• Journal of the Earthquake Engineering Society of Korea
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• v.24 no.2
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• pp.87-93
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• 2020

Seismic responses due to the dynamic coupling between a primary structure and secondary system connected to a structure are analyzed in this study. The seismic responses are compared based on dynamic coupling criteria and according to the error level in the natural frequency, with the recent criteria being reliant on the error level in the spectral displacement response. The acceleration responses and relative displacement responses of a primary structure and a secondary system for a coupled model and two different decoupled models of two degrees-of-freedom system are calculated by means of the time integration method. Errors in seismic responses of the uncoupled models are reduced with the recent criteria. As the natural frequency of the secondary system increases, error in the natural frequency decreases, but seismic responses of uncoupled models can be underestimated compared to that of coupled model. Results in this paper can help determine dynamic coupling and predict uncoupled models' response conservatism.

• Fire Science and Engineering
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• v.20 no.3 s.63
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• pp.1-8
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• 2006

The purpose of this work is to analyze quantitatively if the safety instrumented system(SIS) like the pressure safety valves(PSV) in the processes of ethyl benzene plant have been designed relevantly to the safety integrity level because overpressure in the benzene or ethyl benzene columns causes the explosive reactions, fires and reactor explosions. The safety integrity level(SIL) 3 has been adopted as a target level of SIS based on the general data of the Probability of Failure on Demand of PSV, $1.00E-4{\sim}1.00E-3$. The standard model of the reliability has been set up and then the fault tree analysis of it has been carried out to get the PFD of SIS, and the results show 8.97E-04, 5.37E-04, 5.37E-04 for benzene prefractionator column, benzene column and EB column, respectively. Thus, we conclude that the SIS is designed to fulfill the condition of SIL3, and when the partial stroke test for the control valve are carried out every sixth month, the SIS of each column is expected to increase its reliability up to $22{\sim}27%$.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2107-2118
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• 2020

The nuclear reactor coupled with supercritical carbon dioxide (S-CO₂) Brayton cycle has good prospects in generation IV reactors. Turbomachineries (turbine and compressor) are important work equipment in circulatory system, whose performances are critical to the efficiency of the energy conversion system. However, the sharp variations of S-CO₂ thermophysical properties make turbomachinery performances more complex than that of traditional working fluids. Meanwhile, almost no systematic analysis has considered the effects of turbomachinery efficiency under different conditions. In this paper, an in-house code was developed to realize the geometric design and performance prediction of S-CO₂ turbomachinery, and was coupled with systematic code for Brayton cycle characteristics analysis. The models and methodology adopted in calculation code were validated by experimental data. The effects of recompressed fraction, pressure and temperature on S-CO₂ recompression Brayton cycle were studied based on detailed design of turbomachinery. The results demonstrate that the recompressed fraction affects the turbomachinery characteristic by changing the mass flow and effects the system performance eventually. By contrast, the turbomachinery efficiency is insensitive to variation in pressure and temperature due to almost constant mass flow. In addition, the S-CO₂ thermophysical properties and the position of minimum temperature difference are significant influential factors of cyclic performance.

• Journal of the Korean Society of Safety
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• v.34 no.1
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• pp.27-33
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• 2019

Combustion characteristics of gasoline/ethanol fuel were investigated both numerically and experimentally for vehicle fire safety. The numerical simulation was performed on the well-stirred reactor (WSR) to simulate the homogeneous gasoline engine and to clarify the effect of ethanol addition in the gasoline fuel. The simulating cases with three independent variables, i.e. ethanol mole fraction, equivalence ratio and residence time, were designed to predict and optimized systematically based on the response surface method (RSM). The results of stoichiometric gasoline surrogate show that the auto-ignition temperature increases but NOx yields decrease with increasing ethanol mole fraction. This implies that the bioethanol added gasoline is an eco-friendly fuel on engine running condition. However, unburned hydrocarbon is increased dramatically with increasing ethanol content, which results from the incomplete combustion and hence need to adjust combustion itself rather than an after-treatment system. For more tangible understanding of gasoline/ethanol fuel on pollutant emissions, experimental measurements of combustion products were performed in gasoline/ethanol pool fires in the cup burner. The results show that soot yield by gravimetric sampling was decreased dramatically as ethanol was added, but NOx emission was almost comparable regardless of ethanol mole fraction. For soot morphology by TEM sampling, the incipient soot such as a liquid like PAHs was observed clearly on the soot of higher ethanol containing gasoline, and the soot might be matured under the undiluted gasoline fuel.

• Fire Science and Engineering
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• v.32 no.3
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• pp.42-50
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• 2018

The effect of sub-grid turbulence and combustion models on the mean flame height in a heptane pool fire according to the Fire Dynamics Simulator (FDS) version (5 and 6) based on Large Eddy Simulation (LES) was examined. The heat release rate for the fire simulation was provided through experiments performed under identical conditions and the predictive performance of the mean flame height according to FDS version was evaluated by a comparison with the existing correlation. As a result, the Smagorinsky and Deardorff turbulence models applied to FDS 5 and 6, respectively, had no significant effects on the mean flow field, flame shape and flame height. On the other hand, the difference in pool fire characteristics including the mean flame height was due mainly to the difference in the mixture fraction and Eddy Dissipation Concept (EDC) combustion models applied to FDS 5 and 6, respectively. Finally, compared to FDS 6, FDS 5 provided the predictive result of a significantly longer flame height and more consistent mean flame height than the existing correlation.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.97-106
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• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.185-190
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• 1995

Conventional alarm system has many difficulties in the operator's identifying the plant status during special situations such as design basis accidents. To solve the shortcomings, an on-line alarm annunciator system, called dynamic alarm console (DAC), was developed. In the DAC, a signal is generated as alarm by the use of an adaptive setpoint check strategy based on operating mode, and time delay technique is used not to generate nuisance alarms. After alarm generation, if activated alarm is a level precursor alarm or a consequencial alarm, it would be suppressed, and the residual alarms go through dynamic prioritization which provide the alarms with pertinent priorities to the current operating mode. Dynamic prioritization is achieved by going through the system- and mode-oriented prioritization. The DAC has the alarm hierarchical structure based on the physical and functional importance of alarms. Therefore the operator can perceive alarm impacts on the safety or performance of the plant with the alarm propagation from equipment level to plant functional level. In order to provide the operator with the most possible cause of the event and quick cognition of the plant status even without recognizing the individual alarms, reactor trip status tree (RTST) was developed. The DAC and the RTST have been simulated with on-line data obtained from the full-scope simulator for several abnormal cases. The results indicated that the system can provide the operator with useful and compact information fur the earlier termination and mitigation of an abnormal state.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.697-708
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• 2012

The majority of instrumentation and control (I&C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I&C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I&C systems. Therefore, existing NPPs are replacing the obsolete analog I&C systems with advanced digital systems. New NPPs are also adopting digital I&C systems because the economic efficiencies and usability of the systems are higher than the analog I&C systems. Digital I&C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1791-1798
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• 2019

Instrumentation and Control (I&C) systems of nuclear power plants (NPPs) have been continuously digitalized. These systems have a critical role in the operation of nuclear facilities by functioning as the brain of NPPs. In recent years, as cyber security threats to NPP systems have increased, regulatory and policy-related organizations around the world, including the International Atomic Energy Agency (IAEA), Nuclear Regulatory Commission (NRC) and Korea Institute of Nuclear Nonproliferation and Control (KINAC), have emphasized the importance of nuclear cyber security by publishing cyber security guidelines and recommending cyber security requirements for NPP facilities. As described in NRC Regulatory Guide (Reg) 5.71 and KINAC RS015, challenge response authentication should be applied to the critical digital I&C system of NPPs to satisfy the cyber security requirements. There have been no cases in which the most robust response authentication technology like challenge response has been developed and applied to nuclear I&C systems. This paper presents a challenge response authentication mechanism for a Programmable Logic Controller (PLC) system used as a control system in the safety system of the Advanced Power Reactor (APR) 1400 NPP.