Thermal-Hydraulic Response of a Reactor Core Following Large Break Loss-of-Coolant Accident under Flow Blockage Condition

Abstract

Since the revision of the requirements to consider the effect of fuel burnup on emergency core cooling system performance was proposed, flow blockage in reactor core has been one of the important issues in the thermal-hydraulic analysis of loss-of-coolant accident (LOCA). The present paper describes how much flow blockage would be expected following a large break LOCA based on the actual nuclear design data including the power and burnup of the fuel rods. A system thermal-hydraulic code, MARS-KS, is used for calculation where the burnup specific data of the fuel rods is supported by a fuel performance code, FRACON3. To recover the weakness of the system code in which the flow blockage under multiple rods configuration cannot be automatically simulated in hydraulic calculation, a special modelling scheme is developed and applied to the calculation. The effect of flow blockage on the thermal-hydraulic response of the reactor core is also discussed. To compensate for the uncertainty of the present flow blockage model, additional calculations are attempted for a wide range of the level of blockage.