THERMAL HYDRAULIC MIXING IN THE PRIMARY SYSTEM OF A PRESSURIZED WATER REACTOR DURING HIGH PRESSURE SAFETY INJECTION UNDER STAGNATED LOOP CONDITIONS

One of the key safety issues currently addressed by the Pressurized Water Reactor (PWR) industry is the retention of structural integrity of the pressure vessel during certain high pressure overcooling transients popularly called Pressurized Thermal Shock (PTS). Among the various modes by which a reactor can experience PTS, the high pressure safety injection under stagnated loop conditions is the most severe from the standpoint of overcooling the system. The present work uses a two pronged approach to predict the temperature transients in the primary system under these circumstances. The first approach is to characterize the system response experimentally in a 1/2-scale (of a typical 3000 MW(,th) plant) model facility using brine and fresh water as simulant fluids. The mixing is quantified by the measurement of mean and turbulent concentration and velocity at key locations of interest using electrical conductivity and hot wire probes. Secondly, a computer code is developed which solves the system of governing differential equations using finite difference method. The code is validated against a comprehensive data base which includes present facility and three other facilities in varying scales from 1/5 to full scale. Excellent agreement is demonstrated between the code predictions and experimental data. The code is subsequently employed to predict the temperature transients for commercial reactors, which are presently used by the Oak Ridge National Laboratory as the boundary conditions in their fracture mechanics computations