• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.610-616
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• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• 비파괴검사학회지
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• 제28권1호
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• pp.9-15
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• 2008

증기발생기 전열관에 생성되는 투자율변화에 의해 야기되는 신호왜곡 문제를 해결하기 위한 새로운 탐촉자를 개발하였다. 고리 1 호기 폐전열관에 생성된 자성상을 분리하여 자기이력곡선을 측정하였으며, 자성상이 생성되는 원인을 규명하기 위하여 고온 인장 시험을 수행하였다. 전산모사를 이용하여 탐촉자의 자정상 탐지조건을 결정하였으며, 디지털 신호전송을 위하여 신호처리용 전자회로를 소형화하여 탐촉자 속에 삽입하였다. 본 연구에서 개발된 신형 탐촉자를 이용하여 PVC 신호와 니켈 슬리빙 부위의 결함을 측정하였다. 신형 탐촉자는 보빈 탐촉자와 같이 고속으로 결함을 측정 할 수 있으며, 증기발생기 전열관의 탐상속도와 결함 탐지의 신뢰성을 증진시킬 수 있다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 제36회 하계학술대회 논문집 B
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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.

• 한국소음진동공학회논문집
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• 제13권6호
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• pp.452-459
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• 2003

The characteristics of fluid-elastic instability and effects of turbulent excitations for the KSNP steam generator tubes were investigated numerically. The information for the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the flow-induced vibration(FIV) analysis has been conducted by using the PIAT(program for Integrity assessment of SG tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the SG. To investigate the cause of the concentrated tube plugging zone, the FIV analysis has been performed for various column and row number of the steam generator tubes. From the results of FIV analysis the stability ratio due to the fluid-elastic instability and vibrational amplitude due to the turbulent excitation in the concentrated plugged zone have a trend of larger values than those of the outer concentrated tube Plugging zone.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1669-1679
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• 2017

Background: Steam generator (SG) is one of the significant components in the nuclear steam supply system. A variety of SGs have been designed and used in nuclear reactor systems. Every SG has advantages and disadvantages. A brief account of some of the existing SG designs is presented in this study. A high surface to volume ratio of a SG is required in small modular reactors to occupy the least space. In this paper, performance improvement for SGs of integral small modular reactor is proposed. Aims/Methods: For this purpose, cross-grooved microfins have been incorporated on the inner surface of the helical tube to enhance heat transfer. The primary objective of this work is to investigate thermal-hydraulic behavior of the proposed improvements through modeling in RELAP5-3D. Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure). The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

• Nuclear Engineering and Technology
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• 제53권9호
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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.

• 비파괴검사학회지
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• 제29권1호
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• pp.27-35
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• 2009

본 논문에서는 개선된 특징추출을 이용한 원자력 발전소 증기발생기 세관의 결함 형태 분류에 대한 연구를 수행한다. 본 논문에서는 4가지 축대칭 결함, 즉 I-In 형태, I-Out 형태, V-In 형태, V-Out 형태 결함을 고려한다. 유한요소법에 기초한 수치해석 프로그램을 이용하여 결함의 폭과 깊이를 변화시켜가면서 400개의 와전류탐상시험(ECT) 신호를 생성하였다. 이와 같이 생성된 ECT 신호로부터 새로운 특징을 제안하였는데, 여기에는 최대 임피던스 값을 갖는 점과 최대 임피던스 값의 1/2의 값을 갖는 점 사이의 위상각과 최대임피던스 값을 갖는 점과 최대 임피던스 값의 10%, 20%, 30%, 40%를 갖는 점사이의 위상각들이 포함된다. 또한, 결함형태를 분류하기 위하여 은닉층이 하나인 다층퍼셉트론을 사용하였다. 컴퓨터 모의실험 연구를 통하여 제안된 방법이 최대오차와 평균제곱오차 측면에서 향상된 결함 분류 성능을 얻는다는 것을 보였다.

• 비파괴검사학회지
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• 제29권2호
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• pp.137-144
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• 2009

본 논문에서는 전자기 유한요소 해석을 통하여 원전 증기발생기(SG, steam generator) 세관의 결함 크기 변화에 따른 배열 와전류 프로브의 와전류탐상 특성을 해석하였다. 프로브의 전자기적 특성을 해석하기 위하여 맥스웰 방정식을 이용하여 지배방정식을 유도하였고, 이를 3차원 전자기 유한요소법을 이용하여 문제를 해석하였다. 해석을 위해 선정한 결함은 평저공(FBH, flat bottomed hole) 결함을 선정하였다. FBH결함에 대해 결함의 위치를 관의 외부표면에 존재하게 하고 결함의 깊이는 세관 두께의 20%, 40%, 60%, 80%, 100%로 하였다. 또한 결함의 크기변화 및 시험주파수를 100 kHz, 300 kHz, 400 kHz로 변화시켜 해석하였다. 해석 대상으로는 원자력발전소 증기발생기 세관으로 사용되고 있는 Inconel 600 도체관을 사용하였다. 본 논문을 통하여 결함형상, 깊이 및 크기, 시험주파수의 변화에 따른 탐상신호의 변화를 확인할 수 있었다. 본 논문의 결과는 배열 와전류 프로브의 와전류탐상 신호 평가시 도움이 될 것이다.

• 비파괴검사학회지
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• 제33권6호
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• pp.519-523
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• 2013

A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.