Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Judgment”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A paradigm shift: an introduction to structured decision making for sustainable waste management and remediation - 15649

transport of contaminants and human health risk assessment. Most often these modeling activities are carried out in a conservative fashion, usually beyond even the conservatism that is built into the associated regulations and guidance. Decisions are aimed primarily at compliance with regulatory objectives, rather than optimization. This leads to unnecessarily costly remediation and under-utilization of the nation's relatively few radioactive waste disposal facilities. Risk assessment decisions under CERCLA and RCRA might seem to perform better, but if anything this is because the focus is on shorter-term decision-making because long term fate and transport modeling is not considered necessary in most cases. However, some more complex contamination problems suffer from the same basic issues as radioactive waste disposal in terms of the need for complex fate and transport modeling and subsequent risk assessment. At least CERCLA addresses optimization in the feasibility study, but it does not address how to perform that optimization in the face of uncertainty or stakeholders who have different competing objectives. Other environmental issues that fall outside the realm of DOE Orders, NRC regulations, and regulations such as CERCLA and RCRA, have found a different path towards effective decision-making. This includes land reuse and watershed management decisions, for which there is a focus on sustainability and on stakeholder involvement throughout the decision making process. EPA is using 'structured decision analysis' to achieve the primary goal of maximizing societal welfare. These same tools and approaches can be brought into environmental and waste management decisions that need to be made by DOE. Structured decision analysis provides a formal process for capturing not only the science side of the problem, but also the costs and value judgments of the stakeholders, to reach an optimal solution. But this approach requires a paradigm shift within the DOE and NRC environmental and waste management programs. That is, it requires a willingness to engage a new approach that will provide a path towards effective optimization in the decision-making process for the complex environmental problems faced by DOE and NRC. Because of the current nature of problem solving for these types of problems, which involves many layers of conservatism, this structured decision analysis path towards optimization will also realize substantial cost savings for these programs, while maintaining or improving defensibility and transparency in the decision making process. (authors)

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

Approaches for Communicating Technical Information - 16012

Projects within the nuclear industry are generally complex and associated with high hazards. To achieve success there is often a necessity to apply innovation or adopt longer term strategies and solutions. Because of these long timescales and the fact that some approaches revolve around allowing natural radioactive decay to take place prior to visible action, the industry is sometimes perceived to be technically unable to make progress with its environmental and waste management obligations. Work within the industry additionally takes on a high profile and invariably creates greater levels of anxiety and mistrust in the eyes of stakeholders than comparable projects from non-nuclear related industries. Many aspects of the industry are still seen as secretive, environmentally detrimental and 'risky' and all components of the nuclear life-cycle are therefore open to external scrutiny. Although communication and engagement has increased greatly over the years, especially in developed countries, there is still room for improvement in this field. Such improvement might relate to the frequency and level of engagement, the subject matter or the actual process itself. Communication with stakeholders is required throughout the various phases of the nuclear lifecycle whether the activity relates to mining/milling of uranium, fuel production, operational plant activities, decommissioning, remediation or waste management. The concerns and aspirations of stakeholders during these different activities will vary but need to be adequately addressed. Some organisations now have a basic engagement program and convince themselves that they are carrying out true engagement. But just ticking a box to say that an engagement program is in place does not provide evidence of its appropriateness or success as on many occasions' organisations might still not understand how to adequately engage and communicate technical information. It is not always the information we choose to communicate but that which might be withheld that is often perceived to be most relevant to the stakeholder community. Communicating technical information isn't as straightforward as it sounds, but still has to be undertaken, otherwise trust, understanding and acceptance of chosen options will not be achieved. I utilise what I call the 'ten principles of engagement' which I believe can help when communicating technical Information in relation to nuclear and radiological matters. These are; 1. Develop Trust. 2. Educating people and providing information in a transparent manner. 3. Develop an engagement process that is truly a two way process. 4. Openly discuss risk perception and risk communication. 5. Never try to trivialise risk. 6. Where possible, simplify the language you use. 7. Put background and naturally occurring radioactivity into context. 8. Explain how radioactive materials are often used in everyday life. 9. Try to demonstrate that you too are a member of the public. 10.Walk the talk - Try to undertake what you wish others to undertake. Further research is being undertaken to determine if Mental Models might assist with the communication of nuclear and radiological information. A mental model is an explanation of someone's thought process about how something works in the real world. The mental model approach aims to investigate the judgment and decision-making processes of both expert and lay groups, thus allowing a comparison and visualisation of the different views to be made. It is undoubtedly worthwhile keeping a watching brief on the outcome of such research as we strive to improve our communication with the general public. The aim of this paper is to highlight some simple principles and approaches that can assist with the communication of technical information. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Nuclear, Land marker for a Waste Isolation Site: International Architecture Competition - 18240

The design and construction of a nuclear waste marker must include many architectural considerations. Very little research has been conducted by architects to address this issue to date. Historically, the task of designating a nuclear waste site has been a discussion predominantly contained within the scientific and engineering communities. In an effort to engage architects within this dialogue, arch out loud hosted the NUCLEAR competition as an international, architectural investigation into the design of a nuclear waste site marker. The NUCLEAR competition asks participants to design a marker to designate the site of the Waste Isolation Pilot Plant (WIPP) located near Carlsbad, New Mexico. The marker will be meant to be a means of passive institutional controls (PICs) to deter inadvertent human intrusion upon the site and should last 10,000 years. The NUCLEAR competition defined the purpose of the nuclear waste site marker as a means to communicate the danger of inadvertent intrusion from future generations. The competition requires that entries must be both legible in communication and durable in their construction. The competition brief challenged participants to consider how customs and languages will evolve over time and to predict what motivations future generations might have for digging at the site. Other background information was provided about the site such as how waste was stored and how the unique geology of the site seals the waste. The 1993 Sandia National Laboratories report titled Expert Judgment on Markers to Deter Inadvertent Human Intrusion into the Waste Isolation Pilot Plant [1] was referenced along with the 1992 WIPP Land Withdrawal Act [2], interviews with WIPP scientists [3] involved with the WIPP PICs program and the Nuclear Energy Agency Records Knowledge and Memory working group. Competition participants were given digital 3-D Computer-Aided Design models for reference that included the underground storage portion of the site, existing buildings and the 16-square-mile site boundary identified in the 1992 WIPP Land Withdrawal Act [2]. This paper will: - Provide a detailed description of the submitted proposals including graphic analysis of the visualizations, diagrams, technical drawings, jury comments and statements. - Catalogue and identify the consistent themes within all submissions with possible categories including: 1) overall architectural form and language, 2) integration of written language and symbols within architecture, 3) use of technology and robotics, 4) construction methods and materials. (authors)

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

The IAEA coordinated research project on modular HTGR safety design: status and outlook - 18608

The IAEA Coordinated Research Project (CRP) on Modular High Temperature Gas-cooled Reactor (HTGR) Safety Design started in December 2014 and is now in its second year of implementation. The project investigates and makes proposals on safety design criteria by not only considering the existing water cooled reactors safety requirements for nuclear power plant design but also making use of past and current experience of HTGR licensing in the IAEA member states. Two approaches are being followed to establish the criteria. The first is to develop and apply a safety design process to specific reactor de-signs. The second is to use the existing IAEA specific safety requirements for designs based on water cooled reactor technology and then through expert judgment and review propose changes that is needed to make it applicable to HTGRs. The detailed project content and a description of modular HTGRs and its safety features are given. This informs the comments made on the applicability of the existing requirements and the examples of newly pro-posed HTGR specific safety requirements under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Evaluating a Historical Airborne Release Test with Modern Modeling Methods

Safety analysts throughout the U.S. Department of Energy (DOE) complex rely on the data provided in the DOE Handbook, DOE-HDBK-3010, to determine source terms that may be incorporated into the document safety analyses. Most often, analysts simply take the bounding values due to time constraints or simply to bound calculations. This is a safe approach that helps avoid regulatory critique; however, it may not provide results that are meaningful or relevant to the conditions being evaluated. The derivation of the data, such as airborne release fractions (ARFs) and respirable fractions (RFs) in the Handbook often depend on very limited table-top and bench/laboratory experiments, as well as engineering judgment which may not be well substantiated, and may not be representative of the actual situation. One historical dataset included a test where a solid contaminant was sprinkled over a liquid gasoline pool and set on fire (also referred to as Mishima and Schwendiman's test later in this paper). The airborne release was measured over the duration of the burn, but then collected separately for a time after the burning was completed. It can be assumed that the collected contaminant in the latter period corresponds to re-suspended particles, whereas the earlier period collected contaminant includes a combination of directly released as well as early released re-suspended particles. This scenario was recently modeled using SIERRA/Fuego as part of a parametric sensitivity study, however there were some shortcomings of the analysis methods. First, the code at the time lacked a multi-component particle evaporation model, which is necessary to capture the particle size effect of liquid/solid particles that can dynamically evaporate the liquid component. Second, a particle sticking and resuspension model did not exist at the time of the study. Recent model implementation now permits the study be repeated with the improved physical models. In the earlier work, four potential mechanisms were considered to potentially contribute to the release of the particles. These were: 1. Evaporation Induced Entrainment (EIE); 2. Surface Agitation by Wind; 3. Surface Agitation by Boiling; 4. Residue Entrainment (Resuspension). The sensitivity study evaluated mechanism 1 and 3 in the above list. Parameters varied were selected to be a reflection of a practical range of test conditions. The varied parameters included the boiling duration, fuel height, injected mass, particle size, turbulence intensity, and initial particle height. The boiling duration was the most significant parameter. The most significant mechanism was no.3, Surface Agitation by Boiling. This paper reports the implementation of the new multi-component evaporation model and its effect on the findings from the previous effort. The Mishima and Schwendiman test was re-modeled with the new evaporation model, and the results of the re-evaluation are presented. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Recent Progress on the Stochastic Objective Decision Aide (SODA) Application

Idaho State University, in collaboration with Idaho National Laboratory, is engaged in development of a simple to use software application that utilizes stochastic techniques for the calculation of radiation doses associated with hypothetical radioactive material releases. This work is being performed through a grant from the U.S. Department of Energy (DOE) Nuclear Safety Research and Development (NSR and D) Program. Initial progress on the application was reported previously. Recent additions to the application are reported here. The Stochastic Objective Decision Aide (SODA) provides an easy to use software application to help decision makers support their judgment about the need or lack of need for safety structures, systems, and components. SODA is not intended to replace traditional design basis accident dose calculation methodology, nor is it intended to replace existing software used to perform formal dose consequence calculations. Rather, SODA is intended to supplement information obtained from existing codes to give a decision maker more information about an accident scenario. SODA also provides a simple method to perform parametric or sensitivity studies on individual input parameters. Non-reactor nuclear facilities operated under the approval authority of the US DOE use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge dose evaluation guidelines. Unmitigated events that sufficiently challenge dose evaluation guidelines merit the selection of safety structures, systems, or components (SSCs) to prevent or mitigate the hazard. Dose consequence calculations traditionally involve use of the 'five-factor' formula: ST = MAR.DR.ARF.RF.LPF where ST is the source term (Bq), MAR is the total available material-at-risk (Bq), DR is the damage ratio (no units), ARF is the airborne release fraction (no units), RF is the respirable fraction (no units), and LPF is the leak path factor (no units). Potential radiation doses are then calculated using: CED = χ/Q.BR.ST.DCF (2) where CED is the committed effective dose (Sv), χ/Q is the plume dispersion (s/m{sup 3}), BR is the breathing rate (m{sup 3}/s), ST is the source term (Bq), and DCF is the dose conversion factor (Sv/Bq). Conservative single value input parameters are typically used to represent ARF, RF, LPF and BR. The traditional methodology, while conservative, can lead to skewed conclusions in the balance between cost and risk reduction resulting in over engineered systems with greater design, construction and operating costs. Rather than using a bounding single point value for each parameter in the dose consequence calculation, SODA uses distributions for some or all of the parameters. Each input parameter distribution is stochastically sampled and the resulting dose consequence calculation is repeated many times to develop a dose consequence distribution. The resulting distribution provides more information for the decision maker to rely upon when determining the need for safety SSCs. The SODA application was developed using MATLAB and it incorporates the use of Monte Carlo techniques as well as a graphical user interface (GUI). The application includes user selection of the governing distribution for parameters such as the MAR, DR, ARF, RF, LPF,χ/Q, BR, and DCF. While MATLAB was used for the development work, the application is distributed as a self-contained executable program. Features of the application include pull down menus with available distributions for the various parameters. The user has the option to plot the input parameter distribution that results from the random sampling process. The resulting dose consequence distribution is automatically analyzed to determine its similarity to well know distributions. This was done using the Bayesian Information Criterion method. If the resulting distribution is sufficiently similar to a well-known distribution, the application provides information about the distributions such as the mean value and other parameters. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Study of Whisper Statistics for Criticality Safety Code Validation

Whisper is computational software designed to assist with validation studies with the Monte Carlo particle transport package MCNP. Historical approaches rely on expert based judgment to select benchmarks while Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application model. Whisper will also derive a calculational margin and margin of subcriticality (MOS) that may be used to generate an upper subcritical limit (USL) for a particular application model. Whisper finds benchmark experiments whose isotope sensitivity profiles align best with the MCNP model in question, and assign a correlation coefficient ('ck' in Whisper parlance) to each selected benchmark as a measure of how closely those profiles agree. When ck values are high, there is less uncertainty between model and benchmarks and a higher USL is possible, relative to cases when ck values are low. The default Whisper settings recommends a 100,000 particles per cycle run for 500 active cycles, resulting in 50 M active neutron histories per application run. This is computationally intensive and may not be necessary to produce accurate results. A Whisper statistical study was done varying the neutron histories per cycle from 10,000 to the default 100,000. A lower number of neutron histories will allow Whisper to be used on low memory systems such as laptops. The intent of this paper is to present the changes in the Whisper results as the number of neutrons histories run is varied.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Immersive Visualization for Scientific Data Analysis

We will present the use of immersive visualization at the National Renewable Energy Laboratory (NREL), showcasing how immersive visualization is advancing scientific research and engineering practices and transforming our day-to-day operations. We are leveraging immersive visualization to support scientific discovery and engineering in various domains, including material design, computational fluid dynamics, immersive analytics, grid modernization, digital twins, and situated visualization. We have observed several benefits across four key areas: enhanced spatial judgments, improved understanding through interaction, increased capacity to embed high-dimensional data, and improved collaboration.

immersive analytics↗
DOE OSTI2024

An Approach to Dynamic Human Reliability Analysis and Its Data Collection Framework

Human reliability analysis (HRA) is a method for evaluating human errors in a variety of complex systems such as nuclear power plants, military systems, aircraft, and chemical plants. Most HRA methods currently used by regulatory institutes or utilities are called static HRA and are carried out by simple worksheets or simple calculators. To date, there are many unsolved or intrinsic challenges in static HRA. For example, existing static HRA does not realistically model and evaluate human actions as they would be performed at actual systems. There is no method with HRA to objectively estimate the time required for human actions despite being essential to HRA processes. In addition, many HRA methods still rely on a dataset generated prior to the 1980s, from unrelated industry experience or simply from expert judgment. Accordingly, this study attempted to research how to overcome the challenges of existing HRA via dynamic risk assessment (a.k.a., simulation-based or computation-based risk assessment) techniques. First, this study developed a dynamic HRA method, named as PRocedure-based Investigation Method of EMRALD Risk Assessment – HRA (PRIMERA-HRA). The PRIMERA-HRA mainly concentrates on providing HRA analysts with specific guidelines on how to reasonably model human actions, assign human reliability data and evaluate output of simulation within a dynamic probabilistic risk assessment tool, called as Event Modeling Risk Assessment using Linked Diagrams (EMRALD). Second, this study also developed a module for performance shaping factors (i.e., the key concept in HRA quantification) applicable to dynamic HRA, then implemented it based on PRIMERA-HRA within the EMRALD tool. Third, this study developed an HRA data collection framework to support dynamic HRA, called as Simplified Human Error Experimental Program (SHEEP). Originally, the SHEEP study aimed to support static HRA and its data collection, but recently extended the scope to the new technologies such as dynamic HRA or HRA for advanced reactors. SHEEP focuses on the use of data collected from simplified simulators to complement—but not replace—data collection studies using full-scope simulators and actual operators. To date, many experiments were conducted under the SHEEP framework. Multiple analyses, such as human performance analysis, human error analysis, task complexity analysis, learning effect analysis and time distribution analysis, were also carried out using the collected data. Then, based on the major insights, an approach to inferring full-scope data based on simplified simulator data was proposed. The PRIMERA-HRA and SHEEP research are expected to evaluate human actions more realistically than existing static HRA, provide an opportunity to collect more HRA data with reasonable cost and labor, then contribute to enhance the quality of HRA.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Sensitivity Analysis, Reduced-order Modeling, and Optimization of a Gas-Cooled Pebble Bed Reactor using Equilibrium-Core and DLOFC Performance

This work presents and applies a workflow for performing design optimization on gas-cooled pebble-bed reactors. Based on previous research, a representative equilibrium core of a pebble-bed reactor and a depressurized loss-of-forced-cooling model are created. These applications are built using the Multiphysics Object-Oriented Simulation Environment (MOOSE), specifically utilizing Griffin, Pronghorn, and Bison. After defining design-related parameters and quantities of interest regarding reactor safety and efficiency, this multiphysics model is sampled using the MOOSE stochastic tools module. The result is a comprehensive dataset of configurations, enabling sensitivity analysis and the generation of reduced-order models. Subsequently, the dataset and reduced-order models are employed in an optimization study aimed at maximizing fuel utilization while adhering to safety and operational constraints. The optimization process leads to an improvement of fuel utilization by approximately 10\%, compared to engineering-judgment-based nominal conditions.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2024

An Approach to Dynamic Human Reliability Analysis and Its Data Collection Framework

Human reliability analysis (HRA) is a method for evaluating human errors in a variety of complex systems such as nuclear power plants, military systems, aircraft, and chemical plants. Most HRA methods currently used by regulatory institutes or utilities are called static HRA and are carried out by simple worksheets or simple calculators. To date, there are many unsolved or intrinsic challenges in static HRA. For example, existing static HRA does not realistically model and evaluate human actions as they would be performed at actual systems. There is no method with HRA to objectively estimate the time required for human actions despite being essential to HRA processes. In addition, many HRA methods still rely on a dataset generated prior to the 1980s, from unrelated industry experience or simply from expert judgment. Accordingly, this study attempted to research how to overcome the challenges of existing HRA via dynamic risk assessment (a.k.a., simulation-based or computation-based risk assessment) techniques. First, this study developed a dynamic HRA method, named as PRocedure-based Investigation Method of EMRALD Risk Assessment – HRA (PRIMERA-HRA). The PRIMERA-HRA mainly concentrates on providing HRA analysts with specific guidelines on how to reasonably model human actions, assign human reliability data and evaluate output of simulation within a dynamic probabilistic risk assessment tool, called as Event Modeling Risk Assessment using Linked Diagrams (EMRALD). Second, this study also developed a module for performance shaping factors (i.e., the key concept in HRA quantification) applicable to dynamic HRA, then implemented it based on PRIMERA-HRA within the EMRALD tool. Third, this study developed an HRA data collection framework to support dynamic HRA, called as Simplified Human Error Experimental Program (SHEEP). Originally, the SHEEP study aimed to support static HRA and its data collection, but recently extended the scope to the new technologies such as dynamic HRA or HRA for advanced reactors. SHEEP focuses on the use of data collected from simplified simulators to complement—but not replace—data collection studies using full-scope simulators and actual operators. To date, many experiments were conducted under the SHEEP framework. Multiple analyses, such as human performance analysis, human error analysis, task complexity analysis, learning effect analysis and time distribution analysis, were also carried out using the collected data. Then, based on the major insights, an approach to inferring full-scope data based on simplified simulator data was proposed. The PRIMERA-HRA and SHEEP research are expected to evaluate human actions more realistically than existing static HRA, provide an opportunity to collect more HRA data with reasonable cost and labor, then contribute to enhance the quality of HRA.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Definitions for Testing Whether Evaluated Nuclear Data Relative Uncertainties are Realistic in Size

This document describes various tests that can be used to check whether evaluated relative uncertainties stored in nuclear-data covariances are realistic. To be more specific, these tests check whether nuclear-data uncertainties could be either over- or under-estimated given the input that usually enters the evaluation of nuclear-data mean values and covariances. Warning and error messages on the reliability of nuclear-data uncertainties will result from these tests. If uncertainties of one specific nuclear-data observable trigger warning messages from multiple tests, an evaluator should counter-check the reliability of the relative uncertainties of this particular nuclear-data covariance matrix.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2001

Autogenic Feedback Training Exercise and pilot performance: enhanced functioning under search-and-rescue flying conditions

Studies have shown that autonomous mode behavior is one cause of aircraft fatalities due to pilot error. In such cases, the pilot is in a high state of psychological and physiological arousal and tends to focus on one problem, while ignoring more critical information. This study examined the effect of training in physiological self-recognition and regulation, as a means of improving crew cockpit performance. Seventeen pilots were assigned to the treatment and control groups matched for accumulated flight hours. The treatment group contained 4 pilots from HC-130 Hercules aircraft and 4 HH-65 Dolphin helicopter pilots; the control group contained 3 pilots of HC-130s and 6 helicopter pilots. During an initial flight, physiological data were recorded on each crewmember and an instructor pilot rated individual crew performance. Eight crewmembers were then taught to regulate their own physiological response levels using Autogenic-Feedback Training Exercise (AFTE). The remaining participants received no training. During a second flight, treatment participants showed significant improvement in performance (rated by the same instructor pilot as in pretests) while controls did not improve. The results indicate that AFTE management of high states of physiological arousal may improve pilot performance during emergency flying conditions.

NASA Discipline Neuroscience↗
NASA NTRS2001

Seeing mountains in mole hills: geographical-slant perception

When observers face directly toward the incline of a hill, their awareness of the slant of the hill is greatly overestimated, but motoric estimates are much more accurate. The present study examined whether similar results would be found when observers were allowed to view the side of a hill. Observers viewed the cross-sections of hills in real (Experiment 1) and virtual (Experiment 2) environments and estimated the inclines with verbal estimates, by adjusting the cross-section of a disk, and by adjusting a board with their unseen hand to match the inclines. We found that the results for cross-section viewing replicated those found when observers directly face the incline. Even though the angles of hills are directly evident when viewed from the side, slant perceptions are still grossly overestimated.

NASA Center ARC↗
NASA NTRS2001

Why practice reduces dual-task interference

M. A. Van Selst, E. Ruthruff, and J. C. Johnston (1999) found that practice dramatically reduced dual-task interference in a Psychological Refractory Period (PRP) paradigm with 1 vocal response and 1 manual response. Results from 3 further experiments using the highly trained participants of M. A. Van Selst et al. (1999) support 4 main conclusions: (a) A processing bottleneck exists even after extensive practice; (b) the principal cause of the reduction in PRP interference with practice is shortening of Task 1 bottleneck stages; (c) a secondary cause is that 1 or more, but not all, of the Task 2 substages that are postponed before practice are not postponed after practice (i.e., become automatized); and (d) the extent of PRP reduction with practice depends on the modalities of the 2 responses. A control experiment with 2 manual response tasks showed less PRP reduction with practice than that found by Van Selst et al.

Auditory Perception/physiology↗
NASA NTRS1989

The visually guided control of simulated altitude

Simulated "flights" over three different ground textures were used to examine people's ability to extract optical information useful for active regulation of altitude. The textures were regularly spaced lines as follows: 1) orthogonal to the direction of flight (latitude texture); 2) parallel to the direction of flight (meridian texture); and 3) both parallel and orthogonal (square texture). Visual constant velocity forward flight simulations were displayed on a CRT screen, and subjects asked to maintain one of three initial altitudes using a rate control stick. This task was made difficult by the presence of lateral (irrelevant) and vertical (relevant) "wind gusts." The attitude never varied as winds, forward speed, and vertical rate control resulted in only translational movements. Adjusted root mean square errors (ARMSE) showed altitude regulation was more difficult at higher altitudes and when flying over meridian textures. Refined analysis of a single subject's data showed that this was due both to poorer regulation of the vertical wind disturbance and to a tendency to confuse the lateral wind disturbance for a vertical disturbance.

Vision↗