Validation of DRAGON5 and DONJON5 computer codes with CNPGS Unit-3 cycle-1 design calculations

Abstract

The necessity to replace outdated analogue nuclear system calculations as well as the desire to enhance and guarantee adequate levels of plant availability and safety have led to an increase in the usage of software based design and simulations of nuclear power plants in recent years. In this study, the computer codes DRAGON5 and DONJON5 are validated using design information for 300MWe Pressurized Water Reactor (PWR). The codes are used to perform design analysis of Unit-3 Cycle-1 of the Chashma Nuclear Power Generation Station (CNPGS). The analysis includes calculations for criticality, radial power distribution, reactivity coefficients, and control rod worth under various core conditions. The results are validated against the measured data. The difference in the critical boron concentration (CBC) for Hot Zero Power (HZP) and Hot Full Power (HFP) at All Rod Out (ARO) conditions are 16 pcm and 20 pcm respectively. It has been observed that, the core radial power distribution is in good agreement with the design data with maximum relative errors of 6.2% and 1.5% at HZP and HFP conditions respectively. The validation confirms that, the codes can be employed with confidence for the design calculations of 300MWe PWRs (CNPGS). Furthermore, the 1-D steady state thermal hydraulic analysis is also performed to predict the temperatures of the fuel and coolant at both the average and hot channels at the beginning of life (BOL) and the end of life (EOL). It is found that the peak to average values of fuel centerline temperature at BOL for average and hot channels are 1.084 and 1.17 respectively, while at EOL for average and hot channels are 1.059 and 1.069 respectively, which are in good agreement with the reference peak to average value of 1.6875.