events exceeding design basis or severe external event happen. Several design options have been investigated and auxiliary core cooling system using air as ultimate heat sink has been selected as an additional cooling system regarding system reliability and diversification. Even though the next generation SFR already adopts seismic isolation system, main component designs have been improved considering revised earthquake conditions. For other external events, design measures for various external events are taken into account. Reactor building design has been improved and important safety components are diversified and located separately improving independency. Those design studies and evaluations on the next generation sodium-cooled reactor have contributed to the development of safety design guidelines (SDG) which is under discussion in the GIF framework.

Journal Articles

Development of under sodium viewer for next generation sodium-cooled fast reactors

Aizawa, Kosuke; Chikazawa, Yoshitaka; Ara, Kuniaki; Yui, Masahiro*; Uemoto, Yohei*; Kurokawa, Masaaki*; Hiramatsu, Takashi*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 9 Pages, 2017/06

Inspection in opaque liquid metal coolant is one of important issues for sodium-cooled fast reactors. To facilitate operations and maintenance activities, various under sodium viewers (USVs), including horizontal USVs for obstacle detection for a long distance and imaging USVs for a short and middle distance imaging, have been developed in several research institutes and countries. In this study, an imaging USV for a middle distance, approximately 1 m, has been developed. The USV in this study adopts an optical receiving system which measures the vibration displacement of diaphragm by using a laser as a receiving sensor. This study mainly focuses on the sensitivity improvement for a transmission sensor and the receiving sensor. In addition, an imaging experiment in the water was conducted using the new transmission sensor and receiving sensor. The experimental results showed that the newly developed USV sensors can make higher resolution images of a target than the previous sensors.

Journal Articles

Development of prototype reactor maintenance, 1; Application to piping system of sodium-cooled reactor prototype

Kotake, Shoji*; Chikazawa, Yoshitaka; Takaya, Shigeru; Otaka, Masahiko; Kubo, Shigenobu; Arai, Masanobu; Kunogi, Kosuke; Ito, Takaya*; Yamaguchi, Akira*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

A maintenance management required to prototype nuclear power reactors is proposed. Monitoring and control of sodium impurity and thermal transient are extremely important for sodium boundary maintenance for sodium-cooled fast reactors. At the fast stage of the prototype reactor Monju operation, degradation mechanism on the piping should be demonstrated based on operation experiences. Therefore inspection on a representative position for crack indication and pipe thickness is proposed. Due to less experience of SFR plants, early detection of boundary failure is considered. For a matured operation stage, when degradation mechanism is well demonstrated based on inspection data, inspection cycle could be extended. And for commercial reactors, maintenance without inspection will be established based on accumulated operation experiences including those of the prototype reactor Monju.

Journal Articles

Core performance requirements and design conditions for next-generation sodium-cooled fast reactor in Japan

Oki, Shigeo; Maruyama, Shuhei; Chikazawa, Yoshitaka; Ohtaki, Akira; Kubo, Shigenobu; Hibi, Koki*; Kan, Taro*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

Journal Articles

Development of prototype reactor maintenance, 2; Application to piping support of sodium-cooled reactor prototype

Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji*; Ito, Takaya*; Yamaguchi, Akira*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04

Applications for maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of the piping support could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports.

Journal Articles

Design study on measures to prevent loss of decay heat removal in a next generation sodium-cooled fast reactor

Chikazawa, Yoshitaka; Kubo, Shigenobu; Shimakawa, Yoshio*; Kaneko, Fumiaki*; Shoji, Takashi*; Nakata, Shuhei*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

Sodium-cooled reactor (SFR) has superior characteristics thanks to sodium coolant features such as low pressure and high natural convection capability. Involving lessons learned from the 1F accident, requirements on design base DHRS have been modified. In that modification, safety requirements on design extended conditions have been clarified and sodium temperature criteria have been changed taking into account design margin even for design extended conditions. With the new DHRS configuration including ACS, designs of component cooling water system and emergency power supply have been updated.

Journal Articles

Proposal of maintenance management of nuclear power plants at R&D stage by taking account of their features

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

Hozengaku, 15(4), p.71 - 78, 2017/01

A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. Finally, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.

Journal Articles

Secondary sodium fire measures in JSFR

Chikazawa, Yoshitaka; Kato, Atsushi*; Yamamoto, Tomohiko; Kubo, Shigenobu; Ohno, Shuji; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*

Nuclear Technology, 196(1), p.61 - 73, 2016/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room and air cooler have been analyzed evaluating performances of the candidate sodium fire measures.

JAEA Reports

Maintenance management of nuclear power reactors at the stage of research and development

Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi

JAEA-Research 2016-006, 66 Pages, 2016/07

JAEA-Research-2016-006.pdf:3.4MB

A maintenance management required to nuclear power reactors at the R&D stage was discussed. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R&D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R&D stage was clarified. Next, requirements and consideration for maintenance management was discussed according to the objectives. "Codes for maintenance management of nuclear power plants" and "Guides for maintenance management of nuclear power plants" were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R&D stage were newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented.

Journal Articles

Demonstration of eddy current type flow meter in Monju

Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohashi, Yuko

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.314 - 320, 2016/04

Temperature and flow rate measurement of each fuel subassembly outlet is effective to detect loss of coolant accident (LOCA) and reactivity initiated accident (RIA) early and to understand a thermal hydrodynamic performance in liquid metal fast breeder reactor (LMFBR). This report shows the data of eddy current type flow meters in Monju. High linearity between the signal intensity of each eddy current type flow meter and flow rate of primary sodium was obtained in the flow rate condition of 10$$sim$$100%. In addition, the linearity was also demonstrated in the low velocity region, approx. 0.25 m/s. Fluctuation shown on each eddy current type flow meter was below 0.2 m/s, which is 5 % of the time averaged velocity at the rated condition. Those experimental results show that the eddy current type flow meter can detect the change of relative flow rate.

Journal Articles

Severe external hazard on hypothetical JSFR in 2010

Chikazawa, Yoshitaka; Kato, Atsushi; Hayafune, Hiroki; Shimakawa, Yoshio*; Kamishima, Yoshio*

Nuclear Technology, 192(2), p.111 - 124, 2015/11

 Times Cited Count:1 Percentile:86.4(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.

Journal Articles

Water experiment on phased array acoustic leak detection system for sodium-heated steam generator

Chikazawa, Yoshitaka; Yoshiuji, Takahiro*

Nuclear Engineering and Design, 289, p.1 - 7, 2015/08

 Times Cited Count:4 Percentile:54.36(Nuclear Science & Technology)

A phased array acoustic leak detection system for sodium heated steam generator has been proposed. The major advantage of the new system is it could provide information of acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the new system could separate the target leak source from steam generator background noise. In the previous study, the methodology was proposed and basic performance was confirmed by numerical analysis. However, in the numerical analysis, acoustic transportation through the SG tube bundle was not modeled. In the present study, performance the proposed system has been confirmed in water experiments with mockup tube bundles.

197 (Records 1-20 displayed on this page)