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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Assessment of HRA method predictions against operating crew performance: Part II: Overall simulator data, HRA method predictions, and intra-method comparisons

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.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Assessment of HRA method predictions against operating crew performance: Part I: Study background, design and methodology

This is the first in a series of four papers document ing 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 accident scenarios simulated on nuclear power plant (NPP) simulators. When a met hod was applied by different analyst teams, the predictions of the different teams were also compared against each other. The purpose of the comparisons was to obtain insights concerning the strengths and weaknesses of the HRA methods addressed in the studies as well as the factors contributing to inter-analyst (or intra-method) variability. The background and motivation for the studies is first presented in the article. Then, the overall study design is described with a discussion on high-level tasks and participants. Next, the methodology to achieve the high-level tasks is discussed. Finally, concluding remarks are made concerning lessons learned on benchmarking HRA methods with crew performance of scenarios on NPP simulators.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Assessment of HRA method predictions against operating crew performance: Part III: Conclusions and achievements

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.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Utilizing the SACADA system to collect simulator training data to inform human reliability analyses - 169

Lack of appropriate human performance data for the development of human reliability analysis (HRA) methods has been identified as a key issue contributing to the method-to-method variability among the HRA methods. The United States Nuclear Regulatory Commission (NRC) worked with domain experts to develop the Scenario Authoring, Characterization, and Debriefing Application (SACADA) system to collect nuclear power plants' licensed operator simulator training and examination data to inform HRA. In the past two decades, significant improvements in understanding and modelling human performance in HRA have affected the perspectives of the types of data to be collected for HRA. A significant change is that human performance is strongly dependent on the context of the tasks to be performed. To incorporate this understanding, the SACADA data taxonomy collects the context and actual performance information of the training objective elements (TOEs) in nuclear plant operator simulator training. This provides a framework consistent with HRA to use context to estimate human error probabilities. It is recognized that measuring crew performance is equally important as improving operator training. SACADA is designed to support the nuclear trainers in capturing this important information that is used to identify training weaknesses and drive continuous improvement. This paper discusses the SACADA taxonomy, the SACADA system's functions, and the supporting software design to enable the SACADA system to be practically implemented in the nuclear power plants' operator simulator training program to collect data for tracking operator performance and HRA analysis. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Computerized operator support system and human performance in the control room - 262

An operator aid technology to assist nuclear plant operators maintain situation awareness and detect faults earlier than would be possible using conventional control room technologies is described. Coined the Computerized Operator Support System (COSS), this technology is aimed at the human-machine interface. The COSS has been implemented as a prototype system operating in the control room environment of a full-scale simulator for a commercial nuclear power plant. A human factors assessment that involved two licensed reactor crews was performed to evaluate crew performance. This paper describes the development of the COSS, the underlying fault-detection architecture, and feedback received based on operator-in-the-loop studies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Individual Differences in Human Reliability Analysis

While human reliability analysis (HRA) methods include uncertainty in quantification, the nominal model of human error in HRA typically assumes that operator performance does not vary significantly when they are given the same initiating event, indicators, procedures, and training, and that any differences in operator performance are simply aleatory (i.e., random). While this assumption generally holds true when performing routine actions, variability in operator response has been observed in multiple studies, especially in complex situations that go beyond training and procedures. As such, complexity can lead to differences in operator performance (e.g., operator understanding and decision-making). Furthermore, psychological research has shown that there are a number of known antecedents (i.e., attributable causes) that consistently contribute to observable and systematically measurable (i.e., not random) differences in behavior. This paper reviews examples of individual differences taken from operational experience and the psychological literature. The impact of these differences in human behavior and their implications for HRA are then discussed. We propose that individual differences should not be treated as aleatory, but rather as epistemic. Ultimately, by understanding the sources of individual differences, it is possible to remove some epistemic uncertainty from analyses.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Science-based HRA: experimental comparison of operator performance to IDAC (Information-Decision-Action Crew) simulations

Information-Decision-Action Crew (IDAC) operator model simulations of a Steam Generator Tube Rupture are compared to student operator performance in studies conducted in the Ohio State University’s Nuclear Power Plant Simulator Facility. This study is presented as a prototype for conducting simulator studies to validate key aspects of Human Reliability Analysis (HRA) methods. Seven student operator crews are compared to simulation results for crews designed to demonstrate three different decision-making strategies. The IDAC model used in the simulations is modified slightly to capture novice behavior rather that expert operators. Operator actions and scenario pacing are compared. A preliminary review of available performance shaping factors (PSFs) is presented. After the scenario in the NPP Simulator Facility, student operators review a video of the scenario and evaluate six PSFs at pre-determined points in the scenario. This provides a dynamic record of the PSFs experienced by the OSU student operators. In this preliminary analysis, Time Constraint Load (TCL) calculated in the IDAC simulations is compared to TCL reported by student operators. We identify potential modifications to the IDAC model to develop an “IDAC Student Operator Model.” This analysis provides insights into how similar experiments could be conducted using expert operators to improve the fidelity of IDAC simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Optimization Scheme for Establishing Queues for Shipment of UNF from Shutdown Reactor Sites - 17554

Approximately 14, 815 used nuclear fuel assemblies (5,800 MTHM) are stored in 486 dry storage cask systems located at 12 independent spent fuel storage installations (ISFSIs) located at shutdown reactor sites located across the U.S. that have no access to spent fuel pools. A preliminary evaluation was performed by the University of North Carolina at Charlotte (UNCC) in collaboration with AREVA to establish optimization schemes for establishing queues for the removal of this used/spent nuclear fuel (hence forth identified as UNF) and Greater than Class C low level wastes (GTCC) found at these shutdown reactor sites. These queues assumed the major portion of the shipment of the UNF and GTCC would be by rail, with heavy haul truck or barge potentially utilized to move the UNF from the site to rail for sites not served directly by rail. The destination for these rail shipments were a hypothetical consolidated interim storage facility (CSF) located in West Texas. The queues were developed based on the total costs and time to finish all shipments needed to remove all the UNF from these shutdown sites. Results are presented for optimization schemes based on a continuous flow of shipments using 1, 2, and 3 crews dedicated to individual ISFSIs and discrete shipments using 2 crews capable of performing operations at multiple ISFSIs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Spray Foam Exterior Insulation with Stand-Off Furring - New York

IBACOS and Green Homes America 1 (GHA) collaborated with the New York State Energy Research and Development Authority (NYSERDA) Advanced Buildings Program to research a variation of a commercial curtain wall assembly for residential construction as part of a comprehensive whole-house deep energy retrofit (DER) package. The advantages of this strategy include minimizing the physical connections to each individual existing wall stud, using closedcell spray foam to encapsulate existing siding materials including lead paint, and creating a vented rain screen assembly to promote drying. As Moore (2011) states, "Closed-cell foam acts as air-sealing, insulation, and secondary drainage plane..." Field installations on a limited number of houses indicate that the incremental cost when installed as part of a typical siding and window replacement job adds approximately $10/ft2 of wall area. 2 x 4 framing members were attached to the wall at band joists and top plates using "L" clips, and spray foam was used to insulate the wall after the framing was installed. Windows were installed as the framing was put up, including extension jambs. Key lessons learned include the following: (1) The installed cost of the spray foam system was comparable to the installed cost of the rigid foam and furring strips. (2) Spray foam should be applied in such a way that it locks the framing members in place before the remaining foam is installed in the field of the wall. The spray foam has a tendency to move the new wall framing as it expands. Best practice is to "picture frame" the wall framing with spray foam and then apply the spray foam in the field of each bay. (3) Variations in the existing siding conditions must be accounted for when attaching the ledger boards and wall framing. A best practice is to uniformly shim the ledger boards off the walls so the new framing is a uniform distance from the ledger boards and the inside surface of the interior framing. This is important primarily for window and door installation and the associated extension jambs and trim. (4) From a sequencing perspective, it is advantageous to replace the windows at the same time as the exterior walls are being built. This requires a crew that can perform a number of different tasks (e.g., framing, flashing, window install, trim carpentry) and does not lend itself to subcontracting this approach piecemeal (e.g., framing subcontractor, window installer, drainage plane installer, trim carpenter). (5) Application of the brackets and associated fastener screws was more time consuming than anticipated. (6) It is important for the spray foam contractor to perform all preparatory work, which in this case included placing plastic over the windows and exposed foundation to minimize or eliminate the possibility of damage due to overspray.

Array↗
DOE OSTI2018

Toward a Synthetic Model for Distribution System Restoration and Crew Dispatch

Distribution service restoration (DSR) is critical for improving the resilience and reliability of modern distribution systems by strategically and sequentially energizing the system components and customer loads. Restoring electricity service to affected customers also requires multiple crews with different skill sets to perform multiple tasks that are procedurally interdependent with safety guaranteed. However, in existing DSR practices, switch operations and crew dispatch are scheduled separately, and their interdependence is not fully considered. As advanced technologies are enabling remote communication, control, and dispatch, utilities now desire an integrated DSR framework to achieve seamless coordination among multiple DSR tasks such as switch operation, crew dispatch, and component repair. In this paper, we introduce a synthetic model that integrates the service restoration model and the crew dispatch model based on a universal routing model. The proposed model can provide the estimated time of restoration for each load, the switching sequence for safely operating remotely/manually operated switches, and dispatch solutions for crews with different skill sets. The proposed synthetic model is formulated as a mixed-integer linear programming model, and its effectiveness is evaluated via the IEEE 123 bus test feeder and several large-scale test feeders (EPRI Ckt5, Ckt7, Ckt24, and IEEE 8500 node test feeder.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2016

Continuous Reliability Enhancement for Wind (CREW). Program Update

Sandia's Continuous Reliability Enhancement for Wind (CREW) Program is a follow on project to the Wind Plant Reliability Database and Analysis Program. The goal of CREW is to characterize the reliability performance of the US fleet to serve as a basis for improved reliability and increased availability of turbines. This document states the objectives of CREW and describes how data collected for CREW will be used in analysis. A critical aspect to the success of the CREW project is data input from participating owner/operators. The level of detail and the quality of input data provided dictates the type of analysis that can be accomplished. Options for analysis range from high level availability summaries to detailed analysis of failure modes for individual equipment items. Specific types of input data are identified followed by samples of the type of output that can be expected along with a discussion of benefits to the user community.

17 WIND ENERGY↗
DOE OSTI2019

Weatherization Crew Leader Job Task Analysis

The US Department of Energy's Guidelines for Home Energy Professionals project maintains job task analyses for jobs in the Weatherization Assistance Program and home performance industry. One of those jobs is Crew Leader, defined as: 'A Crew Leader is responsible for supervising and assisting in the retrofitting activities specified in the scope of work. He or she is responsible for interacting with the client plus managing personnel and materials on the job site in a safe and effective manner. The Crew Leader is responsible for quality control, testing procedures, documentation, and conducting a walk through in preparation that all work is completed in a satisfactory manner.' This document is an updated version of the Crew Leader job task analysis.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2014

Modeling and Quantification of Team Performance in Human Reliability Analysis for Probabilistic Risk Assessment

Probabilistic Risk Assessment (PRA) and Human Reliability Assessment (HRA) are important technical contributors to the United States (U.S.) Nuclear Regulatory Commission’s (NRC) risk-informed and performance based approach to regulating U.S. commercial nuclear activities. Furthermore, all currently operating commercial NPPs in the U.S. are required by federal regulation to be staffed with crews of operators. Yet, aspects of team performance are underspecified in most HRA methods that are widely used in the nuclear industry. There are a variety of "emergent" team cognition and teamwork errors (e.g., communication errors) that are 1) distinct from individual human errors, and 2) important to understand from a PRA perspective. The lack of robust models or quantification of team performance is an issue that affects the accuracy and validity of HRA methods and models, leading to significant uncertainty in estimating HEPs. This paper describes research that has the objective to model and quantify team dynamics and teamwork within NPP control room crews for risk informed applications, thereby improving the technical basis of HRA, which improves the risk-informed approach the NRC uses to regulate the U.S. commercial nuclear industry.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

AI Data Quality Monitoring with Hydra

Hydra is an extensible framework for training and managing AI for near real time monitoring that aims to replace the tedious and repetitive data quality monitoring activities the shift crew and online monitoring coordinator typically perform. It continuously scans incoming data in the form of monitoring plots for signs of problems, flagging them for human review. A web app was developed such that experts can efficiently label images for training. Labels are stored in a database for use in training and model validation. Backed up by a comprehensive database, it utilizes an additional web based front-end for viewing the current monitoring status from anywhere in the world. The system has been in production use for the GlueX experiment at Jefferson Lab for more than 2 years with new features still under active development.

Britton, Thomas↗
DOE OSTI2017

HFE in the design of the socio-technical system of the control room of a new generation reactor: issues faced during multistage validation - 126

This paper deals with the contribution of Human Factors and Ergonomics (HF and E) experts to the design of the control room (CR) of a new generation reactor (NGR), our view of the multistage validation (MSV) process and the issues faced during it. Our approach to MSV was to articulate two points of view: the designers' point of view (test and explore various design options) and the regulatory authorities' point of view (validate the performance of the socio-technical system of the CR). To perform this articulation, both tests and validation shared the same method and goals. The main issues faced during this process were to (1) get a coherent and integrated version of the full-scope simulator when the design of the socio-technical system of the CR was still in progress, (2) determine when to stop MSVs, and (3) define a criterion for the sampling of operational conditions. The first issue led us to the conclusion that specific project milestones synchronising the design of every part of the CR have to be planned early in the project in order to mitigate the risk of delaying WSTs (Whole System Tests). Regarding the second issue, we advocate that no further WSTs are required when the last modifications brought to the CR design have no significant impact on crews' activities. Finally, concerning the last issue, we think that performing WSTs in all classes of situations a crew may face during operating the plant is necessary. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Operating displays - new concept: high performance displays - 118

New control rooms in Nuclear Power Plants using digital Distributed Control System (DCS) operated with software displays, are providing thousands of signals to the operating crew. When an event happens, the Flood of information provided to operators could obscure the initiator event increasing the operator response time, work load and complicating the selection of the actions to be performed. There are some initiatives that are trying to minimize the 'avalanche effect' as the alarm engineering filtering and prioritization or diagnostic displays. Likewise, new control rooms are lost the availability to know the plant status just with a quick glance to the panel indicators. Regarding Small Modular Reactors, where in some cases four operators should monitor and control twelve reactors, these problems are even more significant. The challenge is to find a solution, based on current digital systems, that supply information to operators in order to prevent events and take early actions to minimize or avoid any unexpected incident. In order to minimize human errors, Tecnatom has developed a specific set of displays, called 'High Performance Displays' (HPD) based on objectives, work environment limitation, and non-anticipated events. These displays are being design following the HFE standards: applying the NUREG-0711 elements to identify the information that they have to content and the NUREG-0700 to design the interface. The function of these displays is monitoring the plant situation in a way that the operators are able to recognize any potential event at a simple glance but controlling thousands of signals through pattern recognition, symmetries, regular polygons and colors. The objective of the HPD is to provide an integrated function minimizing the tunnel effect and reduce the workload during the diagnosis phase. Moreover, HPD enhance operators to take the right actions to anticipate the response to any event, even avoiding it, if possible This paper analyzes how HPD's have been design as well as presents the results of a preliminary validation of their capabilities. The preliminary validation is a reduce validation in a full scope simulator running two representative scenarios of abnormal and emergency operation. The analysis is focused on the first stage of the event, i.e. the detection phase, HPD's have been installed in a generic full scope simulator of a nuclear power plant, digital controlled. Both scenarios have been performed by two different operator crews. The first operator shift has worked with the conventional human machine interface, that is, displays similar to the ones currently installed in existing nuclear power plant with digital control. The second crew, reproduce the same scenarios but with HPD available. The studio collects preliminary data in order to compare results, draw lesson learned from the experience to improve HPD design, and find solutions in terms of reducing human errors, human workload, and increase the availability of the plant. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Challenges in leveraging existing human performance data for quantifying the IDHEAS HRA method

Our 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. Furthermore, 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗