• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.74-83
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• 2020

A number of so-called 'Type IV' cracking was reported to occur at the welded joints of the P91 steel or P92 steel reheat steam piping systems in Korean supercritical thermal power plants. The reheat steam piping systems are subjected to severe thermal and pressure loading conditions of coolant higher than 570℃ and 4MPa, respectively. In this study, piping analyses and design evaluations were conducted for the piping system of a specific thermal plant in Korea and suggestions were made how structural integrity could be improved so that type IV cracks at the welded joints could be prevented. Integrity evaluations were conducted as per ASME B31.1 code with implicit consideration of creep effects which was used in original design of the piping system and as per nuclear-grade RCC-MRx code with explicit consideration of creep effects. Comparisons were made between the evaluation results from the two design rules. Another approach with modification or reduction of the redundant supports in the piping systems was investigated as a tool to mitigate thermal stresses which should essentially contribute to prevention of Type IV cracking without major modification of the existing piping systems. In addition, a post weld heat treatment method and repair weld method which could improve integrity of the welded joint of P91 steel were investigated.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.636-643
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• 2001

Interference between the crossover leg of the reactor coolant system (RCS) and the pipe whip restraints (PWR) has brought a degradation issue of the integrity of the Reactor Coolant System in Westinghouse type nuclear power plants (NPPs) of Korea. According to the gap Inspect ion carried out during planned overhaul (Year 2000), interference between the crossover leg and the PWR was found in each RCS loop. This plant has had the high vibration problem on the RC pump 'B'. The reason for the high vibration in the RC pump 'B' had been massively surveyed and it was found that the crossover leg of RCS contacted with the PWR in hot condition. Since the contact between the crossover leg and the PWR changes the dynamic characteristics of the piping system for the RCS, this is considered as one reason for the high vibration. And a possibility of overstress on the crossover leg due to the contact with the PWR should be evaluated. Through performing RCS integrity analyses, subsequent actions were initiated to increase the gap between those parts. As the results of the appropriate separation between two parts, it was reported that there was no unusual noise or vibration during plant heat-up. In this paper, the evaluations for the gap between the crossover leg and the PWR and the structural integrity due to loop binding is described.

• Corrosion Science and Technology
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• 제12권3호
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• pp.125-131
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• 2013

The most common pipe wall thinning degradation mechanisms that can occur in the steam and feedwater systems are FAC (Flow Acceleration Corrosion), cavitation, flashing, and LDIE (Liquid Droplet Impingement Erosion). Among those degradation mechanisms, FAC has been investigated by many laboratories and industries. Cavitation and flashing are also protected on the piping design phase. LDIE has mainly investigated in aviation industry and turbine blade manufactures. On the other hand, LDIE has been little studied in NPP (Nuclear Power Plant) industry. This paper presents the development of prediction system for pipe wall thinning caused by LDIE in terms of erosion rate based on air-water ratio and material. Experiment is conducted in 3 cases of air-water ratio 0.79, 1.00, and 1.72 using the three types of the materials of A106B, SS400, and A6061. The main control parameter is the air-water ratio which is defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). The experiments were performed for 15 days, and the surface morphology and hardness of the materials were examined for every 5 days. Since the spraying velocity (v) of liquid droplets and their contact area ($A_c$) on specimens are changed according to the air-water ratio, we analyzed the behavior of LDIE for the materials. Finally, the prediction equations(i.e. erosion rate) for LDIE of the materials were determined in the range of the air-water ratio from 0 to 2%.

• 에너지공학
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• 제27권4호
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• pp.70-79
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• 2018

한국중부발전 제주화력본부에서 방류되는 온배수로부터 열원을 회수하여 약 3 km의 원거리에 위치하고 있는 신촌백합영농조합법인의 시설온실에 냉온수를 공급하기 위한 300 RT 용량의 열펌프 시스템을 구축하였다. 원거리 시설로 온배수열을 공급하기 위한 최적화 방안을 도출하였고, 수송관로의 온도와 누수여부를 모니터링하기 위한 설비들이 설치되었다. 본 논문에서는 제주 시설온실 냉난방을 위한 발전소 온배수 활용 열펌프 시스템의 구성과 주요기기에 대해 기술하였다. 2018년 하절기에 열펌프 시스템의 냉방성능을 평가하기 위한 실증운전을 수행하였다. 냉방운전 시 열펌프 시스템의 주요 위치에서 측정된 유체의 온도와 유량, 그리고 냉방 성능계수를 분석하여, 구축된 열펌프 시스템의 안정성과 냉방성능을 검증하였다.

• International Journal of Air-Conditioning and Refrigeration
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• 제11권2호
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• pp.51-60
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• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. Standard k-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2003년도 ICCAS
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• pp.789-792
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• 2003

Pipe inspection has a great importance to ensure safety for the nuclear power plant. In this paper, we designed visual inspection module for the tube internal, which diameter is 15${\sim}$20mm. And we made inspection module which consisted of CCD camera and light. And the relation between image and real world coordinate is established. Image processing is performed to calculate mapping parameter and analyze the size of defect. For the calculation of mapping parameter, experiment is performed using grid type test pattern. Acquired image is processed to extract image coordinate. Edge detection, thresholding, median filtering and morphology filtering is applied to extract grid pattern. Extracted image coordinate is used to calculate image to real world mapping. Lens distortion was considered and corrected to get exact data. Coordinate transformation data is provided for the users to recognize easily. Experiment was performed using grid type test pattern, we extracted lens distortion parameter and real coordinate of defect point. Radial distortion of lens was corrected but tangential distortion was not considered. As continuum to this study, the tangential distortion of lens is considered and improvement of analy zing technique for the tube internal be explored continuously.

• 설비공학논문집
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• 제15권3호
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• pp.239-245
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• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. $textsc{k}$-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of the main flow in the duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from the main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• 비파괴검사학회지
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• 제28권2호
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• pp.191-198
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• 2008

유체가속부식에 의한 탄소강 배관의 감육은 원자력 발전소 저탄소강 배관의 주요 경년열화 현상으로서, 예상치 못한 배관의 파단을 야기해 발전소의 성능 및 안전을 저해할 수 있다. 최근, 등전위 교번식 직류 전위차법(ES-DCPD, equipotential switching direct current potential drop)을 이용한 배관 감육의 정밀 감시기법이 본 연구자들에 의하여 개발되었다. ES-DCPD 방법은 넓은 배관 영역을 빠르게 검사할 수 있는 방법으로, 넓은 영역의 직관부 감육을 빠르게 검사하는 광역감시법(WiRN, wide range monitoring)과 엘보우 등 곡관부의 감육이 활발한 컴포넌트의 국부적 감육을 비교적 넓은 범위에서 빠르게 스캔하는 협역감시법(NaRM, narrow range monitoring)으로 사용이 가능하다. 광역감시와 협역감시 기법은 초음파검사의 위치 선정파 초음파검사의 검사 누락부에 대한 신뢰성을 개선할 수 있을 것이다. 본 논문에서는 ES-DCPD를 바탕으로 한 새로운 감육 진단 기술을 실험실 환경에서 장기 검증 시험을 수행하여 신호 정밀도를 분석하였고, 결과의 현장 적용성을 논의하였다.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2002년도 춘계학술대회 논문집
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• pp.134-139
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• 2002

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. However, effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. Acoustic emission(AE) has been widely used in various fields because of its extreme sensitivity, dynamic detection ability and location of growing defects. In this study, we investigated failure modes of locally wall thinned pipes and AE signals by bending test. From test results, we could be divided four types of failure modes of ovalization, crack initiation after ovalization, local buckling and crack initiation after local buckling. And fracture behaviors such as elastic region, yielding range, plastic deformation range and crack progress could be evaluated by AE counts, accumulative counts and time-frequency analysis during bending test. It is expected to be basic data that can protect a risk according to local wall thinning of pipes, as a real time test of AE.

• 비파괴검사학회지
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• 제25권2호
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• pp.95-102
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• 2005

원자력 발전소의 많은 배관부는 고온, 고압환경에서 적용되고 있어 환경적 및 기계적 요인에 의하여 부식결함이 빈번히 발생하고 있다. 이와 같은 부식 결함은 초음파 기법 등에 의하여 평가되어야 하고, 본 연구에서는 주사형 레이저 유도 초음파(SLS) 기법을 도입하여 배관부 부식결함의 영상화 기법을 적용하였다. 본 기법은 표면이 거칠거나 배관재와 같은 곡면의 표면에서 적용할 수 있는 장점이 있다. 한편 기존의 주사형 초음파 기법은 초음파 센서와 검사대상체 사이의 초음파 전달 매개체를 확보하기 위하여 시험편이 수침되거나 워터젯을 이용하였으나 주사형 레이저 초음파 기법은 광학적 기법을 이용하여 초음파를 발생시키므로 비접촉 방법에 의한 주사 이미지 획득이 가능하다. 따라서 본 주사형 레이저 초음파 기법은 복잡한 구조물의 검사, 비접촉 원격 및 고화질의 결함 이미지 탐상이 가능하다. 본 연구에서는 배관 결함의 검출능 향상을 위하여 결함 영상 획득에 있어서 다양한 조건의 레이저 유도 초음파 발생 기법을 적용하였고, 배관에 존재하는 응력부식 균열의 결함 영상을 얻게 되었다.