• Proceedings of the KSME Conference
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• 2004.04a
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• pp.96-101
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• 2004

Cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing and valve bodies in light water reactors. These components are subject to thermal embrittlement at the reactor operating temperature. The objective of this study is to summarize the method of estimating ferrite content, Charpy impact energy and J-R curve and to evaluate the thermal embrittlement of the cast austenitic stainless steel piping used in the domestic nuclear power plants. The result of evaluation, two domestic nuclear power plants used CF-8M and CF-8A material has adequate fracture toughness after saturation.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.981-988
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• 2018

An experiment was performed for the OECD/NEA ROSA-2 Project with the large-scale test facility (LSTF), which simulated a steam generator tube rupture (SGTR) accident due to a double-ended guillotine break of one of steam generator (SG) U-tubes with operator recovery actions in a pressurized water reactor. The relief valve of broken SG opened three times after the start of intact SG secondary-side depressurization as the recovery action. Multi-dimensional phenomena specific to the SGTR accident appeared such as significant thermal stratification in a cold leg in broken loop especially during the operation of high-pressure injection (HPI) system. The RELAP5/MOD3.3 code overpredicted the broken SG secondary-side pressure after the start of the intact SG secondary-side depressurization, and failed to calculate the cold leg fluid temperature in broken loop. The combination of the number of the ruptured SG tubes and the HPI system operation difference was found to significantly affect the primary and SG secondary-side pressures through sensitivity analyses with the RELAP5 code.

• Journal of Power System Engineering
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• v.17 no.4
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• pp.17-22
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• 2013

Diesel generator of nuclear power plant has a role for supply of emergency electric power to protect reactor core system in event of loss of off-site power supply. Therefore diesel generator should be tested periodically to verify the function that can supply specified frequency and voltage at design power level within limited time. For this purpose, appropriate maintenances in case that abnormal conditions were found are required in allowed time. In this paper, results of development of engine condition diagnosis technology and study on power plant of its technology for diesel generator are described.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1733-1739
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• 2004

The objective this study is to estimate the feasibility of acoustic emission method for the internal leak from the valves in nuclear power plants. From the experimental results, it was suggested that the acoustic emission method for monitoring of leak was feasible. When the background levels are higher than the acoustic signals from leak, we can detect the leak analyzing the spectrum of the remainders which take the background noise from the acoustic signals.

• Journal of Drive and Control
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• v.9 no.2
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• pp.8-14
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• 2012

Electro-hydraulic spring return actuator(ESRA) is utilized for air conditioning facilities in a nuclear power plant. It features self-contained, hydraulic power that is integrally coupled to a single acting hydraulic cylinder and provides efficient and precise linear control of valves as well as return of the actuator to the de-energized position upon loss of power. In this paper, the algebraic equations of ESRA at steady-state have been developed for the analysis of static characteristics that includes control pressure and valve displacement of pressure reducing valve, flow force on flapper as well as its displacement over the entire operating range. Also, the effect of external load on piston deviation is investigated in terms of linear system analysis. The results of static characteristics show the unique feature of force balance mechanism and can be applied to the stable self-controlled mechanical system design of ESAR.

• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.481-487
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• 2004

Air-operated valve is one of principal valves that are using to control fluid flow in nuclear power plants. A period diagnosis lot safety of power plants is necessary. But there are many difficulties such as economic loss caused by income of high cost devices and a matter hard to deal with users. In this study, we developed the diagnostic system that users of power plants are easy to handle. The diagnostic system is composed of database module, diagnosis test module and evaluation module.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.694-699
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• 2004

The purpose of this study is to estimate the availability of acoustic emission method to the internal leak of the valves at nuclear power plants. The acoustic emission method was applied to the valves at the site, and the background noise was measured for the abnormal plant condition. From the comparison of the background noise data with the experimental results as to relation between leak flow and acoustic signal, the minimum leak flow rates that can be detected by acoustic signal was suggested. When the background levels are higher than the acoustic signal, the method described below was considered that the analysis the remainder among the background noise frequency spectrum and the acoustic signal spectrum become a very useful leak detection method. A few experimental examples of the spectrum analysis that varied the background noise characteristic were given.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.18 no.7
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• pp.603-611
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• 2006

A variety of schemes were sought for a mitigation of thermal stratification phenomenon in the branch piping of domestic nuclear power plant. Several mechanisms of thermal stratification occurrence were introduced in this paper. A number of factors were selected to find out the schemes for thermal stratification mitigation and the computational analysis were performed. The length of vertical branch piping, the diameter, the radius of curvature of the elbow, the direction of connection between main piping and branch piping, the slope of branch piping, the leakage flow rate, the installation of additional valve, the change of the 1st valve position and another branch piping connected with branch piping were mentioned as factors in this paper.

• Journal of the Korean Vacuum Society
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• v.15 no.6
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• pp.594-604
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• 2006

Gas fueling systems operated in the mode of a fixed valve opening at a constant line pressure, and the mode of a constant inlet flow are simulated to establish the relationships between the gas flow pattern and the tube dimensions under various system conditions.

• Journal of Mechanical Science and Technology
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• v.19 no.12
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• pp.2245-2252
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• 2005

Thermal stratification in the branch piping of power plants can be generated by turbulent penetration or by valve leakage. In this study, a numerical analysis was performed to estimate the thermal stratification phenomenon by in-leakage in the SIS branch piping of nuclear power plant. Leakage rate, leakage area and leakage location were selected as evaluation factors to investigate the thermal stratification effect. As a result of the thermal stratification effect according to leakage rate, the maximum temperature difference between top and bottom of the horizontal piping was evaluated to be about 185K when the valve leakage rate was about 10 times as much as the allowed leakage rate. For leakage rate more than 10 times the allowed leakage rate, the temperature difference was rapidly decreased due to the increased mixing effect. In the result according to leakage area, the magnitude of temperature difference was shown in order of $3\%,\;1\%\;and\;5\%$ leakage area of the total disk area. In the thermal stratification effect, according to the leakage location, temperature difference when leakage occurred in the lower disk was considerably higher than that of when leakage occurred in the upper disk.