Numerical Modeling of Delayed-Neutron Precursor Transport in a Sodium-Cooled Fast Reactor

Methods of determining the efficiency of the system that controls the seal-tightness of fuel-rod cladding and localizes FA with leaky fuel rods in a fast reactor are examined. It is shown that the design procedure has significant limitations. A procedure for numerical modeling of the transport of delayed-neutron precursors was developed to take account of the special features of liquid-metal coolant flow. A special computational module FV-BN was developed within the framework of the FlowVision software package. The computational results obtained for the concentration distribution of delayed-neutron precursors are transferred into the deterministic transport code TORT in order to obtain the spatial-energy distribution of the neutron flux density in a three-dimensional geometry. The procedure was verified on full-scale reactor problems by simulating the flow-through parts of the upper mixing chamber of the fast reactor.