ANALYSIS OF FUEL THERMAL CONDITION BASED ON COUPLED NEUTRONIC AND THERMAL-PHYSICAL CALCULATIONS
24th Symposium of AER on VVER Reactor Physics and Reactor Safety (2014, Sochi, Russia)
Reactor dynamics and safety analysis
Abstract
ANALYSIS OF FUEL THERMAL CONDITION BASED ON COUPLED NEUTRONIC AND THERMAL-PHYSICAL CALCULATIONS
Artemov V.G., Artemova L.M., Korotayev V.G., Miheyev P.A., Shemayev Yu.P. FSUE “Alexandrov NITI”, Sosnovy Bor, Russia
ABSTRACT
To assess conservatism in design calculations of fuel temperature condition for the VVER-type reactors, a numerical model has been developed and tested, which provides a subchannel calculation of fuel bundles using capabilities of the KORSAR thermal-hydraulic computer code and the SAPFIR_95&RC_VVER neutronic program package. The approach uses, first, results of power density subchannel simulation performed by the method of superposing the microscopic and macroscopic neutron flux distributions, tested on the VVER steady-state neutronic calculations during SAPFIR_95&RC_VVER verification, and, second, subchannel calculation of the thermal-hydraulic parameters of selected fuel bundles carried out with the use of the KORSAR code.
The following procedure is used for calculating accident scenarios in the proposed analysis:
- calculate the scenario in the complete-core approximation by using the KORSAR/GP code with 3D neutron kinetics; determine the hottest fuel bundles using the uncertainty and sensitivity analysis method; record the thermal-hydraulic and neutronic parameters of these fuel bundles as the scenario proceeds;
- reconstruct the subchannel power density distributions in the selected hottest fuel assemblies based on macroscopic neutron fluxes obtained from a dynamic calculation of the scenario and relative microscopic neutron flux distributions in fuel rods previously calculated for the selected fuel bundles;
- calculate the temperature condition of each individual rod using the KORSAR/GP subchannel model of a fuel bundle based on reconstructed subchannel power density distribution and thermal-hydraulic boundary conditions obtained from complete-core dynamic calculation of the scenario.
Validity of the method for reconstructing power distribution in individual fuel rods is verified against results of test calculations performed by programs using the Monte Carlo method. The fuel bundle subchannel model is verified against results of experiments with electrically heated fuel bundles.
The efficiency of the proposed approach is demonstrated by an example simulation of the VVER-1000 steam generator line break scenario