• Title/Summary/Keyword: nuclear power station

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Development of An Integrated Test Facility (ITF) for the Advanced Man Machine Interface Evaluation

  • Oh, In-Seok;Cha, Kyung-Ho;Lee, Hyun-Chul;Sim, Bong-Sick
    • Proceedings of the Korean Nuclear Society Conference
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    • 1995.10a
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    • pp.117-122
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    • 1995
  • An Integrated Test Facility(ITF) is a human factors experimental environment to evaluate an advanced man machine interface(MMI) design. The ITF includes a human machine simulator(HMS) comprised of a nuclear power plant function simulator, man-machine interface, experiment control station for the experiment control and design, human behavioural data measurement system, and data analysis and experiment evaluation supporting system(DAEXESS). The most important features of ITF is to secure the flexibility and expandibility of Man Machine Interlace(MMI) design to change easily the environment of experiments to accomplish the experiment's objects In this paper, we describe a development scope and characteristics of the ITF such as, hardware and software development scope and characteristics, system thermohydraulic modelling characteristics, and experiment station characteristics for the experiment variables design and control, to be used as an experiment environment for the evaluation of VDU-based control room.

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EVALUATION OF AN ACCIDENT MANAGEMENT STRATEGY OF EMERGENCY WATER INJECTION USING FIRE ENGINES IN A TYPICAL PRESSURIZED WATER REACTOR

  • PARK, SOO-YONG;AHN, KWANG-IL
    • Nuclear Engineering and Technology
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    • v.47 no.6
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    • pp.719-728
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    • 2015
  • Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

Effect of multiple-failure events on accident management strategy for CANDU-6 reactors

  • YU, Seon Oh;KIM, Manwoong
    • Nuclear Engineering and Technology
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    • v.53 no.10
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    • pp.3236-3246
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    • 2021
  • Lessons learned from the Fukushima Daiichi nuclear power plant accident directed that multiple failures should be considered more seriously rather than single failure in the licensing bases and safety cases because attempts to take accident management measures could be unsuccessful under the high radiation environment aggravated by multiple failures, such as complete loss of electric power, uncontrollable loss of coolant inventory, failure of essential safety function recovery. In the case of the complete loss of electric power called station blackout (SBO), if there is no mitigation action for recovering safety functions, the reactor core would be overheated, and severe fuel damage could be anticipated due to the failure of the active heat sink. In such a transient condition at CANDU-6 plants, the seal failure of the primary heat transport (PHT) pumps can facilitate a consequent increase in the fuel sheath temperature and eventually lead to degradation of the fuel integrity. Therefore, it is necessary to specify the regulatory guidelines for multiple failures on a licensing basis so that licensees should prepare the accident management measures to prevent or mitigate accident conditions. In order to explore the efficiency of implementing accident management strategies for CANDU-6 plants, this study proposed a realistic accident analysis approach on the SBO transient with multiple-failure sequences such as seal failure of PHT pumps without operator's recovery actions. In this regard, a comparative study for two PHT pump seal failure modes with and without coolant seal leakage was conducted using a best-estimate code to precisely investigate the behaviors of thermal-hydraulic parameters during transient conditions. Moreover, a sensitivity analysis for different PHT pump seal leakage rates was also carried out to examine the effect of leakage rate on the system responses. This study is expected to provide the technical bases to the accident management strategy for unmitigated transient conditions with multiple failures.

A Comparative Study on Mitigation Alternatives in Response to an Extended SBO for APR1400 Using Systems Engineering (확장된 소내전원 상실 사고시의 대체대응활동 완화를 위한 비교 연구: 시스템 엔지니어링 관점으로)

  • Elaswakh, Islam Sabry;Oh, SJ;Lim, Hak-Kyu
    • Journal of the Korean Society of Systems Engineering
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    • v.12 no.2
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    • pp.91-99
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    • 2016
  • The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

Identification of hydrogen flammability in steam generator compartment of OPR1000 using MELCOR and CFX codes

  • Jeon, Joongoo;Kim, Yeon Soo;Choi, Wonjun;Kim, Sung Joong
    • Nuclear Engineering and Technology
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    • v.51 no.8
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    • pp.1939-1950
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    • 2019
  • The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

Open-Phase Condition Detecting System for Transformer Connected Power Line in Nuclear Power Plant (원자력발전소 변압기 연결 선로 결상 검출 시스템)

  • Ha, Che-Wung;Lee, Do-Hwan
    • The Transactions of The Korean Institute of Electrical Engineers
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    • v.64 no.2
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    • pp.254-259
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    • 2015
  • On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

Development of a Leading Performance Indicator from Operational Experience and Resilience in a Nuclear Power Plant

  • Nelson, Pamela F.;Martin-Del-Campo, Cecilia;Hallbert, Bruce;Mosleh, Ali
    • Nuclear Engineering and Technology
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    • v.48 no.1
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    • pp.114-128
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    • 2016
  • The development of operational performance indicators is of utmost importance for nuclear power plants, since they measure, track, and trend plant operation. Leading indicators are ideal for reducing the likelihood of consequential events. This paper describes the operational data analysis of the information contained in the Corrective Action Program. The methodology considers human error and organizational factors because of their large contribution to consequential events. The results include a tool developed from the data to be used for the identification, prediction, and reduction of the likelihood of significant consequential events. This tool is based on the resilience curve that was built from the plant's operational data. The stress is described by the number of unresolved condition reports. The strain is represented by the number of preventive maintenance tasks and other periodic work activities (i.e., baseline activities), as well as, closing open corrective actions assigned to different departments to resolve the condition reports (i.e., corrective action workload). Beyond the identified resilience threshold, the stress exceeds the station's ability to operate successfully and there is an increased likelihood that a consequential event will occur. A performance indicator is proposed to reduce the likelihood of consequential events at nuclear power plants.

Calculation of fuel temperature profile for heavy water moderated natural uranium oxide fuel using two gas mixture conductance model for noble gas Helium and Xenon

  • Jha, Alok;Gupta, Anurag;Das, Rajarshi;Paraswar, Shantanu D.
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2760-2770
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    • 2020
  • A model for calculation of fuel temperature profile using binary gas mixture of Helium and Xenon for gap gas conductance is proposed here. In this model, the temperature profile of a fuel pencil from fuel centreline to fuel surface has been calculated by taking into account the dilution of Helium gas filled during fuel manufacturing due to accumulation of fission gas Xenon. In this model an explicit calculation of gap gas conductance of binary gas mixture of Helium and Xenon has been carried out. A computer code Fuel Characteristics Calculator (FCCAL) is developed for the model. The phenomena modelled by FCCAL takes into account heat conduction through the fuel pellet, heat transfer from pellet surface to the cladding through the gap gas and heat transfer from cladding to coolant. The binary noble gas mixture model used in FCCAL is an improvement over the parametric model of Lassmann and Pazdera. The results obtained from the code FCCAL is used for fuel temperature calculation in 3-D neutron diffusion solver for the coolant outlet temperature of the core at steady operation at full power. It is found that there is an improvement in calculation time without compromising accuracy with FCCAL.

PMCR-A Power Mapping and Calibration Routing for 600 MWe CANDU-PHW Reactors

  • Oh, Se-Ki;G.Kugler
    • Nuclear Engineering and Technology
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    • v.11 no.4
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    • pp.275-286
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    • 1979
  • In 600 MWe CANDU-PHW reactors PMCR will be used on-site for calibration of the regional overpower system. PMCR will be executed off-line in one of the station computers. The program calculates accurate channel power maps by incorporating a fuel turnup dependent flux to power conversion algorithm. Fuel turnup is calculated by PMCR, hence it is independent of other software. Extensive comparisons with the uniform flux/power conversion approximations were made. Significant improvements in power mapping accuracy are achieved with PMCR.

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Diagnostics of nuclear reactor coolant pump in transition process on performance and vortex dynamics under station blackout accident

  • Ye, Daoxing;Lai, Xide;Luo, Yimin;Liu, Anlin
    • Nuclear Engineering and Technology
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    • v.52 no.10
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    • pp.2183-2195
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    • 2020
  • A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.