• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3236-3246
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• 2021

Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.