Grade Graphite Research Articles

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.

Effect of friction on oxidative graphite intercalation and high-quality graphene formation

Oxidative wet-chemical delamination of graphene from graphite is expected to become a scalable production method. However, the formation process of the intermediate stage-1 graphite sulfate by sulfuric acid intercalation and its subsequent oxidation are poorly understood and lattice defect formation must be avoided. Here, we demonstrate film formation of micrometer-sized graphene flakes with lattice defects down to 0.02% and visualize the carbon lattice by transmission electron microscopy at atomic resolution. Interestingly, we find that only well-ordered, highly crystalline graphite delaminates into oxo-functionalized graphene, whereas other graphite grades do not form a proper stage-1 intercalate and revert back to graphite upon hydrolysis. Ab initio molecular dynamics simulations show that ideal stacking and electronic oxidation of the graphite layers significantly reduce the friction of the moving sulfuric acid molecules, thereby facilitating intercalation. Furthermore, the evaluation of the stability of oxo-species in graphite sulfate supports an oxidation mechanism that obviates intercalation of the oxidant.

Mesoscopic structure features in synthetic graphite

The mesocopic structure features in the coke fillers and binding carbon regions of a synthetic graphite grade have been examined by high resolution transmission electron microscopy (TEM) and Raman spectroscopy. Within the fillers, the three-dimensional structure is composed of crystal laminae with the basal plane dimensions (La) of hundreds nanometres, and thicknesses (Lc) of tens of nanometres. These laminae have a nearly perfect graphite structure with almost parallel c-axes, but their a–b planes are orientated randomly to form a “crazy paving” structure. A similar structure exists in the binding carbon regions, with a smaller La. Significantly bent laminae are widely seen in quinoline insoluble inclusions and the graphite regions developed around them. The La values measured by TEM are consistent with estimates from the intensity ratios of the D to G Raman peak in these regions. Atomistic modelling finds that the lowest energy interfaces in the crazy paving structure comprise 5, 6 and 7 member carbon rings. The bent laminae tend to maintain the 6 member rings, but are strained elastically. We suggest that a 7 member carbon ring leaves a cavity representing an arm-chair graphite edge contributing to the Raman spectra D peak.

Experience of on-site disposal of production uranium-graphite nuclear reactor

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment.

Numerical Simulations of Laser-Driven Cratering Experiments into Porous Graphite

We present the results of an experimental campaign conducted on the LULI2000 laser facility. Semi-infinite targets of a commercial grade of porous graphite were submitted to high-power laser irradiation inorder to generate craters. A 15 ns pulse duration was used along with a focal spot diameter of 900 µm to deliver energies up to 750 J. Numerical simulations of these shots have been performed following a specificmethodology which can be divided in three steps. Firstly, the mechanical loading induced by the laser iscalibrated by simulating the same shot on a thin aluminum target of which free surface velocity is measured byPDV and line-VISAR. Secondly, the same shot is performed on a thin graphite target to validate the materialmodel of graphite. Thirdly, the craterization shot on semi-infinite target is simulated. Numerical results arecompared to experimental measurements of craters obtained using an interferometric profilometer.

Effect of austempering temperature on microstructure and properties of austempered ductile iron

The influence of austempering temperature on microstructure and properties of Austempered Ductile Iron were studied in this paper. The purpose was to improve comprehensive performance of Austempered Ductile Iron for adapting to slurry pump transflux parts. The nodularisation grade of graphite was 2-3 magnitude and the size of graphite was 6-7 magnitude. After austempered at different temperatures for 120 min, the microstructure mainly consisted of acicular ferrite, retained austenite and spheroidal graphite. With increasing austempering temperature, tensile strength improved firstly and then decreased, impact toughness increased, the wear resistance became worse, and the corrosion resistance increased. When the heat treatment process was 900°C × 90 min + 250°C × 120 min, the comprehensive performance was better (tensile strength: 1710 MPa, elongation: 2%, impact toughness: 40 J/cm2, Brinell hardness: 467 HBW, wear weightlessness rate: 0.17 mg/m) than the requirements in ASTM A897-06.

Effect of austempering temperature on microstructure and properties of austempered ductile iron

The influence of austempering temperature on microstructure and properties of Austempered Ductile Iron were studied in this paper. The purpose was to improve comprehensive performance of Austempered Ductile Iron for adapting to slurry pump transflux parts. The nodularisation grade of graphite was 2-3 magnitude and the size of graphite was 6-7 magnitude. After austempered at different temperatures for 120 min, the microstructure mainly consisted of acicular ferrite, retained austenite and spheroidal graphite. With increasing austempering temperature, tensile strength improved firstly and then decreased, impact toughness increased, the wear resistance became worse, and the corrosion resistance increased. When the heat treatment process was 900°C × 90 min + 250°C × 120 min, the comprehensive performance was better (tensile strength: 1710 MPa, elongation: 2%, impact toughness: 40 J/cm2, Brinell hardness: 467 HBW, wear weightlessness rate: 0.17 mg/m) than the requirements in ASTM A897-06.

Structural and mechanical properties of MPG-7 graphite grade

We report the elastic and mechanical properties of MPG-7,8 graphite at room temperatures. Differences in microstructure, elastic and strength characteristics of graphite samples (of two types of graphite) were investigated in detail. The composition and structural properties were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), and Raman microscopy analysis. The second-order elastic constants (elastic modulus) of MPG-7,8 samples were measured by ultrasonic method. The tests of bending strength by three-point bending method were carried out using the Instron 5982 machine. We observed significant differences both in density and in elastic and strength characteristics of these kinds of graphite.

Electrodeposition of Au on basal plane graphite and graphene

Electrodeposition is a useful route to nanoparticle formation that allows control over nucleation and growth by tuning the applied potential. Here, electrodeposition is used to create Au structures on highly oriented pyrolytic graphite and monolayer graphene electrodes. It is demonstrated that spontaneous formation of Au occurs immediately on exposure to the solution, which strongly impacts the cyclic voltammetry and obfuscates the underlying deposition mechanism. Voltammetric measurements are combined with ex situ and in situ electrochemical atomic force microscopy to reveal the processes corresponding to Au deposition and stripping. Deposition on spontaneously formed Au dominates the voltammetry, with the deposition on clean areas of the electrode requiring a much greater overpotential. Additionally, comparison of the extent of deposition and the potentials required for different grades of graphite and monolayer graphene are found to be linked to the density of defects of the electrode. Consideration of such factors is important for obtaining small, highly uniform functionalisation of graphene-based materials suitable for applications.

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

This study investigates the burn-up characteristics and the criticality effect of impurities in the graphite structure of commercial-scale prismatic High Temperature Gas-cooled Reactor (HTGR), and thereby reconsiders the necessity of high-grade graphite material. In an HTGR, the core is filled with the graphite, and the impurities in the graphite have a non-negligible poison effect on the criticality. To account for the effect of the reflector blocks deployed adjacent to the fuel blocks, GTHTR300, commercial-scale HTGR, employed fine purified grade graphite material IG-110. Ideally, the fuel blocks should also employ IG-110; however, for economic purposes they are constructed from an un-purified grade graphite material IG-11. The poisoning effect of the impurity (which behaves like 10B burn-up and is expressed in boron equivalents) decreases exponentially and eventually saturates at 1% of the initial boron equivalent. However, the reactivity worth of the fuel and reflector blocks with 0.03 ppm boron equivalents (equivalent to 1% of IG-11) is negligible (i.e., <0.01%Δ k/kk′). Because the poisoning effect of the impurity mimics that of naturally occurring boron, it was evaluated in whole-core burn-up calculations with the impurities represented by naturally occurring boron.According to the results, the criticality of the commercial-scale HTGR is unaffected by the impurity levels (even in the un-purified grade IG-11) because the impurities burn cleanly until the End of Cycle (EOC). Therefore, the economy of electricity generation by HTGRs can be improved by using the un-purified grade IG-11 instead of the fine purified grade graphite IG-110.

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.

Thermophysical properties of graphite HOPG and HAPG in the solid state and under melting (from 2000 K up to 5000 K)

Experiments with HOPG graphite grade showed that the melting temperature of graphite equals 4800-4900 K and that the melting of graphite is possible only at elevated pressures. The data were obtained for resistivity, specific heat and input (Joule) energy up to 5000 K. HAPG (Highly Annealing Pyrolytic Graphite) is a form of highly oriented pyrolytic graphite. HAPG specimens in the form of strips (thickness 30 microns) were placed in a cell (between two plates of glass-sapphire). The specimen temperature was measured by a high speed pyrometer. The heat of fusion for both graphite grades (heated in a confined volume) was less (and specific heat – higher) than for the case with nearly free expansion. A possible reason for the observed effects is discussed in the report.

The influence of nuclear graphite oxidation on air ingress accident of HTR-PM

The high temperature gas-cooled reactor pebble bed module (HTR-PM) has wide potential applications in the development of advanced nuclear energy. The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a kind of beyond design basis accident of HTR-PM, which will result in air ingress into the reactor core. Even though the rare probability of this accident, it still has got a lot of concerns in the safety analyses of HTR-PM, due to the possible severe consequences of the air ingress accident, which would cause oxidation of the graphite material in the reactor. The oxidation of matrix graphite of spherical fuel elements may lead to the exposure of the TRISO fuel particles and the increase of the radioactivity release. The oxidation of the bottom reflector may cause the decrease of mechanical strength of the nuclear grade graphite. Different kinds of nuclear graphite of the bottom reflector will influence the consequences of the DEGB of the horizontal coaxial gas duct accident. This paper mainly analyzed the influence of the nuclear grade graphite used as the bottom reflector on the consequences of this accident by using the system analysis TINTE code.

An understanding of lattice strain, defects and disorder in nuclear graphite

In this study, microstructural parameters, such as lattice dimension, micro-strain and dislocation density, of different neutron-irradiated graphite grades have been evaluated using the diffraction profiles of X-ray diffraction (XRD) and the scattering profiles of Raman spectroscopy. Using Generation-IV candidate graphite samples (grade PCEA, GrafTech), subjected to neutron irradiation at 900 °C to 6.6 and 10.2 dpa, and graphite samples of similar grain size and microstructure taken from the core of the British Experimental Pile Zero reactor, which have been irradiated at 100–120 °C to 1.60 dpa, an investigation is presented on the effect of irradiation dose and temperature on the aforementioned microstructural parameters. Using two complementary techniques in Raman spectroscopy and XRD, which produced agreeable results, the average lateral crystallite size of the two graphites tested was found to decrease with increasing exposure to fast-neutron irradiation or at lower irradiation temperatures. Conversely, dislocation densities and micro-strains were found to increase following the same changes in irradiation conditions. Supporting evidence for the microstructural information obtained is provided by direct observations made using high-resolution transmission electron microscopy. These images demonstrate the presence of irradiation-induced prismatic edge dislocations as well as features that indicate the presence of basal dislocations. They also provide supporting evidence for the progressive deterioration of the graphitic planes via damage mechanisms as proposed in the literature.

Proton irradiated graphite grades for a long baseline neutrino facility experiment

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140--180 MeV, to peak fluence of $\ensuremath{\sim}6.1\ifmmode\times\else\texttimes\fi{}{10}^{20}\text{ }\text{ }\mathrm{p}/{\mathrm{cm}}^{2}$ and irradiation temperatures between $120--200\text{ }\ifmmode^\circ\else\textdegree\fi{}\mathrm{C}$. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of $\ensuremath{\sim}{10}^{20}\text{ }\text{ }{\mathrm{cm}}^{\ensuremath{-}2}$ where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in microstructural graphite behavior to that under neutron irradiation and the agreement between the fluence threshold of $\ensuremath{\sim}5\ifmmode\times\else\texttimes\fi{}{10}^{20}\text{ }\text{ }{\mathrm{cm}}^{\ensuremath{-}2}$ where the graphite lattice undergoes a dramatic change. The confirmed similarity in behavior and agreement in threshold fluences for proton and neutron irradiation effects on graphite reported for the first time in this study will enable the safe utilization of the wealth of neutron irradiation data on graphite that extends to much higher fluences and different temperature regimes by the proton accelerator community searching for multi-MW graphite targets.

Infiltration of graphite by molten 2LiF–BeF2 salt

With the aim of developing neutron moderator and neutron reflector materials for fluoride-salt-cooled high-temperature reactors (FHRs), static infiltration tests were performed on graphite materials in molten 2LiF–BeF2 (FLiBe) salt, which is a potential primary coolant, at 700 °C and various pressures. The weight gain ratios of four grades (NBG-18, IG-110, NG-CT-10, and NG-CT-50) of graphite after infiltration were measured to determine their infiltration curves. The threshold pressure for FLiBe salt infiltration for the ultrafine-grained graphite (NG-CT-50) was greater than 600 kPa and much higher than those of the other three grades (medium-grained/fine-grained graphites), indicating that this graphite grade more probably resists salt infiltration in FHRs than other grades of graphite. However, if the threshold pressure is exceeded, it has the highest potential capacity for infiltrated salt over the pressures tested. The four grades were also characterized using mercury intrusion porosimetry. It was found that the infiltration curves of these two unwetting liquids were very similar. Scanning electron microscopy characterization showed that the FLiBe salt was distributed relatively uniformly in all four grades, indicating the presence of interconnected networks of open pores throughout the samples. X-ray diffraction patterns showed that infiltration test at high pressure led to an improved structural order and a decreased d-spacing in graphite.

A multi-scale three-dimensional Cellular Automata fracture model of radiolytically oxidised nuclear graphite

A multi-scale approach for fracture simulation, based on the Cellular Automata technique, has been developed and then applied to a nuclear graphite that is used in structural components of the UK Advanced Gas-cooled Reactors (AGR). High resolution X-ray computed tomographs of Gilsocarbon grade graphite, with up to 68% weight loss by radiolytic oxidation, provide quantitative descriptions of the porosity within its constitutive filler particles and their surrounding matrix. The statistical distributions for tensile strength and elastic modulus obtained from small models of the filler and matrix are introduced to a large scale model of the heterogeneous microstructure. These microstructure-derived simulations achieve a good agreement with experimental data. The computationally efficient analysis is then used to investigate the stochastic effects on mechanical properties of possible variations in the microstructure between individual small test specimens. As these may represent material of nominally the same microstructure, there is an apparent increase in the variability of strength and modulus at high weight loss.

Comparison of irradiation behaviour of HTR graphite grades

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

Micro to nanostructural observations in neutron irradiated nuclear graphites PCEA and PCIB

The neutron irradiation-induced structural changes in nuclear grade graphites PCEA and PCIB were investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), Raman spectroscopy, transmission electron microscopy (TEM), selected area electron diffraction (SAED) and electron energy loss spectroscopy (EELS). The graphite samples were irradiated at the Advanced Test Reactor at the Idaho National Laboratory. Received doses ranged from 1.5 to 6.8 displacements per atom and irradiation temperatures varied between 350 °C and 670 °C.XRD and Raman measurements provided evidence for irradiation induced crystallite fragmentation, with crystallite sizes reduced by 39–55%. Analysis of TEM images was used to quantify fringe length, tortuosity, and relative misorientation of planes, and indicated that neutron irradiation induced basal plane fragmentation and curvature. EELS was used to quantify the proportion of sp2 bonding and specimen density; a slight reduction in planar-sp2 content (due to the buckling basal planes and the introduction of non-six-membered rings) agreed with the observations from TEM.

Effects of air flow rate on the oxidation of NBG-18 and NBG-25 nuclear graphite

The effects of air flow rate (FR) (FR range: 1–10 L/min) on the oxidation of NBG-18 and NBG-25 nuclear graphite grades at temperatures between 600 and 1100 °C were studied, in reference to the standard test procedure for measuring oxidation rates of nuclear graphite in air (ASTM D 7542-09). The results showed that the FR effects on oxidation rate (OR) increase with increasing temperature with negligible FR effects at 600 °C for both materials. At high temperatures (>800 °C) there appears to be a two-stage relationship between FR and OR, which corresponds to the transition between reaction rates dominated by chemical kinetics and those dominated by diffusion. The material-specific microstructure appeared strongly influences this transition. The overall OR-FR behaviours of NBG-18 were higher than NBG-25 at 600–800 °C while negligible differences in the OR-FR behaviours between the two grades were observed at 900–1100 °C. The mercury porosimetry data showed that the higher OR-FR behaviours observed in NBG-18 may partly be attributed to the differences in the pore size distribution (open porosity and cumulative pore area) between the grades, especially for the large size pores (diameter ≫ 5 × 103 nm).