Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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.
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.
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.
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：11 Percentile：71.62(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.
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：13 Percentile：73.22(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.
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
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.
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.
Hayakawa, Satoshi*; Watanabe, Osamu*; Ito, Kei; Yamamoto, Tomohiko
Nihon 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.
Hagiwara, Hiroyuki; Yamada, Yumi*; Eto, Masao*; Oyama, Kazuhiro*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
The self-actuated shutdown system (SASS), which is selected for Japan Sodium-cooled Fast Reactor (JSFR), is a passive reactor shutdown system utilizing a Curie point electromagnet (CPEM). With CPEM, an excessive fuel outlet temperature rise is sensed and the control rods are released into the core, and the reactor can be shutdown. Therefore it is important for feasibility of SASS to be established by assuring a quick response of CPEM to the coolant temperature rise. In this paper, a device named "flow collector", which collects flows discharged from six fuel subassemblies surrounding CPEM backup control rods, has been proposed to ensure a shorter response time.
Hayakawa, Satoshi*; Ishikura, Shuichi*; Watanabe, Osamu*; Kaneko, Tetsuya*; Yamano, Hidemasa; Tanaka, Masaaki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
The present methodology was applied to the analysis for the 1/3-scale experiment of the hot-leg pipe of JSFR, and the predicted stress values were compared with the measured stress values. The predicted stress values were underestimated in the case of using the intact pressure fluctuations obtained by the unsteady fluid flow analysis. Therefore, the improvement of the prediction accuracy of the pressure fluctuations on the pipe wall was attempted.
Ono, Ayako; Kobayashi, Jun; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12
Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor. The DHRS of JSFR consists of one unit of DRACS and two units of PRACS. In this study, the sodium experiments were conducted using a sodium test loop PLANDTL in order to investigate the effect of operation mode on transient behavior of thermal hydraulic in PRACS loop. The experimental results revealed the effect of increasing heat removal capacity of PRACS and the forced flow operation in PRACS loop on the thermal transient and natural circulation behavior in PRACS loop.
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.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Nihon Kikai Gakkai Rombunshu, B, 78(787), p.465 - 467, 2012/03
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) for JSFR under natural circulation decay heat removal operations and the sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the evaluation method.
Hamada, Noriaki*; Shiina, Koji*; Fujimata, Kazuhiro*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In a sodium-cooled fast reactor, a vortex cavitation evaluation methodology was developed to predict a possible cavitation generated by vortex at the center of accelerating flow. This methodology was applied to a scaled model experiment, leading to the prospect that the cavitation can be predicted.
Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.
Yamano, Hidemasa; Tanaka, Masaaki; Kimura, Nobuyuki; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Nuclear Engineering and Design, 241(11), p.4464 - 4475, 2011/11
Times Cited Count：19 Percentile：80.18(Nuclear Science & Technology)This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in JSFR, in particular emphasizing on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. Experimental efforts have been made using 1/3-scale and 1/10-sca1e single elbow test sections for the hot-leg piping and 1/4-scale and 1/7-scale triple-elbow test sections for the cold-leg piping. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced the pressure fluctuations onto the pipe. Simulation activities include Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) and Large Eddy Simulation (LES) approaches. Numerical results using the U-RANS approach appear in this paper, indicating its applicability to the hot-leg piping experiments.
Kobayashi, Jun; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki; Watanabe, Osamu*; Oyama, Kazuhiro*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 10 Pages, 2011/10
Design study of an advanced loop-type sodium-cooled fast reactor, JSFR, has been carried out in a frame work of Fast Reactor Cycle Technology Development Project (FaCT) in Japan. As the temperature differences among the control rod channels, blanket assemblies and the core fuel assemblies are 100C centigrade in the maximum, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). In this investigation, a water experiment was conducted using a 1/3 scale 60 sector model of the core and reactor upper plenum. Characteristics of temperature fluctuations near the cold fluid outlets were obtained and it was confirmed that several countermeasures can reduce temperature fluctuations at the bottom of UIS.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/10
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) for JSFR under natural circulation decay heat removal operations and the sensitivity analyses of the uncertainties in the event of loss of external power as the first step to improve the evaluation method.
Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Watanabe, Osamu*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 16 Pages, 2011/09
Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan sodium cooled fast reactor (JSFR). Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. Influences of the pressure loss coefficient in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS due to recovery of natural circulation head via the increase of temperature difference in each loop.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed, in which fully natural circulation system is adopted as the decay heat removal system. A new evaluation method of core hot spot which can be applied to natural circulation decay heat removal has been developed. The new method consists of three-step thermal hydraulics analyses in order to consider the effects of physical phenomena particular to natural circulation, such as inter-fuel-assembly heat transfer and flow redistribution in the core due to buoyancy force. From the viewpoint of calculation cost reduction, we have also developed a simplified model substituting for the third step analysis (subchannel analysis). The new method was applied to the evaluations of a loss-of-external-power event and of a sodium leakage accident in a secondary loop of a large scale reactor.
