193 Emergency Core Cooling and Containment

The design features and operating procedures for a reactor are such that under normal conditions a negligible amount of radioactivity will get into

Emergency Core Cooling and Containment 275

the coolant and find its way out of the primary loop. Knowing that abnormal conditions can exist, the worst possible event, called a design basis accident, is postulated. Backup protection equipment, called engineered safety features, is provided to render the effect of an accident negligible. A loss of coolant accident (LOCA) is the condition typically assumed, in which the main coolant piping somehow breaks and thus the pumps cannot circulate coolant through the core. Although in such a situation the reactor power would be reduced immediately by use of safety rods, there is a continuing supply of heat from the decaying fission products that would tend to increase temperatures above the melting point of the fuel and cladding. In a severe situation, the fuel tubes would be damaged, and a considerable amount of fission products released. In order to prevent melting, an emergency core cooling system (ECCS) is provided in water-moderated reactors, consisting of auxiliary pumps that inject and circulate cooling water to keep temperatures down. Detailed analysis of heat generation and transfer is required in an application to the NRC for a license to operate a nuclear power plant (see References). The operation of a typical ECCS can be understood by study of some schematic diagrams.

The basic PWR reactor system (Fig. 19.8) includes the reactor vessel, the primary coolant pump, and the steam generator, all located within the containment building. The system actually may have more than one steam generator and pump-these are not shown for ease in visualization. We show in Fig. 19.9 the auxiliary equipment that constitutes the engineered safety (ES) system. First is the high-pressure injection system, which goes into operation if the vessel pressure, expressed in pounds per square inch (psi), drops from a normal value of around 2250 psi to about 1500 psi as the result of a small leak. Water is taken from the borated water storage tank and introduced to the reactor through the inlet cooling line. Next is the core flooding tank, which delivers borated water to the reactor through separate nozzles in the event a large pipe break occurs. Such a rupture would cause a reduction in vessel pressure and an increase in building pressure. When the vessel pressure becomes around 600 psi the water enters the core through nitrogen pressure in the tank. Then if the primary loop pressure falls to around 500 psi, the low-pressure injection pumps start to transfer water from the borated water storage tank to the reactor. When this tank is nearly empty, the pumps take spilled water from the building sump as a reservoir and continue the flow, through coolers that remove the decay heat from fission products. Another feature, the building spray system, also goes into operation if the building pressure increases above about 4 psi. It takes water from the borated water storage tank or the sump and discharges it from a set of nozzles located above the reactor, in order to provide a means for

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condensing steam. At the same time, the emergency cooling units of the reactor building are operated to reduce the temperature and pressure of any released vapor, and reactor building isolation valves are closed on unnecessary piping to prevent the spread of radioactive materials outside the building.

We can estimate the magnitude of the problem of removing fission product heat. For a reactor fueled with U-235, operated for a long time at power P0 and then shut down, the power associated with the decay of accumulated fission products is Pf(t), given by an empirical formula such as

For times greater than 10 seconds after reactor shutdown the decay is represented approximately by using A = 0.066 and a = 0.2. We find that at 10 s the fission power is 4.2% of the reactor power. By the end of a day, it has dropped to 0.68%, which still corresponds to a sizable power, viz., 20 MW for a 3000 MWt reactor. The ECCS must be capable of limiting the surface temperature of the zircaloy cladding to specified values; e.g., 2200°F, of preventing significant chemical reaction, and of maintaining cooling over the long term after the postulated accident.

The role of the steel-reinforced concrete reactor building is to provide containment of fission products that might be released from the reactor. It is designed to withstand internal pressures and to have a very small leak rate. The reactor building is located within a zone called an exclusion area, of radius of the order of half a kilometer, and the nuclear plant site is several kilometers from any population center.

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A series of experiments called Loss of Flow Tests (LOFT) was done at Idaho Falls to check the adequacy of mathematical models and computer codes related to LOCA/ECCS. A double-ended coolant pipe break was introduced and the ability to inject water against flow reversal and water vapor determined. Tests showed that peak temperatures reached were lower than predicted, indicating conservatism in the calculation methods.

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