• Proceedings of the KSME Conference
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• 2003.04a
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• pp.654-659
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• 2003

PVAP (Pressure Vessel Analysis Program V1.0) was developed by adopting the finite element analysis program ANSYS V6.0, and Microsoft Visual Basic V6.0 was also utilized for the interfacing and handling of input and output data during the analysis. PVAP offers the end user the ability to design and analyze vessels in strict accordance with ASME Section VIII, Division 2. More importantly, the user is not required to make any design decisions during the input of the vessel. PVAP consists of three analysis modules for the finite element analysis of the primary components of pressure vessel such as head, shell, nozzle, and skirt. In each module, finite element analysis can be performed automatically only if the end user gives the dimension of the vessel. Furthermore, the calculated results are compared and evaluated in accordance with the criteria given in ASME Boiler and Pressure Vessel Code, Section VIII, Division 2. In particular, heat transfer analysis and consecutive thermal stress analysis for the junction between skirt and head can be carried out automatically in the skirt-tohead module. Finally, report including the above results is created automatically in Microsoft Word format.

• 한국기계연구소 소보
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• s.20
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• pp.49-55
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• 1990

Filament-Wound composite pressure vessels offer significant weight saving over the conventional metal pressure vessels. S-2 glass/epoxy pressure vessel was designed, fabricated and tested. It was considered, in the analysis that the matrix and the metal liner of FRP pressure vessel shared the internal pressure. The thicknesses of hoop and helical layers were determinded by the computer program developed for the design of FRP/Metal structural pressure vessel. The computer program increased the structural efficiency about 14% comparing with the netting analysis. The experimental measurement on the deformation by internal pressure agrees with the theoretical analysis within the accuracy of 5%.

• Journal of Welding and Joining
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• v.18 no.4
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• pp.28-37
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• 2000

In this paper the LNG vessel of the Moss type which is capable of lifting 15,261 tons is investigated in the view point of the pressure vessel preliminary design using the finite element method. The Pressure vessel design is based on the equivalent stress levels due to the internal pressure. The finite element model of the spherical pressure vessel is configured using 4 noded quadrilateral shell element. The finite element analysis program NASTRAN and ANSYS 5.5are implemented. The design is compared with the three kinds of the boundary condition : first, where the equator of the pressure vessel is fixed, and where the top and is fixed, and, the bottom end is fixed, respectively. A comparison is presented between the results obtained by the finite element model and by the prototype production model. Additionally just below position(case 1 & case 2) of equator ring was carried out by using ANSYS 5.5. The results show that the vessel design based on the stress is acceptable at the preliminary design.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.275-280
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• 2008

To improve the durability of a hydrogen pressure vessel which is applied high-pressure, it needs the autofrettage process which induces compressive residual stress in the Aluminum liner. This study presents the elasto-plastic analysis to predict the behavior of structure accurately, and the Tsai-Wu failure criterion is applied to predict failure of pressure vessel of Aluminum liner and composite materials. Generally, plastic analysis is more complex than elastic analysis and has much time to predict. To complement its weakness, the AxicomPro(EXCEL program), applied radial return algorithm and nonlinear classical laminate theory (CLT), is developed for predicting results with more simple and accurate than the existing finite element analysis programs.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.12
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Journal of Welding and Joining
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• v.19 no.1
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• pp.104-111
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• 2001

This study presents an analysis method for heat flow and deformation of sheet metal laser welding. A heat source model for 2-dimensional heat flow analysis of laser welding process was suggested in this paper. To investigate the availability of the heat source model, the analysis results were compared and estimated with the results of previous researches. We could get a good agreement between the results of numerical analysis and experiments in the temperature distribution of weldment. Due to the characteristics of welding process, some kinds of deformations are usually generated in a welded structure. Generally, the degree of deformation is dependent on the welding sequence constraints as well as input power Therefore, in this paper we evaluate the deformation of gas pressure vessel according to the welding sequence and input power. In the analysis of weld deformation, 2-dimensional thermo-elasto-plastic analysis was performed for the gas pressure vessel by using a commercial FE program package.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Applied Chemistry for Engineering
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• v.33 no.6
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• pp.653-660
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• 2022

In liquid hydrogen storage tanks, tank damage or leakage in the surrounding pipes possess a major risk. Since these tanks store huge amounts of the fluid among all the liquid hydrogen process facilities, there is a high risk of leakage-related accidents. Therefore, in this study, we conducted a risk assessment of liquid hydrogen leakage for a grid-type liquid hydrogen storage tank (lattice-type pressure vessel (LPV): 18 m³) that overcame the low space efficiency of the existing pressure vessel shape. Through a commercially developed three-dimensional computational fluid dynamics program, the geometry of the site, where the liquid hydrogen storage tank will be installed, was obtained and simulations of the leakage scenarios for each situation were performed. From the computational flow analysis results, the pool formation behavior in the event of liquid hydrogen leakage was identified, and the resulting damage range was predicted.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.74-83
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• 1995

The accurate determination of the fast neutron flux/fluence onto the pressure vessel is an essential part of the reactor pressure vessel surveillance program. It has been reported recently that the iron cross section data in ENDF/B versions III through V might underestimate the flux/fluence of fast neutrons in steel structures such as reactor pressure vessel. In this study, for the comparison of iron data of ENDF/B-IV and VI we produced two 47-group cross section sets, CXFe-IV and CXFe-Ⅵ, which are based on Yonggwang nuclear unit-3/4 model and the iron data of ENDF/B-IV and VI, respectively. A comparison was made of the results obtained from DOT4.3 calculation using CXFe-IV and CXFe-VI. From the results, it was found that the fast flux(E 〉 1.0 MeV), which is important for the pressure vessel embrittlement analysis, increases by about 7.6％ at the inner wall and 20％ at the outer wall of the vessel, if the iron data are used from ENDF/B-VI instead of ENDF/B-IV.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.470-484
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• 1998

An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.