ارزیابی تاثیر خطر تغییرات الزامات، نظارت پرداختن به مدل و پارامتر عدم قطعیت

This paper presents a three steps based approach for the evaluation of risk impact of changes to Surveillance Requirements based on the use of the Probabilistic Risk Assessment and addressing identification, treatment and analysis of model and parameter uncertainties in an integrated manner. The paper includes also an example of application that focuses on the evaluation of the risk impact of a Surveillance Frequency change for the Reactor Protection System of a Nuclear Power Plant using a level 1 Probabilistic Risk Assessment. Surveillance Requirements are part of Technical Specifications that are included into the Licensing Basis for operation of Nuclear Power Plants. Surveillance Requirements aim at limiting risk of undetected downtimes of safety related equipment by imposing equipment operability checks, which consist of testing of equipment operational parameters with established Surveillance Frequency and Test Strategy.

Safe operation of Nuclear Power Plants (NPP) depends on Technical Specifications (TS), so that TS are part of the Licensing Basis (LB) to operate a NPP. They were established taking into account mainly deterministic criteria. The development of Probabilistic Risk Assessment (PRA) and its application since the early 80s to analyze TS changes has brought the opportunity to review TS consistency from a risk viewpoint, i.e. evaluation of risk impact of changes on plant safety on the basis of the risk information provided by the PRA. In particular, main attention has been paid to the role of the Surveillance Frequency (SF) as part of Surveillance Requirements (SR) and of the Completion Time (CT) as part of Limiting Conditions for Operation (LCO). In August 1995, the US Nuclear Regulatory Commission (NRC) adopted a final policy statement on the expanded use of PRA methods in nuclear activities that includes the following [1]. The use of the PRA technology and associated analyses (e.g. sensitivity studies, uncertainty analyses and importance measures) should be used in all regulatory matters to the extent supported by the state-of-art in PRA methods and data and in a manner that complements the NRC׳s deterministic approach. PRA evaluations in support of regulatory decisions should be as realistic as practicable. The Commission׳s safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments. The Nuclear Community has been encouraging the use of PRA to support a risk-informed decision-making framework [2]. In this context, the NRC issued the first draft of Regulatory Guide RG 1.174 in 1998 [3], which remains a major milestone in the NRC initiative to risk-inform the regulations on changes to LB. RG 1.174 introduces the five principles of the risk-informed decision-making to be used for making decisions regarding plant-specific changes to LB. The fourth principle states: “When proposed changes result in an increase in core damage frequency or risk, the increases should be small and consistent with the intent of the Commission׳s Safety Goal Policy Statement”. This principle is the one we are concerned with in this paper. Risk and risk increase are quantified using PRA, i.e. using the Core Damage Frequency (CDF) derived from a level 1 PRA and/or the Large Early Release Frequency (LERF) derived from a level 2 PRA. The Guide defines acceptable ranges of values for the possible increase in CDF and LERF. It also recognizes that the scope, level of detail and technical acceptability of the PRA should commensurate with the application. Caruso et al. [4] presents an approach for using risk assessment in risk-informed decisions on plant-specific changes to licensing basis that is based on the first draft of RG 1.174. Since then, the risk-informed process introduced in RG 1.174 has evolved into a suite of regulatory guides and NUREG reports that define an integrated approach to risk-informed regulation [3], [5], [6], [7], [8], [9] and [10]. Nowadays, there are draft versions of Revision 3 to RG 1.174 (DG-1285) and Revision 2 to RG 1.177 (DG-1287). RG 1.174 [3] presents a framework umbrella for using PRA in risk-informed decision-making on specific changes to licensing basis, while RG 1.177 [5] proposes a more specific approach that focuses, in particular, on plant specific changes to TS, e.g. LCO and SR, which are parts of the licensing basis as introduced early in this section. The original US NRC policy statement in 1995 and the first drafts of RG 1.174 and RG 1.177 in 1998 already established that all sources of uncertainty must be identified and analyzed such that their impacts are understood. Prior work in this field has already faced the problem of addressing uncertainties in reliability and risk based decision making on changes to licensing basis and particularly to TS [11], [12], [13], [14], [15], [16], [17] and [18]. However, comprehensive guidance on the systematic treatment of epistemic uncertainties associated with the specific use of the PRA in risk-informed decision making of changes to LB has expanded mainly in the last years [8], [19] and [20]. Moreover, no specific guidance has been proposed yet for the treatment and analysis of epistemic uncertainties particularly in evaluating the risk impact of changes to TS based on the use of the PRA; therefore, there was a need of adapting the generic guidance for LB changes to this particular PRA based application. This was the aim of the work published in Refs. [21], [22] and [23], which show the origins of a methodology that has evolved into the integrated approach proposed in this paper. This paper presents the framework and proposes three steps based approach, i.e. risk modeling, risk assessment and risk analysis, for the evaluation of risk impact of changes to Surveillance Requirements in TS based on the use of the PRA, which includes identification, treatment and analysis of model and parameter uncertainties in an integrated manner. It is coherent with the integrated approach to risk-informed decision-making defined by the suite of guides and reports introduced above. The paper is organized as follows. Section 2 introduces an overview of a Nuclear Power Plant Technical Specification paying attention to the role of the Surveillance Requirements and Surveillance Frequency. Section 3 presents the framework for the evaluation of risk impact of changes to Technical Specification addressing uncertainties while Section 4 presents three steps based approach proposed for the evaluation of the risk impact of a SR change addressing model and parameter uncertainties. Section 5 presents the results of the example of application that focuses on the evaluation of risk impact of a change to Surveillance Frequency of the Reactor Protection System of a Pressurized Water Reactor Nuclear Power Plant using a level 1 PRA at power and considering internal events only, i.e. adopting the CDF as risk measure. Section 6 presents the concluding remarks.