• 대한기계학회논문집A
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• 제32권8호
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• pp.668-677
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• 2008

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which consideded safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.622-627
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• 1996

Detailed analysis of Feedwater Line Break (FLB) event for the fuel failure point of view are lack because the event was characterized as the increase in reactor coolant system (RCS) pressure. Up to now, the potential of the rapid system heatup case has been emphasized and comprehensively studied. The cooldown effects of FLB event is considered to be bounded by the Steam Line Break (SLB) event since the cooldown effect of SLB event is larger than that of the FLB event. This analysis provides a new possible path which can cause the fuel failure. The new path means that the fuel failure can occur under the heatup scenario because the Pressurizer Safety Valves (PSVs) open before the reactor trips. The 1000 MWe typical C-E plant FLB event assuming Loss of Offsite Power (LOOP) at the turbine trip has been analyzed as an example and the results show less than 1% of the fuel failure. The result is well within the acceptance criteria. In addition to that, a study was accomplished to prevent the fuel failure for the heatup scenario case as an example. It is found that giving the proper pressure gap between High Pressurizer Pressure Trip (HPPT) analysis setpoint and the minimum PSV opening pressure could prevent the fuel failure.

• Nuclear Engineering and Technology
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• 제45권1호
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• pp.73-80
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• 2013

The primary water stress corrosion cracking (PWSCC) of Alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and the pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of steam generators (SG) in Korea. The PWSCC resistance of Ni base alloys which have intergranular carbides is higher than those which have intragranular carbides. Conversely, in oxidized acidic solutions like sodium sulfate or sodium tetrathionate solutions, the Ni base alloys with a lot of carbides at the grain boundaries and shows less stress corrosion cracking (SCC) resistance. The role of grain boundary carbides in SCC behavior of Ni base alloys was evaluated and effect of intergranular carbides on the SCC susceptibility were reviewed from the literature.

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.198-204
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• 1987

발전소 과도현상과 비냉각재 상실사고를 모의할 수 있는 가압경수로발전소 모의코드 MCSIM을 개발하였다. 원자로 냉각재계통은 에너지 방정식과 운동량 방정식을 분리 취급하면서 Drift Flux 2상 유동모델, 적분 운동량 방정식 등을 사용하여 모델링하였다. 증기발생기의 모사는 Pot Boiler 모델을 사용하였고, 2차계통을 위해서는 분리 취급된 정상상태 에너지 방정식과 운동량방정식을 핵출력 계산을 위해서는 점 동특성 방정식을 사용하였다. 현재의 코드성능을 시험하기 위해 완전 냉각재 유동상실사고와 제어봉 집합체 인출 사고를 계산하여 그 결과를 원자력 5/6호기 최종 안전 보고서의 결과와 비교하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.437-443
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• 1997

DYLAM and its related applications are reviewed in detail and found to have many favorable characteristics. Concerning human factor analysis, the study demonstrates that DYLAM methodology represents an appropriate tool to study man-machine behavior provided that DYLAM is used to model machine behavior and an appropriate operator interface human factor model is included. A hybrid model which is a synthesis of the DYLAM model, a system thermodynamic simulation model and a neural network predicative model, is implemented and used to analyze dynamically the CANDU pressurizer system.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1995년도 추계학술발표회 초록집
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• pp.75-81
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• 1995

DYLAM and its related applications are reviewed in detail and found to have many favourable characteristics. Concerning human factor analysis, the study demonstrates that DYLAM methodology represents an appropriate tool to study man-machine behaviour provided that DYLAM is used to model machine behaviour and an appropriate operator interface human factor model is included. A hybrid model which is a synthesis of the DYLAM model, a system thermodynamic simulation model and a neural network predicative model, is implemented and used to analyse dynamically the CANDU pressurizer system.

• Nuclear Engineering and Technology
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• 제27권6호
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• pp.859-869
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• 1995

본 연구는 가압경수로의 부분충수 운전중 잔열제거기능 상실사고 해석시 신뢰성을 확보하기 위해 RELAP5/MOD3.1 코드로 관련 대형 실험을 모의 계산하여, 사고시 예상되는 주요 물리적 현상의 파악과 코드의 예측능력을 평가하는 것이다. 대상 실험으로 선택된 BETHSY Test 6.9a는 이 사고중 증기발생기가 작동하지 않고, 가압기 Manway를 개방한 상태 (Configuration)를 모의한 실험이다. 이 연구 결과는 실제 원전 사고시 예상되는 중요 현상 뿐 아니라, 이에 영향을 미치는 민감한 인자를 파악하여 사고 해석결과의 유효성을 판단하는 데 상당히 기여할 것으로 기대한다. 연구결과 RELAP5/MOD3.1 코드는 대체적으로 계통의 과도기 거동은 타당하게 예측하고 있지만, 모의계산에서 Time-Step이 아주 짧아 막대한 시간이 소요된다는 문제점이 발견되었다. 그 외에도 노심팽창수위 (swollen level)를 과대평가하여 가압기의 수위 및 계통의 압력을 높게 계산하였다. 이로 인해 가압기를 통한 방출량도 과대계산하여 노심노출을 약 400초 빨리 예측하였다. 실험과 코드 예측결과를 종합할 때 가압기 Manway 만의 개방으로는 계통압력이 상승하고, 중력주입냉각수로는 노심수위 회복에 불충분하며, 결국 강제주입에 의해서 노심수위가 회복될 수 있음이 입증되었다.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.37-50
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• 1993

완전급수상실사고 및 복구가정의 사고전개와 열수력학적 거동을 평가하기 위해 RELAP5/MOD3 계산을 수행하고 LOFT L9-l/L3-3 실험 결과와 비교하였다. 또한 본 사고의 주요 열수력 현상에 대한 코드의 예측도를 평가하였다. 본 연구의 결과로서 가압기 수위가 만수위에 도달할 때까지 살수를 이용한 압력 제어, 가압기 압력방출 밸브를 통한 가압방지, 증기발생기 재충수에 의한 이차측 열제거 능력의 재확보, 지속적인 자연순환에 의한 효과적인 일차계통의 냉각등이 이루어 질 수 있고 이 과정중 노심노출은 나타나지 않음이 밝혀졌다. 또한 현재의 현상-중심 비상운전절차서 특히 과압방지성능 및 증기발생기 회복절차 등의 유효성을 입증하기 위해서는 발전소 고유한 평가가 필요함이 밝혀졌다.

• 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.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.