• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.84-91
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• 2017

This paper is proposed to select the optimal finite element type in finite element analysis. Based on the NUREG reports, static analyses were performed using a commercial analysis program, $ABAQUS^{TM}$. In this study, we used a nonlinear kinematic hardening model proposed by Chaboche. The analysis result of solid elements by inputting the same material constants was different from the results of the NUREG report. This is resulted from the difference between shell element and solid element. Therefore, the material constants that have similar result to the experimental result were determined and compared according to element type. In case of using solid element for efficient finite element analysis, we confirmed that the use of C3D8I element type(incompatible mode 8-node linear brick element) leads the accurate result while reducing the analysis time.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.40-47
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• 2017

In this paper, we investigate the feasibility of a strategy of failure detection and identification. The point of proposed strategy includes a pattern extraction approach for failure identification using Auto-Associative Kernel Regression (AAKR). We consider a simulation data concerning 605 signals of a Generic Pressurized Water Reactor(GPWR). In the application, the reconstructions are provided by a set of AAKR models, whose input signals have been selected by Correlation Analysis(CA) for the identification of the groups. The failure pattern is extracted by analyzing the residuals of observations and reconstructions. We present the possibility of extraction of patterns for six failure.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.3
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• pp.1-6
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• 2012

The containment Integrated Leakage Rate Testing(ILRT) of nuclear power plants in Korea is performed in accordance with NSSC(Nuclear Safety and Security Commission) code 2012-16 and ANSI/ANS 56.8-1994. Nuclear power plants in Korea and the United States are to apply same test criteria, ANSI/ANS 56.8-1994, except type A testing time. NPPs in Korea apply 24 hours according to NSSC code 2012-16, but NPPs in United States apply 8 hours according to 10CFR50 App. J for type A test. So, there are many difficulties in order to perform ILRT in Korea. In this study, I review the impact on the ILRT results and the effect of ILRT due to type A testing time. The future, we will continue study to enhance the test reliability and improve these problems.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.8-17
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• 2019

Perforated plates are used for the steam generator tube-sheet and the Reactor Vessel Closure Head in the Nuclear Power Plant. The ASME code, Section III Appendix A-8000, addresses the analysis of perforated plates, however, this analysis is only limited to the flat plate with a triangular perforation pattern. Based on the concept of the effective elastic constants, simulation of flat and spherical perforated plates and their equivalent solid plates were carried out using Finite Element Analysis (FEA). The isotropic material properties of the perforated plate were replaced with anisotropic material properties of the equivalent solid plate and subjected to the same loading conditions. The generated curves of effective elastic constants vs ligament efficiency for the flat perforated plate were in agreement with the design curve provided by ASME code. With this result, a plate with spherical curvature having perforations can be conveniently analyzed with equivalent elastic modulus and equivalent Poisson's ratio.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.1-7
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• 2019

In this paper, the regulations and technical standards related to qualification of safety grade equipment in nuclear power plants are critically reviewed with the qualification procedure in terms of structures, systems, and equipment in nuclear power plants. These facilities should be designed and constructed to protect from natural conditions or disasters and to perform their safety functions even in case of postulated accidents. Equipment Qualification is to demonstrate that the safety related equipment is designed and constructed to perform their safety functions under normal and accident conditions. It is classified into environmental qualification and seismic qualification.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.88-95
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• 2018

In this study, to enhance the seismic performance of nuclear power plants (NPP), a small laminated rubber bearing (LRB) is chosen as a seismic design option of the vulnerable equipment. Prior to the application of equipment base isolation, it is necessary to review the feasibility that the technique contributes enough to the seismic performance of NPP by analysis. At first, some preliminary design of small LRBs for equipment is carried out. Design parameters such as horizontal and vertical stiffnesses, design natural frequencies are checked by calculation and analysis for the four design options considering various upper weights. Performance test of small LRB is to be carried out to verify static performance using the results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.11-15
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• 2018

A development plan for seismic capacity verification procedures of nuclear components based on the elastic-plastic strain (EPS) features is explained in this paper. The EPS methodology is more realistic to assess seismic responses of components to extreme seismic events beyond the safe shutdown earthquake (SSE) than current practices with the criteria of stress limits. The EPS based approach to analyze the seismic capacity of components can reduce over-conservatism in the current stress-based criteria and can incorporate the seismic responses of components deformed in plastic behavior by the motion of extreme earthquake.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.16-23
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• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.10-17
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• 2020

Securing the integrity of steam generator tubes is an essential requirement for safe operation of nuclear power plants. Therefore, tubes that do not satisfy integrity requirements are no longer usable and must be repaired according to the related requirements. In general, the repair criterion is that the damage depth is more than 40% of the tube wall thickness. However, the plugging limit can be changed and be applied, provided a technical proof is given that integrity can be secured against specific degradation at a specific plants and that approval can be obtained from a regulatory agency. A typical example is alternative repair criteria for defects within the tube sheet or tube support plates. In this paper, a background of establishing the plugging limit for steam generator tubes and changes in maintenance criteria are reviewed as examples.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.47-57
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• 2020

A new elastic-plastic-creep constraint parameter is proposed to quantify the effect of constraint on creep crack initiation times. It represents the difference between the transient elastic-plastic-creep crack-tip opening stress and the Riedel-Rice opening stress field in plane strain, which can be determined analytically. Application of the proposed parameter to a large set of creep crack growth test data using C(T) and SEN(B) specimens of Type 316H stainless steel at 550℃ shows that creep crack initiation times can be more accurately characterized by the C⁎-integral together with the proposed parameter.