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Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
The objective of this study is to develop probabilistic risk assessment (PRA) methodology for combination event of low temperature and snow by focusing attention on decay heat removal system (DHRS) of sodium-cooled fast reactor. For this combination event, annual excess probability depending on the hazard intensity was statistically estimated based on the meteorological data. Event tree was developed by considering the impact of low temperature and snow on DHRS: e.g., plug at the air intake of ultimate heat sink and of emergency diesel generator due to accumulated snow, failure of air intake filter due to deposited snow, possibility of freezing of cooling circuits. Recovery actions (i.e., snow removal and filter replacement) were considered in the event tree. Quantification of the event tree showed that dominant core damage sequence is loss of access route for snow removal against the combination event at daily snowfall of 3m/day continued during 24h.
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06
no abstracts in English
Hiroi, Hiroshi*; Arai, Masanobu; Kisohara, Naoyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06
The purpose of fast breeder reactors (FBR) and the role of Monju were discussed in Ministry of education, culture, sports science and technology-Japan (MEXT) after the Fukushima NNP accident. The discussion has concluded that FBRs contribute to energy security and reduction of high-level radioactive waste, and that Monju is to be utilized to demonstrate these usefulness and to implement international contributions. This paper addresses anticipated R&D results from Monju on the basis of the enforcement of new nuclear regulation, the energy situations in Japan and the international status of FBR development and collaborations.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.
Nagatake, Taku; Takase, Kazuyuki*; Yoshida, Hiroyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/06
no abstracts in English
Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/06
Under natural circulation decay heat removal conditions, three characteristic phenomena; flow redistribution in the core as well as in the fuel subassemblies, inter-subassembly heat transfer and gap flow between wrapper tubes of fuel subassemblies are important for assessing the temperature distribution in the core. In order to improve the prediction accuracy, a whole core model which can consider these three phenomena has been incorporated into the plant dynamics analysis code Super-COPD. In this study, analyses of two kinds of sodium experiments were performed to validate Super-COPD with the whole core model, which were focusing on inter-subassembly heat transfer phenomena.
Yamano, Hidemasa; Suzuki, Toru; Kamiyama, Kenji; Kudo, Isamu*
no journal, ,
This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B
C) and stainless steel (SS) under a high temperature condition exceeding 1500
C as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the BC-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a BC pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the BC-SS eutectic in upper part of the solidified test piece due to the density separation.
Takaya, Shigeru; Chikazawa, Yoshitaka; Hayashida, Kiichi*; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi*
no journal, ,
In this paper, a maintenance management suitable to nuclear power plants (NPP) at R&D stage was studied. NPP at R&D stage have several features different from commercial NPP. A maintenance management for NPP at R&D stage should be conducted by taking account of those features. First, objectives of maintenance management of NPP at R&D stage was clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed.
Chikazawa, Yoshitaka; Takaya, Shigeru; Hayashida, Kiichi*; Tagawa, Akihiro; Kubo, Shigenobu; Yamashita, Atsushi*
no journal, ,
A maintenance management suitable to nuclear power plants (NPP) at R&D stage was discussed. Objectives of maintenance management of NPP at R&D stage was first clarified. Next, applicability of codes for maintenance management of commercial NPP to NPP at R&D stage was discussed. Then, requirements and consideration for maintenance management of NPP at R&D stage was proposed. In this paper, an example that the proposal was applied to setting maintenance program of sodium-cooled fast reactor was presented.
Mizutani, Jun*; Ebara, Shinji*; Hashizume, Hidetoshi*; Yamano, Hidemasa
no journal, ,
This study evaluated experimentally the effect of influence of the inflow condition upon the flow field in the cold leg piping of the primary cooling system of a sodium-cooled loop-type fast reactor by using a swirling flow generator. Flow visualization with the two-dimensional particle image velocimetry was conducted to reveal the velocity fields of the flow in the piping, especially fluctuating velocity fields, and the results were compared to those obtained from the fully-developed turbulent inflow case. The experiment showed that the there appeared long-period vortices shedding with weak periodicity from the intrados of the 1st elbow instead of steady vortices shedding from flow separation region. In the 2nd elbow, the periodicity of fluctuating velocities were not largely different from those obtained in the fully-developed turbulent inflow case. The downstream of the 2nd elbow was not largely affected by the inflow condition.