The work is dedicated to the analysis and improvements of methodologies of safety evaluations of Sodium-cooled Fast Reactor (SFR) systems. In particular, the work is built around the SAS-SFR code which is primarily used for the deterministic analysis of the Initiation Phase (IP) of severe accidents in SFRs. It provides comprehensive thermal hydraulics and fuel pin mechanics simulation models for the prediction of steady state characterization of fuel pin configuration during fuel pin burnup cycle and subsequent accident transient behaviour of the entire SFR core including core material melting and relocation phenomena should the fuel pin configuration fail.
Based on existing well-validated computational tools and models for neutron physics, the work aims to improve the existing SAS-SFR capabilities by the application of advanced neutron physics simulation approaches. With regard to the modelling of the IP phenomenology, one important limitation of the currently applied SAS-SFR code is the use of the Point Kinetics (PK) model which considers fixed normalised spatial shape of the core power and neutron flux for the whole range of the ... mehr simulation time. This work aims to overcome this drawback by coupled simulations and application of spatial kinetics neutron physics solution using the PARCS code, which provides spatially dependent power distributions and feedback effects, in particular, for the transient time period characterized by fuel and clad material motion. It is a first of a kind implementation of the spatial kinetics for SAS-SFR which has been practically applied for comprehensive analysis of an Unprotected Loss Of Flow (ULOF) transient of a large commercial SFR core. For the steady state core characterization, a coupled simulation with Monte Carlo neutron physics solution using the MCNP code has been evaluated additionally. Both coupled solutions employ a newly developed methodology for the transfer of the relevant thermal hydraulics core state parameters to the neutron physics evaluation tools, providing an effective basis for comparisons and analysis.
The analysis performed for the SFR core considered in the CP ESFR project and the evaluation of the respective results obtained with the developed numerical tools, demonstrated the extended capability of the improved coupled codes to describe the core behaviour under both steady state and transient conditions with great detail. In addition, the investigations contributed to better understand model limitations and to identify sources of uncertainties related to the complicated calculation routes necessary to be followed during safety evaluations of sodium cooled nuclear reactors.