STEADY STATE AND TRANSIENT THERMALHYDRAULIC ANALYSIS OF PHWR USING COBRA-3C/RERTR

Nuclear cross sections that determine core multiplication strongly depend on core temperature (e.g., the Doppler, moderator density effects etc). On the other hand, since this heat is generated by the neutron flux in the reactor core, the temperature distribution in the core will depend heavily on its neutronic behavior. Fuel centerline temperature could be the limiting constraint on reactor power because of the concern for fuel melting. Likewise, high clad temperature is also a possible limiting factor on reactor power because of the potential degradation of clad material or on-set of critical heat flux phenomenon. An assessment of the steady state and transient thermal hydraulic capabilities of the computer code COBRA 3C/RERTR was made using model for a PHWRs reactor core. The temperature distributions determined for fuel, clad and coolant are compared with analytical results and with the results quoted in safety report. It was found that when the code was run for full power at reduced flow of 70% the bulk coolant temperature remained below the saturation temperature, so there is an adequate design margin is available for safety related scenarios.