• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2953-2959
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• 2021

A new practical modeling of the Norton's power law creep is proposed and implemented to analyze the high temperature behaviors of Alloy 690 SG tube material. In the model, both the stress exponent n and the rate constant B are simply treated as the temperature dependent parameters. Based on the two-step optimization procedure, the temperature function of the rate constant B(T) was determined for the data set of each B value after fixing the stress exponent n value by using the prior optimized function at each temperature. This procedure could significantly reduce the numerical errors when using the power law creep equations. Based on the better description of the steady-state creep rates, the experimental rupture times could also be well predicted by using the Monkman-Grant relationship. Furthermore, the difference in tensile strengths at high temperatures could be very well estimated by assuming the imaginary creep stress related to the given strain rate after correcting the temperature effects on the elastic modulus.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.1
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• pp.9-15
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• 2008

A new ECT probe to separate the ECT signal distortion due to PVC (permeability variation clusters) and ordinary defects created in SG tubes has been developed. The hystersis loops of PVC which are extracted from retired SG (steam generator) tubes of Kori-1 NNP were measured. The tensile tests were performed to identify the mechanism of PVC creation. The conditions detecting the PVC created in 56 tubes were investigated using computer simulation, and the signal processing circuits were inserted in the probe for the digital signal transmission. The new Probe can measure and separate the PVC signal which is created in the SG tubes, and also measures the defects in Ni-sleeving part of SG tubes. furthermore the new ECT probe can measure the defects as fast as bobbin probe, and enhance the testing speed as well as reliability of the defect detection of SG tubes.

• 제어로봇시스템학회:학술대회논문집
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• 1992.10b
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• pp.74-78
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• 1992

This paper describes an intelligent system to automatic evaluation of eddy current(EC) signal for Inspection of steam generator(SG) tubes in nuclear power plant. Some features of the intelligent system design in the proposed system are : (1) separation of representation scheme ,or event capturing knowledge in EC signal and for structural inspection knowledge in SG tubes inspection; (2) each representation scheme is implemented in different methods, one is syntactic pattern grammar and the other is rule based production. This intelligent system also includes an data base system and an user interface system to support integration of the hybrid knowledge processing methods. The intelligent system based on the proposed concept is useful in simplifying the knowledge elicitation process of the rule based production system, and in increasing the performance in real time signal inspection application.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.893-899
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• 2024

In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju". However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H₂), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. At the same time, we observed experimentally that the fine H₂ bubbles exist stably in the liquid sodium, longer than previously expected. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.

• Proceedings of the KIEE Conference
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• 2003.07d
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• pp.2603-2605
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• 2003

증기발생기(SG : Steam Generator)는 다수의 세관으로 구성되어 원자로에서 발생한 열을 이용하여 발전기 터빈을 구동시키는 원동력인 증기를 생성해 주는 기능을 하는 원자력발전소의 핵심 설비이다. 증기 발생기 세관의 건전성을 확보하기 위해 매주기 계획예방 정비, 즉 가동중검사마다 정기적인 와전류검사를 수행하고, 검사결과에 따라 전열관 보수 등과 같은 제반 조치를 취하고 있다. 현재 검사데이터 DB 구축은 일부 발전소에 개발되어 운영중에 있고, 세관 DB와는 별도로 통계정보만을 관리하는 증기발생기 성능관리시스템이 운영되고 있으며, 또한 각 발전소마다 수질을 계측하여 수화학 성분을 감시하는 수질관리시스템이 운용되고 있다. 이러한 이원화된 DB 및 시스템을 통합하고 연계하여 전 원전의 증기발생기를 종합적으로 관리 할 수 있는 시스템의 필요성이 대두되었다. 따라서 본 논문에서는 현장에 보관되어 있는 모든 세관 검사데이터를 취득하여 대용량 데이터베이스를 설계 및 구축하고 이기종의 분산된 수질관리시스템 DB를 연계하여, 증기발생기의 설계/제작부터 검사결과 Mapping, 추이 분석을 통한 수명평가에 이르는 전 과정을 통합 관리한 수 있는 시스템을 개발하고 그 구현방안을 제시한다.

• Proceedings of the KIEE Conference
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• 2005.07b
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• pp.1008-1010
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• 2005

In the examination of steam generator(SG) tube in nuclear power plant, eddy current testing probes play an important role in detecting the defects. Bobbin probe and MRPC probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary MRPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it has excellent detection capability for small cracks, which is hardly detected by bobbin probe. In this paper, for the accurate analysis of experimental ECT signals, construction of MRPC probe signals database according to the variation of defect size is the main purpose. Using 3-D finite element method, ECT signals are analyzed, and signals analysis add according to frequency ingredient. The results, which are analysis and characteristics ion of electromagnetism simulation signals, is databased.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.58-66
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• 2018

This paper represents a systematic approach aimed at improving the performance of the proportional integral (PI) controller for the Advanced Power Reactor (APR) 1400 Feedwater Control System (FWCS). When the performance of the PI controller offers superior control and enhanced robustness, the steam generator (SG) level is properly controlled. This leads to the safe operation and increased the availability of the nuclear power plant. In this paper, a systems engineering approach is used in order to design a novel PI controller for the FWCS. In the reverse engineering stage, the existing FWCS configuration, especially the characteristics of the feedwater controller as well as the feedwater flow path to each SG from the FWCS, were reviewed and analysed. The overall block diagram of the FWCS and the SG was also developed in the reverse engineering process. In the re-engineering stage, the actual design of the feedwater PI controller was carried out using a genetic algorithm (GA). Lastly, in the validation and verification phase, the existing PI controller and the PI controller designed using GA method were simulated in Simulink/Matlab. From the simulation results, the GA-PI controller was found to exhibit greater stability than the current controller of the FWCS.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3225-3234
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• 2022

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• Proceedings of the Korean Information Science Society Conference
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• 2002.10c
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• pp.271-273
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• 2002

증기발생기(SG : Steam Generator)는 다수의 세관으로 구성되어 원자로에서 발생한 열을 이용하여 발전기 터빈을 구동시키는 원동력인 증기를 생성해 주는 기능을 하는 원자력발전소의 핵심 설비이다. 증기발생기 세관의 건전성을 확보하기 위해 매주기 계획예방정비, 즉 가동중 검사마다 정기적인 와전류 검사를 수행하고, 검사결과에 따라 전열관 보수 등과 같은 제반 조치를 취하고 있다. 현재 검사데이터 DB 구축은 일부 발전소에 개발되어 운영 중에 있고, 세관 DB와는 별도로 통계정보만을 관리하는 증기발생기 성능관리시스템이 운영되고 있으며, 또한 각 발전소마다 수질을 계측하여 수화학 성분을 감시하는 수질관리시스템이 운용되고 있다. 이러한 이원화된 DB 및 시스템을 통합하고 연계하여 전 원전의 증기발생기를 종합적으로 관리 할 수 있는 시스템의 필요성이 대두되었다. 따라서 본 논문에서는 현장에 보관되어 있는 모든 세관 검사데이터를 취득하여 대용량 데이터베이스를 설계 및 구축하고 이기종의 분산된 수질관리시스템 DB를 연계하여, 증기발생기의 설계/제작부터 검사결과 Mapping, 추이 분석을 통한 수명 평가에 이르는 전 과정을 통합 관리할 수 있는 시스템을 설계하고 그 구현방안을 제시한다.