Nuclear Graphite Research Articles

Microstructure of Gilsocarbon graphite revealed by a correlative study of optical texture and FIB-TEM

Transmission Electron Microscopy (TEM) alongside Polarized Optical Microscopy (POM), is used to investigate the microstructure of nuclear graphite, specifically Gilsocarbon coke, fillers, binders, and flour. Detailed analyses revealed domain structures, domain boundaries, sub-domains and two types of microcracks: intra-domain and inter-domain cracks. Analysis of TEM images, from samples prepared via Focused Ion Beam (FIB), showed domain sizes in coke filler and flour regions ranged from 1 to 3 μm, whereas binder regions had larger domains of up to tens of micrometers, with correspondingly longer Mrozowski cracks. Within these domains, sub-domain boundaries further constrained Mrozowski cracks, resulting in crack lengths significantly shorter than the domain size. Domain boundaries were classified into small-angle and large-angle boundaries based on orientation differences, with large-angle boundaries arising from multiple small-angle transformations facilitated by polygonal microcracks. This microstructural data obtained from virgin Gilsocarbon graphite provides essential inputs for an experimentally informed model predicting the deformation and fracture properties of this material at the micrometer length scale, which may offer improved insights to enhance our understanding of how these properties may evolve under reactor operating conditions.

Dynamic Deformation in Nuclear Graphite and Underlying Mechanisms.

Time-dependent deformation in nuclear graphite is influenced by the creation and migration of radiation-induced defects in the reactor environment. This study investigates the role of pre-existing defects such as point defect clusters and Mrozowski cracks in nuclear graphite IG-110. Separate specimens were irradiated with a 2.8 MeV Au2+ beam with a fluence of 4.38 × 1014 cm-2 and an 8 MeV C2+ beam with a fluence of 1.24 × 1016 cm-2. Microscopic specimens were either mechanically loaded inside a transmission electron microscope (TEM) or subjected to ex situ indentation-based creep loading. In situ TEM tests showed significant plasticity in regions highly localized around the Mrozowski cracks, resembling slip or ripplocation bands. Slip bands were also seen near regions without pre-existing defects but at very high stresses. Ex situ self-ion irradiation embrittled the specimens and decreased the creep displacement and rate, while heavy ion irradiation resulted in the opposite behavior. We hypothesize that the large-sized gold ions (compared to the carbon atoms) induced interplanar swelling as well as cross-plane channels for increased defect mobility. These findings illustrate the role of pre-existing defects in the dynamic relaxation of stresses during irradiation and the need for more studies into the radiation environment's impact on the mechanical response of nuclear graphite.

Low-Temperature Annealing of Nanoscale Defects in Polycrystalline Graphite

Polycrystalline graphite contains multi-scale defects, which are difficult to anneal thermally because of the extremely high temperatures involved in the manufacturing process. In this study, we demonstrate annealing of nuclear graphite NBG-18 at temperatures below 28 °C, exploiting the electron wind force, a non-thermal stimulus. High current density pulses were passed through the specimens with a very low-duty cycle so that the electron momentum could mobilize the defects without heating the specimen. The effectiveness of this technique is presented with a significant decrease in electrical resistivity, defect counts from X-ray computed tomography, Raman spectroscopy, and nanoindentation-based mechanical characterization. Such multi-modal evidence highlights the feasibility of nanoscale defect control at temperatures about two orders of magnitude below the graphitization temperature.

An insight into annealing mechanism of graphitized structures after irradiation

Graphite components are constantly subjected to a combined actions of annealing and irradiation due to high temperatures and thermal spike caused by irradiation when reactors are operating, resulting in complex microstructures along with matching changes in the material properties and dimensions. This study investigates the evolution process of initial irradiation defects at different annealing temperatures, which is difficult to captured by experiment. The findings indicate that 923K reactor-temperature annealing can also quickly restore isolated self-interstitial atoms and small-sized point defect clusters to intralayer locations on the nanosecond scale. However, multi-interlayer penetration damages and localized point defect aggregation from high-energy irradiation can lead to the disappearance of layered structural information, which allows these damage structures to develop into interlayer dislocations during annealing process. These interlayer dislocations further exacerbate the interlayer expansion of graphite crystal and may elucidate the mechanism behind volume expansion in nuclear graphite within reactors. Fully amorphized regions can also regain a layered structure approximately when guided by residual layered structures at 2000K, while chaotic regions or disordered layered structures form in the absence of guidance. These chaotic regions significantly exacerbate the volume expansion of graphite model, which may be related to the rapid volume increase observed in nuclear graphite components at the end of reactor life. The study provides insights into the transformation of initial irradiation defects into different types of defects under different temperatures and damage states, which serve as a critical foundation for assessing the evolution of irradiation defects in nuclear graphite for reactor applications and annealing studies.

3D analysis of crack propagation in nuclear graphite using DVC coupled with finite element analysis

The initiation and propagation of cracks within nuclear graphite components compromise the lifespan of advanced gas-cooled reactors and even lead to safety incidents. A thorough understanding of fracture properties in nuclear graphite is crucial for the effective management and design of graphite structural components. In this work, an in-situ test of a double cleavage drilled compression (DCDC) specimen was conducted using X-ray tomographic imaging to fully analyze the 3D crack propagation behavior in IG-11 nuclear graphite. “Subvolume Splitting” digital volume correlation (SS-DVC) was employed to accurately quantify the displacement fields surrounding the crack. The entire 3D crack propagation and crack surface around the crack front were precisely captured through 3D imaging processing of the gray level residual (GLR) field. Additionally, crack opening displacements (CODs) and stress intensity factors (SIFs) were determined using finite element (FE) analysis with extended finite element method (XFEM). Experimental results indicate that although mode I crack remains predominant throughout the whole process, complex crack propagation exists around the crack front. The utilization of irregular crack surfaces in FE analysis helps to obtain refined displacement fields and accurate SIFs of the crack front. This work provides comprehensive insights into crack propagation and reveals the influence of complex crack morphology on SIF calculations, facilitating accurate prediction of structural integrity and rational design of graphite components.

Pore structure development of oxidized nuclear graphite

The oxidation-induced microstructure evolution of nuclear graphite (IG-110 and NBG-17) is studied. Graphite samples were oxidized in air at 500 °C. Complicated oxidation paths were observed on the ion-milled surface of oxidized graphite samples. The weight loss of graphite samples during oxidation is mainly attributed to the oxidation paths which are extended and broadened constantly. In contrast, the original gas-escape pores do not show obvious change in sizes during oxidation. The oxidation paths are characterized to be much thinner and more tortuous than the gas-escape pores, which should be taken into account in estimating the oxygen diffusivity in oxidized graphite. The edges of graphite sheets exposed in the oxidation paths could scatter the phonon and electrons strongly. Therefore, the thermal conductivity declines dramatically with oxidation time. At weight loss around 30 %, the thermal conductivity of the oxidized graphite decreased to ∼16.8 % of the corresponding original value.

Generalized denoising network LGCT-Net for various types of ESPI wrapped phase patterns

In this paper, we propose a generalized network based on our proposed Local-Global Channel Transformer (LGCT) module for denoising various types of ESPI wrapped phase patterns (including low-density, medium-density, high-density, variable-density, and discontinuous phase patterns). The Conv + BN + ReLU layer consists of convolution (Conv), batch normalization (BN), and the rectified linear unit (Relu) in series. The generalized network LGCT-Net interleaves four LGCT modules with five Conv + BN + ReLU layers in a dense connection manner. We propose the LGCT Module by stacking three Dilated-Group Convolution blocks (DGC block), a Contextual Transformer block (CoT block), and an Efficient Channel Attention block (ECA block). The LGCT module simultaneously leverages the local context extraction capability of convolutions and the powerful global information extraction capability of a transformer. Additionally, it performs feature extraction in both spatial and channel dimensions. We also create a diverse ESPI wrapped phase pattern denoising dataset with various densities, shapes, noise levels, and discontinuity. We successfully train the LGCT-Net without any preprocessing or postprocessing steps. We evaluate the performance of our method on simulated and experimental ESPI wrapped phase patterns with discontinuity and different densities. Then we compare it with previously published denoising methods PEARLS, HDCNN, ADCNN, and DBDNet quantitatively and qualitatively. The results show that our method facilitates the reduction of speckle noise and the enhancement of fine details while preserving structure and shape, outperforming the compared methods. In the end, we apply our method to dynamic measurements of nuclear graphite ESPI phase patterns at different times. And then performing phase unwrapping on the filtered phase patterns, we achieve successful results.

Surface Fluorination of Nuclear Graphite Exposed to Molten 2LiF-BeF2 (FLiBe) Salt and Its Cover Gas at 700 °C.

This study demonstrates that the reaction of Li2BeF4 (FLiBe) with graphite both in the liquid phase and the gas phase of the molten salt leads to the formation of covalent and semi-ionic carbon-fluorine bonds at the graphite surface and is accompanied by surface microstructural changes, removal of C-O groups, and deposition of metallic beryllium, based on XPS, Raman, and glow discharge mass spectroscopy characterization. At 700 °C, the observed surface density of C-F is higher after 240 h than after 12 h of exposure to molten FLiBe salt; the kinetics of covalent C-F formation is slower than that of semi-ionic C-F formation, and the relative amount of semi-ionic C-F content increases with depth. The graphite sample exposed to the cover gas exhibits less surface fluorination than the salt-exposed sample, with predominantly semi-ionic C-F. Based on these observations and the observed LiF/BeF2 ratio by surface XPS, the hypotheses that fluorination of the salt-exposed graphite occurs via a gas-phase mechanism or that it requires salt intrusion are refuted; future studies are warranted on the transport of C-F semi-ionic and covalent species in graphite at high temperatures.

Static and dynamic fracture toughness of graphite materials with varying grain sizes

Graphite materials play critical roles as moderators, reflectors, and core structural components in high-temperature gas-cooled nuclear reactors. During the reactor operation, graphite materials may experience a variety of loads, including thermal, radiation, fatigue, and dynamic loads, potentially leading to crack initiation and propagation. Therefore, it is imperative to investigate their fracture properties. Despite this, there remains a paucity of comprehensive studies on the fracture toughness of graphite materials with varying grain sizes, particularly concerning dynamic fracture toughness. This study addresses this gap by employing a digital-image-correlation-based virtual extensometer to analyze crack propagation in graphite materials of different grain sizes, enabling precise measurement of crack propagation length and fracture toughness. Findings reveal that static fracture toughness increases with larger grain sizes, while under dynamic loading, smaller grain sizes exhibit greater fracture toughness. Additionally, dynamic fracture toughness shows a near-linear increase with impact speed. Scanning electron microscopy analysis of fracture surface morphology highlights the impact of grain size and impact speed on fracture toughness. Nuclear graphite specimens with larger grain sizes have more irregular grain and pore distributions, enhancing crack deflection and propagation resistance, thereby increasing fracture toughness. The observed loading rate dependence of dynamic fracture toughness is attributed to a gradual transition from intergranular to transgranular fracture modes with increasing impact speed.

Morphological and chemical changes in nuclear graphite target under vacuum and high-temperature conditions

Nuclear-grade graphite is a high-efficiency material, widely used for vacuum applications in nuclear reactors and accelerators as targets facing particle beams. In these contexts, graphite is often exposed to extreme thermal stresses altering its physical and chemical properties. The thermal-induced release of volatile contaminants from targets and the damage of structural components are critical issues that can affect the safety and operation efficiency of beamline facilities. Here, we provide for the first time a detailed picture of the chemical and morphological changes occurring in a nuclear-grade graphite target, obtained through Electrical Discharge Machining (EDM), when exposed in vacuum to high temperatures. The radial temperature gradient induced by the impact of a pulsed energetic (MeV- GeV range) focused particle beams was reproduced by cyclically heating, in the 1300–1800 K temperature range, a disc-shaped graphite target in a vacuum setup. An accurate surface and in-depth chemical analysis of the graphite target was obtained thanks to the high sensitivity (ppm/ppb) of the Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS) technique. The chemical maps clearly show the presence of several metal oxides and impurities in the surface and subsurface regions of the untreated sample. Such contaminants were removed because of the thermal treatment in vacuum more or less efficiently, as demonstrated by Thermogravimetric analysis (TGA), X-ray Photoelectron Spectroscopy (XPS), and ToF-SIMS. However, Raman spectroscopy and SEM-EDS revealed that the high-temperature treatment induces a decrease in the crystallite size of the graphite as well as changes in the target surface porosity with the appearance of microvoids, leading the graphite target to be more prone to the breakage.

Evolution of radiation-induced damage in nuclear graphite – A comparative structural and microstructural study

Graphite, as a material for high-temperature gas-cooled reactors (HTGR), will be exposed to harsh environment. The stability of graphite structure under irradiation is of a key importance for efficiency, reliability and security of the Generation IV nuclear reactors. Three types of nuclear grade graphite were subjected to irradiation in this research – two commercially manufactured (IG-110 and NBG-17) and the laboratory's in-home material (NCBJ). The samples were exposed to 150 keV Ar+ and He+ ions bombardment at 400 °C with fluences ranging from 1E12 to 2E17 ion/cm2 in order to simulate in-reactor conditions. For analysis of the level of structure damage, type of created defects and crystallite size changes under ion irradiation ex-situ Raman spectroscopy was used. The methodology of spectra fitting was developed. Furthermore, SEM observation of irradiated materials was performed. Results showed structural degradation of materials by the means of amorphisation: slight at a low fluence level, rising rapidly at higher irradiation values. Furthermore, stronger structural disorder was found in the materials irradiated with heavier Ar+ ions than with lighter He+. Microstructural evolution of the nuclear graphites aligned with the structural deterioration in its stepwise character.

A new perspective on density and strength loss profiles at the surface of thermally oxidized nuclear graphite

Oxidation of graphite components could influence their designed life in a high-temperature nuclear reactor. The oxidized regions could potentially lower the allowed stress capacity. The American Society of Mechanical Engineers rules for the design and construction of graphite-moderated reactors recommend that subsurface regions that might become excessively damaged by oxidation during reactor operation be identified and excluded from geometry and stress calculations. Identification of oxidation-affected regions is possible, in principle, through complex modeling exercises of reactor behavior during hypothetical accident scenarios coupled with graphite oxidation models, but this procedure may not have the precision needed for informed decisions. This paper proposes an alternate method, based on interpretation of a series of well-designed oxidation experiments, which could augment the designer's tools. The procedure is illustrated by data on oxidation by air of several graphite grades (NBG-18, PCEA, IG-110, R4-650) that are corroborated with independent literature information, when available. The Wichner model for graphite oxidation used for this analysis provides conservative results that could be quickly implemented in the design process.

An In Situ transmission electron microscopy study on the synergistic effects of Au-ion irradiation and high temperature on nuclear graphite microstructure

The combined effects of high-temperature and heavy-ion irradiation on Mrozowski cracks (MC) and nuclear graphite crystallographic dimensions have been studied using in situ heating and in situ ion-irradiation in the transmission electron microscope (TEM). Electron transparent lamella of nuclear graphite, IG-110, was irradiated using a 2.8 MeV Au beam at an ion flux of 3.991 ×1010 ion cm−2 s−1 for 70 min at 800 °C. Upon high-temperature irradiation, Mrozowski crack closure was studied quantitatively. The analysis showed linear, positive expansion of nuclear graphite which is significantly different from the dimensional changes previously reported for low-dose neutron irradiation of nuclear graphite in which the material undergoes negative to positive expansion via a turnaround radiation dose. The trend of the thermal expansion coefficient (CTE) of pristine IG-110 in this study is consistent with previous reports in the 100 °C–800 °C temperature region in which the dimensional change ranges from negative to positive values.

Predicting and visualizing crack propagation in nuclear graphite

Nuclear graphite exhibits a complex fracture behavior due to the presence of defects ranging from nanoscale basal cracks to sub-millimeter scale voids as well as the heterogeneity of constituents. This study aims to develop insights on whether finite element (FE) modeling can effectively reproduce fracture behavior observed in the experiments coupled with micro X-ray computed tomography (CT). Two-phase (graphite-defect) models and grayscale-based models with elastic properties based on constituents were employed, respectively, to compute the stress and strain distribution in an NBG-18 single-edge notched beam under three-point bending. Compared with the two-phase models, the grayscale-based models revealed more variations in the stress and strain distribution because they captured the sub-voxel heterogeneity in the material properties. Particularly, the strain distribution pattern exhibited good agreement with the crack propagation paths shown in the micro-CT images. The results proved the feasibility of predicting and visualizing the crack propagation process using grayscale-based FE models. However, the prediction of crack initiation location using FE modeling remained difficult. The residual stress on the notch surface was measured to be from 1.1 to 2.5 GPa using Raman spectroscopy. It indicated that the large values of highly localized residual stress could be a factor that influenced crack initiation.

Effects of Impurities and Additives on Interactions of Nuclear-Grade Graphite with Molten LiF-NaF-KF

It has been widely acknowledged that the presence of impurities in molten fluoride salt can alter the salt/material interactions. However, the effects of various impurities such as oxides, metal fluorides, reducing impurities, etc., on the behavior of nuclear-grade graphite in molten fluoride salts have not been reported. This study focuses on understanding the effects of various oxidizing and reducing impurities on the wetting and infiltration behavior of molten FLiNaK salt for nuclear-grade IG-110 graphite. Our results suggest that different impurities can cause different effects on nuclear graphite–molten salt interactions, with some impurities leading to significant degradation of the graphite. Our results demonstrate that certain impurities, such as Cr2O3 and CrF3, lead to a limited increase in wetting and infiltration of molten salt into nuclear graphite, while impurities such as Li2O lead to significantly increased wetting and infiltration throughout the cross section of the graphite specimen. Certain impurities, such as Li, can also lead to significant degradation of the graphite in the salt, with the extent of degradation increasing with the increase in the quantity of Li added. Our results also demonstrate that firing of IG-110 graphite at 900°C under a reducing atmosphere made the graphite surface resistant to wetting by molten FLiNaK salt, as compared to the nonfired sample.

Tritium storage and release mechanism in nuclear graphite using first-principles combined thermal desorption theory

Nuclear graphite (NG) is widely applied in nuclear facilities, and its strong interaction with tritium is an essential factor for tritium transport behavior in various reactors and the treatment of radioactive NG waste. Given that NG is a polycrystalline and porous material, interactions between tritium and NG exist in various microstructures, including graphite edges, graphene basal planes, and amorphous structures. In this study, density functional theory which is integrated in Vienna ab initio simulation package (VASP) was combined with thermal desorption theory to elucidate the underlying mechanism of hydrogen isotope storage corresponding to the experimentally observed desorption peaks in NG. The four desorption temperature peaks, ranging from low to high, were mainly attributed to hydrogen isotope desorption from the graphite basal plane surface, amorphous carbon surface, unsaturated armchair and zigzag edges, respectively. The interaction between hydrogen isotopes and carbon nanostructures provides deeper insights into the understanding of tritium enrichment on the NG surface, which is primarily attributed to the adsorption of tritium on graphite basal plane-like surfaces at relatively low temperatures. This study provides a practical method to bridge the gap between the atomic binding structure and desorption temperature peaks from a kinetic perspective, providing insights into temperature-dependent hydrogen desorption in typical carbon structures and deepening our understanding of tritium transport behaviors in advanced nuclear energy systems.

On the reduced damage tolerance of fine-grained nuclear graphite at elevated temperatures using in situ 4D tomographic imaging

Fine-grained nuclear graphite is one of the key structural materials for high temperature gas-cooled reactors as well as several Generation IV nuclear fission reactor designs. However, its deformation and fracture behaviours at elevated temperatures are not well understood. In light of this, the current study focused on investigating the flexural strength and fracture toughness of two fine-grained graphite (SNG623 and T220) using real-time X-ray computed micro-tomography imaging at room temperature, 750 °C and 1100 °C. Specifically, nonlinear-elastic fracture mechanics-based JR(Δa) R-curves at these temperatures were presented with evolution of damage and failure micro-mechanisms, local strain distributions and J-integral fracture analysis, purveying notable findings. Compared to the coarser-grained Gilsocarbon nuclear graphites used in the current UK Advanced Gas-cooled Reactors (AGRs), these modern fine-grained graphites display deficient fracture resistance in the form of far less stable crack growth prior to catastrophic fracture and reduced failure strain at 1100 °C. Moreover, their elevation in strength and toughness at high temperatures is remarkably lower than that of Gilsocarbon graphite. Based on in situ high-temperature Raman spectroscopy mapping, we believe that one of the major causes of this behaviour can be attributed to the smaller magnitude of ‘frozen-in’ residual stress relaxed at elevated temperatures compared with Gilsocarbon graphite.

Preliminary analysis of the irradiation deformation of a typical molten salt reactor graphite component

Nuclear graphite commonly serves as the moderator and constructs the coolant flowing channel in liquid fuel molten salt reactors. Under irradiation and high temperature, graphite components will experience obvious deformation due to irradiation creep, thermal expansion, dimensional change, elastic deformation, and thermo-mechanical properties change. The lifespan of graphite components is a limiting parameter of liquid fuel molten salt reactors, determining by the deformation of the graphite components. An improved approach based on the finite element method for graphite deformation analysis was proposed in this paper. Then this method was utilized to analyze a 10-year deformation history of a typical graphite component. The findings show that the stress-strain field of the graphite component significantly varies with the increasing irradiated neutron fluence. The dimensional change plays a primary role in influencing the deformation of the graphite components, determining the overall deformation trend. The irradiation creep plays a secondary role in influencing the deformation, alleviating the deformation of the graphite components to prolong the graphite lifespan. The improved method, which considers various factors like irradiation creep, thermal expansion, dimensional change, and elastic deformation, tends to provide more accurate results compared to the traditional method that focuses solely on the dimensional changes. Over 10 years, the contraction deformation of graphite components results in a reduction in graphite volume of up to 3.87%. The space for graphite contraction is filled with molten salt, resulting in a maximum increase of 22.49% in fuel volume. The alteration in volume of both graphite and fuel leads to a reduction in reactivity of up to 840 pcm.

MCNN-DIC: a mechanical constraints-based digital image correlation by a neural network approach

Digital image correlation (DIC) is a widely used photomechanical method for measuring surface deformation of materials. Practical engineering applications of DIC often encounter challenges such as discontinuous deformation fields, noise interference, and difficulties in measuring boundary deformations. To address these challenges, a new, to the best of our knowledge, DIC method called MCNN-DIC is proposed in this study by incorporating mechanical constraints using neural network technology. The proposed method applied compatibility equation constraints to the measured deformation field through a semi-supervised learning approach, thus making it more physical. The effectiveness of the proposed MCNN-DIC method was demonstrated through simulated experiments and real deformation fields of nuclear graphite material. The results show that the MCNN-DIC method achieves higher accuracy in measuring non-uniform deformation fields than a traditional mechanical constraints-based DIC and can rapidly measure deformation fields without requiring extensive pre-training of the neural network.

Characterization and ion-irradiation studies of typical microstructures in nuclear graphite

Nuclear graphite is an important structural material for advanced gas cooled reactors and molten salt reactors. Graphite microstructures are always accompanied with typical pores or cracks which plays an important role in nuclear graphite behaviors. A new surface processing method involving ion milling and fast oxidation is used here to characterize typical microstructures from the large-size calcination cracks in fillers to very small quinoline insoluble particles in NBG-17 and NBG-18 graphites. By using this surface processing method and ion irradiation technique, an attempt is made to explore the microstructural basis of graphite irradiation behaviors. Heating-induced relaxation of graphite microstructures is observed, indicating the presence of residual stresses in nuclear graphite. Creased graphite sheets are identified on ion-milled surface and are believed to accommodate a-axis shrinkage of graphite crystallites at the beginning of irradiation, which is consistent with the initial plateau of slow dimensional shrinking. Compared to NBG-17 graphite, NBG-18 is found to have much more calcination cracks which contract during irradiation, resulting in a deeper volume shrinkage of NBG-18. The quinoline insoluble particles containing chaotic structures are shown to severely contract and separate from surrounding structures, leading to voids in between. Both NBG-18 and NBG-17 have abundant quinoline insoluble particles, which might make a contribution to their smaller volume shrinkage than other medium-grain graphite.