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Journal Articles

Thermal-hydraulics to risk assessment; Roles of thermal-hydraulics simulation to risk assessment

Maruyama, Yu; Yoshida, Kazuo

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(7), p.517 - 522, 2021/07

no abstracts in English

Journal Articles

The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.221 - 225, 2018/04

no abstracts in English

Journal Articles

Evaluation of sodium pool fire and thermal consequence in two-cell configuration

Takata, Takashi; Ohno, Shuji; Tajima, Yuji*

Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06

Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR). The present paper deals with the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as a typical one to possess important characteristic of multi-compartment system seen in an actual plant. As a result of the numerical analysis using a lumped-parameter based zonal model safety analysis code SPHINCS, the applicability of the ventilation model implemented in SPHINCS has been demonstrated. It is also investigated that the buoyancy- driven ventilation is dominant in the experiment.

JAEA Reports

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

Journal Articles

Thermal-hydraulics technological strategy roadmap that improves safety of LWRs

Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 58(3), p.161 - 166, 2016/03

no abstracts in English

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 2; Heat transfer and flow visualization experiment by using internally heated annulus

Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11

Journal Articles

Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 1; Outline of the research project

Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11

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:11 Percentile:73.32(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

New AESJ thermal-hydraulics roadmap for LWR safety improvement and development after Fukushima accident

Nakamura, Hideo; Arai, Kenji*; Oikawa, Hirohide*; Fujii, Tadashi*; Umezawa, Shigemitsu*; Abe, Yutaka*; Sugimoto, Jun*; Koshizuka, Seiichi*; Yamaguchi, Akira*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5353 - 5366, 2015/08

Journal Articles

A Rapid evaluation method of the heat removed by a VCS before rise-to-power tests

Takamatsu, Kuniyoshi

Journal of Thermal Science, 24(3), p.295 - 301, 2015/06

 Times Cited Count:1 Percentile:8.21(Thermodynamics)

Before rise-to-power tests, the actual measured value of heat released from the Reactor Pressure Vessel (RPV) or removed by the Vessel Cooling System (VCS) cannot be obtained. It is difficult for operators to evaluate the reactor outlet coolant temperature supplied from the High Temperature Engineering Test Reactor (HTTR) before rise-to-power tests. Therefore, when the actual measured value of heat released from the RPV or removed by the VCS are changed during rise-to-power tests, operators need to evaluate quickly, within a few minutes, the heat removed by the VCS and the reactor outlet coolant temperature of 30 (MW), at the 100% of the reactor power, before the temperature achieves to 967 ($$^{circ}$$C) which is the maximum temperature limit generating the reactor scram. In this paper, a rapid evaluation method for use by operators is presented.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Journal Articles

Master plan and current status for feasibility study on thermal/hydraulic performance of reduced-moderation water reactor

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Misawa, Takeharu; Akimoto, Hajime

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe some advances on experimental and analytical studies. Steady-state and transient critical power experiments have been conducted with two 37-rod bundle test facilities (Gap width between rods: 1.3mm and 1.0mm) and the experimental data reveal the feasibility of RMWR.

Journal Articles

Advances of study on thermal/hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Akimoto, Hajime

Nihon Kikai Gakkai 2005-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.207 - 208, 2005/09

We started R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration and the high void fraction. This presentation shows the advances of thermal/hydraulic feasibility study using large-scale test facility and advanced numerical simulation technology.

Journal Articles

Advances of study on thermal/hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Akimoto, Hajime

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.3mm and 1.0mm) and the experimental data reveal the feasibility of RMWR.

JAEA Reports

Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.

JAERI-Research 2004-008, 383 Pages, 2004/06

JAERI-Research-2004-008.pdf:21.49MB

The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.

Journal Articles

Development of predictable technology for thermal/hydraulic performance of reduced-moderation water reactors, 1; Master Plan

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime

Proceedings of 2004 International Congress on Advances in Nuclear Power Plants (ICAPP '04), p.1488 - 1494, 2004/06

We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR. This series presentation focuses on the feasibility study and shows the R&D plan using large-scale test facility and advanced numerical simulation technology.

JAEA Reports

Nuclear Energy System Department annual report; April 1, 2002 - March 31, 2003

Department of Nuclear Energy System

JAERI-Review 2003-023, 232 Pages, 2003/09

JAERI-Review-2003-023.pdf:16.58MB

The Department has carried out researches and developments (R&Ds) of innovative nuclear energy system and their related fundamental technologies to ensure the long-term energy supply in Japan. The report deals with the R&Ds of an innovative water reactor, called Reduced-Moderation Water Reactor (RMWR), which has the capability of multiple recycling and breeding of plutonium using light water reactor technologies. In addition, as basic studies and fundamental researches of nuclear energy system in general, described are intensive researches in the fields of reactor physics, thermal-hydraulics, nuclear data, nuclear fuels, and materials. These activities are essential not only for the R&Ds of innovative nuclear energy systems but also for the improvement of safety and reliability of current nuclear energy systems. The maintenance and operation of reactor engineering facilities belonging to the Department support experimental activities.

JAEA Reports

Nuclear Energy System Department annual report

Department of Nuclear Energy System

JAERI-Review 2003-004, 236 Pages, 2003/03

JAERI-Review-2003-004.pdf:16.34MB

This report summarizes the research and development activities in the Department of Nuclear Energy System during the fiscal year of 2001 (April 1, 2001 - March 31, 2002). The Department has been organized from April 1998. The main research activity is aimed to build the basis of the development of future nuclear energy systems. The research activities of the Department cover basic nuclear data evaluation, conceptual design of a reduced-moderation water reactor, reactor physics experiments and development of the reactor analysis codes, experiment and analysis of thermal-hydrodynamics, energy system analysis and assessment, development of advanced materials for a reactor, lifetime reliability assessment on structural material, development of advanced nuclear fuel, design of a marine reactor and the research for a nuclear ship system. The maintenance and operation of reactor engineering facilities belonging to the Department are undertaken. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report.

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