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신혜영(Hye-young Shin),나경환(Kyung-hwan Na),박영섭(Young-sheop Park) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5
A methodology for hardness measurement on calandria interior surface of CANDU reactors is developed. This is utilized as a tool for an assessment of neutron irradiation embrittlement effect on the calandria. The best representative measurement points and the most applicable hardness test methods are determined in this study, considering not only operating environments such as neutron flux, pressure and temperature but also a continuous operability of the calandria. These results are applicable to a typical CANDU calandria during the period of a replacement of pressure tubes. Though there inevitably remain some permanent marks on the calandria-tubesheet as a result of indentation for hardness measurement, it has not any negative effects on the integrity of the calandria during the continuous operations. This methodology is useful for assuring the suitability of long term operation of CANDU reactors.
APR1400 미국 설계인증을 위한 종합진동평가 심사서류 작성 방안
고도영(Do Young Ko),김동학(Dong Hak Kim),박영섭(Young Sheop Park) 한국소음진동공학회 2014 한국소음진동공학회 학술대회논문집 Vol.2014 No.10
In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.
외면 보수 용접이 원전 고온관 밀림노즐에서의 결함성장에 미치는 영향
나경환,윤은섭,박영섭,Na, Kyung-Hwan,Yun, Eun-Sub,Park, Young-Sheop 대한용접접합학회 2011 대한용접·접합학회지 Vol.29 No.2
Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.
신혜영(Hye-young Shin),정선미(Sun-mi Jung),박영섭(Young-sheop Park) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11
One of the technical issues addressed in Generic Letter 96-06 by the U.S. NRC is related to the possibility of thermally induced over-pressurization (TIP) of isolated penetration piping through the containment vessel during design basis accidents such as a loss-of-coolant accident (LOCA) or a main steam line break (MSLB). This is a safety-significant issue, because TIP can threaten the integrity of isolated sections of penetration piping, and this in turn can cause a failure of intended isolation function of the containment vessel during the conditions of a LOCA or MSLB. A procedure and a method to evaluate the structural integrity of the isolated piping are established for domestic application by studying the status and trends abroad to deal with this issue. Resolution results in the USA are surveyed and statistically analysed, research outputs are reviewed in order to be utilized in the steps of the evaluation, and the results is described in this paper.
임혁순(Hyuk-Soon LIM),박영섭(Young-Sheop PARK),이광한(Kwang-Han LEE),이석호(Seok-Ho LEE),정대율(Dae-Yul CHUNG) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.4
Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.
APR+ 대체교류발전기의 가스터빈 적용에 대한 민감도분석
문호림(Ho Rim Moon),박범락(Bhum Lak Park),박영섭(Young Sheop Park) 대한기계학회 2012 大韓機械學會論文集A Vol.36 No.1
원자력발전소의 대체교류발전기는 소내정전에 대처하기 위하여 설치되고 있다. 국내 원자력발전소에서 사용되고 있는 대체교류발전기의 구동형식은 대부분 디젤발전기이다. 가스터빈은 디젤발전기보다 구조가 간단하고 유지정비주기가 길다는 장점이 있지만 국내 원자력발전소에 적용사례가 없다는 단점이 있다. 본 논문의 목적은 APR+ 대체교류발전기의 구동형식을 선정하기 위하여 디젤발전기와 가스터빈의 민감도 분석을 하는 데 있다. 이를 위하여 디젤발전기와 가스터빈의 물리적 특성과 국내의 대체교류발전기 적용현황, 그리고US-APWR 의 비상발전기 및 대체교류발전기의 가스터빈 적용사례를 조사하였다. 마지막으로 신뢰도 데이터에 대한 민감도 분석과 신형 노형인 APR+에 가스터빈 대체교류발전기를 적용하여 노심손상빈도의 민감도를 분석하였다. Alternate alternating current (AAC) is used in nuclear power plants (NPPs) in order to cope with station black outs (SBOs). AAC has been provided using diesel engine drive types in Korea’s NPPs. The structure of gas turbine generators (GTGs) is simpler than that of diesel generators (DGs), and GTGs have the advantage of longer maintenance intervals. However, GTG-AAC was not used in NPPs in Korea because of the lack of operation/maintenance experience. The purpose of this paper is to analyze the safety of APR+ considering a diversity of AAC types. This paper analyzes reliability data, mechanical specifications of DGs and GTGs, and the sensitivity of core damage frequency to the ACC type.
윤은섭(Yun Eun Sub),나경환(Na Kyung Hwan),박영섭(Park Young Sheop) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11
In nuclear power plants, nickel base alloys that have high corrosion resistance and good mechanical properties have been used in tubes, small bore penetrations and dissimilar metal weld materials. But, since 2000, several PWSCC (Primary Water Stress Corrosion Cracking) occurrences have been reported worldwide in various pressure boundary components. PWSCC is the mechanism that forms cracks in susceptible materials in the presence of a corrosion environment and tensile stresses. If the materials and environments are fixed, PWSCC characteristics are mainly determined by tensile stresses. In this paper, PWSCC growth assessment of nickel base alloy dissimilar metal welds are introduced. To calculate the total stress, the operating stress and weld residual stress are considered. The FEA (Finite Element Analysis) method is applied to calculate the weld residual stress, and finally PWSCC growth rate is calculated.