A subchannel analysis of CANDU fuel bundle in a crept fuel channel

In a CANDU reactor, the fuel channel diameter increases due to the pressure tube creep resulting from neutron irradiation damage. It is well known that a crept pressure tube leads to a non-uniform coolant flow resulting to a decrease of dryout power. The previous experimental study reported the dryout power and the location the first dryout occurs for the uncrept and crept channel. However, it has the limitation to reveal the detailed thermal-hydraulic characteristics for the local flow field. The subchannel technique is known to be very useful for investigating the thermal behavior of fuel assembly in a nuclear power reactor. In this study, the subchannel analysis has been conducted for the uncrept and crept channel. The objective of this study is to examine the change of thermal-hydraulic characteristics by the pressure tube creep and to assess the thermal margin of each rod in a fuel bundle