Calculation of Power Transients for the SPERT II Nuclear Reactor Core with Inclusion of Nucleate Boiling of the Coolant

摘要

The published data for a series of power transients initiated in the water-cooled core of the SPERT II nuclear reactor were investigated to determine the magnitude of the reactivity feedback effects arising from heat transfer and vapour void formation during subcooled nucleate boiling. This was done using the NAIADQ computer code which describes the thermohydraulic behaviour of the water coolant in the fuelled region of the core in terms of a set of one-dimensional linear, finite difference equations. Nucleate boiling at the fuel surface is represented as an expanding superheated layer of water in which saturated vapour is assumed to be uniformly generated at a non-equilibrium rate. The simulation of the transients, all of which were initiated at ambient temperature, necessitated the calculation of heat transfer from the fuel to the water coolant in the regimes of conduction, convection and surface boiling for a wide range of coolant flow rates during rapidly changing reactor power levels. The calculated variations in the reactor power with time during the transients are in good agreement with the measured data over the range of conditions tested in the experimental investigations. (Atomindex citation 14:793336)