ECCS performance analysis for the integral type PWR

Full text: Full text: For the purpose of seawater desalination and electricity generation the Korea Atomic Energy Research Institute (KAERI) completed the basic design development of an advanced Integral-type Pressurized Water Reactor (IPWR) producing a maximum thermal power of 330MW in March 2002. This plant was designed to produce 40,000 m3/day of potable water by using approximately 10% of the total energy produced and generate about 90 MWe by using the remaining energy. To demonstrate the enhanced performance and safety of the 330 MWt IPWR, a 1/5 scaled-down pilot plant, 65 MWt IPWR, has been developed in Korea. Different from the conventional loop type PWRs, the 65 MWt IPWR contains the reactor coolant and the major primary circuit components, such as the core, two main coolant pumps, 12 steam generator cassettes, and a pressurizer in a single Reactor Pressure Vessel (RPV). Due to this integral arrangement of the primary system the possibility of a large break Loss of Coolant Accident (LOCA) is inherently eliminated and only the small branch line break or leak through a component penetrating the RPV is postulated. Also, the 65 MWt IPWR adopts inherent safety improving features such as a large volume of primary coolant(volume/unit power), substantially large negative moderator temperature coefficients without a soluble boron, a low core power density, a self-controlled N2 gas pressurizer, a canned motor main coolant pump without a pump seal, and a modular helically coiled once-through steam generator cassette. In addition, the 65 MWt IPWR enhances its safety and reliability by adopting the redundant and independent safety systems such as the four independent trains of Passive Residual Heat Removal System (PRHRS) and Safety Injection System (SIS). The PRHRS removes the core decay heat for emergency situations where a normal steam extraction or feedwater supply is unavailable. When the PRHRS actuation signal is generated, the main steam and feedwater isolation valves are closed and the PRHRS isolation valves are opened to connect to the secondary side of the steam generator. Afterwards, the PRHRS cools down the Reactor Coolant System (RCS) by a natural circulation. The SIS compensates for the primary coolant inventory loss in the small break LOCA. To assess the performance of the safety system of the 65 MWt IPWR, small break LOCA break spectrum analyses are performed by using the TASS/SMR code which is a thermal-hydraulic system analysis code implementing the evaluation models required by Appendix K of 10 CFR 50. A guillotine rupture for various branch lines (break inner diameter: 1), a complete rupture of the CEDM housing accompanied by a CEA ejection (equivalent inner diameter: 0.97) and a full inadvertent opening of a POSRV (valve inner diameter: 0.7) are considered in this paper. The analysis results by using conservative initial and boundary conditions and assumptions show that the actuations of the SIS and the PRHRS maintain the RPV coolant level well above the top of the core and adequately remove the long term core decay heat for all cases of the small break LOCAs. Thus, the emergency core cooling performance after a postulated small break LOCA in the 65 MWt IPWR can be assured