• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1772-1777
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• 2007

This paper provides plastic limit loads of pipes with constant-depth, circumferential part-through surface cracks under combined pressure and bending. A key issue is to postulate discontinuous hoop stress distributions in the net-section. Validity of the proposed limit load solutions is checked against the results from three-dimensional (3-D) finite element (FE) limit analyses using elastic-perfectly plastic material behavior.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.5
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• pp.453-460
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• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2188-2197
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• 2022

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.766-775
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• 2015

The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.