Alan M. Kolaczkowski

Expert elicitation approach for performing ATHEANA quantification

Abstract An expert elicitation approach has been developed to estimate probabilities for unsafe human actions (UAs) based on error-forcing contexts (EFCs) identified through the ATHEANA (A Technique for Human Event Analysis) search process. The expert elicitation approach integrates the knowledge of informed analysts to quantify UAs and treat uncertainty (‘quantification-including-uncertainty’). The analysis focuses on (a) the probabilistic risk assessment (PRA) sequence EFCs for which the UAs are being assessed, (b) the knowledge and experience of analysts (who should include trainers, operations staff, and PRA/human reliability analysis experts), and (c) facilitated translation of information into probabilities useful for PRA purposes. Rather than simply asking the analysts their opinion about failure probabilities, the approach emphasizes asking the analysts what experience and information they have that is relevant to the probability of failure. The facilitator then leads the group in combining the different kinds of information into a consensus probability distribution. This paper describes the expert elicitation process, presents its technical basis, and discusses the controls that are exercised to use it appropriately. The paper also points out the strengths and weaknesses of the approach and how it can be improved. Specifically, it describes how generalized contextually anchored probabilities (GCAPs) can be developed to serve as reference points for estimates of the likelihood of UAs and their distributions.

An empirical study of HRA methods - overall design and issues

A diversity of Human Reliability Analysis (HRA) methods are currently available to treat human performance in Probabilistic Risk Assessments (PRAs). This range of methods reflects traditional concerns with human-machine interfaces and with the basic feasibility of actions in PRA scenarios as well as the more recent attention paid to Errors of Commission and decision- making performance. Given the differences in the scope of the methods and their underlying models, there is a substantial interest in assessing HRA methods and ultimately in validating the approaches and models underlying these methods. A significant step in this direction is an international evaluation study of HRA methods, based on comparing the observed performance in simulator experiments with the outcomes predicted in HRA analyses. Its aim is to develop an empirically- based understanding of the performance, strengths, and weaknesses of the methods. This paper presents the overall methodology for this initial assessment study.

Human reliability analysis (HRA) in the context of HRA testing with empirical data

Given the significant differences in the scope, approach, and underlying models of a relatively wide range of existing HRA methods, there has been a growing interest on the part of HRA method developers and users to empirically test the various methods. To this end, there is an ongoing international effort to begin this process by testing the application of HRA methods to nuclear power plant operating crew performance in the HAMMLAB simulators at the Halden Reactor Project in Norway. Initial efforts in designing and implementing these studies have identified a number of issues associated with structuring the studies in order to allow an adequate and appropriate test of the different methods. This paper focuses on issues associated with applying HRA methods in the context of an empirical study, particularly when a research simulator is used for data collection. Example issues include: determining the scope of the analysis when the methods themselves differ in the scope of the HRA processes they address, accounting for differences between the methods in how they use simulator exercises to support the analysis, addressing the impact of experimental controls on application of methods, and given the low probability of human failure events typically modelled in nuclear power plant probabilistic risk/safety assessments (PRAs/PSAs), the need for analysts to present their results in a somewhat different format than they usually do. These types of issues related to applying HRA methods in the context of empirical studies are discussed and resolutions are proposed.

An overview of the evolution of human reliability analysis in the context of probabilistic risk assessment.

Since the Reactor Safety Study in the early 1970s, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood.

Constraints in designing simulator scenarios and identifying human failure events for testing HRA methods

A diversity of different Human Reliability Analysis (HRA) methods are applied in Probabilistic Risk Assessments (PRAs). The Halden Reactor Project has, together with its member organizations, initiated a project to address a desire to test these methods against empirical evidence. In the initial phase of the project, specific accident scenarios have been chosen and human failure events (HFEs) have been identified. As part of these first trials, we are learning about some of the constraints, issues, and characteristics of the scenario designs that have to be dealt with. For instance, scenarios should be PRA/HRA relevant, sufficiently challenging to be useful tests of the HRA methods and yet be plausible, and also be feasible in a simulator setting. We are also learning about issues on how to define the HFEs of interest to ensure usefulness of the results to PRA/HRA. For instance, HFEs must be measurable and defined so that success of the action and failure of the action are clearly distinguishable. Process-based success criteria can for example reflect on several HFEs, and to define success criteria for individual HFEs other measures might be needed. This paper will discuss lessons learned about scenario design and the defining of HFEs to be able to progress toward testing HRA methods using simulations of accident scenarios.

Selection of pressurized thermal shock (PTS) transients to include in PTS risk analyses

Pressurized thermal shock (PTS)-risk-significant events have not typically been included in the probabilistic risk analyses (PRAs) of nuclear power plants (e.g. those used to perform the individual plant evaluations (IPEs)). This paper describes the process used to identify the PTS-risk-significant events to be added to the PRAs of plants being studied in an on-going Nuclear Regulatory Commission (NRC) PTS risk evaluation. The process requires consideration of the five necessary features of PTS events: (1) and (2) fast neutron embrittlement and a crack or flaw in the reactor pressure vessel; (3) rapid cooling of the primary system; (4) a sustained low temperature (<∼ 176.7°C {350°F} for vessels with reference temperature for the nil-ductility transition (RT NDT ) less than 132.2°C {270°F}); and (5) repressurization (or maintenance of high primary system pressure). Without the presence of all five of these features, no event will pose a significant PTS-related risk. This paper assumes the presence of the first two (materials-related) features, and discusses the remaining three (transient-related) features. To date, no radically new initiators or event sequences have been identified. However, since this study considers the added operational difficulties that may be caused by the loss of support systems and the effects on the operators of more complex initiating events, it is believed that compared to previous studies, more realistic PTS risk results will be obtained.

Human Reliability Analysis (HRA) Good Practices

With the expectation that PRA will continue to be used in the commercial nuclear industry in assessing current operating risks, in estimating changes in risk as a result of temporary and permanent plant changes, and as an adjunct to the design process of newer generation plants, it is important that practitioners perform human reliability analysis (HRA) in accordance with good practices and that reviewers recognize the implementation of good practices (or failure to do so) in these analyses. Documents such as the American Society of Mechanical Engineers (ASME) Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications provide requirements for performing a good HRA. However, the requirements do not provide implementation guidance, i.e., they are not as “tutorial” as is necessary to support HRA analysts. This paper provides examples of “good practices” for a HRA as they should be implemented within a broader probabilistic risk assessment (PRA). It also discusses the consideration of “errors of commission” (EOCs) that is not explicitly addressed in the ASME Standard. The good HRA practice guidance focuses on identifying the attributes of a good HRA regardless of the specific methods or tools that are used and does not endorse the use of any particular method.

Methods Advances in EPRI/NRC Fire Risk Re-quantification Project: Modeling of Post-Fire Safe Shutdown in Fire PRA

A significant portion of a fire risk assessment requires examination of the plant response to a postulated set of fire scenarios. The plant response to a given fire scenario depends on the systems damaged by the fire, the plant’s post-fire safe-shutdown design attributes, and operating practices — including applicable emergency procedures and expected operator response to the scenario. This paper presents an overview of the techniques being documented under the EPRI/NRC Fire Risk Re-quantification Project.