• Title/Summary/Keyword: Station Blackout

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Bayes Estimate for the Reliability of Nuclear-Power Plant Emergency Diesel Generator (비상디젤발전기 신뢰도에 대한 베이즈추정)

  • 심규박;류부형
    • Journal of Korean Society for Quality Management
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    • v.25 no.3
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    • pp.108-118
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    • 1997
  • A commercial nuclear power station contains at least two emergency diesel generates(EDG) to control the risk of severe core demage during the station blackout accidents. Therefore the reliability of the EDG's to start and load-run on demend must be maintained at a sufficiently high level. Until now, a simple assessment of start and load-run success rates was used to calculate the EDG's reliability. However, this method has been found to contain many defects. Recently, the work of Martz et al.(1996) proposed the use of the Bayes estimator to find the EDG's reliability. In this paper, we will propose confidence interval for the Bayes estimator, compare the above two methods and, using practical examples, illustrate why the Bayes estimator method is more reasonable in our situation.

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A Study on the Application of Composite Reliability to Estimate the EDG Reliability

  • Shim, Kyu-Bark
    • Journal of Korean Society for Quality Management
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    • v.26 no.4
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    • pp.265-276
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    • 1998
  • A commercial nuclear power station contains at least two emergency diesel generators(EDG) to control the risk of severe core damage during station blackout accidnets. Therefore, thereliability of the EDG's to start and load-run on demand must be maintained at a sufficiently high level. Until now, a simple assessment of start and load-run success rates was used to calculate the EDG reliability. However, this method has been found to contain many defects. Recently, the work of Martz et al.(1996) proposed the use of the Bayes estimator to find EDG reliability. Shim(1996) proposed a confidence interval for the Bayes estimator, compare the above two methods. In this paper, we introduce the notion of "Composite Reliablility" to estimate the reliability of nuclear-power plant EDG, and using practical examples, illustrate which method is more a, pp.opriate in our situation.situation.

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Probabilistic Safety Assessment of Nuclear Power Plants Using Bayes Method

  • Shim, Kyu-Bark
    • Communications for Statistical Applications and Methods
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    • v.8 no.2
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    • pp.453-464
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    • 2001
  • A commercial nuclear power station contains at least tow emergency diesel generators(EDG) to control the risk of severe core damage during station blackout accidents. Therefore, the reliability of the EDG's to start and load-run on demand must be maintained at a sufficiently high level. Probabilistic safety assessments(PSA) are increasingly being used to quantify the public risk of operating potentially hazardous systems such as nuclear power reactors. In this paper, to perform PSA, we will introduce three different types of data and use Bayes procedure to estimate the error rate of nuclear power plant EDG, and using practical examples, illustrate which method is more reasonable in our situation.

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A SE Approach to Predict the Peak Cladding Temperature using Artificial Neural Network

  • ALAtawneh, Osama Sharif;Diab, Aya
    • Journal of the Korean Society of Systems Engineering
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    • v.16 no.2
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    • pp.67-77
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    • 2020
  • Traditionally nuclear thermal hydraulic and nuclear safety has relied on numerical simulations to predict the system response of a nuclear power plant either under normal operation or accident condition. However, this approach may sometimes be rather time consuming particularly for design and optimization problems. To expedite the decision-making process data-driven models can be used to deduce the statistical relationships between inputs and outputs rather than solving physics-based models. Compared to the traditional approach, data driven models can provide a fast and cost-effective framework to predict the behavior of highly complex and non-linear systems where otherwise great computational efforts would be required. The objective of this work is to develop an AI algorithm to predict the peak fuel cladding temperature as a metric for the successful implementation of FLEX strategies under extended station black out. To achieve this, the model requires to be conditioned using pre-existing database created using the thermal-hydraulic analysis code, MARS-KS. In the development stage, the model hyper-parameters are tuned and optimized using the talos tool.

EVALUATION OF PLANT OPERATIONAL STATES WITH THE CONSIDERATION OF LOOP STRUCTURES UNDER ACCIDENT CONDITIONS

  • MATSUOKA, TAKESHI
    • Nuclear Engineering and Technology
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    • v.47 no.2
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    • pp.157-164
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    • 2015
  • Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

Review Criteria for Reliability from Analysis of LOOP frequency in NPPs (소외전원상실사고 빈도수 분석을 통한 원전 신뢰도 검토기준)

  • Moon, Su-Cheol;Kim, Kern-Joong
    • The Transactions of The Korean Institute of Electrical Engineers
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    • v.62 no.3
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    • pp.300-305
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    • 2013
  • LOOP(Loss of Offsite Power) and SBO(Station Blackout) events have been occurring in nuclear power plants should be reviewed and be controlled on important electrical equipments by professional engineer to prevent and to safety improvement from safety assessment and reliability analysis report. LOOP and SBO occasionally happened by internal or external causes. This paper contained that LOOP frequency in the United States NPPs and in the domestic NPPs have compared and analyzed data by the past lessons and probabilistic statistics. Additionally will be installed MG(Mobile Generator) according to the lessons of Fukushima nuclear accident in Japan, which CDF(Core Damage Frequency) and LOOP frequency have reconsidered. And this paper proposed to reduce reliability criteria using PSA(Probabilistic Safety Analysis).

Assessing the Feasibility of an Accident Management Strategy Using Dynamic Reliability Methods

  • Moosung Jae;Kim, Jae-Hwan
    • Nuclear Engineering and Technology
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    • v.29 no.1
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    • pp.1-6
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    • 1997
  • This paper presents a new dynamic approach for assessing feasibility associated with the implementation of accident management strategies by the operators. This approach includes the combined use of both the concept of reliability physics and a dynamic event tree generation scheme. The reliability physics is based on the concept of a comparison between two competing variables, i.e., the requirement and the achievement parameter, while the dynamic event tree generation scheme on the continuous generation of the possible event sequences at every branch point up to the desired solution. This approach is applied to a cavity flooding strategy in a reference plant, which is to supply water into the reactor cavity using emergency fire systems in the station blackout sequence. The MAAP code and Latin Hypercube sampling technique are used to determine the uncertainty of the requirement parameter. It has been demonstrated that this combined methodology may contribute to assessing the success likelihood of the operator actions required during accidents and therefore to developing the accident management procedures.

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A New Dynamic HRA Method and Its Application

  • Jae, Moosung
    • International Journal of Reliability and Applications
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    • v.2 no.1
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    • pp.37-48
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    • 2001
  • This paper presents a new dynamic human reliability analysis method and its application for quantifying the human error probabilities in implementing management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequency used method in PSAs, are discussed. The action associated with implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic obtained, human error probabilities are calculated with respect to the various means and variances of the things. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

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Use of Dynamic Reliability Method in Assessing Accident Management Strategy

  • Jae, Moosung
    • International Journal of Reliability and Applications
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    • v.2 no.1
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    • pp.27-36
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    • 2001
  • This Paper proposes a new methodology for assessing the reliability of an accident management, which Is based on the reliability physics and the scheme to generate dynamic event tree. The methodology consists of 3 main steps: screening; uncertainty propagation; and probability estimation. Sensitivity analysis is used for screening the variables of significance. Latin Hypercube sampling technique and MAAP code are used for uncertainty propagation, and the dynamic event tree generation method is used for the estimation of non-success probability of implementing an accident management strategy. This approach is applied in assessing the non-success probability of implementing a cavity flooding strategy, which is to supply water into the reactor cavity using emergency fire systems during the sequence of station blackout at the reference plant.

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ESTABLISHMENT OF A SEVERE ACCIDENT MITIGATION STRATEGY FOR AN SBO AT WOLSONG UNIT 1 NUCLEAR POWER PLANT

  • Kim, Sungmin;Kim, Dongha
    • Nuclear Engineering and Technology
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    • v.45 no.4
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    • pp.459-468
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    • 2013
  • During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.