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|Title:||STUDIES ON THE FUEL BUNDLE HEAT-UP UNDER FULLY VOIDED CONDITION DURING LOCA IN AN INDIAN PHWR|
|Keywords:||Nuclear Power Plants;Water Reactor;Heat Transfer;Coolant Circulation Pumps;Uranium Oxide|
|Abstract:||Majority of Indian nuclear power plants are PHWRs based on CANDU reactors. These reactors use natural uranium oxide as fuel and are heavy water cooled and moderated. The nuclear fuel is contained in hundreds of horizontal channels which is continuously cooled by the heavy water coolant during its operation. One of the major risk factor in a nuclear power plant operation is Loss of Coolant Accident (LOCA), in which the ability of the coolant employed to transfer heat from the core, is compromised. Such an accident can occur due to breaks in the water circuit or failure of the coolant circulation pumps. This results in core heat-up, leading to its damage to the reactor due to high temperature. In case of an emergency shutdown, even though the fission reaction in the nuclear fuel is halted, the decay heat, which is around 2% of the thermal rating of the reactor, is capable to incur permanent damage to the reactor. Emergency Core Cooling System (ECCS) is employed as a backup measure in such a condition, which replenishes the coolant supply in the heat transport system. In case of multiple failure scenario, such as LOCA with ECCS failure, the decay heat continues to raise the reactor core temperature, eventually leading to the core voiding. In such scenario the convective heat transfer becomes poor and the majority of the heat transfer from fuel bundle takes place by radiation mode. During this abnormal working condition, if the channel pressure is less than 1 MPa, the PT sags and come in contact with the CT. This results in high rate of heat transfer from contact location to moderator. It is therefore very essential to study the temperature behavior of the channel undergoing radiation heat transfer under such accidental scenario. Experiments were carried out with single fuel pin and 19 pin fuel bundle under different test conditions. Experiments with single fuel pin have been conducted to evaluate the variation of total emissivity of pre-oxidized clad (autoclaved) fuel pin of 220MWe Indian pressurized heavy water reactor (IPHWR) at different temperatures ranging from 300ºC – 900ºC under inert atmosphere. The emissivity of fuel clad which is made of Zircaloy-4, is an important parameter for estimating the radiation heat transfer between fuel bundle and pressure tube (PT). From experimental analysis it has been found that the emissivity of the fuel pin decreases with corresponding rise in temperature. Experiments with 19 pin fuel bundle were carried out under two different conditions i.e. under no flow and under flow conditions with fully voided channel undergoing radiation heat transfer. Experiments were performed under no flow conditions to capture the temperature behaviour of channel with two different configurations of PT by changing the eccentricity ii (vertically downward) of PT w.r.t CT at a steady state fuel bundle temperature of 600℃ with PT concentric with CT and at a steady state fuel bundle temperature of 600℃ & 700ºC with PT contact with CT (e = 8.5 mm). During these experiments there was not any flow of Argon gas. Therefore, the heat transfer in the set-up was through conduction, natural convection and radiation. The results showed that the bottom nodes of all the components (Fuel bundle, PT and CT) of the simulated channel was greatly influenced by the PT/CT contact. Moreover, higher temperature were observed at top nodes of the channel. However, no significant variation in temperatures were obtained in fuel bundle and PT in concentric condition. In experiment with 19 pin fuel bundle under flow condition, the argon gas was induced inside the PT and in the annulus region between the PT and CT. The continuous purging of argon gas was done inside the PT so as to avoid any oxidation of 19 pin fuel bundle and PT, while inside the annulus region of PT and CT the argon gas was filled and trapped. Three sets of sub- experiments were conducted under this condition by changing the eccentricity of PT (vertically downwards) with respect to CT i.e. PT concentric with CT (e=0), PT 4 mm eccentric with CT (e=4 mm), and PT contact with CT (e=8.5 mm) respectively at three different steady state fuel bundle temperature of 600ºC, 800ºC and 1100ºC respectively. The experimental results showed that the eccentricity of PT strongly affected the circumferential temperature distribution over the fuel bundle, PT and CT. The variation of temperature along the circumference of fuel bundle and PT at e=0 was insignificant, however, for other positions, this variation in temperature was quite significant. The bottom nodes of fuel bundle showed a decrease in temperature with an increase in eccentricity. Similar observations are also made for PT. On the other hand, the temperature at the bottom nodes of the CT increased for higher eccentricity between PT and CT.|
|Research Supervisor/ Guide:||Kumar, ravi|
|Appears in Collections:||DOCTORAL THESES (MIED)|
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