Assessment of FBR MONJU Accident Management Reliability in Causing Reactor Trips

Abstract

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.

Uncertainty Evaluation of Anticipated Transient without Scram Plant Response in the Monju Reactor Considering Reactivity Coefficients within the Design Range

“…Before reaching 750 °C, the probability of cavitation in the static pressure bearing begins to increase at 650 °C [1]. In other words, it is considered that core damage due to the ULOHS event would be avoided if the PHTS pump inlet sodium temperature stays below 650 °C for a sufficiently long time in terms of reactor operation, 1 h for example.…”

“…In the ULOHS event, the reactivity feedback behavior and the heat transfer behavior of from the fuel pins to the coolant have been evaluated as important phenomena in the PIRT. However, the sensitivity of the heat transfer at a low flow rate has been evaluated to be insignificant [1], and thus the variability is considered only for the reactivity feedback behavior in this paper.…”

“…Among ATWS events, an unprotected loss of heat sink (ULOHS) will have a relatively large contribution to the core damage frequency. This kind of event has significant allowable time until core damage is caused because the core coolant flow can be maintained [1]. Fig.…”

“…In-vessel retention of the molten corium after core damage may not always be possible because of the high temperature of the coolant that results from the loss of the heat sink. Therefore, the uncertainty in the allowable time to core damage in this event has so far been estimated by considering the range of reactivity coefficients [1].…”