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MODELLING OF HEAT TRANSFER PROCESSES IN FUEL CHANNELS OF RBMK-TYPE REACTORS

DOI: 10.1615/ICHMT.2008.CHT.940
17 pages

Eugenijus Uspuras
Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas, Lithuania

Mindaugas Vaisnoras
Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, 3 Breslaujos str., LT-44403, Kaunas, Lithuania

Algirdas Kaliatka
Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas, Lithuania

Sinopsis

Correct evaluation of heat transfer processes plays an extremely important role in the analysis of accident consequences in nuclear power plants. The inadequacy between heat generation in the fuel rods and heat sink leads to the overheating of the core structures in case of reactor accidents. The main phenomenon, occurring in this case is the transition from nucleate coolant boiling to film boiling - Critical Heat Flux. In this paper the issues related to the modelling of CHF and post-CHF heat transfer in RBMK-1500 reactor, using RELAP5 Mod3.2 code, are discussed. The validation process of CHF phenomenon for RBMK-1500 on available experimental investigations, performed in Russian research and design organizations, is presented. The comparison results show that the CHF values calculated with RELAP5 are slightly lower than the experimental ones, if coolant quality is low. It shows that RELAP5 Mod3.2 calculations give a more conservative view. However, in the case of high coolant quality (x > 0.6), the standard approach for CHF calculation can be used since in this region the RELAP5 Mod3.2 code gives a reasonable agreement with the experimental data. Also, the analysis of real fuel channel rupture event in the Leningrad NPP Unit 3 using the RELAP/SCDAPSIM code is presented. The comparison of calculation results and results by other authors shows that the main physical phenomena are adequately modelled, and the developed models are suitable for the analysis of processes in fuel channels and reactor cooling system.

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