DYNAMICAL TWO-PHASE FLOW ANALYSIS
5.5 Base Case
The pressure in the pressurizer, in the secondary side, and in the containment, the mass flow rate of the blowdown, the velocities and void fractions of the break flow geometry, and the pressure loss in the breakflow geometry are shown in Figures 5.2 - 5.6. The mass flow out of the reactor vessel into the containment is restricted by critical flow until t = 4000 s. At that time, a small increase in the mass flow rate can be noticed, as seen in Figure 5.3, and also a sudden decrease in the pressure difference between the containment and the outlet of the break flow geometry, as seen in Figure 5.6.
5.5.1 Coolant Flashing
Coolant flashing occurs due to the de-pressurization of the coolant in the reactor vessel right after the break. When the pressure in the reactor vessel reaches the saturation pressure for the coolant, the coolant saturates.
5.5.2 Natural Circulation
After the pump has coasted down (t RS 100 s after the break), natural circulation in the primary side piping is established due to the temperature differences in the different parts of the piping. The mass flow rate of the natural circulation is msc — 80 kg/s, or around 8% of the mass flow rate during normal operation.
5.5.3 Steam Void Formation in Steam Generator Primary Pip-ing
In the first 800 s of the transient, water and steam are leaking out of the system. During this time (not including the coast-down of the pump) the pressure is decreasing slightly, see Figure 5.2. As the water level in the reactor vessel reaches a point below the outlet of the break, the void fraction in the inlet and outlet of the break geometry increases, and the pressure in the reactor vessel decreases, continuing to a value of ?» 5 bar in RS 6000 s, see Figures 5.5, and 5.2.
After 1000 s, steam is formed in the primary side piping of the steam generator as the pressure and temperature of the coolant reaches the secondary side conditions (20 bar, saturation). Furthermore, the direction of the heat transfer between the primary and secondary side is altered at the same time. The heat transfer rate from the secondary to the primary side is not as efficient, due to the steam in the primary side piping in the steam generator which does not transfer the heat as efficiently as water.
Also, as heat is transferred from the secondary side to the primary side, coolant saturation and steam expansion occur, thereby increasing the pressure and lowering the water level in the hot and the cold leg.
When the water level in the steam generator hot and cold leg is pressed down, the water level in the reactor vessel increases. When the level is low enough, steam slugs can escape, which causes a decrease in pressure, as well as an increase in the water level in the steam generator primary side piping. The level oscillations of the coolant in the reactor vessel causes the instabilities in the break mass flow rate between 1000 and 1600 s. The
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temperatures in the primary and the secondary side of the steam generator are presented in Figure 5.7.
The total volume of the steam generation in the primary side due to the heat transfer from the secondary side of the steam generator is approximately 12 m3, which is equal to a level rise in the reactor vessel of approximately 0.5 m.
In the base case, the secondary steam pressure is not lowered by dumping of the secondary side steam. If the secondary side were to be dumped, the steam generator pressure and temperature would decrease. The steam in the primary side piping would be cooled down, and the water level in the reactor vessel would decrease.
5.5.4 Critical Flow
The flow in the break flow geometry is restricted by critical flow in different sections until t = 4000 s after the break. After 4000 s, the reactor pressure decreases below the value required to maintain the critical flow in the junction which connects the pipe break with the containment. The pressure difference between the containment and the outlet of the break flow geometry instantaneously disappears, see Figure 5.6.
5.5.5 Core Heatup.
After approximately 2700 s, water level decreases to uncover the fuel bundles and the fuel temperature at locations near the top of the core slowly begins to rise.
In Figure 5.8 the cladding temperature rise, as calculated with RELAP5 is presented.
The temperature gradient for the last 200 s of the transient calculated is around 0.1 K/s.
The oscillating behaviour depicted is due to the variations in the calculated heat transfer coefficient by the RELAP5/MOD3 code as the water level oscillates in the top nodes of the fuel rods.
5.5.6 Extension of RELAP5
After 4000 s, the break mass flow rate is no longer critical, and consists entirely of steam. Further calculation of the pressure and the mass flow rate can be performed using the thermal equilibrium equation, combined with the Bernoulli equation (TEENAGE) described in Chapter 4. This way, the time to reach equilibrium between the containment
and the reactor vessel can be calculated. This extension is showed in Figures 5.16, and 5.17. According to this calculation, equilibrium between the reactor vessel and the containment is reached after approximately 16,000 s.
In Figure 6.1 and 6.2, the calculated collapsed levels during the critical flow part of the transient, and after equilibrium is reached are shown, respectively.