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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

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

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

Estimation of mitigation effects of sodium nanofluid for SGTR accidents in SFR

Ichikawa, Kenta*; Kanda, Hironori; Yoshioka, Naoki*; Ara, Kuniaki; Saito, Junichi; Nagai, Keiichi

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07

Studies on the suppression of the reactivity of sodium itself have been performed on the basis of the concept of suspended nanoparticles in liquid sodium (sodium nanofluid). According to the experimental and theoretical results of studies for sodium nanofluid, velocity and heat of sodium nanofluid-water reaction are lower than those of the pure sodium-water reaction. The analytical model for the peak temperature of a sodium nanofluid-water reaction jet has been developed in consideration of these suppression effects by the authors. In this paper, the prediction method for mitigation effects for a damage of adjacent tubes in a steam generator tube rupture (SGTR) accidents is arranged by applying this analytical model for the peak temperature of the reaction jet. On the assumption that the sodium nanofluid is used for the secondary coolant of sodium-cooled fast reactor (SFR), mitigation effects under the design-base accident (DBA) condition and the design-extension condition (DEC) of SGTR are estimated by using this method. As a result, there is a possibility to reduce the number of damaged tubes and to suppress the pressure generated by SGTR accidents by using sodium nanofluid in the secondary coolant.

Journal Articles

Research and development of thick rubber bearing for SFR; Aging properties tests of semi full-scale thick rubber bearing

Watakabe, Tomoyoshi; Yamamoto, Tomohiko; Fukasawa, Tsuyoshi*; Okamura, Shigeki*; Somaki, Takahiro*; Morobishi, Ryota*; Sakurai, Yu*; Kato, Koji*

Nippon Kikai Gakkai Rombunshu (Internet), 83(850), p.16-00444_1 - 16-00444_14, 2017/06

A seismic isolation system composed of a thick rubber bearing and an oil damper has been developed for Sodium-Cooled Fast Reactor. This paper focused on the aging properties of thick rubber bearings, such as basic mechanical properties and ultimate strength. Aging of the rubber bearings was reproduced using thermal degradation based on Arrhenius law.

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

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo

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

Journal Articles

An Experimental study on natural circulation decay heat removal system for a loop type fast reactor

Ono, Ayako; Kamide, Hideki; Kobayashi, Jun; Doda, Norihiro; Watanabe, Osamu*

Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09

 Times Cited Count:4 Percentile:55.34(Nuclear Science & Technology)

Decay heat removal by natural circulation is a significant passive safety measure of a fast reactor against station blackout. The decay heat removal system (DHRS) of the loop type sodium fast reactor being designed in Japan comprises a direct reactor auxiliary cooling system and primary reactor auxiliary cooling system (PRACS). The thermal hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant. The experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, the experiments varying the pressure loss coefficients of the loop as the experimental parameters showed robustness of the PRACS.

Journal Articles

Determination of in-service inspection requirements for fast reactor components using System Based Code concept

Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai

Nuclear Engineering and Design, 305, p.270 - 276, 2016/08

AA2016-0006.pdf:0.51MB

 Times Cited Count:2 Percentile:68.26(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.

JAEA Reports

Determination methodologies for input data including loads considered for reliability evaluation of fast reactor components

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.

Journal Articles

ASTRID nuclear island design; Advances in French-Japanese joint team development of decay heat removal systems

Hourcade, E.*; Curnier, F.*; Mihara, Takatsugu; Farges, B.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1740 - 1745, 2016/04

In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving highlights of ASTRID DHRS current strategy. Focus is made on operating temperature diversification for in-vessel heat exchanger as well as core catcher coolability by an original features such as heat exchanger located within reactor cold pool, whose design was taken over by Japan team since 2014.

Journal Articles

Study on minimum wall thickness requirement for seismic buckling of reactor vessel based on system based code concept

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:81.81(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.

Journal Articles

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*

Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09

 Times Cited Count:8 Percentile:31.87(Nuclear Science & Technology)

A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.

Journal Articles

Determination of ISI requirements on the basis of system based code concept

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.

Journal Articles

Sodium experiments on natural circulation decay heat removal and 3D simulation of plenum thermal hydraulics

Kamide, Hideki; Ono, Ayako; Kimura, Nobuyuki; Endo, Junji*; Watanabe, Osamu*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13), Companion CD (CD-ROM), 11 Pages, 2015/04

Natural circulation decay heat removal is one of the significant issues for fast reactor safety, especially in long term station blackout events. Several sodium experiments were carried out using a 7-subassmbly core model for core thermal hydraulics under natural circulation conditions and for onset transients of natural circulation in a decay heat removal system (DHRS) including natural draft. Significant heat removal via inter-wrapper flow was confirmed in the experiments. Solidification of sodium in an air cooler is one of key issues in loss of heat sink events. Natural circulation characteristics under long-term decay heat removal were also obtained. Multi-dimensional phenomena, e.g., thermal stratification and bypass flow in plenums and/or heat exchangers, may influence the natural circulation. Thus, 3D simulation method was developed for entire region in the primary loop. Comparison of temperature distributions in a DHRS heat exchanger between experiment and analysis was done.

Journal Articles

Application of the system based code concept to the determination of in-service inspection requirements

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.

Journal Articles

Three-dimensional tsunami analysis for the plot plan of a sodium-cooled fast reactor plant

Hayakawa, Satoshi*; Watanabe, Osamu*; Ito, Kei; Yamamoto, Tomohiko

Nippon Kikai Gakkai Rombunshu, B, 79(808), p.2645 - 2649, 2013/12

As the practical evaluation method of the effect of tsunami on buildings, the formula of tsunami force has been used. However, it cannot be applied to complex geometry of buildings. In this study, to analyze the effect of tsunami on the buildings of sodium-cooled fast reactor plant more accurately, three-dimensional tsunami analysis was performed. In the analysis, VOF (Volume of Fluid) method was used to capture free surface of tsunami. At the beginning, it was confirmed that the tsunami experiment results was reproduced by VOF method accurately. Next, the three-dimensional tsunami analysis was performed with VOF method to evaluate the flow field around the buildings of the plant from the beginning of the tsunami until the backwash of that.

Journal Articles

Study on minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling by system based code

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.

Journal Articles

Development of limit state design for fast reactor by system based code

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.

Journal Articles

Global sensitivity analysis for core hot spot evaluation under natural circulation decay heat removal in sodium-cooled fast reactor

Doda, Norihiro; Kamide, Hideki; Ohshima, Hiroyuki; Watanabe, Osamu*

Proceedings of 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9) (CD-ROM), 11 Pages, 2012/09

In the design study for Japan Sodium Cooled Fast Reactor (JSFR), fully natural circulation system is adopted as the decay heat removal system. We have been developing a new evaluation method of core hot spot in transition from rated operation to natural circulation decay heat removal conditions. Since the method is currently based on conservative assumptions and data, there is room for further rationalization of the safety margin which can be achieved by conducting best estimate analyses with confidence and with quantified uncertainty of results. This paper describes a development of PIRT (Phenomena Identification and Ranking Table) and the global sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the evaluation method.

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