Publications

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The Office of Nuclear Regulatory Research (RES) at the US Nuclear Regulatory Commission (USNRC) is sponsoring work in response to a Staff Requirements Memorandum (SRM) directing an effort to establish a single human reliability analysis (HRA) method for the agency or guidance for the use of multiple methods. As part of this effort an attempt to develop a comprehensive HRA qualitative approach is being pursued. This paper presents a draft of the method's middle layer, a part of the qualitative analysis phase that links failure mechanisms to performance shaping factors. Starting with a Crew Response Tree (CRT) that has identified human failure events, analysts identify potential failure mechanisms using the mid-layer model. The mid-layer model presented in this paper traces the identification of the failure mechanisms using the Information-Diagnosis/Decision-Action (IDA) model and cognitive models from the psychological literature. Each failure mechanism is grouped according to a phase of IDA. Under each phase of IDA, the cognitive models help identify the relevant performance shaping factors for the failure mechanism. The use of IDA and cognitive models can be traced through fault trees, which provide a detailed complement to the CRT.

Since the Reactor Safety Study in the early 1970's, 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.

During the 1990's the Electric Power Research Institute (EPRI) developed methods for fire risk analysis to support its utility members in the preparation of responses to Generic Letter 88-20, Supplement 4, "Individual Plant Examination - External Events" (IPEEE). This effort produced a Fire Risk Assessment methodology for at-power that was used by the majority of US Nuclear Power Plants (NPPs) in support of the IPEEE program and several NPPs oversees. Although these methods were acceptable for accomplishing the objectives of the IPEEE, EPRI and the U.S. Nuclear Regulatory Commission (NRC) recognized that these methods require upgrades to support current requirements for Risk-Informed/Performance-Based (RI/PB) applications. In 2001 EPRI and the NRC Office of Nuclear Regulatory Research (RES) embarked on a cooperative project to improve the state-of-the-art in fire risk assessment to support this new risk-informed environment in fire protection. This project produced a consensus document, NUREG/CR-6850 (EPRI 1011989), entitled "Fire PRA Methodology for Nuclear Power Facilities" which addresses fire risk for at-power operations. This report developed: 1) the process for identification and inclusion of the post-fire Human Failure Events (HFEs), 2) the methodology for assigning quantitative screening values to these HFEs, and 3) the initial considerations of performance shaping factors (PSFs) and related fire effects that may need to be addressed in developing best-estimate Human Error Probabilities (HEPs). However, this document does not describe a methodology to develop these best-estimate HEPs given the PSFs and the fire-related effects. In 2007 EPRI and NRC's RES embarked on another cooperative project to develop explicit guidance for estimating HEPs for human error events under fire generated conditions, building upon existing human reliability analysis (HRA) methods. This paper will describe the progress to date on the development and testing of the fire HRA methodology, which includes addressing the range of fire procedures used in existing plants, the range of strategies for main control room abandonment, and the potential impact of fire-induced spurious electrical effects on crew performance. In addition to developing a detailed HRA approach, one goal of the project is to develop a fire HRA scoping quantification approach that allows derivation of more realistic HEPs than those in the screening approach from NUREG/CR-6850 (EPRI 1011989), while requiring less analytic resources than a detailed HRA. In this approach, detailed HRA will be used only for the more complex actions that cannot meet the criteria for the scoping approach.

There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessment (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study is currently underway that compares HRA methods with each other and against operator performance in simulator studies. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies are presented in order to aid in the design of future HRA benchmarking endeavors.

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This report documents state-of-the-art methods, tools, and data for the conduct of a fire Probabilistic Risk Assessment (PRA) for a commercial nuclear power plant (NPP) application. The methods have been developed under the Fire Risk Re-quantification Study. This study was conducted as a joint activity between EPRI and the U. S. NRC Office of Nuclear Regulatory Research (RES) under the terms of an EPRI/RES Memorandum of Understanding [RS.1] and an accompanying Fire Research Addendum [RS.2]. Industry participants supported demonstration analyses and provided peer review of this methodology. The documented methods are intended to support future applications of Fire PRA, including risk-informed regulatory applications. The documented method reflects state-of-the-art fire risk analysis approaches. The primary objective of the Fire Risk Study was to consolidate recent research and development activities into a single state-of-the-art fire PRA analysis methodology. Methodological issues raised in past fire risk analyses, including the Individual Plant Examination of External Events (IPEEE) fire analyses, have been addressed to the extent allowed by the current state-of-the-art and the overall project scope. Methodological debates were resolved through a consensus process between experts representing both EPRI and RES. The consensus process included a provision whereby each major party (EPRI and RES) could maintain differing technical positions if consensus could not be reached. No cases were encountered where this provision was invoked. While the primary objective of the project was to consolidate existing state-of-the-art methods, in many areas, the newly documented methods represent a significant advancement over previously documented methods. In several areas, this project has, in fact, developed new methods and approaches. Such advances typically relate to areas of past methodological debate.

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This paper describes the most recent version of a human reliability analysis (HRA) method called ``A Technique for Human Event Analysis'' (ATHEANA). The new version is documented in NUREG-1624, Rev. 1 [1] and reflects improvements to the method based on comments received from a peer review that was held in 1998 (see [2] for a detailed discussion of the peer review comments) and on the results of an initial trial application of the method conducted at a nuclear power plant in 1997 (see Appendix A in [3]). A summary of the more important recommendations resulting from the peer review and trial application is provided and critical and unique aspects of the revised method are discussed.