• Journal of Energy Engineering
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• v.23 no.2
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• pp.191-198
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• 2014

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. MR programs are developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has developed the MR program for APR1400 safety related systems to establish the advanced maintenance system and will verify the suitability of the MR program through evaluating initial performance. Consequently, it is expected that the safety of the new plant will be improved by developing and implementing the MR program.

• Korean Journal of Medicinal Crop Science
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• v.18 no.5
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• pp.329-337
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• 2010

'Angelicae Pubescentis Radix' (APR) is an important oriental medical preparation. In Korea, Aralia continentalis has been recognized as the source plant of APR. Aralia cordata, which is difficult to distinguish from A. continentalis, and Heracleum moellendorffii, which is frequently used in lieu of A. continentalis, are traded in Korean herbal markets. In contrast, in China, Angelica pubescens is recognized as the source plant of APR. In this study, we devised a method not only to discriminate A. contientalis from A. cordata, but also to discriminate both A. contientalis and A. cordata from H. moellendorffii and A. pubescens. Based on the discrepancy in the sequences of specific regions of ITS, we designed a Cont F/ Cont R primer set to amplify a 173 bp PCR band that appears only in A. continentalis. Additionally, we designed an Ara F/ Ara R primer set to amplify a 278 bp PCR band that appears in both A. continentalis and A. cordata. Using these primer sets and the ST R primer to confirm the PCR amplification results, we developed a simple multiplex PCR method for differentiating A. continentalis from A. cordata and to concurrently differentiate both A. continentalis and A. cordata from other APR herbs.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.1
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• pp.97-102
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• 2012

Alternate alternating current (AAC) is used in nuclear power plants (NPPs) in order to cope with station black outs (SBOs). AAC has been provided using diesel engine drive types in Korea's NPPs. The structure of gas turbine generators (GTGs) is simpler than that of diesel generators (DGs), and GTGs have the advantage of longer maintenance intervals. However, GTG-AAC was not used in NPPs in Korea because of the lack of operation/maintenance experience. The purpose of this paper is to analyze the safety of APR+ considering a diversity of AAC types. This paper analyzes reliability data, mechanical specifications of DGs and GTGs, and the sensitivity of core damage frequency to the ACC type.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2010.10a
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• pp.433-438
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• 2010

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. Nuclear Regulatory Commission Regulatory Guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement, and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors. We analyzed RVI design data of Palo Verde nuclear generating station(U.S.) and Yonggwang nuclear generating station(Korea) and investigated measuring sensors used in both of them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected the most suitable sensors for RVI CVAP in Advanced Power Reactor 1400(APR1400).

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.12
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.343-347
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• 2001

Advanced Pressurized Reactor 1400(APR-1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. The balance of plant (BOP) for the secondary system consists of main steam, feedwater, condensate, turbine generator and auxiliary system. In this paper, we describe the major design features of secondary component, balance of plant configuration, and then the turbine cycle thermal performance evaluation using PEPSE code.

• KOREAN JOURNAL OF CROP SCIENCE
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• v.42 no.4
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• pp.424-428
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• 1997

This study was carried out to investigate the effects of sowing dates, Apr. 25, May 15, June 5 and June 25, on sex ratio, growth and seed yield of "Chonnam species", local variety of hemp plant. The results were following as. : In sex ratio according to the different seeding dates, the revealating rate of female and male plants were highest on June 25, and Apr. 25, respectively. The length of stem, the number of branches, nodes, and seed yield were increased on May 15, but these charateristics were rapidly reduced on June 25 because of short growing period. Yield of seeds was increased by 24％ on May 15 comparing with 1.30ton /ha on Apr. 25.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.41-58
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• 2016

Reliability Centered Maintenance (RCM) introduces a systematic method and decision logic tree for utilizing previous operating experience focused on reliability and optimization of maintenance activities. In this paper RCM methodology is applied on safety injection system for APR-1400. Functional Failure Mode Effects and Criticality Analysis (FME&CA) are applied to evaluate the failure modes and the effect on the component, system and plant. Logic Tree Analysis (LTA) is used to determine the optimum maintenance tasks. The results show that increasing the condition based maintenance will reduce component failure and improve reliability and availability of the system. Also the extension of the surveillance test interval of Safety Injection Pumps (SIPs) would lead to an improved pump's availability, eliminate the unnecessary maintenance tasks and this will optimize maintenance activities.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.505-510
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• 2001

Advanced Pressurized Reactor 1400(APR1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. Due to the increased electric power, In Nuclear Power plant huge quantities of heat are generated in the thermo-dynamic process used for producing electrical energy. So, There is considerationly additional cooling, Heat transfer area and increased cooling water of Heat Exchanger which take care of the different smaller cooling duties within the nuclear power plant. We review applying to PRE instead of Shell-and-Tube Heat exchanger. In this paper, we describe the major design features of PRE, Comparison between a PHE and a Shell-and-Tube Heat Exchanger, and then Applicability of Plate Heat Exchanger in Nuclear Power Plant Component Cooling water systems.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.300-305
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• 2012

The vibration and stress analysis program of comprehensive vibration assessment program (CVAP) is to verify theoretically the structural integrity of reactor vessel internals (RVI) and to provide the basis for selecting the locations monitored in measurement and inspection programs. This paper covers the verification of the vibration and stress analysis methodology of APR1400 RVI CVAP. The analysis methodology was developed to use 3-dimensional hydraulic and structural models with ANSYS and CFX. To validate the methodology, the hydraulic loads and structural reponses of OPR1000 were predicted and compared with the calculated and measured data in the OPR1000 RVI CVAP. Since the results predicted with this methodology were close to the measured values considerably, it was confirmed that the analysis methodology was developed properly.