INL Carbon Characterization Lab
What goes into AGC Graphite Characterization
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Results for “Nuclear Graphite”
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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Semantic Segmentation of Micrographs for Nuclear Fuel Analysis and Degradation Quantification
Not Available
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Rocket Technology Conference
The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the conference, and Mr. Roy V. Humble. Dr. John C. Eward served as general chairman for the conference.
Source record↗
Comprehensive defect evaluation of advanced nuclear fuels using high-resolution acoustic signals and optimized sensor separation
Graphite pebble composite structures based on TRistructural-ISOtropic (TRISO) particles are being developed as core nuclear fuels in advanced power reactors, promising safe operation at increased temperatures. Ensuring the structural integrity of these nuclear fuels requires comprehensive and accurate non-destructive evaluation (NDE) techniques to characterize defects and damage in the pebbles. However, traditional acoustic evaluation methods face limitations in defect characterization due to the highly attenuative, and geometrically and compositionally complex nature of these structures. This study proposes an improved acoustic NDE technique for accurate detection and classification of anticipated relevant defects and damage in graphite pebbles using high-resolution acoustic signals and optimized transmit-receive sensor networks. The proposed approach utilizes a triangular three-sensor network as the base unit, comprising three transmit-receive sensors. The sensor separation distance, as well as acoustic excitation center frequency, pulse-width, and bandwidth are optimized to enhance spatial resolution and improve signal-to-noise ratio, enabling effective characterization of the smallest size and widest range of defects in pebbles. Furthermore, the use of the triangular sensor configuration instead of a more conventional transmit-receive sensor pair expands the inspection region from a one-dimensional linear path to a two-dimensional area, increasing spatial coverage. To mitigate challenges associated with processing of complex acoustic signals arising from high-frequency, high-bandwidth excitation in these structures, a machine-learning-based signal processing algorithm is integrated with the sensor network. In the machine-learning-based algorithm, multi-domain features are extracted from the acoustic signals to capture intricate signal characteristics, significantly improving defect identification and classification compared to traditional approaches. The proposed acoustic NDE technique offers considerable promise for practical and reliable defect/damage diagnostics of advanced nuclear pebble fuels.
42 ENGINEERING↗
Heat and mass transfer rates during flow of dissociated hydrogen gas over graphite surface
To improve upon the performance of chemical rockets, the nuclear reactor has been applied to a rocket propulsion system using hydrogen gas as working fluid and a graphite-composite forming a part of the structure. Under the boundary layer approximation, theoretical predictions of skin friction coefficient, surface heat transfer rate and surface regression rate have been made for laminar/turbulent dissociated hydrogen gas flowing over a flat graphite surface. The external stream is assumed to be frozen. The analysis is restricted to Mach numbers low enough to deal with the situation of only surface-reaction between hydrogen and graphite. Empirical correlations of displacement thickness, local skin friction coefficient, local Nusselt number and local non-dimensional heat transfer rate have been obtained. The magnitude of the surface regression rate is found low enough to ensure the use of graphite as a linear or a component of the system over an extended period without loss of performance.
Nema, V. K.↗
Baseline Graphite Characterization
Baseline Graphite Characterization: Purpose of Baseline, Mechanical Baseline Tests, Physical Properties Tests, Methodology, Baseline progress, Density, Ultimate Tensile Strength, PCEA Ultimate Tensile Strength, Split Disc Tests, Compression Tests, Flexural Tests, and Coefficient of Thermal Expansion.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Tribocorrosion of stainless steel sliding against graphite in FLiNaK molten salt
The molten salt reactor (MSR) concept recently gained renewed interest in developing Generation IV nuclear reactors. One MSR design uses fluoride molten salts to cool tri-structural isotropic particle fuel encapsulated by a carbon matrix into spherical pebbles, which would inevitably contact the reactor's stainless steel container wall during salt circulation. Characterizing this interaction is crucial for reactor safety. Here, in this work, we report the tribocorrosion behavior of graphite sliding against Type 316H stainless steel lubricated by a FLiNaK (LiF:NaF:KF; 46.5:11.5:42 mol %) molten salt in an argon environment. Accelerated wear loss was observed at a higher temperature because of a lower molten salt viscosity and a higher corrosion rate. The graphite had a more rapid material loss at a higher sliding speed than stainless steel because of its higher vulnerability to vibration-induced microfracture. The salt-starved condition caused more material loss than either the no-salt or the salt-flooded condition because neither a graphite transfer film nor stable boundary lubrication could be established at salt starvation. An interesting dual-layer surface film was discovered on the stainless steel worn surface: a chromium-rich top film and a nickel-accumulated but chromium-depleted interlayer. The film composition and structure provide fundamental insights to the mechanochemical interactions between stainless steel and graphite in a molten salt environment.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Oxidation Behavior of Irradiated and Unirradiated NBG-25 Graphite
This presentation discusses The xxidation behavior of irradiated and unirradiated NBG-25 Graphite results of thermogravimetric analysis of same-source specimens from the AGC-1 experiement. Minimal annealing effects, Substantial increase in OR with Irradiation using Conventional Rate Analysis, Apparent Dose Dependency, Examination of Onset (Alternate Analysis) Shows Competing Irradiation Effects, and Similar Test Matrix (without annealing) now under way with NBG-18 Graphite is reviewed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)
To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.
Emrich, William J., Jr.↗
TRISO Spent Nuclear Fuel Recycling or Waste Reduction Using SRNL Vapor Digestion Technology – 25635
There is a renewed interest in advanced reactors, including high-temperature gas cooled reactors (HTGRs). Tri-structural isotropic (TRISO) fuel is being used in many HTGR designs, whether as SMRs or microreactors. However, TRISO-based HTGRs discharge the largest volume of used fuel per megawatt-hour of energy produced compared to other reactors. An order of magnitude reduction or more in the volume of SNF could be realized if the TRISO particles were separated from the graphite moderator and the carbon dispositioned as LLW. The Savannah River National Laboratory (SRNL) has a patented technology readiness level (TRL) 4/5 vapor digestion process for separating nuclear-grade graphite from HTGR SNF. The SRNL process is based on the reaction of NOx species with carbon to form CO2. Because NOx species are several orders of magnitude more reactive with graphite than oxygen, the process can operate at lower temperatures with uncrushed HTGR pebbles or prismatic blocks. Because the fuel elements do not need to be crushed and the graphite is digested using a vapor-based process, the potential for damaging the TRISO particles is much reduced. The DOE Office of Technology Transitions (OTT) is funding SRNL and the University of South Carolina at Columbia to close certain gaps that exist within the technology which impede its direct application to the processing of commercial TRISO-based SNF coming from HTGR advanced reactors.
Pierce, Robert [Savannah River National Laboratory↗
Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment
In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations. The ongoing NTP project, funded by NASA's Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities: (1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options; (2) Engine conceptual design; (3) Mission analysis and engine requirements definition; (4) Identification of affordable options for ground testing; and (5) Formulation of an affordable and sustainable NTP development program. Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klb (sub f) (kip, or 1000 pound-force)-class engine were developed during FYs 13-14 using heritage fuel element designs for both Rover NERVA (Nuclear Engine for Rocket Vehicle Applications) Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms. To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down-selection between GC and cermet fuel was required. An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE (Design Development Test and Evaluation) plan was produced by GRC (Glenn Research Center), DOE (Department of Energy) and industry for the AES program. Assumptions, considerations and key task activities are presented here. At the direction of NASA HQ (32515), NASA and DOE are to work together to formulate a detailed development plan and schedule allowing the affordable development of a small (7.5-16.5 klbf) GC engine for possible flight technology demonstration (FTD) mission within a 10-year timeframe.
Lunar Flyby Mission↗
Oxidation Activities
Slides for Oxidation Activities, oxidation rate, penetration/lathing, and strength after oxidation work. Also includes recent and current oxidation studies, background for rate behavior, strength behavior, density profile, and penetration measurements, oxidation resistant graphite coating development, and irradiated and unirradiated graphite oxidation rate behavior.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Analytical models of hydrogen transport in graphite
The importance of graphite-hydrogen chemical reactions to fusion, fission, and hydrogen storage applications, combined with the rapidly evolving knowledge on the underlying mechanisms, has led to the development of multiple models to describe hydrogen transport in graphite. Significant differences exist among these models, resulting from discrepancies in the modeling assumptions, intended degree of fidelity, and conditions of applicability. This paper attempts at reconciling these apparent differences by providing a comprehensive description of the constitutive equations governing hydrogen transport in graphite at high-temperature, identifying outstanding gaps in knowledge, illustrating how these different models approach them, and proposing alternative analytical formulations grounded on experimental results from hydrogen-graphite studies. Governing equations, closing relations, and simplifying assumptions are discussed for hydrogen transport at the inter-granular and intra-granular level, accompanied by compiled experimental data and illustrated energy diagrams associated to the proposed transport mechanisms. Analytical formulations are provided to reproduce competing hypotheses on the mechanisms, supporting the development of a range of computational models that can enable resolution of outstanding knowledge gaps through comparative testing against experimental data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Computational modeling of graphite degradation in molten salt reactors: Role of infiltration
Molten salt reactors (MSRs) often employ graphite as a moderator and reflector. An important challenge for deploying graphite in these reactors is that, due to limited experimental data, our understanding of graphite’s structural integrity in molten salt environments remains incomplete. Here, this study addresses heat generation from fuel-bearing salt that has infiltrated open pores in the graphite, driven primarily by pressure differentials. This is one of multiple identified physical and chemical mechanisms through which molten salt could potentially degrade graphite. Thermally driven stresses are quantified using the Molten-Salt Reactor Experiment (MSRE) graphite moderator elements as a case study. Finite element simulations predict stress distributions at varying infiltration levels, indicating that thermal stresses increase with higher infiltration. Rare-event simulations using the parallel subset simulation framework identify the combinations and corresponding ranges of input parameters that lead to stresses above a specified threshold. In particular, combinations involving high infiltration amounts, high power density, and low thermal conductivity tend to induce the highest stresses. Under the inputs and assumptions considered in this work, the magnitudes of the thermally driven stresses are quite low, with a very low likelihood of causing failure due to exceeding the graphite’s tensile strength. Additionally, rare-event simulations were performed for two more scenarios: a scaled-up moderator geometry and a localized hotspot in the original geometry. Both cases resulted in increased susceptibility to failure, though not to a detrimental extent. Furthermore, the combined effects of irradiation and infiltration-induced thermal stresses were evaluated. The results showed that thermal stresses from infiltration were negligible compared to those caused by irradiation. The findings of such a study are inherently component-specific, but the methodology presented here could be used for similar assessments of salt-infiltration effects in other graphite components.
36 - MATERIALS SCIENCE↗
Benchmarking GEANT4 and PHITS for 14.8-MeV neutron transport in polyethylene and graphite materials
The feasibility of GEANT4 and PHITS with ENDF/B-VIII.0, JEFF-3.3 and JENDL-4.0 evaluated nuclear data libraries in 14.8-MeV neutron transport simulation are verified by comparing with the MCNP calculations and the experimental data. Here the measured data are from neutron integral experiments including graphite and polyethylene samples. For 6 cm-thick polyethylene and 20 cm-thick graphite samples, the results from GEANT4 and PHITS with three evaluated nuclear data libraries agree well with the MCNP calculations. For a 2 cm-thick graphite sample, the simulated results from GEANT4 and PHITS agree with the MCNP calculations, and they all underestimate the experiment data in below 3 MeV and overestimate them in the energy range of 5–7 MeV. The difference between the simulation results and the experimental data depends on the evaluated nuclear data libraries. In overall, GEANT4 and PHITS are performing reasonable jobs for 14.8-MeV neutron transport simulations.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
The rationale/benefits of nuclear thermal rocket propulsion for NASA's lunar space transportation system
Two nuclear thermal rocket (NTR) technology options are examined - one derived from the graphite-moderated reactor concept developed by NASA and the AEC under the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) programs, and a second concept, the Particle Bed Reactor. The paper also summarizes NASA's lunar outpost scenario, compares relative performance provided by different lunar space transportation system concepts, and discusses important operational issues (e.g., reusability, engine 'end-of-life' disposal, etc.) associated with using this important propulsion technology.
Borowski, Stanley K.↗
Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs
This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.
Cermets↗
Non-destructive structural characterization of graphite components using mechanical resonance and deep learning
As compared to conventional nuclear reactors, microreactors have the potential to significantly reduce construction timelines and capital costs, decreasing the barriers for advanced nuclear reactor technologies. However, the lower power output of these microreactors (typically < 20 MWe) creates challenging economics if operation and maintenance costs cannot be sufficiently reduced. The compact size of these designs presents an opportunity for comprehensive in-situ structural health monitoring to provide real-time feedback in order to reduce operational costs associated with maintenance and downtime. Many microreactor concepts use graphite for both in-core neutron moderation and as a structural material, which has typically required some form of periodic and laborious inspection. This report provides a description and assessment of recent work with graphite to couple acoustic-based experimental measurements and characterization with machine learning models to mature structural health monitoring capabilities and generate benefits for the nuclear microreactor industry. With resilient embedded sensors in development in other programs funded by the US Department of Energy’s Office of Nuclear Energy and elsewhere, the work described herein builds upon previously funded efforts to mature non-destructive testing technology that relates measured vibrational signatures to structural changes, using a combination of new experimental measurements and machine learning processing. Building on past successful demonstrations of predictive workflows to identify structural changes in a hexagonal stainless steel test article with excellent acoustic propagation, we first performed baseline characterization on graphite samples with canonical geometries to ensure compatibility and confidence in the applied techniques for a material with distinctly different mechanical properties. In contrast to efforts in previous years, we worked exclusively with unidirectional vibration data that is more comparable to those expected from the existing embedded sensor technologies which are suitable for deployment in a reactor setting. Established acoustic and modern machine-learning-based characterization approaches were applied to the resulting datasets from these simple geometries. Both approaches were found to be highly capable of detecting even small geometric irregularities amongst nominally identical samples. As such, we then moved to testing these approaches for detection of artificial local stress perturbations introduced into a more complex geometry: a hexagonal block with drilled holes. A main outcome of this work is that a generalizable ML workflow can be used to detect and predict the characteristics of small artificial anomalies in a graphite component with a relevant geometry. While this work was performed using surficial vibration data, we expect the approach to be flexible and viable for other monitoring scenarios, such as those with different arrangements or types of sensor arrays. As compared to previously funded efforts, an existing ML workflow based on neural networks was enhanced through the addition of recently developed Fourier neural operators. As applied to previously collected and new vibration datasets, prediction accuracies of anomaly characterizations were greatly improved with minimal added computational cost. As trained on small durations of vibration data (tens of seconds) collected over a realistic number of locations, the model was able to reliably determine the presence of a subtle stress anomaly and begin to provide location estimates. Such an approach is likely to be viable for more relevant reactor damage scenarios for graphite components, such as progressive crack growth or creep.
36 MATERIALS SCIENCE↗