Simultaneously, void generation and increasing moderator temperature give rise to further development of negative reactivity in core. These minus reactivity acts to restrain the increase in reactor power. This feature of the PWR, intrinsically restraining the power increase due to a disturbance, is called self-regulation and is an important safety characteristic. The LWR core is designed in a way that the power coefficient is always negative. Also as already mentioned, if the reactor condition approaches the acceptable fuel design limit, the reactor will be automatically shut down. The ability to bring the reactor from the critical condition into the subcritical condition within a very short time, is called the shutdown capability of the reactor. The core design provides sufficient shutdown capability to maintain the core under hot shutdown condition even when one of the control rods of the highest worth remains fully out of the core. Moreover, during normal operation, control rods are handled sequentially depending on their function either for control or for safety, with monitoring of their insertion limits.

3. Reactor Coolant Pressure Boundary The reactor coolant pressure boundary which has the same pressure condition of the connoting the reactor and the primary coolant, and causes a loss of coolant accident (LOCA), when breached, acts as a pressure barrier during the normal operation, abnormal transients and abnormal conditions, the coolant pressure boundary, next to the fuel cladding, also acts as a barrier to FP release. If this boundary is breached, the coolant will flow out of the reactor system leading to trouble in removing the decay heat of the reactor core and therefore, it is a particularly important part of the reactor system.

In order to minimize the failure probability of the equipment and piping used in the reactor coolant pressure boundary system and to bring the possibility of occurrence of unusual coolant leakage to an extremely low degree, items such as appropriate material selection, earthquake-resistant design and repression of system over-pressurization are closely considered. Additionally, the system design considers capability to accept in-service inspections during the refueling operation and other reactor outage periods. The boundary system is

also designed and manufactured to have sufficient strength to withstand all design basis transients postulating any kinds of accidents.

Attention is paid in both design and operation to fracture toughness of the ferric steel-made components comprising the reactor coolant pressure boundary system, in order to prevent brittle behavior and occurrence of rapidly propagating fracture under any conditions during normal operation and accidents. For example, when the temperature of the primary coolant system is rising or falling, in addition to confirming the safe temperature domain, an appropriate rate of heating or cooling is also decided in a way to avoid severe operation conditions.

In particular, the increase in ductile-brittle transition temperature (DBTT) of the RV due to irradiation by fast neutrons is thoroughly considered in its material selection, design and manufacturing. Furthermore, test specimens of RV materials, inserted in capsules, are placed in the RV around the core and based on a planned schedule, are removed from the reactor for mechanical tests to confirm safe operation conditions.

In general, if any small crack appears, in piping for example, it is considered as the cause of a detectable leakage and appropriate measures can be taken before the crack gets larger which could result in a serious rupture. Since leakage detection at an early stage is an important factor in preventing accident escalation, any leakage of primary coolant from the reactor coolant pressure boundary is quickly detected by different types of detection systems such as radiation monitors, sump water level gages and condensed water level gages.

4. Engineered Safeguard Systems

Accidents in which the reactor coolant pressure boundary is breached incurring a substantial loss of coolant inventory, are called LOCAs. In such an accident, the ceramic fuel could be considerably damaged resulting in release of large amounts of radioactive materials. In order to prevent LOCAs or mitigate their consequences, engineered safeguard systems, which consist of the emergency core cooling system (ECCS), reactor containment system and containment spray system, are provided to prevent and mitigate such accidents.

NSRA, Japan

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