Heat Transfer Calculation of Simulated Heater Rods throughout Reflood Phase in Postulated PWR-LOCA Experiments

Abstract

A computer code, which calculates the transients of heat flux from simulated nuclear fuel rods by using the transients of rod surface temperature and the heat conduction equation in the rod, was developed in order to investigate the heat transfer modes throughout the reflood phase in PWR-LOCA experiments. The code was applied to the Slab Core Reflood Tests which are part of the Large Scale Reflood Test Program at the Japan Atomic Energy Research Institute. For defining the heat transfer modes during reflood, it is important to obtain accurate heat flux from rod under a wide rod temperature change ranging higher than 1,300 to 300 K and a rapid rod temperature change due to quench, which are principal features in heat transfer during reflood phase. Therefore, the effects of both temperature dependency on physical properties of rod and the axial heat conduction along rod on the heat flux calculation were first investigated. As the results, it was made clear that the temperature dependency on the physical properties should be taken into account and that the effect of axial heat conduction along the rod was negligible except in a very short length of rod at the quench front. The results calculated by the code for the Slab Core Tests when compared with the existing correlations could define the heat transfer modes clearly all through the reflood phase but the recommendations for further investigations were suggested.

KEYWORDS:

• reflood phase
• heat transfer
• calculation code
• temperature dependence
• physical properties
• axial conduction
• slab core reflood test
• accuracy
• fuel rods
• PWR type reactors
• loss-of-coolant
• heat flux