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Conference Article: Codes for 3-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions

Buksha, Yu.K.; Marinenko, E.E.; Touzov, A.A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Abstract

For the analysis of transient and emergency processes during reactor operation it is necessary to have a set of codes, which calculate physical processes with a various degree of accuracy. Codes CORT and BUMT for three-dimensional thermohydraulic calculation of fast reactor core in steady stae, transient and accident conditions are described in this paper. The code CORT calculates thermohydraulics of the whole fast reactor core or group of subassemblies in simplified approximation. The core is described as a set of coupled one-dimensional channels or is divided into a set of ring zones, each of those is also represented by one subassembly (S/A).

The detailed three-dimensional calculation of particular S/A is carried out by code BUMT. For description of S/A thermohydraulics the authors have chosen so called “subchannel model”. In this model the S/A is split into number of channels exchanging one by one with mass, momentum and energy. The coefficients of inter channel exchange are calculated on the basis of empirical correlations. The subchannel model is supplemented by detailed (two-dimensional in each axial cross-section) calculation of fuel pin and S/A wrapper temperatures. For solution of hydrodynamic equations the full-implicit scheme is used. Code BUMT was verified [10] using experimental data for S/A-simulators and results of calculations obtained by other codes.

These codes when used in complex with neutronic code and first circuit thermohydraulic code could describe in detail the thermal state of coolant and performance of fuel pins and construction elements of reactor during steady and transient states of its operation.

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key words: b codes; c codes; fast reactors; hydrodynamics; mathematical models; reactor cores; risk assessment; steady-state conditions; three-dimensional calculations; transients
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:163-174