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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

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

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Development of probabilistic risk assessment methodology of decay heat removal function against combination hazard of low temperature and snow for sodium-cooled fast reactors

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08

Journal Articles

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

Hourcade, E.*; Mihara, Takatsugu; Dauphin, A.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.556 - 561, 2018/04

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 an update concerning ASTRID DHR strategy with description of reference architecture evolution and project objectives. In particular, new developments were made for DHR during normal shutdown and role of Ex-Vessel system. A special focus is made on design process of automatic shutter to hydraulically connect Hot Plenum and cold plenum to enhance primary vessel natural convection.

Journal Articles

Numerical analysis of EBR-II shutdown heat removal test-17 using 1D plant dynamic analysis code coupled with 3D CFD code

Doda, Norihiro; Hiyama, Tomoyuki; Tanaka, Masaaki; Ohshima, Hiroyuki; Thomas, J.*; Vilim, R. B.*

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

In sodium-cooled fast reactors, a natural circulation is expected to remove the core decay heat when the plant gets into a station blackout. From a perspective of reactor safety, the core hot spot temperature arising in the natural circulation should be evaluated accurately. To this end, Japan Atomic Energy Agency is trying to couple a 1-D plant dynamics analysis code Super-COPD and a 3-D CFD code AQUA to solve the thermal-hydraulic field in the whole plant under natural circulation condition. As a validation study, the coupled code was applied to an analysis of EBR-II shutdown heat removal test. The obtained numerical results reasonably agreed with the measured data, which demonstrated the validity of the coupled code.

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

Analysis of natural circulation tests in the experimental fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08

Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.

Journal Articles

Analysis for thermal fluid dynamics in downcomer of JAERI passive safety reactor (JPSR)

Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

Proceedings of 3rd JSME/ASME Joint International Conference on Nuclear Engineering, Vol.2, p.1017 - 1022, 1995/00

no abstracts in English

Journal Articles

Thermal fluid flow analysis in downcomer of JAERI passive safety light water reactor (JPSR)

Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

Prog. Nucl. Energy, 29(SUPPL), p.405 - 412, 1995/00

no abstracts in English

Journal Articles

Design and safety consideration in the High-Temperature Engineering Test Reactor(HTTR)

Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

Energy, 16(1-2), p.449 - 458, 1991/00

 Times Cited Count:3 Percentile:52.11(Thermodynamics)

no abstracts in English

Journal Articles

Core meltdown accident analysis for a BWR plant with MARK I type containment

; ; ; *; *

Source Term Evaluation for Accident Conditions, p.733 - 744, 1986/00

no abstracts in English

Oral presentation

Simplified model for highest cladding temperature evaluation in wire-wrapped fuel rod bundle under natural circulation decay heat removal

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

no journal, , 

We have developed a simplified model for Highest Cladding Temperature Evaluation in Wire-wrapped Fuel Rod Bundle under Natural Circulation Decay Heat Removal. We have demonstrated the case of 271-pin fuel assembly in the external power loss event with the simplified model, and the results were in good agreement with the results by using the subchannel analysis code. Some uncertainties in the fuel assembly, e.g. the variations of fuel pin power and fuel pin position, were considered by setting boundary conditions. The computational cost of the simplified model was about one-thousandth compared with the subchannel analysis model. The simplified model can be applied without unrealistic computational cost to the statistical evaluation of safety criterion under natural circulation decay heat removal, which requires a large number of transient analyses.

Oral presentation

New Reactor Cavity Cooling System (RCCS) having passive safety features, 2; Experimental study on heat transfer characteristics using a scale model

Sato, Noriyasu*; Yamaguchi, Shuhei*; Matsumoto, Tatsuya*; Morita, Koji*; Takamatsu, Kuniyoshi

no journal, , 

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Last year, we built the scale-down heat removal test facility and now conduct experiments to understand the heat-transfer characteristics. As a result, we could assess effect of radiation quantitatively for passive decay heat removal from the RPV.

Oral presentation

Study on heat transfer characteristics of reactor cavity cooling system with passive safety features; Experimental investigation using a scaled model

Yamaguchi, Shuhei*; Sato, Noriyasu*; Matsumoto, Tatsuya*; Morita, Koji*; Takamatsu, Kuniyoshi

no journal, , 

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. In 205, we built the scale-down heat removal test facility and now conduct experiments to understand the heat-transfer characteristics. As a result, we could assess effect of radiation quantitatively for passive decay heat removal from the RPV.

Oral presentation

An Experimental investigation on heat transfer characteristics using a scaled model of a reactor cavity cooling system with passive safety features

Yamaguchi, Shuhei*; Sato, Noriyasu*; Matsumoto, Tatsuya*; Morita, Koji*; Takamatsu, Kuniyoshi

no journal, , 

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Last year, we built the scale-down heat removal test facility and now conduct experiments to understand the heat-transfer characteristics. As a result, we could assess effect of radiation quantitatively for passive decay heat removal from the RPV.

Oral presentation

Investigation on heat transfer characteristics of reactor cavity cooling system for HTGR

Hosomi, Seisuke*; Yamaguchi, Shuhei*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

no journal, , 

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. After building the scale-down heat removal test facility, we conducted experiments and analyses to understand the heat-transfer characteristics. As a result, we could assess effect of radiation and natural convection quantitatively for passive decay heat removal from the RPV.

Oral presentation

Fast Reactor development in Japan

Kamide, Hideki

no journal, , 

Current status of fast reactor development in Japan was outlined and also research strategy of fast reactor in JAEA was stated with recent research results in thermal hydraulics, core safety, and nuclear fuel cycle.

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