• Journal of the Institute of Electronics Engineers of Korea TC
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• v.43 no.12 s.354
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• pp.9-15
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• 2006

In this paper, signal processing techniques for noncoherent impulse-radio-based UWB (IR-UWB) communication system are proposed to provide system implementation of low power consumption and low complexity. The proposed system adopts a simple modulation technique of OOK (on-oft-keying) and noncoherent signal detection based on signal amplitude. In particular, a technique of a novel high speed digital sampler using a stable, lower reference clock is developed to detect nano-second pulses and recover digital signals from the pulses. Also, a 32 bits Turyn code for data frame synchronization and a convolution code as FEC are applied, respectively. To verify the proposed signal processing techniques for low power, low complexity noncoherent IR-UWB system, the proposed signal processing technique is implemented in FPGA and then a short-range communication system for wireless transmission of high quality MP3 data is designed and tested.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.

• Journal of KIISE:Computer Systems and Theory
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• v.36 no.2
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• pp.102-110
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• 2009

The memory wall is the growing disparity of speed between CPU and memory outside the CPU chip. An economical solution is a memory hierarchy organized into several levels, such as processor registers, cache, main memory, disk storage. We introduce a novel memory hierarchy optimization technique in Linux based embedded systems using on-chip SRAM for the first time. The optimization technique allocates On-Chip SRAM to the code/data that selected by programmers by using virtual memory systems. Experiments performed with nine applications indicate that the runtime improvements can be achieved by up to 35%, with an average of 14%, and the energy consumption can be reduced by up to 40%, with an average of 15%.

• Journal of the Korea Society for Simulation
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• v.19 no.1
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• pp.53-61
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• 2010

In wireless sensor networks(WSNs) individual sensor nodes are subject to security compromises. An adversary can physically capture sensor nodes and obtain the security information. And the adversary injects false reports into the network using compromised nodes. If undetected, these false reports are forwarded to the base station. False reports injection attacks can not only result in false alarms but also depletion of the limited amount of energy in battery powered sensor nodes. To combat these false reports injection attacks, several filtering schemes have been proposed. The statistical en-routing filtering(SEF) scheme can detect and drop false reports during the forwarding process. In SEF, The number of the message authentication codes(threshold) is important for detecting false reports and saving energy. In this paper, we propose a dynamic threshold determination method for energy efficient SEF using fuzzy-logic in wireless sensor networks. The proposed method consider false reports rate and the number of compromised partitions. If low rate of false reports in the networks, the threshold should low. If high rate of false reports in networks, the threshold should high. We evaluated the proposed method’s performance via simulation.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Journal of Energy Engineering
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• v.27 no.2
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• pp.14-25
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• 2018

The combined cycle power plant has a cycle of operating the gas turbine with fuel, such as natural gas, and then producing steam using residual heat. The fuel gas is supplied to the gas turbine at a level of 4 to 5 MPa, $200^{\circ}C$ through a compressor and a heat exchanger. In this study, the risk assessment method considering the piping system stress was carried out for safe operation and soundness of the gas fuel supply piping system. The API 580/581 RBI code, which is well known for its risk assessment techniques, is limited to reflect the effect of piping stress on risk. Therefore, the systematic stress of the pipeline is analyzed by using the piping analysis. For the study, the piping system stress analysis was performed using design data of a gas fuel supply piping of a combined cycle power plant. The result of probability of failure evaluated by the API code is compared to the result of stress ratio by piping analysis.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.256-262
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• 2014

Cycle length of domestic nuclear power plants is determined by the demand-supply plan of utility company. The target cycle length is achieved by adjusting the number of feed fuel assembly and fuel enrichment. Traditionally, utility company first select the number of feed fuel assembly and then find out the fuel enrichment to achieve the special cycle length. But it is difficult to find out if this method is most economical than any other combinations of the enrichment and batch size satisfying the same cycle length. In this paper, core depletion calculation is performed to find out the optimum combination of the enrichment and batch size for given target cycle length in terms of fuel cycle cost using commercial core design code; CASMO/MASTER code. To minimize the uncertainty resulting from transition core analysis, levelized fuel cycle cost analysis was applied to the equilibrium cycle core in order to determine the optimum combination. The sensitivity study of discount rate was also carried out to analyze the levelized fuel cycle cost applicable to countries with different discount rates. From the levelized fuel cycle cost analysis results, the combination with smaller batch size and higher fuel enrichment becomes more economical as the discount rate becomes lower. On the other hand, the combination with higher batch size and lower fuel enrichment becomes more economical as the discount rate becomes higher.

• Journal of Energy Engineering
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• v.26 no.3
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• pp.64-70
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• 2017

The concept of Hybrid Safety Injection Tank (Hybrid SIT) proposed by the Korea Atomic Energy Research Institute (KAERI) has been introduced for the purpose of application to the Advanced Power Reactor Plus (APR+). In this study, the SBO situation of the APR+ was analyzed by using the MARS-KS code in order to evaluate whether the operation of the Hybrid SIT has an effect on the cooling performance of the Reactor Coolant System (RCS). According to the analysis, when the actuation valve on the pressure balancing line (PBL) is opened, the Hybrid SIT's pressure rises rapidly, forming equilibrium with the RCS pressure; subsequently, a flow is injected from the Hybrid SIT into the reactor vessel through the direct vessel injection (DVI) line. The analysis showed that it is possible to keep the core temperature below melting temperature during the operation of a Hybrid SIT.

• Journal of Energy Engineering
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• v.29 no.1
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• pp.63-74
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled for decommissioning after permanent shutdown, is expected to generate a large amount of various types of radioactive waste during decommissioning process. For concrete radioactive waste, which is expected to occupy the most amount, it is important to analyze the current waste disposal status and legal limitations and to prepare an appropriate and efficient disposal method. Concrete radioactive waste is waste of various levels, of which the clearance level is bioshield concrete. In this paper, clearance radioactive waste safety evaluation was performed using the RESRAD code, which is a safety evaluation code, based on the activation evaluation results for the wastes with the clearance level. The clearance scenario of the target radioactive waste was selected and the individual's exposure dose was calculated at the time of clearance to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. As a result of the evaluation, the results showed significantly lower results and satisfied the criteria value. Based on the results of this clearance safety assessment, the appropriate disposal method for bioshield concrete, which are the clearance wastes of subject of deregulation, was suggested.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.55-60
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• 2005

The evaluation of neutron flux distribution was performed for the ex-core detector design of SMART-P. DORT and MCNP code were used for the calculation of energy-dependent neutron flux distribution at 100% full power condition. Two code results show that maximum thermal flux appears at the $1^{st}$ water region in IST region and agree within 10% difference. In addition, another evaluation was performed code with assumptions that cote was composed of fission source and control rod without fuel assemblies. These assumptions make neutron count rate to be minimized. As a results, maximum thermal flux showed $6.99{\times}10^{-2}(n/cm^2-sec)$, when the strength of initial fission source was assumed as $1.0{\times}10^8(n/sec)$. The main reason of these results is due to the thermalization of fast neutrons in the water region and thermal flux is proportional to 80% of total neutron flux. Therefore, optimization of filler material of detector guide tube, position of installation and axial length of detector segments is necessary for the design of ex-core detector to enhance the neutron count rate and above results could be used in ex-core detector design as a fluence requirement.