• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.183-183
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.150.2-150.2
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.77-78
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.167.1-167
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• 1999

• 비파괴검사학회지
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• 제29권2호
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• pp.98-103
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• 2009

원자력발전소 증기발생기 전열관에 균열이 발생할 경우 해당 전열관을 관막음하여 안전하게 운영하고 있다. 만약 가동중검사시 균열 검출에 실패할 경우 전열관 파단사고와 같은 대형 사고로 이어질 수 있다. 증기발생기 전열관에는 여러 유형의 균열이 발생하고 있는데, 와전류검사로부터 균열이 확인된 경우 균열의 크기를 평가하여 전열관의 건전성을 평가하여야 한다. 그러나 균열의 깊이평가는 평가방법이 난해하여 평가 결과의 정확도 및 평가자 사이의 일관성이 떨어진다. 본 논문에서는 현재 사용되고 있는 균열깊이 평가방법에 대한 정확도 및 일관성을 확인하고, 보다 신뢰성 있는 평가방법의 개발을 위하여 고리 1호기 구증기발생기를 활용하였다. 국내 유자격 평가자들의 round robin test 결과를 통계적으로 분석하여 균열 유형별 최적의 평가방법을 도출하였다. 본 논문에서 제시된 균열깊이 평가기법은 국내 원전의 증기발생기관리프로그램에 활용되어 원전의 신뢰성 향상에 기여할 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1307-1313
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• 2019

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

• 한국안전학회지
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• 제20권2호
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• 대한기계학회논문집A
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• 제28권2호
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• pp.158-164
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• 2004

Steam generator tubes experience widespread degradations such as stress corrosion cracking, wear, tube rupture, denting, fatigue and so on. The resulting damages can cause tube bursting or leak of the primary water which contains radioactivity Therefore the allowable size of the damage is required to be determined on the maintenance purpose. The burst pressure of a tube with a T-type combination crack consisting of longitudinal and circumferential cracks is obtained experimentally and analytically. Fracture parameters such as stress intensity factor and crack opening angle are investigated. Also the burst pressure for a T-type combination crack is compared with that of a single longitudinal crack to develop a length-based criteria.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1570-1578
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• 2022

During a severe accident, steam generator (SG) tubes undergo rapid changes in the pressure and temperature. Therefore, an appropriate creep model to predict a short term creep damage is essential. In this paper, a novel creep model for Alloy 690 SG tube material was proposed. It is based on the theta (θ) projection method that can represent all three stages of the creep process. The original θ projection method poses a limitation owing to its inability to represent experimental creep curves for SG tube materials for a large strain rate in the tertiary creep region. Therefore, a new modified θ projection method is proposed; subsequently, a master curve for Alloy 690 SG material is also proposed to optimize the creep model parameters, θ_i (i = 1-5). To adapt the implicit creep scheme to the finite element code, a partial derivative of incremental creep with respect to the stress is necessary. Accordingly, creep model parameters with a strictly linear relationship with the stress and temperature were proposed. The effectiveness of the model was validated using a commercial finite element analysis software. The creep model can be applied to evaluate the creep rupture behavior of SG tubes in nuclear power plants.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1314-1323
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• 2018

For nuclear power plants (NPPs) to have long lifetimes, ageing is a major issue. Currently, ageing management for NPP systems is based on correlations built from generic experimental data. However, each system has its own characteristics, operational history, and environment. To account for this, it is possible to resort to prognostics that predicts the future state and time to failure (TTF) of the target system by updating the generic correlation with specific information of the target system. In this paper, we present an application of particle filtering for the prediction of degradation in steam generator tubes. With a case study, we also show how the prediction results vary depending on the uncertainty of the measurement data.