Evaluation of Monju Core Damage Risk With Change of AOT Using Probabilistic Method

Sotsu, Masutake; Kurisaka, Kenichi

doi:10.1115/icone17-75263

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“…Sotsu and Kurisaka estimated the contribution of AOT to the CDF for the Japanese prototype fast breeder reactor MONJU [6] [7]. MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops that can supply 280 MW of electricity.…”

Section: Introductionmentioning

confidence: 99%

Evaluation of MONJU Core Damage Risk due to Control Rod Function Failure

Sotsu

Kurisaka

2012

JPES

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MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops that can supply 280 MW of electricity. The limiting conditions of operation defined in the safety regulations for MONJU given the allowed outage time were evaluated by a probabilistic safety assessment technique in our previous study. If a function failure is found in a control rod, certain measures are required by the safety regulations. In this case, if it is confirmed within 24 h that no other control rods are stuck, reactor operation is allowed to continue. To assess the validity of the 24 h allowable time in view of core damage risk, it is necessary to analyze the conditions to be changed when a stuck rod is discovered. Furthermore, to develop a method for evaluating the probability of a control rod insertion failure, it is necessary to re-estimate the frequency of core damage under control rod insertion failure conditions. This paper describes a method for this re-estimation. The probability of an insertion failure of one control rod has been calculated by applying Jeffrey's noninformative prior distribution by considering insertion times based on the results of a mock-up test. The necessary rod insertion numbers for the main reactor shutdown and backup shutdown systems were considered. The results showed that a completion time of 24 h gives a safety margin comparable to that of the Incremental Conditioned Core Damage Probability, that is, an acceptable risk threshold represented by the U.S. Nuclear Regulatory Commission RG 1.177. Thus, the timeframe defined in the present safety regulations was concluded to be appropriate.

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Section: Introductionmentioning

confidence: 99%