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Takegami, Hiroaki; Terada, Atsuhiko; Onuki, Kaoru; Inagaki, Yoshiyuki
Nihon Kikai Gakkai 2010-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.165 - 166, 2010/09
The Japan Atomic Energy Agency has been conducting R&D on thermochemical water-splitting Iodine-Sulfur (IS) process for hydrogen production to meet massive demand in the future hydrogen economy. A concept of sulfuric acid decomposer was developed featuring a heat exchanger block made of silicon carbide (SiC). Recent activity has focused on the reliability assessment of SiC block. Although knowing the strength of SiC component is important for the reliability assessment, it is difficult to evaluate a large-scale ceramics structure without destructive test. A novel approach for strength estimation of SiC structure has been proposed. In this study, the validity of proposed estimation method was confirmed by comparison with probability of failure, which is conventional method for reliability assessment of ceramic components. The proposal method was effective for strength estimation of a component of large-scale effective volume.
Yan, X.; Sato, Hiroyuki; Tachibana, Yukio
no journal, ,
This paper describes the design of a modular HTGR system cogenerating electricity and hydrogen and discusses the operational feasibility of the system to load follow with the use of a newly proposed control scheme. The system generates electric power by direct cycle gas turbine and hydrogen by a reactor-heated thermochemical process. The reactor heat is transferred to hydrogen plant through a closed heat transport loop. The electricity generated primarily supplies external grid output while meeting power consumption in the reactor and hydrogen plants. For optimum plant economics, the reactor operates in full thermal power at all time and enables electric load follow by varying hydrogen cogeneration rate. The rapid load follow is permitted by the control scheme that adjusts reactor coolant inventory while keeping key reactor plant temperatures constant.
Tanaka, Masaaki; Fujisaki, Tatsuya*; Ohshima, Hiroyuki; Monji, Hideaki*
no journal, ,
Flow-induced vibration in the short-elbow is an important issue in design study of JSFR, because it may affect to structural integrity of the pipe. In this paper, numerical simulations for 1/3-scale water experiment were conducted to validate an in-house LES code for the short-elbow pipe flow and to investigate unsteady flow behavior through the short-elbow. Numerical results showed good agreement with the experimental results. Unsteady flow characteristics in the short-elbow were clarified in relation to the large-scale eddy motion.
Fujita, Kaoru; Yamano, Hidemasa; Kubo, Shigenobu*; Shimakawa, Yoshio*; Ikarimoto, Iwao*
no journal, ,
A self-actuated shutdown system (SASS) is developed to be adapted to JSFR for enhancing safety against anticipated transient without scram (ATWS). The SASS is a passive device, which detach a control rod responding to abnormal increment of coolant temperature. The control rods for backup shutdown system are hold by the electromagnet and the temperature sensitive alloy, which lose the magnetic nature due to the increase of temperature, is installed in the magnetic circuit of SASS. In this paper, the status of SASS development and the result of the validation analysis against ATWS are reported.
Yamano, Hidemasa; Kubo, Shigenobu*; Shimakawa, Yoshio*
no journal, ,
A conceptual design study and related R&D for a large-scale sodium-cooled fast reactor (JSFR) are carried out in the framework of fast reactor cycle technology development project. As a next generation plant, JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability and safety. This paper describes the safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are passive shutdown system and re-criticality free concept against anticipated transient without scram accident in beyond design-based events. Safety analyses were also performed for representative design-based events: control rods withdrawal, primary pump seizure and loss-of-offsite power accidents. All analytical results showed to meet tentative safety criteria with expected reactor trip signals.
Ezure, Toshiki; Kimura, Nobuyuki; Tobita, Akira; Kamide, Hideki
no journal, ,
Cavitation due to sub-surface vortex is one of the significant issues for the design of Japan Sodium Cooled Fast Reactor. This phenomenon has been well-researched for the pump sumps in the turbo machinery field. However, in those researches for the pump sumps, the working fluid is mainly water as well as in the real plant. Thus, there are few works concerning the influences of physical properties of working fluid. In the present study, a fundamental experiment was performed for the influences of fluid viscosity on the sub-surface vortex. The experiment was carried out in a simple cylindrical water tank geometry. The fluid temperature was varied from 10
C to 80C. The experimental results showed that the onset condition of cavitation due to the sub-surface vortex depended on not only cavitation factor but nondimensional circulation.
Ohshima, Hiroyuki; Imai, Yasutomo*
no journal, ,
A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions including fuel deformation. In this study, we focused on SPIRAL, which is one component code of the numerical simulation system and plays the role to simulate detailed local flow and temperature fields in a wire-wrapped fuel pin bundle, and incorporated a new hybrid turbulence model to improve the pressure drop predictability. SPIRAL with the new turbulence model was applied to numerical simulations of flows in a parallel channel and several types of fuel assemblies for code validation. Pressure loss coefficients estimated by the simulations are in good agreement with theoretical and measured data from laminar to turbulent regions.
Yoshikawa, Ryuji; Ohshima, Hiroyuki
no journal, ,
A computer code to analyze flow stability in water-side of steam generator was developed with drift-flux model. The semi-implicit numerical scheme was used, and the flow equations were differenced over the staggered mesh. The total flow rate and the temperature of water flowing into the inlet plenum, the pressure of outlet plenum, and the heat flux through the heat transfer tube were given as boundary conditions for the simulation of water-side flow. The computer code was validated by the simulation of flow instabilities in two parallel channels and comparison with experimental results. The capability of the computer code on prediction of flow oscillation and stable boundary in two parallel channels was confirmed. The sensitivity analysis was also carried out to quantify the impacts of some important parameters on oscillation period and the stable boundary.