Nuclear Grade Graphite Research Articles

Non-destructive Evaluation of Nuclear Grade IG-110 Graphite Using Constant Potential X-Ray

The main structures of the High Temperature Gas-cooled Reactor’s core are made of graphite material. High temperature and the high flux of neutron exposure in the HTGR reactor affects the integrity of graphite material. In order to ensure the integrity of the graphite materials, and in-service non-destructive inspection is needed. Commonly, prior to the operation, the radiography technique is conducted to evaluate the graphite material. The aim of this study is to develop a non-destructive inspection method in order to evaluate the quality of graphite materials. Five specimens made from nuclear grade graphite IG-110 produced by Toyo Tanso, Co. Ltd has been radiographically tested using X-ray constant potential machine to evaluate the graphite material. The parameter of the energy and intensity of X-ray has also been used in this computed radiography technique with a white type of phosphor imaging plate. A scanning machine of HD CR 35 NDT with 50 μm laser size and a 300 rpm scan rate equipped with a 16-bit system has been used to result in the digital image of the graphites. The results show that the increase of energy and intensity of the radiation to some extent produces better image contrast. The radiograph of the specimens also shows that there is no defect in the material. According to these results, it could be concluded that the digital radiography technique is appropriate for evaluating graphite material.

An XPS/UPS study of the surface/near-surface bonding in nuclear grade graphites: A comparison of monatomic and cluster depth-profiling techniques

Samples of highly-orientated pyrolytic graphite (HOPG) and nuclear graphite grades, Gilsocarbon and Pile Grade-A (PGA), were examined using x-ray photoelectron spectroscopy (XPS), ultra-violet photoelectron spectroscopy (UPS) and Raman spectroscopy. The photoelectron spectra was used to characterise the surface and sub-surface, particularly with regards to the sp2 and sp3 carbon bonding content. A peak-fitting methodology was applied and the results were in good agreement with those obtained through analysis of the C KLL spectra. Depth-profiling was performed using both monatomic Ar+ ions and cluster Arn+ ions with the former found to cause unwanted damage to the graphite structure with a dramatic increase in sp3 content from ~11% to ~88% in both nuclear grades in the ion bombarded region. Monatomic Ar+ etching was also found to result in ion implantation, leading to a broadening of the C 1s line and an increase in high energy component around the CO region at ~286.0 eV. These effects were not observed when etching with cluster Arn+ ions. Raman spectroscopy also confirmed the difference in induced damage between Ar+ and Arn+ with measured ID/IG ration, within the damaged region only (R0), values of 1.04 and 0.3 respectively.

Experimental study on the oxidation behavior and microstructural evolution of NG-CT-10 and NG-CT-20 nuclear graphite

NG-CT-10 and NG-CT-20 are newly developed grades of nuclear-grade graphite from China. In this study, their oxidation behaviors were experimentally investigated using thermal gravimetric analysis. Microstructural evolution before and after oxidation was investigated using scanning electron microscope, mercury intrusion, and Raman spectroscopy. The apparent activation energy of NG-CT-10 nuclear graphite is 161.4 kJ/mol in a reaction temperature range of 550–700 °C and that of NG-CT-20 is 153.5 kJ/mol in a temperature range of 550–650 °C. The activation energy in the inner diffusion control regime is approximately half that in the kinetics control regime. At high temperatures, the binder phase is preferentially oxidized over the filler particles and small pores are generated in the binder. No new large or deep pores are generated on the graphite surfaces. Oxygen can diffuse along the boundaries of filler particles and through the binder phase, but cannot diffuse into the spaces between the nanocrystallites in the filler particles. Filler particles are oxidized starting at their outer surfaces, and the sizes of nanocrystallites do not decrease following oxidation.

Investigation on the oxidation behavior and multi-step reaction mechanism of nuclear graphite SNG742

ABSTRACT Gas chromatography and numerical methods were used to investigate the oxidation behavior of SNG742 nuclear grade graphite at 500–1100°C with different oxidizing gas flow rates. A four-step overall reaction mechanism was used to describe the graphite oxidation reaction process, which reasonably explains the peak of CO mole fraction at around 700°C with lower flow rates. The kinetic parameters of apparent oxidation reaction and four-step reaction mechanism of SNG742 graphite were obtained through curve fitting of the measured data. The apparent activation energy of SNG742 nuclear graphite is 200.7 ± 14.2 kJ/mol. The effects of temperature and gas velocity on the oxidation rate were analyzed. The graphite oxidation process was effectively simulated using CFD method with porous media model and the four-step reaction mechanism. The simulation results show that increasing the gas flow rate can reduce the thickness of the diffusion boundary layer and influence the diffusion control regime of graphite oxidation. The four-step overall reaction mechanism can better describe the graphite oxidation process than the one-step overall reaction. The porous medium model can effectively simulate the mass diffusion process and surface chemical reaction in the interior porous region of the graphite.

Microstructure of electrodeposited tungsten coatings after 30 keV helium ion implantation

Compact and smooth tungsten coatings on nuclear grade graphite (IG-430) substrate were successfully prepared by an electrodeposition method from Na2WO4 - WO3 molten salt. The average grain size and the thickness of the prepared coating was about 20.34 μm and 120 μm, respectively. In addition, the electrodeposited tungsten coatings were subjected to ion implantation with 30 keV helium (He) ions to different fluences to evaluate their irradiation resistances. The microstructure evolution of the electrodeposited tungsten coatings was investigated. The results show that all samples suffered surface blistering after ion implantation, and the tungsten coating surface suffered more morphology changes with the increase of ion fluences. Serious blistering, ripple-like structure and grain boundary separation were formed on the coating surfaces after implantation of the highest He fluence of 1 × 1018 cm−2. Moreover, the levels of grain damage were dependent on the tungsten grain orientations. The levels of damage on tungsten grains with low index orientations increased in the order of (100), (110), and (111).

Characterisation of heterogeneities and evaluation of properties of nuclear grade graphite

Graphite is one of the promising candidates to be used in the next-generation first breeder reactors as a moderator against first moving neutron effluence. During interaction with the neutron, substantial change in microstructure and mechanical and thermal properties occur for graphite. This affects its in-reactor performance. The changes depend on the characteristics of as received virgin material. Therefore, it is necessary to characterize as-received graphite accurately. This paper reports the microstructural characterisation of different nuclear graphite. Four different nuclear graphites were obtained from open source through end-user and were examined in the optical microscope. Quantification of microstructural features was carried out by Analysis Five software. A considerable number of pores with cracks were seen within the matrix. The pores and crack size distribution were found in the range of 0-10 µm2 to a 400-1000 µm2 for the investigated samples. The majority of the pores and cracks were seen within the range of 10-100 µm2 area. The total porosity was to the tune of 9.25 - 5.78%. Pores and cracks of a broad variety of the size were found, and which were uniformly distributed. The theoretical density was calculated from the obtained pores and cracks percentage data. There was no correlation found between the pores and the cracks density (no of pores and cracks per mm2) and the theoretical density of the graphite samples. Density was found to be independent of sizes of pores and cracks of nuclear graphite.

Development of a high temperature facility for study of aerosol emission behavior of combustible materials

Aerosol size and number/mass size distribution are vital parameters for evaluating their effect on human health and the environment. To study the aerosol emission behavior of a material, a high temperature facility has been designed. Combustible materials like biomass, wood, agricultural residue as well as commercially used material like graphite can be burnt and characterized using this facility. In the present work, nuclear grade graphite has been tested in the facility, which can attain a maximum furnace temperature of 1350 °C with an air flow rate of 25 lpm. A cooling and dilution mechanism was incorporated to lower the temperature of the aerosol-laden gas to the desired value for aerosol size measurement using spectrometers. Isokinetic sampling has been ensured using a specially designed probe. Losses due to thermophoresis, condensation and coagulation, etc. have also been calculated and are found to less than 5%.

THE POTENTIAL OF INDONESIAN GRAPHITE AS RDE FUEL MATRIX

THE POTENTIAL OF INDONESIAN GRAPHITE AS RDE FUEL MATRIX. The development plan of Ex- perimental Power Reactor (RDE) in Indonesia is non-commercial and leads to the technology type of Pebble Bed Reactor (PBR) - High Temperature Gas Cooled Reactor (HTGR). The fuel used for PBR reactors is kernel dispersed in spherical fuel elements. The matrix used in PBR nuclear fuel is graphite which functions as a neutron moderator, fuel protective material and heat conductor. Domestication of the domestic fuel matrix needs to be conducted to improve national independence. Therefore, it is necessary to do research on the potential of local graphite to be used as RDE fuel matrix. This study focused on the identification and characterization of local and commercial graphite. The results are compared with the literature, how far it is fulfilling nuclear grade graphite for PBR fuel matrix. Characterization of graphite includes phase analysis with XRD, micro- structure with SEM, surface area/porosity, impurities determination with AAS, ICP-OES and NAA, equivalent boron content, carbon content, density, particle size distribution and ash content. The characterization results show that the carbon content obtained was 87.0 ± 4.2% for local graphite and 100% for commercial graphite. Meanwhile, for the purposes of nuclear graphite it requires a carbon content of >99%. The impurity content in local and commercial graphite still does not meet the RDE fuel matrix standard. The results of XRD analysis show that the local graphite phase is the same as the commercial graphite phase, namely the 2H graphite hexagonal crystal system with the lattice group of P 63/mmc. Particle size distribution and surface area of local graphite are higher compared to nuclear graphite literature. The ash content of commercial graphite was 0.236 ± 0.029 and local graphite was 9.587 ± 0.010%. The results of this study indicate that the local graphite from the flotation still requires a further refinement process to obtain local graphite that can be used as a fuel matrix for RDE.

Purification of Indonesian Natural Graphite by Acid Leaching Method as Nuclear Fuel Matrix: Physical Characterization

Graphite matrix in Pebble Bed Reactor (PBR) fuel has an important role not only as neutron moderator and structural material to protect nuclear fuel, but also as heat transfer media. Therefore, the graphite matrix must meet the criteria of physical and chemical properties specified for PBR fuel. This paper focuses on the purification of the Indonesian natural graphite by using hydrometallurgy method with acid treatments. The characteristic of the purified graphite was studied for its specification compliance as a candidate of fuel matrix for PBR type of High Temperature Gas Cooled Reactor (HTGR). Acid and acid mixtures such as HF, HNO3+H2SO4 and HF+HCl+H2SO4 were used for the purification process. Crystal structure examination by X-Ray Diffraction indicates that the graphite sample was 2H poly type with hexagonal crystal structure and lattice group of P 63 m c space group. It was observed that the graphite sample purified by HNO3+H2SO4 mixture had the closest resemblance to single crystalline graphite with a <d002> deviation of 0.94 when compared to perfect graphite crystal. The density of graphite decreases from 2.3273 g/cm3 (before acid treatment) to 2.1808; 2.2203 and 2.2752 g/cm3 after treatment with HF, HNO3+H2SO4 and HF+HCl+H2SO4, respectively. These results are close to the theoretical density value of 2.26 g/cm3. The surface area decreases from 10.346 m2/g to 6.177; 5.831 and 7.63 m2/g for the treated graphite with HF, HNO3+H2SO4 and HF+HCl+H2SO4 respectively. However, these values are still higher than that of nuclear grade graphite (i.e. between 4.80 and 5.55 m2/g). The average diameter size of graphite decreased from 29.65 μm (before treated acid) into 23.12 μm (after treated acid). The Indonesian natural graphite obtained from acid purification treatment is potential to be used as matrix material for PBR - HTGR fuel, but further treatment is necessary.

A new analysis workflow for discrimination of nuclear grade graphite using laser-induced breakdown spectroscopy

Stand-off, in-situ, laser induced breakdown spectroscopy (LIBS) offers a rapid, safe, and cost-effective method for discrimination of radioactive waste materials arising during the operation of nuclear plants and from decommissioning activities. Characterisation of waste materials is a critical activity in understanding the nature of the waste, ensuring hazardous material is managed safely and that waste can be segregated for reuse, recycle or sentenced for appropriate disposal. Characterisation of materials, often in hostile environments, requires the ability to remotely differentiate between materials in terms of their chemical composition and radioactivity.This proposition was tested using a case study on nuclear grade graphite. Graphite has been used extensively as a moderator material in many nuclear reactors. Internationally, over 250,000 tons of various nuclear-grade graphite, and graphite-bearing, materials exist. These are a major issue for nuclear decommissioning and radioactive waste management, due to the long half-lives of the associated 14C and 36Cl isotopes. LIBS offers a method for discrimination of nuclear grade graphites and other carbon and non-carbon-bearing wastes. This paper describes the development of a workflow method, including LIBS measurement analysis, for the discrimination of pre-irradiated nuclear ‘Pile Grade A’ (PGA) graphite moderator rod and domestic lumpwood charcoal, which act as surrogates for nuclear grade graphite and other carbon-bearing wastes. A new analysis workflow comprising the examination of spectral characteristics, multivariate analysis and molecular isotopic spectroscopy is proposed to enable rapid segregation of graphite from a heterogeneous waste stream. Enhanced characterisation techniques have the potential to significantly reduce the cost of decommissioning large parts of legacy nuclear generation plants.

Effects of ion irradiation and oxidation on point defects in IG-110 nuclear grade graphite

Nuclear grade graphite is a kind of key material in the high temperature gas-cooled reactor pebble-bed module (HTR-PM), where nuclear grade graphite acts as the fuel element matrix material, structural material and neutron reflector. In the reactor, the service environment of nuclear grade graphite suffers high temperature and strong neutron radiation. Both neutron radiation and the oxidation by the oxidizing impurities in HTGR coolant can cause the structure to damage and the properties to deteriorate. Therefore, it is of great significance to study the evolution of defects in nuclear grade graphite for improving the reactor safety. The effects of ion irradiation and oxidation on the point defects in IG-110 graphite are studied in this work. The 190 keV He<sup>+</sup> implantation treatments at room temperature with fluences of 1 × 10<sup>15</sup>, 5 × 10<sup>15</sup>, 1 × 10<sup>16</sup> and 1 × 10<sup>17</sup> cm<sup>–2</sup> are performed to induce 0.029, 0.14, 0.29 and 2.9 displacements per atom respectively. Oxidation treatments are performed at 850 ℃ for 10, 15, 20 and 25 min. Different sequences of He<sup>+</sup> ion irradiation and oxidation are performed, which include irradiation only (Irr.), oxidation only (Ox.), irradiation followed by oxidation (Irr.-Ox.), and oxidation followed by irradiation (Ox.-Irr.). Raman spectrum shows that with the increase of ion irradiation dose, the intensity ratio of D peak to G peak (<i>I</i><sub>D</sub>/<i>I</i><sub>G</sub>) first increases and then decreases, implying that the point defects in graphite are induced by ion irradiation and the point defects evolve as dose increases; the degree of graphitization increases after oxidation, implying that the point defects are recovered by the annealing effect at high temperature, and the point defects decrease after oxidation. This makes Ox.-Irr. samples have a lower point defect content than Irr. samples, and leads Irr.-Ox. samples to possess a higher point defect content than Ox. samples. The positron annihilation Doppler broadening tests reveal that there are only point defects after ion irradiation and oxidation have partially recovered point defects. The ion irradiation and oxidation have opposite effects on the evolution of point defect in graphite. The ion irradiation increases the average <i>S</i>-parameter and reduces the average <i>W</i>-parameter, while oxidation reduces the average <i>S</i>-parameter and increases the average <i>W</i>-parameter. The annealing effect at 850 ℃ cannot completely recover the point defects in Irr.-Ox. samples.

Modeling the effects of oxidation-induced porosity on the elastic moduli of nuclear graphites

Oxidation of nuclear graphites produces microstructural changes that affect the elastic moduli of these materials. It is widely accepted that the primary effect of oxidation is to increase porosity, but the related effect on the moduli cannot be explained satisfactorily by simply noting changes to porosity. In this work, models describing the elastic moduli of porous, polycrystalline graphite materials are developed to interpret experimental determinations of Young's modulus and shear modulus in two grades of nuclear graphite – IG-110 and NBG-18 – that were oxidized to produce varying levels of porosity. Experimental measurements were carried out using laser-based ultrasonic methods and were interpreted successfully using models that take into account the effects of preferential oxidation of different elements of the graphite microstructure. The results indicate the importance of the processes that lead to increased porosity since these can heavily influence the nature of the resulting structure-property relationships.

Oxidation resistance of NITE-SiC/SiC composites with/without CVD-SiC environmental barrier coating

In this study, the oxidation resistance of nuclear grade graphite (IG-430U) and NITE-SiC/SiC composites with/without CVD-SiC outer coating were evaluated at various temperatures (400–1000 °C) under high oxygen contents (O2 1% and O2 21%). The effect of CVD-SiC outer coating on the oxidation of NITE-SiC/SiC composite was also investigated in the atmosphere of partial oxygen. The weight change of NITE-SiC/SiC composites was mainly dominated by the active oxidation at the interphase of PyC. The weight change speed of these materials significantly decreased with increasing the oxidation time. The oxygen concentration was not sensitive for the oxidation behavior of NITE-SiC/SiC composites, because these materials were affected by oxygen supply rate and CO2 gas evaporation rate through fine channels between fiber and matrix. The PyC interphase of NITE-SiC/SiC composites with a CVD-SiC outer coating remained after the oxidation test. It could be confirmed that the CVD-SiC environmental coating on the surface of NITE-SiC/SiC was very effective for the suppression of oxidation weight loss.

Effective gaseous diffusion coefficients of select ultra-fine, super-fine and medium grain nuclear graphite

Understanding “Where?” and “How much?” oxidation has occurred in a nuclear graphite component is critical to predicting any deleterious effects to physical, mechanical, and thermal properties. A key factor in answering these questions is characterizing the effective mass transport rates of gas species in nuclear graphites. Effective gas diffusion coefficients were determined for twenty-six graphite specimens spanning six modern grades of nuclear graphite. A correlation was established for the majority of grades examined allowing a reasonable estimate of the effective diffusion coefficient to be determined purely from an estimate of total porosity. The importance of Knudsen diffusion to the measured diffusion coefficients is also shown for modern grades. Knudsen diffusion has not historically been considered to contribute to measured diffusion coefficients of nuclear graphite.

Fluorination of nuclear graphite IG-110 in molten 2LiF-BeF2 (FLiBe) salt at 700 ºC

It is important for the development of advanced nuclear reactors to understand if the exposure to molten fluoride salts changes physical properties of graphite and its ability to absorb tritium. This paper demonstrated that the interaction between high purity nuclear grade graphite IG-110 and molten salt mixture 2LiF-BeF2 (FLiBe) at 700 °C for twelve hours in argon envirnoment resulted in graphite fluorination and changes in graphite microstructure. XPS C1s spectra analysis shows an increase in F functionalization (from 3.5% to 9.0–19.3%), an increase in the sp3/sp2 ratio (from 0.1496 to 0.1798), and F1s spectrum analysis confirms appearance of CF bonds. Raman spectra show an increase in ID/IG ratio from 0.35 to 0.69 and an appearance of peak shoulders characteristic of CF bond formation. GDMS depth analysis from the surface to 5 μm depth shows a higher depth penetration of F compared to that of Li and Be, indicative of surface salt reduction and bulk graphite fluorination. FLiBe reduction by graphite fluorination is not thermodynamically favorable, but certain carbon sites in the graphite are more reactive than the majority of carbon atoms in graphite. We argue that graphite fluorination occurred at these reactive sites, leading to the reduction of salt at the graphite surface. XRD and Raman confirmed microstructural changes. This indicates that the fluorination process may introduce new active sites as existing active sites are consumed, which could be advantageous for tritium chemisorption from FLiBe.

Neutronics Analysis of Small Compact Prismatic Nuclear Reactors for Space Crafts

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235U mass and the total reactor mass.

Experimental study on 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event

The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR) accident scenarios. These include depressurized loss of forced cooling (D-LOFC) events that allow for the possibility of air ingress into the reactor pressure vessel as a result of a break in the helium pressure boundary, which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. To characterize air ingress into the vented low-pressure containment of the next generation nuclear plant HTGR during hypothetical moderate-sized D-LOFC break accidents experimentally, a 1/28 scaled simplified reactor building model was established. A non-dimensional similarity approach was employed for scaling of the experimental facility. Three experiments were designed and conducted to study the dynamic response to the accident scenario. The experimental results suggest an increment in the flow area of the check valve between the reactor cavity (CV1) and steam generator cavity (CV3). Furthermore, qualitative analysis was conducted on the experimental data.

Comparison of NBG-18, NBG-17, IG-110 and IG-11 oxidation kinetics in air

The oxidation rates of several nuclear-grade graphites, NBG-18, NBG-17, IG-110 and IG-11, were measured in air using thermogravimetry. Kinetic parameters and oxidation behavior for each grade were compared by coke type, filler grain size and microstructure. The thickness of the oxidized layer for each grade was determined by layer peeling and direct density measurements. The results for NBG-17 and IG-11 were compared with those available in the literature and our recently reported results for NBG-18 and IG-110 oxidation in air. The finer-grained graphites IG-110 and IG-11 were more oxidized than medium-grained NBG-18 and NBG-17 because of deeper oxidant penetration, higher porosity and higher probability of available active sites. Variation in experimental conditions also had a marked effect on the reported kinetic parameters by several studies. Kinetic parameters such as activation energy and transition temperature were sensitive to air flow rates as well as sample size and geometry.

A multi-technique experimental and modelling study of the porous structure of IG-110 and IG-430 nuclear graphite

In nuclear graphite, the wide range of void sizes precludes a full characterisation of pore volume by means of a single technique. A novel multi-technique approach, consisting of pycnometry, low pressure gas adsorption and mercury porosimetry is presented. The approach is validated for two nuclear-grade graphites designed for use in Generation IV nuclear reactors, namely IG-110 and IG-430. Damage and deformation caused to the structure of the graphite by mercury intrusion is estimated by consecutive intrusion experiments. The damage is assumed to be caused by the highest applied pressures of mercury. It is compensated by substituting that part of the percolation curve with one derived from adsorption measurements. The various measurements are inverse modelled in a way which intelligently bridges the size gap between the techniques. The resulting complete non-hierarchical pore structure covers sizes spanning 4 orders of magnitude. The new approach resolves the long standing issues associated with performing porosimetry on graphitic samples, and fills the gap in knowledge for the assessment of multilevel porosity within graphite. As an example of the possible applications of the resulting void network structure, we calculated the air network flow capacity, related to absolute permeability, for the two graphite samples.

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

Four grades of nuclear graphite were oxidized in helium with traces of moisture and hydrogen in order to evaluate the effects of slow oxidation by moisture on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures, oxidation was always faster than the LH model predicts. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation with temperature which follows the integral Boltzmann distribution function. This suggests deviations from the LH model are apparently caused by activation with temperature of graphite reactive sites, which is probably rooted in specific structural and electronic properties of graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. This new Boltzmann-enhanced Langmuir-Hinshelwood (BLH) model consistently predicts oxidation rates over large ranges of temperature (800–1100 °C) and partial pressures of water (3–1200 Pa) and hydrogen (0–300 Pa). The BLH model can be used for modeling chronic oxidation of graphite components during life-time operation in high- and very high temperature advanced nuclear reactors.