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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 1,657 records · Page 92
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Hanford Double Shell Waste Tank Corrosion Studies (Final Report FY2019)

During fiscal year (FY) 2019, the Savannah River National Laboratory (SRNL) focused on four primary tasks for Hanford Double Shell Tanks (DSTs) corrosion studies. The first task, New Limits, was a continuation from the electrochemical work started in FY16, to expand electrochemical testing to elevated temperatures and elevated hydroxide concentrations and evaluate how these factors will influence pitting corrosion. Using the pitting factor (PF) equation, these new conditions were validated to provide a conservative estimate for the susceptibility of pitting corrosion on legacy carbon steel. The second task, Secondary Liner Corrosion, focused on Vapor Space Corrosion (VSC) and immersion testing studies using two commercially available vapor corrosion inhibitors (VCIs). For this FY, the VCI strategy was applied mid-experiment to determine performance on weathered coupons. The third task, Long-Term Open Circuit Potential (OCP) Drift, was performed using simplified simulated chemistries to evaluate change in pitting corrosion risk due to evolution of corrosion potential for mill-scale and freshly polished 600-grit surfaces. Finally, the fourth task, Microbiologically-Influenced Corrosion (MIC) studies, was focused on the study of leak detection pit (LDP) water sent from the Hanford tank farm for determination of bacteria that can be conducive to corrosion using commercially available kits: BART TM and MICkit® 5. In addition, this FY, a secondary task was added which was a continuation of the study of the Quartz-enhanced Photoacoustic Spectroscopy (QEPAS) system. Work was started but was ultimately not pursued. A summary of each task performed is presented below.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Evaluation of A-105 Waste Properties and Potential Simulants for Confined Sluicing Testing

The alternative retrieval process of confined sluicing is currently being evaluated for retrieving waste from tank A-105 at the Hanford site, which is a single-shell tank known to have previously leaked due to a tear in the tank liner. This report gives a summary of the key waste properties for defining performance based simulants for evaluating confined sluicing and evaluates existing information on A-105 waste to define preliminary estimates for the likely ranges of these key waste properties. Previously developed simulants are assessed, in comparison to A-105 waste, to determine if they are suitable for representing A-105 behavior in planned testing of confined sluicing and/or if new simulants need to be developed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Single-Crystal X-Ray Spectropolarimeter

X-ray spectropolarimetry is a diagnostics technique suitable for measuring intense electron beams and strong magnetic fields in hot, dense plasmas. To increase the accuracy achievable, we investigated the possibility of measuring both orthogonal components of linear polarization simultaneously using a single crystal. We demonstrated experimentally for the first time that crystals with cubic symmetry can be used as polarizing beam splitters. We also found that the polarization-splitting planes in cubic crystals can be accessed in two crystal orientations, a property that can be used to decrease the clutter in polarized spectra. In parallel, we developed a ray-tracing code and are working to add the crystal reflectivity to instrument modeling codes, to help us interpret the spectra and to design spectroscopic and spectropolarimetric instruments. This work is motivated by the need to perform measurements of important quantities in difficult experimental conditions, such as the axial magnetic field at stagnation in magnetized liner inertial fusion loads.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

CY2020 Report for PECOS Activities

All experiments involving the Pecos target chamber in calendar year (CY) 2020 were dedicated to Pre-Heat studies in the context of Magnetized Liner Inertial Fusion (MagLIF). Activities at the target area included actual laser shots but also diagnostic and maintenance, and preparatory work for experiments with cryogenically cooled targets. The latter took up a large part of CY2020’s shot availability, and consequently there were fewer shots performed than in previous years. Since the Z-Beamlet and Z-Petawatt lasers support multiple campaigns, we can only anticipate laser and laser-operator time for one out of three shot windows in a day, and typically not on every day of the week. For that reason, many of our non-shot activities need to be traded against shots as well.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Hanford Double Shell Waste Tank Corrosion Studies FY2020 (Final Report)

For fiscal year (FY) 2020, the Savannah River National Laboratory (SRNL) focused on three experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of deposits on the tanks’ inner sidewalls and bottoms. The second task studied vapor space corrosion (VSC) of the secondary liner using two commercially available vapor corrosion inhibitors (VCIs) at manufacturer recommended dosages. Finally, the third task investigated the long-term open circuit potential (OCP) drift for mill-scale coupons, coupons with polished surfaces, and partial mill-scale coupons that were exposed to tank waste simulants. Additional tests were performed to evaluate the OCP drift as a function of the quantity and composition of organic compounds added to the simulant. A test that was included in the matrix that established the new chemistry limits for tank corrosion, but was previously unreported, was presented and reviewed as well.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Assess and Benchmark Magneto-Inertial Fusion (MIF) Scaling. Final Technical Report for the SNL/Rochester ALPHA Follow-on Project

This project was a follow-on to the Sandia National Laboratories (SNL) and the Laboratory for Laser Energetics (LLE) ARPA-E ALPHA project entitled “Demonstrating Fuel Magnetization and Laser Heating Tools for Low-Cost Fusion Energy”. The primary purpose of this follow-on project was to obtain additional data at the OMEGA facility to help better understand how MagLIF, a platform that has already demonstrated the scientific viability of magneto-inertial fusion, scales across a factor of 1000 in driver energy. A secondary aspect of this project was to extend simulations and analysis at SNL to cover a wider magneto-inertial fusion (MIF) parameter space and test scaling of those models across this wide range of input energies and conditions of the target. This work was successful in improving understanding of how key physics elements of MIF scales and improves confidence in setting requirements for fusion gain with larger drivers. The OMEGA experiments at the smaller scale verified the hypothesis that preheating the fuel plays a significant role in introducing wall contaminants that mix into the fuel and significantly degrade fusion performance. This contamination not only impacts target performance but the optimal input conditions for the target. However, analysis at the Z-scale showed that target performance at high preheat levels is limited by the Nernst effect, which advects magnetic flux from the hot spot, reducing magnetic insulation and consequently reduces the temperature of the fuel. The combination of MagLIF experiments at the disparate scales of OMEGA and Z along with a multiscale 3D simulation analysis has led to new insight into the physical mechanisms responsible for limiting target performance and provides important benchmarks to assess target scaling more generally for MIF schemes. Finally, in addition to the MagLIF related work, a semi-analytic model of liner driven Field Reversed Configuration (FRC) was developed that predicts the fusion gain for such systems. This model was also validated with 2D radiation magneto-hydrodynamic simulations and predicts that fusion gains of near unity could be driven by the Z machine.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Optical Imaging on Z LDRD: Design and Development of Self-Emission and Debris Imagers

We present an overview of the design and development of optical self-emission and debris imaging diagnostics for the Z Machine at Sandia National Laboratories. These diagnostics were designed and implemented to address several gaps in our understanding of visibly emitting phenomenon on Z and the post-shot debris environment. Optical emission arises from plasmas that form on the transmission line that delivers energy to Z loads and on the Z targets themselves; however, the dynamics of these plasmas are difficult to assess without imaging data. Addressing this, we developed a new optical imager called SEGOI (Self-Emission Gated Optical Imager) that leverages the eight gated optical imagers and two streak cameras of the Z Line VISAR system. SEGOI is a low cost, side-on imager with a 1 cm field of view and 30-50 µm spatial resolution, sensitive to green light (540-600 nm). This report outlines the design considerations and development of this diagnostic and presents an overview of the first diagnostic data acquired from four experimental campaigns. SEGOI was fielded on power flow experiments to image plasmas forming on and between transmission lines, on an inertial confinement fusion experiment called the Dynamic Screw Pinch to image low density plasmas forming on return current posts, on an experiment designed to measure the magneto Rayleigh-Taylor instability to image the instability bubble trajectory and self-emission structures, and finally on a Magnetized Liner Inertial Fusion (MagLIF) experiment to image the emission from the target. The second diagnostic developed, called DINGOZ (Debris ImagiNG on Z), was designed to improve our understanding of the post-shot debris environment. DINGOZ is an airtight enclosure that houses electronics and batteries to operate a high-speed (10-400 kfps) camera in the Z Machine center section. We report on the design considerations of this new diagnostic and present the first high-speed imaging data of the post-shot debris environment on Z.

42 ENGINEERING↗
DOE OSTI2021

Application of Shift Ex-Core Calculations to a Detailed MSR Model

This report documents the collaborative research conducted between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) on applying the Monte Carlo neutronics code Shift to calculate the radiation dose rates for a detailed molten salt reactor (MSR) model based on the Molten Salt Breeder Reactor (MSBR). In this work scope, the ANL and ORNL teams worked together and applied Shift to calculate the radiological environments of this detailed MSR facility modeled with explicit geometries. The radiological conditions within the MSR facility were modeled for when the reactor is at two different operation modes: normal full power and drained state. The FW-CADIS hybrid method in Shift was applied successfully to calculate the ex-core neutron and gamma dose rates for the MSBR at full power. Dose rate maps showed that in the current MSBR numerical model, potential pathways exist for neutrons and gammas to stream through the 8-foot concrete shield to reach the top of the MSBR reactor cell. Neutron and gamma dose rates within the drain cell were also calculated for the MSBR at the drained state by integrating the source terms obtained from an ORIGEN-S depletion calculation into the Shift simulation. The results indicate that in the drained state, the delayed gammas from the depleted fuel salt are the main contributors to the dose rates, which suggests that a steel liner in the tank model must be added for calculating the dose rates in the drain cell.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Upgrade of Gamma Spectrometry Systems for ORNL TRISO Fuel PIE

Gamma spectrometry is a key element in much of the post-irradiation examination (PIE) work performed under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program (Demkowicz et al. 2015; Stempien et al. 2021). Gamma spectrometers are integrated into three major capabilities used at the Oak Ridge National Laboratory (ORNL) Irradiated Fuels Examination Laboratory (IFEL) for PIE of tristructural-isotropic (TRISO) coated particles and fuel compacts: the Core Conduction Cooldown Test Facility (CCCTF), the Vertical Counting System (VCS), and the Irradiated Microsphere Gamma Analyzer (IMGA). The CCCTF includes liquid-nitrogen-cooled traps to extract 85 Kr out of the He sweep gas that passes through the furnace in which the fuel compacts are heated during safety testing. Analysis of the 85 Kr activity in the traps is the primary indicator for TRISO failure during safety testing. The VCS is a system used to accurately measure gamma emission from components placed in a lead-shielded chamber. It is used to count the CCCTF deposition cups after removal from furnace. Each cup resides in the CCCTF furnace for typically 12–24 h and is periodically replaced with a fresh cup throughout the safety test. Metallic fission products collect on the water-cooled cups and several gamma-emitting isotopes ( 110 mAg, 134 Cs, 137 Cs, 154 Eu, and 155 Eu) are often measured and provide indication of the retention performance of the TRISO coatings. The VCS is also used to measure the presence of these isotopes on the CCCTF tantalum liner and sweep gas inlet tube for the determination of cup collection efficiency, as well as support other gamma spectrometry needs related to calibration of the 85 Kr fission gas traps and various other special PIE tasks. The IMGA uses gamma spectrometry to measure the inventory of gamma-emitting isotopes in individual TRISO particles. An automated particle handling system within the IMGA hot cell removes each particle from a source vial and positions it in front of a gamma detector, and output from the gamma spectrometer is used by the IMGA software to determine a destination vial such that particles are sorted according to their inventory and retention characteristics. At the conclusion of the AGR-1 and AGR-2 PIE campaigns, the gamma spectrometer systems used at ORNL to support that PIE had reached the end of its life cycle due to gradual obsolescence of the hardware and software. Upgrade of the Canberra Genie 2000 software used by these systems to a Windows 10 version was not a viable option, because the newest Windows 10 version offered by Mirion (the new owner of the Canberra technology) did not include the dynamic-link libraries (DLLs) needed for integration with the custom PIE software used with the CCCTF and IMGA, and Mirion had no current plans for development and release of Windows 10 versions of these DLLs with the Model S560 Genie 2000 Programming Library. Ultimately a switch was made to ORTEC gamma spectrometry systems, which appeared to be a more sustainable solution due to more proactive vendor support. The ORTEC conversion involved replacing the aging detector preamplifier and multichannel analyzer (MCA) hardware, upgrading the obsolete Windows 7 computers to Windows 10 compatible models, adopting ORTEC GammaVision software, and extensive modification of the ORNL-developed Visual Basic .NET (VB.NET) programs that provide the CCCTF and IMGA user interfaces.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Test Report for DPP-1 Regulatory Compliance Testing. Volume 1 - Main Report

Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) for the National Nuclear Security Administration (NNSA), is currently responsible for the regulatory testing of a new Type B fissile material shipping package called the Defense Programs Package (DPP)-1. The DPP-1 is a drum packaging with an inner liner and a removable lid. Both the drum body and the lid are filled with an impact-limiting, thermal-insulating material that protects the inner containment vessel (CV) during normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by 10 US code of Federal Regulations (CFR) 71. In fiscal year 2021, the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP), located at the National Transportation Research Center (NTRC), conducted Type B performance tests as prescribed in Title 10 Code of Federal Regulations Part 71 (10 CFR 71) on seven DPP-1 prototype units. This test report documents test unit (TU) preparation, pretest condition, conditioning of each TU, NCT and HAC testing, and posttest measurements and observations made of the damage resulting from these tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Hanford Double Shell Waste Tank Corrosion Studies (Final Report FY2021)

For fiscal year (FY) 2021, the Savannah River National Laboratory (SRNL) focused on two experimental tasks related to Hanford Double Shell Tank (DST) chemistry and integrity. The first task focused on understanding risk of corrosion due to formation of either continuous layers or discrete patches of deposits on the tanks’ inner sidewalls and bottoms. The second task studied vapor space corrosion (VSC) inhibition of the secondary liner using vapor corrosion inhibitors (VCIs) at dosages less than the manufacturer recommended and a control test in which no VCIs were added.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Atomistic Materials Modeling of High-Pressure Hydrogen Interactions in Ethylene Propylene Diene Monomer (EPDM) Rubber

Elastomeric rubbers serve a vital role as sealing materials in the hydrogen storage and transport infrastructure. With applications including O-rings and hose-liners, these components are exposed to pressurized hydrogen at a range of temperatures, cycling rates, and pressure extremes. Cyclic (de)pressurization is known to degrade these materials through the process of cavitation. This readily visible failure mode occurs as a fracture or rupture of the material and is due to the oversaturated gas localizing to form gas bubbles. Computational modeling in the Hydrogen Materials Compatibility Program (H-Mat), co-led by Sandia National Laboratories and Pacific Northwest National Laboratory, employs multi-scale simulation efforts to build a predictive understanding of hydrogen-induced damage in materials. Modeling efforts within the project aim to provide insight into how to formulate materials that are less sensitive to high-pressure hydrogen-induced failure. In this document, we summarize results from atomistic molecular dynamics simulations, which make predictive assessments of the effects of compositional variations in the commonly used elastomer, ethylene propylene diene monomer (EPDM).

08 HYDROGEN↗
DOE OSTI2021

High-Pressure, Low Temperature Composite Nozzles for Long-Term H2 Dispensing (Final Report)

Final report on evaluation of NanoSonic high pressure hydrogen hoses under CRADA 399. NanoSonic hose liners, fiber winding, and hose end crimp fittings were analyzed for material properties, design, and burst test performance. The hose burst pressures exceeded 100 MPa under -130C sub ambient conditions. Hose crimp fittings were found to have some defects from assembly that impacted burst pressures. The use of digital image correlation provided unique insight to the crimped fitting influence on the composite material and the winding pattern during burst testing.

08 HYDROGEN↗
DOE OSTI2023

Turboexpander for Direct Cooling in Hydrogen Vehicle Fueling Infrastructure

Hydrogen fuel cell electric vehicles (FCEVs) have been identified as one of a few options for zero carbon emissions transportation. A major advantage of FCEVs is that they can fuel quickly and follow a familiar fueling behavior to hydrocarbon-fueled vehicles. Whether light duty or heavy duty, the goal for a hydrogen dispenser is to fuel a vehicle in the same amount of time as the fossil fuel equivalent. When hydrogen is dispensed into the vehicle storage system, however, the temperature rises due to the Joule-Thomson effect and the heat of compression. Typically, vehicles store the compressed hydrogen in composite overwrapped pressure vessels that have a polymer liner with an operational temperature limit of 85°C. This temperature limit can be exceeded during fast fueling if hydrogen is not precooled. Precooling allows for a dispenser to fuel a vehicle at a faster flow rate by preventing the storage tank on the vehicle from overheating. Fueling protocols and requirements are presented in SAE J2601 Fueling Protocols for Light Duty Gaseous Hydrogen Surface Vehicles [1]. A heavy-duty equivalent is under development with similar requirements for precooling. Currently, conventional precooling for light-duty vehicle refueling uses a heat exchanger and chiller to cool the hydrogen gas to -40°C before entering the vehicle. The precooling system represents a significant part of the station capital and operating costs, so if the cost of the precooling system can be reduced by improving its efficiency, the overall station capital and operating cost can be reduced. In this project, National Renewable Energy Laboratory (NREL) and Sandia National Laboratories (SNL) researchers teamed up to investigate the turboexpander precooling application. A turboexpander is a device that places a turbine in a flow path where a pressure differential can be attained. This expansion device will extract work and lower the temperature of the fluid as the pressure reduces. While initial calculations based on established principles showed potential for a turboexpander to generate cooled gas, much work needs to be done to prove the concept. Turboexpanders typically work best under steady state conditions, while the dispenser is a very dynamic flow system. Dynamic turboexpander systems have been proven, such as a turbocharger on a gasoline vehicle. The inlet pressure at a dispenser is also much higher than any other known turboexpander system but should behave similarly to higher density fluids at lower pressures. Having both performed initial calculations, NREL and SNL researchers teamed up to investigate the turboexpander precooling application further. A project was soon built around the idea with SNL performing system modeling using previously proven capabilities and NREL performing hardware characterization with established station capabilities. Creare LLC was contracted as the turbomachinery expert to design and build the concept device. Part way through the project, however, contracting issues with the funding partner caused the project to terminate early before building and characterizing the concept device. While the project could not continue, many key findings were already learned. This paper is a summary of those findings.

08 HYDROGEN↗
DOE OSTI2020

Task 12 PV Sustainability - Human Health Risk Assessment Methods for PV, Part 3: Module Disposal Risks

End-of-life management is important for addressing large future photovoltaic (PV) waste volumes and conserving raw materials for use in new PV modules. In regions without regulatory mandates for PV recycling, end-of-life PV modules can be disposed in accordance with general waste laws. Given the use of various metals as raw materials in semiconductor compounds and electrical contacts in commercial PV modules, some stakeholders have raised concerns regarding potential environmental impacts if PV modules are subject to improper disposal instead of being recycled or disposed in sanitary landfills, as is required in most world regions. To evaluate these concerns, screening-level risk assessment methods are developed herein that evaluate potential human health risks from groundwater and surface (air, soil, surface water) exposure pathways. The methods estimate potential impacts from disposal of end-of-life (EOL) PV modules in non-sanitary landfills under the following worst-case conditions: no leachate collection or groundwater monitoring, no liner for preventing leachate migration, uncovered waste, and lack of stormwater management. Examining worst-case conditions allows the exploration of maximum potential risk to attempt to ensure disposal does not increase health risk above regulatory thresholds. Specifically, this report presents an analysis of potential human health risks associated with non-sanitary landfill disposal for three PV technologies, focusing on release of the highest-prioritized chemical element for each: lead (Pb) in crystalline-silicon (c-Si) PV modules, cadmium (Cd) in thin film cadmium telluride (CdTe) PV modules, and selenium (Se) in thin film copper indium selenide (CIS) PV modules. The prioritization of these chemical elements for analysis is based on stakeholder interest. Because the methodology is chemical-specific, the risk assessment results for these chemicals cannot be directly generalized to other chemicals, although the risk assessment methodology can be applied to other chemicals. If the chemicals chosen are indeed the ones presenting greatest risk, then the results herein should represent the upper bound of health risk from exposure to a single constituent.

14 SOLAR ENERGY↗
DOE OSTI2024

Corrosion Testing of Refractory in Contact with Molten Glasses Designed for Waste Vitrification

It is known that the predictive life of the refractory ceramic liner of nuclear waste glass melters is conservative, as demonstrated by performance of these materials such as in the Defense Waste Processing Facility (DWPF). The motivation for this task is to maximize the useful life of the melters that will be operated at the Waste Treatment and Immobilization Plant (WTP), which will in turn minimize procurement and disposal costs and melter outage times, as well as to identify maximum loadings in the waste glass of those species that corrode melter components. This task was initiated jointly with Pacific Northwest National Laboratory (PNNL) with the objective to develop a methodology and model to enable more accurate prediction of refractory service life under prototypic conditions from laboratory-scale material corrosion tests.

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

Heat Transfer Through a Passive Fire Protective Board from an Impinging Hydrogen Flame

This report documents analysis to determine whether a hydrogen jet flame impinging on a tunnel ceiling structure could result in permanent damage to the Callahan tunnel in Boston, Massachusetts. This tunnel ceiling structure consists of a passive fire protective board supported by stainless steel hangers anchored to the tunnel ceiling with epoxy. Three types of fire protective boards were considered to determine whether heat from the flame could reach the stainless-steel hangers and the epoxy and cause the ceiling structure to collapse. Heat transfer analyses performed showed that the temperature remains constant where the steel hangers are attached to the passive fire protective board. According to these results, the passive fire protective board should provide adequate protection to the tunnel structure in this release scenario. Tunnel structures with similar suspended fire-resistant liner board materials should protect the integrity of the structure against the extremely low probability of an impinging hydrogen jet flame.

08 HYDROGEN↗