Approaches adopted for CHF evaluation during transients using system analysis code RELAP-5 for KKNPP

The Critical Heat Flux (CHF) condition is important in heat-flux-controlled systems such as nuclear reactors since the contributing parameters of Departure from nuclear boiling ratio (DNBR) can threaten the physical integrity of the heated surface (Fuel clad). Kudankulam Nuclear Power Plant (KKNPP) has two operating VVER Reactors with the rated output of 1000 MWe. VVER-1000 is a Pressurized Water Reactor incorporating advanced safety features including passive ones. In Pressurized Water Reactor (PWR),CHF signifies an important thermal-hydraulic safety limit and it needs to be established within prescribed limits for all operating and transient conditions. Various postulated initiating events has been analyzed using Thermal Hydraulic computer code RELAP-5/MOD 3.2 for KKNPP. The calculation is performed from the viewpoint of checking of DNB during the whole transient. The DNBR value observed during the transients is calculated by two ways i.e.RELAP look up tables and externally coupled designer's specified correlations such as Gidropress, Smolin and Kutateladze. This paper covers the approaches adopted far the DNBR evaluation of KKNPP using system analysis code 'RELAP-S' and DNBR estimation and verification with acceptance criterion for governing transients