Graphite Waste Research Articles

Cyclodextrin-based functionalized graphene oxide as an effective corrosion inhibitor for carbon steel in acidic environment

Cyclodextrin-based functionalized Graphene Oxide (CD-GO), was synthesized utilizing cost-effective waste graphite as a source for production of graphene. The obtained novel material, CD-GO, was characterized and evaluated for corrosion inhibition of X60 carbon steels in solutions of HCl at room temperature. The inhibition efficacy of the CD-GO material was evaluated by the use of the weight loss measurement as well as electrochemical techniques. The findings from both were in excellent agreements. CD-GO demonstrated high inhibition efficacy which enhances with increasing CD-GO concentrations. The SEM/EDX, FTIR, and UV–vis surface characterization analysis prove the growth of the protective film of CD-GO on the steel surface. The CD-GO materials were adsorbed on the surface of the steel via oxygen atoms of the CD-GO forming a CD-GO/metal complex on the steel surface. This work is of importance and has dual benefits; graphene was prepared from waste graphite and converted into useful materials used to solve the corrosion problem.

ЭЛЕКТРОХИМИЧЕСКАЯ ОБРАБОТКА ОБЛУЧЕННОГО ГРАФИТА УРАН-ГРАФИТОВЫХ ЯДЕРНЫХ РЕАКТОРОВ

Актуальность исследования обусловлена необходимостью разработки эффективных методов дезактивации облученного графита уран-графитовых ядерных реакторов с целью снижения потенциальной опасности для дальнейшего захоронения. Цель: разработка метода электрохимической дезактивации графитовых радиоактивных отходов в химически агрессивных средах и определение оптимальных параметров ведения процесса. Методы: экспериментальные исследования процесса электрохимической обработки облученного графита в различных электролитах и при различных условиях. Результаты. Показана принципиальная возможность электрохимической обработки графитовых радиоактивных отходов. Представлена конструкция электролизера для проведения процесса обработки, и описана методика проведения эксперимента. Проанализированы электродные процессы, и определены факторы, влияющие на эффективность электрохимической дезактивации. Представлены зависимости скорости разрушения графитовых анодов от плотности тока и различных концентраций кислот. Определено, что наибольшая скорость разрушения электродов наблюдается в концентрированной азотной кислоте. Экспериментально были получены зависимости, отражающие степень очистки графитовых РАО по радионуклидам 60Co, 134+137Cs, 154Eu, 241Am в зависимости от состава электролита. Выявлено, что максимальная степень дезактивации облученного ядерного графита при плотности тока 0,01 А/см2и температуре электролита (16–20) °C достигается в азотной кислоте с добавлением 0,2 М NaF. Проведен сравнительный анализ результатов экспериментов химической и электрохимической обработки графитовых радиоактивных отходов. Показано, что электрохимическая дезактивация графита позволяет снизить активность графита по 60Со в 2–10 раз и по 137Cs в 7–100 раз. При этом в процессе дезактивации не образуется газообразных продуктов реакции, которые необходимо улавливать, создавая ещё большее количество вторичных радиоактивных отходов.

Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal

Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.

About chemical form and binding energy of 14C in irradiated graphite of uranium-graphite nuclear reactors

Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination of possible chemical compounds in which 14C can be found and assessment of fastness of its binding in the structure of irradiated graphite. Indigent and foreign experience of handling graphite radioactive wastes was analyzed, calculations and measurements were performed. Information was provided on the channels of accumulation of 14C in the structure of reactor graphite and it was demonstrated that the largest quantities of the radionuclide in question are generated according to the reaction 14N(n, p)14C. Here, most part of radioactive carbon is generated on 14N nuclei found in the form of impurities in non-irradiated graphite and in the composition of gas used for purging nuclear reactor in the process of operation. 14C radionuclide generated according to 14N(n, p)14C nuclear reaction is localized in the near subsurface graphite layer (in the near subsurface layer of pores) at the depth of not more than 50 nm. Analysis was performed of possible chemical compounds which may incorporate radioactive carbon. It was established that the form of occurrence is determined by the operational properties of specific graphite element in the reactor core. 14C binding energy in the structure of irradiated graphite was evaluated and depth of its penetration in the structure was calculated. It was established that selective extraction of this radionuclide is possible only under elevated temperatures in weakly oxidizing environment which is explained by the binding energy reaching up to 800 kJ/mole in the process of chemical sorption of 14C on the surface of graphite and depth of its occurrence equal to ~ 70 nm in the course of ion implantation. It was demonstrated that radioactive carbon generated according to 13C(n, γ)14C nuclear reaction is uniformly distributed among graphite elements and possesses binding energy ~477 kJ/mole. Its selective extraction is possible only under the condition of destruction of graphite crystalline lattice and organization of the process of isotopic separation. The obtained results allow recommending the most efficient methods of handling irradiated graphite during decommissioning uranium-graphite reactors.

Selective Adsorption of Textile Dyes using Pre-treated Graphite Waste and Graphite/Fe3O4 Composites

In the present study, we describe the application pre-treated graphite waste (PTG) and graphite waste/Fe3O4 composites as adsorbents to remove different textile dyes from water. Selective adsorption of methylene blue (MB), methyl orange (MO), methyl violet (MV), and rhodamine B (RB) textile dyes on PTG and graphite waste/Fe3O4 composites were studied. The removal of dyes in water was assessed to evaluate the selectivity of PTG and graphite waste/Fe3O4 composites. The graphite/Fe3O4 (1:0.5), graphite/Fe3O4 (1:1), and graphite/Fe3O4 (1:2) composites based PTG is abbreviated GF1, GF2, and GF3, respectively, were prepared with different weights of Fe3O4 magnetic nanoparticles with ratio of 1:0.5, 1:1, and 1:2 (w/w%). The adsorption experiment was carried out in batch reactor with adsorbent dosage of 0.1 g and dye concentration of 10 mg/L. The results showed that the optimum adsorption capability of the PTG was 96.3, 89.2, 100, and 75.3 for MB, MO, MV, and RB, respectively. Under comparable experimental condition, the GF2 (2M HCl) showed the optimum removal of dyes of 81.6% (MB), 40.9% (MO), 76.0% (MV) and 20.9% (RB). The adsorption capacity of the dyes is governed by electrostatic interaction between the dyes and the functional groups on the adsorbent. The PTG and the GF2 (2M HCl) composite are promising candidates as low-cost adsorbents for removal of pollutive dyes from waste water.

Preparation and adsorption property of graphene oxide by using waste graphite from diamond synthesis industry

Graphene oxides (GO) were prepared from waste graphite from diamond synthesis industry by using the modified Hummers method. The phase composition, microstructure, specific surface area and pore size distribution of the waste graphite and the as-prepared GO samples were characterized by X-ray diffraction, Fourier transformed infrared spectra, Ultraviolet visible spectroscopy, Raman spectroscopy, field-emission scanning electron microscopy and nitrogen adsorption-desorption tests. The adsorption property of GO to methylene blue (MB) was investigated. The results show that the as-prepared GO has few layers, specific surface area of 116.6 m2 g−1 and a little different structure from those prepared by using natural flake graphite. It exhibited excellent adsorption performance in MB aqueous solution. Adsorption time, GO dosage, initial concentration, and pH value of MB aqueous solution play important roles but the solution temperature had a little effect on the removal rate of MB and the adsorption capacity of GO. Under the condition of a GO dosage of 5.7 mg, room temperature, initial concentration of 12 mg L−1, pH value of 6.5 and adsorption for 40 min, the removal rate of MB and the adsorption capacity of GO reached 96.05% and 379.89 mg g−1, respectively. The adsorption process of MB onto GO met well with a pseudo-second-order kinetic model and Langmuir adsorption isotherm model. The adsorption reaction was spontaneous and purely based on endothermic process.

Thermodynamic Analysis of the Oxidation of Radioactive Graphite in the Na2CO3–K2CO3–Sb2O3 Melt in an Argon Atmosphere

Abstract—Graphite is used as a neutron moderator and reflector. Moreover, graphite can be used as sealants and bearings in reactors. The graphite mass in a reactor is 1–2 ths t. A large amount of radioactive graphite wastes forms when graphite reactors are taken out of service. The existing methods of processing radioactive graphite are based on its isolation from the environment. These methods cannot substantially decrease the volume of radioactive graphite wastes. As a result, the processing of irradiated reactor graphite by oxidation in molten salts can be considered as an alternative reclamation method, which can decrease the volume of radioactive graphite wastes. The oxidation of radioactive graphite in the Na2CO3–K2CO3–Sb2O3 melt in an argon atmosphere is thermodynamically simulated using the TERRA software package. The data obtained are used to analyze the distribution of elements over condensed and gas phases. Heating of the system to 1073 K is found to cause the evaporation of the condensed compounds of antimony and cesium. Upon heating to 1273 K, the condensed compounds of potassium, sodium, and chlorine evaporate. Heating to 1373 K leads to the evaporation of the condensed compounds of nickel. Heating to 1673 K brings about the evaporation of the condensed compounds of uranium, calcium, and strontium. Heating to 1773 K causes the evaporation of the condensed compounds of plutonium, beryllium, americium, and europium. At temperatures above 1773 K, only a vapor–gas phase exists in the system.

The effect of polypropylene fibers on graphite-natural hydraulic lime pastes

A reduction in greenhouse gas emissions has given rise to the search for low-embodied energy materials. Based on this assumption, natural hydraulic lime can be an alternative binder to cement, while the introduction of waste as inert material can be used to reduce the environmental impact of sand. A graphite-NHL composite had been previously analyzed but it showed both a low resistance to cracking and a brittle performance. Hence, in this paper, the effect of polypropylene fibers in this composite has been studied in order to enlarge this composite while improving its performance. Several series of natural hydraulic lime pastes based on NHL5 were prepared with a percentage of fiber ranging from 0% to 2% and graphite from 0% to 25% by weight substitution of natural hydraulic lime. Both the fresh state and the hydric, mechanical and thermal properties were analyzed. Fibers increased the ductile-like behaviour and toughness considerably, as well as its porosity. However, their combination with waste graphite and plasticizer considerably improved the properties of the plain pastes.

The complete oxidation of nuclear graphite waste via thermal treatment: An alternative to geological disposal

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.

Textile Dye Removal from Aqueous Solution using Modified Graphite Waste/Lanthanum/Chitosan Composite

We investigated various pre-treatment processes of graphite waste using thermal, mechanical and chemical methods. The aim of this work is to study the performance of modified graphite waste/lanthanum/chitosan composite (MG) as adsorbent for textile dye removal from aqueous solution. Effect of graphite waste resources, adsorbent size and lanthanum concentration on the dye removal were studied in batch experiments. Selectivity of MG was also investigated. Pre-heated graphite waste (NMG) was conducted at 80°C for 1 h, followed by mechanical crushing of the resultant graphite to 75 μm particle size, giving adsorption performance of ˜58%, ˜67%, ˜93% and ˜98% of the model dye rhodamine B (concentration determined by UV-vis spectroscopy at 554 nm), methyl orange (464 nm), methylene blue (664 nm) and methyl violet (580 nm), respectively from aqueous solution. For this process, the system required less than ˜5 min for adsorbent material to be completely saturated with the adsorbate. Further chemical modification of the pre-treated graphite waste (MG) with lanthanum (0.01 – V 0.03 M) and chitosan (0.5% w/w) did not improve the performance of dye adsorption. Under comparable experimental conditions, as those of the ‘thermal-mechanical-pre-treated-only’ (NMG), modification of graphite waste (MG) with 0.03 M lanthanum and 0.5% w/w chitosan resulted in ˜14%, ˜47%, ˜72% and ˜85% adsorption of rhodamine B, methyl orange, methylene blue and methyl violet, respectively. Selective adsorption of methylene blue at most to ˜79%, followed by methyl orange, methyl violet and rhodamine B with adsorption efficiency ˜67, ˜38, and ˜9% sequentially using MG with 0.03 M lanthanum and 0.5% w/w chitosan.

Preparation and Characterization of Graphite Waste/CeO2 Composites

In this research, the chemical modification of graphite waste with CeO2 was developed and characterized. Graphite waste was pretreated with mechanical to obtain the size 200 mesh (75 μm), and thermal methods at 110°C oven for 6 hours. Here, we demonstrate final properties of graphite before modification (GBM), activated graphite (GA) and graphite/CeO2 composite with variation of 0.5, 1 and 2 g of CeO2 (G0.5; G1; G2). The effect of CeO2 concentration was observed. The presence of cerium in modified graphite samples (G0.5; G1; G2) were analyzed using SEM-EDX. The results show that the best surface area was found in G2 is 26.82 m2/g. The presence of CeO2 onto graphite surface does not significantly increase the surface area of composites.

Graphene Oxide from Carbon Rod Waste

Carbon rods extracted from Zn-C primary battery waste was used as raw material for graphene oxide (GO) synthesis. The synthesis used a modified Hummers method by providing potassium permanganate-sulfuric acid as the oxidizing agent. XRD analysis confirms a significant change between the graphite waste pattern and the produced graphene oxide pattern. A major peak at 2θ 27 ° which present in the graphite waste pattern is disappeared after it converts to the product, as well as a broad peak under 25 ° referring the presence of amorphous carbon. A broad peak at low angle of 12.02 ° dominantly present in the prepared GO pattern as a characteristic peak of GO. Meanwhile, some small peaks at 2θ of 17.76 °, 28.58 °, and 37.28 ° confirming the presence of manganese oxide which was used as oxidizing agent. A sharp peak at 1700 – 1500 cm-1 in the FT-IR spectrum indicates the presence of –C=O group, and at 1600 cm-1 refers to –C=C group. It confirms that this research has produced the targeted GO. Even though, the purity is need to be enhanced by removing the rest of oxidizing agent that still exist in the material.

14C selective extraction from French graphite nuclear waste by CO2 gasification

The decommissioning of French gas-cooled nuclear reactors, all stopped since 1994, will generate 23,000 tons of nuclear waste containing 14C among other radionuclides. This work aims at presenting and testing a selective extraction treatment of this radionuclide from graphite waste by carbon dioxide gasification.In previous works, we highlighted that graphite turns into a heterogeneously degraded carbon with neutron irradiation and 14C could preferentially be concentrated in the most degraded zones. Moreover, these last react selectively with CO2 that would allow the 14C extraction from graphite waste.The results on graphite waste are promising with the quarter of 14C inventory extracted for a relative weight loss of few percent. Up to 68% of 14C inventory was extracted, but with a more important mass consumption. Thus, the potential of this selective 14C extraction would allow imagining alternative scenarios for graphite waste management.

Tailored synthesis of active reduced graphene oxides from waste graphite: Structural defects and pollutant-dependent reactive radicals in aqueous organics decontamination

Anode graphite was recovered from a spent lithium ion battery (LIB) and reutilized as a carbon precursor to obtain graphene-based materials. Characterization results revealed that impurities were removed from the obtained graphite powder by cleansing processes. The as-synthesized reduced graphene oxide (rGO) from the purified graphite (LIB-rGO) demonstrated excellent catalytic ozonation activity against organic pollutants removal. To probe potential catalytic active sites, LIB-rGOs with different defective levels but similar oxygen contents were synthesized. Catalytic ozonation tests revealed that a higher defective level resulted in a greater catalytic activity. Density functional theory (DFT) calculation further demonstrated that ozone molecules could spontaneously decompose into active oxygen species on graphene structural vacancies and edges, which consolidated the role of defective structure in catalytic ozonation activity. Meanwhile, we discovered the pollutant-structure-dependent behavior of dominant reactive oxygen species (ROS) with the aid of radical scavenging tests and electron paramagnetic resonance (EPR) spectra. For phenolic pollutants vulnerable to direct ozone attacking, superoxide radicals (O2−) and singlet oxygen (1O2) were found to be responsible ROS, whereas hydroxyl radicals (OH) were identified as the principle ROS for aliphatic organic pollutants destruction. This study not only put forward a possible way for reutilization of waste LIB anode, but also stepped further for investigating the catalytic ozonation mechanism towards the graphene-based materials including the active sites and the generation of ROS.

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1–100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

Synthesis and characterization of graphite/magnetite composite as low cost potential adsorbent from graphite waste

The comparative properties of pre-treated graphite waste and composite with magnetite nanoparticles were studied. The present work describes the thermal-mechanical method for pre-treated graphite electrode waste and chemical modification on the pre-treated graphite waste with magnetite nanoparticles (Fe3O4). The raw material is graphite electrode waste. The variables of temperature and time that affected the properties of pre-treated graphite waste was also observed. Pre-treated of graphite electrode waste was prepared via thermal process at temperatures of 60, 75, 90°C and various times of 30, 60 and 90 minutes, followed by mechanical crushing of the resultant graphite waste to 75 μm particle size. The synthetic material (graphite waste/Fe3O4 composite) was prepared with hydrochloric acid (0.1 M) for activation of the pre-treated graphite waste, followed by chemical modification with pre-treated graphite waste to Fe3O4 mass ratio 1: 1 (w/w). The experimental results showed that the pre-treated graphite waste are non-porous. Effect of magnetite nanoparticles (Fe3O4) revealed that the surface area of graphite waste/Fe3O4 composite increased from 8.44 m2/g to 64.58 m2/g. The EDX composition of Fe increased from 0.08 to 38.68 wt%, indicated that the modification of Fe3O4 nanoparticles onto graphite waste was carried out successfully. This data is useful for the preliminary treatment process of graphite electrode waste for further applications that require adsorbent preparation and modification.

Innovative radioactive graphite waste management in fission and fusion material decommissioning

Innovative radioactive graphite waste management in fission and fusion material decommissioning

Kinetic and thermodynamic study of adsorptive removal of sodium dodecyl benzene sulfonate using adsorbent based on thermo-chemical activation of coconut shell

Adsorptive separation of Sodium dodecyl benzene sulfonate (SDBS) using thermo-chemically and chemically activated coconut shell based adsorbent as well as using waste graphite powder has been investigated. The dependency of the extent of SDBS removal on pH, concentration, adsorbent dose and temperature has been demonstrated. Under optimized conditions, 92.84%, 39.41% and 34.23% SDBS could be removed in 3h using Phosphoric Acid Activated Coconut Shell-I (PAACS-I), Phosphoric Acid Activated Coconut Shell-II (PAACS-II) and graphite powder respectively. Adsorption kinetics was established to follow pseudo-second order mechanism and isotherm data fitted well with the Langmuir isotherm. Maximum adsorption capacity was estimated to be 28.57mg/g under optimized conditions of pH as 2, adsorbent dose of 3g/L and temperature of 293K. It was also established that adsorption was spontaneous and exothermic. Appreciable activity till 5th reuse cycle confirmed promising nature of the adsorbent.

Thermal oxidation of nuclear graphite: A large scale waste treatment option.

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.