• 한국안전학회지
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• 제21권2호
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• pp.70-79
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• 2006

many plumbing system are needed in the ceiling of the building as it becomes advanced more and more. This leads to make effective space between ceiling level and slab less. Also, piping system is not suitably arranged and operated if it is bent around the columns which they are a lot. But this system can be more effective if it passes through the columns directly. Most people think that those columns should not be damaged with such as holes. But actually this is existed in a hotel building in switzerland. This study is to fing out how much capacity the columns become damaged and low using model size of $20cm{\times}30cm$ rectangular section, and 160cm long, in the structural test. it's compressive strength is focused on $240kg/cm^{2}$ design strength, commonly used in korea. Compressive test for them was done at Hanyang University using UTM one thousand tone(1000t) capacity. Variable numbers for the study are one hole of dia 3cm with distance 20cm or 40cm, two holes of dia 3cm with 20cm and 40cm distance, one hole of dia 5cm with distance 20cm and 40cm, two holes of dia 5cm with 20cm and 40cm distance, me eccentric hole with 20cm and 40cm distance, Normal(without hole). two test specimens of each variable are made for the test. ED5H20 capacity was 16.7% decreased, compared to normal one. While ED5H40 distant 40cm from the end of column top showed 19.5% capacity decrease, compared to normal one. Strain of ED5H20 diameter 5cm, in distance of 20cm form the top of the column was less 5% than the one of diameter 3cm. Finally, conclusions are that in case of hole diameter 3cm, located at 20cm from the end of the column top, capacity was decreased down to 3, percent only compared to the same diameter hole with 20cm distant from the end of it.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.1-6
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• 2011

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established a small-scale gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype made of Hastelloy-X to be tested in the small-scale gas loop. Results from the elastic structural analysis on the PHE prototype were reported in the previous article. In order to investigate the macroscopic structural characteristics and behavior of the PHE prototype under the test condition of the small-scale gas loop far more in detail, elastic-plastic high-temperature structural-analysis of the PHE prototype was carried out in this study.

• Corrosion Science and Technology
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• 제14권6호
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• pp.325-330
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• 2015

Pipe wall-thinning by flow-accelerated corrosion (FAC) is a significant and costly damage of secondary system piping in nuclear power plants (NPPs). All NPPs have their management programs to ensure pipe integrity from wall-thinning. This study analyzed the pipe wall-thinning caused by changing the amine, which is used for adjusting the water chemistry in the secondary system of NPPs. The pH change was analyzed according to the addition of amine. Then, the wear rate calculated in two different amines was compared at the steam cycle in NPPs. As a result, increasing the pH at operating temperature (Hot pH) can reduce the rate of FAC damage significantly. Wall-thinning is affected by amine characteristics depending on temperature and quality of water.

• 해양환경안전학회지
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• 제23권4호
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• pp.400-406
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• 2017

선박용 입형 편흡입 원심 해수펌프의 회전체와 파이프 시스템 내부에 발생한 과도한 손상의 원인을 분석하고자 실제 모델을 역설계 하였다. 이를 위해 전산유체역학(CFD)을 이용하여 해수냉각펌프의 내부유동 분석을 실시하였다. 결과적으로 역설계를 통한 대상펌프의 형상 및 경계조건을 설정하여 설계 운전점인 $125m^3/h$에서의 펌프 효율은 85.3 %, 양정 32.0미터로 계산되었다. 대상 펌프의 최고 효율점은 $150m^3/h$에서 약 86.2 %로 예측되었으나 실제 운전점과는 차이가 있었다. 최저 유량점인 $112.5m^3/h$에서는 저 유량점 특성상 유동이 불안정하여 해석의 수렴이 좋지 않았다. 선박에서 운전 중인 해수펌프 및 파이프 시스템 전반에 걸쳐 진행 중인 공동현상의 원인 규명을 통해 개선방안을 도출하고자 하였으나, 입 출구 전체 파이프 시스템의 계산, 각 지관들에 대한 유량 및 유속 측정 등을 통해 보다 명확한 원인분석을 위한 후속연구가 필요하다.

• 플랜트 저널
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• 제11권1호
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• pp.50-57
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• 2015

연구에서는 국내 D 석탄 화력발전소에서 비회 이송량 35,800 kg/h의 용량으로 운전 중인 비회 이송설비를 대상으로 최대 비회 이송량을 예측하였다. 수평거리 550 m, 수직거리 40 m, 엘보우 9개소, 직경 0.254 m의 이송관으로 구성된 비회 이송관로와 트립(trip) 정압 1,163 mmH2O, 풍량 5,040 m3/h인 용적식 비회 이송 송풍기로 이루어진 비회 이송 시스템에서 최대 비회 이송량은 비회 이송량의 증가에 따른 비회 이송 시스템의 압력 손실과 용적식 비회 이송 송풍기의 트립 정압이 같아질 때이며, 이 조건 하에서 가능한 최대 비회 이송량은 52,600 kg/h로 예상되었다.

• 한국초전도ㆍ저온공학회논문지
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• 제16권3호
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• pp.15-20
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• 2014

The heavy ion accelerator that will be built in Daejeon, Korea utilizes superconducting cavities operating in 2 K. The cavities are QWR (quarter wave resonator), HWR (half wave resonator), SSR1 (sing spoke resonator1) and SSR2. The main role of the cryomodule is supplying thermal insulation for cryogenic operation of the cavities and maintaining cavities' alignment. Thermal and structural consideration such as thermal load by heat leak and heat generation, cryogenic fluid management, thermal contraction, and so on. This paper describes detailed design considerations and current results have being done including thermal load estimation, cryogenic flow piping, pressure relief system, and so on.

• 비파괴검사학회지
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• 제9권1호
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• pp.8-21
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• 1989

This paper presents the ultrasonic characteristics of weldment and detectability of defects of weldment in Austenitic Stainless Steel Type 304 that is composed of mostly coolant piping system in nuclear power plants. The results of this experient show as follows: 1. When the ultrasonic beam detects the defects on the side of base metal and on the opposite side of weldment, the indications which was detected on the screen show different amplitude and different metal path each. 2. The ultrasonically estimated notch depth is generally oversized than actual notch depth. 3. It is easy for the false indication to show up on the screen because of columnar structure of weldment in austenitic stainless steel. 4. The higher frequencies of transducer have more difficulties to detect the defects of the opposite side of weldment because of ultrasonic attenuation in weldment and the longitudinal transmitter-receiver transducer is the most effective in detecting the opposite side defects of weldment.

• 한국CDE학회논문집
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• 제12권1호
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• pp.74-85
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• 2007

During the lifecycle of a nuclear power plant many organizations are involved in KOREA. Korea Plant Engineering Co. (KOPEC) participates in the design stage, Korea Hydraulic and Nuclear Power (KHNP) operates and manages all nuclear power plants in KOREA, Dusan Heavy Industries manufactures the main equipment, and a construction company constructs the plant. Even though each organization has a digital data management system inside and obtains a certain level of automation, data sharing among organizations is poor. KHNP gets drawing and technical specifications from KOPEC in the form of paper. It results in manual re-work of definition and there are potential errors in the process. A data warehouse based on a neutral model has been constructed in order to make an information bridge between design and O&M phases. GPM(generic product model), a data model from Hitachi, Japan is addressed and extended in this study. GPM has a similar architecture with ISO 15926 "life cycle data for process plant". The extension is oriented to nuclear power plants. This paper introduces some of implementation results: 1) 2D piping and instrument diagram (P&ID) and 3D CAD model exchanges and their visualization; 2) Interface between GPM-based data warehouse and KHNP ERP system.

• 비파괴검사학회지
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• 제5권2호
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• pp.22-33
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• 1986

The application and development of on-line monitoring technology of AE to surveillance of crack propagation will contribute to the structural integrity of reactor pressure vessel and piping system. This research has been performed in order to obtain the evaluation technology for source location of AE and the analysis for the AE signal of the welded specimen. AE is detected by 4-channels AE system during pressurization in small pressure vessels. The cracking of artificial defects can be accurately located and categorized in real time. The welded specimens have more events rate and higher amplitude than the weldless less specimens, and the events rate have a peak around the yield point and just before the failure under tensile test.

• 비파괴검사학회지
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• 제12권2호
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• pp.14-20
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• 1992

Stress analysis has been performed using computer program ANSYS in the pressurizer surge line in accordance with ASME Sec. III in order to predict possibility of fatigue failure due to thermal stratification phenomena in pipes connected to reactor coolant system of nuclear power plants. Highly vulnerable area to crack generation has been chosen by the analysis of fatigue due to thermal stress in pressurizer surge line. This kind of result can be helpful to choose the location requiring intensive care during inservice inspection of nuclear power plants.