• Journal of Energy Engineering
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• v.26 no.3
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• pp.71-77
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• 2017

In the event of loss of coolant accident (LOCA) and station black out (SBO) in the primary system of a nuclear reactor, the coolant water should be injected to reactor coolant system (RCS) without any intervention of operators or active components. To satisfy the requirements, hybrid safety injection tank (Hybrid SIT) was suggested by Korea Atomic Energy Research Institute (KAERI). The pressure equalizing time of Hybrid SIT is an important parameter to determine the timing of coolant injection. To predict the pressure equalizing time of the Hybrid SIT, a separate effect test facility was constructed and sensitivity tests were conducted in various conditions. The main parameter determining the pressure equalizing time was obtained from separate effect test (SET) results. The wall of condensation on the inner wall of SIT and direct contact condensation on the water surface affected to the pressure equalizing time very much. In this study, the effect of each condensation phenomena on pressure equalizing time was quantitatively analyzed from results of SET and a prediction method of pressure equalizing time was proposed.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.133-142
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• 2007

The Electric Power Research Institute (EPRI) demonstrated a method for monitoring the performance of instrument channels in Topical Report (TR) 104965, 'On-Line Monitoring of Instrument Channel Performance.' This paper presents the results of several models originally developed by EPRI to monitor three nuclear plant sensor sets: Pressurizer Level, Reactor Protection System (RPS) Loop A, and Reactor Coolant System (RCS) Loop A Steam Generator (SG) Level. The sensor sets investigated include one redundant sensor model and two non-redundant sensor models. Each model employs an Auto-Associative Kernel Regression (AAKR) model architecture to predict correct sensor behavior. Performance of each of the developed models is evaluated using four metrics: accuracy, auto-sensitivity, cross-sensitivity, and newly developed Error Uncertainty Limit Monitoring (EULM) detectability. The uncertainty estimate for each model is also calculated through two methods: analytic formulas and Monte Carlo estimation. The uncertainty estimates are verified by calculating confidence interval coverages to assure that 95% of the measured data fall within the confidence intervals. The model performance evaluation identified the Pressurizer Level model as acceptable for on-line monitoring (OLM) implementation. The other two models, RPS Loop A and RCS Loop A SG Level, highlight two common problems that occur in model development and evaluation, namely faulty data and poor signal selection

• 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.27 no.2
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• pp.165-175
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• 1995

The overpressure protection system for PWR should be designed with sufficient capacity to limit the pressure to less than 110％ of the reactor coolant system design pressure during the most severe abnormal operational transient. In this study, the feasibility of adopting the SEBIM POSRV instead of the current spring loaded pop-opening safety valves to the ABB-CE designed 2825 MWt PWR is investigated for its overpressure protection capability. The required SEBIM POSRV size as well as its opening/closing setpoints are determined through a series of computer analyses using the LTC code which has been used for the overpressure protection analysis for Yonggwang units 3&4. The analysis results show that the overpressure protection system with monobloc SEBIM POS-RV can maintain the RCS pressure below 110％ of the design pressure demonstrating its overpressure protection capability for the ABB-CE designed 2825 MWt PWRs.

• 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)

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.29-35
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• 1996

For the Reactor Coolant System(RCS) flow rate measurement by the secondary calorimetric heat balance method, the coolant temperature of the hot leg is needed. Several Resistance Temperature Detectors(RTD) are installed in the hot leg to measure the temperature, but the average value of RTDs does not correctly represent the energy-averaged(bulk) temperature because of the thermal stratification phenomenon. Therefore some correction is introduced to predict the bulk temperature, but the correction inevitably contains uncertainty because the stratification is not defined well quantitatively yet. Therefore a large uncertainty for the correction has been used for the conservative estimation. But unrealistically large uncertainty causes degradation of the measurement method and yields difficulty to meet the acceptance criterion in start-up flow measurement test. In this paper, an analytical estimation is made on the correction and the related uncertainty using the measured hot leg velocity profile of System 80 reactor flow model test and the measured temperatures of YGN 3&4 and PVNGS 1&2 start-up tests. The results reveal that the magnitude of the correction uncertainty is much smaller than that used in the previous design. Therefore, the confidence on the flow rate measurement method can be improved and the difficulty in start-up flow measurement test can be lessened if the smaller correction uncertainty obtained through this estimation is applied.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.