• Journal of Korean Society on Water Environment
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• v.24 no.6
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• pp.793-800
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• 2008

Recently domestic drinking water industry has recognized membrane-based technology as a promising alternative for water treatment. To ensure successful application of membrane processes, the integrity of membrane systems should be maintained. According to US EPA guidance, the pressure decay test based on the bubble point theory is recommended to detect any membrane defection of which size is close to the smallest diameter of Cryptosporidium oocysts, $3{\mu}m$. Proper implementation of the pressure decay test is greatly affected by initial test pressure, and the interpretation of the test results is associated with upper control limit. This study is conducted to investigate various factors affecting determination of initial test prtessure and upper control limit, including membrane-based parameters such as pore shape correction factor, surface tension and contact angle, and system-based parameters, such as volumetric concentration factor and total volume of system. In this paper, three different hollow fibers were used to perform the pressure decay test. With identical initial test pressure applied, their pressure decay tendency were different from each other. This finding can be explained by the micro-structure disparity of those membranes which is verified by FESEM images of those membranes. More specifically, FESEM images revealed that three hollow fibers have asymmetry, deep finger, shallow finger pore shape, respectively. In addition, sensitivity analysis was conducted on five parameters mentioned above to elucidate their relation to determination of initial test pressure and upper control limit. In case of initial pressure calculation, the pore shape correction factor has the highest value of sensitivity. For upper control limit determination, system factors have greater impact compared to membrane-based parameters.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.299-304
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• 2008

This paper presents plastic limit loads of piping branch junctions with local wall thinning under combined pressure and in-plane bending, based on systematic three-dimensional finite element limit analyses using elastic-perfectly plastic materials. An ideal branch junction without weld or reinforcement around the intersection is considered with two locations of wall thinning; one in the run pipe, and the other in the branch pipe. Based on FE results, effects of thinning geometries on plastic limit moments are quantified and simple approximations of plastic limit loads are proposed.

• Computational Structural Engineering
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• v.10 no.4
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• pp.155-164
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• 1997

A reactor pressure vessel, which contains fuel assemblies and reactor vessel internals, has the thermal stress resulting from the cool-down and heat-up of the vessel wall in combination with the pressure stress from system pressure resulting in large stresses. The combination of the pressure stress and thermal stress along with a decrease in fracture toughness may cause through-wall propagation of a relatively small crack. Therefore, it is necessary to define the relations between operating pressure and temperature during cool-down and heat-up. In this study, theory of fracture mechanics for a pressure/temperature limit curve is investigated and a numerical procedure for generating it is developed. Plant-specific limit curves for the Kori unit 1 plant, the oldest nuclear power plant in Korea, have been obtained for several cooling and heating rates and their results are discussed.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.16 no.2
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• pp.153-163
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• 2003

Performed here is a comparative assessment study for the generation of the pressure-temperature limit curve of the reactor vessel. A round robin problem is proposed using the data available in Korea and all organizations interested in the generation of the pressure-temperature limit curve are invited. The problems consisting of 12 cases for cool-down are solved and their results are compared to generate a reference solution for the reference problem, which will be useful in the evaluation of the generation of the pressure-temperature limit curve in the future.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.505-513
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• 2002

During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.1
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• pp.26-33
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• 2003

Based on detailed FE limit analyses, the present paper provides tractable approximations fer plastic limit pressure solutions fur axially through-wall-cracked pipe; axially (inner) surface-cracked pipe; circumferentially through-wall-cracked pipe; and circumferentially (inner) surface-cracked pipe. In particular, for surface crack problems, the effect of the crack shape, the semi-elliptical shape or the rectangular shape, on the limit pressure is quantified. Comparisons with existing analytical and empirical solutions show a large discrepancy in circumferential short through-wall cracks and in surface cracks (both axial and circumferential). Being based on detailed 3-D FE limit analysis, the present solutions are believed to be the most accurate, and thus to be valuable information not only for plastic collapse analysis of pressurised piping but also for estimating non-linear fracture mechanics parameters based on the reference stress approach.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1824-1829
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• 2007

In the present paper, limit pressures for axial surface cracked pipe are proposed, and a reference stress based J estimation method is also provided based on the proposed limit pressure solutions. Employed cracks are assumed to be constant-depth, internal surface cracks, and wide ranges of variables are considered.

• Journal of Korean Tunnelling and Underground Space Association
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• v.18 no.1
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• pp.53-60
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• 2016

This study analyzed a monitoring data, based on the initial limit values of monitoring in subway, of earth pressure and pore water pressure. The data is obtained from 8 sections of the Seoul metropolitan subway line No 6, 7 and 9 in about 5 years. Also, a research is performed to set up the limit values of management monitoring, which will be applied to management monitoring in tunnel, through comparing the limit values of overseas management monitoring data and that of domestic management monitoring data. And the result obtained from comparison show that the safety phase is 60% of allowable pressure, the attention phase is 80% of allowable pressure and the precision analysis phase is 100% of allowable pressure. Also, we presented a method of management monitoring by the absolute value which can be easily applied easily in practical affairs.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.