• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.839-841
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• 2014

The Nuclear Steam Supply System(NSSS) is designed to provide an integrated approach that includes areas of monitoring relevant to the integrity of the NSSS. LPMS is designed to function as an alarm system by providing sensor channel alarms for the associated subsystems. LPMS is equipped to provide analysis tools for new alarm events, historical events and for historical periodically stored channel data (e.g. waveforms) for most channels. This paper is intended to introduce the diagnosis principle and abnormal symptom of loose parts monitoring system as a monitoring tool in Nuclear Steam Supply System. And also, we are going to introduce signal analysis program in order to perform the actual diagnosis in power plants.

• Proceedings of the Acoustical Society of Korea Conference
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• autumn
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• pp.239-242
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• 2000

In the present study, the influence of car body structures to the noise and vibration characteristics has been sought. The numerical modal analysis for the body-in-white is employed to predict the vibratory response of structure, and then followed by the experimental modal testing to confirm the validity of the model. Using the results of numerical simulations with the designated modal parameters, the optimal structural configuration has been deduced. Special interests have been paid to the sensitivity of sound quality to the structural integrity. Since the structural integrity has a close relationship to the structure-born noise, the substantially low frequency range, which is far below the frequency range almost barely sensible by human auditory organ but still quite influential to overall impression, is especially examined. The subjective assessment agrees with the objective evaluation by means of traditional sound measures as well as psychoacoustic metrics.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2092-2099
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• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.555-561
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• 1994

The program of performing a fast and automatic analysis of gamma ray spectra obtained by a Multi-Channel Analyzer (MCA) is developed for the surveillance of the nuclear fuel integrity. The integrity of the nuclear fuel is confirmed by the measurement of the radiation level of the reactor coolant through the real time monitoring and the periodic sampling analysis. In Yonggwang nuclear power plane 3 and 4, the Process Radiation Monitoring System (PRMS), which is a real time monitoring system, provides a measure of the fuel integrity. Currently, its spectrometer channel can identify only one radionuclide at a time since the signal processing unit of the spectrometer channel is a Single Channel Analyzer (SCA). To improve the PRMS, it is necessary to substitute the MCA for the SCA The program is operated in a real time mode and an on-demand mode, and automatically performed for all procedures. The test results by using the National Bureau of Standards (NBS) mixed standard source are in good agreement with those from Canberra System 100 which is a commercial MCA Consequently, the developed program seems to be employed for automatic monitoring of gamma rays in nuclear power plants.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1273-1279
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• 2010

Pressure tubes at the CANada Deuterium Uranium (CANDU) nuclear power plants are periodically inspected in accordance with the CSA N285.4 code. If flaws that do not satisfy the criteria given in CSA N285.4 are detected, the code permits a fitness-for-service assessment to determine the acceptability of the flawed pressure tubes. In this paper, the Wolsong In-service Evaluation System (WIES) is introduced; this system has been developed for the assessment of the flawed pressure tubes and is based on CSA N285.8. Since the system evaluates the integrity of flawed pressure tubes exactly and promptly during an in-service inspection, it will help in operating the Wolsong nuclear power plants without prolonging the outage period.

• 한국가스학회:학술대회논문집
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• 2001.10a
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• pp.108-113
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.493.2-493
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.322-322
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• 2001

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.636-642
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.267.2-267
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• 1999