• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.257-258
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• 2008

The inrush current of a transformer is a high-magnitude and harmonic-rich current generated when the transformer core is driven into saturation during energizing. The inrush current usually leads to undesirable effects, for example potential damage to the transformer, misoperation of a protective relay, and power quality deterioration in the distribution power system. Inrush current reduction is therefore important for power system operation. In this paper, to reduce the inrush current, the insertion resistance of the Superconducting Fault Current Limiter (SFCL) that is connected in series with the transformer in the distribution system is used. This paper implements the SFCL by using the Electromagnetic Transient Program-Restructured Version (EMTP-RV) to model the SFCL in the distribution system. The simulation results show the beneficial effects of the SFCL for reduction of the inrush current.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3425-3434
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• 2024

The main features of the loading pattern optimization (LPO) problem, such as high-dimensionality, multi-modality, and non-linearity, make it difficult to achieve a truly optimal configuration. In recent years, metaheuristic methods have been successfully used to solve this problem. In this research, a discrete golden eagle optimization (DGEO) algorithm has been developed to solve the LPO problem in the first cycle of VVER-1000 reactor core. To evaluate the proposed algorithm, a linear multi-purpose fitness function has been used to improve the safety parameters of the reactor core by obtaining a flatter power distribution during the first cycle, and also to enhance the economic parameters by increasing the cycle length and reducing the cost of fuel recycling. For this purpose, a FORTRAN program has been written to map the DGEO algorithm for the LPO problem using the PMAX and PARCS core calculation code to compute the fitness function in each iteration. To speed up the calculations, parallel computing has been applied in the written program. The results demonstrated that the loading pattern, which is suggested by the DGEO algorithm, enhances all the safety and economic parameters in the fitness function. Thus, the DGEO algorithm is highly reliable for the LPO problems in the VVER 1000 reactor core.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1178-1183
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• 2004

Temperature distribution and heating characteristic of the panel heater for infrared heating have been investigated. The temperature variation with time is firstly measured with the thermocouple to figure out the response time of the heater to the power input. The heater reaches faster to the steady state in comparison to the ceramic heater. The infrared thermal imaging system is utilized to investigate the temperature distribution over the heater surface. The measured thermal images show that the thermal boundary layer induced by the free convection near the heater surface affects the temperature distribution on the surface. The images also show the fairly good uniformity of the temperature distribution in the core region of the surface.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.9-18
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• 1988

Previously determining the fuel loading pattern is based on the trial and error method. For a candidate pattern, the core analysis is performed and the pattern is examined whether it satisfies the imposed constraints such as the power peaking or not. The pattern, then, is revised by the shuffling of assemblies and the revision is repeated until all of the conditions are met. This method unavoidably requires many iterative diffusion calculations, computing times and accumulated experiences. To overcome these disadvantages, a new method which is called backward diffusion calculation is introduced. If the most desirable power distribution is already known, the optimal loading pattern can be obtained by solving the backward diffusion equation with simple calculation. In this study, the basic equation for the backward diffusion calculation is derived and the optimal power and fuel distributions are searched in one-dimensional cylindrical geometry by using the proposed method. In addition, the basis to determine the optimal power and fuel distributions is suggested for the real core geometry.

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 2009.10a
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• pp.423-427
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• 2009

In the circumstances which a highly reliable operation in electric power facilities of extra high voltage and large capacity is needed, the importance of a preventive diagnostic technology is growing large day and day. The settlement of a preventive diagnostic technology for optimization and efficient management on the electric power facilities like GIS enable the reduce of repair fee, the improvement of safety and the systematic management of electric power facilities. The remaining life prediction of facilities will play a decisive role as a core technology of a preventive diagnostics in the future. And so it is necessary a continuous research and concern for the development of a preventive diagnostic technology hereafter.

• KEPCO Journal on Electric Power and Energy
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• v.5 no.2
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• pp.113-120
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• 2019

We have developed 1 kW-class PEMFC-battery hybrid system independently powering to the building, through the process of system design, current load characteristics analysis, power system configuration for demonstration site and performance evaluation. In order to use the fuel cell and battery as the hybrid type, a control technology for the charging/discharging decision and charging speed of the battery is required rather than using fuel cell. Also output power distribution between PEMFC and the battery is a core of energy management technology. It is confirmed that it is possible to supply independently 1kW powering the building to ensure optimal energy management through the power control experiment of the hybrid system.