Three-Dimensional Analyses of Fluid Flow and Heat Transfer for Moderator Integrity Assessment in PHWR

Abstract

A CANDU reactor has the unique features and the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors. If there is the loss of coolant accident (LOCA) and a coincid ent failu re of the emergency coolant injection (ECI) system, the heavy water mod erator is continuously cooled, providing a heat sink for decay heat produced in the fuel. Therefore, it is one of major concerns to estimate the local subcooling of moderator inside the caland ria vessel under postulated accident in CAN DU safety analyses. The Canadian Nuclear Safety Commission (CNSC), a regulatory body in Canada, categorized the integrity of moderator as a generic safety issue and recommended that a series of experimental works be performed to verify the safety evaluation codes for individual simulated condition of nuclear power plant, comparing with the results of three-dimensional experimental data. In this study, three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL (Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flow rate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. Three flow patterns have been identified in the moderator with flow rate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant.