Uncertainty analysis of rod ejection accident in VVER-1000 reactor

In present work the coupling of neutronics code PARCS and thermo-hydraulics code TRACE was performed with adoption of GRS (Gesellschaft für Anlagen- und Reaktorsicherheit) uncertainty method for rod ejection accident in VVER-1000 reactor core. The determination of input parameters, which represent source of possible uncertainties, for both codes was based on the Phenomenon Identification and Ranking Tables (PIRT) for rod ejection accident in pressurized water reactor (PWR) established by the U.S. Nuclear Regulatory Commission (U.S. NRC). Following input parameters in PARCS were selected: delayed neutron fraction precursor groups, macroscopic cross-section data for density of moderator, temperature of fuel and temperature of moderator. For TRACE code input uncertainty of thermal conductivity of fuel, gap and cladding were defined. The results obtained from uncertainty analyses imply that macroscopic cross-section data of moderator temperature is the most influential uncertainty among the input parameters in a neutronics code. On the other hand, uncertainty of thermal conductivity of uranium fuel significantly affects the results of coupled calculation. The influence of input uncertainties were investigated on the calculated results, mainly for core thermal power, fuel centerline temperature, total reactivity, moderator density, and fuel temperature reactivity.