• Nuclear industry
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• v.16 no.5 s.159
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• pp.72-83
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• 1996

한국원자력안전기술원은 최근 가압경수로형 원자로에서 발생하는 잔열제거펌프의 정지사고를 예방하는 방법을 개발하였다. 이 연구개발은 가압경수로형 원자로의 부분충수운전시 운전의 안전성을 위하여 냉각수의 최대 허용수위를 높이고 최저 요구수위를 낮출 수 있는 방안으로서, 우회유로를 이용한 획기적인 부분충수운전 개선방안이다. 한국원자력안전기술원은 이의 개발과 함께 미국 특허를 취득하였는데, 이는 외국의 기술을 받아들이면서 그의 문제점을 파악하여 해결책을 제시함으로써 국내 원전 관련 기술능력을 크게 제고시킨 것으로 평가된다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.10
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• pp.675-683
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• 2017

In this study, a theoretical analysis was performed to the flow of refilling stage in a safety injection tank, which is the core cooling system of nuclear power plant in an emergency. A theoretical model was proposed with a nonlinear governing equation defining on the flow of the refilling process of the coolant. Utilizing the Taylor-series expansion, the $1^{st}$ - order approximation flow equation was obtained, along with its analytic solution of closed type, which could predict accurately the variations of free surface height and flow rate of the coolant. The availability of theoretical result was confirmed by comparing with previous experimental results.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.185-191
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• 2016

The Hybrid Safety Injection Tank is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer(PZR). In this study, the condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically. The pressure balancing condition was set at the point where the velocity of the Hybrid SIT coolant injected into the Direct Vessel Injection(DVI) line was at or above zero. If the condition was derived from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel, the pressure difference between the pressurizer and SIT is less than 0.07 MPa.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.48-54
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• 1996

가압경수로의 부분충수 운전중 RHR 계통의 기능상실시 사고완화를 위해 가압기 manway와 증기 발생기 출구공동 manway를 동시에 개방한 경우에 대한 실험결과를 CATHAHR2 코드를 이용하여 해석하였다. 해석을 통해 이 경우에 발생하는 물리적 현상을 이해하고 이와 같은 과도기에 대해 코드 예측능력을 평가하므로 써, 실제 원전에서 사고시 적절한 사고대응 방안을 강구하는데 참고가 될 수 있도록 해석적 근거를 제시하고자 한다. 연구결과 CATHARE2 코드는 실험을 통해 관측된 주요 물리적 현상들을 타당하게 예측하였으나, 가압기와 밀림관의 DP를 과대 예측하여 원자로 상부공동의 최대압력을 실험보다 약 7kPa 높게 예측하였다. 노심 노출시간도 노심에서 기포율 분포를 비현실적으로 예측하여 실험보다 약 500초 지연되었다. 실험과 코드의 모의결과를 통하여 노심 노출은 중력주입에 의한 냉각수 보충만으로 충분히 회복될 수 있음을 확인하였다. CATHARE2 코드는 비록 상세한 현상들에 대해 다소 불확실성을 내포하였으나, 전반적인 거동분석에는 타당한 것으로 판단된다. CATHARE 코드는 노심에서 계면 마찰력을 줄임으로써 노심의 차압을 개선할 수 있었고, guide 튜브의 위치를 고온관 중심선과 일치시켜 guide 튜브내 액체의 hold-up 기간을 개선할 수 있었으며, 가압기의 계면 마찰력을 증가시켜서 밀림관에서 "plug and clearing" 현상을 모의할 수 있었다.모의할 수 있었다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.137-149
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• 1996

The present study is to analyze an integral test, BETHSY test 6.9c, which represent loss of RURS accident under mid-loop operation. Both the pressurizer manway and the steam generator outlet plenum manway are opened as vent paths in order to prevent the system from pressurization by removing the steam generated in the core. The main purposes are to gain insights into the physical phenomena and identify sensitive parameters. Assessment of capability of CATHARE2 prediction can be established the effective recovery procedures using the code in an actual plant. Most of important physical phenomena in the experiment could be predicted by the CATHARE2 code. The peak pressure in the upper plenum is predicted higher than experimental value by 7 kPa since the differential pressure between the pressurizer and the surge line is overestimated. The timing of core uncovery is delayed by 500 seconds mainly due to discrepancy in the core void distribution. It is demonstrated that openings of the pressurizer manwey and the steam generator manway can prevent the core uncovery using only gravity feed injection. Although some disagreements are found in the detailed phenomena, the code prediction is considered reasonable for the overall system behaviors.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.859-869
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• 1995

The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.