Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki
Dai-24-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2019/06
In the design study on a sodium-cooled fast reactor (SFR), a numerical simulation code named SYRENA has been developed in Japan Atomic Energy Agency to analyze the behavior of gas bubbles and/or dissolved gas in the primary coolant system. In the present study, the effect of the non-condensable gas entrainment at the free surface on the bubble and the dissolved gas behavior in the primary coolant system were investigated for a typical pool type reactor, and also effect of a dipped-plate (D/P) installed below the free surface in the reactor vessel to suppress the gas bubble entrainment into the primary coolant system was especially investigated. It was clarified that the D/P was influential to the non-condensable gas behavior and the molar flow rate of gas bubbles in the primary coolant system varies depending on the relationship between the gas entrainment rate at the free surface and the exchange flow rate through the D/P.
Pshenichnikov, A.; Kurata, Masaki; Nagae, Yuji
Dai-24-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2019/06
Yamano, Hidemasa; Takai, Toshihide
no journal, ,
It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B
C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. To validate a physical model simulating the eutectic melting reaction and relocation, the visualization experiments of SS-BC eutectic reaction was carried out by contacting SS melts of several kg with a BC pellet heated up to about 1500
C. This paper describes chemical analysis results of eutectic materials formed in the experiment, which are served as interpretation of the experiments and code validation.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
no journal, ,
Sodium fire is one of key issues in sodium-cooled fast reactor plant. JAEA has developed sodium fire analysis codes to evaluate the consequence of sodium fire events. This paper describes a PIRT (Phenomena Identification and Ranking Table) process for sodium fire events. Ranking table for important phenomena and an assessment matrix are completed. As a part of comprehensive validation based on the assessment matrix, an experimental analysis for a sodium spray fire experiment shows good agreement with the experimental result.
Emura, Yuki; Isozaki, Mikio; Matsuba, Kenichi; Kamiyama, Kenji
no journal, ,
In core-disruptive accidents of sodium-cooled fast reactors, molten core materials flow down and penetrate into sodium in a lower part of reactor vessel. In this study, molten stainless steel which is one of components of molten core materials was poured into sodium pool and its penetration behavior into sodium was observed using X-ray system.
Ezure, Toshiki; Onojima, Takamitsu; Tanaka, Masaaki; Kameyama, Yuri*
no journal, ,
As a part of safety enhancement of sodium-cooled fast reactors, experiments was carried out using PLANDTL-2 facility to grasp the core cooling behavior only relying on the natural circulation inside the vessel using dipped-type DHX. As the result, temperature distributions in the core were quantitatively grasped under a situation where cold coolant from DHX penetrated the core. Furthermore, it was observed that temperature distribution at the middle height of core became flat due to the flow redistribution among heating channels resulting from the buoyancy balance between channels.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.