• Nuclear Engineering and Technology
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• v.29 no.6
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• pp.488-499
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• 1997

The measurements of the moderator temperature coefficient (MTC) are performed to demonstrate that the calculational model produces results that are consistent with the measurements. Since negative MTC is also a technical specification value that may limit the cycle length, it is important to measure it as accurately as possible. In this report, preferred choice of test method depending on the time in cycle, best power indication and temperature definition in MTC calculation were determined based on the MTC test results taken during initial startup testing and at 2/3 cycle burnup in the Yonggwang nuclear power plant. The results show that the ratio and rodded methods provided good agreement with the predictions during initial startup testing. However, near end-of-cycle the depletion method gives better results, and so is suggested to be used in the MTC measurements at 2/3 cycle burnup. The use of primary Delta T power as a power indicator in the MTC calculations is highly advisable since it responds with good consistent results very quickly to changes unlike secondary calorimetric power. For the appropriate temperature definitions used in the MTC calculations, it is considered that the arithmetic average temperature measured simply by inlet and outlet thermocouples is preferred. Although volumetric average temperature provides better results, the improvement is not sufficient to compensate for the simplicity of calculations by arithmetic average temperature.

• Journal of the Earthquake Engineering Society of Korea
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• v.11 no.3 s.55
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• pp.63-72
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• 2007

The Emergency Diesel Generator (EDG) is a very important equipment for the safety of a Nuclear Power Plant (NPP). In this study, the operating vibration of three kinds of EDG systems was measured. The target EDG systems are Yonggwang 5 unit and Ulchin 2 and 3 units. The Yonggwang 5 and Ulchin 3 unit EDG systems are the same type but the foundation systems are different. One is an anchor bolt type and the other is a spring and viscous-damper type. The purpose of these measurements is for a verification of the vibration isolation effect depending on the foundation system. As a result, It can be said that the spring and viscous damper system of the EDG performed better for the vibration isolation than that of anchor bolt type.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.69-76
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• 2008

Most pressurized water reactors use Venturi flow meters to measure the feedwater flow rate. However, fouling phenomena, which allow corrosion products to accumulate and increase the differential pressure across the Venturi flow meter, can result in an overestimation of the flow rate. In this study, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate. The available data was divided into two groups by fuzzy c means clustering in order to reduce the training time. The data for training the soft-sensing model was selected from each data group with the aid of a subtractive clustering scheme because informative data increases the learning effect. The proposed soft-sensing model was confirmed with the real plant data of Yonggwang Nuclear Power Plant Unit 3. The root mean square error and relative maximum error of the model were quite small. Hence, this model can be used to validate and monitor existing hardware feedwater flow meters.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1992.10a
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• pp.81-86
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• 1992

원자력발전소 1차 계통 내의 건전성 감시를 위한 설비로는 음향누설 감시계 통(Acoustic Leak Monitoring System: ALMS), 금속파편 감시계통(Loose Parts Monitoring System: LPMS) 및 원자로내부구조물 진동감시계통 (Internals Vibration Monitoring System: IVMS)등이 있다. 현재, 국내의 여 러 원전에는 이들중 일부 계통들이 선택적으로 설치되어 운전중이며, 영광 3,4호기에서는 국내 최초로 이들 3개의 계통을 종합한 핵증기공급계통 건전 성감시계통(Nuclear Steam Supply System Integrity Monitoring System: NIMS)을 설계하였다. 특히, 영광 3,4호기 NIMS에서는 각 계통에 의해 감지 된 1차 계통 내의 이상상태를 하나의 분석컴퓨터(Analysis Computer)를 사 용하여 해석하는 종합결함 탐지해석 기법을 도입하였다.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.285-296
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• 1992

The conceptual design of the Plant Operations Regulator (POR) is presented for the pressurized water reactor plants. The POR is a digital supervisory limitation control system The POR assures that the plant does not exceed the operating limits by regulating the plant operations through monitoring the operating margins of the critical parameters. The POR is aimed at increasing the operating flexibility which allows the nuclear plant to meet the grid demand in very efficient manner. It responds to the grid demand without penalizing plant availability by limiting the load demand or by modifying the plant control schemes when the operating limits are approached or violated. The POR design concepts were tested using simulation responses of the 1000 MWe pressurized water reactors, Yonggwang Units 3 & 4. The simulation results illustrate that the POR can be used to improve operating flexibility.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.555-561
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• 1995

This paper examines the applicability of the modified Borresen's coarse-mesh method(MBSN) to the core power distribution monitoring program development for the Yonggwang nuclear power plant unit 3(YGN 3) which uses fixed incore detectors for monitoring core power distribution. In so doing the modified Borresen's coarse-mesh equations are solved with core internal boundary conditions provided by the fixed incore detectors and three-dimensional core power distributions are com puted for the first-cycle core of the YGN 3 PWR. The results are compared with predictions of the COLSS(Core Operating Limit Supervisory System) which is the axial power shape monitoring program of the YGN 3. It is shown that the modified Borresen's method can reproduce the core axial power shape more closely than the COLSS. Because of other advantages in computing speed and predictive capability, n conclude that the proposed MBSN has a promising practical application for core power distribution monitoring program development.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.183-192
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• 1992

An advanced load-follow operation mode, Mode K, is presented for the Korean Standardized Nuclear Power Plants. The Mode K utilizes a heavy worth bank dedicated to axial shape control independent of the existing regulating banks. In Mode K, the heavy bank provides a wide range of axial shape control and a monotonic relationship between its motion and the axial shape change, which makes it easy to automate axial shape control. The achievement of full automatic reactor power control both for the reactivity and power shape would reduce the burden due to load-follow operation on the operator. Also, it can accommodate the frequen-cy control, which requires the plant to respond to the unexpected demand. The Mode K design concepts were tested using simulation responses of Yonggwang Units 3&4, the reference plants for the Korean Standardized Nuclear Power Plants. The results illustrate that the Mode K is an adequate operation mode to provide practical load-follow capabilities for the Korean Standardized Nuclear Power Plants.

• Korean Journal of Fisheries and Aquatic Sciences
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• v.31 no.2
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• pp.186-194
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• 1998

Species composition and biomass of intertidal benthic algae were studied at the coast of Yonggwang Nuclear Power Plant area and its adjacent stations over 4 seasons (October 1995-August 1996), Of 68 species identified, 7 were Cyanophyta, 12 were Chlorophyta, 14 were Phaeophyta and 35 were Rhodophyta. The largest number of algal species (44) was found at Sangnok, the northernmost station of the study area, whereas the smallest number of species (15) was found from Tongho, ca. 13 km north of the power plant site. Number of species showed highest during the spring (44) and minimum was recorded in autumn (28). Biomass per unit area showed maximum in spring ($189.5\;g\;dry\;wt{\cdot}m^{-2}$ in average) and minimum in winter ($107.9\;g\;dry\;wt{\cdot}m^{-2}$ in average). Biomass values exhibited a wide range of variation among the stations, ranging from a low of $22.0\;g\;dry\;wt{\cdot}m^{-2}$ in annual average at Tongho to a high of $295.7g\;dry\;wt{\cdot}m^{-2}$ in average at Sangnok. Dominant species in biomass were Corallina pilulifera, Sargassum thunbergii, Gymnogongrus flabelliformis and Enteromorpha compressa. There have been little variation in the dominant algal species around the power plant site during the past 10 years and also these algae appeared throughout the west coast of Korea with higher frequency.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.336-347
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• 1997

During postulated severe accidents in Light water Reactors, molten corium which was ejected from the reactor vessel bottom, may erode the concrete basemat of the containment and there by threaten the containment integrity. This study experimentally examines the molten core-concrete interaction (MCC) using 20kg of thermite melt (Fe + $Al_2$O$_3$) and the concrete, used in Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3 & 4) in Korea. The measured data are the downward heat fluxes, concrete erosion rate, gases and particle generation rates during MCCI. Transient results ore compared with those of TURCIT experiment conducted by SNL in USA. The peak downward heat flux to the concrete was measured to be about 2.1㎿/$m^2$. The initial concrete erosion rate was 175cm per hour, decreasing to 30cm per hour. It was shown from the post-test that the erosion was progressed downward up to 18mm in the concrete slug.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.4
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• pp.384-389
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• 2004

As the increase of the operation year of nuclear power plants, the probabilities of the degradation of the major facilities and materials in the nuclear power plants are increased. The integrity of those facilities shall be monitored and verified by the non-destructive examination methods with the regulation codes, so called inservice inspection(ISI). The ISI of Yonggwang unit 6 was performed in four different parts, 1) non-destructive examinations for the components, piping weldments and structures, 2) automated ultrasonic examinations for pressure vessels, 3) visual examinations for the interior structures of the reactor, 4) eddy current examinations for the steam generator tubes. As the results, there was no severe indication and all detected indications were evaluated as non-relavent. Especially for the examinations of the piping weldments, PD(Performance Demonstration) was applied as a W examination method defined in the 1995 edition of ASME Code Sec. XI. The implementation of the PD for the piping weld results in an improvement of the reliability of the UT examinations.