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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O 2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each P O$_2$ , with activation energies of 228 ± 7 kJ/mol for 20 kPa O 2 and 188 ± 8 kJ/mol for 0.2 kPa O 2 . At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O 2 , oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low P O$_2$ , leading to SiO 2 buildup in porous nodules. Here, this phenomenon was not observed at any temperature in 20 kPa O 2 environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Sensitivity analysis of thermal contact conductance modeling to inform MiniFuel irradiation capsule designs

The MiniFuel irradiation platform has been developed by Oak Ridge National Laboratory as a flexible, high-throughput separate effects testing capability within the High Flux Isotope Reactor (HFIR). Finite element thermal models are relied upon to design MiniFuel experiments to achieve a specific time-averaged irradiation temperature for experimental objectives. A previous study identified that uncertainty in the component heat generation rates and thermal contact conductance (TCC) model are the most significant contributors to predicted fuel temperature variance. To address both sources of uncertainty, this work performs sensitivity analysis on the TCC model to identify high-impact, high-uncertainty parameters that contribute to fuel temperature variance. The TCC model is analyzed in increasing detail, first using a standalone Python code, then again after coupling Python to the BISON fuel performance code. Furthermore, the parameters with the largest contributions to fuel temperature variance which can be reduced through design changes are identified as the initial subcapsule gas pressure, contact pressure between the fuel and dish, and the effective surface roughness of the interface. A set of design recommendations for future capsule designs has been established and applied to reduce the previously quantified average fuel temperature uncertainty ranges of ± 40 °C in the HFIR vertical experiment facilities (VXF) and ± 80 °C in the removable beryllium (RB) reflector to approximately ± 32 °C and ± 53 °C, respectively. This equates to a 21 % and 33 % reduction in the uncertainty range of the average fuel temperature for VXF and RB, respectively.

BISON↗
NASA NTRS1992

Progress in SP-100 tribological coatings

The SP-100 Tribology Development program to investigate candidate coating materials for the SP-100 reactor is discussed. The SP-100 will operate at temperatures up to 1500 K in high vacuum. Background information, specific bearing requirements, the basis for coating material selection, testing methods, and initial results of the program are addressed.

Ring, Peter J.↗
DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Water vapor oxidation behaviors of nuclear graphite IG-110 for a postulated accident scenario in high temperature gas-cooled reactors

Water leakage in accidental conditions of high temperature gas-cooled reactors is one of the most critical problems that can compromise the integrity of different nuclear components. In this study, oxidation behaviors of nuclear graphite IG-110 in water ingress accidental conditions were investigated. Mass loss and oxidation rates were evaluated after oxidation tests at temperatures up to 1400 °C in an Ar-20 vol% H 2 O mixed atmosphere. The activation energy decreased from 318.6 to 148.9 kJ/mol with temperature, indicating two different oxidation regimes. The cross-sections of the oxidized samples were systematically characterized. The corresponding logarithmic porosity profiles showed a temperature dependency. Pore formation moved toward near-surface regions with increasing temperature and preferential binder oxidation, with filler particle degradation. Furthermore, oxidant concentration profiles and oxidation depths were estimated using a theoretical model and compared with the experimental results. This work provides important benchmark data and safety analysis guidance for the accident scenario in high temperature gas-cooled reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels

Thermomechanical processing (TMP) of ferritic–martensitic (FM) steels, such as HT9 (Fe–12Cr–1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2024

High Performance Heat Pipe Power Transient Testing at SPHERE Facility

Microreactors are being researched, designed, and built at Idaho National Laboratory (INL). Microreactors are small reactors defined at less than 20MW of power. These reactor concepts are also being looked at throughout industry for various applications. An important aspect of these reactor designs is economic feasibility i.e. lower overnight capital cost. The driving factors for implementing microreactors are quick setup and takedown, minimal operators, and the ability to manufacture them readily and to fit in mid-sized containers for transport. A specific area of research to aid in successful integration of these factors within the designs is passive heat removal of the core’s thermal power. Interest in heat pipes to achieve this passive heat removal has been shown across multiple industry partners. Because of this interest, INL has developed a test facility to facilitate experimental tests for sodium filled heat pipes. INL has developed the Single Primary Heat Extraction and Removal Emulator (SPHERE) facility to run experiments on high performance, sodium filled heat pipes. As mentioned above, heat pipes are passive heat transfer devices. Radially, heat pipes are broken up into an outer wall, a small annular gap, a wick structure, and a centerline gap. They function by utilizing latent heat transfer. Heat pipes are traditionally separated into three regions, an evaporator (heat input), an adiabatic region, and finally a condenser region (heat removal). As heat is being applied to the evaporator, the working fluid undergoes a phase change to a vapor. This phase change causes a differential pressure across the axial length of the pipe driving flow down the center gap of the heat pipe. The vapor flows down past the adiabatic region to the condenser where the heat is removed. This heat removal forces the working fluid to phase change back to a liquid. The wick structure is then utilized to drive the flow back towards the evaporator by capillary forces. This backflow is aided by the annular gap. Because this heat transfer mechanism functions with latent heat transfer, the heat pipe is close to isothermal down the axial length. Heat pipes can operate under a wide range of working fluids. Considerations for these working fluids are primarily driven by operating temperatures amongst other important factors based around overall performance. Sodium filled heat pipes operate from 450°C up to 900°C. This temperature range works well for the current microreactor designs. In conjunction with this experimental capability, INL has developed a modeling software to simulate heat pipe physics within reactor cores. This modeling software is called Sockeye and functions under the established INL Multiphysics Object Oriented Simulation Environment (MOOSE). SPHERE also supports Sockeye development by providing the modeling team with experimental data on an array of setups and operating parameters to support validation efforts. A power transient experiment was performed utilizing the SPHERE facility to continue to aid with Sockeye development. The testing followed a proposed test plan to ramp up and down the temperature of the heat pipe. Sockeye models steady state heat pipe operation with high accuracy, the data provided by the power transient testing aims to assist with the validation efforts and further enhance transient modeling capability of the tool [2].

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

ASME Section III, Division 5, Class A 100,000-hour design data for Alloy 709

This report documents the 100,000 hour, Class A ASME design material data for Alloy 709 based on the extensive Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies qualification test campaign. This report includes design tensile properties, creep rupture data, allowable stresses, isochronous stress-strain curves, buckling charts, and a few additional miscellaneous pieces of design data. Companion work at Oak Ridge National Laboratory and Argonne National Laboratory provide design cyclic data --- fatigue charts and creep-fatigue diagrams --- and an inelastic constitutive model. This work substantially completes the ASME data package for the Alloy 709 Code Case, though the design data will continue to updated as the final tests finish. This report also compares the design performance of Alloy 709 against that of 316H stainless steel to provide a reference for the improved high temperature strength of 709 compared to a reference material for sodium fast reactor construction.

36 MATERIALS SCIENCE↗
NASA NTRS2011

Design of a Mechanical NaK Pump for Fission Space Power

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Mireles, Omar R.↗
DOE OSTI2023

ASME Code Qualification Plan for LPBF 316 SS

This report describes a plan to qualify laser powder bed fusion (LPBF) 316 stainless steel for use with the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code Section III, Division 5 rules for metallic components in high temperature nuclear reactors. Accomplishing this goal would make the material and manufacturing process available to vendors for inclusion in the next generation of advanced, high temperature reactors. The general approach adopted here is to treat LPBF 316 as if it was a completely new material and to develop a plan to qualify the material according to the current ASME practices. One key goal of this work is to explore and develop accelerated qualification approaches that might reduce the time required to qualify new materials by reducing the need for long term testing. However, the qualification plan here does not employ any accelerated qualification approaches to provide a limiting, bounding description of the number, duration, and types of testing required to qualify LPBF 316 without such techniques and to describe a comprehensive dataset that could be used to explore and validate accelerated qualification approaches in the future. The report addresses the fundamental challenges to qualifying Advanced Manufacturing (AM) materials for high temperature applications and summarizes the ASME Section III qualification process as well as current efforts to qualify LBPF and DED 316 for low temperature applications. The report then discusses specific issues, both material and logistical, related to qualifying PBF 316 steel. The final chapters of the report describe a complete test plan designed to generate sufficient data to qualify the material as well as a data management plan for how to store and manage the data to eventually provide the test data packaged needed to qualify the material with ASME.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1983

Clean catalytic combustor program

A combustor program was conducted to evolve and to identify the technology needed for, and to establish the credibility of, using combustors with catalytic reactors in modern high-pressure-ratio aircraft turbine engines. Two selected catalytic combustor concepts were designed, fabricated, and evaluated. The combustors were sized for use in the NASA/General Electric Energy Efficient Engine (E3). One of the combustor designs was a basic parallel-staged double-annular combustor. The second design was also a parallel-staged combustor but employed reverse flow cannular catalytic reactors. Subcomponent tests of fuel injection systems and of catalytic reactors for use in the combustion system were also conducted. Very low-level pollutant emissions and excellent combustor performance were achieved. However, it was obvious from these tests that extensive development of fuel/air preparation systems and considerable advancement in the steady-state operating temperature capability of catalytic reactor materials will be required prior to the consideration of catalytic combustion systems for use in high-pressure-ratio aircraft turbine engines.

Ekstedt, E. E.↗
DOE OSTI2022

Coupled Multiphysics Primary Loop Simulations of the Mk1-FHR in the Virtual Test Bed

To support advanced reactor demonstrations, the Virtual Test Bed (VTB) [1] repository hosts a wide range of challenge problems for showcasing modeling and simulation capabil- ities in support of advanced reactor demonstrations. This document presents a coupled multiphysics model of the Mark 1 pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR). The analysis leverages NEAMS tools (Griffin [2], SAM [3], Pronghorn [4], and the MOOSE [5] heat con- duction module) for core neutronics, thermal hydraulics of the core and primary loop, and multiscale fuel performance simulations. The analysis was entirely created by coupling standalone simulations of the reactor that were previously available on the VTB. All input files and documentation de- veloped for this example are available on the VTB website: mooseframework.inl.gov/ virtual_test_bed/ . This model was featured in the National Reactor Innovation Cen- ter Tech Talk presented in December 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1991

Safety considerations in testing a fuel-rich aeropropulsion gas generator

A catalyst containing reactor is being tested using a fuel-rich mixture of Jet A fuel and hot input air. The reactor product is a gaseous fuel that can be utilized in aeropropulsion gas turbine engines. Because the catalyst material is susceptible to damage from high temperature conditions, fuel-rich operating conditions are attained by introducing the fuel first into an inert gas stream in the reactor and then displacing the inert gas with reaction air. Once a desired fuel-to-air ratio is attained, only limited time is allowed for a catalyst induced reaction to occur; otherwise the inert gas is substituted for the air and the fuel flow is terminated. Because there presently is not a gas turbine combustor in which to burn the reactor product gas, the gas is combusted at the outlet of the test facility flare stack. This technique in operations has worked successfully in over 200 tests.

Rollbuhler, R. James↗
DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents mechanical testing capability demonstration activities performed in fiscal year (FY)2025 at Oak Ridge National Laboratory(ORNL)on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~13GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide(SiC)thermometry (TM) was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target. Mechanical testing at ambient temperature demonstrated the expected irradiation-induced hardening, with yield strength(YS)and ultimate tensile strength(UTS)values increasing substantially relative to unirradiated counterparts. However, this strengthening was accompanied by a reduction in ductility, as indicated by lower uniform and total elongations(UE and TE). Both coated and uncoated claddings exhibited similar mechanical response, though Cr-coated specimens showed surface cracking perpendicular to the loading direction, attributable to the hardness mismatch between the coating and substrate. Fracture in all cases remained ductile, and no coating spallation was observed following HFIR irradiation. Complementary efforts were directed toward the fabrication of test specimens from commercially irradiated cladding (rod 47I, ~31.1 GWd/t average burnup). Axial sectioning and computer numerical control(CNC)machining successfully produced ATT geometries suitable for benchmarking against HFIR-irradiated specimens. This capability enables direct comparison of cladding behavior between test reactor and commercial reactor environments, thereby supporting the validation of HFIR as a surrogate irradiation platform for accident tolerant fuel (ATF) development. Once HFIR irradiations are completed in FY26,the relevant comparison tests will be completed. Collectively, the FY 2025 PIE campaign has provided mechanical performance data for irradiated advanced claddings The demonstrated capabilities support the framework for mechanical testing and further evaluations in subsequent years. These efforts will represent an important contribution toward the licensing and deployment of Cr-coated zirconium alloy cladding as a near-term ATF solution.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Machine-Learning-aided Approach for Predicting the Thermal Expansion Behaviors in Advanced Test Reactor Capsules (NURETH-20 full paper)

Instrumented experiments at test reactors are essential to deploying new advanced reactor systems. Designing new experiments and generating data on specific conditions require both time and cost investment. A high-fidelity model of the experiment environment can be created using finite element analysis software to support the actual experiments, but computation time is still a concern in applying outcomes to real-time usage (e.g., a digital twin). This research proposes a machine-learning-aided approach to temperature and displacement predictions, based on the thickness of the outer gas gap on the experimental capsule used for the in-pile demonstration of a novel thermal conductivity probe in the Advanced Test Reactor. The capsule consisted of U10Zr fuel, a rodlet, sodium, and inner and outer capsules. There were gas gaps between the fuel and rodlet and between the inner and outer capsule. The learning data consisted of an experimental capsule’s radial distributions of temperature and displacement, as obtained from Abaqus and the physical features. For the first step, temperature was predicted using three positional parameters. Then the displacement was predicted using six different positional parameters. Each physical feature was normalized to be both nondimensional and standardized. The temperature and displacement predictions showed good agreement in all cases involving interpolation and extrapolation. Also, data similarity enhancement increased the similarity between training and target data increasing the predictive accuracy of machine-learning models. In some cases of extrapolation, the accuracy of the machine-learning model showed limited performance, but still data similarity enhancement improved the accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-2 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE work on fuel from the second test irradiation, Advanced Gas Reactor-2 (AGR-2), began at INL in July 2014. This work scope includes Oak Ridge National Laboratory (ORNL) providing technical input, performing PIE testing and analysis, and contributing expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Towards a NEAMS-based high-fidelity model of the MARVEL reactor

This report outlines the progress of Idaho National Laboratory in developing a high-fidelity and high-resolution model of the Microreactor Applications Research Validation and Evaluation reactor. The model was developed under the Nuclear Energy Advanced Modeling and Simulation microreactor application driver at Idaho National Laboratory. The overarching objective of this activity is the development of a high-fidelity multiphysics MARVEL model using NEAMS tools, and to verify and validate NEAMS tools against MARVEL reference simulation and experimental data, respectively. This is a unique opportunity to conduct multiphysics analysis on a soon-to-be-deployed microreactor. This multiphysics model developed under the NEAMS-funded INL microreactor application driver leverages three single-physics models coupled via the MOOSE’s MultiApp and Transfer systems. The latter systems enable in-memory data transfer between MOOSE-based and MOOSE-wrapped applications. The first single-physics model, that functions as main application, leverages Griffin to model the neutron transport in the core through the discontinuous finite element (DFEM) discrete ordinates solver (SN). Several optimization flags that were developed by the Griffin developer team were beta-tested to enhance the solver’s performance. These include the combined use of using_average_xs and update_averaged_xs_on that enable to avoid expensive on-the-fly cross sections evaluations at each linear iterations in favor of evaluations of the macroscopic cross sections at each Picard iteration. The second single-physics model uses BISON to handle solid heat transfer and asymptotic hydrogen redistribution analysis in the fuel. While the model returns consistent results for the temperature and hydrogen distribution in the fuel, a mismatch was noticed in the calculated temperature in the reflector due to the value of the gap conductance used in our model. Ongoing investigations are being performed to assess the origin of this discrepancy. Finally, the System Analysis Module (SAM) was used to model the flow of the sodium-potassium eutectic in the primary loop. A first verification was also performed showing good agreement in terms of mass flow rate and inlet temperature. All mesh files were generated using the MOOSE Reactor module, removing the need for external meshing tools. Notably, this workscope represents one of the initial applications of the MOOSE Reactor module for modeling highly irregular geometries. The use of the reactor module significantly streamlined the mesh generation process. The full multiphysics mode, that combines all the single physics models, was leveraged to conduct initial steady-state multiphysics simulations to compute power, and temperature distribution in the reactor. Initial testing was performed for transient simulations as well. In this case, the new checkpoint restart capability for eigenvalue calculations was tested showing the capability for streamlined restart of transient calculations. Future work will focus on improving the fidelity of the model by performing comprehensive code-to-code comparisons. For instance, the full-core Griffin neutronics model will be benchmarked against MCNP reference results, that were provided by the MARVEL design team. Additionally, the SAM T/H model will be verified against reference RELAP-5 results for selected accident scenarios. Besides code-to-code verification exercises, the model fidelity will be improved by replacing the single-channel SAM model with a more complex SAM-Pronghorn coupled model, in which the sub-channel capability is deployed to obtain radial temperature resolution in the coolant. This model will be developed in synergy with the NEAMS thermal hydraulics team.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2019

Experimental Measurements of the High-Temperature Oxidation of Carbon Fibers

We carried out laboratory experiments to study the decomposition of carbon fibers at high temperature and studied the material used in charring carbon phenolic ablators for planetary probe heatshields. Porous plug samples were exposed to controlled rates of molecular oxygen, carbon dioxide, and carbon monoxide at furnace-heated flow-tube reactor temperatures between 700 and 1500 K and pressures between 2000 and 6200 Pa. We used calibrated mass spectroscopy to make time-resolved quantitative measurements of decomposition products of the gases downstream of the test sample. Absolute and differential pressure measurements were used to determine the high-temperature permeability of the material and to monitor changes during material decomposition, and we characterized the microstructure of the porous sample using scanning electron microscopy. We found substantial carbon fiber oxidation for O2 flows, resulting in increasing CO/CO2 production ratios with increasing temperature. The CO2 was shown to react with carbon fibers following a Boudouard reaction at temperatures above 1200 K.

Francesco Panerai↗