Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system
- Indira Gandhi Centre for Atomic Research, Kalpakkam (India)
Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Systems Technology; American Nuclear Society (ANS), La Grange Park, IL (United States); American Institute of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers (ASME), New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
- OSTI ID:
- 107783
- Report Number(s):
- NUREG/CP-0142-Vol.2; CONF-950904-Vol.2; ON: TI95017078; TRN: 95:021405
- Resource Relation:
- Conference: 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995; Other Information: PBD: Sep 1995; Related Information: Is Part Of Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11; Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)]; PB: 795 p.
- Country of Publication:
- United States
- Language:
- English
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