Graphite Components Research Articles

Depth resolved analysis of hydrogen in W7-X graphite components using laser-induced ablation-quadrupole mass spectrometry (LIA-QMS)

A deep understanding of the plasma-wall interaction processes in fusion devices like Wendelstein 7-X is necessary for an efficient plasma operation and a long lifetime of the plasma-facing components.In this work we present an approach employing residual gas analysis after picosecond laser-induced ablation (ps LIA-QMS) of graphite limiter tiles, exposed in the first plasma operational phase of Wendelstein 7-X, for depth-resolved and quantitative hydrogen content analysis. A series of poloidal and toroidal locations are analyzed at three of the five limiters, showing up to 2.3 × 1022 hydrogen atoms/m2 in net-deposition areas after a total plasma exposure of about 311 s in mixed hydrogen and helium operation. Shallow implantation of hydrogen is observed in erosion zones, where a low fuel content is present due to the high surface temperature during plasma operation. The hydrogen content spans between (1.1 and 3.7) × 1021 hydrogen atoms/m2 in the net-erosion areas. Moreover, oxygen has been analyzed and its appearance in both the implantation and deposition zone was verified. Results are compared to thermal desorption spectrometry and to simultaneously performed laser-induced breakdown spectroscopy (LIBS) measurements.

A Review of Finite Element Method Models for Nuclear Graphite Applications

Graphite components serve as structural support and as a neutron moderator in multiple nuclear power stations and is also a prospective material in Generation IV reactor concepts. The environment (e.g. high temperature and intense radiation) within a graphite core changes the geometry, microstructure and material properties of the graphite components, that could limit the lifetime (i.e. safe operation) of a reactor. Some of the assessments made for graphite reactor core components include Finite Element Method (FEM) simulations. These simulations are used to estimate the stress state and strain state and probability of failure of irradiated graphite components. FEM also forms part of the validation of measurements or as a complementary technique to experimental procedures in graphite research. A comprehensive overview of FEM methodologies is provided in this review, including: FEM software, basics steps to set up a FEM model, and analytical solutions used to validate simulations for this field of study. The objective of this paper is to inform researchers in the different available FEM techniques and tools used in the research of nuclear graphite.

Fracture study in notched graphite specimens subjected to mixed mode I/II loading: Application of XFEM based on the cohesive zone model

The aim of the present research is to predict the fracture load of key-hole and U-notched graphite components under mixed mode I/II loading by means of the extended finite element method (XFEM) based on the linear cohesive zone model (CZM). For this purpose, two series of fracture experiment conducted on several key-hole notched rectangular graphite plates, and also U-notched Brazilian disk graphite specimens and reported previously in the literature are presented. Then, the experimentally obtained fracture loads of the tested graphite specimens for both fracture test series are theoretically estimated by means of XFEM in combination with CZM. Since the graphite specimens have perfectly brittle behavior, the linear elastic notch fracture mechanics (LENFM) assumption is utilized. For different notch geometries, the experimental and theoretical results are plotted in the form of the variations of the fracture load versus the notch tip radius. Finally, it is shown that XFEM-CZM approach could successfully predict the experimentally obtained fracture loads of the tested specimens.

Stress analysis of the TMSR graphite component under irradiation conditions

TMSR uses nuclear graphite as a neutron moderator, a reflector, and the structural material, and utilizes molten salt as a coolant. When running normally, the graphite components are immersed in the molten salt. Thus, the nuclear graphite comes into direct contact with the molten salt, which infiltrates the open pores of the nuclear graphite. This infiltration may influence the stress analysis of the graphite component. In this study, a User Material subroutine was used to analyze the stress distribution of the graphite component, both with and without molten salt infiltration. Many influence factors were taken into consideration, such as the dose gradient, the shape of the permeation zone, and the permeation area. The results show that the dose gradient, shape, and area of the permeation zone all significantly influence the stress distribution. Furthermore, the results of the stress analysis indicate that for a regular graphite component with a square cross section, the peak maximum principal stress value occurs at the center of the cross section, and the symmetry of the maximum principal stress distributions was modified by quarter circle and half ellipse permeation zones.

Integrated plasma facing component calorimetry for measurement of shot integrated deposited energy in the NSTX-U.

The upgrade to the National Spherical Torus eXperiment (NSTX-U) [J. Menard et al., Nucl. Fusion 52, 083015 (2012)] increases the injected neutral beam power up to 12 MW and the plasma current up to Ip = 2 MA for plasma durations up to 5 s. The graphite plasma facing components have been re-designed to handle greater heat and energy fluxes than were seen in NSTX using a castellated design. We present the experimental testing and validation of a castellated graphite target, similar to the prototype tile design, instrumented with thermocouples at various depths in the castellation. During testing, incident heat flux is provided by a programmed electron beam system and surface temperatures are measured via infrared thermography directly viewing the target surface. It was found that the thermocouple response scaled linearly with the measured surface temperature rise regardless of thermocouple depth in the castellation. A sensitivity of 14.3 °C/kJ of deposited energy was found when treating individual castellations as a semi-infinite solid.

Heat Treatment of Irradiated Graphite in an Oxidizing Atmosphere

It was demonstrated that thermal treatment and decontamination methods can be used for the processing of irradiated nuclear graphite. Special features of the treatment of the irradiated graphite surface with an oxidizing agent were considered. A mathematical model of the interaction of a gaseous oxidizing agent with irradiated graphite was described with consideration for the release of stored Wigner energy. It was established that the qualitative and quantitative compositions of gaseous reaction products depend on thermal and gas-dynamic process conditions. The mathematical models and calculation algorithms proposed can be used for the process optimization of the heat treatment of irradiated graphite for the selective recovery of a number of radionuclides (mainly, 14C and 36Cl), which are concentrated in a thin near-surface layer of graphite components.

Integration of fiber Bragg grating temperature sensors in plasma facing components of the WEST tokamak.

Plasma Facing Components (PFC) temperature measurement is mandatory to ensure safe high power and long pulse tokamak operation. IR thermography systems which are widely used in magnetic fusions devices become challenged with the choice of tungsten as a PFC material in the ITER tokamak, mainly due to emissivity uncertainties and reflection issues in a hot environment. Embedded temperature measurements are foreseen to cross-check the IR thermography measurements. Fiber Bragg grating sensors are potentially of great interest for this application because they are immune to electromagnetic interference and allow the measurement of a large number of temperature spots on a single fiber. Four optical fiber temperature sensing probes, each of them including 11 regenerated fiber Bragg gratings equally spaced by 12.5 mm (equivalent to one ITER-like tungsten monoblock), have been specifically designed and manufactured for the WEST project (W-tungsten Environment and Steady State Tokamak). The four probes are embedded in W-coated graphite components at two different distances from the surface, 3.5 mm and 7 mm, to cover a wide range of temperatures up to 900 °C. This paper addresses the design and integration issues and the qualification and performance assessment performed in the laboratory. It also shows the first measurements of this new diagnostic achieved in a tokamak environment during baking of the machine and during early diverted plasma exposure.

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.

Experimental and computational correlation of fracture parameters KIc, JIc, and GIc for unimodular and bimodular graphite components

The influence of bimodularity (different stress ∼ strain behaviour in tension and compression) on fracture behaviour of graphite specimens has been studied with fracture toughness (KIc), critical J-integral (JIc) and critical strain energy release rate (GIc) as the characterizing parameter. Bimodularity index (ratio of tensile Young's modulus to compression Young's modulus) of graphite specimens has been obtained from the normalized test data of tensile and compression experimentation. Single edge notch bend (SENB) testing of pre-cracked specimens from the same lot have been carried out as per ASTM standard D7779-11 to determine the peak load and critical fracture parameters KIc, GIc and JIc using digital image correlation technology of crack opening displacements. Weibull weakest link theory has been used to evaluate the mean peak load, Weibull modulus and goodness of fit employing two parameter least square method (LIN2), biased (MLE2-B) and unbiased (MLE2-U) maximum likelihood estimator. The stress dependent elasticity problem of three-dimensional crack progression behaviour for the bimodular graphite components has been solved as an iterative finite element procedure. The crack characterizing parameters critical stress intensity factor and critical strain energy release rate have been estimated with the help of Weibull distribution plot between peak loads versus cumulative probability of failure. Experimental and Computational fracture parameters have been compared qualitatively to describe the significance of bimodularity. The bimodular influence on fracture behaviour of SENB graphite has been reflected on the experimental evaluation of GIc values only, which has been found to be different from the calculated JIc values. Numerical evaluation of bimodular 3D J-integral value is found to be close to the GIc value whereas the unimodular 3D J-value is nearer to the JIc value. The significant difference between the unimodular JIc and bimodular GIc indicates that GIc should be considered as the standard fracture parameter for bimodular brittle specimens.

Experimental study to estimate the surface wear of nuclear graphite in HTR-PM

The aim of this study was to simulate the surface wear of graphite components in a pebble-bed nuclear reactor caused by the moving fuel balls and, thus, to help evaluate the integrity of the whole graphite structures over their service life. To this end, wear tests were conducted on the chewing machine with a specially designed load jig. The test specimens included graphite plates and graphite spheres, to represent the graphite bricks and fuel spheres, respectively. The main parameters, including volume loss of graphite plates and spheres, micro-morphology of the wear surface and weight and particle size of graphite dust generated, were studied in this work.

A preliminary study on seismic behavior of the graphite reflector in molten salt reactor

Graphite reflector submerged in molten salt in TMSR has important influences on the evaluation of structural stability of the reflector and the safety of the reactor because of the collision and fluid–solid interaction. In this paper, an approach is proposed for simulating the non-linear responses of the reflector under seismic excitation. And then, with this approach, a finite element model of the reflector in TMSR is built to investigate its dynamic characteristics and the seismic time history responses. From the results of the seismic time history analysis, the displacements of the control rod channel, which are required for the evaluation the safety of the reactor, are obtained as the boundary condition for the control rod drop test. At the same time, the impact forces between the components are provided that is required for the stress evaluation of graphite components.

Exploring the scope of capacitance-assisted electrochemical carbon dioxide capture.

Optimisation of a capacitance-assisted electrochemical carbon-capture process is facilitated by the physical separation of the graphite and aluminium anode electrodes. This facilitates graphite electrode recycling and enables high current and increased aluminium surface area experiments which fix carbon at a higher rate and the same cell-voltage. Quantification of the H2 cathode byproduct shows that this process could be a net energy producer if recycled aluminium is used as the sacrificial anode.

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.

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.

Estimation of Inductively Coupled Plasma Etching Damage of Boron‐Doped Diamond Using X‐Ray Photoelectron Spectroscopy

Diamond has attracted considerable attention as an advanced‐generation material for power devices because it shows high breakdown characteristics and carrier mobility. For the development of high‐performance diamond power devices, each interfacial atomic structure must be optimized to encourage carrier conduction. Advanced etching processes without material damage could solve this issue. In this study, the inductively coupled plasma (ICP) etching, a basic device processing technique for diamond power devices is focused on. The electrical properties of diamond are deteriorated under repeated processing for mesa‐structured devices; therefore, it is estimated the ICP etching effect on diamond surfaces by angle‐resolved X‐ray photoelectron spectroscopy. At first, damage to the etched diamond surface is observed. The 2‐nm thickness of the damaged layer is reduced by decreasing the set bias voltage. However, graphite components appear with etching at lower bias voltages. Because annealing and mixed‐acid treatments effectively remove these graphite components, a three‐step method of zero‐bias‐voltage ICP etching, annealing, and mixed‐acid treatment is used for the damage reduction process. A clean surface is obtained using hydrogen plasma treatment as a finishing process.

A Low O/Si Ratio on the Surface of Mercury: Evidence for Silicon Smelting?

AbstractData from the Gamma‐Ray Spectrometer (GRS) that flew on the MErcury Surface, Space ENvironment, GEochemistry, and Ranging spacecraft indicate that the O/Si weight ratio of Mercury's surface is 1.2 ± 0.1. This value is lower than any other celestial surface that has been measured by GRS and suggests that 12–20% of the surface materials on Mercury are composed of Si‐rich, Si‐Fe alloys. The origin of the metal is best explained by a combination of space weathering and graphite‐induced smelting. The smelting process would have been facilitated by interaction of graphite with boninitic and komatiitic parental liquids. Graphite entrained at depth would have reacted with FeO components dissolved in silicate melt, resulting in the production of up to 0.4–0.9 wt % CO from the reduction of FeO to Fe0—CO production that could have facilitated explosive volcanic processes on Mercury. Once the graphite‐entrained magmas erupted, the tenuous atmosphere on Mercury prevented the buildup of CO over the lavas. The partial pressure of CO would have been sufficiently low to facilitate reaction between graphite and SiO2components in silicate melts to produce CO and metallic Si. Although exotic, Si‐rich metal as a primary smelting product is hypothesized on Mercury for three primary reasons: (1) low FeO abundances of parental magmas, (2) elevated abundances of graphite in the crust and regolith, and (3) the presence of only a tenuous atmosphere at the surface of the planet within the 3.5–4.1 Ga timespan over which the planet was resurfaced through volcanic processes.

Finite element modelling of multilayer Advanced Gas-cooled Reactor bricks and creep interaction

The structural integrity of nuclear graphite bricks is important in Advanced Gas-cooled Reactors (AGRs) as they not only provide moderation but channels for fuel, cooling and control rods. AGR graphite moderator bricks are subjected to fast neutron irradiation and radiolytic oxidation during reactor operation, leading to component dimensional and material properties changes. With irradiation ageing the graphite components deform in the axial and radial directions causing changes to bore diameter and brick length. Furthermore, deformation at the brick end faces (called dishing) can lead to the formation of axial gaps between the bricks which can then lead to fuel channel bowing. However, it is believed that irradiation creep reduces the potential size of these axial gaps and retains bricks in contact to some extent. Therefore, as the reactors age it is important to understand the nature of this contact behaviour between bricks in the fuel channels. This paper focuses primarily on the finite element modelling of contact behaviour between bricks and the effects of irradiation-induced creep on contact conditions and hence, ovality of the brick bore. Multilayer fuel brick models and a brick with a rigid body on the top surface have been modelled. The results show that multilayer models are required to understand the contact conditions between the bricks throughout the life of a reactor.

Static Multiaxial Fracture Behavior of Graphite Components: A Review of Recent Results

The problem of mixed mode (I+III) brittle fracture of polycrystalline graphite is investigated systematically here for the first time. The present study considers cylindrical specimens weakened by circumferential notches characterized by different acuities. A new complete set of experimental data is provided considering different geometrical configurations by varying the notch opening angle and the notch tip radius. The multiaxial static tests have been performed considering different values of the mode mixity ratio. A criterion based on the local Strain Energy Density previously applied by the same authors only to pure modes of loading is extended here to the case of tension and torsion loadings applied in combination.

Impact of gap size uncertainty on calculated temperature uncertainty for the advanced gas reactor experiments

Research being conducted on tristructural-isotropic fuel development and qualification involves seven advanced gas reactor (AGR) experiments that were planned to provide fuel qualification data to support the licensing and operation of the high-temperature gas-cooled reactor. Each AGR test consists of multiple independent capsules containing fuel compacts placed in one or more graphite cylinders shrouded by a stainless-steel shell. These capsules are instrumented with thermocouples embedded in the graphite holder, enabling temperature control. The desired fuel temperature is maintained by variation of the neon/helium gas mixture in response to feedback from thermocouple readings. In the absence of direct measurements, the commercial finite-element heat transfer code ABAQUS was used to predict fuel temperatures. Recognizing inherent uncertainties in the simulation model due to complex physical mechanisms, capsule geometries, and material properties, comprehensive temperature uncertainty quantification was performed. The uncertainty results reveal that the uncertainties in gap sizes are among the most influential factors contributing to calculated temperature uncertainty. The gap size uncertainties originate from a lack of direct experimental data for accurate assessment of dimensional change rates of fuel compacts and graphite components due to complex irradiation-induced material shrinkage or swelling. The study described here focuses on the impact of the gap size uncertainties based on the post-irradiation examination metrology data on calculated temperature uncertainty.

Fabrication of large-sized TaC-coated carbon crucibles for the low-cost sublimation growth of large-diameter bulk SiC crystals

Low-cost TaC-coated graphite components (SinTaC) fabricated via wet powder forming and sintering has been proposed to reduce the production cost and improve the crystal quality of SiC wafers. However, the sizes of the SinTaC components are limited by the available sizes of graphite materials with coefficients of thermal expansion (CTEs) that match that of the TaC layer, hindering their application in the production of large-diameter SiC wafers. Here we demonstrate the scaling up of the sizes of SinTaC components through the reselection of new graphite materials based on available maximum size and CTE. The large-sized SinTaC components fabricated from the optimal reselected graphite material were tested in the growth of SiC via sublimation. The results confirm the enhanced durability of the large-sized SinTaC components compared to the graphite ones. Furthermore, the SinTaC components are reusable in multiple growth cycles.