Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

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概要

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• This paper describes a method and application of quantitatively evaluating Accident Management (AM) reliability upon a reactor trip failure for the MONJU fast breeder reactor using a PSA technique. The present method comprises an allowable time estimation that is based on plant transient response analysis using the Super-COPD code that was developed for use in best estimates of the plant dynamics of MONJU and in estimating failure probability of operator’s actions in AMs within the allowable time based on time records obtained from simulator training. Application of this method to MONJU resulted in the estimation that the allowable time for an unprotected loss-of-heat sink event would be more than the longest observed time of 326 s. The corresponding operation failure probability would be less than 0.1 even after taking the uncertainty into consideration. Combining this with a level 1 PSA revealed that the total frequency of core damage accompanying a reactor trip failure at MONJU could be decreased by at least 50 percent due to the reactor trip AM.

• 一般社団法人 日本原子力学会の論文
• 2010-10-01

著者

• Kurisaka Kenichi

  Safety Evaluation Group Fbr System Engineering Unit Advanced Nuclear System Research And Development

• SOTSU Masutake

  Plant Technology Assessment Group, FBR Plant Engineering Center, Japan Atomic Energy Agency

• Sotsu Masutake

  Plant Technology Assessment Group Fbr Plant Engineering Center Japan Atomic Energy Agency

• SOTSU Masutake

  Plant Dynamics Analysis Group, FBR Plant Engineering Center, Japan Atomic Energy Agency

関連論文

• Evaluation of MONJU Core Damage Risk with Change of AOT Using Probabilistic Method
• Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips
• Evaluation of MONJU Core Damage Risk due to Control Rod Function Failure

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