• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.7-12
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• 2008

The purpose of this study is to present a methodology to estimate the capacity of the pressure relief valve which prevents overpressure of the pressure vessel in a cold state. In this methodology, the transient behavior of the flow rate through the pressure relief valve and the pressure inside the pressure vessel are considered. The result of this study shows the followings; The more the relief valve capacity is considered in excess, the more the initial relief flow rate and the initial pressure inside the pressure vessel are high and low respectively. When the relief valve capacity is determined properly, the pressure inside the pressure vessel maintains almost the same value, so the ASME code requirement will be met.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.492-500
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• 2004

The purpose of this study is localization of safety relief valves fur Nuclear Service through technical development with overall design, fabrication, inspection, capacity certification test and functional qualification test of safety relief valves in accordance with KEPIC MN Code(or ASME Sec.III ). The safely relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. But we're depending on technology of the other country up to the present time. Because we don't have our own technologies, we have been spent the great time and money on installing and repairing safety relief valve at nuclear power plant. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.34-45
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• 2001

We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.

• 한국안전학회지
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• 제16권4호
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2003년도 하계학술대회 논문집 B
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• pp.702-704
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• 2003

In this paper, analysis of bobbin probe signal in steam generator tube with bulge defect on CE system 80 nuclear power plant is represented. The CE system 80 steam generator is adopted in ULJIN-4 nuclear power plant. From Maxwell's equation, the electromagnetic governing equation for eddy current problem is derived and by performing the finite element formulation the 3-dimensional finite element code with brick element is developed. For the ease of the comparison the numerical results with experimental ones, the calculated signals are adjusted by using the ASME standard 100[%] through hole signal. For analysis of the effect of variation of the bulge depth on the impedance signal 0.2[mm] and 0.4[mm] depth of bulge defect signals are calculated and analyzed. As the depth of the bulge defect is increased, the magnitude of the signal is increased, too. But the rate of the increment of the signal is less than that of the depth of defect. From the result of this paper, we can obtained the information of the effect of bulge defect on the impedance signal.

• 소성∙가공
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• 제19권2호
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• pp.88-94
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• 2010

The next generation of diesel engine can operate at high injection pressure up to 1,800bar. The common rail pipe must have higher internal strength because it is directly influenced by the high-pressure fuel. Folding defects in the Common rail pipe can not ensure the structural safety. Therefore, Preform design and fatigue-life analysis are very important for preventing the head of the common rail pipe from folding in the heading process and for predicting fatigue life according to the amount of folding. In this study, a closed form equation to predict fatigue life was suggested by Goodman theory and pressure vessels theory in ASME Code in order to develop an optimization technique of the heading process and verified its reliability through fatigue-structural coupled field analysis. The results calculated by the theory were in good agreement with those obtained by the finite element analysis.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.525-525
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• 2005

This study is for the seismic analysis and the structural integrity evaluation of the box beam for supporting nuclear fuel-transfer-basket of the HANARO reactor pool. For performing the seismic analysis and evaluating the structural integrity in air or submerged condition, the finite element model of the fuel-transfer-basket and its supporting box beam(the coupled model) was developed. The hydrodynamic effect is also considered by using added mass concept. The seismic response spectrum analyses of the coupled model under the design floor response spectrum loads of Safe Shutdown Earthquake(SSE) were performed. Through the numerical experiments, the analysis results show that the stress values of the coupled model lot the structural integrity are within the ASME Code limits.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.546-554
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• 2023

The cyclical behavior of Alloy 617 was examined at 25 ℃ and high temperatures of 800, 850, 900, and 950 ℃ in air to obtain its fatigue life curves. The specimens tested at 25, 800, and 850 ℃ cyclically hardened, whereas those tested above 900 ℃ cyclically softened from the first cycle, that is, their fatigue life was reduced at high temperatures owing to loss of strength. Parameters of the typical Coffin-Manson-Basquin relationship were determined for each test temperature. Interestingly, no significant difference in fatigue life was observed for the specimens tested in the range of 800-950 ℃. Owing to the similarity in fatigue life, we determined fatigue strength and fatigue ductility exponents that could be applied for this temperature range. The parameters obtained were close to the universal slopes, although the fatigue ductility exponent was slightly different. The proposed fatigue life curves were compared with those presented in ASME code.

• 대한기계학회논문집A
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• 제39권4호
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• pp.405-413
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• 2015

최근 미국에서는 가동기간이 오래된 원전 매설배관에서 부식 및 침식에 의해 삼중수소 누설로 지하수가 오염되는 사례가 급증하고 있다. 따라서, 현재 원전 안전등급 매설배관으로 사용되고 있는 금속재료의 배관을 대신해서 부식 및 침식 등의 열화 손상에 대한 저항성이 우수한 고밀도 폴리에틸렌(HDPE) 배관을 ASME Code Class 3 안전계통 배관으로 사용하기 위한 연구가 수행되고 있다. 본 연구에서는 발전소 가동 중 매설배관에 가해질 수 있는 하중과 온도 범위를 바탕으로 HDPE 배관 융착부에 대한 인장 시험과 저속균열성장 (SCG) 시험을 수행하였다. 시험 결과로 얻은 SCG 시험편의 파단면을 분석하여 HDPE 재료의 파손 기구를 파악하였다. 이를 바탕으로 3D 유한요소 해석을 이용하여 균열이 있는 HDPE 재료가 버틸 수 있는 한계하중에 대한 검증을 수행하였다.

• Journal of Radiation Protection and Research
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• 제38권2호
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• pp.100-105
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• 2013

우리나라는 상용 및 연구용 원자로 수출, 그리고 국내 원자력 발전소의 추가 건설로 인하여 원자력 분야 전문 인력의 수요가 급증하고 있다. 이에 따라 원자력 인력 양성이 중요한 현안으로 대두되고 있다. 원자력 관련 주요 기관은 교육에 대한 절차와 자원들이 체계적으로 갖추어져 있지만 중소기업은 규모가 영세한 여건 때문에 교육이 어려울 수밖에 없는 실정이다. 본 연구에서는 '교육의 체계적인 접근법(Systematic Approach to Training: SAT)'을 도입하여 교육과정을 개발하고자 하였다. 이에 따라 중소기업을 대상으로 설문조사를 하였으며 그 결과를 바탕으로 방사선 분야 교육과정으로서 '방사선 기초 시범 교육'을 개발하고 운영한 결과를 요약하였다. '방사선 기초 시범 교육'은 기대감, 만족도, 강사 역량 등에서 4.0 (5.0 만점 기준) 이상의 높은 교육 참여자 평가 결과를 나타냈다. 수요 분석에 기반을 둔 '방사선 기초 시범 교육' 과정 개발 경험은 향후 수요 분석에서 나타난 원자력 발전 분야 및 ASME code 등의 교육과정 개발에 활용될 것이다.