Nuclear Graphite Research Articles

Microstructural changes in nuclear graphite induced by thermal annealing

Irradiation-induced property change in nuclear graphite is particularly important when considering the in-service lifetimes of graphite components. In other works, it has been shown that annealing of irradiated graphite above the irradiation temperature may heal atomic-level defects and thus reverse some amount of physical property change. In this work, virgin nuclear graphite IG-110 was annealed at 2500 °C to observe any microstructural changes due solely to high-temperature thermal annealing. The results shown in this study suggest that fullerene-like defects, which are not observed in as-prepared nuclear graphites, will arise due to thermal annealing near graphitization temperature. Consequently, such defects may directly contribute to non -recoverable physical property change which has previously been observed in irradiated nuclear graphites. • Non-recoverable physical property changes to nuclear graphite. • High-temperature thermal annealing of nuclear graphite. • Microstructural changes induced solely by high-temperature thermal annealing. • Fullerene-like defects characterized via transmission electron microscopy.

Surface modification of ET-10 graphite by helium irradiation and preoxidation: Morphology, microstructure and chemical states

The synergistic effect of neutron irradiation and oxidising impurities inside reactors has a considerable influence on the modification and performance of nuclear graphite. We investigated the modification of the surface morphology, structure and functional groups of preoxidised ET-10 graphite by He+ beam irradiation. A distinct lamellar structure after high-temperature preoxidation indicates the depletion of the binders. Raman parameters |q−1| and ID1/(ID1+ID2) were calculated to reflect the variation in defect and crystallite size, and the crystallite size (La) of ET-10 graphite was reduced from 12.66 nm to 1.4 nm after irradiation. The high-temperature oxidation, on the one hand, increased La by the annealing recovery of defects; on the other hand, led to faster surface oxidation, destroying crystals and reducing La. Furthermore, changes in the electronic structure and elemental composition were qualitatively and quantitatively measured by X-ray photoelectron spectroscopy. The irradiation reduction occurred mainly at 5 × 1014 He+/cm2, where the ratio of O-containing groups was restored from the unirradiated 47% to 27%, which was close to the level of the graphite matrix. Lastly, while previous researches suggest that nuclear stopping power has a similar effect. The irradiation reduction is found to be related the energy transferred by the electron interaction combined with SRIM simulations.

Perspective on “code qualifying” new graphite grades for use in advanced nuclear reactors*

The American Society of Mechanical Engineers (ASME) publishes the Boiler and Pressure Vessel (BPV) Code, which include guidance for the safe development, construction, and operation of boilers and pressure vessels. ASME BPV Code Section III “Rules for Construction of Nuclear Facility Components” Division 5 focuses on “High Temperature Reactors”. Subsection HH, subpart A lists the different materials properties that are to be measured and how those properties change due to different environmental conditions (oxidation and irradiation damage) for a graphite to be accepted for use in a high temperature reactor core (i.e., “Code Qualified”). Currently there are no nuclear graphite grades that are “Code Qualified” (i.e., a reactor designer can select a graphite grade and build their reactor without any additional testing), which is due in part to development of new graphite grades in the last 20 years and the lack of comprehensive programs needed to produce the data for the code cases. This perspective is going to discuss the requirements, as called out in the ASME BPV Code, that are necessary to “code qualify” a nuclear graphite grade but will primarily focus on the practical and technical challenges associated with irradiation-induced property changes and how to address these to assist with getting graphite ready for use in advanced nuclear reactors. These same technical challenges can be expected to arise for other materials being developed for advanced reactor concepts.

Measurement of Helium and Xenon Knudsen permeability in graphite IG-11

Understanding the flow and diffusion (transport) of gases through nuclear graphite is of interest in both the prismatic and pebble bed high-temperature gas reactors (HTGRs), for normal operation and under accident conditions, for example, as related to graphite oxidation. Both the laminar and turbulent gaseous transport are of interest. It has also been noted that nuclear graphite can have pore sizes that can possibly lead to non-continuum transport, as the pore size can be comparable to the gaseous mean free path. We report measurements and analyzes of experiments performed with IG-11 graphite in the pressure range of 10 Pa to 14,700 Pa, 293 K, using Helium, and research grade xenon (99.999%), respectively. These values are in fair agreement with those reported for different graphites and gases by other investigators, but are also higher by six orders of magnitude than those reported for Kr permeability in a developmental very low permeability graphite-HXT-90. Our measured ratio of the Helium and Xenon permeabilities is about 4, different from mXemHe≈5.72 that corresponds to diffuse reflection at the graphite surface for both gases. Simple analysis indicates that the two gases have slightly different accommodation coefficients (about 25%) with the graphite surface. However, there could also be other reasons, such as experimental uncertainties or the need for a more detailed analysis.

Self-lubrication of nuclear graphite in argon at high temperature

Tribology studies in argon of nuclear graphite previously degassed at 600 °C show lower friction and wear at 600 °C than at room-temperature: the COF decreases from 0.55(14) to 0.33(5) and the specific wear rate decreases from 0.4(3) to 0.06(3) μg/Nm. Microstructural characterization of the wear spots via digital, polarized, and electron microscopy and Raman spectroscopy suggests formation of a Tribo-film formed by fracture perpendicular to the basal planes that exhibits crystallite alignment. The improved self-lubrication at high temperature results from the presence of a thicker and more continuous Tribo-film, attributed to the increase with temperature in the tensile strength and in the anisotropy of the chemical reactivity of graphite crystallites.

Estimation of a Fission Product Concentration Distribution in a Spherical Body from Surface Release Measurements

The diffusion of fission products (FPs) in reactor materials affects the nuclear source term. The diffusion coefficient itself is measured through various techniques. In the release method, it is of interest to know the initial distribution of the FPs in nuclear graphite or other materials from an exterior measurement like mass surface flux or cumulative mass release. In this paper, a Fredholm integral of the first kind is considered, relating the initial distribution to the cumulative release fraction of a diffusant from a spherically symmetric body. The Tikhonov regularization, conjugate gradient least-squares (CGLS) method, and algebraic reconstruction technique (ART) with nonnegativity and conserved mass constraints were compared to fractional release data from a simulated linear profile using data for Cs diffusion in a 0.32-cm sphere NBG-18 at 1090 K. The Tikhonov regularization was shown to provide a better estimation of the initial linear distribution than the CGLS and ART methods. The performance of the Tikhonov regularization was further examined with simulated constant, quartic, and exponential initial distributions. The Tikhonov regularization provided a reasonable recovery of the exponential profile, while the estimation of the linear, constant, and quartic profiles suffers from several issues.

Effects of irradiation on nano-pore phenol-formaldehyde resin infiltrated IG-110 graphite

• The PF filler in pores of IG-110 shrinks after irradiation. • The irradiated densified isostatic graphite has smaller change ratio of L a . • The irradiation behavior depends on the initial graphitization degree of graphite. Iso-molded nuclear graphite (IG-110) was infiltrated with phenol–formaldehyde (PF) resin to get the densified isostatic graphite with different bulk density (IG-110-PF, 1.86 g/cm 3 and IG-110-PF-D, 1.91 g/cm 3 ). The IG-110-PF and IG-110-PF-D samples were irradiated with 7 MeV Xe 26+ to study their surface topography and microstructure evolution. The impregnation process of PF resin affects the initial graphitization degree of graphite, which lead to the difference in irradiation behavior. Due to the irradiation-induced graphitization, the PF filler (carbonized products of PF resin infiltrated into pores) in pores of IG-110 shrinks after irradiation, which confirmed by the smaller change ratio of crystallite size along the a-axis direction (L a ) in IG-110-PF and IG-110-PF-D after irradiation than IG-110. L a of IG-110-PF-D decreased with the dose increase, resulting from the introduction of irradiation-induced in-plane defects, which is similar to IG-110.The increase in L a of PF filler in IG-110-PF and decrease of increase rate in dislocation density after irradiation with surface irradiation dose of 0.02–0.11dpa, provide supporting evidence for the change of PF filler. The difference in irradiation behavior confirms that disordered structure of PF filler and initial crystallinity play an important role in microstructure evolution.

Simulation calculation of the influence of interstitial atoms on the desorption behavior of tritium in nuclear graphite

In order to explore the effect of interstitial atoms on tritium desorption behavior of nuclear graphite, the adsorption and desorption behaviors of hydrogen atoms on interstitial atoms were simulated by first-principles. The results indicated that the interstitial atoms can gradually adsorb 3 hydrogen atoms until the sp3 hybrid structure is formed. The CH3 structure with dangling bonds tends to desorb as ionic CH3, which will combine with free H or other CH3 to form CH4 or C2H6. The CH (or CH2) structure tends to polymerize with other CH (or CH2) structures to form C2H2 (or C2H4), which reaction energy barrier is slightly lower than that of the CH3 structure; Although the migration energy of the interstitial atoms with H adsorbed on the graphite flakes is not much different from that of the interstitial atoms without H, the energy barrier of their recombination with single vacancies is significantly greater than that of the interstitial atoms without H. Therefore, it can predict that the large number of interstitial atoms in decommissioned nuclear graphite will become important sites for chemisorption of tritium. In addition, the reaction energy barrier of the interstitial atoms possessed large adsorption energy with tritium gas desorption is significantly higher than that with carbon containing tritium products (including C2T2, C2T4, CT4 or C2T6). This work will provide technical guidance and support for the heat treatment of the adsorbed tritium in decommissioned nuclear graphite.

Synthesis of needle coke by co-carbonization of coal liquefaction pitch and refined soft pitch

ABSTRACT Needle coke was a type of engineering material and function artificial carbon material that was extensively utilized as a raw material for the synthesis of ultra-high power electrodes and nuclear graphite. The characteristics of the pitch have a considerable influence on the generation of needle coke of high quality. To study the potential of coal liquefaction pitch (CLP) for the synthesis of high-quality needle coke, refined soft pitch was utilized as an additive to generate needle coke via the co-carbonization method. The optical microscope (OM), scanning electron microscopy (SEM), X-ray diffraction, Raman spectrum, and curve-fitted method were used to analyze the microstructure, true density, carbon crystalline distribution, and micro-strength of the generated needle coke, respectively. Factually, the blending of CLP and refined soft pitch was an essential method to adjust the thermal stability of the pitch mixtures. In addition, co-carbonization of CLP and refined soft pitch was an effective technique for preparing needle coke of high quality with a true density of 2.206 g/cm−3 and a fibrous structure and leaflet structure content of 94.44%.

Multiscale characterization and comparison of historical and modern nuclear graphite grades

Beginning with Chicago Pile I, graphite has been used as a moderator material in nuclear power stations and is considered a potential material for use in future Generation IV advanced reactors. The microstructure of graphite is responsible for much of its mechanical and thermo-physical properties, and how it responds to irradiation. To understand graphite microstructure, it is necessary to understand its porosity at the macro- and micro-scales; and to understand its porosity, it is necessary to characterize the morphological connectivity of the void content and the two main phases of graphite: filler and binder. Here, using several microscopy and analytical techniques, a detailed examination of the heterogeneity, microstructure and pore structure of different graphite grades and their binder and filler phases is presented. Significant differences were found between coarser and finer nuclear grades. Coarse grades have a more diverse range of filler particles, pores and thermal cracks. Finer grades have a more well-defined pore size distribution, fewer variations of filler particles sizes and do not contain as many large thermal cracks. Fine grades tend to have a well-connected network of pores whereas coarser grades contain a larger content of closed porosity. The framework developed within this work can be applied and used to assess the various graphite grades that would down-select materials for specific use in graphite moderated reactor designs.

Diffusion of cesium in oxidized and unoxidized IG-110 nuclear graphite

Time-release diffusion measurements of cesium have been conducted over the temperature range 1073 K – 1973 K on oxidized and unoxidized IG-110 graphite. Four cesium concentrations were tested to investigate the concentration dependence of the diffusion coefficient. Two levels of oxidation were tested and compared to unoxidized concentration-matched sets to explore the effects of graphite oxidation. The results demonstrate that cesium diffusion coefficient in unoxidized IG-110 graphite is independent of concentration within the range 34 – 163 µg Cs/ggraphite. Above this, the effective cesium diffusion coefficient changes with concentration. The diffusion coefficient was increased by a factor of 2–12 in the oxidized set with 7.8% mass loss. These results can be used to aid predictive modeling of cesium diffusion in HTGR cores.

The Wigner energy and defects evolution of graphite in neutron-irradiation and annealing

The annealing process of nuclear graphite accompanying the release of Wigner energy was studied in this research. IG-110U and ETP-10 graphite were irradiated at 200 °C by neutrons with low-dose (1.38 × 1019 n/cm2) and high-dose (1.92 × 1020 n/cm2), respectively. Their Wigner energy release at 100–550 °C was measured based on a high-precision differential scanning calorimeter. Characteristics of Wigner energy release spectrum were investigated. The activation energy of the energy release peak (170–230 °C) for low-dose graphite was 1.48 eV, which was related to the decomposition of interstitial clusters and the behaviour of subsequent interstitial pairs. Furthermore, a comprehensive analysis of Wigner energy release combined with thermal conductivity and multi-scale structural parameters during the neutron irradiation and annealing process is given.

Determination of Impurities in Graphite Using Proton Induced Gamma Ray Emission, Total Reflection X-ray Fluorescence and Instrumental Neutron Activation Analysis

Determination of impurities in graphite is very important for its quality control, as their presence even at trace level can affect the performance of graphite in various applications. Graphite with equivalent boron content (EBC) less than 5 mg kg-1 is considered as nuclear grade. Elements with high neutron absorption cross section (boron and rare earths) contribute significantly to EBC. Non-destructive method is preferred as there is no sample processing and probability of loss of volatile elements while digestion. Proton Induced Gamma Ray Emission (PIGE), Instrumental Neutron Activation Analysis (INAA) were utilized for the non- destructive determination of impurities in both nuclear and commercial grade graphite. Low Z elements like Li, B, F, Na, Al and Si were detected in graphite by PIGE whereas Na, K, Sc, Cr, Mn, Fe, Co, Zn, Rb, Zr, Sb, Cs, La, Ce, Nd, Sm, Eu, Tb, Yb, Hf, Ta, Th were determined using INAA. Few elements like Ca, Ti, V, Ni, Sr and Pb remained undetected by both the non-destructive techniques. These elements were determined by Total Reflection X-ray Fluorescence (TXRF) after digestion of the graphite samples by dry ashing. Combinations of these techniques were utilized to get maximum information regarding the impurities present in graphite.

Determination of nuclear graphite impurities by prompt gamma activation analysis to support decommissioning operations

Elemental composition of non-irradiated nuclear grade graphite was determined by Prompt Gamma Activation Analysis (PGAA) to quantitively assess some neutron activation precursors. The knowledge of these data is paramount for an accurate radiological characterization of the material before decommissioning of graphite-moderated nuclear reactors. Bulk results of most of the nuclides were consistent with benchmark Inductively Coupled Plasma–Mass Spectrometry (ICP-MS) and literature data. In the case of 14N and 35Cl, depth distribution profiles were observed and quantified for the first time. These outcomes could shine a light on 14C and 36Cl depth distribution profiles and fractional release during leaching experiments.

Using porous random fields to predict the elastic modulus of unoxidized and oxidized superfine graphite

Nuclear graphite is a candidate material for Generation IV nuclear power plants. Porous materials such as graphite can contain complex networks of pores that influence the material's mechanical and irradiation response. A methodology known as the random finite element method (RFEM) was adapted to create synthetic microstructures and predict the influence of porosity on the elastic properties of graphite during oxidation. RFEM combines random field theory and the finite element method in a Monte Carlo framework to estimate the mechanical response of a given grade of graphite. In this research, the random fields were verified through experimental characterization to predict the elastic response of three nuclear graphite grades, ETU-10, IG-110, and 2114. Finite element models (FEM) were generated using segmentations of x-ray computed tomography (XCT) data known as image-based models (IBMs) to validate and compare with the RFEM results and better understand the effects of uniform oxidation in these graphite grades. The RFEM predictions appear to correlate well with the experimental values of the measured Young’s modulus of the three graphite grades and display the same trends as IBMs.

Explosibility of nuclear graphite measured in a 1 m3 chamber

Introduction. Nuclear graphite poses a threat due to the formation of the graphite dust – air mixture (GDAM) during the dismantling of decommissioned nuclear reactors. However, there is no clear answer to the question on the GDAM explosibility. A review of international studies suggests that GDAM is either inexplosive or its explosibility is weak (Phylaktou H.N. et al., 2015). In this paper, the authors advance arguments for the explosion safety of GDAM.Selected research result. The authors considered a well-known result of a study on the combustion of GDAM with an average particle size of 5 μm, the concentration of about 450 g/m3 in a 1.138 m3 chamber, and an ignitionsource made by Fr. Sobbe GmbH («Sobbe 10 kJ»). The maximum overpressure ΔPmax was 0.47 bar in the chamber, and it fitted the case of an explosive air suspension, according to EN 14034-3 (1 bar = 100 kPa).Interpretation of the research result. Pressure oscillograms were compared for the following two cases: the case of the maximum manifestation of the GDAM explosion hazard (ΔPmax = 0.47 bar; dP/dt|max = 3.8 bar/s) and the case of combustion of an ignition source in the absence of air suspension (ΔPmax = 0.027 bar; dP/dt|max = 2.7 bar/s). The comparison shows that the first 20 ms of a pressure change inside the chamber is mainly due to the combustion of the ignition source: the characteristic values ΔP = 0.03 bar and (dP/dt) ≈ 3.8 bar/s are close to the «Sobbe 10kJ» combustion index in the absence of GDAM. A further increase in ΔP is accompanied by the constant or sharply decreasing value of (dP/dt), which means a monotonous decrease in the flame velocity and proves the incombustibility of GDAM.Conclusions. Due to the smallness of ΔPmax, GDAM can be considered nonexplosive under normal atmospheric conditions. Dependency diagrams, relating the pressure of combustion products and its growth to time offer important information about the combustion of the air suspension in explosion chambers under the condition of a low dust explosion hazard.

Preliminary Study on Oxidation of Nuclear Grade Graphite

Preliminary Study on Oxidation of Nuclear Grade Graphite

Calibration of a Finite Element Forward Model in Eddy Current Inspection

We report on the use of a novel constrained optimisation algorithm for calibrating the finite element model in an eddy current inspection application. An accurate finite element forward model is often important in such eddy current applications for training neural networks or as part of an iterative solver. However, the subject of calibration of the model has not received much attention in the literature to date. We consider a multi-frequency, eddy current depth profiling application, which is important for non-destructive testing in the nuclear industry. In the optimisation algorithm, we use a Levenberg-Marquardt algorithm and a bisection search to ensure constraints are satisfied, coupled with a truncated Gradient and Hessian method. We calibrate two types of finite element models, one using a filament representation for the coils and the other using a full 3D approach. The results show the feasibility of using the constrained non-linear optimisation algorithm for tuning the finite element model parameters. The mean signal-noise ratio after tuning on a truncated spectrum was 29.47 dB for the 3D model and 28.96 dB for the filament; in contrast the mean SNR using the measured coil parameters was 1.48 dB and 4.98 dB for the two uncalibrated models, respectively. The results show that a filament model is competitive with a 3D model of the coils (within the nuclear graphite application); therefore, a computationally faster filament model can be used with minimal effect on accuracy.

The structure evolution in neutron-Irradiated nuclear graphite and post-annealing

The neutron irradiated microstructure and performance of nuclear graphite as a neutron moderator and structural materials are directly related to the safety and longevity of reactors. Nuclear-grade isotropic graphite IG-110U and ETP-10 were irradiated at 200 °C with low-dose (1.38 × 1023 n/m2, 0.02dpa) and high-dose (1.92 × 1024 n/m2, 0.25dpa) neutron and high-temperature annealed at high temperatures up to 1400 °C, with the whole process analyzed using X-ray diffractometer and CARBONX program. After neutron irradiation of both doses, the c-axis lattice constant <c> became significantly larger, while the a-axis lattice constant <a> only changed a little. Although the crystallite size along c-axis Lc and the number of internal atom layers M both increased significantly, the crystallite size along a-axis La changed slightly for the low-dose samples. However, all three parameters decreased significantly after high-dose irradiation. During the subsequent high-temperature annealing, there were "recovery" points where both the microscopic parameters and macroscopic dimensions were restored to the initial unirradiated value and kept mostly constant. The movement and recombination of interstitial atoms, vacancies and their clusters played a major role in this annealing process. However, the crystallite sizes of two graphites irradiated with high-dose neutrons did not reached their initial values even with annealing at 1400 °C, mainly due to the unrecoverable defects, such as interstitial loops, pores, and microcracks.

SiC/MoSi2-SiC-Si Oxidation Protective Coatings for HTR Graphite Spheres with Residual Si Optimized.

SiC/MoSi2-SiC-Si coatings for nuclear graphite spheres with different Si-Mo ratios were prepared through two-step pack cementation. XRD, SEM and EDS techniques were used to analyze the composition and microstructure of the coatings. The oxidation resistance performance of the composites at 1773 K, in static air, was investigated. The results showed that the SiC-MoSi2-Si coating could be divided into a denser inner layer and a loose outer layer, as free Si would infiltrate into the inner micropores of the coating under capillary force. When the Si/Mo ratio of the second pack cementation was 7:1, the thickness of the denser inner layer basically reached the maximum and exhibited excellent oxidation resistance ability, with a weight gain of 0.19% after 200 h oxidation. The performance improvement was analyzed as a result of the addition of SiC and C powder in the pack cementation process, effectively increasing the phase interfaces to relax the thermal stress in the coating. With different Si-Mo ratios, the content of residual Si and the formation rate of SiO2 glass layer on the coating surface were also different, thus affecting the anti-oxidation performance. The main reactions occurring at different stages of the oxidation curve were also discussed.