Jeffrey C. Joe

The Measure of human error: Direct and indirect performance shaping factors

The goal of performance shaping factors (PSFs) is to provide measures to account for human performance. PSFs fall into two categories—direct and indirect measures of human performance. While some PSFs such as “time to complete a task” are directly measurable, other PSFs, such as “fitness for duty,” can only be measured indirectly through other measures and PSFs, such as through fatigue measures. This paper explores the role of direct and indirect measures in human reliability analysis (HRA) and the implications that measurement theory has on analyses and applications using PSFs. The paper concludes with suggestions for maximizing the reliability and validity of PSFs.

Simplified Expert Elicitation Procedure for Risk Assessment of Operating Events

This report describes a simplified, tractable, and usable procedure within the US Nuclear Regulator Commission (NRC) for seeking expert opinion and judgment. The NRC has increased efforts to document the reliability and risk of nuclear power plants (NPPs) through Probabilistic Risk Assessment (PRA) and Human Reliability Analysis (HRA) models. The Significance Determination Process (SDP) and Accident Sequence Precursor (ASP) programs at the NRC utilize expert judgment on the probability of failure, human error, and the operability of equipment in cases where otherwise insufficient operational data exist to make meaningful estimates. In the past, the SDP and ASP programs informally sought the opinion of experts inside and outside the NRC. This document represents a formal, documented procedure to take the place of informal expert elicitation. The procedures outlined in this report follow existing formal expert elicitation methodologies, but are streamlined as appropriate to the degree of accuracy required and the schedule for producing SDP and ASP analyses.

Early-Stage Design and Evaluation for Nuclear Power Plant Control Room Upgrades

As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate operator performance with these systems as part of a verification and validation process. While there is regulatory and industry guidance for some modernization activities, there are no well defined standard processes or predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages. This paper proposes a framework defining the design process and metrics for evaluating human system interfaces as part of control room modernization. The process and metrics are generalizable to other applications and serve as a guiding template for utilities undertaking their own control room modernization activities.

Extracting and converting quantitative data into human error probabilities

This paper discusses a proposed method using a combination of advanced statistical approaches (e.g., meta-analysis, regression, structural equation modeling) that will not only convert different empirical results into a common metric for scaling individual PSFs effects, but will also examine the complex interrelationships among PSFs. Furthermore, the paper discusses how the derived statistical estimates (i.e., effect sizes) can be mapped onto a HRA method (e.g. SPAR-H) to generate HEPs that can then be use in probabilistic risk assessment (PRA). The paper concludes with a discussion of the benefits of using academic literature in assisting HRA analysts in generating sound HEPs and HRA developers in validating current HRA models and formulating new HRA models.

Human factors issues for multi-modular reactor units

Smaller and multi-modular reactors (MMR) will be highly technologically-advanced systems allowing more system flexibility to reactor configurations (e.g., addition/removal of reactor units). While the technical and financial advantages of such systems may be numerous, MMR presents many human factors challenges that may pose vulnerabilities to plant safety. An important human factors challenge in MMR operation and performance is the monitoring of data from multiple plants from centralized control rooms where human operators are responsible for interpreting, assessing, and responding to different system states (e.g., simultaneously monitoring refueling at one plant while vigilant to another plant’s normal operating state). Furthermore, the operational, safety, and performance requirements for MMR can significantly change current staffing models and roles, the mode in which information is displayed, and the approach for conducting procedures and training. Consequently, addressing human factors concerns in MMR is essential in reducing plant risk.

Human Performance Modeling for Dynamic Human Reliability Analysis

Part of the U.S. Department of Energy’s (DOE’s) Light Water Reactor Sustainability (LWRS) Program, the Risk-Informed Safety Margin Characterization (RISMC) Pathway develops approaches to estimating and managing safety margins. RISMC simulations pair deterministic plant physics models with probabilistic risk models. As human interactions are an essential element of plant risk, it is necessary to integrate human actions into the RISMC risk framework. In this paper, we review simulation based and non simulation based human reliability analysis (HRA) methods. This paper summarizes the foundational information needed to develop a feasible approach to modeling human interactions in RISMC simulations.

Development of a Technical Basis and Guidance for Advanced SMR Function Allocation

This report presents the results from three key activities for FY13 that influence the definition of new concepts of operations for advanced Small Modular Reactors (AdvSMR: a) the development of a framework for the analysis of the functional environmental, and structural attributes, b) the effect that new technologies and operational concepts would have on the way functions are allocated to humans or machines or combinations of the two, and c) the relationship between new concepts of operations, new function allocations, and human performance requirements.

PROOF OF CONCEPT FOR A HUMAN RELIABILITY ANALYSIS METHOD FOR HEURISTIC USABILITY EVALUATION OF SOFTWARE

An ongoing issue within human-computer interaction (HCI) is the need for simplified or “discount” methods. The current economic slowdown has necessitated innovative methods that are results driven and cost effective. The myriad methods of design and usability are currently being cost-justified, and new techniques are actively being explored that meet current budgets and needs. Recent efforts in human reliability analysis (HRA) are highlighted by the ten-year development of the Standardized Plant Analysis Risk HRA (SPAR-H) method. The SPAR-H method has been used primarily for determining humancentered risk at nuclear power plants. The SPAR-H method, however, shares task analysis underpinnings with HCI. Despite this methodological overlap, there is currently no HRA approach deployed in heuristic usability evaluation. This paper presents an extension of the existing SPAR-H method to be used as part of heuristic usability evaluation in HCI.

Human Reliability Analysis in the U.S. Nuclear Power Industry: A Comparison of Atomistic and Holistic Methods

A variety of methods have been developed to generate human error probabilities for use in the US nuclear power industry. When actual operations data are not available, it is necessary for an analyst to estimate these probabilities. Most approaches, including THERP, ASEP, SLIM-MAUD, and SPAR-H, feature an atomistic approach to characterizing and estimating error. The atomistic approach is based on the notion that events and their causes can be decomposed and individually quantified. In contrast, in the holistic approach, such as found in ATHEANA, the analysis centers on the entire event, which is typically quantified as an indivisible whole. The distinction between atomistic and holistic approaches is important in understanding the nature of human reliability analysis quantification and the utility and shortcomings associated with each approach.

Light Water Reactor Sustainability Program Operator Performance Metrics for Control Room Modernization: A Practical Guide for Early Design Evaluation

As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate the operator performance using these systems as part of a verification and validation process. There are no standard, predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages of a new system. This report identifies the process and metrics for evaluating human system interfaces as part of control room modernization. The report includes background information on design and evaluation, a thorough discussion of human performance measures, and a practical example of how the process and metrics have been used as part of a turbine control system upgrade during the formative stages of design. The process and metrics are geared toward generalizability to other applications and serve as a template for utilities undertaking their own control room modernization activities.