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Journal Articles

10.2.2 Outline of fast reactor and development status in the world

Kamide, Hideki

Genshiryoku No Ima To Ashita, p.265 - 268, 2019/03

no abstracts in English

Journal Articles

The Development status of Generation IV reactor systems, 7; Sodium-cooled Fast Reactor (SFR)

Kamide, Hideki; Ito, Takaya*; Kotake, Shoji*

Nippon Genshiryoku Gakkai-Shi, 60(9), p.562 - 566, 2018/09

Sodium-cooled Fast Reactors (SFRs) have significant characteristics on sustainability as follows, highly effective utilization of Uranium resource, burning of TRU long-life nuclide, e.g., Plutonium, reduction of volume and toxicity of high level radioactive waste of spent fuels. SFRs are one of promising concepts at a step of demonstration phase of development. BN-800 in Russia has already started commercial operation. This is a great step toward the commercialization of SFRs. Russia stated that SFR entered the step of commercial use and next step was demonstration of safety and economy of SFRS by means of operation of BN-1200. Construction of a demo reactor of 600 MWe started in China. In India, operation of PFBR is planned near future and also constructions of 6 units of commercial reactors are also planned. In this report such SFR development plans of oversea countries are summarized including development status and future direction in Japan,

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

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

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

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

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Journal Articles

Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

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

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.

Journal Articles

Influence of inlet velocity condition on unsteady flow characteristics in piping with a short elbow under a high-Reynolds-number condition

Ono, Ayako; Tanaka, Masaaki; Kobayashi, Jun; Kamide, Hideki

Mechanical Engineering Journal (Internet), 4(1), p.16-00217_1 - 16-00217_15, 2017/02

In the design of the Advanced Sodium-cooled Fast Reactor in Japan, the Reynolds number in the primary hot leg (H/L) piping reaches 4.2$$times$$10$$^{7}$$. Furthermore, a short elbow is used in the H/L piping to achieve a compact plant layout. In the H/L piping, flow-induced vibration is a concern due to the excitation force caused by pressure fluctuation in the short elbow. In this report, the influence of inlet velocity condition on the unsteady velocity characteristics in the short elbow was studied by controlling the flow patterns at the elbow inlet. Measured velocity distributions indicated that the inlet velocity profiles affected a circumferential secondary flow, which then affected an area of flow separation at the elbow. It was also found that the velocity fluctuation at low frequency components observed upstream of the elbow could remain in downstream of the elbow though its intensity was attenuated.

Journal Articles

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Nuclear Engineering and Design, 312, p.30 - 41, 2017/02

 Times Cited Count:1 Percentile:64.68(Nuclear Science & Technology)

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Journal Articles

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor; Influence of flow collector of backup CR guide tube

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 5 Pages, 2016/11

JAEA has been conducting a design study for an advanced large-scale sodium-cooled fast reactor (SFR). Hot sodium from the fuel subassembly can mix with the cold sodium from the control rod (CR) channel at the bottom of Upper Internal Structure (UIS). Temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of UIS. JAEA had performed a water experiment to examine countermeasures for the significant temperature fluctuation generated at the bottom of SFRs UIS. Meanwhile, a self-actuated shutdown system (SASS) is equipped in a backup control rod (BCR) channel to ensure reactor shutdown. The BCR guide tubes have a flow guide structure "flow-collector" to provide reliable operation of SASS. Flow-collector may affect the thermal mixing behavior at the bottom of the UIS. This study has investigated the influence of the flow- collector on characteristics of the temperature fluctuation around the BCR channels.

Journal Articles

Development of core hot spot evaluation method of a loop type fast reactor equipped with natural circulation decay heat removal system

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

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

A natural circulation decay heat removal system is adopted in the design of an advanced loop type fast reactor in Japan. For the core structural integrity, we have developed a new evaluation method for the core hot spot temperature during natural circulation decay heat removal operations. In the method, safety analyses are performed with the plant dynamics models that can consider characteristic thermal-hydraulic phenomena under natural circulation conditions. In addition, the core hot spot temperature is estimated with its uncertainty quantified in the statistical manner. This paper describes the evaluation method and also the application results to a loss of offsite power event.

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:2 Percentile:57.25(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

Influence of fluid viscosity on vortex cavitation at a suction pipe inlet

Ezure, Toshiki; Ito, Kei; Kameyama, Yuri*; Kamide, Hideki; Kunugi, Tomoaki*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 15(3), p.151 - 158, 2016/09

no abstracts in English

Journal Articles

Study on behavior of vortex cavitation around suction pipes in sodium-cooled fast reactor geometry

Ezure, Toshiki; Ito, Kei; Kamide, Hideki; Kunugi, Tomoaki*

Thermal Science and Engineering, 24(3), p.31 - 38, 2016/07

Journal Articles

A Study on the thermal-hydraulics in the damaged subassemblies under the operation of decay heat removal system

Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04

Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.

Journal Articles

Benchmark analysis of EBR-II shutdown heat removal test-17 using of plant dynamics analysis code and subchannel analysis code

Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04

Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.

Journal Articles

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.8141 - 8155, 2015/08

In this paper, the authors focus on four kinds of thermal-hydraulic issues associated with the SDC, i.e. fuel assembly thermal-hydraulics, natural circulation decay heat removal, thermal striping phenomena, and core disruptive accidents, and provide a description of their evaluation method developments including verification and validation and necessary experimental studies for the Japan Sodium-cooled Fast Reactor (JSFR). These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all phenomena envisioned in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing down of knowledge/technologies.

Journal Articles

Proposal of benchmark problem of thermal striping phenomena in planar triple parallel jets tests for fundamental code validation in sodium-cooled fast reactor development

Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08

Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.

Journal Articles

Validation of plant dynamics analysis code using shutdown heat removal test-17 performed at the EBR-II

Ohira, Hiroaki; Doda, Norihiro; Kamide, Hideki; Iwasaki, Takashi*; Minami, Masaki*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2585 - 2592, 2015/05

IAEA's Coordinated Research Project on Benchmark Analyses of Shutdown Heat Removal Test (SHRT) performed at the Experimental Breeder Reactor-II (EBR-II) has been carried out since 2012. The benchmark specifications were provided by the Argonne National Laboratory (ANL) and the model development for thermal-hydraulics codes and/or plant dynamics codes has been conducted by participating organizations. The experimental data were also provided by the ANL after the calculations have been performed as the blind simulation. JAEA participated in this benchmark analyses, and the plant dynamics analysis code; Super-COPD was applied to the SHRT-17 simulation. The calculated inlet temperature of the high pressure plenum agreed well with the test data in all simulation time. Although the Z-pipe inlet temperature and the IHX intermediate outlet temperature had some discrepancy in the first 400 sec. caused by larger mass flow rate of the primary pump and the perfect mixing model of upper plenum, these temperatures and the flow rate agreed well with the measured data after 400 sec. Hence it was concluded the present analytical model could predict the natural circulation in good accuracy.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 1; Overview

Kamide, Hideki; Ando, Masato*; Ito, Takaya*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

JAEA, JAPC and MFBR have been conducted design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lesson learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, in the frame work of generation IV international forum (GIF), the design study is focusing on the design measures against sever external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated.

Journal Articles

Water experiments on thermal striping in reactor vessel of Japan Sodium-cooled Fast Reactor; Countermeasures for significant temperature fluctuation generation

Kobayashi, Jun; Ezure, Toshiki; Kamide, Hideki; Oyama, Kazuhiro*; Watanabe, Osamu*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. High cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60$$^{circ}$$ sector model of the reactor upper plenum. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP.

365 (Records 1-20 displayed on this page)