Neutron kinetics modelling for simulations of loss of coolant accidents in the nuclear power plants

Abstract

The code used by Framatome to predict the progression of a LOCA is the system scale thermal-hydraulic code CATHARE (Code Avancé de THermoHydraulique pour les Accidents deRéacteurs à Eau). It calculates the full primary and secondary circuits including the core and the fuel elements. CATHARE currently utilizes a 0D neutronic model that solves the PointKinetics Equation (PKE) to determine the evolution of instantaneous fission power.
This approach is suitable for transients where the core's moderator density changes rapidly in a quasi-uniform manner, such as in large break LOCA scenarios (from the initial situation witha core full of liquid to a sudden and complete voiding leading to a full vapour environment). However, during Intermediate Break (IB) scenarios, with slower dynamics, the assumption of auniform core's moderator density is no longer valid. This assumption results in a significant underestimation of void antireactivity in the upper part of the core and a slight overestimation atthe bottom. Thus, using PKE for IB-LOCA leads to an overestimation of the fission power, and the more heterogeneous the core, the higher the conservatism of this hypothesis is expected.
In fact, the specific application of IB-LOCA involves a precise neutronic calculation in a strongly diphasic fluid environment which is first due to the uncovering of the core and then to itsreflooding by the safety injections. The presented work relies on the development of a fine 3D coupling between neutronics and thermal-hydraulics at the assembly scale plunged into a fullreactor simulation. Such a development goes beyond the known limitations of current neutronics/thermal-hydraulics couplings (dealing with low void fraction situations) which are notsuitable for LOCA safety studies.