A methodology for the estimation of release of fission products during LOCA with loss of ECCS

A Loss of Coolant Accident (LOCA) in a nuclear reactor along with the failure of the Emergency Core Cooling System can cause sustained voiding of the core. In such a situation the core experiences very low flow which leads to poor heat removal from the reactor core. The heat to be removed from the core includes stored heat, heat generated due to metal water reaction at high temperatures, decay heat etc. The poor heat removal leads to heating of the fuel pins to high temperatures. The heating of fuel pins is further enhanced due to metal-water reaction at high temperatures. These high temperatures of the fuel pins can lead to fission product release, which is transported into the Primary Heat Transport (PHT) system and can enter the containment through the break. Analysis is involved due to the complexity of the system and the phenomena to be simulated. In this paper a multistage analysis methodology is presented that involves the development and application of a number of computer programs to model the various phenomena involved. The computer code PHTACT computes the activity release from the fuel as a function of fuel temperatures and cladding oxidation, its distribution into the PHT system and release into the containment. Computation of thermal hydraulic parameters during LOCA is done using the thermal hydraulic analysis code RELAP5. The detailed simulation of fuel pin temperatures is done using computer code HT/MOD4. The convective boundary conditions required for the code are obtained from RELAP5. Creep deformation is considered in the computation of dimensional changes of the coolant channel and estimation of flow blockage due to clad ballooning. The progression of various reaction layers due to high temperature reaction between fuel and clad and clad and steam is also computed, which affects the structural strength of the clad. Different approaches are possible and analysis can be carried out in different phases depending upon the complexities to be considered. Details of an analysis carried out for the proposed Indian Advanced Heavy Water Reactor (AHWR) and some results are presented in the paper. (author)