Pre-conceptual thermal–hydraulics and neutronics studies on sodium-cooled oxide and carbide cores

Abstract

The Generation IV International Forum (GIF) has among its main goals the excellence in safety and reliability of the proposed innovative nuclear systems. The development of computational tools that are able to assist the design and safety analysis of these innovative reactor concepts is crucial. In this line, the JRC–IET is developing a static and dynamic integrated safety analysis platform with the objective to perform an integrated core and safety analysis of nuclear reactor systems. The first application of this platform has been made in the framework of the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) that is also part of EURATOM contribution to GIF. Two core designs have been currently proposed for the 3600MWth sodium-cooled reactor concept based on oxide and carbide fuel respectively.Using the integrated safety analysis platform, static calculation on neutronics (incl. reactivity coefficients) and thermal–hydraulic characteristics for both oxide and carbide reference cores have been conducted. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the tools and the models applied in the study together with the relevant results for the oxide and carbide core.