• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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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.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1996.04a
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• pp.27-39
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• 1996

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized as average and linearly varying(through the thickness) stresses. The resulting stresses are interpreted per Section 111 of the ASME Boiler and Pressure Vessel Code from which the Primary(B$_1$) and Secondary(C$_1$) stress indices for pressure, the Primary(B$_2$) and Secondary(C$_2$) stress indices for moment are developed. Several analysis were peformed on various structural geometries in order to determine empirical relationships for the stress indices as a function of dimensionless ratios.

• Computational Structural Engineering
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• v.9 no.3
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• pp.115-123
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• 1996

A finite element analysis of a trunnion pipe anchor is presented. The structure is analyzed for the case of internal pressure and moment loadings. The stress results are categorized into the average (membrance) and the linearly varying(bending) stresses through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary (B/sub 1/) and Secondary(C/sub 1/) stress indices for pressure, the Primary(B/sub 2R/, B/sub 2T/) and Secondary(C/sub 2R/, C/sub 2T/) stress indices for moment are developed. Several analyses were performed for various structural geometries in order to obtain empirical representation for the stress indices in terms of dimensionless ratios.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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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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• v.51 no.6
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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.

• The KSFM Journal of Fluid Machinery
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• v.17 no.1
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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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• v.29 no.5
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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.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.30 no.3
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• pp.207-214
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• 2017

High voltage electric power transmitter GIS(Gas Insulated Switchgear) above 72.5kV needs to satisfy domestic Korean peninsular standard(ES-6110-0002) in KEPCO with respect to normal and special operation conditions which include internal gas pressure, dead weight, wind and seismic load. Some other requirements not described in Korean standard can be applied from other international standards such as IEC(International Electronical Committee) 62271-203 and 62271-207. The GIS is a kind of pressure vessel structure made of aluminum and filled with SF6 gas of internal pressure 0.4~0.5MPa. Finite element analysis of GIS is performed with such operational loads including seismic loading and the stability and reliability is determined according to ASME BPVC(Boiler and Pressure Vessel Code) SEC. VIII standard where the allowable stress level of the pressure vessel is suggested. The result shows that the stress of GIS is satisfied the allowable stress level and the safety factor is about 2.3 for Korean peninsular standard.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1994.10a
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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.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.17 no.4
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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.