• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.1-10
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• 2016

The number of nuclear power plants operating in Korea was 24 as of year 2015. Nine units out of 24 units have been operated for a period over 20 years. Kori unit 1 has been in operation for 40 years, and an extended operation for Wolsong unit 1 was decided in 2015. There has been reported some crackings in reactor internals in PWR have been reported in Europe, USA, Japan and Korea, and some of them were replaced with new one. Repair and replacement technologies for the reactor internals have been developing in order to meet the regulatory requirements for long term operation in Korea. The technologies will also be used for the exported nuclear units. It is required to review degradation history of the reactor internals worldwide as a part of the degradation management program development. Schematics of reactor internals designed and supplied by Westinghouse, Framatome and Combustion Engineering are described herein. Materials degradation history of reactor internals of PWR plants in USA, Japan and Europe is surveyed and summarized. Some events from Korean plants are also described. Aging management strategy for the internals is suggested.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.49-55
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• 2016

This paper analyzed a root cause of abnormality in the temperature and vibration at generator-side bearing of motor generator set (MG Set), which is a power supply system to control element drive mechanism (CEDM) of nuclear power plants (NPPs), and modified the design of roller-type and sealing method to improve the abnormalities. From the inspection of MG Set and analysis of temperature variation during service, it was found that the abnormal temperature transition was basically associated with original design of generator-side bearing, whose roller was axially restrained by inner race, and that the abnormal vibration level was caused by inserting small chips of cage and V-ring, which were generated due to the abnormal temperature transition at roller bearing. Type of bearing and sealing method were modified based on these analyses. The temperature and vibration level measured at roller bearing showed that the modifications clearly improved the operational stability of MG Set.

• 한국압력기기공학회 논문집
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• 제5권2호
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• pp.13-19
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• 2009

As the operating time in nuclear power plants (NPPs) increases, the integrity of nuclear components may be continually degraded due to aging effects of systems, structures and components. Recently, a number of NPPs are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. Therefore, it is beneficial to build a monitoring system to measure an aging status. In this paper, the Aging Monitor (AM) based on lots of aging database obtained from the operating plants and research results on the aging effects was developed to monitor, manage and evaluate the aging phenomena systematically and effectively in NPPs. The AM for the CANDU is divided into 6 modules: (1) Aging Alarm/Coloring Monitor, (2) Aging Database, (3) Aging Document, (4) Real-time Integrity Monitor, (5) Surveillance and Inspection Management System, and (6) Continued Operation and Periodic Safety Review (PSR) Safety Evaluation. The proposed system is expected to provide the integrity assessment for the major mechanical components of an NPP under concurrent working environments.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.117-124
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• 2020

3D printing is a technology that has significantly grown in recent years, particularly in the aerospace, defense, and medical sectors where it offers significant potential cost savings and reduction of the supply chain by allowing parts to be manufactured on-site rather than at a distance supplier. In nuclear industry, 3D printing technology should be applied according to the manufacturing trend change. For the application of 3D printing technology to the nuclear power plant, several problems, including the absence of code & standards of materials, processes and testing & inspection methods etc, should be solved. Preemptively, the improvement of reliability of 3D printing technology, including mechanical properties, structural performance, service performance and aging degradation of 3D printed parts should be supported. These results can be achieved by collaboration of many organizations such as institute, 3D printer manufacturer, metal powder supplier, nuclear part manufacturer, standard developing organization, and nuclear utility.

• 한국압력기기공학회 논문집
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• 제11권2호
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.17-26
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• 2011

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.28-35
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• 2021

The flow accelerated corrosion (FAC) is one of significant aging and degradation mechanism and can affect structural integrity of CANDU feeder pipes. Pipe burst can occur under normal operation pressure (min. 10 MPa) if wall-thinning of the feeder pipe due to FAC is accumulated. Previous studies considered simple shapes of feeder pipe with local wall-thinning in order to conservatively assess structural integrity of wall-thinned feeder pipe. In this paper, a new FE model is developed, having an actual shape of the feeder pipe (double bent) as well as the actual wall-thinning shape and location based on the in-service inspection result. Then, the burst pressure assessment of the wall-thinned feeder pipe is performed using lower bound limit load analysis considering elastic-perfectly plastic material. In addition, an improved formulation to predict the burst pressure of the wall-thinned feeder pipe is presented and the safety margin is compared with an existing assessment method.

• 대한기계학회논문집B
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• 제34권4호
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• pp.437-442
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• 2010

고온, 고압의 유체가 흐르는 탄소강 배관에서는 유동가속부식으로 인한 배관감육 현상이 발생할 수 있다. 화력 및 원자력발전소에서 유동가속부식으로 인한 배관 손상시 고비용의 보수와 발전 정지를 유발할 뿐 아니라 발전소 신뢰도 및 안전성에 영향을 미칠 수도 있다. CHECWORKS 프로그램은 국내 발전소에서 유동가속부식에 의한 배관 손상을 예방하기 위하여 배관 두께검사 데이터를 평가하고 검사 계획을 수립하는데 이용되어 왔다. 그러나 상기 프로그램은 원전 2차측 배관 모두를 데이터베이스화한 후에 배관라인 그룹별로 유동가속부식 손상을 예측하기 때문에 국부적으로 감육에 민감한 부위를 찾는데 어려움이 있다. 본 논문에서는 CHECWORKS 프로그램을 이용하여 해석을 수행하고 수치해석을 통하여 검증할 수 있는 방법론을 기술하였다. 또한 국내 원전 2개의 배관 라인그룹에 대하여 CHECWORKS 프로그램을 이용한 유동가속부식 민감 부위를 FLUENT를 이용한 수치해석 결과와 비교하였다.

• 대한기계학회논문집B
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• 제39권2호
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• pp.199-206
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• 2015

본 연구에서는 인공신경망을 이용하여 배관이나 배관요소의 작동상태를 예측할 수 있는 진단방법을 제안한다. 입자영상유속계 기술을 이용하여 얻어진 배관의 검사부위의 진동에 의한 이동량을 인공신경망의 학습용으로 사용한다. 측정시스템은 카메라, 조명, 인공신경망이 탑재된 호스트컴퓨터로 구성된다. 구축된 모니터링시스템이 제대로 작동하는지 이미 알고 있는 진동원(2개의 휴대폰)에 대하여 적용하였다. 진동가속도의 최소값, 최대값, 평균값을 인공신경망의 학습에 사용해 본 결과, 평균값이 진동상태의 실시간 모니터링에 적합함을 확인하였다. 구축된 진단시스템은 실제 가스배관의 작동상태에 대하여 모니터링 가능함이 확인되었다.

• 대한기계학회논문집A
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• 제37권5호
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• pp.649-655
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• 2013

이종금속용접부에 대한 실제 용접 공정 중 용접부에서 결함이 발견되면 이를 제거하고 보수용접이 수행된다. 일반적으로 보수용접을 수행하면 용접부에서 인장 잔류응력이 크게 증가될 수 있는 것으로 알려져 있다. 따라서 Alloy 82/182를 사용하여 보수용접이 수행된 이종금속용접부의 일차수 응력부식 균열 현상을 평가하기 위해서는 보수용접에 의한 용접부의 응력 변화를 정확하게 평가해야 한다. 본 논문에서는 비선형 유한요소해석을 수행하여 보수용접에 의한 원자력 이종금속 맞대기 용접부의 응력 분포를 평가하였다. 특히 보수용접 공정 모사를 위한 여러 유한요소 해석방법이 이종금속용접부의 응력분포에 미치는 영향을 평가하였다.