Code assessment and validation is one of the most relevant research lines in thermal hydraulics best estimate codes. During last decades, Nuclear Energy Agency (NEA) Organization for Economic Co-operation Development (OECD) have sponsored dozens experimental projects this field. Most them were compiled CSNI Validation Matrix 1996. Several been promoted new century as SETH, PKL, PKL-2, PKL-3 PKL...

A RELAP5-3D model has been developed for the RBMK Kursk 1 NPP. The primary features of this model are the three-dimensional neutronics model of the reactor core and the user subroutine that generates the neutron cross-sections. This subroutine contains the same neutron cross-section libraries that are presently being used in the Russian STEPAN/KOBRA calculations. The implementation of the neutr...

When a nuclear power plant is in shutdown conditions for refuelling, the reactor coolant system water level is reduced. This situation is known as mid-loop operation, and the residual heat removal (RHR) system is used to remove the decay power heat generated in the reactor core. In mid-loop conditions, some accidental situations may occur with not a negligible contribution to the plant risk, an...

The heat transfer model of the RELAP5/MOD3.1 (R5M3) code was extensively reviewed and assessed. The most important deficiency of the current version of the code was attributed to its treatment of the transition boiling heat transfer regime. The current transition boiling model significantly underpredicts the heat transfer rate to the liquid phase. Since at low quality conditions the liquid boil...

Abstract The present paper deals with the assessment of original and a modified version RELAP5/MOD3.3 against OSU Multi Application Small Light Water Reactor (OSU-MASLWR). new implemented features regard suitable correlations for heat transfer coefficient evaluation in helical geometry. Furthermore, two different modelling Helical Coil Steam Generator (HCSG) are assessed. In first approach, HCS...

The development and deployment of gas-cooled reactors (GCRs) require flexible modelling simulation tools for a detailed understanding the system behavior in normal operation accidental scenarios. Amongst identified safety-relevant scenarios reactors, outlet plenum flow distribution during operations was ranked to be high importance with low knowledge level phenomenon identification ranking tabl...