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The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code (SBC) by considering the safety point of view. This new method utilizes analysis models of a probabilistic safety assessment (PSA), and the reliability target is derived from (1) the proposal to a quantitative safety goal that was published by the nuclear safety commission (NSC) of Japan; (2) the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF) that were determined in the Fast Reactor Cycle Technology Development (FaCT) project by Japan Atomic Energy Agency (JAEA). This method was applied to determine the reliability target of the structures and components which constitute the reactor cooling system in the Japan sodium-cooled fast reactor (JSFR). The risk from the reactor is shown by the sum of combination of various elements in the PSA analysis model. Those elements include dynamic failures and static failures of the structures and components, and human errors. However, the present study focuses on the sequences including the static failure, and the probability of dynamic failures and human errors in those sequences is conservatively assumed as a unity. It was confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determine the reliability target associated with a random failure of the structures and components, and also confirmed that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.

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