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At least 325 records · Page 18

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NASA NTRS1986

The analysis of airline flight records for winds and performance with application to the Delta 191 accident

Advanced analytical methods applied to airline digital flight records and ATC radar tracking data are providing new insights into atmospheric disturbances that cause serious operating problems for aircraft. Performance-based methods for estimating unmeasured flow angles and the possible influence of rain on lift and drag, and a state-estimation procedure for determining winds along the flightpath are reviewed. The methods are applied to data records from the Delta Flight 191 wind-shear accident at the Dallas/Ft. Worth Airport (August 2, 1985) which involved an L-1011 airliner on final approach. The results indicate that the aircraft encountered a strong downlow followed by a strong outflow accompanied by large and rapid changes in the vertical wind. This pattern of low-level winds is consistent with the model that treats a downburst as a vortex ring. Although rain was present in the downburst, the analysis showed no evidence of performance loss. The results obtained from analyses of airline turbulence encounters provide a means for simulating turbulence phenomena and for training pilots for future encounters.

Bach, R. E., Jr.↗

DOE OSTI2024

Nuclear safety Enhanced: A Deep dive into current and future RAVEN applications

As the horizon of nuclear energy expands with the advent of small modular reactors, IV generation reactors, and fusion reactors, there is a growing perspective that the licensing process could benefit from a more comprehensive approach. Moving beyond traditional deterministic and PRA analysis might pave the way for a novel safety analysis paradigm propelled by the increasing computational power at our disposal. This paper explores different methodologies that can improve the outcomes of nuclear safety analysis. These range from uncertainty quantification techniques, aimed at enhancing the precision of safety margins, to deploying dynamic event trees by driving system code simulations, capturing the potential evolutions of severe accidents. These methodologies introduce innovative dimensions to safety analysis, considering the consequences of postulated events and the dynamics of accident sequences. However, they also bring forth challenges, especially in managing the complexity and sheer volume of potential scenarios. The paper touches upon some strategies to counter these challenges, emphasizing the importance of adaptability and continuous evolution in the face of emerging nuclear safety concerns. Additionally, the paper sheds light on the need for advanced tools to apply these methodologies. Among these tools is RAVEN, an open-source software designed for parametric and probabilistic analyses. Its core components, including distribution, sampler, and reduced order model, enable various applications, from risk assessment and mitigation to dynamic learning and plant control logic simulations.

97 - MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Sensitivity Analysis of Effectiveness of FLEX Strategies Using a CDF-Based Importance Measure Under Accident Conditions

One of the lessons learned from the Fukushima Daiichi nuclear power plant (NPP) accident is strengthening the station blackout (SBO) mitigation capabilities by enhancing defense in depth for all existing and new NPPs. One of the possible remedies is diverse and flexible coping strategies (FLEX). The objective of this study is to address the benefits of FLEX in various accident scenarios in terms of both risk and cost. FLEX was originally devised against SBO accidents. In this research, we investigate the fundamental plant responses against accidents considering two fundamentally different cases: accidents that lead to high pressure on the primary side and accidents that lead to low pressure on the primary side. Several uncertainties are associated with the characteristics of the FLEX portable equipment. Specifically, the time for FLEX deployment may depend on several factors such as type of accident, point of injection, availability of safety systems, battery backup timings, and human actions. This study utilizes a dynamic risk assessment framework to analyze accident scenarios and suggests a novel importance measure, which is a cumulative distribution function– based importance metric that characterizes the influence of input distribution on complete output distribution. The importance of the existing and newly developed FLEX strategy based on risk significance is illustrated with examples. The suggested measure provides clear insight into how FLEX affects risk of the whole system and additional risk margins thanks to new safety systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2009

Hubble Space Telescope Crew Rescue Analysis

In the aftermath of the 2003 Columbia accident NASA removed the Hubble Space Telescope (HST) Servicing Mission 4 (SM4) from the Space Shuttle manifest. Reasons cited included concerns that the risk of flying the mission would be too high. There was at the time no viable technique to repair the orbiter s thermal protection system if it were to be damaged by debris during ascent. Furthermore in the event of damage, since the mission was not to the International Space Station, there was no safe haven for the crew to wait for an extended period of time for a rescue. The HST servicing mission was reconsidered because of improvements in the ascent debris environment, the development of techniques for the astronauts to perform on orbit repairs to damage thermal protection, and the development of a strategy to provide a crew rescue capability. However, leading up to the launch of servicing mission, the HST crew rescue capability was a recurring topic. For HST there was a limited amount of time available to perform a crew rescue because of the limited consumables available on the Orbiter. The success of crew rescue depends upon several factors including when a problem is identified, when and to what extent power down procedures are begun, and where the rescue vehicle is in its ground processing cycle. Severe power downs maximize crew rescue success but would eliminate the option for the orbiter servicing the HST to attempt a landing. Therefore, crew rescue success needed to be weighed against preserving the ability of the orbiter to have landing option in case there was a problem with the rescue vehicle. This paper focuses on quantification of the HST mission loss of crew rescue capability using Shuttle historical data and various power down capabilities. That work supported NASA s decision to proceed with the HST service mission, which was successfully completed on May 24th 2009.

Hamlin, Teri L.↗

NASA NTRS2010

Hubble Space Telescope Crew Rescue Analysis

In the aftermath of the 2003 Columbia accident, NASA removed the Hubble Space Telescope (HST) Servicing Mission 4 (SM4) from the Space Shuttle manifest. Reasons cited included concerns that the risk of flying the mission would be too high. The HST SM4 was subsequently reinstated and flown as Space Transportation System (STS)-125 because of improvements in the ascent debris environment, the development of techniques for astronauts to perform on orbit repairs to damaged thermal protection, and the development of a strategy to provide a viable crew rescue capability. However, leading up to the launch of STS-125, the viability of the HST crew rescue capability was a recurring topic. For STS-125, there was a limited amount of time available to perform a crew rescue due to limited consumables (power, oxygen, etc.) available on the Orbiter. The success of crew rescue depended upon several factors, including when a problem was identified; when and what actions, such as powering down, were begun to conserve consumables; and where the Launch on Need (LON) vehicle was in its ground processing cycle. Crew rescue success also needed to be weighed against preserving the Orbiter s ability to have a landing option in case there was a problem with the LON vehicle. This paper focuses on quantifying the HST mission loss of crew rescue capability using Shuttle historical data and various power down strategies. Results from this effort supported NASA s decision to proceed with STS-125, which was successfully completed on May 24th 2009.

Hamlin, Teri L.↗

DOE OSTI2025

Risk Analysis of Radiological Release from Pu-238 Targets During Manual Handling

Pu-238 Isotope Production Targets are routinely installed in the Advanced Test Reactor (ATR) core, transferred, and stored in the spent fuel canal. These evolutions involve manual handling and manipulation of the targets underwater using long handled tools. The ATR Safety Analysis Report (SAR) postulates a design basis accident which results in damage from manual manipulation of targets, and radiological consequences must be determined for receptors inside the reactor facility, as well as public receptors. This presentation presents the analysis used to determine the radiological consequences due to potential target damage in the ATR canal. The analysis considered radionuclide release fractions, damage ratios for handling evolutions, and entrainment of radionuclides in the canal water.

Advanced Test Reactor↗

DOE OSTI2023

SCALE Modeling of the Sodium Cooled Fast-Spectrum Advanced Burner Test Reactor

This report documents the modeling and simulation of a sodium-cooled fast reactor (SFR) as part of a U.S. Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with the Oak Ridge National Laboratory code SCALE and the Sandia National Laboratories (SNL) code MELCOR. Based on publicly available benchmark specifications, a fully heterogeneous 3D SCALE model of the 250 MWth Advanced Burner Test Reactor (ABTR) was developed to demonstrate SCALE’s capabilities for full-core reactivity analysis, fuel inventory prediction, and decay heat analysis of an SFR. The benchmark specifications contain modeling details for the ABTR core at the beginning of equilibrium cycle (BOEC) at operating conditions; they were derived from a 2006 preconceptual design report produced by Argonne National Laboratory. The ABTR was designed to demonstrate reactor-based transmutation of transuranics, that is, to “burn” transuranics recovered from light-water reactor (LWR) spent fuel. The ABTR’s fuel is designed to operate in 4 month cycles using uranium/transuranic (U/TRU) metallic fuel, with a TRU content of approximately 20%, at a conversion ratio of approximately 0.6. Various reactivity calculations were performed with SCALE for the ABTR and, where possible, compared with results available in the open literature. Additionally, SCALE was used to perform a full-core depletion calculation over the 4 month cycle to obtain the nuclide inventory at the end of equilibrium cycle (EOEC). These nuclide inventories, decay heat, power profiles, and reactivity feedback coefficients at EOEC represent the initial conditions for analyzing severe accident scenarios with MELCOR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2000

U.S. Civil Rotorcraft Accidents, 1963 Through 1997

Narrative summary data produced by the U.S. National Transportation Safety Board (NTSB) has been obtained and analyzed for all 8,436 U.S. civil registered rotorcraft accidents which occurred from mid-1963 through 1997. This analysis was based on the NTSB's assignment of each mishap into one of 21 "first event" categories. The number of U.S. civil registered rotorcraft as recorded by the Federal Aviation Administration (FAA) for the same period has also been obtained. Taken together, these data indicate the civil rotorcraft accident rate (on a per 1,000 registered rotorcraft basis) has decreased by almost a factor of 10 (i.e., from 130 accidents per 1,000 rotorcraft in 1964 to 13.4 per 1,000 in 1997). Analysis of the mishap data indicates over 70% of the rotorcraft accidents were associated with one of the following four NTSB "first event" categories: 2408 Loss of engine power (28.5%); 1,322 In-flight collisions with objects (15.7%); 1,114 Loss of control (13.2%); 1,083 Airframe/component/system failure or malfunction (12.8%).

Harris, Franklin D.↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the Advanced Test Reactor

The Advanced Test Reactor (ATR) is a light water reactor with aluminum-clad driver fuel. A primary mission of the ATR is to support the next generation of nuclear reactors. This support necessarily requires irradiation of advanced materials such as sodium, fuel salts, and metal eutectics. Irradiation of advanced materials in the ATR environment presents a challenge when completing accident analyses and demonstrating compliance to the Safety Analysis Report (SAR). Many advanced materials have the possibility to react with the ATR protective barriers such as the cladding or primary coolant system (PCS) boundary during postulated accident scenarios. Further, molten fuel experiments fall outside of the standard regulatory framework for dose consequence analyses. ATR is currently developing new safety analysis methods to support irradiation of advanced materials. The primary considerations for this development are 1) experiment containment design requirements, 2) primary coolant system response to an experiment containment failure, and 3) dose analyses for molten fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the ATR

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1972

Study of the application of ERTS-A imagery to fracture-related mine safety hazards in the coal mining industry

The author has identified the following significant results. Various data compilation and analysis activities in support of ERTS-1 imagery interpretation are in progress or are completed. These include the compilation of mine accident data, areas of mine roof instability and the analysis of high altitude color infrared photography and low altitude color and color infrared photography which was acquired by NASA in support of the project. The photography reveals that many fracture lineaments are detectable through a varied thickness of glacial till. These data will be compiled on a series of 1:250,000 scale base maps and evaluated for a correlation between fracture zones and mine accidents and rooffalls. Due to high occurrence of cloud cover in the project area and to the delay in imagery shipments, little progress has been made in the analysis of ERTS-1 imagery.

Wier, C. E.↗

DOE OSTI2021

Quantitative Risk Analysis of High Safety Significant Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants using IRADIC Technology

This report documents the activities performed by Idaho National Laboratory (INL) during fiscal year (FY) 2021 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) Risk Assessment project. In FY-2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a strong technical basis to support effective, licensable, and secure DI&C technologies for digital upgrades/designs. An integrated risk assessment technology for the DI&C systems (IRADIC technology) was proposed for this strategy, which aims to (1) provide a best-estimate risk-informed capability to quantitatively and accurately estimate the safety margin obtained from plant modernization, especially for the High Safety Significant Safety-related (HSSSR) DI&C systems, (2) develop an advanced risk assessment technology to support transition from analog to DI&C technologies for nuclear industry, (3) assure the long-term safety and reliability of vital HSSSR DI&C systems, (4) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals and deal with the expensive licensing justifications from regulatory insights, the IRADIC technology is instructive for nuclear vendors and utilities to effectively lower the costs associated with digital compliance and speed industry advances by: (1) defining an integrated risk-informed analysis process for DI&C upgrade, including hazard analysis, reliability analysis, and consequence analysis, (2) applying systematic and risk-informed tools to address common cause failures (CCFs) and quantify responding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the individual level, system level, and plant level, (4) providing insights and suggestions on designs to manage the risks; thus, to support the development, licensing, and deployment of advanced DI&C technologies on nuclear power plant (NPPs). In this report, an approach for performing software CCF analysis, given limited data, is developed and demonstrated using a case study of a highly redundant digital reactor trip system. Consequence analysis is also performed based on different accident scenarios. Results indicate that plant modernization including the improvement of HSSSR DI&C systems will make great benefits to plant safety by providing more safety margins to accident management. In addition, a novel approach is proposed in this report for the quantification of software hazards when sufficient operational and testing data available. The method incorporates software development quality as well as strong analysis techniques to identify and link software defects to potential failure modes. The approach includes both semantic and test-based analysis to detect failures that can exist in different stages of the software development life cycle. This method is applied to an advanced human system interface relevant to reactor trip safety developed from the APR 1400 design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1996

Loss of Situation Awareness in Pilots: Analysis of Incident Reports

Approximately 75% of all aviation accidents and incidents are attributable to human failures in monitoring, managing, and operating system. Tactical decision errors were found to be a factor in 25 of 37 major US air transport accidents between 1978 and 1990. These two facts demonstrate the inability of some pilots to maintain situation awareness. Situation awareness (SA) is defined as 'the perception of elements in the environment within a volume of time and space, the comprehension of their meaning, and the projection of their status in the near future". Thus, when a pilot loses SA, he or she is unable to ether perceive, comprehend, or project the status of the aircraft. In pilots terms, he or she has 'fallen behind the airplane'. Our study this summer involved an analysis of 190 NASA Aviation Safety Reporting System (ASRS) reports.

Eric B. Villeda↗

NASA NTRS2016

Differences in Characteristics of Aviation Accidents during 1993-2012 Based on Flight Purpose

Usually aviation accidents are categorized and analyzed within flight conduct rules (Part 121, Part 135, Part 91) because differences in accident rates within flight rules have been demonstrated. Even within a particular flight rule the flights have different purposes. For many, Part 121 flights are synonymous with scheduled passenger transport, and indeed this is the largest group of Part 121 accidents. But there are also non-scheduled (charter) passenger transport and cargo flights. The primary purpose of the analysis reported here is to examine the differences in aviation accidents based on the purpose of the flight. Some of the factors examined are the accident severity, aircraft characteristics and accident occurrence categories. Twenty consecutive years of data were available and utilized to complete this analysis.

Evans, Joni K.↗

NASA NTRS1982

Analysis of the nature and cause of turbulence upset using airline flight records

The development and application of methods for determining aircraft motions and related winds, using data normally recorded during airline flight operations, are described. The methods are being developed, in cooperation with the National Transportation Safety Board, to aid in the analysis and understanding of circumstances associated with aircraft accidents or incidents. Data from a recent DC-10 encounter with severe, high-altitude turbulence are used to illustrate the methods. The analysis of this encounter shows the turbulence to be a series of equally spaced horizontal swirls known as 'cat's eyes' vortices. The use of flight-data analysis methods to identify this type of turbulence phenomenon is presented for the first time.

Parks, E. K.↗

DOE OSTI2025

The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Application of SCALE to Molten Salt Fueled Reactor Physics in Support of Severe Accident Analyses

As part of a US Nuclear Regulatory Commission–sponsored project to assess the modeling and simulation capabilities for accident progression, source term, and consequence analysis for advanced reactor technologies with SCALE and MELCOR, SCALE was used for the modeling and simulation of a molten salt-fueled reactor (MSR). SCALE capabilities for the modeling of MSR physics were demonstrated based on the Molten Salt Reactor Experiment (MSRE). Of primary interest were the determination of the system’s nuclide inventory, as well as the inventories in the various regions of the loop, considering that the fuel is continuously pumped through the system. This report contains discussions on the following: 1. Determination of the system-average fuel salt inventory considering fission gas removal in the off-gas system and noble metal removal through plating out at the heat exchanger using recent enhancements in SCALE’s depletion sequence TRITON, 2. Assessment of the nuclide spatial distribution throughout the loop using SCALE’s depletion solver ORIGEN, 3. Calculation of the core’s power profile, flux profile, temperature reactivity coefficients, and xenon reactivity using full-core calculations with SCALE’s Monte Carlo code KENO-VI. The results obtained with SCALE were post-processed to provide the MELCOR team with the core inventory and decay heat of the system, as well as the inventory and decay heat of individual regions in the loop, a zone-wise power profile, temperature feedback coefficients, and the xenon worth.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Enhanced Safety of Heavy-Duty Vehicles on Highways through Automatic Speed Enforcement – A Simulation Study

Highway safety remains a significant concern, especially in mixed traffic scenarios involving heavy-duty vehicles (HDV) and smaller passenger cars. The vulnerability of HDVs following closely behind smaller cars is evident in incidents involving the lead vehicle, potentially leading to catastrophic rear-end collisions. Here, this paper explores how automatic speed enforcement systems, using speed cameras, can mitigate risks for HDVs in such critical situations. While historical crash data consistently demonstrates the reduction of accidents near speed cameras, this paper goes beyond the conventional notion of crash occurrence reduction. Instead, it investigates the profound impact of driver behavior changes within desired travel speed distribution, especially around speed cameras, and their contribution to the safety of trailing vehicles, with a specific focus on heavy-duty trucks in accident-prone scenarios. To conduct this analysis, we utilize SUMO, an open-source microscopic traffic simulator, enabling realistic simulations of real-world driving behaviors with and without speed cameras. The simulations also consider empirical distributions of desired headway gaps and reaction times exhibited by human drivers. This study offers a fresh perspective on highway safety by examining how the influence of automatic speed enforcement systems on driver behavior enhances the safety of HDV when faced with challenging situations. By shedding light on these dynamics, we contribute valuable insights to the ongoing research efforts on using technology to create safer roadways, ultimately promoting safer coexistence between heavy-duty and passenger vehicles.

33 ADVANCED PROPULSION SYSTEMS↗