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PNC TN9410 98-044, 47 Pages, 1998/06
Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast breeder reactors (LMFBRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Therefore the in-vessel components located in the core outlet region, such as upper core structure (UCS), flow guide tube, C/R upper guide tube, etc., must be protected against the stationary random thermal process which might induce high-cycle fatigue. In this study, thermal striping conditions at the tee junction in the MONJU EVST system (maximum temperature difference : 110 
C, Velocity ratio between main and branch pipes : 0.25) were investigated numerically by the use of computer programs. From the investigations, the following results have been obtained: (1) Effects of the secondaly flows generated by the existence of 90 elbow located at upstream position of the tee junction were negligeble, because the flow velocity in the main pipe is 0.25 of the flow velocity in the branch pipe. (2) A ration between maximum and effective amplitudes of the temperature fluctuations calculated by the DINUS-3 code was 3.18. It was concluded that the value 6.0 as the ratio used in the integrity evaluation of the EVST system is a coservative side. (3) There was a limit in ability of a time-averaged multi-dimensional code AQUA, in the evaluation of thermal striping phenomena with recirculation flows. One of the reasons was considered that the local equilibrium of turbulence flows was not established in this tee junction problem.
Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki
no journal, ,
Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure, which was achieved through probabilistic risk assessment for the EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. The safety strategy for the EVST involves whole core refueling-early transfer of all core fuel assemblies into the EVST-assuming a severe situation that results in sodium level reduction leading finally to the top of the reactor core fuel assemblies in a long time. This study introduces the success criteria mitigation along the decay heat decrease over time. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, a probability analysis for human error, and quantification of accident sequences. The fuel damage frequency of the EVST was evaluated to be approx. 10
/year. This study also quantitatively showed the effectiveness of design improvement.
Nakamura, Yoshihide; Sawazaki, Hiromasa; Morioka, Tatsuya; Uchida, Takenobu; Sato, Takeshi; Shiotani, Hiroki; Kisohara, Naoyuki
no journal, ,
With the liquid metal sodium cooled fast reactor, sodium has to be maintained and managed at a high purity to prevent the channel blockage by the precipitation of impurities and the material corrosion with sodium. At Monju, primary heat transport system, secondary heat transport system, and ex-vessel fuel storage tank sodium cooling system are corresponding to this clean-up system. We evaluate the sodium purity control condition of these facilities for more than 20 years by the measurement of the plugging meter and chemical analysis, and confirm that we could maintain the purity with a value lower enough than the management targeted value.
