As we saw in Chapter 7, the spent fuel from the reactor may be stored under­water in cooling ponds for typically 10 years or more. The current stage of de­velopment of the nuclear program in Europe, the United States, and Japan is such that final decisions about the next phase of spent fuel management, namely engineered surface storage, need to he taken in the next few years.

It would be possible, for instance, to continue with away-from-reactor un­derwater storage, perhaps with the fuel contained in an additional “bottle" to prevent the spread of contamination within such a large water basin store.

Alternatively, a dry storage system could he adopted. Essentially, two differ­ent dry storage systems have been developed: the cask or container system and the modular vault diy store. In Canada development of cliy store containers for CANDU fuel has been in process for 20 years. The latest design of dry storage container is shown in Figure 8.4. It consists of a box 25 m x 2 m x 3 5 m high, constructed from inner and outer steel shells filled with heavy concrete. It weighs 53 tons and contains some 384 CANDU spent fuel bundles: total mass,

7.3 tons. These are stored horizontally in four racks. The Canadian nuclear sta­tion at Pickering near Toronto will ultimately have 700 such containers storing nearly 5000 tons of fuel, making it, when complete the world’s largest dry store. For more highly rated PWR spent fuel Germany has developed a container using ductile cast iron (CASTOR) 2.4 m diameter. x 4.8 m high, weighing 100 tons when loaded and containing either 33 PVR spent fuel elements or 74 spent fuel elements—15 tons of spent fuel.

An alternative dry storage system is the modular vault dty store illustrated in Figure 8.5. In this concept the spent fuel is contained in individual vertical

sealed fuel storage tubes retained within a concrete vault that can be con­structed in modules. Air is drawn in by natural circulation between the array of storage tubes and is discharged via an outlet duct. The coolant air does not come in contact with the spent fuel and therefore neither it nor the concrete structure becomes contaminated. A facility of this type has been constructed to store the spent fuel from the gas-cooled reactor at Fort St. ViJ. in. ‘This particular

lesign has 45 fuel storage tubes, each holding six fuel elements. The fuel is noved in and out of the storage tubes by a fuel handling machine moved by he building crane. A modular vault dry store has been considered by Scottish..Juclear for dry storage of AGR fuel. Designs for both an 800-ton and a 1200-ton :apacity have been prepared.

Storage in these dry stores would continue for 50-100 years, during which he level of radioactivity gradually decays (Figure 8.2), as does the rate of heat )roduction (Figure 7.5). Surface storage in this form for extended periods is ad­vantageous since natural convection cooling can be arranged and the packages monitored systematically. Ultimately, the rate of heat generation will become low enough to permit storage without special arrangements for natural convec­tion cooling. At this stage long-term disposal may be considered.

The concepts being considered for ultimate disposal of spent or unre­processed fuel include disposal to underground salt formations or within hard rock geological formations.

2.1 INTRODUCTION

In Section 1.4 we briefly introduced the fission process and explained that it leads to the generation of heat within the nuclear fuel. This heat can be used to generate electrical energy in a nuclear power station. In this chapter we shall further explore this heat generation process and discuss the aspects of nuclear reactor design concerned with removing and utilizing the heat.

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^^).

The next generation of nuclear power plants will incorporate design features that

will eliminate or reduce the challenges to the various containment barriers or mit­igate the consequences of failure. One example is the design features included on

the European pressurized water reactor (EPR). The EPR design includes:

• The elimination of situations where the degradation of the core occurs with the primaiy circuit still at high pressure. This is achieved by high-reliability secondaiy side-decay heat removal systems but also by means of rapid depressurization via the pressurizer relief valves.

• The elimination of direct containment heating via the depressurization facility.

• The limitation of the containment pressure increase using a dedicated spray heat removal system that can subcool the water and return the pressure to atmospheric. The containment design pressure of 7.5 bars allows 12-24 hours after the accident before it is necessary to use the spray system.

• The provision of a double-wall containment with collection of all leaks in the interwall space where a lower pressure is maintained.

• The prevention of hydrogen explosions by reducing the hydrogen concentra­tion using catalytic recombiners together with selectively placed igniters.

• Accommodation of the consequences of an instantaneous full cross section rnpture of the reactor pressure vessel at a pressure of 20 bars via careful design of the layout.

• Provision to cope with molten fuel coming from a failed pressure vessel lower head, first, without a “steam explosion” and, second, preventing interaction with the containment concrete.

This is accomplished as shown in Figure 6.2 by connecting the reactor cavity

to a dedicated molten core spreading chamber via a refractory lined melt dis­charge channel. The spreading chamber has a large area (150 m2) and is nor­mally sealed from the reactor cavity by a steel plate. This plate resists melt-through for a limited time in order to accumulate the molten fuel in the cavity. The spreading compartment is connected via pipes to the refueling water storage tank in the containment. These pipes are normally closed by fusible plugs. This ensures that the water floods the spreading chamber only after the melt has been spread over the area of the chamber.