• 한국지진공학회논문집
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• 제24권4호
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• pp.179-187
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• 2020

Seismic qualification of equipment including piping is performed by using floor response spectra (FRS) or in-structure response spectra (ISRS) as the earthquake input at the base of the equipment. The amplitude of the FRS may be noticeably reduced when obtained from coupling analysis because of interaction between the primary structure and the equipment. This paper introduces a method using a modal synthesis approach to generate the FRS in a coupled primary-secondary system that can avoid numerical instabilities or inaccuracies. The FRS were generated by considering the dynamic interaction that can occur at the interface between the supporting structure and the equipment. This study performed a numerical example analysis using a typical nuclear structure to investigate the coupling effect when generating the FRS. The study results show that the coupling analysis dominantly reduces the FRS and yields rational results. The modal synthesis approach is very practical to implement because it requires information on only a small number of dynamic characteristics of the primary and the secondary systems such as frequencies, modal participation factors, and mode shape ordinates at the locations where the FRS needs to be generated.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.579-586
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• 2011

During a reactor normal operation, a primary coolant was designed to remove the fission reaction heat of the reactor. When one pump is failure and the other pump shall supply the cooling water to cool the reduced power, it is necessary to estimate how much flow will be supplied to cool the reactor. We carried a flow net work analysis for two parallel pumping system as based on the piping net work of the primary cooling system in HANARO. As result, it is estimated that the flow of one pump increased than the rated flow of the pump below the cavitation critical flow.

• 한국안전학회지
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• 제14권1호
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• pp.25-32
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• 1999

The objective of this paper is to evaluate the fracture resistance characteristics of SA508 CL.1a carbon steel, TP347 stainless steel and their associated welds manufactured for primary coolant system of Ulchin 3,4 nuclear power plants. The effect of various parameters such as pipe size, welding method, chemical composition, crack plane orientation, metallography and fractography on the material properties were discussed. Test results showed that the effect of pipe size on fracture toughness is negligible while the effect of welding method on fracture toughness is significant. In addition, the drop of fracture toughness in the field fabrication weld of TP347 stainless steel is probably due to the large amount of $\sigma$-phase precipitated on the $\delta$-ferrite boundary and the large size dimples.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• 비파괴검사학회지
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• 제17권1호
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• pp.11-22
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• 1997

원자력발전소 압력용기 및 배관은 많은 용접부를 포함하고 있으며 용접부내 결함은 크기, 위치 및 형태에 따라 압력용기 및 배관의 건전성에 커다란 영향을 미친다. 따라서 주요 압력용기와 배관의 용접부에 대해서는 가동 전 중 검사시 초음파 탐상시험을 실시하여 그 건전성을 확인하고 있다. 초음파 결함 신호로부터의 결함 분류는 비파괴 평가에 있어 매우 중요하며 초음파 형상 인식 방법이 적당하다. 본 논문에서는 탄소강 압력용기 용접부에 내재하는 결함으로부터 얻어진 초음파 결함 신호의 형상 인식을 위한 절차로써 데이터 수집, 특징 추출, 특징 선택 및 결함 분류를 하였으며, 결함 분류에 있어 결함의 종류를 크게 선형(linear)과 체적(volumetric)의 두 종류로 분류함에 있어 퍼지이론을 적용하여 퍼지이론을 적용한 초음파 형상 인식 기법의 가능성 및 효율성을 제시하였고 그 결과 기존의 분류기(classifier)들에 비해 보다 우수한 결과를 얻을 수 있었다.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.28-32
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• 2019

In nuclear power plant operation and spent fuel canisters, it is necessary to provide a sound technical basis for the safety and security of long-term operation and storage respectively. Recently, the peening technology is being discussed and the technology will be adopted to ASME Section III, Division 1, Subsection NX (2019 Edition). The peening is prohibited in current edition, but it will be approved in 2019 Edition and adopted. However, Surface stress improvement techniques such as the peening is used to mitigate SCC susceptible in operating nuclear plants. Although the peening will be approved to ASME CODE, there are no performance criteria listed in the 2019 edition. The Korean International Working Group (KIWG) formed a new Task Group named "Advanced Surface Stress Improved Technology". The task group will develop a CODE CASE to address PWSCC(Primary Water Stress Corrosion Cracking) and CISCC(Chloride Induced Stress Corrosion Cracking) for new ASME Section III components. TG-ASSIT was started to make peening performance criteria for ASME Section III (new fabrication) applications. The objective of TG-ASSIT is to gain consensus among the relevant Code groups that requirements/mitigation have been met.

• 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.

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.78-82
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• 2010

The feeder pipe of Main Primary Heat Transfer System in Wolsong Nuclear Power Plant was inspected by the Ultrasonic Phase Array technique in 2010. It is the first time to apply this method to the construction at Nuclear Power Plant in Korea. The time required for UT technique is less than RT method. The UT method doesn't need to evacuate personnel who works nearby inspecting area and doesn't need to wait developing of film. For these reasons, the UT method is the fastest method among the volumetric inspections. As a result of the examination, it became clear that main defect of the feeder pipe is the Lack of fusion in the welded area. Moreover, the rate of defect was reduced gradually as improvement of welder's skill. If welding machine has problem, the defect has tended to same pattern(occurred same position in the welding area) but these defects were founded without specific rules. For these reasons, the creation of defect is dependent on the skill of worker not on the automatic welding machine. This evaluation of defect signal and collecting data would be useful to further examination in ISI.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.14-20
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• 2011

The end of design lifetime for Wolsong unit 1 will be reached on 20th November in 2012. So the license renewal documents for the continuous operation of Wolsong unit 1 is under reviewing now. Major components of primary system such as pressure tubes, feeder pipes including delayed neutron monitoring system tubing are being replaced and many components of secondary system are also being repaired. In this paper, the assessment on the wear degradation of delayed neutron monitoring system tubing(on the other hand, DN tube was called) was performed for the ageing management of the same component. The wear defects of this component was one of causes that resulted in heavy water leakage accidents. Therefore design specifications of Wolsong uint 1 and heavy water leakage accidents of pressurized heavy water reactors were reviewed and causes of wear defect for DN tubes were analyzed. Wear propagation equations based on the heavy water leakage history were made and the proper repairing time was possible to be expected if the continued operation was considered. Finally design change items of DN tubes that were conducted for the long term operation of Wolsong unit 1 are introduced.

• 한국화재소방학회논문지
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• 제16권1호
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• pp.24-29
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• 2002

본 연구는 배관에서의 정확한 압력손실과 유량을 측정하기 위하여 스프링클러 시스템중 격자배관방식에 대한 수리계산 프로그램을 개발하는데 그 목적이 있다. 본 프로그램 개발은 배관망 해석방법 등 여러 가지 자료로 근거하며, 격자배관방식이 국내에서 현행 시행되고 있는 규약배관방식의 단점을 보안하여 스프링클러 시스템을 성능기준화재안전설계로 이끌어 내기 위함이다. 소화설비의 전산화로 인한 작업의 편리성과 정확한 계산을 측정함으로서 전진적인 화재안전설계를 이룩하게 될 것이다. 이러한 프로그램의 개발은 미국 등에서 먼저 이루어져 오고 있으며 국내에서의 소방에 대한 발전이 한 단계 발전되는 계기를 마련하게 될 것이다.