• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2009년 추계학술발표대회
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• pp.100-100
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• 2009

본 논문에서는 원자력발전소 1차 계통의 스테인리스강 저합금강 이종금속용접부 및 스테인리스강 동종용접부의 잔류응력을 평가하고 스테인리스강 용접부의 응력부식균열 민감성에 대해 고찰하였다. 노즐 안전단의 이종금속용접부 및 안전단 배관의 동종용접부 제작 및 소재가공에 의행 생성되는 잔류응력을 예측하기 위해 열 탄소성 유한요소법 수치해석을 수행하였으며, 용접공정과 함께 표면의 잔류응력에 기여하는 절삭 및 연삭가공과 소재의 담금질 공정을 열 탄소성적으로 모사하였다. 전산해석 결과, 스테인리스주강의 담금질 잔류응력은 무시할 수 없는 상당한 크기이므로 배관 용접잔류응력 평가 시 소재의 담금질 효과를 고려해야 할 것으로 판단된다. 이종금속 용접과 동종금속 용접공정이 보수용접 없이 정상적인 절차(내면에서 외면으로 적층)로 완성된다면, 냉각재 환경에 노출되는 용접부 내면의 잔류응력은 재료의 응력부식균열 민감성에 영향을 주지 않을 것으로 판단된다. 한편, 안전단 배관 동종용접부의 연삭가공에 의해 내면의 잔류응력이 크게 상승하는 것으로 예측되었으므로, 내면의 연삭가공 이후 표면잔류응력 완화처리(예, 버핑)가 필요하다.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.297-310
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• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.

• Journal of Advanced Marine Engineering and Technology
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• 제19권1호
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• pp.28-35
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• 1995

The fluid flow for a energy transfer is essential for the design and operation of power plants, petrochemical plants and ships including a process. When the operating conditions of a plant are changed or any transitional event occured, the flow controls of a fluid must be performed to follow the new operating state or mitigate the results of a event. Generally these flow controls to accommodate the new operating state of a plant are made by the use of various valves. The refore the design of valves and the related techniques are very important to the system and component designs. However the system and component design are not familiar with the practical theory of the valve since the derivative procedures of the flow equations in a valve are difficult and it is not easy to found the theoretical foundamentals and informations about the design of a valve from the present references. In this study the flow equations applicable to a valve for liquid are theoretically derived in detail. And the definition of valve reynolds number and its boundary values between the tubulent and laminar flow is described compared with the values of a circular pipe flow.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2001년도 추계학술대회(한국공작기계학회)
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• pp.304-310
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• 2001

Flow pulsation often causes vibration and noise in piping systems and therefore has been a troublesome concern for fluid system engineers. According to frequency increase in this paper under the influence wave form of velocity in springly flow and viscosity are drop coefficient of viscosity become increase so that impedance and resistance. The transient variations of flow rate are measured by a modified impedance tube method which is realized by virtue of the present analytical technique. At pipe line in order to eliminate vibration, confirm happened intermittently impedance characteristics. We make a test and frequency analysis and have to minimize obstructive component at hydraulic circuit.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 1997년도 제26회 추계학술대회
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• pp.232-238
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• 1997

In this paper, a theoretical method is presented to design a hydraulic control valve system that consist of an important component in the hydraulic snubber. The hydraulic snubber is used essentially to support the piping systems at power plants. To calculate the force due to pressure drop and flow rate in the valve orifice and by-pass hole, Bernoulli equation is used. The Reynolds equation are numerically analyzed in the clearance gap between the valve cone and valve seat to estimate the friction force and leakage flow rate. Based on the detailed theoretical data, we developed successfully the hydraulic snubber for power plants.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.37-42
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• 2010

The stay cylinder weld at the steam generator of Korean Standard Nuclear Power Plants is safety class I component and is subjected to be inspected by the volumetric examination such as ultrasonic method. As accessibility of this area is limited due to the narrow space and high radiation, the existing manual inspection method involves various difficulties. Moreover operators may be exposed to internal contamination by contaminated dust during the surface buffing process to improve the inspection reliability of this area. Recently the new automatic inspection system for stay cylinder welds has been developed. The inspection system basically consists of a driving assembly, data acquisition device and signal processing units. The driving assembly is classified by 1) the scanner for inspecting and buffing the weld, 2) pillars for guiding the scanner and 3) the base frame for loading and supporting pillars. The scanner has 4 sensor modules to inspect in 4 refracted angles and 4 incident directions. These components can be inserted into the skirt of the stay cylinder through the manway hole and assembled easily by one-touch in the skirt. Data acquisition device and signal processing units developed in previous works are also newly upgraded for better processing of data analysis and evaluation. The system has been successfully demonstrated not only in the mock-up but also in the field. In this paper, newly developed inspection system for the stay cylinder weld of the steam generator is introduced and their field applications are discussed.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.40-45
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• 2008

There exist many types of pipe and component fatigue through vibrations, thermal fatigues or shifting. In some cases of thermal stratification/thermal fatigue, pipes & components are receiving thermal stress by means of material expansion and shrinkage by continuous thermal repetitive variation. Small cracks initially occur on the inside surface by thermal stress. These cracks grow in depth the pipe wall and finally come to a rupture. Pipe parts of susceptibility to thermal stratification and thermal fatigue are now being examined by conventional UT(ultrasonic test) as volumetric examination. It is difficult to fully satisfy the code & standards requirements because 3" weldolet weldments of RCS 16" pipe to 3" branch connection lines have complex structural shape. To solve the problems of conventional UT examination, we made a realistic mock-up and UT calibration block. We performed a simulation of phased array UT utilizing CIVA as NDE(Non-Destructive Examination) simulation software. Also we designed phased array UT transducer and wedge, optimal frequency by using simulation data. We performed phased array UT experiment through mock-up including artificial flaws(notch). The phased array UT technique is finally developed to improve the reliability of ultrasonic test at RCS 16" pipe to 3" branch connection weld.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.1-18
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• 2017

세계적으로 원자력발전소의 안정적 운영 및 안전성 확보를 위해 수명기간 중 주요 기기 및 배관의 실제 운전 과도상태를 체계적으로 관리하고, 피로 손상의 정량적 평가 및 관리를 위한 체계적인 시스템이 요구되고 있는 실정이다. 이에 본 논문에서는 원자력발전소의 안전등급 1 설비에 대한 피로 평가요건을 분석하였고, 피로 감시방법 및 절차와 웹 기반으로 개발된 피로 감시 시스템인 NuFMS 개발 및 검증 내용을 기술하였다. NuFMS는 설계 시 고려한 과도상태 발생 횟수 대 비발전소의 특정 운전 시점에서의 실제 발생 횟수를 비교하여 안전 여유도의 정량적 확인이 가능하며, 누적피로사용계수 도출을 통해 정확한 피로영향 분석뿐만 아니라 손상 관리가 가능하다. 이와 같이 NuFMS의 적용을 통해 원자력발전소 기기 및 배관의 피로 건전성을 확인하고 운영 신뢰도를 향상시킬 수 있으며, 발전소의 안전성 유지 및 운영비용 절감 등의 효과를 기대할 수 있다. 따라서 향후 국내 전 원전에 NuFMS를 확대 적용할 예정이며, 이러한 기술의 해외 수출을 적극 추진 중이다.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• 한국지반환경공학회 논문집
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• 제11권9호
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• pp.15-25
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• 2010

본 연구는 중앙심벽형 필댐인 주암댐과 임하댐을 대상으로 자동계측에 의해 연속적으로 생산되는 침투수량 계측자료에 대하여 분석을 통하여 필댐 고유의 특성에 따른 장기 침투특성과 댐의 안전관리 방법을 검토하고자 하였다. 필댐의 침투수량 계측값에는 내재 하는 강우 성분 등의 외부 요인의 영향으로 직접적으로 이상 누수의 발생을 검출하는 것은 어렵다. 이 때문에, 종래 저수위와 강우량을 고려하는 중회귀분석 등에 의해 누수량을 추정하는 방법이 적용되어 왔으나, 강우 성분의 추정 오차가 상대적으로 크고 정밀도가 불량한 것으로 알려졌다. 본 논문에서는 강우 성분의 분리해석을 통해 직접적으로 강우 성분에 영향을 받지 않는 저수지 수위에 연동하는 댐별 침투거동을 평가함과 아울러 분석대상 댐의 지형적, 수리지질학적 특성을 반영한 3차원 수치해석을 실시하여 계측 침투수량 자료와 비교하였다. 2개 대상댐의 침투거동은 각각의 고유한 특징을 가지고 있으며, 장기적으로 침투수량의 감소를 보여주고 있어 안정적인 상태로 나타났다. 또한, 수문곡선분리법은 침투수 안전관리 방법으로 적용가능한 것으로 판단되었다.