Thermalhydraulics and Neutronics Studies on the FP7 CP-ESFR Oxide and Carbide Cores
NURETH-14 - 2011 September 25-30


Presented at:
NURETH-14
2011 September 25-30
Location:
Toronto, Canada
Session Title:
D8-3 Multi-scale multi-physics couplings; Multiphysics applications to new generation reactors

Authors:
Luca Ammirabile (European Commission, JRC, Institute for Energy)
Haileyesus Tsige-Tamirat (European Commission, JRC, Institute for Energy)
  

Abstract

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. 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 ESAP models applied in the study together with the relevant results for the oxide and carbide core.

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