• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1436-1445
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• 2021

To obtain the dynamic characteristics of reactor secondary circuit under transient conditions, the system analysis program was developed in this study, where dynamic models of secondary circuit were established. The heat transfer process and the mechanical energy transfer process are modularized. Models of main equipment were built, including main turbine, condenser, steam pipe and feedwater system. The established models were verified by design value. The simulation of the secondary circuit system was conducted based on the verified models. The system response and characteristics were investigated based on the parameter transients under emergency shutdown and overload. Various operating conditions like turbine emergency shutdown and overspeed, condenser high water level, ejector failures were studied. The secondary circuit system ensures sufficient design margin to withstand the pressure and flow fluctuations. The adjustment of exhaust valve group could maintain the system pressure within a safe range, at the expense of steam quality. The condenser could rapidly take out most heat to avoid overpressure.

• 한국유체기계학회 논문집
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• 제16권4호
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• pp.35-43
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• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• 한국자기학회지
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• 제15권5호
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• pp.292-297
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• 2005

원자력 발전소 증기발생기의 전열관은 전열면으로서의 역할과 방사능 차단벽의 역할로서 중요하며, 증기발생기 튜브의 폭발은 원자력 발전소 사고에 관계된다. 증기발생기 전열관 재료로는 Incone1600이 사용되고 있으며, 이 재료가 금속이면서 비자성이기 때문에 와전류 탐상법으로 전열관의 결함을 탐지하고 있다. 본 연구에서는 탐상 감도를 향상시키기 위하여 차등형 와전류 탐촉자를 개발하였으며, 개발된 차등형 와전류 탐촉자의 성능검사를 위해 Inconel600과 자기적 성질이 비슷하고 구하기 쉬운 SUS304로 가로 100mm, 세로 100mm, 그리고 두께 10mm인 평판에 선형결함과 원형결함을 가공하여 기준시편으로 제작하였다. 제작된 차등형 와전류 탐촉자를 사용하여, 자화주파수 50kHz, lift-off 0.4mm에서 직경이 0.25mm이고, 깊이가 0.2mm 크기의 결함까지 측정이 가능하였다.

• 한국전산유체공학회지
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• 제7권3호
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• pp.35-43
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• 2002

This paper addresses a numerical method for predicting transient temperature distributions in the wall of a curved pipe subjected to internal laminar thermally-stratified flow. A simple and convenient numerical method of treating the unsteady conjugate heat transfer in non-orthogonal coordinate systems is presented. Numerical calculations are performed for the transient evolution of thermal stratification in two curved pipes, where one has thick wall and the other has so thin wall that its presence can be negligible in the heat transfer analysis. The predicted results show that the thermally stratified flow and transient conjugate heat transfer in a curved pipe with a finite wall thickness can be satisfactorily analyzed by the present numerical method, and that the neglect of wall thickness in the prediction of pipe wall temperature distributions can provide unacceptably distorted results for the cases of pipes with thick wall such as safety related-piping systems of nuclear power plant.

• 대한기계학회논문집A
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• 제27권1호
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• pp.8-17
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• 2003

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear Power Plant. In Pipes of energy Plants, sometimes, the local wall thinning may result from severe erosion-corrosion (E/C) damage. However, the effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. From the tests, fracture behaviors and fracture strength of locally thinned pipe were manifested systematically. The observed failure modes were divided into four types; ovalization. crack initiation/growth after ovalization, local buckling and crack initiation/growth after local buckling. Also, the strength and the allowable limit of piping system with local wall thinning were evaluated.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.74-78
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• 2015

Nuclear power plant steam generator that is one of the main component has several thousands of thin tubes. And the steam generator tube is subject to damage because of the severe operation conditions such as the high temperature and pressure. Therefore periodic inspections are conducted to ensure the integrity of steam generator component. Hanul unit 3 also has been inspected in accordance with in-service inspection program and is scheduled to be replaced for exceeding the plugging rate which was recommended by manufacturer. During the steam generator replacement activity, we found several clustered porosity on inner surface of main feed water pipe. Additionally crack-like indications were found at weld interface between base material and weld of main feed water pipe. This paper describes the field experience and visual testing results for inner surface of main feed water pipes. The destructive test result had shown that these indications were porosities which were caused by manufacturing process not by operation service.

• 비파괴검사학회지
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• 제32권3호
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• pp.284-290
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• 2012

Infrared thermography uses infrared energy radiated from any objects above absolute zero temperature, and the range of its application has been constantly broadened. As one of the active test techniques detecting radiant energy generated when energy is applied to an object, ultrasound infrared thermography is a method of detecting defects through hot spots occurring at a defect area when 15~100 kHz of ultrasound is excited to an object. This technique is effective in detecting a wide range affected by ultrasound and vibration in real time. Especially, it is really effective when a defect area is minute. Therefore, this study conducted thermography through lock-in signal processing when an actual defect exists inside the austenite STS304 seamless pipe, which simulates thermal fatigue cracks in a nuclear power plant pipe. With ultrasound excited, this study could detect defects on the rear of a pipe by using an aluminium reflector. Besides, by regulating the angle of the aluminium reflector, this study could detect both front and rear defects as a single infrared thermography image.

• 한국산학기술학회논문지
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• 제14권1호
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• pp.33-38
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• 2013

여러 종류의 증기발생기 검사정비 로봇의 엔드이펙터 모션 구동에 전부 사용할 수 있도록 ARM Cotex M3-107 MCU 기반의 제어기와 엔드이펙터 모션 프로그램 생성 응용소프트웨어로 구성된 범용 엔드이펙터 모션구동 제어시스템을 개발하였다. 범용 제어시스템을 적용하여 엔드이펙터의 직선이송 및 회전이송의 위치 결정의 오차는 무시할만한 수준이며, 재현성은 0.04% 오차를 보여줌으로써 실제로 사용 가능한 범용 엔드이펙터 모션구동 제어시스템을 개발하였다.

• 에너지공학
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• 제11권1호
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• pp.18-25
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• 2002

침부식 (FAC, Flow-Accelerated Corrosion)에 의한 감육 문제는 원자력 발전소 배관관리에 있어서 매우 중요하다. 특히 FAC는 배관 내부 유체의 pH, 용존산소 농도, 유체 온도, 유속 및 습증기 분율 등과 배관의 형상 및 재료 등의 특정 조건에서만 발생하므로, FAC 문제를 관리하기 위해서는 체계적인 접근이 필요하다. 본 연구에서는 국내 특정 CANDU원전의 2차계통 배관을 대상으로 관련 데이터베이스 구축, 구축된 데이터베이스를 이용한 FAC감육율의 예측 및 배관 잔여수명의 평가 등을 수행하였다. 또한 FAC 발생기구 및 FAC에 영향을 주는 요인에 대해서도 조사하였다. 습분분리기와 플래시탱크 사이 배관 라인의 해석 예로부터 FAC 문제를 관리하는 방안을 제시하였다. 제시된 방안은 국내 다른 원자력발전소의 배관 관리에도 활용될 수 있을 것이다.

• Smart Structures and Systems
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• 제9권5호
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• pp.461-472
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• 2012

This study proposes an optics-based active sensing system for continuous monitoring of underground pipelines in nuclear power plants (NPPs). The proposed system generates and measures guided waves using a single laser source and optical cables. First, a tunable laser is used as a common power source for guided wave generation and sensing. This source laser beam is transmitted through an optical fiber, and the fiber is split into two. One of them is used to actuate macro fiber composite (MFC) transducers for guided wave generation, and the other optical fiber is used with fiber Bragg grating (FBG) sensors to measure guided wave responses. The MFC transducers placed along a circumferential direction of a pipe at one end generate longitudinal and flexural modes, and the corresponding responses are measured using FBG sensors instrumented in the same configuration at the other end. The generated guided waves interact with a defect, and this interaction causes changes in response signals. Then, a damage-sensitive feature is extracted from the response signals using the axi-symmetry nature of the measured pitch-catch signals. The feasibility of the proposed system has been examined through a laboratory experiment.