Before the accident at Three Mile Island, most attention was focused on the posmlated large-break LOCA. However, the Three Mile Island incident sharply focused attention on the fact that a small break (typically up to sizes where the reactor remains pressurized despite the break, say, up to 12-cm-diameter holes) in the primary circuit was, in fact, much more likely. At Three Mile Island this small break was due to a stuck-open power-operated relief valve. It could, however, also have occurred as a result of the break in one of the large number of small pipes attached to the primary circuit. Figure 4.20 shows a histogram of the number of pipes attached to the circuit as a function of pipe size and cross­sectional area. Also shown is the percentage area relative to the main coolant pipes. Figure 4.20 is for a German P’^TC. design, but the result is likely to be much the same for other designs.

The most important difference between small-break and large-break acci­dents is that in the former the reactor depressurizes relatively slowly. Reactor pressure as a function of time after the break is shown for various break sizes in Figure 4.21. Since the core may remain at a high pressure in a small-break LOCA, it is not possible to activate the accumulator or low-pressure injection system until late in the accident.

A typical sequence for a small-break LOCA is illustrated in Figures 4.22 to

.Intermediate

breaks yi Large breaks

5 20

Fi^^e 4.20: Connection pipe diameter/cross section/percentage spectrum of a P’^TC. Solid lines: primary loop system; dashed lines: pressurizer.

Fi^^e 4.21: Primary pressure versus time for small-break LOCAs in P’^TC. (O) Pri­mary temperature 175°C; (□) reflood tank empty. (Two HPI pumps; reflood tanks 4 x 286 m3; no LPI pumps.)

4.26. As with the large-break LOCA, the most serious effects are found when the break is in the reactor inlet pipe (the cold legs).

Following the initiation of the break, the pressure falls and the reactor trips. As the pressure falls below about 100 bars, the high-pressure injection system comes on. The pressure continues to fall to around 70 bars, when the hottest liquid in the circuit starts to vaporize and produce steam. First, the water in the pressurizer vaporizes. As the saturation condition is reached throughout the hotter parts of the primary circuit, steam bubbles form and because the pumps are stopped, will settle out in the upper part of the reactor as shown in Figure 4.22 There has been considerable controversy about whether to leave the cir­culating pumps operating or to stop them during a LOCA. If left on, they may assist in circulating liquid through the core, promoting its cooling. On the other hand, they may aid the loss of fluid by pumping it out through the breach. Cur­rent rules for operating P^WR indicate that the pumps should be stopped, this being considered to give the balance of advantage in general.

As a result of depressurization, steam forms and collects in the upper head of

the reactor but cannot escape via the breach because of the arrangement of the pipework. However, the loss of water is quite rapid, though it removes a rela­tively small amount of energy from the system. The water drains down to the level of the water inlet-outlet pipes on the reactor pressure vessel (the “noz­zles”) in about 250 s, during which the pressure may be maintained at a high level and still inhibit the actuation of the accumulators and LPIS.

During this phase, the steam generators are voided (i. e., their original water content is lost through the break) on the primary side, the steam from the reac­tor having access to the steam generators, as illustrated in Figure 4.23. Obvi­ously, the steam generators represent a potential heat rejection source, with the steam from the reactor core being condensed in the steam generator tubes and flowing back down the tubes and into the core Figure 4.23). However, this ben­eficial action is conditional on the secondary side being at a sufficiently low pressure (and corresponding low saturation temperature) to allow heat to be extracted via the secondary circuit. Assume, for instance, that the secondary — side pressure remains at its normal operating value of 70 bars (1000 psia) while

the steam from the primary system is arriving at the steam generators also at 70 bars. In this situation the saturation temperatures are identical and no conden­sation can take place. Therefore, it is imperative in this type of accident that sec­ondary-side cooling or depressurization is carried out.

Because of the continuing loss of water from the system, the core begins to dry out from the top downward (so-called core uncovery), as illustrated in Fig­ure 4.24. The system is still pressurized due to the formation of steam, which cannot escape through the cold leg break, being blocked by the water in the vessel and the pump. The pump has a U-bend under it (the pump loop sea[), and this remains full of water and blocks the steam from flowing from the ves­sel through the steam generator and pump to the break. The level in the pump side of the loop seal near the break (i. e., left-hand side of the loop seal at the left in Figure 4.25) is roughly equal to the level of water in the vessel since the two levels are connected by the (voided) steam generator and pump. Only when the levels reach the bottom of the U-bend can the steam pass the loop seal. At this point, steam from the core passes through the pump and out along the cold leg to the break. This results in a rapid depressurization. The water in

the remaining part of the core partially vaporizes, and the mixture of water and steam bubbles formed rewets the upper part of the core. As depressurization proceeds, the core may be dried out again (as in a large-break LOCA). How­ever, the depressurization permits actuation of the accumulator and LPIS sys­tems, and these rapidly reflood the core and bring it to a cold condition. In the longer term (usually longer than 350 s) heat is extracted in the way illustrated for the large LOCA in Figure 4.17.

Figure 4.27 shows the variation of water level and fuel clad temperature for small-break LOCAs with two different equivalent break diameters. The level of the two-phase mixture in the vessel is shown. Normally, the cooling is good for regions of the core that are in contact with this mixture, the core being over­heated above this level. The mixture level is much higher than the level of the liquid without steam bubbles would be. The latter level is referred to as the col­lapsed liquid level, and the phenomenon of increase of level due to the pres­ence of the bubbles is termed level swell. A similar phenomenon occurs in dispensing glasses of beer.

4.3.4 Alternative ECCSs

The descriptions above apply mainly to a typical U. S. P^WR where the emer­gency core cooling water is injected into the coolant inlet pipes (the cold legs) only. Alternative ECCSs have been used, the most important being that used in the German P^WR where the emergency core cooling water is injected into both the cold legs and the hot legs (the coolant outlet pipes from the reactor vessel). Such combined injection is claimed to offer advantages in more rapid quench­ing of the core and in lower peak cladding temperatures during a large-break LOCA. In the case of the small-break LOCA, it is claimed that faster depressur­ization occurs, allowing early actuation of the accumulator and LPIS systems.