On a worldwide basis, the P^^ is the most common power-generating reactor. It is appropriate, therefore, to deal with the various operating states and postu­lated accident conditions for this reactor in some detail, using the framework laid out in Section 4.1.

2.1.1 Operating States of the PWR

The situation in nor-^l operation of a P^^ is illustrated in Figure 4.4. The pri­mary circuit consists of a pump that passes water at 292°C from the steam gen­erator through the reactor core, where it is heated to 325°C (it does not boil at this temperature since it is at high pressure). This hot water passes back through the U-tubes in the steam generator, where it cools down to 292°C; the water on the secondary side of the steam generator is boiled to generate steam, which passes out of the containment to the turbines, is subsequently con­densed, and returns through the feedwater pump to the secondary side of the steam generator. Also shown in Figure 4.4 are the various circuits for emer­gency core cooling water injection into the primaty circuit (i. e., the emergency core cooling system). These consist of:

1. The accumulators. These are large vessels containing water that are pressur­ized with nitrogen gas. They are connected to the primary circuit via auto­matic valves, which open if the primary circuit pressure falls below a preset level (typically 40 bars).

2. A high-pressure injection system (HPIS). This allows pumping of water into the system at pressures of about 100 bars, though normally at a relatively low rate.

3. A low-pressure injection system (LPIS). This allows water to be pumped at a high flow rate into the reactor, provided the reactor is at a low enough pres­sure (typically below 30 bars).

The combination of emergency core cooling injection systems thus allows a response to a variety of reactor depressurization and loss-of-coolant accidents.

If water escapes from the primary circuit, it collects in a sump at the bottom of the containment vessel and may be recirculated from there through the ECCS pumps back into the primary circuit. In the LPIS, the flow passes back to the re­actor through a heat exchanger, where it is cooled by the component cooling water system (CCWS). This provides a means of long-term decay heat removal from the reactor in the event of a loss-of-coolant accident. Note that the LPIS pumps can also be used to inject a spray of water into the containment to con­dense any steam present in the containment, thereby reducing the containment pressure in the event of an accident.

It is helpful in discussing P’^TC. operational states to represent the operation in terms of a pressure/temperature map as illustrated in Figure 4.5. The solid line in Figure 4.5 represents the saturation temperature (or boiling point) as a function of pressure. The P’^TC. must operate at a temperature to the left of this line to ensure that steam is not formed in the reactor. Figure 4.5 presents the operating conditions, showing the inlet and outlet temperatures at the operating pressure. The reactor pressure control is achieved in the pressurizer (see Figure 4.4) by having a body of liquid in contact with vapor at the saturation pressure. By raising or condensing steam within the pressurizer, the reactor circuit (which is connected to the pressurizer) is maintained at a fixed pressure. Thus, in terms of Figure 4.5, the pressurizer operates on the saturation curve as shown.

The reactor may reach the saturation condition by either increasing temper­ature or decreasing pressure. The most common way to reach saturation condi­tions is by depressurization, as illustrated. If the depressurization occurs by

means of a leak from the primary circuit, the initial rate of depressurization is extremely high. Once saturation conditions are reached, however, the rate of depressurization is much slower and may even reverse, with the reactor in­creasing in pressure for a short time.

Start-up and shutdown of the reactor must be carried out very carefully to avoid transients that would bring the reactor into a saturated state, with conse­quent vapor generation. It is also very important to avoid pressurizing the reactor vessel at too low a temperature. Doing this may cause existing small and insignif­icant defects in the vessel to extend and form significant cracks. The zone shown on the left-hand side in Figure 4.5 must also be avoided during operational tran­sients such as start-up and shutdown. Thus, there is a “window" for operation that is bounded at both low temperature and high temperature as illustrated. In prac­tice, the reactor is brought to its operating condition rather slowly over a period of about 24 h. A controlled return to cold shutdown also takes about 24 h.

The upset operating states of a P^^ can be categorized as follows:

1. Upsets leading to a change in the primary-side cook}nt inventory. This could be (as illustrated in Figure 4.6) a loss of fluid through a relief valve or through some other service line to the reactor. The primary-side inventory may also be increased by inadvertent pumping of water into the circuit through the high-pressure charging pumps. In the latter case, the pressurizer may become totally flooded with water and pressure control may be lost.

2. Upsets in the secondary-side heat removal capability. This could include loss of feedwater supply or changes in feedwater temperature, maloperations of the main steam-isolating valves, a turbine trip, or maloperation of pressure­regulating valves and/or safety valves (see Figure 4.7).

3. Other upset conditions (see Figure 4.8). These include inadvertent malopera — tion of the control rod system and the possibility of a trip on one of the main reactor coolant pumps.

Emergency events in a P^^ include (as illustrated in Figure 4.9) stuclr-^^^n pressure relief valves, a small break in the steam line, a small break in the pri­mary circuit inlet pipe, and a loss of flow on all the reactor coolant pumps.

Limiting faults (defined in Section 4.1) in a P^^ system are illustrated in Fig­ure 4.10 and include a large break in the outlet steam line, a large break in the inlet primary circuit pipe, a steam generator tube rupture, the seizing up of the rotor on one of the main coolant circulating pumps, and the failure of a control rod mechanism housing (a control rod ejection accident). Of these, perhaps the most famous and most widely considered is the primary circuit inlet pipe failure (the design base accident for the P^^).