Popis: |
To ensure the safe operation of nuclear reactors, a neutronic code is usually used to analyze the neutronics and safety-related parameters of the nuclear reactor. For the neutronic analysis, standardized benchmark problems for well defined nuclear reactor conditions are selected and studied, with the purpose of obtaining qualitative and quantitative measures of the reliability and accuracy of the simulation methods, and cross-section data libraries, respectively. Benchmarking is achieved by comparison with: other experimental data or simulation code results. After the benchmarked simulation method is achieved, the neutronic evaluations are performed with more assurance and accuracy of the results. The aim of this study is to calculate the integral parameters: ⍴28, δ25, δ28 and C* of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) PWR core at hot zero power state using open-source Monte Carlo code OpenMC. The PWR BEAVRS benchmark is based on the, full PWR core design, and operational data for a 4-loop Westinghouse pressurized water reactor. The first fuel cycle at hot zero power state was simulated using the ENDF/B-VII.1 data library and the nuclear data processing code NJOY is used to generate a point wise cross section data. The fission and capture reaction rates of the U-235 and U-238 were obtained for two groups; an epi-thermal and thermal groups with the energy ranges; 100E-3 MeV > Eepi-thermal > 6.25E-7 MeV and 6.25E-7 MeV > Ethermal > 0.0 eV respectively. The results of effective multiplication factor (), control rod worths and pin by pin power distribution in the whole core are compared with experimental data provided with benchmark specification and a good satisfactory result is obtained. It is concluded that this study demonstrates that the Monte Carlo code OpenMC can be used for precise neutronics analysis under a wide range of nuclear reactor operating conditions, including nuclear fuel burnup. |