• KEPCO Journal on Electric Power and Energy
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• v.1 no.1
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• pp.1-8
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• 2015

When the line and switchgear of the substation system are disconnected, ferro-resonance can occur. This happens even if the capacitive reactance and inductive reactance are not equal, which are not common resonance conditions. Resonance conditions vary depending on the busbar configuration environment. Although the damping resistance method applying the existing saturable reactor to cope with ferro-resonance has been successfully applied on site, there can be loss of normal function during long-term operation. The reason is because the rise in the operating frequency of saturable reactors means the saturation number is increased. Therefore, it can no longer function as saturable reactor since the resistor having inadequate capacity is burned out. To address this problem, in this paper, an EMTP-based simulation test was performed by designing and applying a dual reactor method, which adds an extended divergence reactor to the 1st side of the VT. The test result confirms that when the divergence reactor is inserted, the voltage and current values obtained at the 1st side and 2nd side of the VT as well as current values of divergence reactor part were stabilized from the transient phenomena and return to normal values. When compared with existing measures, although this method is similar in adding having a reactor added to a system regarding ferro-resonance, it has the advantage of being able to prevent ferro-resonance in advance since the reactor is added before the system is saturated. In addition, because it does not use damping resistance, it can extend the equipment life and stabilize its operation. Therefore, there are a lot of differences in terms of its operating characteristics and achivement of goal between the conventional method and new divergence reactor method.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4560-4570
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• 2022

Dynamic fault tree (DFT) and its related research methods have received extensive attention in safety analysis and reliability engineering. DFT can perform reliability modelling for systems with sequential correlation, resource sharing, and cold and hot spare parts. A technical modelling method of DFT is proposed for modelling ship collision accidents and loss-of-coolant accidents (LOCAs). Qualitative and quantitative analyses of DFT were carried out using the cutting sequence (CS)/extended cutting sequence (ECS) method. The results show nine types of dynamic fault failure modes in ship collision accidents, describing the fault propagation process of a dynamic system and reflect the dynamic changes of the entire accident system. The probability of a ship collision accident is 2.378 × 10^-9 by using CS. This failure mode cannot be expressed by a combination of basic events within the same event frame after an LOCA occurs in a marine nuclear reactor because the system contains warm spare parts. Therefore, the probability of losing reactor control was calculated as 8.125 × 10^-6 using the ECS. Compared with CS, ECS is more efficient considering expression and processing capabilities, and has a significant advantage considering cost.

• Journal of Microbiology and Biotechnology
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• v.12 no.1
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• pp.14-17
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• 2002

To increase the efficiency of a two-stage anaerobic wastewater digestion system, various polymers were added to the methanogenic reactor as supports. The addition of polyurethane addition (6%, w/v) to the methanogenic reactor facilitated the organic loading rate (2-day Hydraulic Retention Time), higher than that of the conventional methanogenic reactor (6-day HRT). During the operation of the polyurethane-added reactor, a significant decrease in the organic mass in the effluent (COD 5-6 kg/l) was achieved, compared to that of the conventional reactor (COD 15-20 kg/l). The methane gas production rate also improved about 3-fold in the polyurethane-added reactor. More biomass was found to accumulate in the polyurethane-liquid phase (volatile solid, 26-28kg) than in the free-liquid phase (volatile solid, 5- 7 kg/l) after 90 days of operation. A scaled-up experiment with a polyurethane-added 2.5-1 reactor confirmed the previous results, and no adverse effects such as plugging or channeling due to decreased efficiency was observed even after 4 months of operation.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2004.05a
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• pp.395-400
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• 2004

This study is that STEP based Data Repository of APR1400 Nuclear Power Plant Reactor Coolant System is developed. The STEP based Data Repository is accessed by Web-based and an attribute data of Reactor Coolant System Equipment is offered. Also, a P&ID drawing file & 3D CAD Model of Reactor Coolant System is loaded. The P&ID drawing file of Reactor Coolant System Equipment Model is connected with 3D CAD Model file. This 2D/3D CAD Model connected Prototype system confirms a real layout of Reactor Coolant System.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1995.10a
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• pp.61-67
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• 1995

The steam generator, one of the major components in the reactor coolant system, plays an important role in transferring the thermal energy made in the reactor during normal operation to the secondary side and producing steam to drive turbine. A hydraulic snubber system is used in order to protect the steam generator under the dynamic loading condition and to absorb the thermal expansion transmitted by the reactor coolant piping due to high temperature and pressure during normal operation. In this study, the model for a geometrical linkage system is presented to analyze the snubber stroke of the steam generator and the parameters in the snubber stroke analysis are investigated. A method to analyze lever ratio of the linkage system which is required in the process of determining the snubber stiffness value is also presented. To discuss the validation of the suggested analysis, the analysis results are compared with the measured data during the hot functional test for the standardized 1000 Mwe pressurized water reactor plant under the construction.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.582-582
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• 2012

Characteristics of a fusion-driven transmutation reactor was investigated. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

• Earthquakes and Structures
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• v.12 no.6
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• pp.653-663
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• 2017

A seismic margin assessment evaluates how much margin exists for the system under beyond design basis earthquake events. Specifically, the seismic margin for the entire system is evaluated by utilizing a systems analysis based on the sub-system and component seismic fragility data. Each seismic fragility curve is obtained by using empirical, experimental, and/or numerical simulation data. The systems analysis is generally performed by employing a fault tree analysis. However, the current practice has clear limitations in that it cannot deal with the uncertainties of basic components and accommodate the newly observed data. Therefore, in this paper, we present a Bayesian-based seismic margin assessment that is conducted using seismic fragility data and fault tree analysis including Bayesian inference. This proposed approach is first applied to the pooltype nuclear research reactor system for the quantitative evaluation of the seismic margin. The results show that the applied approach can allow updating by considering the newly available data/information at any level of the fault tree, and can identify critical scenarios modified due to new information. Also, given the seismic hazard information, this approach is further extended to the real-time risk evaluation. Thus, the proposed approach can finally be expected to solve the fundamental restrictions of the current method.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.397-397
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• 2011

In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the radial build of TF coil and the shield play a key role in determining the size of a reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with one-dimensional radiation transport code. Conceptual design study of a compact superconducting LAR tokamak reactor with aspect ratio less than 2.5 was conducted and the optimum radial build was identified. It is shown that the use of an improved shielding material and high temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at low aspect ratio, and that by using an inboard neutron reflector instead of breeding blanket, tritium self-sufficiency is possible with outboard blanket only and thus compact sized reactor is viable.