• Transactions of the Korean Society of Automotive Engineers
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• v.14 no.4
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• pp.1-11
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• 2006

The aim in this study is to develop the combined EGR system with a non-thermal plasma reactor for reducing exhaust emissions and improving fuel economy in turbo intercooler ECU common-rail diesel engines. In this study, the characteristics of soot, CO and $CO_2$ emissions under four kinds of engine loads are experimentally investigated by using a four-cycle, four-cylinder, direct injection type, water-cooled turbo intercooler ECU common-rail diesel engine with a combined plasma exhaust gas recirculation(EGR) system operating at three kinds of engine speeds. The EGR and non-thermal plasma reactor system are used to reduce $NO_x$ emissions, and the non-thermal plasma reactor and turbo intercooler system are used to reduce soot and THC emissions. The plasma system is a flat-to-flat type reactor operated by a plasma power supply. The fuel is sprayed by pilot and main injections at the variable injection timing between BTDC $15^{\circ}$ and ATDC $1^{\circ}$ according to experimental conditions. It is found that soot emissions with increasing EGR rate are increased, but are decreased as the applied electrical voltage of the non-thermal plasma reactor is elevated at the same engine speed and load. Results also show that CO and $CO_2$ emissions are increased as EGR rate is elevated, and CO emissions are increased, but $CO_2$ emissions are decreased as the applied electrical voltage of the non-thermal plasma reactor is elevated at the same engine speed and load.

• Transactions of the Korean Society of Automotive Engineers
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• v.14 no.3
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• pp.10-21
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• 2006

The aim in this study is to develop the combined EGR system with a non-thermal plasma reactor for reducing exhaust emissions and improving fuel economy in turbo intercooler ECU common-rail diesel engines. At the first step, in this paper, the characteristics of performance and $NO_x{\cdot}THC$ emissions under four kinds of engine loads are experimentally investigated by using a four-cycle, four-cylinder, direct injection type, water-cooled turbo intercooler ECU common-rail diesel engine with a combined plasma exhaust gas recirculation(EGR) system operating at three kinds of engine speeds. The EGR system is used to reduce $NO_x$ emissions, and the non-thermal plasma reactor and turbo intercooler system are used to reduce THC emissions. The plasma system is a flat-to-flat type reactor operated by a plasma power supply. The fuel is sprayed by pilot and main injections at the variable injection timing between BTDC $15^{\circ}$ and ATDC $1^{\circ}$ according to experimental conditions. It is found that the specific fuel consumption rate with EGR is increased, but the fuel economy is better than that of mechanical injection type diesel engine as compared with the same output. Results show that $NO_x$ emissions are decreased, but THC emissions are increased, as the EGR rate is elevated. $NO_x$ and THC emissions are also slightly decreased as the applied electrical voltage of the non-thermal plasma reactor is elevated. Thus one can conclude that the influence of EGR in $NO_x$ and THC emissions is larger than that of the non-thermal plasma reactor, but THC emissions are greatly influenced by the non-thermal plasma reactor as the EGR rate is elevated.

• Journal of the Korean Society of Safety
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• v.29 no.6
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• pp.166-171
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• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.167-168
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• 2012

Analysis on structural effects which are reduction of friction coefficient and increase of tendon area by HDPE greased and large size strand in post-tensioning system of reactor containment building was carried out. Effective ratio of tendon force increases 67% to 83% by HDPE greased strand and vertical, horizontal internal section forces increased maximum 51%, 41% respectively. Tendon quantity could be reduced 30% by large size and HDPE greased strand that can maintain safety of ultimate internal pressure same as at present.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.20-29
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• 1991

The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU longrightarrow FBR strategy is to be the best option for Korea. However, APWR, CANDU longrightarrow Passive Safe Reactor(PSR）longrightarrowFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.110-114
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• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.379-379
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• 2000

The robot systems used in nuclear power plants need to be both reliable and safe. As a part of the "Validation of nuclear safety-grade equipment" project, we established reliability analysis program and performed a number of analysis using conventional reliability analysis techniques. This paper describes the procedures, techniques, and results of the analysis utilized in our project. In addition, the paper includes current status of reliability analysis techniques and the summary of foreign case studiesse studies

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 1998.04a
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• pp.147-154
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• 1998

The purpose of this paper is to propose the draft guidelines of seismic isolation design of Liquid Metal Reactor (LMR) using high damping laminated rubber bearings. The scopes of guidelines include design requirements of a seismically isolated system and components, seismic isolator, isolation system, interface system between seismic isolation and non-seismic isolation part, qualification and acceptance tests of seismic isolator, seismic isolation reliability, and seismic safety and monitoring system. Proposed guidelines shall be revised to extend to general design guideline for nuclear facilities by further research and discussions.

• Proceedings of the KIEE Conference
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• 1999.07g
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• pp.3175-3177
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• 1999

In this study, the wireless communication system configuration and major equipments for KNGR will be introduced. Also the reliability and safety impact on the KNGR will be checked by reviewing /testing the EMI/RFI conditions of wireless communication system, The elements of the wireless communication system are radio exchange, base stations, portable telephone handsets.