Aging Effect on the Fuel Behaviors for CANDU® Fuel Safety Analysis

Abstract

Recent safety analysis for the refurbished Wolsong 1 considered the aging effect of the reactor by assuming that the core is the 11 EFPY (effective full power year) state. Here, 11 EFPY core state means that the core was assumed to be operated for 11 EFPY after the refurbishment of the Wolsong 1. From the site operation data and thermal-hydraulic analysis results, the aging data for major components of reactor and heat transport system were derived. The derived data include the roughness data of feeder pipe, steam generator, end-fitting and pressure tube, creep data for pressure tube, orifice degradation, and steam generator fouling, and so on. The reason that the safety analysis was carried out for the aged core is just to obtain the conservative results in terms of the safety. However, a detail comparison of the safety analysis results for the fresh core and the aged core has not been assessed for the specific postulated accident scenario. In this study, fuel safety analyses were carried out for both fresh and 11 EFPY aged cores in order to compare the aging effect on the fuel safety analysis. The safety analysis was conducted for a stagnation feeder break accident which is one of the single channel accidents when the other channels remain intact in the CANDU® core. A feeder break can be occurred in any of 380 channels in the reactor at any time during the reactor’s operating life. Because of this, a feeder break is assumed to occur in the high-powered ‘limiting’ channel of which power is a limiting operational condition of 7.3 MW for the conservative safety assessment. From the circuit and channel thermal-hydraulic analysis, the critical break size of the inlet feeder pipe which causes the most stagnation flow in the channel was determined for each fresh and aged core. Fuel analyses were carried out for two steps: the first analysis evaluated the fission product inventory and its distribution for the normal operating condition and the second step assessed the fission gas release after the feeder break for each critical break size of the fresh and aged cores. Through the analysis results, the aging effect on the fuel safety will be assessed and also it will be checked whether the aged core assumption is meaningful for the safety analysis of the design basis accidents.