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Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
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. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition 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 an appropriate numerical models for the direct heat exchanger (DHX).
PNC TN1100 93-008, 26 Pages, 1992/12
FBR development activities in Japan have been performed by the government in cooperion with private enterprises. The prototype reactor "MONJU" is now undergoing functnal testing, and the first demonstration plant is in the conceptual design stage. R for commercial plants has been conducted for several years. Commercial plants are quired to be superior to LWRs with regard to economy, safety, and reliability. Accoingly, the Power Reactor & Nuclear Fuel Development Corporation and the Japan Atomicower Company set up specific R&D goals for commercialization, identified plant conces, and planned the necessary R&D activities. In order to make the concepts a realit both government and private enterprises must play a roll in developing and demonstring FBR technologies through construction and operation of prototype and demonstrati plants. In addition, they must perform FBR optimization activities, such as an enhced safety core and a passive decay heat removal system etc., according to lo
Tanaka, Masaaki; Amano, Katsunori*; Ishikawa, Nobuyuki; Nabeshima, Kunihiko; Ohshima, Hiroyuki; Oyama, Kazuhiro*; Nakamura, Hironori*; Ichihara, Takashi*
no journal, ,
In the investigation of measures to enhance safety of a sodium cooled fast reactor, conceptual design of a test apparatus for sodium experiment named AtheNa-RV/DHRS has been conducted at JAEA in corporation with MFBR for feasibility study of the diverse systems for decay heat removal under various operating conditions including the sever accident. Important thermal-hydraulic problems to be studied using AtheNa-RV/DHRS and significant points in the design of it were investigated based on knowledge from the existing investigations regarding decay heat removal systems and results of the preliminary numerical simulation of AtheNa-RV/DHRS in temporal design.
Tanaka, Masaaki; Ezure, Toshiki; Ishikawa, Nobuyuki; Nabeshima, Kunihiko; Oyama, Kazuhiro*; Nakamura, Hironori*; Ichihara, Takashi*
no journal, ,
In the investigation of measures to enhance safety of a sodium cooled fast reactor, conceptual design of a test apparatus for sodium experiment named AtheNa-RV/DHRS has been conducted for feasibility study of the diverse systems for decay heat removal under various operating conditions including the sever accident. According to the knowledges of important thermal-hydraulic problems to be studied using AtheNa-RV/DHRS and significant points in the design of it, major components consisting of the simulated core and the direct heat exchanger (DHX) was preliminary designed.
Tanaka, Masaaki; Ezure, Toshiki; Ishikawa, Nobuyuki; Miyakoshi, Hiroyuki; Shimizu, Ryo*; Nakamura, Hironori*; Oyama, Kazuhiro*
no journal, ,
Conceptual design of a sodium test facility AtheNa-RV/DHRS has been conducted for feasibility study of the diverse systems in decay heat removal under various operating conditions including severe accident. In accordance with the requirements for an integrated effect test facility, sodium test loop layout and the reactor vessel with in-vessel direct heat exchangers are proposed.
Tanaka, Masaaki; Kikuchi, Norihiro; Hamase, Erina; Murakami, Satoshi*; Fujisaki, Tatsuya*; Imai, Yasutomo*
no journal, ,
For safety enhancement of sodium-cooled fast reactor, 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 method. The numerical estimation method which can predict thermal hydraulic phenomena in the reactor vessel during DHRS operation is necessarily required. In this study, appropriate modeling to predict the inter wrapper flow which is caused in the gap between fuel subassemblies is preliminary investigated. The validation of the numerical results is carried out in comparison with the measured temperature data in the scaled sodium test facility named PLANDTL-1. Through the simulation, potential applicability of the gap model using the correlation equations to simulate the thermal hydraulics behavior of the inter-wrapper flow in the core is indicated.
Tanaka, Masaaki; Kikuchi, Norihiro; Doda, Norihiro; Hamase, Erina; Imai, Yasutomo*
no journal, ,
During the decay heat removal operation including accident management in sodium-cooled fast reactor, appearance of the recirculation flow in the vertical direction with the upward and the downward flows due to the buoyancy and the gravity forces respectively is indicated in the fuel subassemblies. In this paper, the numerical simulation of the sodium test for a grid-spacer type fuel subassembly with a planer blockage at force convection condition was carried out. From the comparison of the temperature distribution between the measurement and the numerical results by ASFRE, potential capability of ASFRE to apply the recirculation flow region was indicated.
Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
no journal, ,
For safety enhancement of sodium-cooled fast reactor, 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 methods. The numerical estimation method which can predict thermal hydraulic phenomena in the reactor vessel during DHRS operation is necessarily required. In this study, the numerical analysis of a sodium test for decay heat removal conducted at the scaled sodium experimental facility named PLANDTL-2 was implemented to validate the RV-CFD model.
