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Irradiation damage in graphite due to fast neutrons in fission and fusion systems
... The data are in good agreement with the ASTMrecognized values (replotted from fig. 8) to within the measurement error. The realworld AGR data necessarily combine multiple irradiation phenomena, including a range of fast neutron irradiation fluences and radiolytic oxidation at the operating temperatures between 356 C and 426 C. Reviews of data from material test reactor experiments are conflicting and suggest that irradiation damage may have a small or negligible effect on the specific heat capacity: provided the measurements are made at temperatures below the irradiation temperature, the specific heat increases, 45 but decreases for graphite irradiated at temperatures above 300 C. 46 In both cases the effect was small so these trends do not disagree with the AGR data. ...
... In contrast, full recovery was reported for various carbonaceous materials (including highly oriented pyrolytic graphite) irradiated up to 450 C and tested prior to 1970. 46,49 For PGA graphite, a formula describing the recovery was derived by Gray, based on the di-vacancy concentrations. 46 The ability to recover indicates that the loss in conductivity results from the formation of defect clusters (or similar features) that disrupt the lattice planes, rather than, say, grain recrystallization or impurity migration. ...
... 46,49 For PGA graphite, a formula describing the recovery was derived by Gray, based on the di-vacancy concentrations. 46 The ability to recover indicates that the loss in conductivity results from the formation of defect clusters (or similar features) that disrupt the lattice planes, rather than, say, grain recrystallization or impurity migration. This observation may help to design graphites suitable for retaining their favorable thermal conductivity. ...
Polycrystalline graphite has a unique combination of high-temperature properties that has made it the material of choice for many industrial applications. Several nuclear reactor designs that operate between 500°C and 1,000°C include graphite components. These components must maintain their integrity even at the 1,800°C they could be exposed to during an accident. The operational behavior of these graphites during both proof testing of as-manufactured material and postirradiation examination must be determined by measuring physical, mechanical, and thermal properties. For reasons of expense and practicality the properties are measured in (or near to) ambient conditions. It is essential that the measured properties may be extrapolated reliably to high temperatures. Laboratory testing at elevated temperatures therefore provides data for (1) defining temperature-dependent extrapolation curves, (2) informing conceptual models that help to establish confidence in ambient-temperature test methods, and (3) inputs into numerical simulations of operating conditions. The properties of interest for this paper are selected on the basis of current ASTM standards to include those most relevant to current and future fission reactor operation. The effects of fast neutron irradiation on the high-temperature behavior are presented in general terms, and the conventional understanding of the mechanisms behind both the inert and irradiated behavior are outlined. Areas for further research are then highlighted, the findings of which would support design, qualification, operation, and safety monitoring of graphite-moderated nuclear reactors.
... CFC material considerations for applications in nuclear reactor environments were addressed in [11,12]. Relevant to the behavior of CFCs are studies conducted and reports on nuclear grade graphite [4,[13][14][15][16][17][18][19][20][21][22][23] focusing on the microstructural and physical property evolution as a function of temperature and fluence. The characterization is complemented by the understanding of thermal shock effects on graphitic materials (including CFC), reported in refs. ...
... Visual examination of the AC-150K CFC following the low-temperature irradiations (Phase A) revealed important structural degradation of the CFCs. Figure 7 shows low resolution- before important micro and macroscopic degradation is observed [15][16][17][18][19]21]. However, this threshold has not, to the authors" best knowledge, been observed in CFCs. ...
... As is shown in Figure 10 (a), the irradiation temperature plays an important role in the radiationinduced dimensional changes. The larger deviation from the unirradiated state exhibited by the lower fluence/annealing temperature combination (which indicates annealing of interstitials leading to effective shrinkage of the macroscopic structure above the irradiation temperature) is in-line with the changes in the lattice parameter observed in irradiated graphite [15,17,18,21]. ...
Carbon fiber composites (CFC) are, among other applications, already in use as low-impedance collimating elements for intercepting TeV-level protons machines, like the CERN Large Hadron Collider. Towards a comprehensive understanding of these materials’ properties and behavior, a series of radiation damage campaigns have been undertaken in order to quantitively investigate the response of AC-150K CFC to high proton fluences. The CFCs dimensional changes, stability, micro-structural evolution, and structural integrity below and above 5 × 10²⁰ p/cm² fluence threshold were studied. The irradiating particles were protons of kinetic energies in the range of 130-200 MeV. The effects of the irradiation temperature (∼90-180°C and 280-500°C) combined with the proton fluence on the dimensional stability and the microstructural evolution of this anisotropic two-dimensional composite structure were studied using precision dilatometry and X-ray diffraction. Our results shed light on the fluence-based limitations for these composite materials. Furthermore, we compare these results with reference unirradiated graphite, and we discuss the similarities and differences in the dimensional changes and post-irradiation annealing properties.
... A large body of research has emerged over the past seven decades regarding graphite and the effects of thermal and fast neutrons on its micro-and macrostructure as well as its thermophysical properties [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16][17][18][19]. Changes in physical and mechanical properties, swelling, and annealing as well as evolution of the graphite lattice structure were addressed. ...
... Studies on the lattice parameter evolution under neutron and/or electron irradiation were reported in Refs. [7][8][9][10][11][12][13][14][15][16]. Nuclear graphite irradiated at high temperatures was studied in Ref. [17]. ...
... Shown in Fig. 5(b) [15] are macroscopic dimensional changes in isotropic graphite for several irradiation temperatures. Similar to what has been established by several studies [1][2][3][6][7][8][13][14][15][16] at the lattice level as a result of the stored energy release that essentially begins at ∼200°C, the two distinct temperature regimes (below 200°C and above 200°C) are also observed for the macroscopic behavior of isotropic graphite [ Fig. 5(b)]. At irradiation temperatures below 200°C, the isotropic graphite tends to swell rather than shrink, a process typically observed in other nonisotropic grades prompted by the filling, initially, of the pores and Mrozowski cracks. ...
The NT-02 neutrino physics target made of the isotropic graphite grade produced neutrinos for the MINOS and MINERVA high-energy physics experiments. The segmented, 95-cm-long NT-02 target was bombarded with a 340 kW, Gaussian 1.1 mm sigma beam of 120 GeV protons reaching 6.516×1020 protons on target and a peak fluence of 8.6×1021 protons/cm2. Reductions in detected neutrino events during the experiment were attributed to radiation-induced damage on the target material leading to the NT-02 target replacement. With future neutrino physics targets aiming at the multimegawatt power regime, identifying life expectancy or fluence thresholds of target materials is of paramount importance, and, therefore, pinpointing the exact cause and target failure mode triggering the neutrino yield reduction is critical. To help unravel the effects of the 120 GeV beam on the isotropic graphite structure at the microstructural or lattice level, x-ray beams from National Synchrotron Light Source II were utilized to study failed in-beam as well as intact NT-02 target segments. The primary objective was to arrive at a scientifically sound explanation of the processes responsible for the target failure by correlating macroscopic observations with microstructural analyses. Results from transmission electron microscopy studies were integrated in assessing the microstructural evolution. The x-ray diffraction study revealed (a) the diffused state reached by the graphite microstructure within the 1σ of the beam where the graphite lattice structure transforms into a nanocrystalline structure, a finding supported by electron microscopy examination, thus providing an indication of the fluence threshold, and (b) the dominant role of the irradiation temperature profile exhibiting a high gradient from the beam center to the heat sink and aggravating the damage induced in the microstructure by the high proton fluence. The effects of the 120 GeV protons on the isotropic graphite target structure are corroborated by observed damage induced by 160-MeV protons and by fast neutrons to comparative doses on similar graphite, an assessment that will aid the design of next-generation megawatt-class neutrino targets. © 2019 authors. Published by the American Physical Society. Published by the American Physical Society under the terms of the »https://creativecommons.org/licenses/by/4.0/» Creative Commons Attribution 4.0 International license. Further distribution of this work must maintain attribution to the author(s) and the published article's title, journal citation, and DOI.
... Some of these defects can be removed by annealing [3], such as the infamous Wigner-energy release [4], while others persist. Over time, the accumulating defects significantly alter the microstructure, reducing the degree of crystalline order [5,6] and creating cross-links between layers. These changes affect physical properties such as the thermal conductivity [7] and Young's modulus [8]. ...
... Some of this variation can be attributed to orientation effects, with the most extreme value being 780 eV reported by Zobelli et al. [12] for displacement along the bond axis. Notably, the value of 60 eV for E d used by the UK nuclear industry is at the upper end, a value which is "certainly too high," according to an International Atomic Energy Agency technical document [6]. The same document comments on large reported variations of E d with crystallographic directions, a situation it describes as "not satisfactory." ...
... Reading from left to right and then line by line, the references for all data refer to Refs. [6,11,[14][15][16][17][18][19][20][21][22][23][24][25][26][27][28][29][30][31]. ...
Molecular-dynamics simulations are used to compute the threshold displacement energy (Ed) in a series of progressively damaged graphite structures. The Ed values are obtained by a statistically robust probabilistic method using a large number of primary-knock-on-atom events at energies up to 100 eV. No sharp threshold for Ed is observed, and a number of possible definitions are considered. For pristine graphite, the best estimate of Ed is 24 eV. Ed decreases with increasing irradiation damage, dropping by nearly a factor of 2 at a dose of one displacement per atom. For a fully disordered amorphous-carbon structure, Ed is around 5 eV. This evolution of Ed is an important missing ingredient in current estimates of radiation doses in nuclear reactors, which assume Ed is constant over the reactor lifetime, despite substantial structural evolution.
... graphite, C=C composites) as targets for long beam exposure. The works of Professor Kelly on irradiation damage in graphite are presented in [3,[10][11][12][13] where dimensional changes in graphite, including thermal expansion coefficient and fractional changes in Young's modulus as a function of neutron dose, are summarized. Graphite crystal growth in the hexagonal axis and basal plane as a function of neutron dose and irradiation temperature are listed. ...
... All four grades exhibit an increase of CTE due to proton irradiation. Figure 10(b) (after Marsden [11]) shows change of CTE induced by neutron irradiation at different temperatures, including temperatures as low as 200°C. Shown in Fig. 10(b) is the increase of CTE occurring in irradiated pile graphite for the lower temperatures exposed to neutron fluences Figure 11 shows the influence of proton fluence on the CTE of three of the four graphite grades studied (POCO, IG 430 and SGL). ...
... Specifically, based on results presented in [3] the strength and Young's modulus of graphite irradiated with neutrons increase up until ∼10 20 n=cm 2 for irradiation performed at room temperature and they start decreasing at higher fluences with changes being smaller at higher temperatures. Also reported in [11] for Pile Grade A and Gilsocarbon graphite grades under neutron irradiation is the presence of two distinct regions about the 300°C irradiation temperature. Below 300°C the Young's modulus increases by a factor of ∼3, peaks and then decreases, followed by an increase and later a catastrophic decrease and disintegration. ...
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 ∼6.1×1020 p/cm2 and irradiation temperatures between 120–200 °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 ∼1020 cm−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 ∼5×1020 cm−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.
... During operation, graphite is exposed to neutron radiation and temperature gradients that could cause irradiation and radiolytic oxidation. These occurrences will affect material properties such as weight loss, porosity changes and thermal conductivity which can potentially lead to material and structural failure [5] . ...
... Future reactors such as VHTR are expected to operate at higher temperatures than the current reactors and its core outlet temperature (COT) will be about 10 0 0 °C [5][6][7][8] . In order to enhance thermal efficiency of future nuclear reactors, extensive research has been carried out, aiming to provide an improved design for future reactors [6][7][8][9][10] . ...
... Finally, K RD depends on the neutron irradiation, as the latter plays a significant role in producing various types of defects, which are more effective in scattering phonons. Therefore, many studies [5,13,15] emphasise that the change in conductivity is only caused by the phonon scattering in lattice defects. This phonon scattering increases with temperature. ...
The graphite bricks of the UK carbon dioxide gas cooled nuclear reactors are subjected to neutron irradiation and radiolytic oxidation during operation which will affect thermal and mechanical material properties and may lead to structural failure. In this paper, an empirical equation is obtained and used to represent the reduction in the thermal conductivity as a result of temperature and neutron dose. A 2D finite element thermal analysis was carried out using Abaqus to obtain temperature distribution across the graphite brick. Although thermal conductivity could be reduced by up to 75% under certain conditions of dose and temperature, analysis has shown that it has no significant effect on the temperature distribution. It was found that the temperature distribution within the graphite brick is non-radial, different from the steady state temperature distribution used in the previous studies To investigate the significance of this non-radial temperature distribution on the failure of graphite bricks, a subsequent mechanical analysis was also carried out with the nodal temperature information obtained from the thermal analysis. To predict the formation of cracks within the brick and the subsequent propagation, a linear traction-separation cohesive model in conjunction with the extended finite element method (XFEM) is used. Compared to the analysis with steady state radial temperature distribution, the crack initiation time for the model with non-radial temperature distribution is delayed by almost one year in service, and the maximum crack length is also shorter by around 20%.
... This agrees well with the work of Zobelli [24] who reported a value of 22.2 eV. Using density functional theory, Kelly et al. calculated a value of 23 eV perpendicular to the layer [25]. It is interesting here to note the extreme anisotropy: 23 eV perpendicular to the layer, compared with 780 eV in the direction of a C-C bond in the layer [26]. ...
... The IAEA TECDOC 1154 [25] appears to endorse an effective value of 40 eV for isotropic fluxes previously proposed by Thrower and Mayer [27], and remarks that the value of E d used in the UK of 60 eV is 'certainly too high'. Nevertheless, in that same review [27], there is reference to Reynolds' argument [28] that a practical way to account for annealing effects during irradiation and for overlapping of collision volumes at doses above 10 18 n cm −2 (E > 1 MeV) is to employ the higher value of 60 eV. ...
... Before describing the cascades, we first observe that the EDIP threshold for graphene based on a single (001) trajectory is 20.9 eV, which is encouragingly close to the value of 22−23 eV given by ab initio calculations [23][24][25] and 23.6 eV from experiment [21,22]. ...
The environment dependent interatomic potential (EDIP) including Ziegler–Biersack–Littermark (ZBL) interactions for close encounters is applied to cascades starting from a host atom and from an interstitial atom. We find the room temperature displacement threshold to be 25 eV, increasing to 30 eV at 900 K. The latter correlates well with the measured threshold for vacancy production. Additionally, divacancy production is found to occur, including interlayer divacancies from around 60 eV. The data suggest a new, continuous damage function applies, where the threshold region depends on the square root of the primary knock-on atom (PKA) energy in excess of the threshold, evolving to a linear dependence on PKA energy.
... Below 400 o C, graphite oxidation is negligible. The prevailing oxidation modes are categorized based on temperature as follows: [23][24][25][26] Mode ( ...
... Such increase in temperature increases the reaction rate constants, and the mobility and diffusion of the oxygen molecules and other oxidants from the bulk gas to the graphite surface. The heterogeneous corrosion reactions are: 23,34 ) ...
... For example, at 800 o C and 0.1 atm, the relative rates for the C-O 2 (5b), C-H 2 O (5f), C-CO 2 (5d) and C-H 2 (5g) reactions are 10 5 , 3, 1 and 3 x 10 -3 , respectively. 24 Homogeneous reactions of gaseous species are: 23 Reactions (6a) and (6b) occur in the boundary layer and the bulk multi-species gas mixture. The first reaction competes for oxygen, and the second competes for water vapor in the bulk gas. ...
A massive air or steam ingress in High Temperature Reactors (HTRs) nominally operating at 600-950 o C is a design-basis accident requiring the development and validation of graphite oxidation and erosion models to examine the impact on the potential fission products release and the integrity of the graphite core and reflector blocks. Nuclear graphite is of many types with similarities but also differences in the microstructure, volume porosity, impurities, type and size of filler coke particles, graphitization and heat treatment temperatures, and the thermal and physical properties. These as well as the temperature, types and partial pressures of oxidants affects the prevailing oxidation mode and kinetics of the oxidation processes of graphite in HTRs. This paper reviews the graphite crystalline structure, the fabrication procedures, characteristics, chemical kinetics and modes of oxidation of nuclear graphite for future model developments.
... Reaktionopeus on pieni alle 350 • C lämpötilassa, ja vasta noin 400 • C lämpötilassa reaktio tulee merkittäväksi. [47] Hiilidioksidin kanssa graitti reagoi reaktiolla ...
... Reaktionopeus on hyvin pieni vielä 625 • C lämpötilassakin, eikä reaktiosta ole merkittävää haittaa vielä 675 • C lämpötilassakaan. [47] Vesihöyry reagoi graitin kanssa reaktioilla ...
... Pyrolyyttinen hiili höyrystetään suoraan halutulle pinnalle [6]. TRISO-partikkeleissa kiderakenne tulisi olla mahdollisimman isotrooppinen [47]. Pyrolyyttinen hiili estää ohuinakin kerroksina kaasun läpäisemisen [6]. ...
... Some general sources of information on the basic defect physics and processes involved in irradiation of graphite can be found in the key books and articles by Simmons [8], Reynolds [9,10], Gittus [11] and Kelly [1,12,13], and Thrower and Mayer [14], the last of these reviewing many observations and theories on point defects up to 1978. Also, Engle and Eatherly [15] have reviewed the irradiation behaviour of graphite at high temperatures. ...
... While the behaviour of aggregates under irradiation exhibits some notable differences in behaviour to single crystals, the variation at the level of our interest is not believed to be significant. Cases are highlighted where differences arise but a fuller discussion of specific property changes in reactor-grade graphites can be found in [1] (and in references within) and [12,13,15,[23][24][25]. ...
... We must draw a distinction between cluster defects produced by accretion and fed by progressive thermal migration and those produced during the collision cascade. Several authors [1,13,55] refer to the existence of a majority population of interstitials in a bound, immobile configuration of 4 AE 2 atoms which have no appreciable affinity for mobile interstitials and are thus stable with respect to further aggregation. These were originally identified from neutronscattering studies on irradiated graphite [56,57] and have been proposed as the main contributor to c-axis dimensional change (see section 5.3). ...
This article discusses the nature of radiation defects in graphite, reviewing past and recent developments in understanding their structure, interactions and effect on physical properties. The principal focus is on behaviour at the atomic and microstructural level, with an interest both in understanding graphite moderator damage in nuclear reactors and building a foundation for the range of emerging technological applications of defect-engineered graphitic materials. It is hoped that this article will both clarify the picture that has emerged over the last 50 years and provide a useful background to ongoing efforts.
... Xiaowei, et al., explained that Blanchard reported that the oxidation of artifi graphite occurs by gas diffusing through pores in the temperature range of 600-900 [38,39]. Cured graphite blocks have low porosity and relatively fewer large pores, lead to excellent oxidation resistance. ...
... After the oxidation of HR2-G and C-HR2-G, the bulk density decreased by 1.30 Xiaowei, et al., explained that Blanchard reported that the oxidation of artificial graphite occurs by gas diffusing through pores in the temperature range of 600-900 • C [38,39]. Cured graphite blocks have low porosity and relatively fewer large pores, leading to excellent oxidation resistance. ...
The purpose of this study is to improve the oxidation resistance of graphite blocks after graphitization at 2800 °C by introducing a curing process of phenolic resin, used as a binder to control the pore size. Using the methylene index obtained from FTIR, the curing temperature was set to 150 °C, the temperature at which cross-linking most highly occurs. Graphite blocks that had undergone curing, and were carbonized with a slow heating rate, showed increased mechanical and electrical properties. Microstructural observation confirmed that the curing process inhibited the formation of large pores in the graphite block. Therefore, the cured graphite block showed better oxidation resistance in air than a non-cured graphite block. Oxidation of the graphite block was caused by pores created by pyrolysis of the phenolic resin binder, which acted as active sites.
... 化已经有了比较全面的研究结果 [4,5] . 在辐照损伤 程度以及石墨抗辐照性能的评价方面, Zhai等 [6] 和Zeng等 [7] 系统研究了石墨的拉曼光谱受重离子 辐照的影响, 付晓刚等 [8] 提出石墨气孔形貌变化可 作为石墨辐照性能的评价方法, Burchell等 [9] 提出 电阻率可作为研究石墨在辐照和退火后的缺陷演 化的指标. 尽管辐照损伤的机理基本确认为辐照产 生的点缺陷(间隙原子和空位), 结构性能的变化源 自点缺陷的移动、聚合、复合等演化行为 [10] , 对石 墨中缺陷的模型也有一些讨论 [11] , 但对原子尺度 上石墨缺陷演化的研究还不充分. ...
... 物, 在更高的温度下才能被去除 [9,33] . ...
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+ implantation treatments at room temperature with fluences of 1×10^15, 5×10^15, 1×10^16 and 1×10^17 cm-2 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+ 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 (ID/IG) 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 S-parameter and reduces the average W-parameter, while oxidation reduces the average S-parameter and increases the average W-parameter. The annealing effect at 850℃ cannot completely recover the point defects in Irr.-Ox. samples.
... Graphite oxidation is known to consist of several kinetic regimes, sometimes also referred to as 'oxidation modes', beginning at temperatures > 400˚C. There is a wealth of literature in this area [35][36][37] and therefore only a short explanation will be recapped for scientific completeness. The modes are classified as follows: ...
... The observed dramatic jump in oxidation rate implies that the transition between modes B and C is particularly sudden and temperature sensitive. Interestingly, the increase in rate between 600-700˚C is relatively small and may indicate an extension of the chemical controlled regime to temperatures closer to 700˚C, rather than the widely reported 600˚C [32,35,40] suggesting that, with these specific conditions, the temperature range for mode B is quite narrow, approximately between 700-800˚C. These differences in observed transitions are a result of many influencing factors such as microstructure, sample geometry, density and impurity content, which can vary widely between researchers, experimental set-up and graphite grade [41]. ...
This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400–1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700–800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000–1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.
... Most irradiated graphite oxidation data comes from studies in the United Kingdom on the Magnox and Advanced Graphite Reactor (AGR) designs 52,53 . These designs use a CO 2 coolant that dissociates under irradiation, creating constant low levels of oxygen within the graphite microstructure at normal reactor core temperatures. ...
... A characteristic release peak at around 200°C was found for some graphite grades irradiated at a low temperature. This peak shifts to higher temperatures and diminishes in magnitude (to almost undetectable levels), as the irradiation temperature increases 52 . Graphite irradiated at temperatures at or above 250°C should not accumulate significant Wigner energy stored in the crystal structure. ...
The passively safe modular High Temperature Gas-cooled Reactor (mHTGR) design is one of the
primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. While the nuclear grade graphite core provides neutron moderation, high temperature stability, thermal conductivity, and passively safe reactivity control for these high temperature reactors, graphite is a carbonaceous material and this has generated a persistent concern that the graphite core components could actually burn during accident
conditions. This report directly addresses the issue of uncontrolled, self-sustained oxidation (burning) of graphite and demonstrates that burning is not possible for high purity, nuclear graphite.
... The chemical form of surface 14 C in irradiated graphite is the subject of a separate study and future publication. Gasification of graphite occurs via a 3-step mechanism: (1) the primary adsorption of oxygen onto the graphite surface, (2) the surface reaction between adsorbed oxygen and active carbon sites, and (3) desorption of the newly formed carbon-oxygen species [8] . Oxygen species adsorption and interaction with carbon active sites occurs spontaneously . ...
... Nonetheless, at a magnification of 250x, there is as notable, qualitative difference between the morphologies before (Figure 2, Left) and after irradiation (Fig. 2, Right). This difference is attributed to radiation damage that causes fragmentation of the surface and is consistent with previous findings on irradiation damage to graphite [8]. The small number and size of pores in unirradiated NBG-18 graphite (Figure 3 , Left) are a result of the manufacturing process. ...
Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the , which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide (). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].
... The idea of such scaffolding structures has been previously demonstrated by da Silva et al. 23 for carbon nanotube bundles (CNBs) where covalent bonds construct strong cross-links between individual nanotubes, leading to a sizable increase in the shear modulus of CNBs. It has been shown 23,24,25,26,27,28 that the covalent cross-links in carbon based materials can be introduced via high energy defects produced by irradiation 29,30,31 . ...
In the present study we investigate the irradiation-defects hybridized graphene scaffold as one potential building material for the anode of Li-ion batteries. Designating the Wigner V22 defect as a representative, we illustrate the interplay of Li atoms with the irradiation-defects in graphene scaffolds. We examine the adsorption energetics and diffusion kinetics of Li in the vicinity of a Wigner V22 defect using density functional theory calculations. The equilibrium Li adsorption sites at the defect are identified and shown to be energetically preferable to the adsorption sites on pristine (bilayer) graphene. Meanwhile the minimum energy paths and corresponding energy barriers for Li migration at the defect are determined and computed. We find that while the defect is shown to exhibit certain trapping effects on Li motions on the graphene surface, it appears to facilitate the interlayer Li diffusion and enhance the charge capacity within its vicinity because of the reduced interlayer spacing and characteristic symmetry associated with the defect. Our results provide critical assessment for the application of irradiated graphene scaffolds in Li-ion batteries.
... Under irradiation nuclear graphite structure expands in the c-direction and contracts in the direction parallel to the basal planes, which is often called the 'swelling' of nuclear graphite [5,8,15]. In earlier reports, this change in dimension has been attributed to displacement of carbon atoms upon neutron irradiation [16][17][18]. ...
The combined effect of high-temperature and heavy-ion irradiation on Mrozowski cracks (MC) and nuclear graphite crystallographic dimensions have been studied using in situ heating and in situ ion-irradiation in the transmission electron microscope (TEM). Electron transparent lamella of nuclear graphite, IG-110, was irradiated, using a 2.8 MeV Au beam at an ion flux of 3.991 x 1010 ion cm ⁻² s ⁻¹ for 70 minutes at 800 o C. Upon high-temperature irradiation, Mrozowski crack closure was studied quantitatively. The analysis showed linear, positive expansion of nuclear graphite which is significantly different from the dimensional changes previously reported for low-dose neutron irradiation of nuclear graphite in which the material undergoes negative to positive expansion via a turnaround radiation dose. The trend of the thermal expansion coefficient (CTE) of pristine IG-110 in this study is consistent with previous reports in the 100-800 o C temperature region in which the dimensional change ranges from negative to positive values.
... While the correlation of the damage induced in graphite, IG-430 or any other grade, from protons or fast neutrons may not yet be fully understood, the experimental data to-date [12][13][14] provide evidence of damage similarities on the basis of fluence. Consequently, the large body of work on irradiated graphite generated to-date addressing the physio-mechanical changes on several graphites may come into play in evaluating the IG-430 grade The behavior of graphite under neutron irradiation has been explored over the past several decades in [15][16][17][18][19][20][21][22][23][24][25] In these studies, emphasis has been given to neutron-induced amorphization [18] , dimensional changes [19 , 20 , 22] both macroscopic and lattice, swelling [21] , Young's modulus and strength evolution [16 , 20 , 22] , and thermal conductivity [23] . In [18] experimental evidence is reported that irradiation at 150 °C can produce large crystal dimensional changes in relatively small doses. ...
... As seen in Fig. 12b where the (0 0 2) reflections are compared, the effect is similar to what has been observed in graphite (i.e. intensity reduction due to irradiation accompanied with broadening and shifting [19,18,22,23,24]) but at a much smaller degree. Particularly the shifting and broading of the (0 0 2) reflection, indicative of amorphisization and breaking into smaller crystallites ocurring in the graphite structure, is very limited implying resistance to radiation damage. ...
Hexagonal boron nitride was irradiated with 140 MeV protons to a fluence of ~6 10²⁰p/cm² at Tirr ~200 °C. Isotropic graphite was also irradiated alongside h-BN under similar conditions and fluence to enable direct comparison of the two similar structures. The effects of proton irradiation on dimensional stability and microstructure were studied using precision dilatometry and energy dispersive Xray diffraction techniques revealing by direct comparison to graphite that h-BN can better resist radiation damage from bombardment when irradiated with protons impinging normal to the crystallographic planes or along the crystallographic c-axis. X-ray diffraction experiments also revealed a preferred orientation of the crystallites in bulk samples near the sample surface, an orientation influenced by irradiation. Thermal studies using differential scanning calorimetry and thermogravimetric analysis to 740 °C, augmented by precision dilatometry, provided evidence of subtle phase transitions attributed to residual w-BN in the matrix. Irradiation appears to induce shifting of such transitions.
... These impurities can be introduced into the cooling gas from the cooling gas filling, which maintains the operating pressure, in the form of contamination by maintenance or by other chemical processes [9]. The following are some possible reactions that can occur in the reactor core at high temperatures (with a maximum fuel temperature in the core of 900 • C) [21]. ...
Previous studies on the safety of gas-cooled high-temperature reactors (HTR) have analyzed the corrosion and oxidation behavior of the primary circuit components under normal and accident conditions. Through the use of graphite components, graphite particles can be formed by mechanical and chemical means whose influence on the structural change of metal surfaces must be analyzed in a comprehensive manner. The dust resuspension and deposition in tank geometry (DRESDEN-TANK) test facility was set up to thermally anneal metallic samples (Alloy 800H, Inconel 617) loaded with graphite particles under typical HTR conditions (helium, 750 °C, 6 MPa) for the investigation of interactions over a long-term range. In addition to the carrying out of a description of the processes occurring on the material surface, the gaseous reaction products have been analyzed. The results show that the presence of graphite particles in the near-surface layer has a significant impact on corrosion processes due to thermally-induced interactions. In this case iron and chromium are degraded in the metallic alloys, which leads to a structural change in the near-surface layer. Furthermore, the graphite particles significantly influence the formation of the oxide layers on the alloys; for example, they influence the formation speed of the layer and the layer height. The originally deposited particles thus exhibit a chemically-altered composition and a different geometric shape.
... In case of higher temperatures, the external mass transfer of gas species (O 2 , CO, CO 2 ) into the surface becomes limited and reaction remains restricted to the outer surface only (Chi and Kim, 2008;Contescu et al., 2008;Hu et al., 2014). Blanchard reported the range of in-pore diffusion-controlled regime as 600-900°C (Kelly, 2000). The transition temperature gets affected by density, impurities and the microstructure of graphite (Hinssen et al., 2008). ...
Graphite is used as a structural material, neutron moderator and reflector in some designs of nuclear reactors. The core of a prismatic or pebble bed type High Temperature Reactor (HTR) is also comprised of graphite components. In addition to HTRs, it is also used as a structural material in the Molten Salt Breeder Reactor (MSBR). It’s response to postulated accidental condition is a critical aspect of probabilistic safety assessment studies for such futuristic reactor designs. Generation of graphite aerosol particles has been shown as a possible hazard in the case of an air ingress accident for HTRs. These aerosol particles act as a carrier for radioactive source term and their characteristics decide the impact of accident. This study aims to measure and interpret number and mass characteristics of graphite particles generated by exposing graphite to varying high temperature conditions. An exclusive sophisticated facility is developed, tuned and employed for measuring the aerosol characteristics at extreme temperature and concentration conditions. Apart from measuring number and mass size distributions of generated particles at different temperatures and flows, emission of CO and CO2 was also monitored. Particle concentration and CO gas concentration was found to be maximum at 700 °C, the transition temperature for the generation process. Evaporation-condensation and transport of unburnt carbon particles could be linked to the observed particles in nucleation and accumulation mode size ranges, respectively. Mechanism of particle generation and graphite particle transformation at different temperatures is also discussed.
... It takes several forms, and the differences in their microscopic and mesoscopic structures cause variations in their physical properties. For example, graphite that has been radiation damaged by high-energy neutrons has attracted considerable interest over the years (Kelly et al., 2000). Various models have been proposed for the radiation-induced defects (Telling & Heggie, 2007). ...
In the small-angle scattering from as-prepared and neutron-irradiated highly oriented pyrolytic graphite samples, a new type of streak pattern is observed near the -type double Bragg scattering. This newly observed scattering is assigned as the double scattering whose first scattering is diffuse scattering near the 00l Bragg scattering and whose second scattering is the Bragg scattering. Specular reflection from the surfaces of the crystallites or microcracks is also observed as a streak. These two types of scattering from the neutron-irradiated sample show oscillatory behavior, unlike that from the non-irradiated samples. The scattering is analyzed using the sharp-boundary model, and is explained by the assumptions that the non-irradiated sample has interfaces with a width of about 1 nm and the neutron-irradiated sample has a slab-like structure with a thickness of about 14 nm.
... There are studies that reported the oxidation behaviour of carbon material in HTGR, and some simulation models for the oxidation processes have been established [14][15][16]. Almost all of these studies have shown that the porosity and pore structure of the carbon materials have an important influence on the oxidation processes. The report of International Union of Pure and Applied Chemistry (IUPAC) about the gas absorption and desorption processes in porous materials has pointed out that the pore size and pore structure are the key factors in determining the shape of gas adsorption isotherm [17]. ...
Many tons of porous carbon materials (including BC and IG-110) are contained in HTGR, which are serving as structural material and fuel matrix material. These materials would absorb moisture and other impurities when exposed to the environment, and these impurities (especially moisture) absorbed in the carbon material must be removed before the reactor operation to prevent corrosion reaction at high temperature (more than 500°C). As the pore microscopic structure characteristic is the significant factor affecting the gas adsorption and flow in the porous materials, the detailed 3D pore structures of the carbon materials (BC and IG-110) in HTGR were studied by Micro-XCT and HPMI methods in this paper. These pore structure characteristics include pore geometry, pore size distribution, and pore throat connectivity. The test results show that the pore size distribution of BC material is wide, and the pore diameter is obviously larger than that of IG-110. Pore connections in BC show radial shape connections at some special points, and the pore connectivity in IG-110 is very complex and presents a huge complex 3D pore network.
... These gaseous impurities can be derived from refueling, maintenance and chemical reactions (Kissane, 2009). The following presents the possible reactions that can happen in the reactor core at high temperatures (Kelly et al., 2000). ...
Carbonaceous dust can bring some potential threats to the safe operation of the high temperature gas-cooled reactor. Predictive models based on physical generation mechanism have produced errors in regards of actual observations. Thus, the chemical generation mechanism is worth investigating. This paper mainly analyzes the formation of carbonaceous dust by chemical vapor deposition with carbon monoxide and hydrogen mixture as feed gas. The adopted substrate is made of Inconel 617 alloy. The effect of hydrogen content and reaction temperature on the generation of carbonaceous dust have been investigated. The morphology and micro-structure of the deposited carbonaceous dust have been characterized by scanning electron microscope, energy dispersive spectrometer, and Raman spectrum. The experimental results show that the production of the car-bonaceous dust is almost in parabolic relationship with H 2 flow rate, while for the influence of reaction temperature , the result is much closer to exponential relationship. In addition, two kinds of carbonaceous dust particles, larger agglomerate particles and small isolated particles, have been observed.
... When the temperatures are higher than 1 173 K, the oxidation belongs to the regime III ( boundary layer regime) . Meanwhile, Blanchard et al. [15] gave another classification in which the range of in鄄pore diffusion controlled regime was 873鄄 1 173 K. Taking the difference of components be鄄 tween the MG and polycrystalline graphite into con鄄 sideration, temperatures such as 798, 823, 873, 923 and 973 K were chosen to study the oxidation behav鄄 ior of MG. ...
The effects of temperature on the oxidation behavior of the A3-3 matrix graphite (MG) in the temperature range 798-973 K in air with a flow rate of 100 ml/min to burn-offs of 10-15 wt%, were investigated by a home-made thermo-gravimetric experimental setup. The oxidation rate (OR) increases significantly with the temperature. The OR at 973 K is over 70 times faster than at 798 K. The oxidation kinetics of A3-3 MG in air at temperatures up to 973 K is in the reaction control regime, where the activation energy is 176 kJ/mol and the Arrhenius equation could be described as: OR=2.9673×108·exp(-21124.8/T) wt%/min. The relatively lower activation energy of MG than that of structural nuclear graphite indicates that MG is more easily oxidized.
... Graphite has been used as a moderator and a reactor core structural material since the beginning of the development of nuclear reactors [1]. It is being extensively used in gas-cooled reactors and is currently considered as a potential moderator/structuralmaterial for the next generation very high temperature reactors (VHTRs). ...
We have examined irradiation induced creep of graphite in the framework of transition state rate theory. Experimental data for two grades of nuclear graphite (H-337 and AGOT) have been analyzed to determine the stress exponent (n) and activation energy (Q) for plastic flow under irradiation. We show that the mean activation energy (Q) lies between 0.15–0.32 eV with a mean stress-exponent of 1.0±0.2. A stress exponent of unity and the unusually low activation energies strongly indicate a diffusive defect transport mechanism for neutron doses in the range of 3 to 4×1022 n/cm2.
... The two-stage model of fracture, despite its simplicity and clarity, does not apply as a rule to complex heterogeneous systems under the combined action of the irradiation, high temperature and mechanical stress. In this situation it is requires a totally different approach, and this approach has been developed in a lot of studies, primarily by Kelly B.T. [59], as well as in a number of papers by Panyukov S.V. et al.6061. Bв tСО Дθ0Ж «....РrapСТtО Тs ЯТОаОН as tСО poХвМrвstaХ consisting from the ordered crystallite, so one can describe the radiation-ТnНuМОН ОППОМts» (ПТР.1.14). ...
This review is devoted to the application of the graphite and graphite composites in the science and technology. The structure and an electrical properties, the technological aspects of producing the high-strength artificial graphite and dynamics of its destruction are considered. These type of graphite are traditionally used in the nuclear industry, so author concentrates on the actual problems of the application and testing of the graphite materials in the modern science and technology.
Translated from chapters 1 of the the monograph (by Zhmurikov E.I., Bubnenkov I.A., Pokrovsky A.S. et al. Graphite in Science and Nuclear Technique// eprint arXiv:1307.1869, 07/2013 (BC 2013arXiv1307.1869Z).
... Graphite is of concern for some reactor types like the British AGR. Therfore graphite has been frequently investigated and the mechanisms of the damage which graphite undergoes on neutron irradiation are quite well understood [62]. However, many processes have not been correlated with the properties of the pristine graphites. ...
Irradiation damage is one of the most important damage mechanisms for nuclear materials. Neutrons transfer their energy to atoms which start to jump creating vacancies and interstitials being responsible for formation of defect clusters or microstructural changes (segregations, phase reactions). Nuclear reactions or transmutation can create alpha particle emitters which leads to helium gas which has to be accomodated by the material. All these effects can significantly deteriorate materials properties and limit the life-time of components. In the first part of this chapter an introduction into the most important radiation damage effects will be given. In the second part the consequences of irradiation damage (hardening, embrittlement, segregation, swelling, radiation creep) of components for current and future nuclear plants will be discussed.
... In the standard model, variation in the lattice dimension in irradiated graphite arises from the strain induced by point defects [20,21]. Some authors believe interstitials located between graphene layers increase the lattice parameter and promote the swelling along the c-axis [17] while others suggest the clusters of n carbon atoms where n ¼ 4 AE 2 perform this role [18]. The overwhelming majority of c-axis lattice parameter changes with irradiation disappears at about 300°C. ...
Molecular dynamics simulations using empirical potentials reveal HOPG graphite’s response to irradiations. Two different methodologies: displacement cascades and Frenkel pair accumulations, probe the primary damage and dose effect, respectively. This work reveals that in HOPG graphite primary knock-on atoms with initial energies less than 40 keV do not induce amorphisation by direct impact. Rather, defects stabilise and persist after a single irradiation event. However, amorphisation occurs via the accumulation of defects mimicking multiple events. Before amorphisation the graphite structure undergoes three stages of evolution characterised by (i) an increase in point defects; (ii) a wrinkling of graphene layers pinned by small amorphous pockets; and (iii) a full amorphisation of the structure via percolation of the small amorphous pockets. This structural evolution gives way to an irradiation induced volume change of the HOPG graphite. In the first stage, interstitials contribute, as expected, to c-axis swelling, while vacancies contribute to basal plane shrinkage. Subsequently, rippling of the graphene layers induces the overall volume to change. A power law relation illustrates the relation between the c-axis swelling and the basal-plane shrinkage as a function of the irradiation dose.
... Pile Grade A (PGA) graphite is used as a moderator in Magnox reactors. Bricks are manufactured by extrusion of a mixture of needle coke filler particles, in the size range 0.1 to 1.0mm, with coke flour and coal pitch binder, as described previously [1,2,3,4]. The extrusion process causes the needle-like coke particles to align with the extrusion axis, thus producing a material with orthotropic properties. ...
This study aims to develop computer models, with a microstructure representative of the PGA graphite, to contribute to the understanding of the relationship between the amount of porosity, the load-displacement behaviour and crack propagation. The project is in two linked parts, the first provides a model of the porous graphite which is then introduced into a lattice type finite element model to provide the load-displacement and crack propagation predictions. Microstructures consisting of matrix and pores with added aligned filler particles, typical of needle coke, were studied. The purpose was to isolate the effect of filler particles on fracture strength and the fracture path. In the paper crack paths and fracture mechanisms are discussed for different amounts of porosity and various filler particle arrangements.
... During graphite manufacture, which involves several high temperature heating cycles (Marsden and Hall, 2012), internal stresses (Kuroda et al., 2005;Rand, 2012) are generated at the micro-scale in the constituent phases of graphite grades. The manufacturing stresses retained in bulk graphite billets as internal (residual) stresses and may cause cracking (Hodgkins et al., 2006;Kelly, 2000;Kuroda et al., 2005). In addition, such unintended residual stresses in artificially manufactured graphite grades may limit their service life depending up on the neutron irradiation temperature and doses. ...
Micro-Raman microscopy technique is applied to evaluate unevenly distributed residual stresses in the various constituents of polygranular reactor grades graphite. The wavenumber based Raman shift (cm−1) corresponds to the local residual stress and measurements of stress dependent first order Raman spectra in graphite have enabled localized residual stress values to be determined. The bulk polygranular graphite of reactor grades – Gilsocarbon, NBG-18 and PGA – are examined to illustrate the residual stress variations in their constituents. Binder phase and filler particles have shown to be under compressive and tensile stresses, respectively. Among the studied graphite grades, the binder phase in Gilsocarbon has the highest residual stress and NBG-18 has the lowest value. Filler particles in Gilsocarbon have the highest residual stress and PGA showed the lowest, this is most likely due to the morphology of the coke particles used in the manufacturing and applied processing techniques for fabrications. Stresses have also been evaluated along the peripheral of pores and at the tips of the cracks. Cracks in filler and binder phases have shown mixed behaviour, compressive as well as tensile, whereas pores in binder and filler particles have shown compressive behaviour. The stresses in these graphitic constituents are of the order of MPa. Non-destructive analyses presented in this study make the current state-of-the-art technique a powerful method for the study of stress variations near the graphite surface and are expected to increase its use further in property determination analysis of low to highly fluence irradiated graphite samples from the material test reactors.
... Needle coke graphite is the most porous species [32]. Samples were cut from a 24 cm cube provided by the RA-6 Nuclear Reactor, of the Centro Atómico Bariloche, Argentina. ...
... In general, when oxidation rate is faster, but the supply of oxidant is limited (as in normal operating conditions) the penetration depth is expected to narrow. Blanchard [31] identified two oxidation regimes where the extent of gasification is limited by the rate of supply of oxidant (rate balance controlled regime) or by the fixed amount of oxidant available in a closed system, for example after blowing down and recharging the coolant (mass balance controlled regime). In these conditions, which are typical for normal VHTR operation, the location and the rate of oxidation is controlled by the balance between the rate of inwards diffusion of the oxidant and the rate of in-pore gasification by which the oxidant is consumed. ...
Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction (i.e. slow, continuous, and persistent) of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors' knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam (Velasquez, Hightower, Burnette, 1978). Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades that are being considered for use in gas-cooled reactors. The Langmuir-Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800-1100 degrees C), and partial pressures of water (15-850 Pa) and hydrogen (30-150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity between the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.
... The present theoretical three-dimensional structures of the heteroatom-doped graphenes are in excellent agreement with The structural properties such as lattice constants and heteroatom−carbon bond lengths for pristine and 11 heteroatom-doped graphenes are listed in Table 1. The available experimental data 28,29 and previous theoretical results 25,30 are also included in Table 1. The agreement between the present results and previous work is very good. ...
Low contact barrier electrodes and various field-emitting devices require a tunable work function and graphene is a dream material for these applications. In this work, the theoretical investigations on the variation of the work function for monolayer graphene doped with different kinds of atoms from groups IIA-VIA of the Periodic Table are reported. The geometry, density of states, dipole moment, and work function of each heteroatom-doped graphene are calculated using ab initio density functional theory with a dispersion correction. The obtained formation energy of the heteroatom-doped graphenes is in the order: N < B < P < O < S < Si < As < Se < Ge < Al < Ga. The work functions without an electric field abide by a periodic law in terms of doping atoms except for O-doped graphene. The calculated results demonstrate that the work functions of all heteroatom-doped graphenes are a linear function of the applied external electric field intensity and the slopes of the lines deviate from the ideal value in different extent, which is mainly dependent on the polarization of the heteroatom-carbon bonds and the production of the induced dipole moments of the doped graphenes. The present calculated results make known that the graphene work function can be tinkered up from 0.5 to 8.5 eV by using different kinds of doping atoms of groups IIIA-VIA elements and applying an electric field with various intensities. Such a wide range of adjustable work function makes graphene a very promising material for contact electrodes and field-emitting devices.
... The average oxidation rate for each material at 600 °C is shown in Table 1. Within the literature [7] oxidation rates for graphite at these temperatures are described as 'negligible' and considered not to be problematic at 675 °C. Our measured rate at 600 °C for Gilsocarbon is 3.7 Â 10 À5 %/min, and although extremely small, this cannot reasonably be considered negligible over the lifetime of plant graphite components. ...
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.
... These layers are stacked either in the ABAB sequence leading to the hexagonal 2H structure or in the ABCABC arrangement for the rhombohedral 3R structure. Normally, highly ordered or highly oriented graphite has the 2H hexagonal structure but even high quality samples still contain a non-negligible fraction of the 3R rhombohedral phase [4][5][6]. This is because graphite cannot be produced out of the melt at ambient pressure. ...
The atomic structure of several nuclear graphite samples, an essential moderator material for nuclear reactors, has been investigated by X-ray diffraction. The patterns were analyzed by the conventional Rietveld refinement approach as well as by the CARBONXS model, which takes into account disorder and stacking faults. The refined parameters compiled with those from literature reveal a generic picture for the structure of all graphite specimens.
This paper discusses an artificial neural inversion approach for non-destructive testing of the graphite moderator bricks that make up the cores of Advanced Gas-Cooled Reactors (AGR). The study employed fully connected feedforward neural networks consisting of four hidden layers and trained with backpropagation approach using Levenberg-Marquardt optimisation algorithm. The approach is based on multi-frequency (MF) eddy current (EC) measurements, and combinations of simulated and measured datasets from a laboratory sample and one of the operating reactor core, along with different regularisation parameters were used to train and test the networks. Various types of artificially generated errors were added to the data during training procedures, which in turn allowed for error tolerance and improved the generalisation of the ANNs to unseen test datasets. First, the ANN was tested using unseen simulated data, followed by the measurements collected from laboratory sample and one of the operating reactor core. The first test from the unseen simulated data showed mean profile error ranging between 1.30% and 8.20%, whereas the profiles estimated from reactor core measurements showed mean profile error ranging between 1.84% and 17.80% when compared with the resistivity measurements from trepanned graphite sample taken out of the reactor core. Further comparison of the network outputs against the profiles estimated using traditional iterative inversion algorithm indicates reasonable agreement between the two approaches with the exception of one case, but the solution time for the ANN was found to be over three orders of magnitude faster than the iterative inversion algorithm.
Multiple length-scale microstructural characterisation has been conducted on four different grades of unirradiated fine-grained graphite: CL 2020, IG 430, SGL R7650 and POCO ZXF-5Q, which are the potential candidates as high-power proton beam intercepting materials (secondary particle-production targets). Micrometre and sub-micrometre sized porosity structures have been characterised by micro-X-ray computed tomography (µXCT) and focused ion beam-scanning electron microscopy tomography (FIB tomography). Results show that these four grades of graphite have distinct porosity structures with many interconnected pores; POCO-ZXF-5Q has the most uniform porosity size and spatial distribution out of the four at these length scales examined. As a direct measurement of the total number and volume of nano-sized Mrozowski cracks is currently not readily feasible, the characteristic crystalline size La, which reflects the coherent scattering length at nano-scale was evaluated using micro-Raman spectroscopy. It has been found that a significant reduction in average characteristic crystalline size had been induced by sample preparation. Further measurements on freshly fractured sample surfaces excluding these damages suggested that IG430 graphite has the largest average crystallite size (∼290 nm) while POCO-ZXF-5Q has the smallest average crystallite size (∼80 nm) with a narrow distribution. These results from un-irradiated graphite were used to further understand the proton irradiated behaviour of these four grades of graphite.
This paper presents a novel system for mapping the through-wall electrical conductivity profiles of graphite moderator bricks that make up the core of an Advanced Gas-Cooled Reactor. Multi-frequency eddy current measurements were used to reconstruct the conductivity profiles as a function of depth from the bores of different graphite bricks. The study was carried out using a modified Levenberg Marquardt algorithm, along with a finite element based forward model. First, the algorithm was tested using a laboratory graphite sample with known electrical conductivity profile. Secondly, real datasets from one of the operating Advanced Gas-Cooled Reactors was collected and then used for reconstruction. The results in this study were compared with measurements from trepanned graphite samples taken from the reactor cores, and showed reasonable agreement in multiple cases, suggesting that this method could be a viable tool for non-destructively assessing the condition of the graphite core in Advanced Gas-Cooled Reactors.
Molecular dynamics simulations were performed on pristine graphite using an adaptive intermolecular reactive empirical bond-order (AIREBO) potential, which introduced various concentrations of point defects and different sizes in defect clusters. Induced by the accumulation of mono-vacancy and self-interstitial atoms (SIA), the changes to the geometry, density, Young’s modulus, fracture strength, and coefficient of thermal expansion (CTE) of graphite were simulated over a range of defect concentrations and cluster sizes. We found that accumulated Frenkel pairs lead to clearly-observed changes to the geometric and elastic properties in graphite crystal, but a relatively negligible effect on CTE. As expected, changing size of defect clusters with a given concentration relates to the variation of material properties. The property changes simulated in the present work are qualitatively consistent with the experimental observations. The results are expected to provide some insights into the link between cascade-induced point defects and experimentally irradiation-assisted material property changes.
This paper introduces a novel approach, developed through an academic and industrial collaboration, to the thermal treatment of nuclear graphite waste arising as a result of reactor decommissioning and oxide fuel assembly dismantling. A crucial part of the process is the thermal oxidation of the graphite via a plasma furnace. Laboratory scale treatment of the graphite found the oxidation rate to increase with temperature, with a significant increase in the CO/CO2 production ratio at T > 1000 °C. There was also a linear increase of oxidation rate with air flow rate, up to 100 ml/min, after which, the process became less efficient due to oxygen wastage. Effects of graphite particle size, over the range 0.5–10.0 mm, was found to be relatively small and importantly the effect of the graphite provenance on the oxidation rate was also found to be minimal. Treatment of virgin and irradiated graphite analogues, under the same conditions, showed little difference in oxidation behaviour, providing confidence that this process could be scaled up and used effectively in the disposal of reactor cores. Scale-up of this work was carried out on a “nuclear-ready” full-scale industrial facility, demonstrating graphite feeding, furnace operation and graphite destruction successfully. Experiments showed that comparable conditions between lab scale and pilot-scale showed similar oxidation behaviour, with 71.6 kg (equating to 90%) of graphite gasified in 6 h, giving an oxidation rate of ∼12 kg/h. Engagement with UK regulators has indicated that it is likely to be desirable to further investigate the possibility that isolation and confinement of graphite-derived ¹⁴C (t1/2 = 5730 yrs) in a carbon sequestration scheme may be an improvement on the baseline strategy, which is to store it in its solid form as untreated graphite in a geological repository.
IG-110 graphite samples were polished and irradiated with Xe ions at various fluences, then annealed at high temperatures up to 1100 °C. After irradiation, small hills were found on the polished surfaces, indicating an anisotropic swelling induced by irradiation. Around 30% swelling at a fluence of 2 × 10¹⁵ ions/cm² was characterized using atomic force microscopy. Severe swelling of the graphite crystallites caused stresses between adjacent crystallites, but leaved no intergranular cracks on the polished surface, which was ascribed to irradiation-induced creep of graphite. The pore morphology was affected by the anisotropic swelling. We found many contracted pores but only one expanded pore, indicating a decreased porosity induced by irradiation. After annealing at 1100 °C, TEM characterization showed clearly increased lattice order and decreased width of the (002) diffraction arc, indicating the annihilation of dislocations and recovery of basal plane rotations. Annealing-induced recrystallization of damaged graphite led to recovery of the crystallites' swelling and many small cracks appearing on the samples' surfaces.
This paper presents a summary of scientific results obtained during the last few years in the National Research Centre “Kurchatov Institute” (NRC KI). The main aims of this research are the development of a new experimental method that can be applied for the investigation of the influence of a high level of radiation damage on the plasma erosion effects of irradiated materials (graphite materials and tungsten) at different temperatures and the investigation of hydrogen isotope accumulation in these materials relevant to fusion reactor conditions (in ITER).
A high level of radiation damage in these materials (0.1 to 10 displacements per atom) was achieved by irradiating them with fast charged particles at the NRC KI cyclotron, simulating fast neutron irradiation in a fusion reactor. The plasma erosion effects in irradiated and nonirradiated materials were compared using the linear plasma simulator LENTA at NRC KI.
The performed investigations have shown that the erosion factor of irradiated graphite materials is increased and the tungsten surface structure is changed due to the accumulation of radiation damage. No influence of irradiation on tungsten erosion rate was observed in these experimental tests. The accumulation of hydrogen isotopes and helium concentrations were measured in the irradiated tungsten.
The performed work and the obtained results suggest a new promising experimental method for the experimental investigation of plasma effects on fusion structural materials at different irradiation temperatures and different radiation damage levels.
In this study, a heat transfer simulation of a UO2 particle-graphite system in highly enriched nuclear fuel at the Transient Reactor Test Facility (TREAT) was performed using the finite element method. Different factors that can impact fuel performance were modeled and implemented in the simulated micro-scale UO2 particle-graphite system. The fission fragments caused an irradiation-induced degradation of the thermal conductivity of the graphite, which added major heat resistance to the irradiated system. The effect of graphite quality and irradiation on the UO2 particles has also been evaluated, but neither has an impact as pronounced as the fission fragment damage to the graphite. By combining these factors, the dynamic temperature profiles were obtained, and the limitations on particle size in the irradiated and unirradiated UO2 particle-graphite systems have been determined.
This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behaviour of 36Cl. This radionuclide is mainly issued through the neutron activation of 35Cl by the reaction 35Cl(n, γ)36Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36Cl, we implanted 37Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature and electronic and nuclear irradiation improve this effect.
In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 °C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300°C, and would be more efficient in dry inert gas than in humid gas.
Due to the layered nature of graphite, the migration and interaction of point defects in the graphite crystal structure are highly anisotropic, and it is usually assumed that individual mobile lattice vacancies are confined to diffuse on a single plane. We present the results of ab initio calculations based on density functional theory which demonstrate that vacancies can, in fact, move between adjacent planes when they interact with one another via relatively low energy pathways, often with barriers of less than 1 eV. These interlayer transition mechanisms can significantly alter both the kinetics of point-defect aggregation and coalescence and also the resultant morphologies of multivacancy complexes that form as a result of migrating vacancies interacting.
9,10-anthraquinone (AQ) and its derivatives, i.e., benzofuro[5,6-b]furan-4,8-dione (BFFD), benzo[1,2-b;4,5-b']dithiophene-4,8-dione (BDTD) and pyrido[3,4-g]isoquinoline-5,10-dione (PID) are environmentally friendly and cheap electrode materials. The calculated binding energies of the four organic molecules are in the order: BFFD < BDTD < AQ < PID on monolayer graphene and BFFD < BDTD < PID < AQ on monolayer h-BN. The physisorption causes a work function shift relative to the isolated graphene or h-BN nanosheet in the order: AQ < BDTD < BFFD < PID on both the graphene and h-BN nanosheets. The strong physisorption suggests that the solubility of the four organic compounds in the electrolte solutions can be reduced by binding them to a graphene or h-BN nanosheet, making the organic compound-graphene or organic compound-h-BN composite a promising electrode material for lithium-ion batteries.
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