• Journal of KIISE:Software and Applications
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• v.29 no.12
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• pp.860-872
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• 2002

Fault tree analysis is the most widely used saftly analysis technique in industry. However, the analysis is often applied manually, and there is no systematic and automated approach available to validate the analysis result. In this paper, we demonstrate that a real-time model checker UPPAAL is useful in formally specifying the required behavior of safety-critical software and to validate the accuracy of manually constructed fault trees. Functional requirements for emergency shutdown software for a nuclear power plant, named Wolsung SDS2, are used as an example. Fault trees were initially developed by a group of graduate students who possess detailed knowledge of Wolsung SDS2 and are familiar with safety analysis techniques including fault tree analysis. Functional requirements were manually translated in timed automata format accepted by UPPAAL, and the model checking was applied using property specifications to evaluate the correctness of the fault trees. Our application demonstrated that UPPAAL was able to detect subtle flaws or ambiguities present in fault trees. Therefore, we conclude that the proposed approach is useful in augmenting fault tree analysis.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.291-295
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• 2007

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL consists of In-Pile Test Section (IPS) and Out-Pile System (OPS). FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. Task Force Team was organized to do a construction systematically and the communication between members of the task force team was done through the CoP(community of Practice) notice board provided by the Institute. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. Without a sweet of the workers of the participating company of HEC(Hyundae Engineering Co, Ltd), HDEC(HyunDai Engineering & Construction Co. Ltd), equipment manufacturer, and the task force team, it is not possible to install the FTL facility within the planned shutdown period. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

• Journal of KIISE:Computer Systems and Theory
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• v.34 no.1
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• pp.10-18
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• 2007

Mobile communication devices such as PDAs, cellular phones, etc., need to perform several kinds of computation-intensive functions including H.264 encoding/decoding and 3D graphics processing. In this paper, new reconfigurable architecture is described, which can perform either motion estimation for H.264 or rendering for 3D graphics. The proposed motion estimation techniques use new efficient SAD computation ordering, DAU, and FDVS algorithms. The new approach can reduce the computation by 70% on the average than that of JM 8.2, without affecting the quality. In 3D rendering, midline traversal algorithm is used for parallel processing to increase throughput. Memories are partitioned into 8 blocks so that 2.4Mbits (47%) of memory is shared and selective power shutdown is possible during motion estimation and 3D graphics rendering. Processing elements are also shared to further reduce the chip area by 7%.

• Journal of Mechanical Science and Technology
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• v.15 no.3
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• pp.291-299
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• 2001

Recently, severe wear on the shutdown rod cladding of Ulchin Nuclear Power Plant #1, #2 were observed by the Eddy Current Test(E.C.T.). In particular, the wear at the sixth card location was up to 75%. The test results indicated that the Flow Induced Vibration(F.I.V.) might be the cause of the fretting wear resulting from the contact between Rod Cluster Control Assemblies(RCCAs) and their spacing cards(guide plates) arranged in the guide tube. From reviewing RCCAs fretting wear repots and analyzing the general characteristics of F.I.V. mechanism in the reactor, geometric layout and flow conditions around the control rod, it is concluded that the turbulence excitation is the most probable vibration mechanism of RCCA. To identify the governing mechanism of RCCA vibration, an experiment was performed for a representative rod position in which the most serious fretting wear experienced among the six rod positions. The experimental rig was designed and set up to satisfy the governing nondimensional numbers which are Reynolds number and mass damping parameter. The vibration amplitude measurement by the non-contact laser displacement sensor showed good agreements in the frequency and the maximum wearing(vibration) location with Ulchin E.C.T. results and Framatome report, respectively. The sudden increase in the vibration amplitude was sensed around the 6th guide plate with mass flow rate variation. Comparing the similitude rod behaviour with the idealized response of a cylinder in flow induced vibration, it was found that he dominant mechanism of vibration was transferred from turbulence excitation to periodic shedding at the mass flow ate 90ι/min. Also the critical velocity of the vibration in RCCAs was determined and the vibration can be prevented by reducing the bypass flow rate below the critical velocity.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.354-358
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• 2016

Background: To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and Discussion: It was found that the leak rate is $0.1%{\cdot}day^{-1}$ of air, $0.004%{\cdot}day^{-1}$ of noble gas and $3.7{\times}10^{-5}%{\cdot}day^{-1}$ of aerosol during typhoon passing. The air leak rate of $0.1%{\cdot}day^{-1}$ can be converted into $1.36m^3{\cdot}hr^{-1}$, but the design leak rate in HANARO safety analysis report was considered as $600m^3{\cdot}hr^{-1}$ under the condition of $20m{\cdot}sec^{-1}$ wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

• Journal of IKEEE
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• v.15 no.4
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• pp.319-323
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• 2011

In this paper presents a new model of dual-mode synchronous buck converter with dynamic control for mobile applications was proposed. The proposed circuit can operate at 2.5MHz with supply voltage 2.5V to 5V for low ripple and minimum inductor and capacitor size, which is suitable for single-cell lithium-ion battery supply mobile applications. For high efficiency, the proposed circuit adopts synchronous type and dynamic control. The proposed circuit is designed by using the device parameter of TSMC 0.18um BCD process and the performance is evaluated by Cadence spectre. Experimental board level results show the maximum conversion efficiency is 96% at 100mA load current.

• Journal of KIISE:Computing Practices and Letters
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• v.8 no.1
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• pp.46-61
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• 2002

Among the many phases involved in software development, requirements analysis phase in generally considered to play a crucial role in determining the overall software quality. Therefore, many software development companies manages the phase as one of the important phase. Especially, safety assurance through requirements analysis for safety-critical systems is quite demanding, and national and international bodies routinely require safety demonstration. Among various approaches, inspection and formal methods are generally shown to be effective. In this paper, we propose a formal verification procedure for SCR(Software Cost Reduction)-style SRS(Software Requirements Specification) using the PVS specification and verification procedure and applied this procedure to an industrial system such that a shutdown system for Wolsung nuclear power plant. This system had been verified through inspection not formal verification. The application of formal methods is rare in Korea, so it is very important to experiment about formal verification to industrial systems.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.11
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• pp.907-914
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• 2014

The objective of this study was to increase the generating efficiency of concentrated photovoltaics (CPV) by using hybrid solar tracking. Further, the proposed system was demonstrated to have the ability to extract thermal energy from a concentrated photovoltaic system by using thermal absorbers containing heat pipe, which could then be used for a heating system or hot-water supply. The average electrical efficiency was 16 during the day, and the average thermal efficiency was 62. Therefore, this system demonstrated a total efficiency (electrical thermal) of 78. All the processes, i.e., tracking of the sun, calculation of the sun's position, reinstatement of the heating device toward the east for tracking on the next day, and system shutdown, were programmed using Simulink. A parametric analysis of the heat pipe, concentration ratio, and inlet velocity was also performed in terms of the operating temperature of the CPV and the outlet temperature. The simulation and experimental results for the thermal absorber were found to be in good agreement.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.121-126
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• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.