• Proceedings of the Korean Geotechical Society Conference
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• 2009.09a
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• pp.524-528
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• 2009

An influence factors for soundness evaluation of river levee include resistibility and embankment for piping of ground consisting embankment in case piping, permeability coefficient of ground, height of embankment, the width of crest, material characteristics of embankment and foundation ground, shape of embankment slope, an influence for penetration of rainfall or river water in case slope stability. In this study, it was operated a feasibility investigation of existing design result, stability evaluation for permeability coefficient use and permeability coefficient change of foundation ground to investigate an influence in line with permeability coefficient change for result of river levee penetration analysis. The evaluation results of influence factors, the permeability coefficient used in design and it was evaluated influence in safety factor of piping. After the evaluation of influence factors, the permeability coefficient used in the design appears with the fact that differs in a design report about same soil, Accordingly, the stability investigation of embankment by application of literature data can affect stability evaluation results by change factors like a permeability coefficient, void ratio. It should be certainly used material properties by a test in soundness evaluation of river levee.

• Journal of the Korean Institute of Gas
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• v.18 no.6
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• pp.14-20
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• 2014

In this study, the stress analysis for the piping support design is performed as per the rules of the ASME Code, Section III, subsection NF-Component Support which provides a simplified method of design analysis for piping support. This method makes use of simple equations and conservative allowable stress limits for design and service loadings. For the base plate, code equation is satisfied within the allowable limits. Both anchor bolts and pipe strap are governed by the their interaction equations. The stresses resulting from various loadings and their combinations are within the allowable limits specified in the above mentioned ASME Code. Thus, it was proved that the structural integrity of the pump assembly was satisfactory.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.5
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• pp.381-388
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• 2009

In some portions of nuclear piping systems, stratification phenomena may occur due to the density difference between hot and cold stream. When the temperature difference is large, the stratified flow under diverse operating conditions can produce high thermal stress, which leads to unanticipated piping integrity issues. The objectives of this research are to examine controvertible numerical factors such as model size, grid resolution, turbulent parameters, governing equation, inflow direction and pipe wall. Parametric three-dimensional computational fluid dynamics analyses were carried out to quantify effects of these parameters on the accuracy of temperature profiles in a typical nuclear piping with complex geometries. Then, as a key finding, it was recommended to use optimized mesh of real piping with the conjugated heat transfer condition for accurate thermal stratification analyses.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.237-248
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• 2013

Thermal stratification has continuously caused several piping failures in nuclear power plants since the early 1980s. However, this critical thermal effect was not considered when the old nuclear power plants were designed. Therefore, it is urgent to evaluate this unexpected thermal effect on the structural integrity of piping systems. In this paper, the thermal effects of stratified flow in two different safety injection piping systems were investigated by using a coupled CFD-FE method. Since stratified flow is generally generated by turbulent penetration and/or valve leakage, thermal stress analyses as well as CFD analyses were carried out considering these two primary causes. Numerical results show that the most critical factor governing thermal stratification is valve leakage and that temperature distribution significantly changes according to the leakage path. In particular, in-leakage has a high possibility of causing considerable structural problems in RCS piping.

• Tunnel and Underground Space
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• v.32 no.1
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• pp.30-58
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• 2022

The deep geological repository for high-level radioactive waste disposal is a multi barrier system comprised of engineered barriers and a natural barrier. The long-term integrity of the deep geological repository is affected by the coupled interactions between the individual barrier components. Erosion and piping phenomena in the compacted bentonite buffer due to buffer-rock interactions results in the removal of bentonite particles via groundwater flow and can negatively impact the integrity and performance of the buffer. Rapid groundwater inflow at the early stages of disposal can lead to piping in the bentonite buffer due to the buildup of pore water pressure. The physiochemical processes between the bentonite buffer and groundwater lead to bentonite swelling and gelation, resulting in bentonite erosion from the buffer surface. Hence, the evaluation of erosion and piping occurrence and its effects on the integrity of the bentonite buffer is crucial in determining the long-term integrity of the deep geological repository. Previous studies on bentonite erosion and piping failed to consider the complex coupled thermo-hydro-mechanical-chemical behavior of bentonite-groundwater interactions and lacked a comprehensive model that can consider the complex phenomena observed from the experimental tests. In this technical note, previous studies on the mechanisms, lab-scale experiments and numerical modeling of bentonite buffer erosion and piping are introduced, and the future expected challenges in the investigation of bentonite buffer erosion and piping are summarized.

• Journal of the Korean Society for Precision Engineering
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• v.28 no.2
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• pp.233-238
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• 2011

In piping design of nuclear power plant facilities, the load stress according to self-weight is important for design values in test run(shutdown and starting). But sometimes it needs more studies, such as seismic analysis of an earthquake of power plant area and fatigue life and stress of thermal expansion and anchor displacement in operating run. In this paper, seismic evaluations were performed to nuclear piping system of Shin-Kori NO. 3&4 being built in Pusan lately. Results of seismic analysis are evaluated on basis of KEPIC MN code. The structural integrity on RCB piping system was proved.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.732-740
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• 2022

A piping system stress analysis need to be re-performed for structural integrity assessment after reinforcement of a pipe with significant wall thinning. For efficient stress analysis, a one-dimensional beam element for the wall-thinned pipe with reinforcement needs to be developed. To develop the beam element, this work presents analytical equations for elastic stiffness of the wall-thinned pipe with reinforcement are analytically derived for axial tension, bending and torsion. Comparison with finite element (FE) analysis results using detailed three-dimensional solid models for wall-thinned pipe with reinforcement shows good agreement. Implementation of the proposed solutions into commercial FE programs is explained.

• Journal of the Korean Geosynthetics Society
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• v.14 no.4
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• pp.1-9
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• 2015

According to national regulations and its commentary, such as Rivers Design Criteria & Commentary (KWRA, 2009), Foundation Structure Guideline and its Commentary(MLTM, 2014 and KGS, 2009), the integrity evaluation of river levee includes slope stability evaluation of both riverside/protected low-land and piping stability evaluation with respect to foundation and levee body along with water level conditions. In this case the design hydro-graph can be the most important input factor for the integrity evaluation, however it is fact that the national regulations do not provide any proper determination methods regarding hydro-graph. The authors thus executed an integrity evaluation of sluice gate in levee by changing each hydro-graph factor, including rising ordinary water level, lasting flood water level, falling water level, and flood frequency, in order to suggest a determination method of reasonable hydro-graph. As a result, the authors suggested that at least over 57 hours of rising ordinary water level and over 53 hours of lasting flood water level should be considered for the design hydro-graph of sluice gate in levee at Mun-san-jae.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.636-643
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• 2001

Interference between the crossover leg of the reactor coolant system (RCS) and the pipe whip restraints (PWR) has brought a degradation issue of the integrity of the Reactor Coolant System in Westinghouse type nuclear power plants (NPPs) of Korea. According to the gap Inspect ion carried out during planned overhaul (Year 2000), interference between the crossover leg and the PWR was found in each RCS loop. This plant has had the high vibration problem on the RC pump 'B'. The reason for the high vibration in the RC pump 'B' had been massively surveyed and it was found that the crossover leg of RCS contacted with the PWR in hot condition. Since the contact between the crossover leg and the PWR changes the dynamic characteristics of the piping system for the RCS, this is considered as one reason for the high vibration. And a possibility of overstress on the crossover leg due to the contact with the PWR should be evaluated. Through performing RCS integrity analyses, subsequent actions were initiated to increase the gap between those parts. As the results of the appropriate separation between two parts, it was reported that there was no unusual noise or vibration during plant heat-up. In this paper, the evaluations for the gap between the crossover leg and the PWR and the structural integrity due to loop binding is described.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.44-50
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• 2011

Turbulent mixing of hot and cold coolants is one of the possible causes of high cycle thermal fatigue in piping systems of nuclear power plants. A typical situation for such mixing appears in turbulent flow through a T-junction. Since the high cycle thermal fatigue caused by thermal striping was not considered in the piping fatigue design in several nuclear power plants, it is very important to evaluate the effect of thermal striping on the integrity of mixing tees. In the present work, before conducting detailed evaluation, three thermal striping evaluation methodology suggested by EPRI, JSME and NESC are analyzed. Then, a by-pass pipe connected to the shutdown cooling system heat exchanger is investigated by using these evaluation methodology. Consequently, the resulting thermal stresses and the fatigue life of the mixing tee are reviewed and compared to each other. Futhermore, the limitation of each methodology are also presented in this paper.