• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.71-76
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• 2019

Actual stress distributions in the nozzle of a pressure vessel may not be in plane strain condition, implying that the crack-tip constraint condition may be relaxed in the nozzle. In this paper, a methodology for evaluating the fracture toughness of the ASME Code is presented considering the relaxation of the constraint effect in the nozzle of the reactor pressure vessel. The crack-tip constraint effect is quantified by the T-stress. The equation, which represent the relation between the fracture toughness in the lower constraint condition and the plane strain fracture toughness, is derived using the T-stress. This equation is similar to the method for evaluating the fracture toughness of the Master Curve for low constraint conditions. As a result of evaluating the fracture toughness considering the constraint effect in the reactor inlet, outlet and direct injection nozzles using the proposed equation, it was confirmed that the fracture toughness in the nozzles is higher than the plane strain fracture toughness. Applying the proposed evaluation methodology, it is possible to reflect the relaxation of the constraint effect in the nozzles of the reactor pressure vessel, therefore, the safe operation area on the pressure-temperature limit curve can be prevented from being excessively limited.

• 한국산학기술학회논문지
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• 제21권1호
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• pp.487-493
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• 2020

최근 환경부의 강화된 규제로 건축물 아파트 등 산업시설을 도장함에 있어 페인트를 유압에 의해 자동으로 이송시켜 스프레이식으로 도장하는 기존 에어리스 방식의 사용을 금지하였다. 따라서 기존의 에어리스 도장장치에 스프레이 대신 부착하여 사용할 압송롤러개발이 필요한데, 현재 개발되어 판매 중인 압송롤러는 롤러를 구성하는 SUS배관 및 드럼베이스의 홀 설계에 대한 자세한 연구 사례가 없어 그 효율성이 의심이 되는 바이다. 특히, 롤러를 이용한 도장 시롤러 양단에 페인트가 충분히 공급되지 못하면 도장 후 붓을 이용한 재작업이 필요하게 되므로, 롤러 양단에 페인트가 잘 유입될 수 있도록 드럼베이스 홀을 분포시킬 필요가 있다. 본 연구에서는 SUS배관 및 드럼베이스에 가공되는 홀의 형태가 페인트의 유입양상에 미치는 영향을 알아보고 보다 효율적인 홀 설계 방안을 제시하기 위하여 SUS배관 및 롤러에 대하여 전산유체해석을 수행하였다. 해석결과 SUS배관에서의 홀 배치는 배관에 균일하게 배치하는 것이 균형 있게 페인트를 유입시키기에 좋고, 드럼베이스의 경우는 홀을 드럼베이스 양단에 밀집하여 가공하면 페인트가 롤러 양단에 더 잘 유입시킬 수 있음이 검증되었다.

• 한국재난정보학회 논문집
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• 제18권3호
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• pp.550-559
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• 2022

연구목적: 파이프는 기계, 전자, 전기, 플랜트 등 많은 산업 분야에서 응용기기로 널리 사용되고 있으며, 소방, 화학 등 안전 관련 분야에서도 널리 사용되고 있습니다. 제품의 다양화에 따라 배관 분야에서도 기술의 중요성이 높아지고 있습니다. 특히 기존 동관을 스테인리스강으로 변경하는 경우 구조해석이나 유동 해석을 통해 안전성과 유동특성을 평가할 필요가 있다. 연구방법: 자체 개발한 일체형 인서트형 커넥터인 6.35 소켓 모델의 유동 안전성은 CFD 해석을 이용하여 유동유발진동(FIV) 평가 과정의 4단계를 통해 진행하였다. 연구결과: 배관계 벽면에 작용하는 압력변동의 진폭은 3,780Pa이하의 수준으로 형성되며, 이는 냉매 배관의 운전압력이나 설계응력과 비교했을 때 매우 작은 수준의 압력으로, 난류에 의한 진동이 배관의 구조안전성에 미치는 영향은 미미한 수준인 것으로 나타났다. 결론: 유동 해석을 통하여 후크조인트 체결 시 발생하는 압력 및 진동 등을 검토한 결과 일반적인 배관계의 고유진동수가 50Hz 이하에서 형성되는 것을 고려했을 때, 난류유동에 의해 유발되는 진동이 배관계의 공진을 발생시킬 가능성은 희박한 것으로 판단된다.

• 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.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.42-47
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• 2011

The purpose of this study is to investigate the effects of through thickness on the mechanical properties and microstructural features in Mod. 9Cr-1Mo steels for RPVs. The microstructures at all locations were typically tempered martensite, but small amount of delta ferrite was observed at the center region. The prior austenite grain size increased with the depth from the surface. The yield strengths of center and 1/4T location were higher than that of surface by 30MPa. The impact toughness of center was low compared to those of other specimens. Also, upper shelf energy was low at the center. The toughness deterioration in center might be caused by larger size of the prior austenite grains and existence of the delta ferrite.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.120-125
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• 2007

Put Abstract text here The strain distribution measurement for wall thinned pipe bends by ESPI is presented. Defect types observed in the steel piping in the nuclear power plants (NPP) are the crack at the weld part and the wall thinning defect in the pipe bends. Especially, the wall thinning defects in the pipe bends due to the flow-accelerated corrosion (FAC) is a main type of defects observed in the carbon steel piping system. ESPI is one of the optical non-destructive testing methods and can measure the stress and the strain distribution of the object subjected by the tensile loading or the internal pressure. In this paper, the strain distribution of the wall thinned pipe bends due to the internal pressure will be measured by ESPI technique and the results are discussed. From the results, the size of the wall thinning defect can also be measured approximately.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.24-31
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• 2018

The mechanical design procedure is studied for the PCHE(printed circuit heat exchanger) with electrochemical etched flow channels. The effective heat transfer plates of PCHE are assembled by diffusion bonding to make a module. PCHE is widely used for industrial applications due to its compactness, cost efficiency, and serviceability at high pressure and/or temperature conditions. The limitations and technical barriers of PCHE are investigated for application to nuclear components. Rules for design and fabrication of PCHE are specified in ASME Section VIII but not in ASME Section III of nuclear components. Therefore, the calculation procedure of key dimensions of PCHE is defined based on ASME section VIII. The effective heat transfer region of PCHE is defined by several key dimensions such as the flow channel radius, edge width, wall thickness, and ridge width. The mechanical design procedure of key dimensions was incorporated into a program for easy use in the PCHE design. The effect of assumptions used in the key dimension calculation on stress values is numerically investigated. A comparative analysis is done by comparing finite element analysis results for the semi-circular flow channels with the formula based sizing calculation assuming rectangular cross sections.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.