Nuclear Grade Graphite Research Articles

Effects of specimen size on the flexural strength and Weibull modulus of nuclear graphite IG-110, NBG-18, and PCEA

Changes in flexural strength and Weibull modulus due to specimen size were investigated for three nuclear graphite grades, IG-110, NBG-18, and PCEA, using four-point-1/3 point (4-1/3) loading with specimens of three different sizes: 3.18 (Thickness)×6.35 (Width)×50.8 (Length), 6.50(T)×12.0(W)×52.0(L), 18.0(T)×16.0(W)×64(L) (mm) (total: 210 specimens). Results showed some specimen size effects were grade dependent: While NBG-18 (a) showed rather significant specimen size effects (37% difference between the 3T and 18T), the differences in IG-110 and PCEA were 7.6–15%. The maximum differences in flexural strength due to specimen size were larger in the PCEA and NBG-18 having larger sized coke particles (medium grain size: >300μm) than the IG-110 with super fine coke particle size (25μm). The Weibull modulus showed a data population dependency, in that it decreased with increasing numbers of data used for modulus determination. A good correlation between the fracture surface roughness and the flexural strength was confirmed.

Experimental investigation and CFD analysis on cross flow in the core of PMR200

The Prismatic Modular Reactor (PMR) is one of the major Very High Temperature Reactor (VHTR) concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear grade graphite. However, the shape of the graphite blocks could be easily changed by neutron damage during the reactor operation and the shape change can create gaps between the blocks inducing the bypass flow. In the VHTR core, two types of gaps, a vertical gap and a horizontal gap which are called bypass gap and cross gap, respectively, can be formed. The cross gap complicates the flow field in the reactor core by connecting the coolant channel to the bypass gap and it could lead to a loss of effective coolant flow in the fuel blocks. Thus, a cross flow experimental facility was constructed to investigate the cross flow phenomena in the core of the VHTR and a series of experiments were carried out under varying flow rates and gap sizes. The results of the experiments were compared with CFD (Computational Fluid Dynamics) analysis results in order to verify its prediction capability for the cross flow phenomena. Fairly good agreement was seen between experimental results and CFD predictions and the local characteristics of the cross flow was discussed in detail. Based on the calculation results, pressure loss coefficient across the cross gap was evaluated, which is necessary for the thermo-fluid analysis of the VHTR core using a lumped parameter code.

Total hemispherical emissivity of Inconel 718

Total hemispherical emissivity for Inconel 718 was measured in anticipation of its application in Very High Temperature Gas Reactors (VHTRs). A majority of current emissivity data for Inconel 718 is in the form of spectral measurements. The data presented here were obtained with an experimental apparatus based on the standard ASTM C835-06 for total hemispherical emittance. Measurements of Inconel 718 were made for four different surface types including: (i) ‘as-received’ from the manufacturer, (ii) oxidized in air and humidified helium, (iii) sandblasted with aluminum oxide powder, and (iv) with a thin coating of nuclear grade graphite powder (grade NGB-18). The emissivity for the ‘as-received’ sample ranged from 0.21 to 0.28 in the temperature interval from 760K to 1275K. Oxidation in air at 1073K resulted in an increase in emissivity into the range from 0.2 at 650K to 0.52 at 1200K. There was no dependence on the oxidation times studied here. Oxidation with humidified helium at 1073K produced less of an increase in emissivity than the oxidation in air but there was an increase up to the range from 0.2 at 600K to 0.35 at 1200K. Sandblasting of Inconel 718 was also observed to increase the emissivity up to the range from 0.43 at 780K to 0.53 at 1270K when 60 grit sized powder was used and up to the range from 0.45 at 683K to 0.57 at 1267K when 120 and 220 grit sized powders were used. Coating of graphite powder onto the ‘as-received’ Inconel 718 showed no increase in the emissivity whereas the coating of graphite powder onto the 220 grit sandblasted Inconel 718 did show an increase in emissivity up to the range from 0.60 at 650K to 0.70 at 1200K. This was apparently due to better adhesion of graphite on the roughened surface.

Simulation of graphite oxidation in oxygen at 400–800°C

A computer model of nuclear-grade graphite is proposed that allows a simulation of graphite-sample oxidation to be performed (with an error 7%) taking into account the granulometric composition, porosity, and density. The results from simulating the GMZ graphite oxidation in the regime of chemical kinetics showed a satisfactory agreement with experimental data.

Identification and location of 14C-bearing species in thermally treated neutron irradiated graphites NBG-18 and NBG-25: Pre- and post-thermal treatment

Recent studies have been performed to determine the effectiveness of thermal treatment as a method for removing 14C contamination from irradiated graphite surfaces. Samples of two grades of irradiated nuclear graphite (NBG-18 and NBG-25) were thermally treated to determine the amount of 14C contamination on irradiated graphite surfaces. The results of these analyses indicate that specific chemical forms of 14C (namely, 14CO and 14CO2) may be selectively removed based on the temperature used during thermal treatment. Characterization studies utilizing various surface analysis techniques (XPS, SIMS, SEM/EDS) were employed to investigate the chemical speciation, bond structure, and morphology of the surfaces of pre- and post-thermally treated irradiated graphite.

Model for neutron total cross-section at low energies for nuclear grade graphite

At subthermal neutron energies, polycrystalline graphite shows a large total cross-section due to small angle scattering processes. In this work, a new methodology to determine pore size distributions through the neutron transmission technique at subthermal energies is proposed and its sensitivity is compared with standard techniques. A simple model based on the form factor for spherical particles, normally used in the Small Angle Neutron Scattering technique, is employed to calculate the contribution of small angle effect to the total scattering cross-section, with the width and center of the radii distributions as free parameters in the model. Small Angle X-ray Scattering experiments were performed to compare results as a means to validate the method. The good agreement reached reveals that the neutron transmission technique is a useful tool to explore small angle scattering effects. This fact can be exploited in situations where large samples must be scanned and it is difficult to investigate them with conventional methods. It also opens the possibility to apply this method in energy-resolved neutron imaging. Also, since subthermal neutron transmission experiments are perfectly feasible in small neutron sources, the present findings open new possibilities to the work done in such kind of facilities.

First principles study of oxidation behavior of irradiated graphite

The relationship between nuclear graphite microstructure and its oxidation resistance underlines the importance of comprehensive oxidation characterization studies of the new grades of nuclear graphite. Periodic DFT calculations are performed to model oxidation behavior of irradiated graphite. O2 molecules adsorbed on perfect and defective graphite surfaces are calculated. The adsorptive energy of O2 on defective graphite adsorption site with one carbon atom missing is approximately 10 times as strong as that on a defect-free perfect graphite surface. Monovacancy and divacancy on graphite surface can easily chemisorb O2 molecule compared to perfect surface. Two oxidation processes including CO and CO2 formation steps are analyzed. For symmetric monovacancy defect, three dangling C atoms are unsaturated and exhibit high adsorption ability, as well as reconstructed monovacancy and divacancy defects. These vacancy defects in irradiated graphite decrease oxidation resistance of nuclear graphite.

Molten FLiNaK salt infiltration into degassed nuclear graphite under inert gas pressure

Infiltration of molten FLiNaK salt into degassed nuclear graphite samples under inert gas pressure was studied. The weight gain of different grades (2020, 2114, IG-110, NBG-8, G1 and G2) of nuclear graphite during infiltration with different pressures was measured. Molten salt infiltration was compared with mercury intrusion porosimetry where it was found that mercury infiltration was a useful predictor of the threshold pressure and infiltration volume per gram graphite for molten salt infiltration. The distribution and morphology of salt in the graphite were observed by scanning electron microscopy, with very little difference between the molten salt content at the center and edge of samples for samples infiltrated at pressure higher than the threshold pressure. Increased molten salt infiltration with increased pressure resulted from the occupation of smaller pores and full occupation of the larger irregular pores. The similarity of weight gain between molten salt infiltration equilibrated at 20 and 100h showed 20h was adequate to obtain equilibrium.

Electron irradiation of nuclear graphite studied by transmission electron microscopy and electron energy loss spectroscopy

Structural and chemical bonding changes in nuclear graphite have been investigated during in-situ electron irradiation in a transmission electron microscope (TEM); electron beam irradiation has been employed as a surrogate for neutron irradiation of nuclear grade graphite in nuclear reactors. This paper aims to set out a methodology for analysing the microstructure of electron-irradiated graphite which can then be extended to the analysis of neutron-irradiated graphites. The damage produced by exposure to 200keV electrons was examined up to a total dose of approximately 0.5dpa (equivalent to an electron fluence of 5.6×1021electronscm−2). During electron exposure, high resolution TEM images and electron energy loss spectra (EELS) were acquired periodically in order to record changes in structural (dis)order and chemical bonding, by quantitatively analysing the variation in phase contrast images and EEL spectra.

Simulating Neutron Radiation Damage of Graphite by In-situ Electron Irradiation

Radiation damage in nuclear grade graphite has been investigated using transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS). Changes in the structure on the atomic scale and chemical bonding, and the relationship between each were of particular interest. TEM was used to study damage in nuclear grade graphite on the atomic scale following 1.92×108 electrons nm−2 of electron beam exposure. During these experiments EELS spectra were also collected periodically to record changes in chemical bonding and structural disorder, by analysing the changes of the carbon K-edge. Image analysis software from the 'PyroMaN' research group provides further information, based on (002) fringe analysis. The software was applied to the micrographs of electron irradiated virgin 'Pile Grade A' (PGA) graphite to quantify the extent of damage from electron beam exposure.

Dynamic Tensile Characteristics of Nuclear-Grade Graphite IG-11

The effect of strain rate up to nearly = 102/s on the tensile stress-strain properties of isotropic fine-grained nuclear-grade graphite IG-11 was investigated. Cylindrical tensile specimens machined out of graphite bars were used in both static and dynamic tests. The dynamic tensile stress-strain curves up to fracture were determined using the split Hopkinson bar (SHB). The low and intermediate strain-rate tensile stress-strain relations up to fracture were measured on an Instron 5500R testing machine. It was demonstrated that the ultimate tensile strength increases slightly, while the fracture strain and absorbed energy up to fracture decrease dramatically with increasing strain rate. Macro and microscopic examinations revealed a slight difference in the fracture surfaces between the static and dynamic tension specimens.

Removal of carbon-14 from irradiated graphite

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. On of the isotopes of great concern for long-term disposal of irradiated graphite is carbon-14 (14C), with a half-life of 5730years. Study of irradiated graphite from some nuclear reactors indicates 14C is concentrated on the outer 5mm of the graphite structure. The aim of the research presented here is to develop a practical method by which 14C can be removed. In parallel with these efforts, the same irradiated graphite material is being characterized to identify the chemical form of 14C in irradiated graphite.A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam®, were exposed to liquid nitrogen (to increase the quantity of 14C precursor) and neutron-irradiated (1013neutrons/cm2/s). During post-irradiation thermal treatment, graphite samples were heated in the presence of an inert carrier gas (with or without the addition of an oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon only were performed at 900°C and 1400°C to evaluate the selective removal of 14C. Thermal treatment also was performed with the addition of 3 and 5vol% oxygen at temperatures 700°C and 1400°C. Thermal treatment experiments were evaluated for the effective selective removal of 14C. Lower temperatures and oxygen levels correlated to more efficient 14C removal.

Determination of concentrations of trace elements in nuclear grade graphite by charged particle activation analysis

A programme has been undertaken to develop an in-house standard reference material of graphite using various analytical techniques including nuclear techniques like charged particle activation analysis (CPAA) and neutron activation analysis (NAA). The concentrations of 13 elemental impurities viz Ca, Ti, V, Cr, Fe, Ni, Cu, Zn, Ga, Ge, Sr, Zr, Mo have been determined at mg kg−1 to μg kg−1 levels in graphite by CPAA through instrumental approach using 13 and 18 MeV proton beam from cyclotron at Variable Energy Cyclotron Centre, Kolkata. The analytical results of CPAA have been validated by analyzing the same graphite samples by instrumental NAA. The validation of our analytical results has also been performed by the Z-score tests through the determination of concentrations of the elements of interest in a certified reference material (CRM) taking another CRM as standard. The experimental detection limits of all elements determined in graphite material by CPAA are also reported.

Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

Oxidation rate of nuclear-grade graphite IG-110 in the kinetic regime for VHTR air ingress accident scenarios

The oxidation rates of nuclear-grade graphite IG-110 in the kinetically-controlled temperature regime of graphite oxidation were predicted and compared in Very High Temperature Reactor air ingress accident scenarios. The oxidative mass loss of graphite was measured thermogravimetrically from 873 to 1873K in 100% air (21mol%). The activation energy was found to be 222.07kJ/mol, and the order of reaction with respect to oxygen concentration is 0.76. The surfaces of the samples were characterized by Scanning Electron Microscopy, Energy Dispersive Spectroscopy, Fourier Transform Infrared Spectroscopy and X-ray Photoelectron Spectroscopy before and after oxidation. These results are compared with those available in the literature, and our recently reported results for NBG-18 nuclear-grade graphite using the same technique.

Chemical kinetics parameters and model validation for the gasification of PCEA nuclear graphite

A series of gasification experiments, using two right cylinder specimens (∼12.7×25.4mm and 25.4×25.4mm) of PCEA nuclear graphite in ambient airflow, measured the total gasification flux at weight losses up to 41.5% and temperatures (893–1015K) characteristics of those for in-pores gasification Mode (a) and in-pores diffusion-limited Mode (b). The chemical kinetics parameters for the gasification of PCEA graphite are determined using a multi-parameters optimization algorithm from the measurements of the total gasification rate and transient weight loss in experiments. These parameters are: (i) the pre-exponential rate coefficients and the Gaussian distributions and values of specific activation energies for adsorption of oxygen and desorption of CO gas; (ii) the specific activation energy and pre-exponential rate coefficient for the breakup of stable un-dissociated C(O2) oxygen radicals to form stable (CO) complexes; (iii) the specific activation energy and pre-exponential coefficient for desorption of CO2 gas and; (iv) the initial surface area of reactive free sites per unit mass. This area is consistently 13.5% higher than that for nuclear graphite grades of NBG-25 and IG-110 and decreases inversely proportional with the square root of the initial mass of the graphite specimens in the experiments. Experimental measurements successfully validate the chemical-reactions kinetics model that calculates continuous Arrhenius curves of the total gasification flux and the production rates of CO and CO2 gases. The model results at different total weight losses agree well with measurements and expand beyond the temperatures in the experiments to the diffusion-limited mode of gasification. Also calculated are the production rates of CO and CO2 gases and their relative contributions to the total gasification rate in the experiments as functions of temperature, for total weight losses of 5% and 10%.

The thermal performance of fuel matrix material in a CO2 atmosphere

The thermal oxidation performance of a semi-graphitic fuel matrix-material has been compared to two grades of nuclear graphite between 600°C and 1200°C in flowing CO2. Fuel matrix material is used to produce compacts or pebbles containing TRISO coated particle fuel for High Temperature Reactors (HTRs). The A3-27 fuel matrix-material grade was compared to NBG-18 and Gilsocarbon nuclear graphite grades. At 1200°C temperatures A3-27 appears to be more reactive than NBG-18, but less so than Gilsocarbon. At 600°C the oxidation rate of A3-27 is comparable to that of NBG-18, but both are significantly higher than that of Gilsocarbon. It is concluded that the comparable thermal oxidation behaviour of graphite and fuel-matrix material suggests that operating temperatures in a CO2 cooled reactor fuelled with TRISO coated particle fuel would not need to be reduced below those considered acceptable for the use of nuclear graphite.

Computational and experimental prediction of dust production in pebble bed reactors—Part I

This paper describes the computational modeling and simulation, and experimental testing of graphite moderators in frictional contacts as anticipated in a pebble bed reactor. The potential of carbonaceous particulate generation due to frictional contact at the surface of pebbles and the ensuing entrainment and transport into the gas coolant are safety concerns at elevated temperatures under accident scenarios such as air ingress in the high temperature gas-cooled reactor. The safety concerns are due to the documented ability of carbonaceous particulates to adsorb fission products and transport them in the primary circuit of the pebble bed reactor, thus potentially giving rise to a relevant source term under accident scenarios. Here, a finite element approach is implemented to develop a nonlinear wear model in air environment. In this model, material wear coefficient is related to the changes in asperity height during wear. The present work reports a comparison between the finite element simulations and the experimental results obtained using a custom-designed tribometer. The experimental and computational results are used to estimate the quantity of nuclear grade graphite dust produced from a typical anticipated configuration. In Part II, results from a helium environment at higher temperatures and pressures are experimentally studied.

Fracture behavior of nuclear grade graphite under mixed mode I-II

Mixed mode I and mode II fracture toughness values of five grades of nuclear graphites were measured at room temperature using a centrally notched disk specimen. The specimens were compressed diametrically with a crosshead speed of 0.125 mm/ min in ambient air. The fracture loads of the nuclear graphites decreased as the notch inclination angle increased under mixed mode I-II loading. The nuclear grade graphites with the higher bulk density and coarser coke particles showed higher mode I and mode II fracture toughness. The ratios of KIIc/KIc were in the range of 1.4 ~ 1.5 regardless of the coke type and size

Oxidation rate of nuclear-grade graphite NBG-18 in the kinetic regime for VHTR air ingress accident scenarios

One of the most severe accident scenarios anticipated for VHTRs is an air ingress accident caused by a pipe break. Graphite oxidation could be severe under these conditions. In this work, the oxidation rate of NBG-18 nuclear-grade graphite was studied thermogravimetrically for different oxygen concentrations and with temperatures from 873 to 1873K. A semi-empirical Arrhenius rate equation was developed for the temperature range of 873–1023K. The activation energy of NBG-18 was 187kJ/mol and the order of reaction was 1.25. The penetration depth of oxidant was about 3–4mm for NBG-18 oxidized at 973K. Increased porosity and changes in external geometry became more prominent at higher temperatures from about 1173 to 1873K. The surface of oxidized NBG-18 was characterized by SEM, EDS, FTIR and XPS.