• Journal of the Computational Structural Engineering Institute of Korea
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• v.13 no.3
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• pp.295-304
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• 2000

To evaluate the structural integrity for the strainer under seismic loading the seismic analysis and design were performed for T-type strainer in accordance with ASME, Section Ⅲ, Class 3(ND). Since there are no specified design requirements for the strainer in ASME Code, the strainer body was analysed according to ND-3500, valve design. Flanged joints connected with PCS piping were designed according to ND-3658.3. And the criteria for the cover flange was governed by the Appendix XI. Both a frequency analysis and an equivalent static seismic analysis of the strainer were carried out using the finite element computer program, ANSYS. The frequency analysis results show the fundamental natural frequency is greater than 33Hz, thus justifying the use of the equivalent static analysis through which membrane and bending stresses are obtained in the critical points near the branch connection area. The results of the seismic evaluation fully satisfied with the structural acceptance criteria of the ASME Code. Accordingly the structural integrity on the strainer body and flanges were proved.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.29 no.1
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• pp.1-8
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• 2016

In this paper, we perform structural analysis for a base frame which is used to support a motor for large centrifugal compressor drives and a safety assessment according to the concrete placement. First, the structural analysis about four loading conditions for the motor base frame was conducted and the structural safety was evaluated through both the maximum distortion energy theory and Mohr-Coulomb theory. It was possible to perform a more reasonable safety evaluation against local stresses occurring at the discontinuous portion of the fragile structural members by applying the safety assessment through ASME VIII Div. 2. In addition, the motor base frames with and without the internal concrete placement were quantitatively compared by the structural analysis and safety evaluation using ASME code and it was found to improve the structural integrity due to the concrete placement.

• 열병합발전
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• s.35
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• pp.4-11
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• 2003

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.70-77
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• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• Journal of the Korean Society of Safety
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• v.21 no.6 s.78
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• pp.20-25
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• 2006

In the design of class 1 components to apply ASME code section III NB, a fatigue is considered as one of the important failure mechanisms. Fatigue analysis procedure and standard fatigue design curve(S-N curve) is suggested in ASME code, which had to be performed to meet the integrity of components at the design step. As the plant life extension for operating power plants and the long-lived plant design, however, are being progressed, the fact which the existing ASME fatigue design curve can not consider fatigue effects sufficiently comes to the fore. To find the technical solution for these problems, a number of researches and discussion are continued up to now. In this study, the detailed fatigue analyses using the 3 dimensional modeling for the fatigue-weakened components were performed to develop the optimized fatigue analysis procedure and their results are compared with other reference solutions.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.20-24
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• 2015

During refueling Outage, In-service inspections(ISIs) for the Nuclear Power Plant components are mandatory requirement in accordance with ASME Code Sec. XI. Especially, in current ultrasonic testing is one of the most important NDT techniques that are used for volumetric examination methods for nuclear power plant components, and accurate sizing of flaw indication by UT is essential to assure the integrity of the components. However, ASME code specifies minimum requirement for vessel examination procedure, and so far many different flaw sizing approaches have been tried to apply. Through the Round Robin Test(RRT), the accuracy of ultrasonic flaw sizing using DAC techniques was measured with the mock-ups simulating typical pressure vessel welds. These mock-ups contain artificially introduced flaws of known size and location. This paper shows experimental comparison data on the accuracy of techniques using such as 6dB drop, 50%DAC, 20%DAC and 20%DAC with beam spread correction, and also shows that diverse DAC techniques can be effectively applied to the assessment of the flaw sizing for pressure vessel welds in the stage of welding and fabrication.

• The KSFM Journal of Fluid Machinery
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• v.8 no.3 s.30
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• pp.16-25
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• 2005

The purpose of this study is localization of safety relief valves for Nuclear Service. The safety 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. We developed design technology used FEM ' CFM about safety relief valve for Nuclear Service according to ASME (or KEPIC) Code and KHNP's Technical Specification. To prove validity of a design technology, actually, we manufactured and inspected and tested the sample products designed according to a developed technology. The capacity qualification test was achieved according to requirement of ASME(or KEPIC) Code by NBBI and the functional qualification test was achieved according to ASME QME-1 for operating condition in technical specification of KHNP by NLI. 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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• v.56 no.9
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• pp.3764-3774
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• 2024

In accordance with Regulatory Guide 1.207, Rev.1, fatigue assessments must be conducted considering the influence of primary coolant environment in nuclear reactors. Environmental fatigue, resulting from corrosion in the primary coolant, is evaluated in air fatigue life assessments through the application of an environmental fatigue correction factor. This environmental fatigue correction factor depends on sulfur content, operating temperature, dissolved oxygen, and strain rate. It remains constant for sulfur content, operating temperature, and dissolved oxygen, while strain rate introduces potential errors based on the analysis method. The current fatigue evaluation procedure for air, following ASME B&PV Code Sec.III, NB-3200, employs elastic analysis with a simplified elastic-plastic correction factor(Ke). However, Ke factor is considered excessively conservative, prompting less conservative alternatives proposed by JSME, RCC-M, ASME Code Case N-779. This study applied both ASME Ke and JSME Ke for fatigue evaluations considering environmental effects. Additionally, fatigue assessments accounting for elastic-plastic effects were conducted using Neuber and Glinka methods, compared with actual experiments. The analysis systematically examined changes in fatigue life and the environmental fatigue correction factor due to plastic effects in environmental fatigue evaluations.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.310-319
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• 1997

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 (membrane) stress, the linearly varying (bending) stress and the peak stress through the thickness. The resulting stresses are interpreted per Section III of the ASME Boiler and Pressure Vessel Code from which the Primary(B$_1$), Secondary(C$_1$) and Peak(K$_1$) stress indices for pressure, the Primary (B$_2$), Secondary(C$_2$) and Peak(K$_2$) stress indices for moment are developed. Based on the comparison between stress value by stress indices derived in this paper and stress value represented by the ASME Code Case N-391-1, the empirical equations for stress indices are effectively used in the piping stress analysis. Therefore, the use of empirical equations can simplify the procedure of evaluating the local stress in the piping design stage.