Estimation of minimum critical heat flux ratio in highest power channel of 700 MWe PHWR using computer code CORETHERM

For safe and reliable operation of a reactor, it is important to predict the detailed flow and temperature distributions in the thermal hydraulic design of a reactor core. Critical Heat Flux (CHF) is one of the vital parameter for the thermal design of a fuel bundle. CHF is the maximum heat flux beyond which the surface temperature rises sharply and corresponding limiting power is called the critical power which is an important consideration for the thermal design of nuclear fuel bundle. Critical Heat Flux Ratio (CHFR) is the ratio of CHF to operating heat flux of fuel bundles. The nuclear fuel bundle must be operated below the CHF limit during normal operation and Anticipated Operational Occurrence (AOO). The acceptance criteria for Critical Heat Flux Ratio under normal operating limit is 1.3 and under AOO limit is 1.1. 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. It has 392 channels divided in two loops and each channel is having 12 fuel bundles in series. In PHWR operation, during transients, channel flow varies with changes of reactor and channel power. In addition, flow blockage in a channel could be caused by small objects which could be components in the PHT system (viz. valve parts, washers, bolts, nuts, etc.) or their fragments which have been dislodged, or they could be objects left inside the system inadvertently during construction. During blockage in coolant channel, mass flux reduces, coolant quality increases and pressure distribution changes in the heated section of the coolant channel. Complete flow blockage spectrum covers rate of closure of area as well as extent of area. The flow could fall suddenly or it could fall gradually, and the end-point flow could be anywhere between normal and very low flow. The in-house developed code CORETHERM (CORE THERMal Hydraulics), which is based on single channel lumped parameter approach, has been used to estimate Minimum Critical Heat Flux Ratio (MCHFR) for the highest power channel of 700 MWe IPHWR. The code models a single channel from inlet header to outlet header including feeders and end-fittings and fuel bundles. The code estimates steady state thermal hydraulic parameters in the channel at various powers and flow combinations using reactor inlet header and outlet header pressures as boundary condition. It can also estimate MCHFR during credible channel flow blockage scenario and pressure tube diametrical creep. This paper brings out CHF and CHFR at different locations of fuel bundles of highest power channel of 700 MWe PHWR. The obtained MCHFR value has been compared with the international sub channel analysis code and is found to be in good agreement with it. (author)