Consistent simulation of thermal and neutronic parameters in a core of supercritical water reactor

The paper deals with extended validation of the FRC-SCP module and its coupling to the neutron transport code KIR. The software module FRC-SCP performs thermohydraulic simulation of a coolant under supercritical pressure (SCP) and a cooled fuel rod. The software package KIR solves accurately steady-state and time-dependent neutron transport equations using Monte Carlo method. SCP coolant under pseudophase transition exhibits specific behavior as its density and heat transfer coefficient change significantly. Existing feedback dependencies on coolant density and fuel temperature are important for the nuclear safety analysis of the reactor. To solve thermohydraulic and neutron transport problems consistently the codes FRC-SCP and KIR are combined into the KIR-TH complex. We discuss various iteration procedures to solve the coupled problems. Iterations involve heat generation profile as well as thermal state of a fuel rod and a coolant including its density. Test calculations for a fuel rod cooled by SCP water show that the KIR-TH complex is capable of coupled thermal and neutron transport simulation. The paper also covers the Grass heat transfer correlation which is built up-on physical considerations and its application to heat transfer analysis. We present the extended validation against experiments on the heat transfer to SCP water in heated tubes for various thermal parameters. It shows a good agreement between calculations and measurements under normal heat transfer conditions when Kurganov, Deev, or Grass correlations are used. The study reveals the possibility of consistent thermal and neutron simulation for SCP water reactor including the conditions of pseudophase transition.