Nuclear reactor, accidents transient cooling

Nuclear power plant systems may be classified as "Frontline" and "Support. . iccurding to their. service in an accident. Frontline systems are the engineered safety systems that deal directly with an accident. Support systems support the frontline systems. Accident initiators are broadly grouped as loss of cooling accidents (LOCAs) or transients. In a LOCA, water cooling the reactor is lost by failure of the cooling envelope. These are typically classified as small-small (SSLOCA), smalt (SLOCA), medium (MLOCA) and large (LLOCA). 

The appropriate combination of inherent safety features and engineered ones should be necessary to prevent the extension of anticipated transients and also postulated accidents within the system. As an example of inherent safety features, one should mention Doppler reactivity that is effective against criticality events, which is a basis to ensure the reactor shut down function. Natural circulation is also an inherent safety feature to enhance the cooling capability of the reactor core. The nuclear energy system should be designed in such a way that severe core damage leading to the release of massive amounts of radioactive materials could be avoided. 

SAPHER, D., DZODZO, M, GREENSPAN, E., Applying the DSNP modular modelling system to transient and accident simulations of lead cooled reactors, ICONE-9 (Proc. 9 Int. Conf on Nuclear Engineering, Nice, France, April 8-12, 2001), ICONE9-661, ASME. 

In the event of a LOCA, the nuclear reaction in the core would be automatically cut off by the loss of the moderation associated with the coolant, while in the transient case the probability of failure of the shutdown systems is sufficiently low that continued operation at power is not a significant contributor to a core melt situation. For both classes of accident, therefore, the important requirement is the maintenance of a cooling capability sufficient to remove decay heat (see Table 12.7) from the reactor core. The emergency core cooling systems (ECCS) for the PWR and BWR are described below. 

Y. Okano, S. Koshizuka, K. Kitoh and Y. Oka, Flow-Induced Accident and Transient Analyses of a Direct-Cycle, Light-Water Cooled, Fast Breeder Reactor Operating at Supercritical Pressure, Journal of Nuclear Science and Technology, Vol. 33(4), 307-315 (1996)... 

K. Kitoh, S. Koshizuka and Y. Oka, Pressure and flow-induced accident and transient analyses of a direct-cycle, supercritical-pressure, light-water-cooled fast reactor, Nuclear Technology, Vol. 123, 233-244 (1998)... 

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