• 한국안전학회지
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• 제21권4호
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• pp.42-48
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• 2006

The strategies of maintenance and operation are usually established based on the number of flaws and their size distribution obtained from nondestructive inspection in order to preserve safety of the plant. But non destructive inspection results are different from the physical flaws which really exist in the equipments. In case of a single inspection, it is easy to estimate the number of physical flaws using the POD curve. However, we may be faced with some difficulties in obtaining the number of physical flaws from the periodic in-service non destructive inspection data. In this study a simple method for estimating the number of physical flaws from periodic in-service nondestructive inspection data was proposed. In order to obtain the flaw growth history, the flaw growth was simulated using the Monte Carlo method and the flaw size and the corresponding POD value were obtained for each flaw at each periodic inspection time. The flaw growth rate used in the simulation was statistically calculated from the in-service inspection data. By repeating the simulation numerous flaw growth data could be generated and the effective POD curve was obtained as a function of flaw size. From the effective POD curve the number of physical flaws was obtained. The usefulness and convenience of the proposed method was evaluated from several applications and satisfactory results were obtained.

• Corrosion Science and Technology
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• 제3권3호
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• pp.112-117
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• 2004

This work investigated the material properties of Ni-P-B alloy electrodeposits obtained from a Ni sulfamate bath as a function of the contents of the P and B sources($H_3PO_3$ and dimethyl amine borane complex(DMAB), respectively) with/without additives. Chemical composition, residual stress, microstructure and micro hardness were investigated using ICP(inductively coupled plasma) mass spectrometer, flexible strip, XRD, TEM and micro Vickers hardness tester, respectively. From the results of the compositional analysis, it was observed that P and B are incorporated competitively during the electrodeposition and the sulfur from the additive is codeposited into the electrodeposit. The measured residual stress value increased in the order of Ni, Ni-P, Ni-B and Ni-P-B electrodeposits indicating that boron affects the residual tensile stress greater than phosphorus. As the contents of the alloying element sources of P and B increased, crystallinity and the grain size of the electrodeposit decreased. The effect of boron on crystallinity and grain size was also relatively larger than the phosphorus. It can be explained that the boron with a smaller atomic radius contributes to the increase of residual stress in the tensile direction and the larger restraining force against the grain growth more significantly than the phosphorus with a larger atomic radius. Introduction of an additive into the bath retarded crystallization and grain growth, which may be attributed to the change of the grain growth kinetics induced by the additive adsorbed on the substrate and electrodeposit surfaces during electrodeposition.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2010년도 춘계학술발표대회 초록집
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• pp.43-46
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• 2010

As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

• Corrosion Science and Technology
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• 제11권3호
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• pp.96-102
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• 2012

0.1 M NaOH 용액에 PbO첨가양이 증가함에 따라 Alloy 600에 형성되는 산화막의 부동태 피막 특성이 열화되었다. 또한 뚜렷한 2중층 구조의 산화막이 점차 사라지고, 산화막내 존재하는 납의 양이 증가하였다. 산화막 내부 납의 양이 증가함에 따라 산화막 내부 니켈의 결핍이 점차 커졌다. 납에 의해 산화막의 부동태 특성이 약화됨에 따라, 응력부식균열 저항성 또한 급감하였을 것으로 판단된다.

• 플랜트 저널
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• 제9권4호
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• pp.26-30
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• 2013

유동가속부식에 의해 배관이 파손된 500 MW급 A 복합발전소 배열회수보일러 저압증기발생기 배관을 모델로 삼아 배관급수 내의 용존산소 부족이 유동가속부식의 주요 원인임을 도출하고 용존산소를 증가시키기 위해 적용된 하이드라이진 차단 수처리에 대한 적용효과를 분석하였다. 수처리 적용 1년 후 급수의 용존산소는 0.15 ppb에서 3~5 ppb로 상승되고, 산화환원전위도 -245 mV에서 170 mV로 산화성으로 상승되었다. 또한 유동가속부식에 의한 부식생성물인 철분함유량은 18.5 ppb에서 5~7 ppb로 감소되었다. 따라서 하이드라이진 차단 수처리로 급수의 용존산소가 증가되며 유동가속부식에 의한 배관의 부식생성물인 철분함유량이 감소됨을 확인하였다.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1446-1453
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• 2021

In the past, several experimental investigations aiming at characterizing the natural circulation (NC) behavior in test facilities were carried out. They showed a variety of flow patterns characterized by an inverted U-shape of the NC flow curve versus primary mass inventory. On the other hand, attempts to reproduce such curves using thermal-hydraulic system codes, showed 10-30% differences between the measured and calculated NC mass flow rate. Actually, the used computer codes are generally based upon nodalization using single U-tube representation. Such model may not allow getting accurate simulation of most of the NC phenomena occurring during such tests (like flow redistribution and flow reversal in some SG U-tubes). Simulations based on multi-U-tubes model, showed better agreement with the overall behavior, but remain unable to predict NC phenomena taking place in the steam generator (SG) during the experiment. In the current study, the CATHARE code is considered in order to assess a NC characterization test performed in the four loops PKL facility. For this purpose, four different SG nodalizations including, single and multi-U-tubes, 1D and 3D SG inlet/outlet zones are considered. In general, it is shown that the 1D and 3D models exhibit similar prediction results up to a certain point of the rising part of the inverted U-shape of the NC flow curve. After that, the results bifurcate with, on the one hand, a tendency of the 1D models to over-predict the measured NC mass flow rate and on the other hand, a tendency of the 3D models to under-predict the NC flow rate.

• 한국안전학회지
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• 제36권3호
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• pp.74-80
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• 2021

Korea Atomic Energy Research Institute (KAERI) has been operating an integral effects test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), according to APR1400 for transient experimental and design basis accident simulation. Moreover, based on the experimental data, the domestic standard problem (DSP) program has been conducted in Korea to validate system codes. Recently, through DSP-05, the performance of the passive auxiliary feedwater system (PAFS) in the event of multiple steam generator tube rupture (MSGTR) has been analyzed. However, some errors exist in the reference input model distributed for DSP-05. Furthermore, the calculation results of the heat loss correlation for the secondary system presented in the technical report of the reference indicate that a large difference is present in heat loss from the target value. Thus, in this study, the reference model is corrected using the geometric information from the design report and drawings of ATLAS. Additionally, a new heat loss correlation is suggested by fitting the results of the heat loss tests. Herein, MSGTR-PAFS accident analysis is performed using MARS-KS 1.5 with the improved model. The steady-state calculation results do not significantly differ from the experimental values, and the overall physical behavior of the transient state is properly predicted. Particularly, the predicted operating time of PAFS is similar to the experimental results obtained by the modified model. Furthermore, the operating time of PAFS varies according to the heat loss of the secondary system, and the sensitivity analysis results for the heat loss of the secondary system are presented.

• 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.