Graphite Waste Research Articles

Search for Solving the Problem of Conditioning the Reactor Graphite

Spent reactor graphite constitutes a significant part of the accumulated solid radioactive wastes, and a search for its isolation from the biosphere is a topical task. Radioactive nuclides are listed in the paper, the activity of which determines the degree of contamination of the graphite stack and other construction elements of decommissioned uranium-graphite reactors. The suggested technical solutions must be based on a differentiated approach to various kinds of a graphite waste, which differ in the specific activity and spectrum of radionuclides. A preliminary fractionation (assortment) will make it possible to many times diminish the amount of waste delivered to the point of deep burial. A search for approaches to graphite processing must be subordinate to the problem of finding the less expensive technologies. Any procedure aimed to process graphite to remove 14C and 36Cl into a separate product will only make larger the total waste volume and require an additional expenditure. The development of even the most effective methods cannot provide the economic effect that can be obtained if the regulations on the content of impurities in graphite are reconsidered and a safe burial of the reactor graphite in near-surface storages is substantiated.

2.0-V flexible all-solid-state symmetric supercapacitor device with high electrochemical performance composed of MWCNTs-WO3-graphite sheet

Simple and scalable fabrication of 3D nanostructure MWCNTs-WO3-graphite sheet with favorite flexibility and excellent mechanical strength is demonstrated. The flexible sheet was fabricated by employing second-hand and waste graphite electrodes through uniform adding of Zn nanoparticles (ZnNPs) into matrix of multi-walled carbon nanotubes-tungsten oxide-graphite sheet followed by selective dissolving of ZnNPs from the sheet. The fabrication condition of the MWCNTs-WO3-graphite sheet was optimized for its application in flexible supercapacitor device. The MWCNTs-WO3-graphite sheet exhibited attractive electrochemical behaviors including a low ohmic drop and a high areal capacitance of 114 mF cm−2 (103 F g−1) at 1.7 mA cm−2 (1.5 A g−1) in aqueous 1.0 M H2SO4 electrolyte. All-solid-state symmetric supercapacitor device based on flexible MWCNTs-WO3-graphite sheets showed a high energy density of 57 μWh cm−2, a high operation potential of 2.0 V, and significant cycling stability with 96% areal capacitance retention after 2000 continuous cycles.

Numerical investigation of thermal gasification in enclosed screw-conveyor reactor for removal of 14C from irradiated graphite waste

ABSTRACT Investigated numerically are the effects of the rotational speed of the screw conveyor and the reactor temperature on the changes in the diameter of graphite particles passing through a laboratory-scale enclosed screw-conveyor reactor. Also investigated experimentally are a thermal gasification of graphite particles to carbon–oxygen compounds to determine a kinetic reaction model and kinetic parameters. The non-Newtonian fluid-like flow of graphite particles in the enclosed screw-conveyor reactor is numerically simulated by solving the momentum equation using a recently developed viscoplastic model with a rotating reference frame. The numerical approach coupled with the determined kinetic model of the gasification reaction of graphite powder successfully predicted the motion of the graphite particles and changes in their diameter flowing through the reactor. For a reactor temperature of 1073 K, the size of the feed graphite particles was predicted to decrease by 2% to 30% depending on the rotational speed of the screw conveyor. The predictions of the numerical simulations can be used to optimize the operating conditions of the enclosed screw-conveyor reactor for decontaminating irradiated graphite waste by removing the radiocarbons, which are located mainly on the graphite surface.

Characterisation of RBMK-1500 graphite: A method to identify the neutron activation and surface contamination terms

In this study, we provide the results of radiological characterization of the RBMK-1500 graphite using modeling and nuclear spectrometry; we also introduce a method to identify neutron activation and surface contamination terms for the spent graphite waste. The simplified model of the reactor core created by programme package SCALE 6.2 for 4 × 4 fragment of the core was used for simulation of neutron activation of graphite impurities in the RBMK-1500 reactor. Calculations were supplemented by the non-destructive measurements of gamma-ray emitting nuclides and destructive analysis of selected samples of the graphite stack of Ignalina NPP Unit 1 reactor. Our analysis demonstrates that the partial contribution of different contamination sources can be identified by combining modeling and measurements. The findings on radionuclides such as 14C, 36Cl, 60Co, 134Cs, 137Cs, 154Eu as well as actinides in Ignalina NPP Unit 1 graphite having a significant impact on graphite activity and radiological characterization are discussed. The proposed method is also applicable for identification of contamination source in the other activated components of the reactor core.

The green exfoliation of graphite waste and its suitability for biosensor applications.

This work is concerned with the bio-exfoliation of graphite using a soil bacterium. The isolated bacterium showed a detectable ability to oxidize and change its physical appearance and chemical structure. Multiple characterization procedures were used to study the physical and chemical changes. Raman and FTIR spectroscopy proved that the isolate G3 partially exfoliated the graphite into multi-layer sp2 graphitic layers. Scanning electron microscopy (SEM) proved that there was a change in morphology between untreated graphite waste and that manipulated by bacteria. Cyclic voltammetry results proved that the green exfoliated graphite (GEG) was suitable for use in biosensor applications and showed a noticeable ability to detect methanol, even at lower concentrations. These findings are considered as promising for the biological manipulation of graphite waste for environmental purposes. In addition, it is proved that the bacterial transformation of graphite into other GEG structures occurs without needing the chemically hazardous methods that are currently applied.

Content of Long-Lived Radionuclides of Carbon-14 and Chlorine-36 in Reactor Graphite and in the Biosphere (Is there a Problem with Carbon-14 and Chlorine-36 when It Comes to the Processing of Reactor Graphite?)

The amounts of technogenic carbon and chlorine contained in reactor graphite and in the biosphere are compared. The total amount of 14C on Earth and in the atmosphere, according to a rough estimate, is 2.0 × 108 Ci, while in all reactor graphite in Russia this value is no more than 3 × 107 Ci. The amount of 36Cl in reactor graphite is approximately (0.2–5) × 103 Ci, which is many times less than its content in the ocean, 7.16 × 108 Ci. Therefore, when disposing of reactor graphite waste in a near-surface storage, the yield of 14C and 36Cl will not lead to significant disproportions in their accumulation in the biosphere

Synthesis, characterization, and performance of graphene oxide and phosphorylated graphene oxide as additive in water-based drilling fluids

Graphene oxide (GO) and phosphorylated graphene oxide (PGO) have been synthesized from graphite electrode waste (GrW) or spent pot lining (SPL) from industrial aluminum wastes. The graphite waste was successfully recovered from the industrial aluminum wastes by alkaline and acid leaching method. The GO was then synthesized according to modified Hummers method in room temperature. The addition of oxygen into graphite structure was indicated by the formation of new vibrational bands at 3344, 1727, and 1069 cm−1 in the infrared (IR) spectrum, and the increase of oxygen content was confirmed by energy dispersive X-ray (EDX). The formation of single or few layers of graphene is confirmed through X-ray diffraction (XRD) analysis which show a down shift of 2θ peak and longer interlayer distance. Additional phosphate group to graphene oxide in functionalization process, in order to enhance the dispersion of graphene in water, was confirmed by scanning electron microscopy-energy dispersive X-ray (SEM-EDX). The effect of heat treatment on phosphorylated graphene oxide (PGO) was investigated by calcination at different temperatures (200–500 °C). The Fourier transform infrared (FTIR) spectroscopy and EDX results indicated that heat treatment decreased the oxygen containing group in PGO. An addition of GO as additive in water-based drilling fluids resulted in better rheology properties (plastic viscosity (PV) reduced from 10 to 7 cp and yield point (YP) increased from 11 to 15 lbs/100 ft2), an increase in fluid loss control performance (filtrate volume decreased from 6 mL to 3.6 mL and filter cake thickness reduced from 1.06 to 0.33 mm), and an increase in lubricity ability (lubricity coefficient reduced from 0.158 to 0.119). In the alkaline environment, the bentonite particles are negatively charged, and GO may also turn to be negatively charged. The repulsive static interaction might reduce the plasticity. The zeta potential of the GO and PGO was measured to be −38.90 and −41.23 mV, indicating that both are not easily agglomerated. The additional PGO as additive in water-based drilling fluids showed a negative impact due to significant pH decreased of drilling fluid.

Synthesis and Characterization of Graphite Oxide, Graphene Oxide and Reduced Graphene Oxide from Graphite Waste using Modified Hummers’s Method and Zinc as Reducing Agent

Synthesis and Characterization of Graphite Oxide, Graphene Oxide and Reduced Graphene Oxide from Graphite Waste using Modified Hummers’s Method and Zinc as Reducing Agent

One-pot redox synthesis of graphene from waste graphite of spent lithium ion batteries with peracetic acid assistance

In order to meet increasingly urgent application requirements for energy storage and aviation materials, it is necessary to facilitate the rapid, low-cost, large-scale preparation of high-quality graphene. In this study, reduced graphene oxide (rGO) was prepared from the waste graphite of spent lithium batteries by a one-pot redox reaction using a ternary system contains sulfuric acid, potassium permanganate and peracetic acid. Peroxyacetic acid was first used in the system to effect oxidation and reduction preparation of graphene to avoid additional reduction steps. XRD, XPS, SEM, TEM, FT-IR, and Raman spectroscopy analyses showed that reduced graphene oxide samples synthesized from waste graphite (L-rGO) and commercial graphite (C-rGO) have similar microscopic morphologies and structural characteristics. Therefore, lithium ion battery waste graphite is a good potential resource for the preparation of high-quality rGO. The carbon-to-oxygen ratios of L-rGO and C-rGO were 1.79 and 4.25, respectively, indicating that peracetic acid effectively reduced graphene oxide to rGO. The yield of rGO was calculated to be 61.2%, which means that rGO can be produced on a large scale using this method.

Determining the porosity and water impregnation in irradiated graphite

Impurities in nuclear graphite can become neutron-activated during operation, generating radionuclides and leading to modifications in the microstructure of the graphite due to fast neutrons. The long-term disposal of nuclear graphite waste requires collecting data on the behaviour of long-lived radionuclides in disposal conditions (water-saturated). The release of radionuclides in solution depends on a number of physicochemical processes such as water ingress into the graphite structure (radionuclides sites), as well as the solubilisation and transport of radionuclides in solution through the graphite pores. Analysis of the impregnation (impregnation speed, impregnation rate) of water in the porous graphite environment represents one of the main parameters that will greatly influence the physicochemical processes controlling the release of radionuclides in solution. For this reason, the impregnation of irradiated samples from the G2 and St Laurent A2 (SLA2) gas-cooled graphite-moderated reactors was studied, as was the distribution of the porosity. Results show that the geometric density of the samples decreases after irradiation, which is expressed as an increase in the total porosity or more precisely in the open porosity mainly due to CO2 radiolytic corrosion. This means that irradiated graphite is mainly a macroporous solid. Nuclear graphite becomes a hydrophilic material with its open porosity filled by water in just a few days. This is due to the presence of functional groups on the pore surface.

Effect of Nano-Copper Additives on the Physical Properties of Porous Alumina Ceramics using Graphite Waste as a Pore-Forming Agent

Solid state and sacrificial techniques were used to prepare the porous alumina ceramics. The target of the present research is to investigate the effect of Cu metal addition of nanoscale particle size on the physical properties of porous alumina ceramics using graphite waste as a pore-forming agent. X-Ray Diffraction (XRD), a Field Emission Scanning Electron Microscope (FESEM) and a Transmission and Electron Microscope (TEM) were used to analyse the microstructure and ceramic phases. Different ratios of Cu metal were added (3, 6, 9 and 12 wt.%) at different ratios of graphite waste. The results of this investigation show that with increasing ratios of Cu metal, the porosity decreased while the shrinkage and densities increased. Increases in the density and decreases in the porosity of porous alumina ceramics and the formation of the tenorite (CuO) phase due to sintering at high temperature (1600°C) may lead to improving the strength of porous alumina ceramics. Some potential applications include gas filtration components and as thermal insulation materials.

Corrosion Behaviour of Mg Alloys in Alkaline Solutions and in Alkali-Activated Binders

The reprocessing of spent fuels from UNGG (Uranium Natural Graphite Gas) nuclear reactors in France generates cladding wastes mainly composed of Mg alloys. The management of these wastes involves their immobilization in hydraulic binders with high pH interstitial solutions (12.5<pH<13.7). In the case of magnesium corrosion, the choice of these materials for immobilization and storage is not straightforward due to the characteristic reactivity of these alloys, which is caused by several aspects [1]. First, magnesium corrosion is strongly influenced by the pH and the composition of the electrolyte (the pore solution of hydraulic binder). Furthermore, the electrochemical potential of Mg places this metal as an anode with respect to several materials [2] (e.g. residual graphite materials initially present in the fuel assemblies or steel from the container), which can create galvanic couplings and accelerate corrosion. Previous work has found low corrosion rates of Mg-Zr alloys and low H2 releases produced when Mg is immobilized in a Na-geopolymer mortar in the presence of fluorides (magnesium corrosion inhibitor) [3]. Similarly, there was a low release of H2 from corrosion during the galvanic coupling between Mg-Mn and graphite, as well as the formation of thinner corrosion product layers when Mg-Mn is immobilized in an alkali-activated slag mortar [4]. These results have contributed to the choice of these two materials as the most appropriates for embedding these wastes. Nevertheless, the electrochemical mechanisms responsible for this weaker corrosion, compared to conventional cementitious matrices, must be identified in these two reference scenari. The purpose of this work is to investigate the corrosion processes that occur once the Mg alloys are embedded in their immobilization matrices. The goal is to understand the phenomena that lead to low corrosion rates in these materials, in the case of general and galvanic corrosions. From these guidelines, a study of the cementitious matrices (porosity, resistivity and gas diffusion) and its interstitial solutions (pH, ionic species and conductivity) was carried out in order to identify the main parameters that can limit the corrosion rate of magnesium. The corrosion behavior of Mg alloys against these conditions has been tested in model solutions and in each cementitious binders using electrochemical techniques (OCP, ZRA, EIS and polarization curves), gravimetry and surface characterization (XRD, SEM/EDS and FIB/STEM). Both matrices are relevant for the storage of Mg wastes and have a thermodynamically favorable pH for magnesium passivation due to the precipitation of Brucite (Mg(OH)2) on its surface. The high resistivity of the alkali-activated slag and the possible consumption of O2 by the sulphides of the pore solution are favorable aspects to reduce galvanic corrosion rate. In the case of the Na-geopolymer/NaF, the oxygen concentration can also limit the rates of galvanic corrosion, since, during storage, oxygen access could be restrained. However, the presence of corrosion inhibitors (silicates and fluorides) is mainly responsible for the reduction of Mg corrosion, because of a protective film formed at the MgZr/Na-geopolymer interface. This corrosion rate may vary with the concentration of fluoride available in solution and is probably related to the nature of the corrosion product formed on the magnesium surface. General and galvanic corrosion in the Na-geopolymer/NaF mortar were quantified by gravimetric measurements and in both cases the nature of the corrosion products was characterized by scanning electron microscopy (SEM/EDS), X-ray diffraction (XRD) and FIB/STEM (Focused Ion Beam/Scanning Transmission Electron Microscope). The results revealed the stratification of the corrosion products found at the Mg-Zr/Na-geopolymer interface and were in good agreement with the experiments performed in the model alkaline solution. Magnesium corrosion in the Na-geopolymer/NaF mortar is rather limited by the chemistry of the interstitial solution. [1] K. W. Guo , “A Review of Magnesium/Magnesium Alloys Corrosion and its Protection”, Recent Patents on Corrosion Science, vol. 2, pp. 13-21, 2010. [2] G. L. Song and A. Atrens, “Corrosion Mechanisms of Magnesium Alloys”, Adv. Eng. Mater., vol. 1, no. 1, pp. 11–33, 1999. [3] D. Chartier, B. Muzeau, L. Stefan, J. Sanchez-Canet and C. Monguillon, “Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag”, Journal of Hazardous Materials, vol. 326, pp. 197-210, 2017. [4] D. Lambertin, F. Frizon and F. Bart, “Mg-Zr alloy behaviour in basic solutions and immobilization in Portland cement and Na-geopolymer with sodium fluoride inhibitor”, Surface and Coating Technology, vol. 206, pp. 4567-4573, 2012.

Electrochemical treatment of irradiated nuclear graphite

The paper is dedicated to the issues of accumulation of irradiated reactor graphite and possible options for handling it. There are described the advantages of the electrochemical method of decontamination of radioactive waste widely used in the nuclear industry. The use of this method to reduce the potential hazard of irradiated reactor graphite is proposed. The processes occurring during the removal of radioactive contamination from the surface of graphite radioactive waste are described. An experimental setup is presented to assess the possibility of using the method of electrochemical decontamination of irradiated graphite. The results of the determining the electrolyzer current-voltage curve and the dissolution rate of the electrodes made of irradiated graphite are presented. According to the results of experimental studies, data on the decontamination coefficients of irradiated graphite were obtained for various radionuclides (60Co, 137Cs, 154Eu, 152Eu) in HNO3, H2SO4, H2O, H2O2, HNO3 + KMnO4, H2SO4 + KMnO4 under various process modes. The evaluation of temperature fields inside the electrolyzer was carried out. The dependence of the removal efficiency of long-lived 14C radionuclide on the total mass loss of irradiated graphite was obtained.

The electrochemical deconsolidation mechanism of graphite matrix in HTGR spherical fuel elements

In High-Temperature Gas-cooled Reactor (HTGR)s, coated fuel particles are dispersed in a graphite matrix, which forms both compact and spherical fuel elements. As for the post irradiation examination (PIE) of HTGR irradiated fuels, the electrochemical deconsolidation method in which graphite matrix is disintegrated into graphite powder is one of the feasible processing methods. The graphite matrix is composed of natural graphite, artificial graphite and glassy carbon from carbonized resin, which have different effects on electrochemical deconsolidation. In this work, synthesized graphite matrix specimens were produced by intentionally varying the resin binder content in the graphite matrix raw materials from 5% to 50%, in order to investigate the electrochemical deconsolidation mechanism of graphite matrix in HTGR fuel elements. The results showed that increasing the glassy carbon content increased the degree of chemical oxidation, with more pronounced production of graphite oxide. The driving force for graphite matrix deconsolidation is proposed to be a synergetic effect of the expansion stress around closed- and through-hole microstructure originating from oxidatively destroyed graphite layers. This work will not only help to more thoroughly understand the electrochemical deconsolidation procedure of HTGR fuels, but also contribute to the development of a feasible method to manage radioactive graphite waste.

Perbedaan Jenis Crucible Pada Peleburan Sampah Aluminium Terhadap Waktu Peleburan Aluminium

Abstrak&#x0D; &#x0D; Sampah aluminium merupakan sampah padat yang sulit diurai oleh lingkungan. Waktu penguraian aluminium berkisar 80-100 tahun. PT. Combiphar merupakan industri farmasi yang kegiatan produksinya menghasilkan sampah aluminium. Setiap minggunya timbulan sampah aluminium mencapai 45 kg. Variasi jenis crucible, dapat menghasilkan waktu yang berbeda pula. Jenis crucible yang digunakan yaitu stainless steel dan grafit. Jenis crucible tersebut memiliki konduktivitas termal yang baik. Jenis penelitian ini adalah bersifat eksperimen dengan desain penelitian post test without control. Sampah aluminum berasal dari PT. Combiphar dengan jumlah sampel yang digunakan yaitu 2,07 kg. Berdasarkan hasil penelitian dari 9 kali pengurangan untuk 2 jenis crucible yaitu jenis crucible stainless steel dan jenis crucible grafit diketahui bahwa jenis crucible grafit lebih cepat meleburkan sampah aluminium dengan rata-rata waktu 45.78 menit dalam 9 kali pengulangan sedangkan jenis crucible stainless steel memiliki rata-rata waktu peleburan 58.56 menit dalam 9 kali pengulangan.

Synthesis and studying of reduced few-layered graphene coatings in gas sensor applications

In this work reduced few-layered graphene (rFLG) nanoparticles were synthesized using electrochemical pulse exfoliation method from waste graphite crucibles. The regular change in voltage polarity in the synthesis process ensures both the separation of graphite in layers and the reduction of graphene oxide. A method for synthesizing free-standing rFLG and nitrocellulose (NC) composite film has been developed involving creation of rFLG-NC ink that can be deposited on various substrates. It has been observed that a successful synthesis of a free-standing composite coating is possible with the mass ratio of rFLG:NC at least 9:1 of which resistivity is on the order of approximate 10 ohm-centimetres in dependence from temperature and surrounding atmosphere. Sensor electrode was prepared by simple dip-coating method and the response to humidity, organic solvent vapours and nitrogen containing gases was measured and conclusion made that rFLG-NC coating is sensitive to many kinds of gases (cross-sensitive) what may result in false detecting or can be used to multi-sensor chips for artificial olfaction devices.

SYNTHESIS OF GRAPHENE OXIDE AND REDUCED GRAPHENE OXIDE FROM GRAPHITE FOIL INDUSTRIAL WASTES

Natural and synthetic graphite, graphite intercalated with different compounds and expanded graphite are used for the obtaining of graphene oxide (GO) and reduced grapheme oxide (rGO). In this work, it is confirmed that powdered graphite foil wastes (pGFW) can be successfully used for the synthesis of GO and rGO. During the implementation of certain technological operations, the unused graphite foil pieces (wastes) are left, which are unfit for exploitation purposes. Practically these are expanded graphites and their chemical oxidation to GO or obtaining of graphene can be conducted using known methods with some adjustments. A fraction with a particle size of <140µm was obtained by wet and dry grinding, which further was purified in order to remove various types of impurities. As the EDX analysis shows, the powder consists of carbon and oxygen. Oxidation of powders was conducted using the same methods as in case of different types of graphite precursors. The paper presents results obtained in pGF oxidation using low-temperature (~0°C; KMnO 4 –NaNO 3 –H 2 SO 4 ) and relatively high-temperature (~50°C; KMnO 4 –H 2 SO 4 ) modes. In case of low-temperature mode oxidation of pGFW the C/O ratio (at.%) is 61:38. In case of their reduction with ascorbic acid the C/O ratio is 81:19. The method of synthesis of GO and its separation from the reaction mixture were partially corrected. Sulfuric acid and ions (K + ,Na + , and Mn +2 ) can be removed using 5-fold decanting (2 times H 2 O, 3 times HCl solution 5%). A 5% solution of HCl precipitates GO-flakes in 7–10min and, thus, the process of removing the main impurities is accelerated. From decanted solutions GO was reduced to the rGO with ascorbic acid at 80°C. By the high-temperature treatment of rGO received from graphite foil wastes graphene is obtained with a defective structure.

A new analysis workflow for discrimination of nuclear grade graphite using laser-induced breakdown spectroscopy

Stand-off, in-situ, laser induced breakdown spectroscopy (LIBS) offers a rapid, safe, and cost-effective method for discrimination of radioactive waste materials arising during the operation of nuclear plants and from decommissioning activities. Characterisation of waste materials is a critical activity in understanding the nature of the waste, ensuring hazardous material is managed safely and that waste can be segregated for reuse, recycle or sentenced for appropriate disposal. Characterisation of materials, often in hostile environments, requires the ability to remotely differentiate between materials in terms of their chemical composition and radioactivity.This proposition was tested using a case study on nuclear grade graphite. Graphite has been used extensively as a moderator material in many nuclear reactors. Internationally, over 250,000 tons of various nuclear-grade graphite, and graphite-bearing, materials exist. These are a major issue for nuclear decommissioning and radioactive waste management, due to the long half-lives of the associated 14C and 36Cl isotopes. LIBS offers a method for discrimination of nuclear grade graphites and other carbon and non-carbon-bearing wastes. This paper describes the development of a workflow method, including LIBS measurement analysis, for the discrimination of pre-irradiated nuclear ‘Pile Grade A’ (PGA) graphite moderator rod and domestic lumpwood charcoal, which act as surrogates for nuclear grade graphite and other carbon-bearing wastes. A new analysis workflow comprising the examination of spectral characteristics, multivariate analysis and molecular isotopic spectroscopy is proposed to enable rapid segregation of graphite from a heterogeneous waste stream. Enhanced characterisation techniques have the potential to significantly reduce the cost of decommissioning large parts of legacy nuclear generation plants.

The exfoliation of irradiated nuclear graphite by treatment with organic solvent: A proposal for its recycling

Abstract For the past 50 years, graphite has been widely used as a moderator, reflector and fuel matrix in different kinds of gas-cooled reactors. Resulting in approximately 250,000 metric tons of irradiated graphite waste. One of the most significant long-lived radioisotope from graphite reactors is carbon-14 (14C) [1] with a half-life of 5730 years, this makes it a huge concern for deep geologic disposal of nuclear graphite (NG). Considering the lifecycle of NG a number of waste management options have been developed, mainly focused on the achievement the radiological requirements for disposal [2] . The existing approaches for recycling depend on the cost to be economically viable. In this new study, an affordable process to remove 14C has been proposed using samples taken from the Nuclear Power Plant in Latina (Italy) which have been used to investigate the capability of organic and inorganic solvents in removing 14C from exfoliated nuclear graphite, with the aim to design a practicable approach to obtain graphite for recycling or/and safety disposed as L& LLW.

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.