Abstract
A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.
Highlights
Integral pressurized water reactor (IPWR) is being considered as one of the next-generation advanced nuclear reactors designed to be inherently safe by naturally and physically passive mechanisms
passive residual heat removal system (PRHRS) is expected to safely remove the core decay heat, only through natural circulation, in case of both station blackout accident and long-term cooling for repair or refueling
It can be seen that the core power and the heat transferred to the secondary side of the steam generator decrease rapidly at the initial stage and at about 20 seconds they begin to decrease slowly
Summary
Integral pressurized water reactor (IPWR) is being considered as one of the next-generation advanced nuclear reactors designed to be inherently safe by naturally and physically passive mechanisms. The primary coolant system components of the IPWRs, composed of the core, the pressurizer, the main coolant pumps (MCPs), and the once-through steam generators (OTSGs), are housed in the reactor pressure vessel (RPV). One of the very important design features of the IPWRs is the simplifications and improvements in the safety systems. Such passive safety systems as passive residual heat removal system (PRHRS) are employed to accomplish the inherent safety functions and mitigate the consequences of the postulated accidents. PRHRS is expected to safely remove the core decay heat, only through natural circulation, in case of both station blackout accident and long-term cooling for repair or refueling
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