SPLOSH II A DYNAMICS PROGRAMME FOR NUCLEAR - THERMAL – HYDRODYNAMIC BEHAVIOUR OF WATER-COOLED REACTORS

摘要

A dynamics code is described that solves the two-group neutron diffusion equations simultaneously with the thermal and the hydraulic equations for an average channel of a water-cooled reactor. Other reactor channels can be represented as 'slaves', which have no feedback to the average channel. The fission power at any axial station in a slave channel is related to that in the average by prescribed time-dependent factors, and the hydraulic flow is determined from pressure-drop requirements dictated by the performance of the average channel. A finite difference model of the fuel element and can represents the behaviour of the fuel temperatures and surface heat flux. The representation of the hydraulic circuit has been made sufficiently general that the code is applicable to B.W.R., P.W.R. and pressure tube reactor designs.