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DOE OSTI2025

SAS4A/SASSYS-1 Validation with FFTF Tests

The FFTF liquid sodium cooled test reactor was built to assist the development and testing of advanced fuels and material for fast reactors. The goals of the FFTF Passive Safety Testing program included confirming the safety margins of FFTF as a liquid metal reactor, providing data for computer code validation, and demonstrating the inherent and passive safety benefits of its specific design features. The program included thirteen unprotected loss of flow without scram tests. In preparation for the Passive Safety Testing program, a series of individual reactivity feedback tests were performed which were designed to simulate and validate specific features and reactivity feedbacks of the FFTF core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS1982

High-pressure LOX/hydrocarbon preburner injector investigation

The NASA-MSFC program to furnish a high-pressure LOX/hydrocarbon turbine drive combustor technology base for future booster engine programs is discussed. The program evaluated the LOX/RP-1 and LOX/methane preburner and gas generator performance and gas property prediction methods for fuel-rich and oxidizer-rich operating conditions. A small-scale hardware test program, which included 42 hot-fire tests of fuel-rich LOX/methane and LOX/RP-1 and oxidizer-rich LOX/methane, supplied information to reduce uncertainties in prediction techniques. The hot-fire results show that theoretical computer model features should be modified to support curves generated from experimental data.

Huebner, A. W.↗

NASA NTRS2010

Simulated Service and Stress Corrosion Cracking Testing for Friction Stir Welded Spun Form Domes

Damage tolerance testing development was required to help qualify a new spin forming dome fabrication process for the Ares 1 program at Marshall Space Flight Center (MSFC). One challenge of the testing was due to the compound curvature of the dome. The testing was developed on a sub-scale dome with a diameter of approximately 40 inches. The simulated service testing performed was based on the EQTP1102 Rev L 2195 Aluminum Lot Acceptance Simulated Service Test and Analysis Procedure generated by Lockheed Martin for the Space Shuttle External Fuel Tank. This testing is performed on a specimen with an induced flaw of elliptical shape generated by Electrical Discharge Machining (EDM) and subsequent fatigue cycling for crack propagation to a predetermined length and depth. The specimen is then loaded in tension at a constant rate of displacement at room temperature until fracture occurs while recording load and strain. An identical specimen with a similar flaw is then proof tested at room temperature to imminent failure based on the critical offset strain achieved by the previous fracture test. If the specimen survives the proof, it is then subjected to cryogenic cycling with loads that are a percentage of the proof load performed at room temperature. If all cryogenic cycles are successful, the specimen is loaded in tension to failure at the end of the test. This standard was generated for flat plate, so a method of translating this to a specimen of compound curvature was required. This was accomplished by fabricating a fixture that maintained the curvature of the specimen rigidly with the exception of approximately one-half inch in the center of the specimen containing the induced flaw. This in conjunction with placing the center of the specimen in the center of the load train allowed for successful testing with a minimal amount of bending introduced into the system. Stress corrosion cracking (SCC) tests were performed using the typical double beam assembly and with 4-point loaded specimens under alternate immersion conditions in a 3.5% NaCl environment for 90 days. In addition, experiments were conducted to determine the threshold stress intensity factor for SCC (K1SCC) of Al-Li 2195 which to our knowledge has not been determined previously. The successful simulated service and stress corrosion testing helped to provide confidence to continue to Ares 1 scale dome fabrication.

Stewart, Thomas J.↗

DOE OSTI2022

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. Tests included the use of fuel additives (Sn, Sb, and Pd) for reducing the gettering of lanthanide fission products, zirconium liners to mitigate fuel-cladding chemical interaction, and annular fuel geometry for sodium-free fuel designs. The first cycle of FAST testing was completed and a series of rodlets were removed for the initial post irradiation examination (PIE) of FAST rodlets. The rodlet irradiation conditions was evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % heavy metal [%FIMA]) control rodlets and a helium bonded annular rodlet (4.7 %FIMA). Initial non-destructive PIE has been completed and includes, visual inspection, neutron radiography and gamma scanning of the FAST rodlets. Radiography confirmed the integrity of the experiments as well as showing the complete filling of the annulus in the annular fuel at a modest burnup (4.7 %FIMA) and potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior with the exception of the ceasium in the He-bonded annular fuel. Future destructive PIE will be necessary in order to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1982

Assessment of a 40-kilowatt stirling engine for underground mining applications

An assessment of alternative power souces for underground mining applications was performed. A 40-kW Stirling research engine was tested to evaluate its performance and emission characteristics when operated with helium working gas and diesel fuel. The engine, the test facility, and the test procedures are described. Performance and emission data for the engine operating with helium working gas and diesel fuel are reported and compared with data obtained with hydrogen working gas and unleaded gasoline fuel. Helium diesel test results are compared with the characteristics of current diesel engines and other Stirling engines. External surface temperature data are also presented. Emission and temperature results are compared with the Federal requirements for diesel underground mine engines. The durability potential of Stirling engines is discussed on the basis of the experience gaind during the engine tests.

Cairelli, J. E.↗

DOE OSTI2022

Estimation of China Fusion Engineering Test Reactor performance and burning fraction in different pellet fueling scenarios by a multi-species radial transport model

Tritium self-sufficiency in future deuterium–tritium fusion reactors is a crucial challenge. As an engineering test reactor, the China Fusion Engineering Test Reactor requires a burning fraction of 3% for the goal to test the accessibility to the future fusion plant. To self-consistently simulate burning plasmas with profile changes in pellet injection scenarios and to estimate the corresponding burning fraction, a one-dimensional multi-species radial transport model is developed in the BOUT++ framework. Here several pellet-fueling scenarios are then tested in the model. The results show that the increased fueling depth improves the burning fraction by particle confinement improvement and fusion power increase. Nevertheless, by increasing the depth, the pellet cooling-down may significantly lower the temperature in the core region. Taking the density perturbation into consideration, the reasonable parameters of the fueling scenario in these simulations are estimated as pellet radius ${r}_{{\rm{p}}}=3\,{\rm{mm}},$ injection rate $=\,4\,\mathrm{Hz},$ and pellet injection velocity $=\,1000\mbox{-}2000\,{\rm{m}}\,{{\rm{s}}}^{-1}$ without drift or $450\,{\rm{m}}\,{{\rm{s}}}^{-1}$ with high-field-side drift.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Local Power Impact Experiment Design for a New Fuel Type for use in the Advanced Test Reactor

The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. Reactor physics safety evaluations currently use Monte Carlo for the 21st Century (MC21), a continuous-energy Monte Carlo radiation transport code [3]. Existing MC21 models of the ATR and ATRC cores have a validation basis for use in neutronics analyses with HEU fuel. The models are used to support safety analyses that include comparisons to the safety requirements for the reactors. However, the use of the LOWE element in the ATR and ATRC is not currently covered by the current model validation basis. To deploy the new fuel type, extensive computational reactor physics support is necessary to support the use of LOWE in the ATR and ATRC. Therefore, LOWE requires a rigorous validation basis, aligned with that of HEU fuel, that takes advantage of the existing software tools and processes currently used for the ATR and ATRC. The experiment to validate of the MC21 models for determining power, the Power Impact Validation Experiment, will consist of two flux runs in the ATRC, one with fully HEU loading and one with a single LOWE element. Both flux runs will be instrumented with 20 sets of azimuthal fission wires and 3 sets of axial fission wires, as shown in Figure 4. Standard flux run methodology will be used [4]. Power Impact Validation Experiment data will be compared against MC21 calculated data, both for absolute fission rate accuracy and to determine the relative change in fission rates between the two runs. The results of the Power Impact Validation Experiment and subsequent evaluations will provide the validation basis for MC21 for use with LOWE elements. Key features of the Power Impact Validation Experiment include: (1) Two flux runs to allow for LOWE perturbed measurements to be compared to already validated measurements taken from a full core of HEU fuel, (2) Optimization of instrumentation to balance analytical needs with practical considerations (e.g., limited time window to count beta particles from fission products), and (3) Standard ATRC core loading, including both driver positions and flux traps, to minimize cost while remaining representative of typical ATR core loading.

42 ENGINEERING↗

DOE OSTI2024

Thermal Conductivity Measurements on Irradiated TRISO Fuel at IMCL

INL is leading the development and testing of accident tolerant fuels: fuels that resist melting down and that contain radioactive byproducts, preventing their release into their environment. One such fuel, called TRi-structural ISOtropic particle fuel (TRISO) See Fig 1. Understanding how heat propagates through a fuel is essential to fuel development and qualification. When irradiated in a reactor the properties that affect heat transport change. Our objective is to quantify the change in thermal properties of TRISO after irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Evaluation of the Dissolution Behavior of L-Bundle End Caps and HFIR Fuel Carriers

The H-Canyon facility is currently using the 6.1D dissolver for the dissolution of Material Test Reactor (MTR) fuel and the 6.4D dissolver for the dissolution of High Flux Isotope Reactor (HFIR) fuel using mercury-catalyzed HNO₃ dissolution flowsheets. The processing strategy for both dissolvers involves the dissolution of multiple charges of fuel per batch. After the designated heating cycle, the dissolvers are opened, and the charging wells are probed to determine if the MTR or HFIR fuel has dissolved. If undissolved fuel fragments are beyond a certain height, the dissolver must be closed and heated for an additional amount of time to dissolve the remaining material. In recent MTR fuel dissolutions, “high probes” (i.e., excessive undissolved material) were frequently observed, which resulted in extended dissolution times. The suspected cause of the high probes was the incomplete dissolution of the L-Bundle End Cap, rather than the fuel or fuel bundles. The End Cap is hypothesized to be binding in the insert well and not dropping into the acid as the L-Bundle and fuel dissolve. Once the End Cap is dislodged by the probe and drops into the acid, the dissolution rate of the End Cap appears to be significantly reduced compared to the dissolution rate of the fuel and other parts of the L-Bundle. A similar issue has also been observed with the lifting bail on the HFIR fuel carriers. During HFIR fuel dissolutions, the lifting bail on the outer carrier has resulted in high probes due to incomplete dissolution. In one case, a partially dissolved bail was caught in one of the insert well holes which prevented the probe from going to the bottom of the well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

PBX-9501 High Explosive Infrared Thermography Test Report

High explosive safety requirements put in place by the Department of Defense necessitate that Los Alamos National Laboratory verify that high melting explosive-based plastic bonded explosive 9501 will not present a health or safety risk to personnel when exposed to various accident scenarios. The JT5R-14C test is being developed to validate this requirement in a scenario where PBX-9501 is exposed to a hydrocarbon fuel fire. Preliminary testing is required to establish benchmark data on the burning characteristics of PBX-9501. The PBX-9501 High Explosive Infrared Thermography test was designed to provide the benchmark burning characteristics of PBX-9501 in a simulated fuel fire environment using propane burners as a fuel source. The PBX-9501-HEIT was successfully completed using a 20.32 cm tall cylindrical sample of PBX-9501 with a radius of 10.16 cm, a volume of 6,589.63 cm 3 , a density of 1.83 g/cm 3 , and a mass of 1.2 kg. The PBX-9501-HEIT test was captured and recorded video sequence using a Telops Inc., Multispectral Infrared camera. The infrared sequence was reviewed to identify key events throughout the test and serves as a visual tool to cross reference against future testing to aid in the identification of PBX-9501 high explosive ignition and deflagration characteristics. The data obtained in the thermographic sequence confirmed that the temperatures emitted by the PBX-9501 were higher than the temperatures of the propane burner ignition source. The propane burner flames had a temperature range of 700-850 °C. The cylindrical sample of PBX-9501 burned at an average temperature of 1000-1200 °C on the surface with off-gassing temperatures averaging 1100-1250 °C. The heat map of the PBX-9501 obtained through the test was between 250-550 °C higher than the propane gas temperatures. The temperature differential between the PBX-9501 and the propane flame allowed for the identification of key burning features.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

NASA NTRS1982

Additional experiments on flowability improvements of aviation fuels at low temperatures, volume 2

An investigation was performed to study flow improver additives and scale-model fuel heating systems for use with aviation hydrocarbon fuel at low temperatures. Test were performed in a facility that simulated the heat transfer and temperature profiles anticipated in wing fuel tanks during flight of long-range commercial aircraft. The results are presented of experiments conducted in a test tank simulating a section of an outer wing integral fuel tank approximately full-scale in height, chilled through heat exchange panels bonded to the upper and lower horizontal surfaces. A separate system heated lubricating oil externally by a controllable electric heater, to transfer heat to fuel pumped from the test tank through an oil-to-fuel heat exchanger, and to recirculate the heated fuel back to the test tank.

Stockemer, F. J.↗

DOE OSTI2025

Fracture toughness measurement of ceramics with fuel pellet geometry

The crack initiation of nuclear fuel pellets, driven by extreme temperature gradients, is a critical factor impacting reactor performance. Although this fact is widely acknowledged, there is limited fracture toughness data for specimens with representative geometry and fabricated using relevant processing methods. Here, to address this, a new chevron-notched short-rod specimen configuration with diameter and length both equal to 10 mm, representative of a conventional fuel pellet and tested in tension, was developed in an attempt to validly measure Mode I fracture toughness, K Ic . Test calibration was carried out by testing polycrystalline ZrO 2 , SiC, and Si 3 N 4 ceramics fabricated with conventional fuel pellet geometry and known fracture toughnesses. The results suggest that the method described in this paper produces valid K Ic measurements using conventional ceramic fuel pellets.

Ceramics↗

DOE OSTI2022

MOOGLE: A Multi-Objective Optimization tool for three-dimensional nuclear fuel assembly design

MOOGLE is a new genetic algorithm methodology for the three-dimensional design of nuclear fuel assemblies. MOOGLE uses common fuel rod types as the decision variable to develop a suite of three-dimensional fuel assemblies to provide optimized solutions to the design problem. Pressurized Water Reactor (PWR) fuel assemblies were optimized using IFBA and gadolinium (Gd 2 O 3 ) as burnable poisons in order to compare how burnable poison choice affects optimization results. Boiling Water Reactor (BWR) fuel bundles were also optimized using three unique fuel rod palettes to study how the size of the design space affects optimization results. Burnable poison analysis showed that utilizing IFBA and Gd 2 O 3 as burnable poisons produced the best and widest range of optimized solutions. Further, BWR fuel bundle optimization results indicate that the inclusion of additional fuel rod types produced a wider solution space but did not improve optimization results for regions explored using fewer unique fuel rods. These tests demonstrate MOOGLE's ability to analyze the tradeoffs between the inclusion of different fuel elements and their effects on assembly performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS2004

A Summary of DOD-Sponsored Research Performed at NASA Langley's Impact Dynamics Research Facility

The Impact Dynamics Research Facility (IDRF) is a 240-ft.-high gantry structure located at NASA Langley Research Center in Hampton, Virginia. The IDRF was originally built in the early 1960's for use as a Lunar Landing Research Facility. As such, the facility was configured to simulate the reduced gravitational environment of the Moon, allowing the Apollo astronauts to practice lunar landings under realistic conditions. In 1985, the IDRF was designated a National Historic Landmark based on its significant contributions to the Apollo Moon Landing Program. In the early 1970's the facility was converted into its current configuration as a full-scale crash test facility for light aircraft and rotorcraft. Since that time, the IDRF has been used to perform a wide variety of impact tests on full-scale aircraft, airframe components, and space vehicles in support of the General Aviation (GA) aircraft industry, the U.S. Department of Defense (DOD), the rotorcraft industry, and the NASA Space program. The objectives of this paper are twofold: to describe the IDRF facility and its unique capabilities for conducting structural impact testing, and to summarize the impact tests performed at the IDRF in support of the DOD. These tests cover a time period of roughly 2 1/2 decades, beginning in 1975 with the full-scale crash test of a CH-47 Chinook helicopter, and ending in 1999 with the external fuel system qualification test of a UH-60 Black Hawk helicopter. NASA officially closed the IDRF in September 2003; consequently, it is important to document the past contributions made in improved human survivability and impact tolerance through DOD-sponsored research performed at the IDRF.

Jackson, Karen E.↗

NASA NTRS2004

A Summary of DOD-Sponsored Research Performed at NASA Langley's Impact Dynamics Research Facility

Jackson, Karen E.↗

DOE OSTI2018

Control of Ionomer Distribution in Roll-to-Roll Coated Fuel Cell Catalyst Layers

Fuel cell catalyst layers are a complex mixture of catalyst particles and ion-conducting polymers. The performance of catalyst layers is highly dependent on microstructure, which must balance proton and electron conduction with gas transport. As fuel cells transition from the laboratory to industrial production for vehicles and other applications, there is a need to understand how manufacturing processes impact the microstructure and, ultimately, the performance of catalyst layers. In this work, we have focused on the roll-to-roll coating of catalyst layers for proton exchange membrane fuel cells. The catalyst layers are responsible for the electrochemical reactions that enable power production. In these devices the catalyst is commonly a Pt nanoparticle supported on porous carbon blacks. The ion conducting polymer, or ionomer, is typically a perfluorosulfonic acid. The most common being Nafion. To fabricate the catalyst layer, the catalyst particles are dispersed in an ink with the ionomer and cast to form the catalyst layer. Commonly, laboratory samples are prepared using spray coating or hand painting, which effectively coat many thin layers to achieve the desired catalyst layer thickness. This creates catalyst layers with uniform distributions of materials throughout the thickness of the full catalyst layer. In contrast, roll-to-roll methods, like slot die, gravure or screen printing, coat the catalyst layer as a single wet film. As the film dries, ink constituents are able to segregate, leading to a heterogeneous distribution of materials through the thickness of the layer. It is also known that the choice of solvent and drying rate can affect the morphology of the catalyst layer. Thus, there is a need to understand how ink formulation and the drying process influence the distribution of materials and catalyst layer morphology. Here, we have explored the influence of solvent and drying rate on morphology of roll-to-roll coated catalyst layers. We show the development of the Kelvin probe method as a rapid screening technique for qualitative analysis of ionomer distribution. We also utilize nano X-ray computed tomography to visualize electrode structure and to quantify particle-size and pore-size distributions, thickness-dependent ionomer distribution, tortuosity, and effective transport properties. We find that solvent has a strong influence on ionomer distribution, with less of an effect on porosity. Conversely, drying temperature has a strong influence on porosity, but less influence on ionomer distribution. Finally, we utilize in situ fuel cell performance testing and other advanced diagnostics to quantify the impact of catalyst layer properties on fuel cell performance and demonstrate that roll-to-roll coating is capable of coating high performance catalyst layers in multi-meter lengths.

14 SOLAR ENERGY↗

DOE OSTI2019

Minimizing Transuranic Waste Generated in Hot Cells at Oak Ridge National Laboratory Through Use of a Non-Destructive CO{sub 2} Pellet Cleaning System - 19577

Wastes generated in shielded hot cells containing materials contaminated with transuranic (TRU) waste present unique problems in waste disposal at Department of Energy (DOE) sites. Oak Ridge National Laboratory (ORNL) routinely processes materials contaminated with TRU waste from the High Flux Isotope Reactor (HFIR) and from other sources (both on and off-site) in their hot cell facilities, generating wastes that require eventual disposal. The waste streams generated by these processes are often high activity and exhibit high dose rates, driving them into the remote-handled (RH) category at the Waste Isolation Pilot Plant (WIPP). Reduction of the TRU waste footprint is desirable from a disposal cost perspective due to constraints with RH waste processing and a lack of space availability at the WIPP. Historically TRU waste generated at ORNL has been processed by the TRU Waste Processing Center (TWPC), which belongs to the Department of Energy's Office of Environmental Management (DOE-EM). Now that Oak Ridge legacy TRU waste is reaching its final stage, DOE-EM has turned over the mission of processing newly-generated TRU waste streams to the DOE Office of Science at ORNL. To reduce the volume of potential TRU waste occupying valuable space in the ORNL Hot Cells, a non-destructive CO{sub 2} pellet cleaning system will be used to capture particulate contaminants removed from the surfaces of waste inside the ORNL Non-Reactor Nuclear Facilities (NNFD) Hot Cells at the Irradiated Fuels Examination Laboratories (IFEL). The particulate will be segregated by particle size and captured for disposal later. This action will separate some of the TRU isotopes (particulate) from the larger waste volume (debris waste items), resulting in both lower waste disposal costs and reduced TRU waste volumes. This will be accomplished through a specialized cyclone separator. Secondary wastes are usually generated through other methods of decontamination, such as water jets and abrasive agents. These secondary wastes are absent with the CO{sub 2} pelletized system, due to sublimation of the dry ice upon impact. This is an important consideration with IFEL, a facility whose primary mission is to perform a series of R and D testing on nuclear fuel rods. Liquid is not a feasible method of decontamination as the IFEL's drains in the Hot Cells were plugged years ago due to Resource Conservation and Recovery Act (RCRA) regulations. Constraints at IFEL, such as criticality safety concerns with use of hydrogenous materials, are also alleviated with the non-destructive CO{sub 2} pellet cleaning system. These are a couple of the considerations why this method of decontamination is preferred. Due to concerns with oxidizers at WIPP grouting will be performed on the contaminated particulate produced during the trapping by the cyclone separation phase. This method will comply with mitigating oxidizing properties in TRU waste, required by the recently published Basis of Knowledge (BoK) document as part of Acceptable Knowledge (AK) by WIPP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1931

A Method for Reducing the Temperature of Exhaust Manifolds

This report describes tests conducted at the Langley Memorial Aeronautical Laboratory on an "air-inducting" exhaust manifold for aircraft engines. The exhaust gases from each cylinder port are discharged into the throat of an exhaust pipe which has a frontal bellmouth. Cooling air is drawn into the pipe, where it surrounds and mixes with the exhaust gases. Temperatures of the manifold shell and of the exhaust gases were obtained in flight for both a conventional manifold and the air-inducting manifold. The air-inducting manifold was installed on an engine which was placed on a test stand. Different fuels were sprayed on and into the manifold to determine whether the use of this manifold reduced the fire hazard. The flight tests showed reductions in manifold temperatures of several hundred degrees, to values below the ignition point of aviation gasoline. On the test stand when the engine was run at idling speeds fuels sprayed into the manifold ignited. It is believed that at low engine speeds the fuel remained in the manifold long enough to become thoroughly heated, and was then ignited by the exhaust gas which had not mixed with cooling air. The use of the air-inducting exhaust manifold must reduce the fire hazard by virtue of its lower operating temperature, but it is not a completely satisfactory solution of the problem.

Schey, Oscar W↗