• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 춘계 학술발표회 제16권1호
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• pp.412-415
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• 2004

This study was performed to verify the behaviors of fiber Bragg grating (FBG) sensors attached to the containment structure in the nuclear power plant as a part of structural integrity test which demonstrates that the structural response of the non-prototype primary containment structure is within predicted limits plus tolerances when pressurized to $115\%$ of containment design pressure, and that the containment does not sustain any structural damage.

• 방사성폐기물학회지
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• 제18권2_spc호
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• pp.237-245
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• 2020

사용후핵연료를 저장하는 볼트체결 저장용기의 격납경계를 형성하는 주된 구성요소는 금속 밀봉재이다. 이러한 금속 밀봉재는 열과 방사선에 의해 그 성질이 저하된다. 또한, 금속 밀봉재가 강한 열에 장기간 노출되면 크리프 현상이 발생한다. 이러한 크리프는 밀봉시스템에 응력 이완을 가져와서, 결국 밀봉재의 건전성을 해치게 된다. 이러한 응력 이완은 금속 밀봉재의 밀봉성능 저하로 이어지며, 저하의 정도에 따라 저장용기의 누설을 야기할 수 있다. 또한, 볼트 체결력의 감소도 밀봉성능 저하에 영향을 미친다. 본 논문에서는 금속 밀봉재의 격납건전성과 볼트체결력 감소를 평가하기 위해 수행한 가속화 시험의 결과에 대하여 기술한다. 전 시험기간 동안 각 시편에서의 누설률, 볼트 변형률, 금속 밀봉재 주변 온도를 계측하여 분석하였고, 금속 밀봉재는 저장기간 50년 동안 격납건전성을 유지함을 입증하였다. 또한, 가속화 시험의 타당성에 대해서 기술하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기설비전문위원
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• pp.73-75
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• 2005

A number of the power cables arc installed in the containment of the nuclear power plant. According to the IEEE Standard 835, the calculation of the temperature rise shows the operation possibility of power cables in the containment. In this paper, we expect the integrity of the power cables by using the calculation of the temperature rise and the development of the lifetime extension of the cables.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.69-74
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• 2023

The paper presents the results investigating the effect of aircraft engine impact location on the intended function evaluation results of spent nuclear fuel transport cask. As a result of the investigation, it is found that the structural integrity is maintained as the maximum accumulated equivalent plastic strain is below the acceptable criterion regardless of the collision location. It is identified that when the aircraft engine collided with the upper part of the transport cask without considering impact limiter the containment performance is weakened compared to when the aircraft engine collided with the central part.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.597-604
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• 2013

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 추계 학술발표회 제16권2호
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• pp.643-646
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• 2004

The posttensioning system of nuclear containment have to be verified its structural integrity by the periodic inspection because the structural behavior of the containment is changed by the variation of the physical property of concrete and tendon as time passes. In this study a program 'SAPONC-CANDU' which is able to monitor and analysis the micro structural behavior of the domestic CANDU type containment at all times was developed. The readings of vibrating-wire strain gauges embedded into the concrete of containment were used as input data for operating the program. This program provides the long-term prediction values and bands of the concrete strain due to the time dependent factors of the concrete and tendon of the domestic CANDU type containment.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.468-482
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• 2001

As a process of Direct Containment Heating (DCH) issue resolution for Korean Standard Nuclear Power Plants (KSNPs), a containment load/strength assessment with two different approaches, the probabilistic and the deterministic, was performed with all plant-specific and phenomena-specific data. In case of the probabilistic approach, the framework developed to support the Zion DCH study, Two-Cell Equilibrium (TCE) coupled with Latin Hypercubic Sampling (LHS), provided a very efficient tool to resolve DCH issue. In case of the deterministic approach, the evaluation methodology using the sophisticated mechanistic computer code, CONTAIN 2.0 was developed, based on findings from DCH-related experiments or analyses. For three bounding scenarios designated as Scenarios V, Va, and VI, the calculation results of TCE/LHS and CONTAIN 2.0 with the conservatism or typical estimation for uncertain parameters, showed that the containment failure resulted from DCH loads was not likely to occur. To verify that these two approaches might be conservative , the containment loads resulting from typical high-pressure accident scenarios (SBO and SBLOCA) for KSNPs were also predicted. The CONTAIN 2.0 calculations with boundary and initial conditions from the MAAP4 predictions, including the sensitivity calculations for DCH phenomenological parameters, have confirmed that the predicted containment pressure and temperature were much below those from these two approaches, and, therefore, DCH issue for KSNPS might be not a problem.

• 센서학회지
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• 제22권1호
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• pp.71-75
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• 2013

Nuclear containment building is used as second blockage to protect us from a radiation leakage caused by the natural disaster or any accidents, so it's safety is important and must be kept with continuous surveillance. In this study, we measured the strain of a nuclear containment building's wall by using FBG sensor and investigated the structural safety of a nuclear containment building. 50 FBG strain sensors and 18 FBG strain sensors were attached on the side wall and upper dome of a nuclear containment building, respectively. We measured the strains of the outside concrete wall during the Structural Integrity Test (SIT) of a nuclear containment building. The strain of an upper dome was larger than that of a side wall, about $200{\mu}{\varepsilon}$. And the very small strain was measured at vertical direction of a side wall. These experimental results were used to evaluate the structural health of nuclear containment building.

• 대한토목학회논문집
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• 제26권5A호
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• pp.901-908
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• 2006

안전성 관련 구조물인 원자력 격납건물은 시간의 흐름에 따라 콘크리트와 텐던의 물리적 성질 변화로 구조거동의 미세한 변화를 가져오기 때문에 주기적 점검을 통한 구조건전성 검증이 필요하다. 본 연구에서는 국내 부착식 텐던 격납건물인 CANDU형의 월성 원전을 대상으로 미세 구조거동 분석이 가능한 'SAPONC-CANDU' 프로그램을 개발하였으며, 이는 온도와 시간종속성 영향인자들 즉, 크리프, 건조수축, 텐던의 인장력 하에서 격납건물 콘크리트 속에 매립되어 있는 진동식 와이어 변형률 게이지의 변형률 변화량에 대한 예측값을 계산하는 알고리즘에 기초한다. 개발된 프로그램의 구동을 위해서 변형률 게이지의 계측값이 입력데이타로 사용되고 최종적으로 각각의 변형률 게이지에 대해서 변형률 변화량의 예측값, 예측선, 예측폭이 그래프 형태로 제공되기 때문에 국내 원자력발전소 CANDU형 격납건물의 구조건전성을 평가하는 현장 관리자가 이를 손쉽게 활용할 수 있다.