Level 1 probabilistic safety assessment to support the design of the CEA 2400 MWth gas-cooled fast reactor

作者：Bassi C^*; Azria Ph; Balmain M

来源：Nuclear Engineering and Design, 2010, 240(11): 3758-3780.

DOI：10.1016/j.nucengdes.2010.09.003

As part of the design of a 4th generation reactor, the integration of safety in the early phase of the concepts is expected. To date, probabilistic insights are increasingly employed in the safety demonstration in combination with the deterministic approach (e.g. to identify the sequences of complex failures, to justify the categorization of situations) and used, even at an early stage of design, to identify the reliability of systems and equipment to handle the safety objectives (expressed in terms of core damage frequency targets). Within this frame, the CEA has undertaken to assess the benefit of developing a probabilistic model to support the design of the 2400 MWth gas-cooled fast reactor.
In the building process of this level 1 probabilistic safety assessment, a first phase consisted in making a preliminary model that took only into account families of initiating events that were defined for the design of the decay heat removal dedicated loops, namely the loss of coolant accidents (representative of medium pressure situations) and loss of off-site power/station black-out transients (representative of high pressure situations). Owing to the results obtained with this preliminary L1 PSA model, it emerged that an increased reliability of the DHR function in high pressure conditions (i.e. characterized by IEs not associated to the loss of integrity of the helium pressure boundary) is suitable to reduce the overall core damage frequency. The track was therefore chosen to require the use of normal loops as first line of provision of the DHR function, possibly including components or particular operating modes related to the secondary and tertiary circuits. In addition, this final L1 PSA model is characterized by success criteria based on transient calculations performed with the CATHARE2 code and to a perimeter extended to all representative internal IEs at full operating power.
This paper presents the building process and the main results related to these two successive L1 PSA models. Finally, useful insights are translated into GFR design improvements that are leading to an overall CDF at full operating power that satisfies nowadays with the probabilistic target defined for 3rd generation reactors, being at least the objective for 4th generation reactors.

• 出版日期2010-11
• 单位中国地震局