• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.476-485
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• 1999

In order to evaluate the gravity-injection capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Yong Gwang Units 3&4 were reviewed. The six cases of possible gravity-injection paths from the refueling water tank (RWT) were identified and the thermal-hydraulic analyses were performed using the RELAP5/MOD3.2 code. The core cooling capability was significantly dependent on the gravity-injection path, the RCS opening, and the injection rate. In the cases with the pressurizer manway opening higher than the RWT water level, the coolant was held up in the pressurizer and the system pressure continued increasing after gravity-injection. The gravity injection eventually stopped due to the high system pressure and the core was uncovered. In the cases with the injection path and opening on the same leg side, the core cooling was dependent on whether the water injected from the RWT passed the core region or not. However, in the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. In addition, from the sensitivity study on the gravity-injection flow rate, it was found that about 54 kg/s of injection rate was required to maintain the core cooling and the core cooling could be provided for about 10.6 hours after event with that injection rate from the RWT. Those analysis results would provide useful information to operators coping with the event.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.647-652
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• 1998

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identified the possible even scenarios following the loss of shutdown cooling. The Thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior, From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determined the containment closure time to prevent the uncontrolled released of fission products to atmosphere, These data provide useful information to the abnormal procedure to cope with event.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.258.2-258
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• 2001

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.1-11
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• 1989

Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.531-540
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• 1998

The DYLAM-3 code which overcomes the limitation of event tree/fault tree was applied to LOOP (Loss of Off-site Power) in the mid-loop operation employing HEPs (Human Error Probabilities) supplied by the ASEP (Accident Sequence Evaluation Program) and the SEPLOT (Systematic Evaluation Procedure for Low power/shutdown Operation Task) procedure in this study. Thus the time history of core uncovery frequency during the mid-loop operation was obtained. The sensitivity calculations in the operator's actions to prevent core uncovery under LOOP in the mid-loop operation were carried out. The analysis using the time dependent HEP was performed on the primary feed & bleed which has the most significant effect on core uncovery frequency. As the result, the increment of frequency is shown after 200 minutes duration of simulation conditions. This signifies the possibility of increment in risk after 200 minutes. The primary feed & bleed showed the greatest impact on core uncovery frequency and the recovery of the SCS (Shutdown Cooling System) showed the least impact. Therefore the efforts should be taken on the primary feed & bleed to reduce the core uncovery frequency in the mid-loop operation. And the capability of DYLAM-3 in applying to the time dependent concerns could be demonstrated.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.257-264
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• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)