Emergency Core Cooling System

The Emergency Core Cooling System (ECCS) is composed of a series of systems which are designed to safely shutdown a nuclear reactor during accident conditions. Under normal conditions heat is removed from a nuclear reactor by condensing steam after it passes through the turbine. The condensed steam (water) is then fed back into the reactor in a BWR or back through the heat exchanger in a PWR which keeps the reactor core at a constant temperature. During an accident the condenser is not used so alternate methods of cooling are required to prevent damage to the nuclear fuel.

Systems

In most plants ECCS is composed of the following systems:

High Pressure Coolant Injection System (HPCI)

This system consists of a pump or pumps which have sufficient pressure to inject coolant into the reactor vessel while it is pressurized. It is designed to monitor the level of coolant in the reactor vessel and automatically inject coolant when the level drops below certain setpoints. This system is normally the first line of defense for a reactor since it can be used while the reactor vessel is still pressurized.

Depressurization System

This system consists of a series of valves which open to vent steam into a primary containment structure which depressurizes the reactor vessel and allows lower pressure coolant injection systems to function.

Low Pressure Coolant Injection System (LPCI)

This system consists of a pump or pumps which inject additional coolant into the reactor vessel once it has been depressurized.

Internal/Core Spray System

This system consists of a series of pumps and spargers (special spray nozzles) which spray coolant into the primary containment structure. It is designed to condense the steam in primary containment structure to prevent it from rupturing.

These systems allow the plant to respond to a variety of accident conditions and at the same time creates redundancy so that the plant can still be shutdown even if one or more of the systems fails to function.

References

American National Standard, ANSI N18.2, “Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,” August 1973. IEEE 279, “Criteria for Protection Systems for Nuclear Power Generating Stations.”