Analysis of Fuel Rod Behavior Under Steady State and Transient Condition by Modification of THEATRe Code

Abstract

The consistent and precise safety evaluation is important to ensure the reactor safety in accidents and transients. For that purpose, analysis is to be performed by applying the important nuclear fields simultaneously. Various nuclear codes have been developed for safety assessment of nuclear power plant. Generally, these codes are designed for some specific analysis such as thermal hydraulic, reactor kinetics, fuel performance, etc. The code designed for specific analysis, normally, approximate the effect of other disciplines which may not provide more consistent and accurate results. This study is related to the modification of THEATRe code by incorporating the new fuel behavior models of FRAPTRAN code. The THEATRe code considers the original fuel pin dimensions in normal, transient, and accident situation. The modified version of THEATRe is capable to simulate the thermal behavior of the fuel, cladding, and gas-gap according to the reactor operating conditions and during accidents/transients including fuel with different burnups. The AP1000 reactor is considered for comparison of fuel rod parameters under steady state and transient conditions. When Reactivity Initiated Accident (RIA) is simulated for transient analysis of fuel rod behavior, the results show that the centerline fuel temperature is lower in modified THEATRe code because of better heat transfer from the reduced gap width. The RIA simulation results illustrate that the modified THEATRe code provides slightly lower increase in fuel rod centerline temperature. Modified THEATRe code can estimate the change in fuel pin gap width in case of normal operation and also in transient conditions which is relevant to the realistic calculations of fuel rod parameters in a thermal hydraulic code.