Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Accident Analysis”

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 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. Further, these IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

A density functional theory analysis of the adsorption and surface chemistry of inorganic iodine species on graphitea

In the event of a nuclear accident, fission products may be released into the environment. The release of 131 I is of particular concern to human health. Iodine can be captured using a number of materials and frequently, this is accomplished with activated carbon impregnated with organic bases. Previous studies have used DFT and the graphite (0001) surface as a surrogate for adsorption, those studies focus on the species I • , I 2 , and CH 3 I. In this work we perform an ab initio study of the adsorption onto the surface of a graphite sheet of I 2 , CH 3 I, and inorganic acidic iodine species (HI, HOI, HIO 2 , and HIO 3 ), which were selected to examine the possible effect of oxidation state on adsorption. The PBE exchange-correlation functional with D3 dispersion was employed. It was found that for molecular iodine, the iodine atoms tended to either situate above the center of a hexagonal site on the graphite or directly atop a carbon atom with the lighter components resting closer to the graphite. For each species the relative binding energies spanned the range of 21–33 kJ mol -1 and graphite-iodine distance was in the range of 3.52–3.93 Å. In all cases we found no significant charge transfer between the iodine species and the graphite, thus we conclude that all the iodine species studied undergo strong physisorption to the graphite.

Ritzmann, Andrew M.↗
NASA NTRS1987

Structural analysis of a bolted joint concept for the Space Shuttle's solid rocket motor casing

The Space Shuttle Challenger accident is thought to have been caused by the failure of one of the tang-clevis joints which join together the casing segments of the Solid Rocket Motors (SRM). Excessive displacement between the tang and clevis, possibly unseating the O-ring seals, may have initiated the resulting accident. An effort was undertaken at NASA's Langley Research Center to design an alternative concept for mating the casing segments. A bolted flanged joint concept was designed and analyzed to determine if the concept would effectively maintain a seal while minimizing joint weight and controlling stress levels. It is shown that under the loading condition analyzed the seal area of the joint remains seated. The only potential stress problem is a stress concentration in the flange at the edge of the bolt hole, which is highly localized. While heavier than the existing joint, this concept does have some advantages which make the bolted joint an attractive alternative.

Lindell, Michael C.↗
NASA NTRS1987

Structural analysis of a bolted joint concept for the space shuttle's solid rocket motor casing

The Space Shuttle Challenger accident is thought to have been caused by the failure of one of the tang-clevis joints joining together the casing segments of the Solid Rocket Motors (SRM). Excessive displacement between the tang and clevis, possibly unseating the O-ring seals, may have initiated the resulting accident. An effort was made at NASA Langley Research Center to design an alternative concept for mating the casing segments. A bolted flange joint concept was designed and analyzed to determine if the concept would effectively maintain a seal while minimizing joint weight and controlling stress levels. It is shown that under the loading conditions analyzed the seal area of the joint remains seated. The only potential stress problem is a stress concentration in the flange at the edge of the bolt hole, which is highly localized. While heavier than the existing joint, this concept does have some advantages making the bolted joint an attractive alternative.

Lindell, Michael C.↗
NASA NTRS1937

Aircraft accidents.method of analysis

This report is a revision of NACA-TR-357. It was prepared by the Committee on Aircraft Accidents. The purpose of this report is to provide a basis for the classification and comparison of aircraft accidents, both civil and military.

Source record↗
DOE OSTI2025

Gas-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the pebble-bed high-temperature gas-cooled reactor (PB-HTGR) reference plant models for the United States Nuclear Regulatory Commission. These models serve as the foundation for the future development of detailed design evaluation models based on license applications. The reference plant models have been developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, which is the code suite proposed for non-light-water reactor systems safety analysis. It incorporates various simulation tools developed by the Nuclear Energy Advanced Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of the work that was performed was to assess BlueCRAB’s level of readiness for modeling a PB-HTGR. To do so, we first developed numerical models in BlueCRAB that include the key physics for this technology to ensure an adequate level of fidelity for modeling PB-HTGR core performance and for performing multiphysics simulations for equilibrium core conditions and different accident scenarios. Then we simulated transient scenarios, including depressurized and pressurized loss of forced cooling accidents, over-cooling, and control rod withdrawal events with delayed and prompt supercritical reactivity insertions. The analysis in this report includes comparisons of the 2D thermal fluid porous media models in Pronghorn and SAM, and comparisons of coupled SAM/Griffin/SAM and coupled Pronghorn/Griffin for depressurized and pressurized loss of forced cooling, over-cooling, and control rod withdrawal events. In addition, we compare 3D, 2D, and 0D/PKE neutronic models for the two control rod withdrawal scenarios with coupled Pronghorn/Griffin. The comparisons show that the BlueCRAB models lead to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields, such as neutron flux, power, temperature, density, pressure, and velocity, are within the expected ranges, and their distributions can be explained by the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. This work further highlights the need for flexible tools with various levels of fidelity to cover the breadth and depth of needs that may arise in future technical evaluations of the PB-HTGR. We believe that the BlueCRAB capabilities will be a significant asset for confirmatory analyses in order to resolve important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1996

An Independent Evaluation of the FMEA/CIL Hazard Analysis Alternative Study

The present instruments of safety and reliability risk control for a majority of the National Aeronautics and Space Administration (NASA) programs/projects consist of Failure Mode and Effects Analysis (FMEA), Hazard Analysis (HA), Critical Items List (CIL), and Hazard Report (HR). This extensive analytical approach was introduced in the early 1970's and was implemented for the Space Shuttle Program by NHB 5300.4 (1D-2. Since the Challenger accident in 1986, the process has been expanded considerably and resulted in introduction of similar and/or duplicated activities in the safety/reliability risk analysis. A study initiated in 1995, to search for an alternative to the current FMEA/CIL Hazard Analysis methodology generated a proposed method on April 30, 1996. The objective of this Summer Faculty Study was to participate in and conduct an independent evaluation of the proposed alternative to simplify the present safety and reliability risk control procedure.

Ray, Paul S.↗
DOE OSTI2022

Semi-Annual Report for Modular Integrated Gas High Temperature Reactor Development during Performance Period October 2021 - March 2022

Modular Integrated Gas High Temperature Reactor (MIGHTR) is being designed by a multidisciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed MIGHTR in 3 years and support its commercialization as a safe and low-cost HTGR. Argonne National Laboratory is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. Work performed by Argonne also includes the thermal hydraulics analysis of the primary heat transport system including reactor pressure vessel (RPV) internals for normal operation, shutdown, and accident conditions. This first semi-annual report summarizes the progress made at Argonne on the two tasks since the start of the project. For the RCCS design task, high-level design work was performed including identification of design requirements for the conceptual MIGHTR RCCS and initial design calculations. The baseline dimensions of the MIGHTR RCCS were derived from scoping analysis results and preliminary steady state performance was estimated. At the initial stage of the primary system analysis task, the primary system thermal fluids model for the MIGHTR core was outlined to use a simplified two-dimensional model to predict the temperature distribution and coolant pressure losses in the fuel assembly blocks.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Comparative analysis of temperature dependent properties of commercial nuclear fuel pellet and surrogates undergoing cracking: A review

A variety of normal and accident scenarios can generate thermal stresses enough to cause cracking in light water reactor (LWR) fuel. To better understand this behavior, cracking experiments have been carried out to induce thermal gradients in fuel pellets via induction heating and direct resistance heating. In this study, Ceria (CeO 2 ) and yttria stabilized zirconia (YSZ) pellets were chosen as a surrogate fuel material for uranium dioxide (UO 2 ). Induction heating was done using copper coils and molybdenum susceptors which heated the surrogates to a threshold temperature that is sufficiently high for the fuel material to conduct current. Thereafter, direct resistance heating was achieved by passing current through the specimen using a DC power supply to introduce volumetric heating to replicate reactor operating conditions. Simultaneous real-time dual imaging of the ceria pellet surface has been designed using optical and infra-red camera system to capture images of cracks and full-field temperature gradients on pellet. It was observed that YSZ pellets need a much higher voltage-current for volumetric heating and do not exhibit the cracking pattern as ceria and UO 2 , hence YSZ was ruled out for further consideration. Ceria was found to be an appropriate surrogate for UO 2 . Furthermore, the experimental set up and test conditions demand the need to primarily understand the thermo-physical, mechanical and optical properties of UO 2 and the surrogates. Thus, it is essential to have an in-depth knowledge about the various temperature dependent properties of UO 2 and CeO 2 for establishing a comparative analysis between the surrogate and UO 2 for conducting experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

University of Missouri Research Reactor (MURR) LEU Fuel Fluid-Structure Interaction Analysis

The University of Missouri Research Reactor (MURR ® ) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that are actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M 3 ) Reactor Conversion Program to convert from highly enriched uranium (HEU, ≥ 20 wt.% U-235) to low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. In this work, Fluid-structure interaction (FSI) analysis at the fuel element level, as compared to the fuel plate level of the previous work. is performed which models all components of the MURR LEU fuel element, including fuel plates and the supporting structures e.g., side plates, end fittings, and combs. Therefore, the effect of supporting structures on the coolant flow distribution, the fuel plate deflection, and the resulting coolant channel gap reduction can be evaluated. In addition to the nominal element geometry and flow rate, the tolerances in the geometry dimensions of coolant channel and plate thickness, the effect of a comb on plate deflection, and the uncertainty of the flow rate per element are considered in this work. The effect of comb on plate displacement is quantified through two bounding cases: the case assuming a perfect bond between the comb and plates and the case neglecting the comb effect. Note that in this analysis, the FSI has been decoupled from the other structural effects caused by irradiation (e.g., swelling and irradiation creep). Assessment of combined effects is planned for a later stage of this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Spatiotemporal features of traffic help reduce automatic accident detection time

Quick and reliable automatic detection of traffic accidents is of paramount importance to save human lives in transportation systems. However, automatically detecting when accidents occur has proven challenging, and minimizing the time to detect accidents (TTDA) by using traditional features in machine learning (ML) classifiers has plateaued. We hypothesize that accidents affect traffic farther from the accident location than previously reported. Therefore, leveraging traffic signatures from neighboring sensors that are adjacent to accidents should help improve their detection. We confirm this hypothesis by using verified ground-truth accident data, traffic data from radar detection system sensors, and light and weather conditions and show that we can minimize the TTDA while maximizing classification performance by considering spatiotemporal features of traffic. Specifically, we compare the performance of different ML classifiers (i.e, logistic regression, random forest, and XGBoost) when controlling for different numbers of neighboring sensors and TTDA horizons. We use data from interstates 75 and 24 in the metropolitan area that surrounds Chattanooga, TN. Our results show that the XGBoost classifier produces the best results by detecting accidents as quickly as 1.0 min after their occurrence with an area under the receiver operating characteristic curve of up to 83% and an average precision of up to 49%. We describe limitations, open challenges, and how the proposed framework can be used for quicker operational accident detection.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 (Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that is used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility simulates nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field may be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility was designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary coolant and secondary coolant system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. To date a test matrix has been completed utilizing the primary system of THETA. These tests, along with computational fluid dynamics and systems code models, determined the heat transfer across the core barrel and intermediate heat exchanger outlet was too great to effectively represent scaled thermal hydraulic phenomena of a liquid metal cooled reactor. Therefore, a significant effort was made to remove the primary system from the METL 28” test vessel #4 and clean the residual sodium from the primary system to facilitate upgrades. Thermal insulation was then incorporated in the core barrel and intermediate heat exchanger outlets. The primary system was then replaced, and a series of tests were performed to assess the performance of the thermal insulation. With primary system testing and upgrades complete, the secondary system could then be brought online. The tube side of the shell-and-tube intermediate heat exchanger was installed onto the primary system flange to begin installing the secondary system. The support structure for the secondary system was then erected on the METL mezzanine alongside the THETA primary system to facilitate installation of the secondary system components (sodium-to-air heat exchanger, flowmeter and pump). The piping and expansion tank were welded into the secondary system. Non-destructive examination of the secondary system welds was completed in order to satisfy ASME B31.3 pipe code for class M process fluids. The heating system and insulation were then added to prepare the system to be filled with sodium. The ancillary electrical equipment was installed which included the pump control box, blower VFD, pipe heater control system, etc. The secondary system will be filled, and a test matrix will be completed in early FY2024.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1973

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

The author has identified the following significant results. The most important result to date is the demonstration of the special value of repetitive ERTS-1 multiband coverage for detecting previously unknown fracture lineaments despite the presence of a deep glacial overburden. The Illinois Basin is largely covered with glacial drift and few rock outcrops are present. A contribution to the geological understanding of Illinois and Indiana has been made. Analysis of ERTS-1 imagery has provided useful information to the State of Indiana concerning the surface mined lands. The contrast between healthy vegetation and bare ground as imaged by Band 7 is sharp and substantial detail can be obtained concerning the extent of disturbed lands, associated water bodies, large haul roads, and extent of mined lands revegetation. Preliminary results of analysis suggest a reasonable correlation between image-detected fractures and mine roof fall accidents for a few areas investigated. ERTS-1 applications to surface mining operations appear probable, but further investigations are required. The likelihood of applying ERTS-1 derived fracture data to improve coal mine safety in the entire Illinois Basin is suggested from studies conducted in Indiana.

Wier, C. E.↗
NASA NTRS1991

Does CRM training work?

Formal cockpit resource management training in crew coordination concepts increases the percentage of crews rated as above average in performance and decreases the percentage of crews rated as below average.

NASA Discipline Space Human Factors↗
NASA NTRS2002

Feral Hogs Management at Merritt Island National Wildlife Refuge: Analysis of Current Management Program

This ST1 Technical Memorandum (TM) summarizes a two-month project on feral hog management in Merritt Island National Wildlife Refuge (MINWR). For this project, feral hogs were marked and recaptured, with the help of local trappers, to estimate population size and habitat preferences. Habitat covers included vegetation cover and Light Detection and Ranging (LIDAR) data for MINWR. In addition, an analysis was done of hunting records compiled by the Refuge and hog-car accidents compiled by KSC Security.

Rosenfeld, Arie↗
NASA NTRS2007

Image Analysis Based on Soft Computing and Applied on Space Shuttle During the Liftoff Process

Imaging techniques based on Soft Computing (SC) and developed at Kennedy Space Center (KSC) have been implemented on a variety of prototype applications related to the safety operation of the Space Shuttle during the liftoff process. These SC-based prototype applications include detection and tracking of moving Foreign Objects Debris (FOD) during the Space Shuttle liftoff, visual anomaly detection on slidewires used in the emergency egress system for the Space Shuttle at the laJlIlch pad, and visual detection of distant birds approaching the Space Shuttle launch pad. This SC-based image analysis capability developed at KSC was also used to analyze images acquired during the accident of the Space Shuttle Columbia and estimate the trajectory and velocity of the foam that caused the accident.

Dominquez, Jesus A.↗
DOE OSTI2022

Overview of recent SCALE activities for Non-LWR inventory and decay heat analysis

In 2019, the US Nuclear Regulatory Commission initiated a project for the development and assessments of non-light-water reactor (non-LWR) accident progression using the SCALE and MELCOR simulation tools. SCALE simulations are used to generate nuclide inventories, full-core power distributions, decay heat, and kinetics parameters to initialize MELCOR simulations of severe accident scenarios. Five non-LWR concepts were studied: high-temperature gas-cooled reactor (HTGR), heat pipe reactor (HPR), high-temperature fluoride salt-cooled reactor (FHR), molten salt-fueled reactor (MSR), and sodium-cooled fast reactor (SFR). This paper summarizes the SCALE results obtained in 2021 for the first three non-LWR concepts, compares characteristics and results to common LWRs, and provides the strategy for the analysis of the remaining two non-LWRs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗