• Journal of the Korean Society of Safety
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• v.31 no.6
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Transactions of the Korean Society of Automotive Engineers
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• v.14 no.2
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• pp.39-48
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• 2006

An engine cooling system affects overall performances of an engine which has been recently requested higher power in more confined engine room. The design of efficient cooling system demands a great effort to effectively correlate with each components, such as water jacket, radiator, coolant pump, cooling fan, etc. Thus, the aim of this study is to provide the design tool of the cooling system in the early design stage by enabling for the designer to accurately predict the engine cooling performances. This user-friendly design tool has various ways to assemble each components and control the running condition with related database. The present design tool was simulated and compared with experimental data. As a result, the inlet and outlet temperature of the radiator agree very well with experiments. It was concluded that the present design tool could be effectively used for the design of the engine cooling system.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.2
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• pp.1-6
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• 2019

The objective of this study was to investigate performance characteristics of thermal management system(TMS) in a fuel cell electric vehicle with 100kW Fuel Cell(FC) system. In order to build up analytic modelling for TMS, each component was installed and tested under various operating conditions, such as water pump, radiator, 3-Way valve, COD heater, and FC stack etc. and as the results of them, correlations reflecting component's characteristics with flow rate, air velocity were developed. Developed analytic modelling was carried out under various operating conditions on the road. To verify modelling's accuracy, after prediction for optimum coolant flow rate was fulfilled under certain operating conditions, such as FC system, water pump speed, opening of 3-way valve, and pipe resistance, analytic and experimental values were compared and good agreement was shown. In order to predict cold-start operating performance for analytic modelling, coolant temperature variation was analyzed with $-20^{\circ}C$ ambient temperature and duration was predicted to rise in optimum temperature for FC. Because there is appropriate temperature difference between inlet and outlet of FC stack to operate FC system properly, related analysis was performed with respect to power consumption for TMS and heat rejection rate and performance map was depicted along with FC operating conditions.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.127-134
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• 2017

The reactor internals shall be secured in safety and structural integrity under various vibrational loading conditions. Thus, U.S. NRC, Regulatory Guide 1.20 requires the evaluation for the reactor internals due to acoustic induced vibration including the response to the reactor coolant pump pressure pulsation. This paper suggests a methodology to develop an analytical model of the core support barrel accounting for the fluid around the structure and to analyze the responses to the pump pulsation loads using acoustic structure interaction analysis. The analysis results were compared with those of US Palo Verde 1 CVAP and showed a good agreement. Thus, it is expected that the suggested methodology could be an efficient way to evaluate the response of the core support barrel to the pump pulsation loads.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2009.06a
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• pp.361-361
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• 2009

KEPCO High Temperature Superconducting (HTS) cable system rated with $3\Phi$, 22.9kV, 1250A was laid in 2006, and the long term test is in progress. The HTS cable system with the cooling system has been operated below cryogenic temperature. That environment exposes the system to the thermo-mechanical stress due to the significant temperature difference, and the cooling system has moving parts for the forced circulation of the coolant. Therefore the HTS cable system experiences thermal fatigue and moving part such as liquid nitrogen pump need a regular replacement every 5000 hours Building the assessment criterion, the maintenance procedure was established and regular preventive maintenance was done; improvement of the termination structure and the replacement of the bearing of liquid nitrogen pump. Following the proper process, the reliability assessment test including He leakage detection and the stability of flow rate was performed. This paper describes the process and result of the first regular maintenance of KEPCO HTS cable system

• The Transactions of the Korean Institute of Electrical Engineers B
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• v.50 no.10
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• pp.483-493
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• 2001

The 3-phase induction is used for the MCP(main coolant pump) and the pure water is used for lubrication of bearing because of the difficulty of repair. Therefore the type of motor is the canned-motor that is welded by sealed can to prevent the stator and rotor from the lubricating water. A lot of Eddy currents are produced in the can because of the conductivity of material. And these eddy currents in the can are the most important cause that decrease the efficiency of induction motor. Therefore we have to find the method to decrease these eddy currents in the can for the improvement of efficiency of motor. In this paper, we proposed the method of design and analysis of canned-motor using equivalent circuit with consideration of can loss for the improvement of efficiency of motor.