• Nuclear Engineering and Technology
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• v.36 no.4
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• pp.364-373
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• 2004

This study investigates the fluidelastic instability characteristics of helical steam generator type tubes used in operating nuclear power plants. To obtain a natural frequency, corresponding mode shape, and participation factor, modal analyses using various conditions are performed for helical type tubes. Investigated are the effects of the number of turns, the number of supports, and the status of the inner fluid on the modal and fluidelastic instability characteristics of the tubes, which are expressed in terms of the natural frequency, the corresponding mode shape, and the stability ratio.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.837-842
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• 2000

In the coiling process of helical steam generator tubes of integral reactor SMART, a considerable amount of spring back, which induces dimensional inaccuracy and difficulty in fabrication, has been arised. In this research, an analytical model was derived to evaluate the amount of the spring back for steam generator tubes. The model was developed on the basis of beam theory and elastic-perfectly plastic material property. This model was extended to consider the effect of plastic hardening and the effect of the tensile force on the spring back phenomena. Parametric studies were performed for various design variables of steam generator tubes in order to minimize the spring back in the design stage. A sensitivity analysis has shown that the low yield strength, the high elastic modulus, the small helix diameter, and the large tube diameter result in a small amount of the spring back. The amount of the spring back can be controlled by the selection of adequate design values in the basic design stage and reduced to an allowable limit by the application of the tensile force to the tube during the coiling process.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2100-2106
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• 2020

This paper newly proposes an efficient analytic springback prediction method to predict the final dimensions of bent cylindrical tubes for a helical tube steam generator in a small modular reactor. Three-dimensional bending procedure is treated as a two-dimensional in-plane bending procedure by integrating the Euler beam theory. To enhance the accuracy of the springback prediction, mathematical representations of flow stress and elastic modulus for unloading are systematically integrated into the analytic prediction model. This technique not only precisely predicts the final dimensions of the bent helical tube after a springback, but also effectively predicts the various target radii. Numerical validations were performed for five different radii of helical tube bending by comparing the final radius after a springback.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.22 no.3_1spc
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• pp.573-579
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• 2013

Once-through helically coiled steam generator tubes subjected to external pressure are of interest because of their application to advanced small- and medium-sized integral reactors, in which a primary coolant with a relatively higher pressure flows outside the tubes, while secondary water with a relatively lower pressure flows inside the tubes. Another notable point is that the values of the mean radius to thickness ratio of these steam generator tubes are very small, which means that a thick-walled cylinder is employed for these steam generator tubes. In the present paper, the maximum allowable pressure of helically coiled and thick-walled steam generator tubes with through-wall cracks under external pressure is investigated based on a detailed nonlinear three-dimensional finite element analysis. In terms of the crack orientation, either circumferential or axial through-wall cracks are considered. In particular, in order to quantify the effect of the crack location on the maximum external pressure, these cracks are assumed to be located in the intrados, extrados, and flank of helically coiled cylinders. Moreover, an evaluation is also made of how the maximum external pressure is affected by the ovality, which might be inherently induced during the tube coiling process used to fabricate the helically coiled steam generator tubes.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.14 no.4
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• pp.327-335
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• 2004

This study investigates the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the helical type tubes with various conditions. The wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting-wear characteristics of the tube.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.355.2-355
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• 1999

The spnng back phenomena occurring in the coiling process of a steam generator tube induces the dimensional inaccuracy and makes the coiling procedure difficult. In this research, an analytical model was developed to evaluate the amount of the spring back for SMART steam generator tubes. The model was developed on the basis of beam theory and elastic-perfectly plastic material property. This model was extended to consider the effect of plastic hardening and the effect of the tensile force on the spring back phenomena. Parametric studies were performed for various design variables of steam generator tubes in order to minimize the spring back in the design stage. A sensitivity analysis has shown that the low yield strength, the high elastic modulus, the small helix diameter, and the large tube diameter result in a small amount of the spring back. The amount of the spring back can be controlled by the selection of adequate design values in the basic design stage and reduced to an allowable limit by the application of the tensile force to the tube during the coiling process.rocess.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.323-323
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• 2004