AREVA NP's Advanced Thermal Hydraulic Methods for Reactor Core and Fuel Assembly Design

摘要

The main objective of the Thermal Hydraulic (TH) analysis of reactor core and fuel assembly design is the determination of pressure loss and critical heat flux (CHF). Especially the description of the latter effect with corputer codes requires the modeling of a large variety of physical phenomena (I - phase and 2 - phase flow) over a wide range of geometrical scales given by the fuel assembly (length of the assembly: 4m) and its components as well as by the physical proecesses (few mm or sub mm). To cover the full range of needed TH analysis different codes are used and developed at AREVA NP. The main work is performed with sub - channel and computational fluid dynamics (CFD) codes. Successful code development is strongly dependent on the availability of high quality data of relevant flow conditions. For this reason, AREVA NP operates a unique fleet of world wide TH test facilities providing high quality data in the 1 - phase and 2-phase domain for sub - channel codes, as well as for CFD development. This paper gives an overview of recent advances in the TH methods and codes development of AREVA NP. Some applications which have been made possible by these improvements will be shown to underline their value for reactor operation of FA design. It is the ain of presenting the efforts undertaken by AREVA NP to update and to improve its methods and codes with the objective of being able to provide more comprehensiveTH analysis.