• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• 한국전산유체공학회지
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• 제17권4호
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• pp.1-8
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• 2012

During a reactor normal operation, two parallel 50% capacity cooling pumps circulate primary coolant to remove the fission reaction heat of the reactor through heat exchangers cold by a cooling tower. When one pump is failure, the other pump shall continuously circulate the coolant to remove the residual heat generated by the fuels loaded in the reactor after reactor shutdown. It is necessary to estimate how much flow rate will be supplied to remove the residual heat. We carried out a flow network analysis for the parallel primary pumps based on the piping network of the primary cooling system in HANARO. As result, it is estimated that the flow rate of one pump increased about 1.33 times the rated flow of one pump and was maintained within the limit of the cavitation critical flow.

• 대한기계학회논문집A
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• 제37권1호
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• pp.47-54
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• 2013

본 논문에서는 Alloy 82/182를 용접재로 이용한 원자로 배관 이종금속 맞대기 용접부(Dissimilar Metal Butt Weld)에서의 PWSCC에 의한 균열성장 거동을 평가하였다. 이를 위해 먼저 유한요소 응력해석을 수행하여 이종금속용접부에서의 응력분포를 결정하였으며, 이때 이종금속용접 및 동종금속용접에 의한 용접잔류응력 외에 수압시험과 정상운전 조건도 고려하여 기계적 하중에 의한 응력 재분배를 고려하였다. 최종적으로 이와 같이 구한 응력 분포를 바탕으로 PWSCC에 의한 축방향 및 원주방향 가상 균열의 균열성장 거동을 평가하여 PWSCC 균열 성장량을 계산하였다. 본 논문의 결과는 향후 PWSCC에 의한 원자로 배관 이종금속 맞대기 용접부의 균열성장 거동 예측에 적용될 수 있다.

• 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.

• 한국기계가공학회지
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• 제17권4호
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• pp.23-31
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• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

• 대한기계학회논문집A
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• 제32권9호
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• pp.770-781
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• 2008

In nuclear power plants, ferritic low alloy steel components were connected with austenitic stainless steel piping system through alloy 82/182 butt weld. There have been incidents recently where cracking has been observed in the dissimilar metal weld. Alloy 82/182 is susceptible to primary water stress corrosion cracking. Weld-induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operating. This paper presents residual stress computation performed by 6" safety & relief nozzle. Based on 2 dimensional and 3 dimensional finite element analyses, effect of welding variables on residual stress variation is estimated for sensitivity analysis.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1995년도 춘계학술대회논문집; 전남대학교, 19 May 1995
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• pp.318-321
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• 1995

원자력 발전소의 안전성 보장 및 신뢰성 향상을 위하여 시운전 단계에서 원자력 발전소내 안전등급에 해당하는 배관계통의 상태 확인을 위하여 각종시험을 하도록 되어있다. 특히 새로운 설계기념, 크기 또는 용량을 갖는 원자로 모델에 대해서는 필수적으로 건전성 평가를 하게 되었다. 이를 위해 발전소 건설기간에 시행하는 고온 기능시험 중에 원자로 주변 주요 시스템인 원자로 냉각재 루프 계통에 대한 건전성 확인을 위해 압전형 고온 가속도 센서를 이용하여 정상운전상태의 진동을 측정하여 시스템 진동거동을 규명하였다. 배관시스템의 일상운전상태는 유체의 흐름과 기기운전이 일정한 정상상태와 펌프의 기동 또는 정지 및 밸브의 급격한 개폐등으로 발생하는 과도상태로 나눌 수 있다. 따라서 두 가지 상태의 진동을 측정해야 한다. 배관계통은 정상운전 상태로 설계수명을 유지할 수 있어야 하므로 정상진도잉 최소화 되어야 한다. 진동 평가기준은 배관재질의 응력(S/N 커브) 곡선을 참조하여 설계수명내에 손상이 일어나지 않도록 재료의 허용응력을 산정하고 이를 진동변위로 환산하여 정한 것이며 이 값에 측정 데이타를 비교하여 1차계통 배관의 건전성을 확인하였다.