Primary Pipe Rupture Accident Analysis for the Clinch River Breeder Reactor

Analyses of primary pipe rupture accidents in the Clinch River Breeder Reactor were carried out with Brookhaven National Laboratory-modified versions of the DEMO code. The thermal transient responses of the core and radial blanket were calculated for a large number of initial conditions and plant configurations. These include studies of variations of pipe break size and operating power. Pipe ruptures commencing from two-loop initial operating conditions are presented. The sensitivity of the thermal response to variations in particular parameters within the estimated ranges of their uncertainties has been studied. Conditions under which sodium boiling in the core occurs are delineated.