Subchannel Thermohydraulic Calculations for Fuel Subassembly of Reactor Core on Supercritical Water
UDC: 621.039.5: 536.24.08
Results of thermohydraulic calculations with subchannel codes SUP and MIF-SKD are presented in the paper. Code SUP allows the thermohydraulic parameters of the part of reactor core to be estimated. Input neutron-physical parameters are calculated with code ACADEM. Code MIF-SKD gives us possibility to predict local thermohydraulic parameters in separate fuel subassembly. The code has been verified on the data gained in experiments on water in round tube and on freon in pin bundle.
Thermohydraulic parameters of the part of reactor core and fuel subassembly of VVER-SKD by the power 1700 MW for different coolant flows within reactor core (one- and double-thread) are presented here.