• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.410-423
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• 2000

A two-dimensional numerical model is developed and applied to the LAVA-EXV tests performed at the Korea Atomic Energy Research Institute (KAERI) to investigate the external cooling effect on the thermal margin to failure of a reactor pressure vessel (RPV) during a severe accident. The computational program was written to predict the temperature profile of a two-dimensional spherical vessel segment accounting for the conjugate heat transfer mechanisms of conduction through the debris and the vessel, natural convection within the molten debris pool, and the possible ablation of the vessel wall in contact with the high temperature melt. Results of the sensitivity analysis and comparison with the LAVA-EXV test data indicated that the developed computational tool carries a high potential for simulating the thermal behavior of the RPV during a core melt relocation accident. It is concluded that the main factors affecting the RPV failure are the natural convection within the debris pool and the ablation of the metal vessel, The simplistic natural convection model adopted in the computational program partly made up for the absence of the mechanistic momentum consideration in this study. Uncertainties in the prediction will be reduced when the natural convection and ablation phenomena are more rigorously dealt with in the code, and if more accurate initial and time-dependent conditions are supplied from the test in terms of material composition and its associated thermophysical properties.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.44-50
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• 2014

There were many flaw issues of reactor vessel head penetration in USA fleets. USNRC issued 10CFR50.55a to implement reactor vessel head penetration ultrasonic examination performance demonstration(PD) in US for enhancement of inspection reliability. After September 2009, all US utilities inspected their RVHP with PD qualified system. Korea Hydro and Nuclear Power Company(KHNP) have developed reactor vessel head penetration performance demonstration system for ultrasonic test to apply for pressurized light-water reactor power plants in accordance with 10CFR50.55a since September 2011. RVHP configuration surveying and analysis, code requirement analysis, and performance demonstration specimen design were performed up to this day. Fingerprinting of manufactured specimen, development of test data management program, development of operation procedure, input of flawed data, and development of final report will be performed for the next step. This paper describes the development status of the performance demonstration system for reactor vessel head penetration ultrasonic examination in Korea.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.622-627
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• 2001

The object of this paper is to compare stress intensity factor that be calculated by Raju-Newman's equation, finite element method, and Vessel INTegrity analysis inner flaws(VINTIN) program for longitudinal semi-elliptical cracks in cylindrical vessel under inner pressure. For this, three-dimensional finite-element analyses were performed to obtain the stress intensity factors for various surface cracks with t/R = 0.1. The finite element meshes were designed for various crack shapes with t/R of 0.1. The crack depth to thickness ratio, a/t, was set to 0.2 and 0.5 matching Raju-Newman's equation. The crack depth to length ratio, a/c, was set to 0.2 and 0.4 in the same way and 0.33 was added to extend the range of crack configuration. Finite Element Analyses(FEA) were performed using the commercial FEA program ABAQUS. The results showed that the Raiu-Newman solutions were about 4-10% lower than FEA's using symmetric model of one-eighth of a vessel and close to those of FEA using symmetric model or one-forth or a vessel. Ana VINTIN solutions were nearly equal to those or Raju-Newman.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.58-63
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• 2019

This paper presents preliminary analysis and results on IASCC sensitivity of a core shroud in the reactor pressure vessel. First, neutron irradiation flux distribution of the reactor internals was calculated by using the Monte Carlo simulation code, MCNP6.1 and the nuclear data library, ENDF/B-VII.1. Second, based on the neutron irradiation flux distribution, temperature and stress distributions of the core shroud during normal operation were determined by performing finite element analysis using the commercial finite element analysis program, ABAQUS, considering irradiation aging-related degradation mechanisms. Last, IASCC sensitivity of the core shroud was assessed by using the IASCC sensitivity definition of EPRI MRP-211 and the finite element analysis results. As a result of the preliminary analysis, it was found that the point at which the maximum IASCC sensitivity is derived varies over operating time, initially moving from the shroud plate located in the center of the core to the top shroud plate-ring connection brace over operating time. In addition, it was concluded that IASCC will not occur on the core shroud even after 60 years of operation (40EFPYs) because the maximum IASCC sensitivity is less than 0.5.

• Asia-pacific Journal of Multimedia Services Convergent with Art, Humanities, and Sociology
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• v.7 no.2
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• pp.427-435
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• 2017

Most of the domestic city buses are equipped with the pressure vessels subjected to internal pressure applied by compressed natural gas. Pressure vessels subjected to internal pressure are used in various forms and purposes. Fuel is explosive and has flammable high pressure. The damage of the pressure vessel causes many property damage and loss of life. Safe design for pressure vessel is always necessary. Due to these reasons, many studies using finite element analysis have been conducted. In this paper, the stresses of cylindrical vessel and spherical dome were analyzed using ANSYS, a finite element analysis software. In order to verify the validity of the analysis, a model with a perfectly spherical shape of the dome was designed and observed. Based on the ASME standard in used, stress distribution was also analyzed for models designed with compressed natural gas(CNG). The FEM analysis software agreed with the theory when the dome shape was perfectly spherical. The model designed based on the ASME specification theory, stress concentration occurred in the knuckle part.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.656-660
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• 2002

The fiber reinforced composite material is widely used in the multi-industrial field where the weight reduction of the infrastructure is demanded because of their high specific modulus and specific strength. Pressure vessels using this composite material in comparison with conventional metal vessels can be applied in the field where lightweight and the high pressure is demanded from the defense and aerospace industry to rocket motor case due to the merits which are energy curtailment by the weight reduction and decrease of explosive damage precede to the sudden explosion which is generated by the pressure leakage condition. In this paper, for nonlinear finite element analysis of E-glass/epoxy filament winding composite pressure vessel receiving an internal pressure, the standard interpretation model is developed by using the ANSYS 5.7.1, the general commercial program, which is verified as the accuracy and useful characteristic of the solution based on Auto LISP and ANSYS APDL. Both the preprocessor for doing exclusive analysis of filament winding composite pressure vessel and postprocessor that simplifies result of analysis have been developed to help the design engineers.

• Journal of Ocean Engineering and Technology
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• v.19 no.1 s.62
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• pp.95-99
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• 2005

This paper presents the optimum design of cylindrical shell under external pressure loading. Two kinds of material, Al7075-T6, Ti-6Al-4V, are considered. For each material, the design variable is a thickness of the unstiffened parallel middle body shell, and the state variable, constraint, is hoop stress and the object .function is total weight of the cylindrical shell. Optimization is performed by conventional FE Program, ANSYS. In addition, buckling analysis is performed for the middle body of the cylindrical shell. Finally, we calculates the payload of the cylindrical shell to keep neutral buoyancy with optimized thickness in deep-sea applications.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.22 no.5
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.

• Journal of the Korean Society for Precision Engineering
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• v.15 no.11
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• pp.251-257
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• 1998

동적 파괴인성치 측정시스템과 동적 2차원 유한요소해석 프로그램을 개발하여 원자로압력용기에 사용하는 강(SA508 cl.3, SA516 gr.70)의 동적 파괴인성치와 동적 균열정지인성치를 평가하고 이에 대한 유용성을 확인하였으며, 이 시스템 을 이용하여 재료의 동적 파괴특성을 규명하였다. SA508 cl.3와 SA516 gr.70의 동적 균열전파속도(a)에 대응하는 동적 응력확대계수 (K(a))에 대한 실험식을 얻었으며, 동적 응력확대계수와 동적 균열전파속도와의 관계는 전형적인 "$\Gamma$" 형으로 나타남을 확인하였다.