• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.53-60
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• 2015

This study was investigated to develop process technology of laser cladding with austenite stainless steel for Weld Inlay repair of dissimilar metal weld in reactor vessel in/outlet nozzles. Weld Inlay experiments were performed by laser cladding repair system consisting of common manipulator, laser apparatus and welding process scheduler, etc. Single pass welding experiments were conducted in order to obtain the optimum welding process parameters for filler wires of ER309L and Alloy 52M before multi-layer laser cladding. Based on the above obtained results, multi-layer laser cladding experiments were carried out, and welding qualities for weld specimens were estimated by PT, OM, SEM and EDS analysis. Consequently, it was revealed that multi-layer laser cladding on austenite stainless steel using filler wires of ER309L and Alloy 52M could be possible to meet ASME Code standard without any weld defect.

• Journal of the Korean Society for Precision Engineering
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• v.26 no.12
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• pp.110-116
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• 2009

Thick-walled cylinders, such as a cannon or nuclear reactor, are autofrettaged to induce advantageous residual stresses into pressure vessels and to increase operating pressure and the fatigue lifetimes. As the autofrettage level increases, the magnitude of compressive residual stress at the bore also increases. However, the Bauschinger effect reduces the compressive residual stresses as a result of prior tensile plastic strain, and decreases the beneficial autofrettage effect. The purpose of the present paper is to predict the accurate residual stress of SNCM8 high strength steel using the Kendall model which was adopted by ASME Code. The uniaxial Bauschinger effect test was performed to decide BEF, then this constant was used in calculation. There were some differences between theoretical solution and modified solution.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.569-580
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• 2014

Seismic isolation is a viable strategy for protecting safety-related nuclear structures from the effects of moderate to severe earthquake shaking. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities. The performance expectations identified in the NUREG and ASCE 4 for seismic isolation systems, and superstructures and substructures are described in the paper. Robust numerical models capable of capturing isolator behaviors under extreme loadings, which have been verified and validated following ASME protocols, and implemented in the open source code OpenSees, are introduced.

• International Journal of Naval Architecture and Ocean Engineering
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• v.9 no.5
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• pp.509-524
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• 2017

Experiments and a numerical simulation were conducted to investigate the deformation and impact behavior of a corroded pipe, as corrosion, fatigue, and collision phenomena frequently occur in subsea pipelines. This study focuses on the deformation of the corrosion region and the variation of the geometry of the pipe under impact loading. The experiments for the impact behavior of the corroded pipe were performed using an impact test apparatus to validate the results of the simulation. In addition, during the simulation, material tests were performed, and the results were applied to the simulation. The ABAQUS explicit finite element analysis program was used to perform numerical simulations for the parametric study, as well as experiment scenarios, to investigate the effects of defects under impact loading. In addition, the modified ASME B31.8 code formula was proposed to define the damage range for the dented pipe.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.24 no.11
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• pp.777-783
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• 2012

The hardness and strength test was performed to make the manufacturing process of SA213 P92 boiler pipe steel. And the microstructure change was studied to find out the cause of room temperature property of P92 steel, ie, low hardness and strength property. The room temperature property of P92 steel depends on the improper normalizing and cooling rate. Especially, Ferrite was formed and the steel had low hardness when the temperature was decreased slowly under the cooling rate $1^{\circ}C$/min after normalizing at the temperature around $A_{c1}$ to $A_{c3}$. The critical heat treatment temperature and cooling rate was over $900^{\circ}C$ and over $10^{\circ}C$/min to satisfy the minimum yield and tensile stress which was laid down by ASME Code.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2001.04a
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• pp.481-488
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• 2001

The HANARO in-chimney bracket was proposed as a structure which supports the guide tubes of irradiation facilities at the irradiation sites of CT, IR and OR4/5 in HANARO core for the reduction of flow-induced vibration and seismic response of the irradiation facilities. For the evaluation of the structural integrity of the in-chimney bracket, its finite element model is developed. The seismic response analysis was performed for the in-chimney bracket and related reactor structures, under the response spectrum of OBE and SSE. The analysis results show that stress values of the in-chimney bracket and reactor structures for the seismic loads are within the ASME code limits. It is also confirmed that its fatigue usage factor is much less than 1.0. For the verification of the implementation effects of the in-chimney bracket, the vibration level of the guide tube of the instrumented fuel assembly, which is subjected to fluid-induced vibration, was measured and analyzed. The vibration analysis results demonstrate that the vibration level of the instrumented fuel assembly has been remarkably reduced after installing the in-chimney bracket. Therefore, when the in-chimney bracket is installed at the reactor chimney, any damage on the structural integrity is not expected.

• Journal of Welding and Joining
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• v.29 no.2
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• pp.64-71
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• 2011

In this paper, crack propagation analyses in the inner diameter (ID) repair weld of the dissimilar metal weldment of a nozzle were performed using a finite element alternating method (FEAM). To calculate the theoretical solution for the crack tip stress intensity factor, a weak type singular integral equation consisted of crack surface traction and dislocation density function was constructed and solved in conjunction with the FEAM. A two-dimensional axisymmetric finite element nozzle model was prepared and ID repair welding was simulated. An initial crack, 10% depth of weld thickness, was assumed and crack propagation trajectory from the initial crack to the 75% depth of thickness was calculated using the FEAM. Crack growth versus time curve was also calculated and compared with the curves obtained from ASME code method. With the method constructed in this paper, crack propagation trajectory and crack growth time were calculated automatically and effectively.

• Proceedings of the KSME Conference
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• 2001.06b
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• pp.404-409
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• 2001

Seismic qualification of the Air Cleaning Units for nuclear power plant Ulchin 5&6 has been performed with the guideline of ASME Section III and IEEE 344 code. By using the structural and geometrical similarity analysis, the three models to be analyzed is condensed into a single model and, at the same time, the excitation forces and other operating loads for each model are encompassed with respect to different loading conditions. As the fundamental frequencies of the structure are found to be less than 33Hz, which is the upper frequency limit of the seismic load, response spectrum analysis using ANSYS is performed in order to combine the modal stresses within the frequency limit. In order to confirm the structural and electric stability of the major components, modal analysis theory is adopted to derive the required response spectrum at the component locations. As the all combined stresses obtained from the above procedures are less than allowable stresses and no mechanical or electrical failures are found from the seismic testing, the authors confirm the safety of the nuclear equipments Air Cleaning Units studied in this paper.

• 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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.545-553
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• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.