This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: β – delayed neutron fraction precursors and transport, absorption, ν-fission, κ-fission, down scattering for moderator temperature, moderator density and fuel temperature. As a neutronics code, PARCS in conjunction with thermo-hydraulics code TRACE was applied. Uncertainty analyses were carried out by the code DAKOTA where coupled PARCS/TRACE calculations were run 146 times in order to meet a two-sided 95/95 confidence interval. In the thermo-hydraulic code thermal power of the reactor core and inner temperature of the fuel rod were studied. This research shows that the most influential parameters associated with cross-section parametrization are moderator temperature and fuel temperature.
- Coupled calculation
- Cross-section uncertainty
- RIA accident
ASJC Scopus subject areas
- Nuclear Energy and Engineering