The classical design basis accident for a pressurized-water reactor is the large-break loss-of-coolant accident (LOCA). It is assumed that in this accident one of the inlet pipes from the circulating pump to the reactor vessel is com-

pletely broken and moved apart to allow free discharge of the primary coolant from both broken ends. This kind of break is called a “double-ended guillotine” break or a “200%” break. Because this break is commonly believed to represent about the worst accident that could happen to a water reactor circuit, it has been chosen as the basis for the design of the emergency response systems.

The sequence of events following the break is shown in Figures 4.13 to 4.17, which illustrate the situation in the whole reactor circuit. A more detailed illus­tration of the events within the reactor vessel itself is given in Figure 4.18. The main phases are as follows:

1. Blowdown phase. Under normal operation (Figures 4.4 and 4.18a), water flows through the inlet pipes (the cold legs) to the reactor vessel, down the annular space around the core, up through the core, and out through the vessel outlet pipes (hot legs) to the steam generator. When a large break oc­curs in one of the cold legs, the contents of the reactor vessel and primary loops are blown down through the break as illustrated in Figure 4.13 and 4.18b. After a very rapid initial depressurization, the pressure falls more slowly due to the creation of a two-phase steam-water mixture in the vessel and circuit, the mass flow of such a mixture through a break being much lower than that for a single-phase liquid. After about 10 s, the pressure has

Refill phase time = 30-40 seconds

Fi^^e 4.16: Large LOCA: reflood phase (40-250 s).

fallen to that for initiation of flow from the high-pressure injection system and the accumulators into the ECCS line in the cold legs.

2. Bypass phase. After the initiation of the ECCS, starting with the HPIS and the accumulators, there is still a significant upward flow of steam in the down­comer annulus through which the cooling water normally flows. This steam flow prevents the accumulator ECCS water from entering the region of the vessel below the core (the lowerplenum), and the water simply bypasses the

upper part of the inlet annulus and flows out through the break, as illustrated in Figures 4.14 and 4.18c.

3. Refill phase. Filling of the lower plenum (the refill phase—Figures 4.15 and 4.18c) begins after further depressurization, when the steam flow up the an­nulus has dropped to a sufficiently low value that it can no longer restrict the ingress of ECCS water. By this stage, the LPIS system will also have been ini­tiated. In a typical PWR, refilling of the lower plenum starts about 23 s after the initial break, and it takes 17 s to fill the lower plenum with liquid, ending this refill phase of the accident.

4. Reflood phase. At a very early stage in the blowdown phase, the core has dried out and the fuel element temperatures rise rapidly, after which they fall relatively slowly due to the existence of substantial steam flows in the core. Typically, the fuel element temperatures rise to around 1000°C. This leads to rupture of the fuel elements, which release gaseous fission products into the primary circuit and, via the break, into the containment vessel. The behavior of typical fuel elements as a function of temperature is shown in Table 4.2. When the lower plenum is filled, the reflood phase begins (Figures 4.16 and 4.184), with the fuel elements beihg rewetted from the bottom upward. Es­sentially, a constant liquid head is Maintained in the inlet annulus during this phase, with excess ECCS water overflowing through the break as illustrated. As the fuel elements rewet, a considerable volume of steam is formed and entrained liquid droplets flow before the rewetting front and pass into the upper plenum. The steam-droplet mixture passes from the upper plenum, through the steam generator, through the circulating pump, and back into the cold leg, flowing out through the break. The water droplets tend to evap­orate in the steam generator due to the backflow of heat from the secondary — side (still hot) fluid. The resistance presented by the outflow route causes a back pressure in the upper plenum, which restricts the rate at which the re­flood can take place. This phenomenon is often referred to as steam bind­ing. The highest resistance of the upper plenum, through the steam generator and circulating pump, to the break would occur when all of the droplets issuing from the core passed to the steam generator and the circu­lating pump rotor was locked stationary. However, the resistance is much re­duced, and the flooding rate greatly increased, if the droplets deposit out on the upper plenum structures and thus are not carried out of the vessel, and if the pump rotor is still rotating.

5. Long-term cooling. In the long term, the situation is as illustrated in Figure 4.17. Water is passed to the unbroken cold leg from the LPIS injection pump and maintains a head of liquid that drives water through the core by natural circulation. Steam maybe generated in the core and may escape with the overflow water through the break as illustrated. This generated steam is condensed by sprays in the containment, which are also fed from the LPIS pump.

In carrying out the design of a P^WR two types of calculations are usually em­ployed, based on an evaluation model or on best-estimate methods. With an evaluation model, the various phenomena are represented by equations and as­sumptions that are postulated to give the worst conceivable result. For instance, it is normally assumed that there is no penetration of ECCS water during the blowdown phase. In best-estimate methods, the best available physical models are used for the various phenomena and an attempt is made to calculate the sys­tem behavior on the basis of these models. It should be pointed out, however, that the calculation of two-phase flows, particularly for the rapid transient condi­tions and large pipe sizes encountered in reactors, is still at an uncertain stage. As explained in Chapter 3, two-phase flows are very complex and in many re­spects poorly understood. It would be unsatisfactory to rely on two-phase mod­eling as a basis for reactor design. Some critics claim that the uncertainties in two-phase flow predictions imply that the reactor is unsafe. We do not share this view. From our long experience, we would agree with the assessment of the current state of modeling of two-phase flows but disagree that the reactor design is based on the results of such modeling. This design must satisfy very conserv­ative criteria that do not depend on knowing about the details of two-phase flow behavior.

Figure 4.19 shows the variation of peak clad temperature as a function of time calculated from the evaluation and best-estimate models, respectively. The con­tinuous line was calculated by using the conservative evaluation approach; the best-estimate values are shown as error bars.

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