INVESTIGATIONS ON CONVECTIVE HEAT TRANSFER FROM CHANNEL ASSEMBLY TO MODERATOR DURING LOCA IN AN IPHWR

Abstract

India has pressurized heavy water reactors (PHWR) with capacities of 220 MWe and 540 MWe, and the 700 MWe capacity reactor is under construction. The 392 coolant channel assemblies that make up these 700 MWe PHWRs are positioned horizontally within the Calandria vessel. A pressure tube (PT) made of Zr-2.5% Nb alloy and a calandria tube (CT) made of Zr-2 alloy are the two concentric tubes that make up each coolant channel. Axially loaded into the PT is a series of fuel bundles with a configuration of 37 fuel elements. The cooling environment of the channel deteriorates in a postulated accident scenario, such as a loss of coolant accident (LOCA) concurrent with the loss of emergency coolant injection system (LOECIS), because steam is the only coolant available to remove the decay heat of the fuel. Due to this intense heating, the channel's structural integrity may be compromised (for example, by the PT sagging, ballooning, or both, and bursting of clad tubes). Safety analysis of postulated initiating events is part of nuclear reactor design. All studies conclude that the moderator acts as a heat sink and is capable of removing decay heat, thus limiting the fuel heat-up. The convective heat transfer is found to be the governing heat transfer resistance in the decay heat transfer flow path from the fuel bundle of the channels to the moderator. Thus, the removal of decay heat, after a shutdown of the reactor, from channel to moderator plays a significant role in determining the fuel bundle temperatures. This thesis aims to determine the temperature of the fuel channel and the distribution of heat transfer for various channel configurations at different subcooling temperatures of the moderator. An experimental facility was developed to investigate the temperature behaviour of the PT and CT under heat-up conditions at different subcooling temperatures (65 ℃, 75 ℃, and 85 ℃) of the moderator using the 37 Fuel Pin Bundle Simulator (FPBS). At four different steady-state centre pin temperatures of FPBS, i.e. 450 ℃, 600 ℃, 750 ℃, and 900 ℃, respectively, these experiments were carried out under the condition by varying the eccentricity of PT (vertically downwards) with respect to CT, i.e. PT concentric with CT (e = 0), and PT in contact with CT (e = 8 mm). All the experiments were carried out under no flow condition; however, to prevent any oxidation of the 37 FPBS and PT, argon gas was continuously purged from the PT, and it was filled and trapped inside the annulus region of the PT and CT. Two-dimensional numerical simulations of the experiments were also performed using ANSYS Fluent 2020 R2. The Discrete Ordinates (DO) model was used to solve the radiation heat transfer equation in the numerical simulations. The experimental findings demonstrated the circumferential temperature distribution over PT and CT were significantly impacted by the eccentricity of PT.