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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Approach to using 3D laser-induced breakdown spectroscopy (LIBS) data to explore the interaction of FLiNaK and FLiBe molten salts with nuclear-grade graphite

Nuclear graphite has historically been a key component of many nuclear reactor designs and has emerged as key to numerous advanced nuclear reactor design concepts. Molten salt reactors (MSRs) are one broad group of advanced reactor designs currently being pursued by industry for commercialization. Several MSR designs under consideration use graphitic materials that directly interface with a molten salt, whether it is a fuel salt, coolant salt, or both. Therefore, the interaction of graphite materials with molten salts must be understood. To gain this required understanding, a range of data is needed including porosity, strength, and composition as a function of different salt exposure parameters. In this study, a laser-induced breakdown spectroscopy (LIBS) measurement and data analysis methodology was developed to obtain spatially resolved elemental composition information for graphite samples exposed to a molten fluoride salt. Traditional univariate emission line analysis of atomic, ionic, and molecular optical emission signals was coupled via correlation analysis with spectral decomposition of the data using principal component analysis. Elemental depth profiling and elemental mapping were also performed to visualize salt–graphite interactions. LIBS was demonstrated to be useful for measuring key analytes such as fluorine and hydrogen, which are troublesome for other analysis techniques. Evidence for complex behavior was found, thereby demonstrating the usefulness of the developed approach for future systematic studies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Nondestructive Evaluation of Nuclear-Grade Graphite

Nondestructive Evaluation of Nuclear Grade Graphite Dennis C. Kunerth and Timothy R. McJunkin Idaho National Laboratory Idaho Falls, ID, 83415 This paper discusses the nondestructive evaluation of nuclear grade graphite performed at the Idaho National Laboratory. Graphite is a composite material highly dependent on the base material and manufacturing methods. As a result, material variations are expected within individual billets as well billet to billet and lot to lot. Several methods of evaluating the material have been explored. Particular technologies each provide a subset of information about the material. This paper focuses on techniques that are applicable to in-service inspection of nuclear energy plant components. Eddy current examination of the available surfaces provides information on potential near surface structural defects and although limited, ultrasonics can be utilized in conventional volumetric inspection. Material condition (e.g. micro-cracking and porosity induced by radiation and stress) can be derived from backscatter or acousto-ultrasound (AU) methods. Novel approaches utilizing phased array ultrasonics have been attempted to expand the abilities of AU techniques. By combining variable placement of apertures, angle and depth of focus, the techniques provide the potential to obtain parameters at various depths in the material. Initial results of the study and possible procedures for application of the techniques are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

A phase-field approach to model multi-axial and microstructure dependent fracture in nuclear grade graphite

The fracture behavior of nuclear grade graphites is strongly influenced by underlying microstructural features such as the character of filler particles, and the distribution of pores and voids. These microstructural features influence the crack nucleation and propagation behavior, resulting in quasi-brittle fracture with a tortuous crack path and significant scatter in measured bulk strength. This paper uses a phase-field method to model the microstructural and multi-axial fracture in H-451, a historic variant of nuclear graphite that provides the basis for an idealized study on a legacy grade. The representative volume elements are constructed from randomly located pores with random size obtained from experimentally determined log-normal distribution. The representative volume elements are then subjected to simulated multi-axial loading, and a reasonable agreement of the resulting fracture stress with experiments is obtained. Finally, quasi-brittle stress-strain evolution with a tortuous crack path is also observed from the simulations and is consistent with experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reconstruction of Fission Product Distribution from Tomographic Scans in TRISO Fuel Graphitic Matrix and Nuclear Grade Graphites

An image reconstruction method was developed to rectify shortcomings of earlier methods that became apparent as destructively sampled data became available. This reconstruction method was applied to the tomographic gamma scans of nuclear graphite and graphitic matrix samples from AGR-3/4. There is generally agreement between profiles measured via destructive sampling of these rings and profiles from tomographic reconstruction.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Reconstruction of Fission Product Distribution from Tomographic Scans in TRISO Fuel Graphitic Matrix and Nuclear Grade Graphites

An image reconstruction method was developed to rectify shortcomings of earlier methods that became apparent as destructively sampled data became available. This reconstruction method was applied to the tomographic gamma scans of nuclear graphite and graphitic matrix samples from AGR-3/4. There is generally agreement between profiles measured via destructive sampling of these rings and profiles from tomographic reconstruction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Comparison between the Strength Levels of Baseline Nuclear-Grade Graphite and Graphite Irradiated in AGC-2

This report details the initial comparison of mechanical strength properties between the cylindrical nuclear-grade graphite specimens irradiated in the second Advanced Graphite Creep (AGC-2) experiment with the established baseline, or unirradiated, mechanical properties compiled in the Baseline Graphite Characterization program. The overall comparative analysis will describe the development of an appropriate test protocol for irradiated specimens, the execution of the mechanical tests on the AGC-2 sample population, and will further discuss the data in terms of developing an accurate irradiated property distribution in the limited amount of irradiated data by leveraging the considerably larger property datasets being captured in the Baseline Graphite Characterization program. Integrating information on the inherent variability in nuclear-grade graphite with more complete datasets is one of the goals of the VHTR Graphite Materials program. Between “sister” specimens, or specimens with the same geometry machined from the same sub-block of graphite from which the irradiated AGC specimens were extracted, and the Baseline datasets, a comprehensive body of data will exist that can provide both a direct and indirect indication of the full irradiated property distributions that can be expected of irradiated nuclear-grade graphite while in service in a VHTR system. While the most critical data will remain the actual irradiated property measurements, expansion of this data into accurate distributions based on the inherent variability in graphite properties will be a crucial step in qualifying graphite for nuclear use as a structural material in a VHTR environment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

HIGH TEMPERATURE ANNEALING OF IRRADIATED NUCLEAR GRADE GRAPHITE

Previous work has shown that the properties of nuclear grade graphite are substantially affected by neutron irradiation. Understanding the details of how these changes occur and exactly what the damage consists of will enable prediction of the property changes as a function variables like dose, temperature, composition, micro and macrostructure. Other work has shown these atomic level changes or damage can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. Experiments were carried out at the Idaho National Laboratory to investigate how the properties of irradiated graphite recover. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young’s modulus and electrical resistivity between annealing temperatures of 500°C and 2380°C. Graphite grades NBG-18, IG-110 and PCEA are considered in both stressed and unstressed irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High temperature annealing of irradiated nuclear grade graphite

Previous work has shown that the material properties of nuclear grade graphite are substantially affected by the atomic and microstructural changes that occur during neutron-irradiation. The parameters that play a role in these changes are irradiation dose, temperature, graphite composition and the initial microstructure of the graphite. Understanding the details of how these material changes occur and exactly what the changes consists of will enable the prediction of the material property changes as a function of these variables. Other work has shown these atomic level and microstructural changes can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. Here in this work, experiments were carried out to investigate how the properties of irradiated graphite recover when heated. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron-irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young's modulus and electrical resistivity between annealing temperatures of 500 °C and 2380 °C. Graphite grades NBG-18, IG-110 and PCEA are considered that were both stressed and unstressed during irradiation.

36 MATERIALS SCIENCE↗
NASA NTRS2011

Statistical Models of Fracture Relevant to Nuclear-Grade Graphite: Review and Recommendations

The nuclear-grade (low-impurity) graphite needed for the fuel element and moderator material for next-generation (Gen IV) reactors displays large scatter in strength and a nonlinear stress-strain response from damage accumulation. This response can be characterized as quasi-brittle. In this expanded review, relevant statistical failure models for various brittle and quasi-brittle material systems are discussed with regard to strength distribution, size effect, multiaxial strength, and damage accumulation. This includes descriptions of the Weibull, Batdorf, and Burchell models as well as models that describe the strength response of composite materials, which involves distributed damage. Results from lattice simulations are included for a physics-based description of material breakdown. Consideration is given to the predicted transition between brittle and quasi-brittle damage behavior versus the density of damage (level of disorder) within the material system. The literature indicates that weakest-link-based failure modeling approaches appear to be reasonably robust in that they can be applied to materials that display distributed damage, provided that the level of disorder in the material is not too large. The Weibull distribution is argued to be the most appropriate statistical distribution to model the stochastic-strength response of graphite.

Nemeth, Noel N.↗
DOE OSTI2018

Oxidation Kinetics by Water Vapor of Nuclear Graphite Grade 2114

This report presents measurement results of oxidation rates by water vapor of nuclear graphite grade 2114 manufactured by Mersen, USA. This report completes kinetic study of chronic oxidation of grade 2114, of which partial data were compiled as a Letter Report in July 2017. The research is part of a broader program at Oak Ridge National Laboratory (ORNL) aiming at development of a comprehensive model for chronic oxidation of nuclear graphite in conditions relevant to operation of gas-cooled high-temperature reactors (HTGR). Grade 2114 is the fourth nuclear graphite investigated under this program. Experimental data measured at ORNL were analyzed in collaboration with Dr. Robert Mee (University of Tennessee, Knoxville) and graphite-specific parameters in the oxidation rate equation were determined. It was shown, once again, that the new Boltzmann-enhanced Langmuir-Hinshelwood (BLH) model recently proposed by ORNL is a better choice than the classical Langmuir-Hinshelwood (LH) model for prediction of chronic oxidation of graphite components during normal operation of HTGRs.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Dynamic Modeling and Characterization of Nuclear-grade Graphite

Idaho National Labs serves as the spearhead for many innovative energy solutions to the world's energy crisis. One such solution is the INL's Microreactor which is designed to deploy to extreme/remote environments where other sources of power are either unavailable or unreliable. In order to best design these energy solutions for their operational environments, it is crucial to understand how the design, components, and materials will respond to the environmental conditions. One key material in these innovative designs is a nuclear-grade graphite known as PCEA. This study examines the behavior of PCEA graphite under dynamic loading, similar to that which may occur in extreme environments. The objective is to characterize the dynamic behavior and produce an accurate, reliable constitutive model suitable for use in simulation tools such as INL's MOOSE. Graphite specimens were tested using a Split Hopkinson Pressure Bar (SHPB) to administer the dynamic compressive load. The SHPB was charged at various pressures to produce a range of strain rates on the material in compression. Data was acquired via strain gauges on the SHPB setup, from which stress, strain, and time data were collected. Analysis revealed the stress-strain behavior of the material as well as insights into the material behavior's relationship to strain rate. Further work must continue to characterize the various other dynamic behaviors of the material which will combine to create a substantially trustworthy constitutive model for this grade of nuclear-grade graphite. Ultimately, this will allow for realistic simulation of the material in reactor designs, allowing for prediction of design weaknesses and leading to improved designs for increased resilience, security, and reliability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

In situ transmission electron microscopy of electron-beam induced damage process in nuclear grade graphite

Atomic level processes involved in the swelling and crack-closing in nuclear grade graphite under electron irradiation have been observed in real-time using transmission electron microscopy. Noise-filtered lattice images show the formation of vacancy loops, interstitial loops and resulting dislocations with unprecedented clarity. The dislocation dipoles formed via vacancy loops were found to undergo climb resulting in extra basal planes. Concurrent EELS studies showed a reduction in the atomic density because of the breakage of hexagonal carbon rings. The formation of new basal planes via dislocation climb in addition to the bending/breaking of basal planes leads to swelling and closing of micro-cracks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Study of Tritium Transport in Nuclear-Grade Graphite and Molten Fluoride Salt Systems

High-temperature salts are proposed as coolant in recent reactor technologies including Fluoride Salt-Cooled High Temperature Reactor (FHR) with fluoride salt as primary and secondary loop coolant, Molten Salt Reactors (MSRs) with fuel salt, and high-magnetic-field fusion machines with salt coolant blanket. The baseline coolant is {sup 7}Li{sub 2}BeF{sub 4} (flibe) due to its good thermal-hydraulic and neutronic properties. Tritium produced in flibe due to neutron irradiation makes the issue of tritium transport an area of great interest. In the temperature range of 427- 704 deg. C, tritium readily diffuses through metallic reactor structural materials, which makes tritium control and capture a key technical challenge. It is anticipated that tritium can be captured by graphite, as graphite material showed a high affinity for tritium in the Molten Salt Reactor Experiment (MSRE) due to the porous and irregular crystallite microstructure. Two types of nuclear grade graphite which used as core components for FHR: matrix graphite, which encapsulates the TRISO (Tri-structural-isotropic) fuel particles in FHR, and high purity, isotropic nuclear grade graphite IG-110 which can be used as moderator and reflector in reactor core for both fission and fusion reactor. The microstructure of matrix graphite and IG-110 is different from that of matrix graphite, due to the raw material and manufacturing processes. There are three phenomena important for predicting tritium transport in graphite-salt systems: tritium species diffusion in molten flibe; tritium species transport at the salt-graphite interface; and tritium species transport in graphite. In this summary microstructure characterization of IG-110 and matrix graphite, characterization of graphite-salt interface as well as techniques for trace tritium detection in graphite are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2012

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

Nemeth, Noel N.↗
DOE OSTI2023

FINITE ELEMENT MODEL MESH REFINEMENT EFFECTS ON QUALIFICATION OF NUCLEAR GRADE GRAPHITE COMPONENTS

The American Society of Mechanical Engineers (ASME) provides the full and simplified design-by-analysis probabilistic assessments for determining acceptance of nuclear grade graphite core components. The assessments can be characterized by three parts: (1) a component stress distribution, often determined by a finite element (FE) model; (2) a Weibull probability density function (pdf) that characterizes the experimental tensile strength distribution; and (3) the post-processor, which combines the FE model and the Weibull strength distribution in accordance with the full and simplified assessments to determine component acceptance. It is known that the level of mesh refinement in FE models can affect the modeled component’s calculated stress distribution. Depending on the component geometry, the stress distribution may converge with sufficient refinement. It was previously unknown whether the acceptance decision resulting from the full and simplified assessments might change even with sufficient mesh refinement. This study explores that question using experimental strength results for a dog-bone geometry for two graphite grades, IG-110 and PCEA. The simplified assessment has two criteria that must be met, the first limits the combined membrane stress by the allowable stress and the second limits the peak equivalent stress by the allowable stress scaled by the ratio of flexural to tensile strength. In the application of the simplified assessment, convergence of the peak equivalent stress required extreme mesh refinement, however, the acceptance decision was not affected. It is hypothesized that more complex geometries with stress concentrations may present mesh refinement effects on the simplified assessment acceptance decision. Mesh refinement did affect the acceptance decision in the full assessment for the applied pressure loadings in this study. This work suggests component stress distribution convergence is not a sufficient criteria for POF convergence in the full assessment and that mesh refinement should continue until the POF has converged, especially where the resulting POF is bordering the SRC acceptable POF limit.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Understanding FLiNaK Salt Intrusion Behavior on Nuclear-Grade Graphite via Neutron Tomography

Graphite is an essential material as a neutron moderator in molten salt reactors (MSRs). To understand the impact of salt on the graphite structure to develop structural materials for MSRs, molten salt intrusion behavior on nuclear-grade graphite was studied. The graphite samples, IG-110 and PCEA, were tested for infiltration with LiF-NaF-KF (FLiNaK) at 750°C, 5 bar for 12 h, and, after the intrusion experiment, the graphite was analyzed by neutron imaging. The graphite and Li from FLiNaK showed great contrast in the neutron attenuation coefficient. Thus, the salt behavior in the graphite structure has been visualized for the first time without damaging the sample. The 3D image of the graphite was reconstructed after a neutron computed tomography scan, and the average salt coverage distribution of the XY surface in different depths was obtained from the reconstructed 3D images.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluating alternate test techniques to characterize mechanical properties in nuclear-grade graphites - 18421

The deployment of graphite as a core structural material in the very high temperature reactor (VHTR) design presents a number of challenges with regard to accurate property and performance assessments. In the United States Advanced Reactor Technologies (ART) program, candidate VHTR graphite grades are being evaluated not only with respect to irradiated property values, but also with a focus on comprehensive data sets that explicate the property variability inherent to nuclear-grade graphites. The large data sets being compiled on baseline physical and mechanical properties provide a means not only to quantify variability within and between major graphite grades, but also to evaluate test techniques utilized on irradiated specimens that are severely limited in size and geometry. Detailed appraisals of emerging mechanical test techniques that either (a) fall outside of the prescribed test size and configuration constraints in order to be assembled into irradiation capsule such as the three-point bend test on sub-sized cylinders or (b) are not yet fully evaluated and standardized by ASTM International such as the Brazilian disc splitting tensile strength test can be more accurately generated when a broad comparison is possible that leverages multiple mechanical tests on full-size specimens using approved standards. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗