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A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters. [ABSTRACT FROM AUTHOR]

The core cycle length is crucial parameter for effective and economic operation of a nuclear power plant. Nuclear industry has great emphasis on improving the feasibility and sustainability of nuclear power generation. The cost of the electricity can be reduced by increasing the cycle length without violating the safety margins. Various methods and strategies have been examined to extend a Pressurized Water Reactor (PWR) core cycle length with optimized loading patterns. The objective of this work is to analyse the feasibility of the system to extend the cycle length of Chashma Nuclear Power Generating Station (CNPGS) unit-3, currently operating in Pakistan from 12 months to ∼18 months. To achieve the objective, fuel enrichment is increased from 3.4% to 4.1% with Gd 2 O 3 as burnable poisons to suppress the excess reactivity. The multi-cycle analyses showed promising results and the core cycle length has been extended to ∼18-month for CNPGS unit-3 keeping power peaking factor, temperature coefficients and discharge burnup within design limits. The method can be confidently applied at any cycle of the power plant. Additionally, fuel and moderator temperatures at average and hot channels at beginning of life (BOL) and the end of life (EOL) are predicted using the 1-D steady state thermal hydraulic analysis. The peak to average temperatures at BOL and EOL are found well below the reference value (1.6875). • The cycle length of CNPGS unit-3 is extended from 12 months to ∼18 months. • The fuel enrichment is increased from 3.4% to 4.1% and Gd 2 O 3 is used as the burnable poison. • The multi-cycle analyses is performed using DRAGON5 and DONJON5 computer codes. • Various neutronic parameters are calculated under various core conditions and kept within design limits. • The 1-D steady state thermal hydraulic analysis is performed at average and hot channels at BOL and EOL. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]