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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

First-principles modeling for dislocation motion of HEA alloys

Tsuru, Tomohito; Itakura, Mitsuhiro; Yuge, Koretaka*; Aoyagi, Yoshiteru*; Shimokawa, Tomotsugu*; Kubo, Momoji*; Ogata, Shigenobu*

Proceedings of 4th International Symposium on Atomistic and Multiscale Modeling of Mechanics and Multiphysics (ISAM-4) (Internet), p.59 - 62, 2019/08

High entropy alloys (HEAs) are chemically complex single- or multi-phase alloys with crystal structures. There are no major components but five or more elements are included with near equiatomic fraction. In such a situation, deformation behavior can no longer be described by conventional solid solution strengthening model. Some HEAs, indeed, show higher strengthening behavior and anomalous slip. However, the mechanisms of these features have yet to be understood. In the present study, we investigate the core structure of dislocations in BCC-HEAs using density functional theory (DFT) calculations. We found that core structure of a screw dislocation is identified as is the case with common BCC metals. On the other hand, dislocation motion should be different from pure BCC metals because of chemical and configurational disorder around dislocation core. We confirmed the specific feature of dislocation motion in HEAs by two-dimensional Peierls potential surface.

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

Coolability evaluation of debris bed on core catcher in a sodium-cooled fast reactor

Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05

Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.

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

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

Progress of design and related researches of sodium-cooled fast reactor in Japan

Kamide, Hideki; Sakamoto, Yoshihiko; Kubo, Shigenobu; Oki, Shigeo; Ohshima, Hiroyuki; Kamiyama, Kenji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures in order to strengthen safety of a fast reactor since the Great East Japan Earthquake. This paper describes the progress of design study and research and development related to safety enhancement and the severe accident countermeasures. For the purpose of strengthening of decay heat removal function, several researches have been carried out on the decay heat removal in a core disruptive accident (CDA), diversity and applicability of decay heat removal systems, and thermal hydraulic evaluation methods. In order to elucidate the behavior of molten fuel during CDA, some in-pile and out-of-pile tests has been performed by international collaboration including basic experiments. Core design was also improved from the viewpoint of preventing the occurrence of severe accident.

Journal Articles

Study on safety design concept for future sodium-cooled fast reactors in Japan

Kubo, Shigenobu; Shimakawa, Yoshio*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

This paper describes safety design concept for future sodium-cooled fast reactors (SFRs) in Japan, which is based on the safety design criteria and safety design guidelines developed in the auspices of the international forum of generation IV nuclear energy systems. Inherent and/or passive design features are utilized based on SFRs characteristics such as low pressure, high thermal inertia of the system. Lessons learned from the Fukushima Dai-ichi accident is one of important issues to be incorporated into the safety design concept. In order to realize commercial SFRs in the future, robust and rational safety design should be pursued by integrating various factors into the design and limiting additional specific systems, structures and components. Existing engineering principle for the design and manufacturing of SFR's components, and innovative technologies introduced in the FaCT project are keys to achieve the safety concept.

Journal Articles

SAS4A analysis study on the initiating phase of ATWS events for generation-IV loop-type SFR

Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.

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

Safety evaluation of self actuated shutdown system for Gen-IV SFR

Saito, Hiroyuki*; Yamada, Yumi*; Oyama, Kazuhiro*; Matsunaga, Shoko*; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

A self-actuated shutdown system (SASS) is a passive device, which can detach a control rod for reactor shutdown in response to excessive increase in coolant temperature. Since a detachment temperature, which triggers release of a control rod, and a response time are identified as important parameters for validity analyses, this study focused on investigation of the response time and the detachment temperature, and safety analysis to see feasibility of the SASS in low power. For this purpose, design modifications were made to shorten the response time and three-dimensional thermal-hydraulic analysis in a low power operation was carried out in order to confirm the response time. The resulting detachment temperature level is lower than previous studies, leading to improved safety parameters. Based on improved parameters, a safety analysis to see feasibility of the SASS in low power was carried out. From this safety evaluation, it was confirmed that core damage can be prevented by the SASS with flow collector in the case of LOF type ATWS event.

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

Experiments EAGLE project for fast reactor safety; A Joint-research program with the Republic of Kazakhstan (NNC/RK)

Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu

Enerugi Rebyu, 36(11), p.46 - 49, 2016/11

no abstracts in English

Journal Articles

Event sequence analysis of core disruptive accident in a metal-fueled sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Kubo, Shigenobu; Ueda, Nobuyuki*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 10 Pages, 2016/11

In this study, the event sequence analysis of CDA in a large metal-fueled SFR has been performed in order to investigate reactivity progression and molten fuel relocation behavior in the metal-fueled SFR. The initiating phase analysis during the CDA initiated by unprotected loss-of-flow accidents has been conducted using the CANIS code, which showed a small power peak. Using the initial conditions based on the initiating phase analysis, the SIMMER-III code was applied to a whole-core scale analyses to clarify the event sequence including the reactivity progression and molten fuel relocation. As a result, recriticality in the whole core analysis resulted in a very mild energy release. The mild energy release in the metal-fueled core can be attributed to the small specific heat of metal fuel and the large prompt negative reactivity feedback mechanism.

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

Development of safety design guideline for generation-IV sodium-cooled fast reactors

Kubo, Shigenobu

Nippon Genshiryoku Gakkai-Shi, 57(10), p.667 - 671, 2015/10

no abstracts in English

133 (Records 1-20 displayed on this page)