Nuclear Graphite Research Articles

Characterization of heterogeneity in NBG-18 nuclear graphite microstructure by correlative analysis of optical texture and focused ion beam transmission electron microscopy observations

NBG-18 is a nuclear graphite that has been considered for use in high-temperature gas-cooled reactors planned by South Africa and China. Although many characterization experiments have been performed on NBG-18 nuclear graphite, some structural features leading to microscopic heterogeneity have not yet been noticed. In this study, a detailed correlative analysis of optical texture and results from focused ion beam transmission electron microscopy (FIB-TEM), Raman mapping, and atomic force microscopy (AFM) was performed. First, various optical texture elements were identified, including aspects of the fine optical structure (granular domains and isotropic matrix) of the NBG-18 filler. A similar dual structure was also observed in the binder (anisotropic components in an isotropic matrix), and occasionally flow-type binder structures were seen. Then, guided by the large-area optical microscopy maps, various elements were explored in detail using advanced microscopy (FIB-TEM, AFM) and spectroscopy (Raman) techniques. Thus, unexpected structural properties were revealed (local isotropy and anisotropy, orientations, lattice sizes, rosettes, local atomic order and disorder), the origins (e.g., changes during carbonization process) and future evolution (e.g., changes induced by electron irradiation) of which were discussed. Thus, this study explores a new level of microstructure complexity and heterogeneity of NBG-18 nuclear graphite.

Diffusion of fission products in nuclear graphite: A review

• A literature review of the metallic FPs diffusion in nuclear graphite is presented. • The diffusion theories applied in the diffusivity determination are summarized. • Measurements of the diffusion coefficient of metallic FPs are discussed. • The variables affecting the diffusion coefficients of Cs, Sr, and Ag are discussed. Nuclear graphite in high-temperature gas-cooled reactors (HTGRs) acts as a moderator, a structural material, as well as an accommodation for fission products (FPs). Given the release of FPs under normal and accident conditions, the diffusion of FPs in nuclear graphite should be elucidated for the estimation of source term and reactor safety. However, the existing experimental data of diffusion are enriched in data of the nuclear graphite types (i.e., H451 and A3), that may not apply to next-generation nuclear graphite. The proposed diffusion theory and experimental systems are conducive to developinging a novel experimental method for the more accurate collection of the diffusion data and an insight into FPs diffusion in graphite. For this reason, this paper reviews the information regarding the diffusion of metallic FPs (e.g., Cs, Sr and Ag) in nuclear graphite. The diffusion theory adopted to calculate the diffusion coefficient is summarized. Emphasis is placed on the measurements for determination the diffusion coefficient and the discussion of detection limits, accuracy and experimental methods. Given the diffusion coefficient data, the variables of diffusion are addressed. The mentioned results lay a basis for gaining insights into the diffusion theory and can be exploited by employing with a valid measurement method to determine the diffusion coefficients of FPs in various nuclear graphite materials under a particular range of conditions.

The pores evolution in 3D view of nuclear graphite under ion beam irradiation

• FIB-SEM tomography show off the pores evolution of ion irradiated nuclear graphite in 3D view. • 3D reconstruction for the quantitative statistic of porosity. • Typical microcracks are extracted and displayed in 3D perspectives. • Reveal the closure of microcracks caused by ion irradiation on a 3D scale. As an outstanding material for moderator and reflector, graphite is broadly applied in nuclear reactors. Irradiation will change the pore structure and content, which is intimately related to the properties of graphite. However, the irradiation effect of nuclear graphite on pores requires a more profound understanding, especially the evolution on a three-dimensional(3D) view. In this work, FIB-SEM tomography intuitively reveal the pore structure and porosity evolution of ion irradiated graphite grades IG110 in a 3D view, which was irradiated with 30 MeV Ni 5+ ion at 400 °C. Besides, the quantitative statistic of porosity indicates a negative correlation between porosity with the ion irradiation damage and confirms the consistency of ion deposition sites and irradiation damage peak. Moreover, individual microcracks of the different ion irradiated depths and virgin graphite are extracted and displayed in various 3D perspectives, which exhibit a much more complex morphology than lenticular structure and demonstrates the closed trend of the microcracks under irradiation. Through the display of 3D pore structure and quantitative statistics of porosity, aims to provide a further understanding of ion irradiation behavior of nuclear graphite.

The distribution and selective decontamination of carbon-14 from nuclear graphite

• A detailed 14 C release profile from irradiated graphite is constructed. • A reducing gradient of 14 C from gas-facing surfaces is observed and discussed. • The 14 C-enriched surface layer persists deeper than any carbonaceous deposit. • The 14 C release profile is repeatable for a given graphite source and location. This work investigates 14 C release behaviour from UK Magnox reactor sourced irradiated graphite under relatively low temperature oxidation conditions: a 1% oxygen in argon atmosphere at temperatures ranging from 600°C to 700°C, and durations from 4 to 120 hours. A method is used for construction of a detailed 14 C release profile with mass loss. The 14 C release profile is predictable between samples, and an empirical release profile is derived to predict releases to higher mass losses than in this work. An accelerated 14 C release rate at small mass losses is observed, indicative of a 14 C-enriched surface region with a depleting concentration gradient into the material, and the source terms and complexities associated with predicting where the 14 C arose from are discussed. Selective decontamination of a fraction of the 14 C is possible, with the limitation of a reducing efficiency with mass loss and time. This work demonstrates a method for determining the distribution of 14 C between near-surface regions and the bulk material in differing sources of irradiated graphite, and provides data in support of making an informed assessment for the adoption of low temperature thermal treatment and isotopic reduction of irradiated graphite wastes.

Simulation of crack propagation behavior of nuclear graphite by using XFEM, VCCT and CZM methods

In order to study the fracture performance of nuclear graphite, numerical simulations were performed on single-edge notched graphite beam subjected to three-point bending by using XFEM, CZM and VCCT in ABAQUS. The dependence of peak load Pc on the element size was analyzed and numerical results indicated that the Pc is sensitive to the mesh size using the three methods and it is the most sensitive using VCCT. For the three models, the influences of critical parameters such as maximum principal stress σmax for the XFEM, τc nominal interfacial strength for the CZM and critical strain energy release rate GIc and GIIc for the VCCT, on the load-displacement curves were evaluated. Furthermore, by comparing with the experimental results measured by ESPI technique, the reliability of the finite element model was validated and appropriate values of the parameters for fracture simulation of IG11 graphite by using the three models were proposed. The influences of mesh size and above fracture parameters on the critical crack length ac were analyzed. Finally, through the strain field at the crack tip, the lengths of the crack in graphite were determined.

Microstructural Deformation and Modulus Changes of Nuclear Graphite due to Hydrostatic Pressure Loading

For the first time, the influence of pore closure on elastic modulus in the UK's Magnox and Advanced Gas Cooled Reactors (AGRs) graphite was determined by subjecting the graphite samples to high hydrostatic pressure and dynamic loading. Two grades of nuclear graphite were tested; anisotropic Pile Grade A (PGA) which was used in the Magnox reactors, and semi-isotropic Gilsocarbon as used in the AGRs [1] . The volumetric strain was measured in two directions during compression and separately the velocity of sound through the samples was also determined as a function of confinement pressure. Under hydrostatic loading, the stiffness of PGA graphite was reduced after a few percent of volume strain before increasing again after about ∼20% volumetric strain. Gilsocarbon showed similar behaviour to PGA at lower volumetric strain (∼10 to 13%) however due to Gilsocarbon having a higher density and lower porosity compared to PGA, the response was generally stiffer than PGA. During unloading, significant hysteresis was observed in both graphite grades, with the sample volume almost fully recovering when the pressure was completely removed. Micromechanical models are used to explain the poro-elastic response of graphite during hydrostatic compaction, and to make comparisons and extend data on the influence of pore change due to oxidation on modulus.

OXIDATION BEHAVIOR AND MICRO STRUCTURE ANALYSIS OF NUCLEAR GRAPHITE IG-110 AT 520℃ UNDER AIR ENVIRONMENT

Graphite IG-110 is a nuclear graphite structural and moderator material that has been used for high temperature gas cooled reactors (HTGR). Under normal operating conditions or accidental entry of air or water (air ingress or water ingress), a nuclear graphite. Therefore, the aim of this study is to investigate the oxidation resistant and microstructure change behavior of graphite IG-110 at high temperature under air environment. The sample of IG-110 was tested using Magnetic Suspension Balance (MSB) to analyze the weight change by in-situ for 420 minutes at a temperature of 520 o C. Morphological and microstructure analysis was carried out by optical microscope, SEM-EDS (Scanning Electron Microscope –Energy Dispersive X-ray Sprectroscope) and XRD (X-Ray Diffractometer). The results showed that Graphite IG-110 has a change in surface structure caused by the reaction of the material with oxygen in air at high temperatures. Furthermore, the crystal size of the material structure was slightly change. However, in general, the corrosion rate of graphite IG-110 at a temperature of 520℃ under the air environment is relatively low. So that if graphite IG-110 is exposed to air at a temperature of 520℃ for several hundred minutes in a nuclear reactor estimated does not suffer serious damage

A direct dc arc emission spectrometry determination of chlorine impurities in nuclear grade graphite

In the present study, the direct current arc atomic emission spectrometry (dc arc AES) method was proposed for direct determination of chlorine impurities in nuclear-grade graphites and graphite foils in the range of 1‐–100 ppm in order to meet nuclear power industry requirements. The procedure is based on near-infrared (NIR) AES approach for determination of non-metal impurities, contrary to the most common vacuum ultraviolet AES approach. Carbon dc arc in argon atmosphere was used as radiation source. This radiation source employs standard extra high purity 6 mm diameter standard shaped graphite electrodes and standard vertical spectroanalytic stand, but employs several improvements as well: special protective fused quartz tube, and specially built fiber optic refocuser. Registration of spectra was performed via a novel compact tunable spectrometer with linear complementary metal oxide semiconductor array (CMOS) detector. As the work in near-infrared spectral range is hampered by spectral interference factors, measures were taken to counter them. Due to specifics of the radiation source, the excitability of chlorine was studied both in air and argon in presence of carbon, as well as in its absence. Artificial standards for calibration of the spectrometer were prepared by mixing extra high purity NaCl and graphite powder and evaluated by two independent approaches: indirectly by dc arc AES determination of Na, and directly by water extraction and photometrical determination of Cl. The characteristics of evaporation and excitation in dc arc of chlorine from both standards and real nuclear graphite samples were studied due to the differences in the nature of chlorine source in them. The detection limit of was determined 0.8 μg/g. The whole cycle of measurement, including sample preparation is about 10 min.

A comparison of non-linear optimisation algorithms for recovering the conductivity depth profile of an electrically conductive block using eddy current inspection

This paper compares 15 different non-linear optimisation algorithms for reconstructing the depth profile of electrical conductivity for a conductive block of material using multi-frequency measurements from an eddy current probe. Synthetic data was generated using a finite element model, where the underlying conductivity profile was randomly generated and smoothed, noise was then superimposed on the synthetic data. The compared algorithms were designed such that optimal performance was achieved for each algorithm by using the knowledge of the true underlying conductivity profile used to generate the synthetic data. The most competitive algorithm after ten iterations was determined to be dependent on the a priori estimate and level of noise, where the Broyden–Fletcher–Goldfarb–Shanno, augmented Levenberg-Marquardt and Levenberg-Marquardt are each top of different case studies. The augmented Levenberg-Marquardt and Levenberg-Marquardt were then employed on measured data, where the algorithms used a novel update procedure (in the nuclear graphite application) for the regularisation parameter. The study considers the conductivity profile of graphite blocks of the type used in the nuclear industry.

Effects of damage evolution on tensile strength measurement of nuclear graphite material by ring compression test

• Effects of damage evolution on tensile strength measurements were studied. • Damage evolution should be considered if an indirect method is used for strength measurement. • Uniaxial tensile strength can be measured via the diametrical ring compression test. Because of the limitations of the direct tension method, indirect methods such as Brazil tests or ring compression tests are widely applied to measure the tensile strength of brittle materials. However, in most cases of indirect method application, empirical equations that have been derived based on the assumption of a linear elastic material are used, and the effects of damage on the tensile strength measurements are ignored. Thus, in this study, ring-shaped nuclear graphite IG-11 specimens were subjected to diametrical compression tests to study the effects of damage evolution on tensile strength measurements. By analysing the experimental results and the damage evolution law obtained via an inverse method and finite element analysis, the true uniaxial tensile strength of IG-11 was accurately obtained. The results indicate that damage plays an important role in the tensile strength measurement of the nuclear graphite material when an indirect method is used. This study is also useful as a reference for obtaining accurate tensile strength predictions for other types of brittle materials.

Oxidation Experiments and Kinetics Analysis of Nuclear Graphite ET-10 by Gas Analysis and Microstructure Observation

Nuclear graphite can be used in fission and fusion systems due to its excellent nuclear performance and mechanical properties where the ability of oxidation resistance is usually concerned. Although the excellent performance of new graphite ET-10 was revealed by previous experiments regarding the accident conditions of a fission reactor, further studies are needed to oxidize the graphite under the conditions recommended by the ASTM D7542 standard. A test facility was designed and developed to oxidize the cylindrical specimen with a 10 L/min airflow. According to oxidation rates and microstructures of specimens, the chemical kinetics-controlled regime was determined as 675–750 °C, where the activation energy was obtained as 172.52 kJ/mol. The experiment results revealed the excellent ability of graphite ET-10 for oxidation resistance with lower oxidation rates and longer oxidation times compared with some mainstream graphite. The main reasons are the low contents of some impurities and the binder and the low active surface area due to the non-impregnation baking process undertaken to produce graphite with coal tar pitch coke. It should be noted that the evolution of oxidation behavior at the bottom part of the specimen (facing the airflow) was quicker than that at the upper part of the specimen. We also suggest that the abundance of oxygen supply and the good linearity of the Arrhenius plot are prerequisites of the chemical kinetics-controlled regime rather than sufficient conditions.

Current Status of Purification of Indonesian Natural Graphite as a Candidate for Nuclear Fuel Matrix by Hydrometallurgy and Pyrometallurgy Methods

Indonesia has natural graphite reserved in Sulawesi, Sumatera and Kalimantan islands. The highest graphite content was observed in Sanggau Region, West Kalimantan. Indonesian natural graphite has the potential to be used as fuel matrix in Pebble Bed Reactor (PBR) type - High Temperature Gas Cooled Reactor (HTGR). To increase self-reliance of graphite supply and decrease graphite import for fuel matrix purposes, it is necessary to master the purification technology of graphite. Graphite matrix in Pebble Bed Reactor (PBR) fuel has an important role. It is not only used as neutron moderator and structural material to protect nuclear fuel but also as heat transfer media. Therefore, the graphite matrix must meet the physical and chemical criteria specified for PBR fuel. This paper focuses on current status of purification of Indonesian natural graphite as a candidate for nuclear fuel matrix using hydrometallurgy and pyrometallurgy. Acid and acid mixtures such as HF, HNO3+H2SO4 and HF+HCl+H2SO4 were used for the hydromet-allurgy purification process, while Arc Plasma System were used for pyrometallurgy methods. Characterization for the purified graphite includes physical and chemical properties. The result showed that Indonesian natural graphite obtained from hydrometallurgy was classified into low nuclear grade graphite but it still could be used as nuclear fuel matrix PBR - HTGR. The graphitization degree needs to be increased and the impurity content needs to be decreased for the purified Indonesian graphite using hydrometallurgy method. The Indonesian natural graphite obtained from preliminary pyrometallurgy methods has 96.78 % graphite content according to SEM (surface) observation and the graphite content needs to be increased into nuclear grade graphite.

Oxidation behaviors of matrix-grade graphite during water vapor ingress accidents for high temperature gas-cooled reactors

Water ingress into cores of high temperature gas-cooled reactors (HTGRs) can cause serious safety problems in graphitic components, especially for matrix graphite, which is more vulnerable than nuclear graphite. In this study, oxidation behaviors of a matrix-grade graphite, ARB-B1, were investigated under simulated water ingress accidental conditions. The oxidation tests were conducted in water vapor/helium mixed atmospheres at elevated temperatures up to 1200 °C. Oxidation rates and activation energy were evaluated by mass loss measurements. The activation energy was 121.5 kJ/mol in the considered temperature range, indicating only one oxidation regime−in-pore diffusion assisted by cracks. The cracks originated from the fabrication process, acting as penetration paths for the oxidant, played an important role in the oxidation behaviors. Ungraphitized binder oxidation occurred at all temperatures, while apparent oxidation of filler particles started from 1100 °C. Moreover, oxidation effects on compressive properties and hardness of the matrix graphite were discussed in terms of oxidized microstructures. This first detailed study on matrix graphite oxidation in water vapor provides important kinetics knowledge and microstructural and mechanical data for accidental scenarios of HTGRs as well as for development of new matrix graphite materials.

A statistical analysis of pores and microcracks in nuclear graphite

Microstructure characterization is of great value to understanding nuclear graphite's properties and irradiation behavior. However, graphite is soft and could be easily damaged during sample preparation. A three‐step polishing method involving mechanical polishing, ion milling, and rapid oxidation is proposed for graphite. Ion milling is adopted to remove the broken graphite pieces produced by mechanical polishing. Rapid oxidation is then adopted to remove irradiation‐induced damage layer during ion milling. The Raman spectra show very small G peak width and very low ID/IG ratio after rapid oxidation. The microcracks that were conventionally observed via a transmission electron microscope can be observed on rapid‐oxidized surface in a scanning electron microscope. By digital image processing, the microcracks along with the gas‐escape pores in nuclear graphite IG‐110 are statistically analyzed. Porosity's distributions on crack (pore) size (spanning from 10 nm to 100 μm) are given, which could help to understand and simulate graphite's performances in reactors.

In Situ Mechanical Loading and Neutron Bragg-edge Imaging, Applied to Polygranular Graphite On IMAT@ISIS

BackgroundBragg edge imaging have seen significant developments in the last decade with the availability of new time-resolved detectors, however, there have been no studies of changes in local coherent scattering from grain reorientation and deformation with load. Such damage accommodation mechanism may occur in (quasi)-brittle materials.ObjectiveWe developed a novel method using in-situ Bragg imaging at the ISIS spallation neutron and muon source on the IMAT (Imaging and MATerials science and engineering) instrument using an energy-resolved detector setup. We collected and analysed data of a proof-of-concept experiment demonstrating the use of the method.MethodsWe have developed a loading apparatus that addresses the constraints posed by Bragg imaging, allowing us to resolve features in the material microstructure. We use energy-resolved neutron imaging to obtain images in energy bins and we have developed a set of codes to register and correlate these images, as well as detect changes in local coherent scattering, in situ.ResultsPreliminary results from this method on Gilsocarbon nuclear graphite allow qualitative observation of local changes in Bragg contrast, which may be due to deformation or grain reorientation.ConclusionsWe have demonstrated that we can track changes in local coherent scattering under mechanical load, with sufficient resolution to track features with a size above 100 microns. This method, apparatus and accompanying codes may be used on the IMAT instruments by users interested to better understand deformation in their materials.

Structural and radiological characterization of irradiated RBMK-1500 reactor graphite

This study aims to characterize the irradiated RBMK-1500 nuclear graphite in terms of both structural and radiological properties. The experimental results of morphological and structural analysis of the irradiated graphite samples by using SEM, Raman spectroscopy as well as the theoretical evaluation of primary displacement damage are presented. Moreover, the experimental and theoretical evaluation of the neutron flux is provided and the presence of several γ emitters in the analyzed graphite samples is assessed. Furthermore, the improved version of rapid analysis method for 14C activity determination is applied and the experimentally obtained results are compared with calculated ones. Results indicate that structural changes are uniform enough in all the analyzed samples. However, the distribution of radionuclides is non-homogeneous in the irradiated RBMK-1500 reactor graphite matrix. The comprehensive understanding of both structural and radiological characteristics of nuclear graphite is very important when dealing with decision about irradiated graphite waste management strategy or treatment options prior to its final disposal.

Analysis and evaluation of the irradiated dimensional and volume change behavior of nuclear graphites

Literature on the changes in the specimen dimensions and volume after high energy neutron irradiation of more than a dozen graphites were examined. Some of these data have been previously used to design graphite components in gas cooled reactors. The dimensional change behavior of a graphite is rather unique and depends on both the irradiation temperature and accumulated irradiation dose. The characteristics of the dimensional and volume change behavior of several graphites were analyzed and evaluated from engineering design consideration of graphite components. Specifically, six parameters were derived from the dimensional and volume change behavior. These parameters can serve as “figures-of-merit” (FOM) to enable engineering judgement on the irradiation-induced dimensional and volume change behavior of nuclear graphites, and are referred to as “derived figure-of-merit (DFOM).” The DFOMs also provide engineering bases to examine the opportunity to select several competing graphites for fulfilling the design and operation of components in reactor service. The utility of these DFOMs can be realized potentially in designing inservice inspection and component condition monitoring (CCM) protocols that are effective and efficient both technically and economically. The DFOMs, in combination, also can guide the reactor operator to arrive at optimum component replacement schedule for graphite components, which ensure maintaining graphite components functional requirements and related reactor safety issues.

Neutron thermalization in nuclear graphite: A modern story of a classic moderator

Graphite has been of great interest since the dawn of the nuclear era. In this work, we will focus on the thermal neutron scattering of two types of nuclear graphite that have different microstructures, namely, the historic PGA which has a very anisotropic medium-grained needle coke, and the new G347A which is a near-isotropic fine-grained graphite. We report the results of room temperature inelastic neutron scattering measurements for the scattering function, double differential scattering cross-section, and phonon density of states. The obtained results are discussed in conjunction with results extracted from first-principles lattice dynamics simulations of crystalline graphite, using the incoherent approximation and by also including the coherent-one phonon contribution. Excellent agreements have been found between the measured and calculated scattering functions and phonon densities of states. The inclusion of the coherent-one phonon scattering shows a significant improvement of the calculated scattering functions, double differentials, and inelastic scattering cross-sections. We also highlight the physical concerns about the current thermal neutron scattering ENDF/B-VIII of nuclear graphite library and its corresponding phonon density of states. This work aims to provide a new presentation of neutron thermalization by graphite and offers a new set of high quality experimental and computational data for consideration.

Nuclear graphite strength degradation under varying oxidizing conditions

The oxidized residual strength for two nuclear graphite grades, fine-grained IG-110 and medium-grained NBG-18, were determined over a mass loss range of 0% to 10% for three different oxidation temperatures. This study addresses the critical licensing issue of determining oxidized strength of graphite core components within the design rules of the American Society of Mechanical Engineers (ASME) Boiler Pressure and Vessel Code over a range of oxidation temperatures from 550 °C to 750 °C. Results generally demonstrate low-temperature oxidation to have a 30% greater strength reduction than similar mass losses at higher oxidation temperatures. These differences are attributed to whether the graphite oxidizes primarily within the kinetic-controlled regime or the diffusion-controlled regime which dictate the level of oxidation penetration depth into the interior microstructure. Finally, this study presents a new test method for determining mechanical compression strength from previously oxidized graphite material. This test procedure is based upon two approved American Society for Testing and Materials (ASTM) standards, ASTM D7542 and C695, and combines the requirements for both standards. It is hoped that this new test method can assist ASME BPVC users in determining oxidized residual strength since there are currently no approved standard for conducting such tests.

Probing basal planes and edge sites in polygranular nuclear graphite by gas adsorption: Estimation of active surface area

Polygranular nuclear graphite, manufactured at 2800–3000 °C from a carbonized filler and binder mix, has high graphitization degree, crystalline structure close to perfect graphite, and about 20% porosity. The pore surfaces expose large regions of rough, defective prismatic edges of graphite crystallites which are the locus of graphite materials surface sites active for oxidation, chemisorption, and electron transfer. However, we show that high-resolution N2 and Kr first monolayer adsorption on polygranular graphite (P/P0 < 0.015) occurs in many ways like adsorption on graphitized carbon blacks. This proves the presence of energetically homogeneous basal planes domains in graphite porosity, which was not fully acknowledged before. Using classical analysis methods (Langmuir, Hill-de Boer, adsorption potential distribution) we quantity the basal plane area (BPA) of several polygranular graphite types and correlate it with their microstructure. We propose that gas adsorption is uniquely positioned to reliably estimate the active surface area of graphite by subtracting BPA from the BET total surface area. Direct estimation based on adsorption is preferable to indirect calculations based on microstructural information.