• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1919-1920
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• 2006

This paper deals with the design review of a power converter topologies for CEDMCS (Control Element Drive Mechanism Control System). The CEDMCS provides the control signals and motive power to operate the CEDMS. The CEDM's raise and lower the CEAs (Control Element Assemblies) in the reactor core. The CEAs are constructed with the Boron-10 isotope which has a high microscopic cross section of absorption for thermal neutrons. This characteristic causes the addition of negative reactivity when a CEA is inserted and positive reactivity when it is withdrawn from the reactor core.

• Proceedings of the KIEE Conference
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• 2005.07a
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• pp.664-666
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• 2005

본 논문에서는 장거리 초고압 송전선로에서의 단상재폐로시에 발생가능한 과전압과 2차아크전류를 EMTP를 이용하여 예측계산 하였다. 예측계산결과로부터 과전압을 억제하고 재폐로 동작이 가능한 최적의 중성점 리액터와 400 kV 급 수평배열 1회선 선로를 단상으로 재폐로 하는데 필요한 재폐로 무전압시간을 구하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.61-66
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• 1996

High pressure. high temperature steam/water blowdown test loop has been constructed. The loop simulates a pressurizer. depressurizalion system and In-Containment Refueling Water Storage Tank (IRWST) with full pressure and temperature conditions. and will be used to generate data for development of an optimal sparser as well as for design of safety/automatic depressurization system. In addition. experiments for reactor safety and pressurizer thermal hydraulics are scheduled. In this paper. general description of the Blowdown and Condensation (B&C) Loop will be given together with the test program.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.437-442
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• 1997

To determine the optimum pressurizer water level during normal operation for Kori unit 1, performance and safety analysis are peformed. The methodology is developed by evaluating "decrease in secondary heat removal" events such as Loss of Normal Feedwater accident. To demonstrate optimum pressurizer level setpoint, RETRAN-03 code is used for performance analysis. Analysis results of RETRAN following reactor trip are compared with the actual plant data to justify RETRAN code modelling, The results of performance and safety analyses show that the newly established level setpoints not only improve the performance of pressurizer during transient including reactor trip but also meet the design bases of the pressurizer volume and pressure. pressure.

• International Journal of Fuzzy Logic and Intelligent Systems
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• v.3 no.1
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• pp.112-120
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• 2003

This paper reports the development work carried out in respect of a proposed application of Neural Network approach for the Korean Next generation Reactor (KNGR) now referred as APR-1400. The emphasis is on establishing the methodology and the approach to be adopted towards realizing this application in the next generation reactors. Keeping in view the advantages and limitation of Artificial Neural Network Approach, the role of ANN has been limited to plant status or to be more precise plant transient monitoring. The simulation work carried out so far and the results obtained shows that artificial neural network approach caters to the requirements of plant status monitoring and qualifies to be incorporated as a part of proposed operator support systems of the referenced nuclear power plant.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.291-292
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• 2007

This paper deals with the design of control cabinet based on safety PLC for Control Rod Control System(CRCS). The CRCS controls the operation of the CRDMs(Control Rod Drive Mechanisms). The CRDM moves the control rods which regulate the reactor power. vertically in the reactor core. The Control Cabinet in CRCS makes and conveys control signals to the power cabinet which provides power to the CRDM. We designed the Control Cabinet, based on POSAFE-Q, safety PLC. The application programs working in PLC can be programmed by pSET(POSAFE-Q Software Engineering Tool), Identified Development Environment.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.88-93
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• 1996

The initial condition should be consistent with real reactor core state for the simulation of the core transient. The initial xenon distribution, which cad not be measured in the core, has a significant effect on the transient with xenon dynamics of PWR. In the simulation of the transient stating from non-equilibrium xenon state, the accurate initialization of the non-equilibrium xenon distribution is essential to predict the core transient behavior. In this study, the xenon initialization method to predict the core transient more accurately was developed through the first-order perturbation theory of the relationship between simulated power and measured power distribution and verified by the application of the simulation for a startup test of Yonggwang Unit 3.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.244-252
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• 2021

Wind plays an important role in cases of unexpected radioactive pollutant dispersion, deciding distribution and concentration of the leaked substance. The accurate prediction of wind has been challenging in numerical weather prediction models, especially near the surface because of the complex interaction between turbulent flow and topographic effect. In this study, we investigated the characteristics of atmospheric dispersion of radioactive material (i.e. ¹³⁷Cs) according to the simulated boundary layer around the HANARO research nuclear reactor in Korea using the Weather Research and Forecasting (WRF)-Mesoscale Model Interface (MMIF)-California Puff (CALPUFF) model system. We examined the impacts of orographic drag on wind field, stability calculation methods, and planetary boundary layer parameterizations on the dispersion of radioactive material under a radioactive leaking scenario. We found that inclusion of the orographic drag effect in the WRF model improved the wind prediction most significantly over the complex terrain area, leading the model system to estimate the radioactive concentration near the reactor more conservatively. We also emphasized the importance of the stability calculation method and employing the skillful boundary layer parameterization to ensure more accurate low atmospheric conditions, in order to simulate more feasible spatial distribution of the radioactive dispersion in leaking scenarios.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.53-62
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• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.