This is the last in a series of three papers documenting two large-scale human reliability analysis (HRA) empirical studies – the International HRA Empirical Study and the US HRA Empirical Study. The goal of the two studies was to develop an empirically-based understanding of the performance, strengths, and weaknesses of HRA methods by comparing HRA method predictions against actual operator performance in simulated accident scenarios on nuclear power plant (NPP) simulators. This paper first addresses areas where there is convergence between the two studies and where differences lie. Then it summarizes the combined insights and conclusions, including key findings on HRA in general through lessons learned about the HRA methods assessed in the studies and specific recommendations for improving guidance, practice and methods. Then it discusses the relevance and usefulness of simulator data for HRA in general. Finally, it presents the key achievements and overall conclusions of the two studies taken together.
This is the first in a series of three papers documenting two large-scale human reliability analysis (HRA) empirical studies – the International HRA Empirical Study and the US HRA Empirical Study. The two studies are the first major efforts in recent years to benchmark HRA methods by comparing HRA method predictions against actual operator performance in responding to accidents simulated on nuclear power plant (NPP) full-scale simulators. The studies aimed to gain knowledge and insights concerning the strengths and weaknesses of the studied HRA methods and the factors contributing to inter-analyst (or intra-method) variability. In addition, the studies also compared the results of the same HRA method applied by different analysis teams. This paper provides the background and motivation of the studies, the overall study design, the simulation scenarios and human failure events to be analyzed, and concluding remarks concerning lessons learned on benchmarking HRA methods with crew performance of scenarios on NPP simulators.
This is the third in a series of four papers documenting two large-scale human reliability analysis (HRA) empirical studies – the International HRA Empirical Study and the US HRA Empirical Study. Here, the goal of the two studies was to develop an empirically-based understanding of the performance, strengths, and weaknesses of HRA methods by comparing HRA method predictions against actual operator performance in simulated accident scenarios on nuclear power plant (NPP) simulators. However, since in most cases only a single HRA team applied a given method in the International study, it was often difficult to separate analyst effects from variability in results related to the methods themselves. Since at least two HRA teams used each of the HRA methods in the US Study, intra-method comparisons were performed to identify method strengths and weaknesses independent of analyst specific effects where possible. This paper first summarizes the intra-method comparison results from the U.S. Study. Then, it discusses the reasons for the observed HRA predictive differences and the underlying methodological and guidance limitations that permitted the differences to arise. In the fourth paper, this information is combined wit h the results of the comparisons of method predictions to the actual crew data, from both the International and U.S. Studies, to develop the final conclusions about overall strengths and weaknesses of HRA methods.
This is the second in a series of four papers documenting two large-scale human reliability analysis (HRA) empirical studies – the International HRA Empirical Study and the US HRA Empirical Study. The goal of the two studies was to develop an empirically-based understanding of the performance, strengths, and weaknesses of HRA methods by comparing HRA method predictions against actual operator performance in simulated accident scenarios on nuclear power plant (NPP) simulators. The first paper (Part I), provided background in formation for the studies and an overview of their design and methodology. This paper first briefly describes the scenarios simulated in the studies and the associated human failure events (HFEs) addressed in the HRA analyses. Then, it discusses the overall simulator data followed by observations on the operating crew performance in the scenario simulations. Lastly, it presents some quantitative comparisons of the HRA methods’ predictions with the simulator data.
The Integrated Human Event Analysis System for nuclear power plant internal events at-power application (hereafter "IDHEAS AT-POWER") is a new human reliability analysis (HRA) method developed by the U.S. Nuclear Regulatory Commission (NRC) in collaboration with the Electric Power Research Institute (EPRI). It was developed to provide a structured approach to the qualitative and quantitative analysis of operator actions during internal, at-power nuclear power plant events. The IDHEAS AT-POWER method was tested to evaluate whether its guidance can be practically applied to produce consistent HRA results. This paper presents study findings and final conclusions on the method performance. Lessons learned on study methodology and recommendations for method improvement are also presented.
This article documents an exploratory study for collecting and using human performance data to inform human error probability (HEP) estimates for a new human reliability analysis (HRA) method, the IntegrateD Human Event Analysis System (IDHEAS). The method was based on cognitive models and mechanisms underlying human behaviour and employs a framework of 14 crew failure modes (CFMs) to represent human failures typical for human performance in nuclear power plant (NPP) internal, at-power events [1]. A decision tree (DT) was constructed for each CFM to assess the probability of the CFM occurring in different contexts. Data needs for IDHEAS quantification are discussed. Then, the data collection framework and process is described and how the collected data were used to inform HEP estimation is illustrated with two examples. Next, five major technical challenges are identified for leveraging human performance data for IDHEAS quantification. These challenges reflect the data needs specific to IDHEAS. More importantly, they also represent the general issues with current human performance data and can provide insight for a path forward to support HRA data collection, use, and exchange for HRA method development, implementation, and validation.
Disposal overpacks are proposed as an element of the engineered barrier system for direct disposal of spent nuclear fuel in dual-purpose canisters (DPCs) [1]. DPCs are currently licensed for storage and transport, but not disposal. In the DPC disposal system, overpacks would provide long-term containment, and conversely, they would keep groundwater from flooding DPCs. Without flooding, DPCs can never achieve nuclear criticality because they are under-moderated.
Human reliability analysis (HRA) is used in the context of probabilistic risk assessment (PRA) to provide risk information regarding human performance to support risk-informed decision-making with respect to high-reliability industries. The IntegrateD Human Event Analysis System (IDHEAS) is a new HRA method developed for internal, at-power nuclear power plant (NPP) events. It was motivated by the intention to reduce unnecessary and inappropriate variability in HRA results and improve the reliability of human error probability (HEP) estimates. The method has a strong foundation in human performance and cognitive psychology theories, and employs a cause-based quantification model. This paper documents a study conducted to pilot test IDHEAS to (1) identify issues that needed to be addressed and (2) provide feedback to refine the method before the method was finalized. An introduction on IDHEAS is provided first. Then sample IDHEAS analysis results are presented for illustration purposes. Next, insights from the testing in terms of method strengths and weaknesses are discussed, which is followed by concluding remarks.
United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.