• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.177-186
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• 2021

Radioactive materials from nuclear power facilities can be released into the atmosphere through various channels. Recently, the dispersion of radioactive materials has become critical issue in Korea after Kori Unit 1 and Wolsong Unit 1 were permanently shut down. In this study, annual atmospheric dispersion factors were compared based on the continuous release and purge release using the XOQDOQ computer program, a method for calculating atmospheric dispersion factors at commercial nuclear power stations. The meteorological data analyzed in this study was based on the Shin Kori nuclear power meteorological tower which has the largest operating nuclear power plants in Korea, for three years (from 2008 to 2010). The analysis results of the dispersion factor of the radioactive material release obtained using the XOQDOQ program showed that the difference between the continuous release and purge release was within two times. This study will be valuable helpful for revealing the uncertainty of the predictive atmospheric dispersion factor to achieve regulation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.315-334
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• 2014

Wolsong Low- and Intermediate-level radioactive waste (LILW) disposal center has two different types of disposal facilities and interacts with the neighboring Wolsong nuclear power plant. These situations impose a high level of complexity which requires in-depth understanding of phenomena in the safety assessment of the disposal facility. In this context, multidimensional radionuclide transport model and hydraulic performance assessment model should be developed to identify more realistic performance of the complex system and reduce unnecessary conservatism in the conventional performance assessment models developed for the $1^{st}$ stage underground disposal. In addition, the advanced performance assessment model is required to calculate many cases to treat uncertainties or study parameter importance. To fulfill the requirements, this study introduces the development of two-dimensional integrated near-field performance assessment model combining near-field hydraulic performance assessment model and radionuclide transport model for the $2^{nd}$ stage near-surface disposal. The hydraulic and radionuclide transport behaviors were evaluated by PORFLOW and GoldSim. GoldSim radionuclide transport model was verified through benchmark calculations with PORFLOW radionuclide transport model. GoldSim model was shown to be computationally efficient and provided the better understanding of the radionuclide transport behavior than conventional model.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.579-585
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• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Proceedings of the KIEE Conference
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• 1999.07e
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• pp.2110-2113
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• 1999

This paper describes the initial characteristics of turbine generators at the Taean thermal power plant #4 and Wolsong nuclear power plant #2. The turbine generators had been in service for two years. The insulation diagnostic tests included measurements of insulation resistance, polarization index, ac current, dissipation factor($tan{\delta}$) and partial discharges (PD). The values of ac current and tan a were measured by Schering bridge. PD measurements were conducted using digital PD detector. The variation of $tan{\delta}$ and PD was confirmed in two generator stator insulations.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.58-67
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• 1999

The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsong 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using Statechart-based Formalism and Stalemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language though we manually made Software Requirement Specification(SRS) for safety-critical software using statechart-based formalism. Most of the phases of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering (CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.315-320
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• 1998

In this work, an automatic software verification method for Nuclear Power Plant (NPP) protection system is developed. This method utilizes Colored Petri net (CPN) for modeling and Prototype Verification system (PVS) for mathematical verification. In order to help flow-through from modeling by CPN to mathematical proof by PVS, a translator has been developed in this work. The combined method has been applied to a protection system function of Wolsong NPP SDS2(Steam Generator Low Level Trip)and found to be promising for further research and applications.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.561-569
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• 1993

DOT4.2-QAD-CG coupling method was used to analyze the dose rates outside the side and the bottom shield system of CANDU 6 plant. The average dose rates at the main airlock and the new fuel loading area are approximately 6 $\mu$Sv/h as it is required. The calculated dose rates have a good agreement with the measurements at the operating CANDU 6 plant. The method used in this paper can be applied to the radiation shielding analysis of Wolsong 2, 3, and 4 CANDU 6 type plants which will be constructed in the near future.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.157-171
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• 1999

In this work, an automatic software verification method for Nuclear Power Plant (NPP) protection system is developed. This method utilizes Colored Petri Net (CPN) for system modeling and Prototype Verification System (PVS) for mathematical verification. In order to help flow-through from modeling by CPN to mathematical proof by PVS, an information extractor from CPN models has been developed in this work. In order to convert the extracted information to the PVS specification language, a translator also has been developed. ML that is a higher-order functional language programs the information extractor and translator. This combined method has been applied to a protection system function of Wolsong NPP SDS2(Steam Generator Low Level Trip). As a result of this application, we could prove completeness and consistency of the requirement logically. Through this work, in short, an axiom or lemma based-analysis method for CPN models is newly suggested in order to complement CPN analysis methods and a guideline for the use of formal methods is proposed in order to apply them to NPP Software Verification and Validation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.4
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• pp.495-499
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• 2010

In this paper, the regulatory guidelines for the continued operation of the CANDU reactor in Korea were introduced. Wolsong Unit 1, which is a CANDU 600 reactor in Korea, will reach its design life of 30 years in 2012. A licensee who wants to operate a nuclear power plant beyond its design life should submit reports of periodic safety reviews (PSRs) conducted on the basis of 11 safety factors. In addition, the licensee should provide the following: (1) scoping and screening results for aging management, (2) aging management program, (3) TLAA, including the continued operation term, (4) operation-experience feedback, and (5) important safety-research results. In this study, 54 regulatory guidelines for the five above-mentioned items for the CANDU reactor in Korea were developed.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.