• Journal of KSNVE
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• v.5 no.3
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• pp.361-371
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• 1995

The vibration of reactor internals should be monitored and diagnosed for the early detection of the failure of reactor pressure vessel. This can be performed by analyzing the time-history signals from the excore neutron flux detertors. The conventional method is an on-demand system which generates power spectra through Fast Fourier Transform(FFT) algorithm. The operator can make his own decision to detect abnormal vibration using these spectra. This post- processing method, however, requires special expertise in the reactor noise analysis and signal processing for random data. It may mislead the operator into erroneous decision-making, if he is a novice in reactor noise analysis. Hence this study is focused on the automated monitoring and diagnosis procedure for the reactor noise analysis, especially on the Fuzzy algorithm to recognize the pattern of the vibration of Core Suport Barrel. The excore neutron signals of Yonggwang Nuclear Power Plant unit 3 is acquired and analyzed using conventional FFT spectra and tested to adopt the Fuzzy method. An Automated Monitoring and Diagnosis System for CSB Vibration using this Fuzzy method is proposed. Furthermore, vibration data for CSB of Youggwang Nnclear Power Plant unit 3 is presented.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.502-506
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• 1994

The design of safety computer system is described along with the case of software design and testing in the Core Protection Calculator System (CPCS). The application of computer system in safety system requires not only hardware qualification but thorough testing on software to verify its correctness and completeness. The testing on software for CPCS is performed by comparing the outputs of two versions of code. One is implemented in assembly language and the other is in Fortran. The testing is performed in sequencial and overlapping manner. Phase I test verifies that each software module is implemented correctly by executing every branch. Phase II test verifies that the integrated software is complete, meeting its requirements specification and also the integrated system meet its requirement and timing constraints. Through these testing, the Yonggwang Nuclear Power Plant Units (YGN) 3 and 4 CPCS software is verified to be correct and complete, and the integrated system is designed as in its requirements specification.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.271-279
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• 2003

Soil blocks were taken into culture boxes from 12 paddy fields within 5 km radii of Yonggwang and Ulchin NPPs and $^{90}Sr$ was applied to the surface water at a pre-transplanting stage and $1{\sim}2$ days before the start of heading. Following the pre-transplanting application, transfer factors were investigated for $2{\sim}4$ years. In the year of application, transfer factors $(m^2\;kg^{-1}-dry)\;of\;^{90}Sr$ applied before transplanting, showing no regionally distinguishable trend, varied with soils by a factor of about 2 with averages of $2.6{\times}10^{-4}$ for hulled seeds and $1.3{\times}10^{-2}$ for straw Transfer factors of $^{90}Sr$ applied shortly before heading were about 2 times greater than those applied before transplanting. Transfer factors tended to decrease with increasing soil pH and exchangeable Ca. Generic values of $^{90}Sr$ transfer factors in the year of deposition were proposed for the Korean paddy fields. In the second year compared with the first year, the transfer factor decreased more in Ulchin soils, which were on the whole higher in sand content, than in Yonggwang soils. For Yonggwang soils as a whole, the annual decrease in transfer factor was well described by an exponential equation with a half-life of about 2.2 years.

• Korean Journal of Fisheries and Aquatic Sciences
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• v.29 no.1
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• pp.97-106
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• 1996

Macroalgal community was analysed from December 1993 to October 1994 in Yonggwang vicinity, western coast of Korea. A total 51 species (12 green, 11 brown and 29 red algae) of marine algae were identified. Among four localities, the number of species observed was the highest as 34 species at Shimwon and the least as 31 species at Sunchanggum and Gamakdo. Seasonally, the number of species observed was the highest as 42 species in winter and the least as 18 species in summer. The species showing relatively high important value were Enteromorpha compressa, Sargassum thunbergii, Corallina pilulifera and Carpopeltis affinis, which were all common to four investigated localities. Seasonal and regional fluctuations of mean biomass was $66.0\~820.0\;g-wet\;wt/m^2$ at Hyanghado, $248.3\~886.3\;g-wet\;wt/m^2$ at Sunchanggum, $154.5\~510.2\;g-wet\;wt/m^2$ at Gamakdo and $85.0\~451.9\;g-wet\;wt/m^2$ at Shimwon, respectively. The flora investigated could be classified into six functional groups such as coarsely branched form $(41.2\%)$, sheet form $(25.5\%)$, filamentous form $(19.6\%)$, thick leathery $(7.8\%)$, crustous form $(3.9\%)$ and jointed calcarious form algae $(2.0\%)$. At the effluent area of the nuclear power plants, the algal composition of functional groups may affect species composition due to thermal pollution.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.229-239
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• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.118-120
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• 2004

The non-destructive examinations for Yonggwang unit 6 was conducted in four different fields, these are 1)all non-destructive inspections for components, piping weldments and structures, 2)automated ultrasonic inspection for pressure vessels weldments. As the results, there were no big indications, and all indications detected during inspection were evaluated as the metallurgical and geometrical non-reinvent indications form weldments. Especially for the weldment of pipes, PD(Performance Demonstration) was applied as a UT inspection method according to 1995 edition of ASME code Sex. XI, this resulted in improvement of the reliability of UT inspection.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1995.10a
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• pp.201-207
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• 1995

A series of vibration assessment programs has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibration prior to its commercial operation. The structural analysis was done to provide the basis for measurement and the theoretical evidence for the structural integrity of the reactor internals. The actual flow induced hydraulic loads and reactor internals vibration response data were measured and recorded during pre-core hot functional testing of the plant. Then, the measured data have been reduced and analyzed, and compared with the analysis results such as the frequency contents, stresses, strains and displacements. It is concluded that the structural analysis methodology performed for vibration response of the reactor internals due to the flow induced vibration is appropriately conservative, and also that the structural integrity of YGN 4 reactor internals to flow induced vibration is acceptable for long term operation.

• Journal of Radiation Protection and Research
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• v.26 no.1
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• pp.19-26
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• 2001

In order to assess the atmospheric dispersion for the accidental releases of nuclear power plants, in calculating X/Q values in the XOQAR and PAVAN codes which are based on Reg. Guide 1.145, the X/Q and frequency values are plotted on log-normal paper. Starting with the highest X/Q value of this plot, the codes compare the slope of the line drawn from this point to every other point within an increment containing ten X/Q values. If there are fewer than ten values, only the number available are used. The coefficients that produce the line with the least negative slope are saved. The end point of this line is used as the next starting point, from which slopes to the points within the next increment, containing ten X/Q values, are compared. The X/Q values corresponding to the cumulative frequency values 0.5%, 5% or 50% are calculated to search for the $0{\sim}2$ hour X/Q value that tends to be a very conservative value. In this work, a fuzzy logic inference method is used for nonlinear interpolation of the X/Q values versus the cumulative frequency. The fuzzy logic inference method is known to be a food technique for nonlinear interpolation. The proposed method was applied to a potential accidential radioactive release of the Yonggwang nuclear power plant, which gives more realistic X/Q values.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.311-318
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• 2010

U. S. NRC Regulation Guide 1.52 regulating ESF ACS in nuclear power plants has been revised to revision 3. To apply reduction of operability test time, allowance of alternative challenge agents for in-place leak test of HEPA filters, and upgrade of Methyl Iodide penetration acceptance criterion in activated carbon performance test suggested in Reg. Guide 1.52(Rev.3) on Yonggwang units 5 and 6 ESF ACSes, technical feasibility study was carried out with on-site experiments as well as experiments with a lab-scale model. It was confirmed that the moisture in the system returned to the level before the test in 1 or 4 days even though the moisture was removed during the operability test lasting more than 10 hours. Therefore, it is appropriate to perform monthly operability test in 15 minutes just long enough to check the operability of equipment. To change challenge material for in-place HEPA filter leak test, size of aerosol, production rate, and leak detection capability were compared for DOP and PAO. It was concluded that PAO can be substituted for DOP in nuclear power plants. The upgrade of Methyl Iodide penetration acceptance criterion from 0.175 % to 0.5 % in active carbon filter bed deeper than 4 inches was to conform to the change of activated carbon performance test method to ASTM D3803(1989). It was confirmed that Methyl Iodide penetration acceptance criterion of 0.5 % under $30^{\circ}C$, relative humidity 95 % condition was conservatively good enough for testing performance of active carbon insitu. The licence change of Yonggwang units 5 and 6 has been completed based on this study.

• Journal of Radiation Protection and Research
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• v.29 no.2
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• pp.117-127
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• 2004

A dynamic compartment model was developed to evaluate the transport of accidently released radionuclides onto rice-fields. In the model, the surface water compartment and shoot-base absorption were introduced to account for the effect of irrigation, which is essential to a rice cultivation. The soil mixing by plough and irrigation before transplanting rice was also considered, and the rate of root-uptake and shoot-base absorption were modeled in terms of the function of biomass. In order to test the validation of the model, it was applied to the analysis of some simulated $^{137}Cs$ deposition experiments that were performed while cultivating rice in a greenhouse using soils sampled from rice-fields around Kori, Yonggwang and Ulchin nuclear power plants. The model prediction was generally agreed within about one order of magnitude with experimental data.