• Proceedings of the KIPE Conference
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• 2012.11a
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• pp.85-87
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• 2012

This paper describes the design and analysis of an integrated transformer magnetic core which comprises of two different power cores with PFC inductor and LLC resonant transformer magnetic cores. The equivalent magnetic circuit modeling approach is employed to analyze the variations in coupling coefficient and inductance in terms of air gaps under the operations of the respective power core. Simulation and experimental studies are performed with a fabricated prototype integrated core and their results are discussed.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2005.03a
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• pp.453-460
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• 2005

In this study, the seismic fragility analysis of a base isolated condensate storage tank installed in the nuclear power plant. The condensate storage tank is safety related structure in a nuclear power plant. The failure of this tank affect significantly to the core damage frequency of the nuclear power plants. The seismic analysis of the liquid storage tank was performed by the simple calculation method and the dynamic time storage analysis method. The convective and impulsive fluid mass is modeled as added masses proposed by several researchers. To evaluate the effectiveness of the isolation system, the comparison of HCLPF and core damage frequencies in non-isolated and isolated cases are carried out. It can be found from the results that the seismic isolation system increases the seismic capacity of a condensate storage tank and decreases the core damage frequency significantly.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.41-47
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• 1995

A parametric study bas been performed for the various refueling schemes of CANDU 6 reactor loaded with reference DUPIC fuel. The optimum discharge burnup was determined such that the peak bundle power is minimized for the equilibrium core. Based on the results of instantaneous core calculation using patterned random age distributions, it was decided to perform the refueling simulations only for 2-bundle and 4-bundle shift refueling schemes. The 600 FPD simulation has shown that the operational margins of the channel and bundle power to the license limits are 7.9% and 17.1%, respectively, for 2-bundle shift refueling scheme. The 4-bundle shift refueling scheme also satisfies the license limits and the operational margins of the channel and bundle power are 7.1% and 9.8%, respectively. The result of refueling simulation indicate the possibility of using reference DUPIC fuel in current CANDU 6 reactor.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.679-686
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• 2000

Reactor Noise is defined as the fluctuations of measured instrumentation signals during full-power operation of reactor which have informations on reactor system dynamics such as neutron kinetics, thermal-hydraulics, and structural dynamics. Reactor noise analyses of ex-core neutron detector signals have been performed to monitor the vibration modes of reactor internals such as fuel assembly and Core Support Barrel in Yonggwang 3&4 Nuclear Power Plant. A real time mode separation technique have been developed and applied for the analyses. It has been found that the first vibration mode frequency of the fuel assembly was around 2.5 Hz, the beam and shell mode frequencies of CSB(Core Support Barrel) 8 Hz and 14.5 Hz, respectively. Also the analyses data base have been constructed for the continuous monitoring and diagnose of the reactor internals.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.141-147
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• 2017

Mechanical shim is an advanced technology for reactor power and axial offset control with control rod assemblies. To address the adverse accuracy impact on the ex-core power range neutron flux measurements-based axial offset control resulting from the variable positions of control rod assemblies, the lead-lag-compensated in-core self-powered vanadium detector signals are utilized. The prompt ${\gamma}$ current of self-powered detector is ignored normally due to its weakness compared with the delayed ${\beta}$ current, although it promptly reflects the flux change of the core. Based on the features of the prompt ${\gamma}$ current, a method for configuration of the lead-lag dynamic compensator is proposed. The simulations indicate that the method can improve dynamic response significantly with negligible adverse effects on the steady response. The robustness of the design implies that the method is of great value for engineering applications.

• Journal of international Conference on Electrical Machines and Systems
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• v.3 no.1
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• pp.78-82
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• 2014

Power Transformers with more than one secondary winding are not uncommon in industrial applications. But new classes of applications where very large number of independent secondaries are used are becoming popular in controlled converters for medium and high voltage applications. Cascade H-bridge medium voltage drives and Pulse Step Modulation (PSM) based high voltage power supplies are such applications. Regulated high voltage power supplies (Fig. 1) with 35-100 kV, 5-10 MW output range with very fast dynamics (${\mu}S$ order) uses such transformers. Such power supplies are widely used in fusion research. Here series connection of isolated voltage sources with conventional switching semiconductor devices is achieved by large number of separate transformers or by single unit of multi-secondary transformer. Naturally, a transformer having numbers of secondary windings (~40) on single core is the preferred solution due to space and cost considerations. For design and simulation analysis of such a power supply, the model of a multi-secondary transformer poses special problem to any circuit analysis software as many simulation softwares provide transformer models with limited number (3-6) of secondary windings. Multi-Secondary transformer models with 3 different schemes are available. A comparison of test results from a practical Multi-secondary transformer with a simulation model using magnetic component is found to describe the behavior closer to observed test results. Earlier models assumed magnetising inductance in a linear loss less core model although in actual it is saturable core made-up of CRGO steel laminations. This article discusses a more detailed representation of flux coupled magnetic model with saturable core properties to simulate actual transformers very close to its observed parameters in test and actual usage.

• Journal of the Korean Society of Systems Engineering
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• v.18 no.2
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• pp.58-74
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• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• Progress in Superconductivity and Cryogenics
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• v.12 no.4
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• pp.24-30
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• 2010

The renewable energy source is considered as a good measure to cope with the global warming problem and the fossil energy exhaustion. The construction of electric power plant such as an offshore wind farm is rapidly increasing and this trend is expected to be continued during this century. The bulky and long distance power transmission media is essential to support and promote the sustainable expansion of renewable energy source. DC power cable is generally considered as the best solution and the demand for DC electric power has been rapidly increasing. Especially, the high temperature superconducting (HTS) DC cable system begins to make a mark because of its advantages of huge power transmission capacity, low transmission loss and other environmental friendly aspects. Technical contents of DC HTS cable system are very similar to those of AC HTS cable system. However the DC HTS cable can be operated near its critical current if the heat generation is insignificant, while the operating current of AC HTS cable is generally selected at about 50~70% of the critical current because of AC loss. We chose the specifications of the cable core of 'Tres Amigas' project as an example for our study and investigated the heat generation when the DC HTS cable operated near the critical current by some electric and thermal analyses. In this paper, we listed some technical issues on the design of the DC HTS cable core and described the process of the cable core design. And the results of examination on the current capacity, heat generation, harmonic loss and current distribution properties of the DC HTS cable are introduced.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.452-459
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• 2023

The core power control is an important issue for the study of dynamic characteristics in China initiative accelerator driven subcritical system (CiADS), which has direct impact on the control strategy and safety analysis process. The CiADS is an experimental facility that is only controlled by the proton beam intensity without considering the control rods in the current engineering design stage. In order to get the optimized operation scheme with the stable and reliable features, the variation of beam intensity using the continuous and periodic control approaches has been adopted, and the change of collimator and the adjusting of duty ratio have been proposed in the power control process. Considering the neutronics and the thermal-hydraulics characteristics in CiADS, the physical model for the core power control has been established by means of the point reactor kinetics method and the lumped parameter method. Moreover, the multi-inputs single-output (MISO) logical structure for the power control process has been constructed using proportional integral derivative (PID) controller, and the meta-heuristic algorithm has been employed to obtain the global optimized parameters for the stable running mode without producing large perturbations. Finally, the verification and validation of the control method have been tested based on the reference scenarios in considering the disturbances of spallation neutron source and inlet temperature respectively, where all the numerical results reveal that the optimization method has satisfactory performance in the CiADS core power control scenarios.