EXPERIMENTAL STUDY ON POSTULATED LOCA IN INDIAN HEAVY WATER NUCLEAR REACTORS

Abstract

The Indian fleet of operating commercial nuclear reactor consists of two BWRs (Boiling Water Reactors), eighteen PHWRs (Pressurised Heavy Water Reactors) and two VVERs (1000 MW each unit) reactor with total installed capacity of 6780 MW. The AHWR (Advance Heavy Water Reactor) is in design and development stage that burns thorium as fuel in the core. The IPHWRs are based on CANDU type reactors that burn natural uranium oxide as fuel and heavy water cooled and moderated. The reduction in flow rate of coolant inside the primary circuit due to Loss of Coolant Accident (LOCA) leads to one of the major risk, in which, heat carrying capacity of the coolant is compromised. The LOCA can occur due to large or small break in the primary circuit or due to pump failure. The reduction in heat carrying capacity of the coolant leads to heat up of reactor core and can damage the reactor core. The reduction in pressure inside the channel, due to break in primary heat transport channel, leads to formation of water vapor. The clad tube material (Zr-2.5% Nb) oxidizes under presence of high temperature steam and hydrogen gas is generated, which is explosive in nature. The decay heat (around 2% of the thermal rating) of the reactor is capable to damage to the nuclear reactor if not removed from the core. To avoid such a scenario, the Emergency Core Cooling System (ECCS) is employed as a backup measure, which, reestablished the coolant supply in the HTS (heat transport system) and carry away the decay heat from the fuel bundle. The emergency shutdown scenario becomes terrible when, multiple failures occur, such as LOCA along with the failure of ECCS. The fall in pressure inside channel and reduction in coolant flow, eventually lead to the boil-off condition and finally ends with the core voiding. The mode of heat transfer change with change in thermal hydraulic condition of the channel. The convection mode remains the dominating during normal operating conditions, but in case of the boil-off condition or core voiding, the radiation mode of heat transfer becomes the dominant mode of heat transfer, which carries heat from fuel bundle to moderator. Under boil-off condition of channel, if pressure inside the channel fall below 1 MPa, then PT sags and come in contact with the CT. The contact of PT to CT causes high heat flux from contact area to moderator. The high rate of heat flux through contact area can damage CT due to metallurgical failure of CT/PT. So the study of various modes of heat transfer and thermal condition of the CT, PT and fuel bundle is imperative. ii An experimental facility for simulated LOCA scenario of 220 MWe PHWR has been developed in the Nuclear Thermal hydraulics and Reactor Safety Laboratory (NTRSL). The single channel of IPHWR consists of calandria tube, pressure tube and simulated 19 pins fuel bundle. The experiments were conducted at three different temperatures 600°C, 900°C and 1100°C with three eccentric position of PT (e = 0 mm, 4 mm and 8.5 mm) under flow of steam and argon mixture. The PT eccentric positions were changed with respect to CT vertical. The argon gas acts as a carrier of the steam during experiment and also helps to dilute the hydrogen gas formed due to oxidation of clad tube of the fuel bundle. The experimental result shows that the eccentricity greatly affects the temperature distribution of the nuclear channel. The bottom side nodes of the fuel bundle show significant drop in temperature when the CT and PT come in contact at all three temperatures. The maximum and minimum temperature exists diametrically opposite to each in calandria tube and pressure tube. The second experimental facility was developed to understand the rewetting phenomena in the AHWR (Advance Heavy Water Reactor). The vapour film formed between the surface of heated clad tube and coolant, when surface attained the minimum film boiling temperature. The contact between coolant and surface re-established when the coolant breaks the stable vapour film, and this process of quenching is known as rewetting. The direct contact of heated surface and coolant enhances the heat removal rate from heated surface. The LOCA scenario in AHWR is similar to as described as above, in which, the coolant supply inside the primary heat transport circuit decrease which leads to rise of temperature of the nuclear channel. The heat up of the clad tube due to decay heat can cause melt down of the nuclear core, so, to avoid such scenario, emergency core cooling water introduced in radial jet form into the nuclear core through the central water tube. An experimental facility has been created to investigate single channel of AHWR with postulated LOCA has been developed. The outer surface of the pressure tube was completely insulated to resist radial heat loss. The 54 pins fuel bundle consists of three concentric rings: inner, middle and outer ring having 12, 18 and 24 numbers of clad tube pins. The central water tube, which is placed at the center of the fuel bundle, receive water from emergency core cooling system is able to produce radial jet effect on the clad tube and quench it while there is LOCA. The experiments were conducted at four different initial temperatures 450°C, 500 °C, 550 °C and 600 °C and 90 lpm water supply. The heating of the fuel bundle was done in complete vapor environment to simulate the LOCA condition under boil-off condition. The iii experiments results show that the water jet injection during rewetting at different axial and circumferential location causes different rewetting periods of each pin. The inner ring fuel pins went under rewetting immediately with the initiation of ECC, whereas, the outer ring takes more time to rewet. It is also observe that the jet facing nodes on the pins rewet faster than the nodes in opposite direction at the same axial elevation except outer ring pins.