Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: The status of studies on fast reactor core thermal hydraulics at PNC

Nishimura, M.; Ohshima, H.; Kamide, H.; Yamaguchi, K.; Yamaguchi, A. (O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, O-arai, Ibaraki-ken (Japan))

Abstract

An outlook was addressed on investigative activities of the fast reactor core thermal-hydraulics at Power Reactor and Nuclear Fuel Development Corporation. Firstly, a computational modeling to predict flow field under natural circulation decay heat removal condition using multi-dimensional codes and its validation were presented. The validation was carried out through calculations of sodium experiments on an inter-subassembly heat transfer, a transient from forced to natural circulation and an inter-wrapper flow. Secondly, experimental and computational studies were expressed on local blockage with porous media in a fuel subassembly. Lastly, information was presented on an advanced computational code based on a subchannel analysis code. The code is under the development andextended to perform whole core simulation.

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key words: after-heat removal; calculation methods; computer codes; fast reactors; heat transfer; many-dimensional calculations; natural convection; pnc; power reactors; reactor cores; simulation; sodium
Reference:
Technical committee meeting on methods and codes for calculations of thermohydraulic parameters for fuel, absorber pins and assemblies of LMFR's with traditional and burner cores. Obninsk (Russian Federation) 27-31 Jul 1998
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--1157, pp:28-45