Upgrade of the Sub-channel Thermal-hydraulic Analysis Code COBRA-EN for Super-critical Water Reactors

Super-Critical Water Reactors (SCWRs) have been outlined as a possible option for next generation nuclear reactor technology. Fuel assembly design represents a crucial aspect for the success of this concept. Large density variations, low moderation, heat transfer enhancement and deterioration, all play a key role in the core design parameters. Current qualified sub-channel thermal-hydraulic analysis codes are not able to run under super-critical water conditions. At JRC-IE the sub-channel thermal-hydraulic analysis code COBRA-EN has been upgraded to work above the critical pressure. The water properties package of the code is based on the IAPWS Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam. The heat transfer and pressure drop correlations for the super-critical region have also been integrated. As part of the code assessment, both a hexagonal and a square fuel assembly configuration have been analysed. The code has also been applied, coupled with MCNP, to investigate the impact of the use of hydride fuel in super-critical fuel assembly. Analyses performed included steady-state density distribution, pressure drop, axial and radial coolant and fuel rod temperatures. COBRA-EN effectively captured the trends seen in similar studies with acceptable accuracy. Future activities will include the implementation of new features (wall, wire wrap model) and the validation of the code against experimental data.

AMMIRABILE Luca;  TSIGE-TAMIRAT Haileyesus; 

2009-11-13

Atomic Energy Society of Japan

JRC52439

https://publications.jrc.ec.europa.eu/repository/handle/JRC52439,