• Nuclear Engineering and Technology
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• 제27권4호
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• pp.470-482
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• 1995

2025 MWt 가압경수로인 울진 3, 4호기에는 설계기준초과사고인 완전급수상실사고를 완화하기 위하여 안전감압계통이 채택되었다. 본 논문은 울진 3, 4호기의 안전감압계통의 방출유량을 결정하기 위한 해석방법 및 결과에 대하여 논의하였다. 안전감압계통의 방출용량을 다음과 같은 두가지의 설계요건에 따라 결정하였다 : 1) 두 개의 고압안전주입펌프 중 하나의 펌프만이 작동하고 운전원이 안전감압계통의 한 계열의 감압경로를 가압기안전밸브가 열리자마자 개방하였을 경우 노심노출을 방지하여야 한다 2) 두 개의 고압안전주입펌프가 모두 작동하고 두 계열의 안전감압경로를 가압기안전밸브가 열린 후 30분 뒤에 개방하였을 경우 노심노출을 방지하여야 한다. CEFLASH-4AS/REM 전산코드의 모델 및 해석 결과의 타당성을 검토하기 위하여 REL-AP5/MOD3를 이용한 해석을 수행하였다. 운전원의 복구과정이 없을 경우와 운전원이 충전 및 유출운전에 의해 사고를 완화하는 경우의 완전급수상실사고 경위에 대해 수치모사를 수행하였다. 두 사고 경 위에 대해 CEFLASH-4AS/REM에 의해 예측된 원자로계통의 주요 열수력학적 거동이 RELAP5 /MOD3에 의한 결과와 정성·정량적으로 잘 일치하는 것을 알 수 있었다. 결론적으로 울진 3, 4호기에 대해 완전급수상실사고시 안전감압계통을 이용한 충전 및 유출운전에 의해 잔열제거 및 일차계통냉각재 재고량 유지가 성공적으로 이루어짐을 수치모사를 통해 확인 할 수 있었다.

• Journal of Mechanical Science and Technology
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• 제20권7호
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• pp.1030-1042
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• 2006

An experimental study has been performed to investigate the off-take phenomena at the header-feeder systems (horizontal header pipe with multiple feeder branch pipes) in a CANDU (CANadian Deuterium Uranium) reactor with the branch orientation varies ${\pm}36^{\circ}\;or\;{\pm}72^{\circ}$. In order to evaluate the applicability of the conventional correlations used in the safety analysis code, RELAP5-Mod3, the test facility is designed with the 1/2 scale of the. CANDU 6. It was found that the data set for the top, bottom and side branches are in a good agreement with the correlations used. However, for the specific angled branches, ${\pm}36^{\circ}\;and\;{\pm}72^{\circ}$, the onsets of off-take data and quality data showed large deviation with the conventional model inside RELAP5-MOD3. Furthermore, based on the uncertainty analysis, the conventional 2.5 power law needs to be modified. The present experimental data set can be useful for the construction of the general correlation considering the arbitrary branch orientation.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• Nuclear Engineering and Technology
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• 제31권5호
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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.

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.58-74
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• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• 한국안전학회지
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• 제27권5호
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• pp.229-234
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• 2012

Operator's action time is evaluated from MAAP4 analysis used in conventional probabilistic safety assessment(PSA) of a nuclear power plant. MAAP4 code which was developed for severe accident analysis is too conservative to perform a realistic PSA. A best-estimate code such as RELAP5/MOD3, MARS has been used to reduce the conservatism of thermal hydraulic analysis. In this study, operator's action time of core cooling recovery operation is evaluated by using the MARS code, which its Fussell-Vessely(F-V) value was evaluated as highly important in a small break loss of coolant(SBLOCA) event and loss of component cooling water(LOCCW) event in previous PSA. The main conclusions were elicited : (1) MARS analysis provides larger time window for operator's action time than MAAP4 analysis and gives the more realistic time window in PSA (2) Sufficient operator's action time can reduce human error probability and core damage frequency in PSA.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.17-28
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• 1999

A thermal-hydraulic analysis is conducted on the loss-of-coolant-accident (LOCA) during shutdown operation of YGN Units 3/4. Based on the review of plant-specific characteristics of YGN Units 3/4 in design and operation, a set of analysis cases is determined, and predicted by the RELAP5/MOD3.2 code during LOCA in the hot-standby mode. The evaluated thermal-hydraulic phenomena are blowdown, break flow, inventory distribution, natural circulation, and core thermal response. The difference in thermal-hydraulic behavior of LOCA at shutolown condition from that of LOCA at full power is identified as depressurization rate, the delay in peak natural circulation timing and the loop seal clearing (LSC) timing. In addition, the effect of high pressure safety injection (HPSI) on plant response is also evaluated. The break spectrum analysis shows that the critical break size can be between 1% to 2% of cold leg area, and that the available operator action time for the Sl actuation and the margin in the peak clad temperature (PCT) could be reduced when considering uncertainties of the present RELAP5 calculation.

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.94-107
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• 2022

Accidents prevention and mitigation is the highest priority of nuclear power plant (NPP) operation, particularly in the aftermath of the Fukushima Daiichi accident, which has reignited public anxieties and skepticism regarding nuclear energy usage. To deal with accident scenarios more effectively, operators must have ample and precise information about key safety parameters as well as their future trajectories. This work investigates the potential of machine learning in forecasting NPP response in real-time to provide an additional validation method and help reduce human error, especially in accident situations where operators are under a lot of stress. First, a base-case SGTR simulation is carried out by the best-estimate code RELAP5/MOD3.4 to confirm the validity of the model against results reported in the APR1400 Design Control Document (DCD). Then, uncertainty quantification is performed by coupling RELAP5/MOD3.4 and the statistical tool DAKOTA to generate a large enough dataset for the construction and training of neural-based machine learning (ML) models, namely LSTM, GRU, and hybrid CNN-LSTM. Finally, the accuracy and reliability of these models in forecasting system response are tested by their performance on fresh data. To facilitate and oversee the process of developing the ML models, a Systems Engineering (SE) methodology is used to ensure that the work is consistently in line with the originating mission statement and that the findings obtained at each subsequent phase are valid.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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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.