Experimental investigation of heat transfer during LOCA with failure of emergency cooling system

Paper presented at the 5th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, South Africa, 1-4 July, 2007.The pressure tubes (PT) in the Indian Pressurized Heavy Water Reactor (PHWR) are kept concentrically inside horizontal calandria tube (CT), which is submerged in a moderator pool. During postulated loss-of-coolant accident (LOCA) without emergency core cooling system, the temperature of PT could rise significantly. At elevated temperature, the weight of the PT with the weight of fuel pins could cause it to sag. Before sagging heat transfer from the PT to CT is mainly by convection. Once the PT sags and touches the CT, heat transfer mode changes to conduction. Direct contact between the PT and CT increases heat transfer and checks the temperature rise of the PT and limit the fuel heatup and subsequent degradation of the reactor core. An experimental set-up is designed and fabricated at Indian Institute of Technology Roorkee (IITR) to simulate the LOCA in the Indian PHWR. From the experimental investigation, it is found that the contact between the PT and CT occurred at around 300 seconds after the initiation of the simulation with corresponding temperature of the PT at around 680°C. The experiment was continued after the contact between the PT and CT and it is found that the temperature rise of the PT was within 800°C. This work demonstrates the inherent safety feature of Pressurised Heavy Water Reactor.cs201

ASTEC adaptation for PHWR limited core damage accident analysis

International audienceUnder limited core damage accidents (LCDAs) of Pressurized Heavy Water Reactor (PHWR), coolable geometry of the channel might be retained thanks to the presence of moderator heat sink. Indeed, the pressure tube is amenable to creep deformation at high temperature due to internal pressure and fuel bundles weight. Partial or complete circumferential contact between pressure tube and calandria tube aids heat dissipation to the moderator. A new module has been developed by Bhabha Atomic Research Centre (BARC) for simulating this phenomenon which is specific to horizontal-type of reactors. It requires additional calculation of pressure tube sagging/ballooning and temperature field in the circumferential direction. The module is well validated with available experimental results concerning pressure tube deformation and the associated heat transfer in the area of contact. It is then used in analysing typical LCDAs scenarios in Indian PHWR under low and medium internal pressure conditions. This module is implemented in the ASTEC IRSN-GRS severe accident code version under development and will thus be available in the next major version V2.1. © 2013 Elsevier B.V