• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 춘계학술발표회논문집(2)
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• pp.943-949
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• 1995

원자력 발전소의 설비중 가압기 밀림관 노즐의 피로 잔존 수명에 대해 연구하였다. 원자력발전소 운전중 발생하는 각종 천이상태에 의해 밀림관 노즐에 작용하는 열응력과 역학적 응력을 상용 유한요소법 펙케지를 이용하여 계산하였다. 계산된 응력값들과 ASME Boiler and Pressure Vessel Code를 이용하여 가압기 밀림관 노즐의 피로 잔존 수명을 평가하였다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1996년도 봄 학술발표회 논문집
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• pp.27-39
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• 1996

배관을 구속시키기 위한 원통형 배관 지지대 (Trunnion Pipe Support)가 부착된 배관의 응력해석 을 위하여 유한요소해석을 사용하였다. 해석결과로 부터 얻어진 응력은 두께에 대한 평균 및 선 형 응력으로 분류 되었으며, 분류된 응력값은 압력에 의한 일차응력계수(B$_1$)와 이차응력계수(C$_1$), 모멘트에 의한 일차응력계수(B$_2$)와 이차응력계수(C$_2$)를 추정하기 위하여 ASME Code에 정의된 것과 일치하게 해석되었다. 무차원의 함수로써 응력계수에 대한 경험식을 개발하기 위하여 여러 모델의 해석을 수행하였다.

• 전산구조공학
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• 제9권3호
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• pp.115-123
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• 1996

배관을 구속시키기 위한 원통형 배관 지지대(Trunnion Pipe Support)가 부착된 배관의 응력해석을 위하여 유한요소해석을 사용하였다. 해석결과로 부터 얻어진 응력은 두께에 대한 평균(막응력) 및 선형 응력(굽힘응력)으로 분류되었으며, 분류된 응력값은 압력에 에 의한 일차응력계수(B/sub 1/)와 이차응력계수(C/sub 1/), 모멘트에 의한 일차응력계수(B/sub 28/, B/sub 2T/)와 이차응력계수(C/sub 28/, C/sub 2T/)를 추정하기 위하여 ASME Code에 정의된것과 일치하게 해석되었다. 무차원의 함수로써 응력계수에 대한 경험식을 개발하기 위하여 여러 모델의 해석을 수행하였다.

• 대한기계학회논문집A
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• 제27권5호
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• pp.724-733
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• 2003

Fatigue design rules for welds in the ASME Boiler and Pressure Vessels Code are based on the use of Fatigue Strength Reduction Factors(FSRF) against a code specified fatigue design curve generated from smooth base metal specimens without the presence of welds. Similarly, stress intensification factors that are used in the ASME B3l.1 Piping Code are based on component S-N curves with a reference fatigue strength based on straight pipe girth welds. But the determination of either the FSRF or stress intensification factor requires extensive fatigue testing to take into account the stress concentration effects associated with various types of component geometry, weld configuration and loading conditions. As the fatigue behavior of welded joints is being better understood, it has been generally accepted that the difference in fatigue lives from one type of weld to another is dominated by the difference in stress concentration. However, general finite element procedures are currently not available for effective determination of such stress concentration effects. In this paper, a mesh-insensitive structural stress method is used to re-evaluate the S-N test data, and then more effective method is proposed for pressure vessel and piping fatigue design.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1681-1688
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• 2019

In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

• 한국유체기계학회 논문집
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• 제17권1호
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• pp.5-11
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• 2014

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

• Journal of Advanced Marine Engineering and Technology
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• 제29권5호
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• pp.571-578
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• 2005

Stresses of main stop valve and control valve casings for the steam turbines of power plants are analyzed by the finite element method. The stress intensity is obtained to check the results on the basis of the design criteria of ASME boiler and pressure vessel code. To verify accuracy of the finite element analysis. analyzed stresses are compared with those measured during the hydrostatic pressure test. Stress category drawings. which play an important role in evaluating casting defects, are produced from the analysis results, and important points in casting of the valve casings are discussed in terms of the stress category.

• 한국전산구조공학회논문집
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• 제30권3호
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• pp.207-214
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• 2017

국내의 72.5kV 이상, 주파수 60Hz의 송배전설비인 옥내 및 옥외용 가스절연개폐장치(GIS)는 내진 안전성에 대해 국가에서 정한 한전표준규격(ES-6110-0002)을 만족해야 한다. 이 규격에서 명시되지 않은 사항은 IEC 62271-203, 62271-207 등의 관련 기기 규격에 준한다. 한전표준규격에서 기기는 정상사용상태와 특수사용상태에서 건전성이 유지되어야 한다. 안전성 판단을 위해 ASME BPVC SEC.VIII 내압용기 설계 기준에 의해 A6061-T6 재질의 GIS에 대한 정상사용상태 기준과 국내 한전표준규격과 국외 IEC 62271-207에 의한 특수사용상태 기준(지진)에 대한 총체적 응력상태를 판단하였다. 한전표준규격 기준(0.22g) 적용시, 최종응력이 알루미늄인 Part A는 78.2MPa, Part D2의 경우 102.3MPa로, ASME 허용응력 값 181.5MPa를 만족하고 있다. IEC 62271-207 High 0.5g의 경우에도 최종응력은 Part A는 90.5MPa, Part D2는 103.8MPa이다. 본 연구 결과, 72.5kV GIS는 한전표준규격의 구조안전성과 내진성능을 충분히 만족함을 보이고 있다. 내진해석으로 내진시험을 수행할 수 없는 대형 전력기기의 내진성능 실증에 활용될 수 있을 것으로 기대된다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1994년도 가을 학술발표회 논문집
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• pp.129-136
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• 1994

Spherical storage tank for chemical plant is analyzed for the loads and their combinations in accordance with Section Ⅷ, Division 2 of the ASME Boiler and Pressure Vessel Code. Design Analysis of Butene storage tank is carried out by utilizing 3-dimensional plate and beam elements of a general purpose finite element program. Two separate 3-D finite element models are used; one for the global analysis of the entire spherical storage tank, the other for the local analysis of junction part and its vicinity of shell-to-supporting structures. The analysis is focused on the equator plate in the shell and the junction part of shell-to-supporting structures.

• 한국기계가공학회지
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• 제17권4호
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• pp.23-31
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• 2018

Stress of a steam separator, equipment of the high-pressure (HP) evaporator for a HRSG, was analyzed and evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the analysis results of the piping system model of the HP evaporator, reaction forces of the riser tubes connected to the steam separator, i.e., nozzle loads, were derived. Next, a finite element model of the steam separator was constructed and analyzed for the design pressure and the nozzle loads. The results show that the maximum stress occurred at the bore of the riser nozzle. The primary membrane stresses at the shell and nozzle were found to be less than the allowable stress. Next, the steam separator was analyzed for the steady-state operating conditions of operating pressure, operating temperature, and nozzle loads. The maximum stress occurred at the bore of the riser nozzle. The primary plus secondary membrane plus bending stress at the shell and nozzle was found to be less than the allowable stress.