Irradiated Graphite Research Articles

Decontamination performance of metallic radionuclides in irradiated graphite via a fluidized bed reactor

Metallic radionuclides in irradiated graphite waste are extremely hazardous to the biosphere. In consideration of its simple process, the chlorination roasting could be a promising option for the decontamination of metallic radionuclides from irradiated graphite waste. A novel numerical model for the removal of metallic radionuclides in irradiated graphite via a bubbling fluidized bed reactor was built up to evaluate the decontamination performance of chlorination roasting in large-scale treatment. The gas-solid flow was coupled with the reaction kinetics determined by experimental results and quantum chemistry calculation. A bubbling fluidized bed reactor with treatment capacity of 36–360 kg h−1 was designed, which was in terms of the effects of temperature, carrier gas velocity, HCl content and feed rate on the decontamination efficiency (DE). Temperature and HCl content had significant effects on DE. DE increased from 12 % to 100 % with rising temperature from 1173 to 1573 K. DE grew from 62 % to 99 % with increasing HCl content from 10 % to 20 %. Machine learning model was applied to directly determine the decontamination performance of the fluidized bed reactor, which provided reliable assessment in the chlorination roasting process. This study would be beneficial to large-scale treatment of metallic radionuclides in irradiated graphite via a fluidized bed reactor.

Roadmap of Graphite Moderator and Graphite-Matrix TRISO Fuel Management Options

Most high-temperature reactors use graphite as a moderator and structural material. This includes high-temperature gas-cooled reactors with helium cooling and TRi-structural ISOtropic (TRISO) fuel particles embedded in graphite, as well as fluoride salt–cooled high-temperature reactors with clean salt coolant and TRISO fuel particles embedded in graphite and thermal spectrum molten salt reactors with a graphite moderator and fuel dissolved in the salt. The largest volume radioactive waste stream from these reactors is the irradiated graphite. We describe herein a roadmap for management of these graphite wastes that contain radioactive 14C, tritium, and other radionuclides. There may be some graphite wastes with sufficiently low radioactivity levels that can be treated as nonradioactive waste and managed like other graphite waste. Management options for the graphite include (1) direct disposal, (2) recycled back to the reactor or other nuclear applications, and (3) oxidizing the graphite with release as an effluent or underground sequestration of the carbon dioxide. Cosequestration of this carbon dioxide with carbon dioxide from industrial, biological, and cement production processes can isotopically dilute the 14C before sequestration to eliminate the possibility of exceeding individual radiation exposure limits. We also describe options for processing graphite-matrix TRISO fuel, including separating the bulk graphite to reduce the volumes of used fuel for disposal or processing to recover fissile materials. The inventories of radioactive isotopes in different carbon wastes vary by many orders of magnitude; thus, there is no single economic option for the management of all graphite waste.

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

Generation of plasma from multiple targets by laser ion source for TIARA ion implanter

We investigated a laser ion source for use as an ion source for the ion implanter in Takasaki Ion Accelerators for Advanced Radiation Application to increase the number of ion species that can be produced and to enable rapid switching of ion species. Herein, we analyzed the charge state distribution of ions in a laser plasma generated through irradiation of graphite, titanium, copper, and tantalum targets with 30 mJ Nd:YAG laser to determine the particle number of low-charge ions per pulse at 1.1 m from the target. The particle number of singly charged ions could be increased by adjusting the laser power density. For carbon, titanium, and tantalum, the particle numbers were in the order of 1010 ions per pulse, whereas for copper, the particle numbers were in the order of 109, lower than for other target materials.

An In Situ transmission electron microscopy study on the synergistic effects of Au-ion irradiation and high temperature on nuclear graphite microstructure

The combined effects 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 ×1010 ion cm−2 s−1 for 70 min at 800 °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 °C–800 °C temperature region in which the dimensional change ranges from negative to positive values.

Chlorination mechanism of typical metallic radionuclides by HCl on irradiated graphite surface from first principles

Radioactive activity of metallic radionuclides accounts for a considerable proportion in irradiated graphite from nuclear reactors. Chlorination roasting is a promising method to remove metallic radionuclides in decommissioned irradiated graphite. The decontamination effect and chlorination mechanism of 90Sr and 60Co by HCl over an irradiated graphite surface were estimated by density functional theory. 90Sr and 60Co could be chlorinated via X(signifying metallic radionuclide)→XCl→XCl2 reaction pathway. The rate-determining step appeared where metallic radionuclides desorbed from the irradiated graphite surface. The chlorination energy barriers of 90Sr and 60Co were 236 and 575 kJ/mol, respectively, implying that chlorination roasting method for decontaminate of 90Sr and 60Co in the irradiated graphite was feasible. This work is beneficial to the application of chlorination roasting for removal of metallic radionuclides in irradiated graphite.

One-pot in situ synthesis of expandable graphite-encapsulated paraffin composites for thermal energy storage

To mitigate the intermittency of renewable energy generation and achieve net-zero greenhouse gas emissions, a cyclically stable and energy-efficient thermal energy storage system (TESS) is increasingly required. To prevent leakage and enhance thermal conductivity and stability, TESS is typically fabricated by impregnating energy storage materials (ESMs) into porous conductive additives. However, the impregnation-based encapsulation process is both time-consuming (several to tens of hours) and energy-intensive, and the incorporation of additives reduces the proportion of ESMs, resulting in decreased energy capacity. In this study, we present an ultrafast (several minutes) and energy-efficient one-pot encapsulation method for in situ encapsulation of paraffin wax (PW) within the pores of expanded graphite (EG) via microwave irradiation of expandable graphite (Eg) in molten PW. The resulting binary PW/EG composites exhibit shape stability and no leakage even at 92 wt% PW loading. The PW/EG composite with 10 wt% EG demonstrates a thermal conductivity of 3.7 W/mK (16.2 times higher than that of pure PW), a thermal release/absorption efficiency of around 99.5 %, and almost 100 % capacity retention after thermal cycling. Additionally, we demonstrate that integrating PW/EG TESS with thermoelectric (TE) modules can increase TE power generation by more than fivefold, highlighting their potential for energy storage and conversion as well as mitigating waste heat source intermittency.

Characterization and ion-irradiation studies of typical microstructures in nuclear graphite

Nuclear graphite is an important structural material for advanced gas cooled reactors and molten salt reactors. Graphite microstructures are always accompanied with typical pores or cracks which plays an important role in nuclear graphite behaviors. A new surface processing method involving ion milling and fast oxidation is used here to characterize typical microstructures from the large-size calcination cracks in fillers to very small quinoline insoluble particles in NBG-17 and NBG-18 graphites. By using this surface processing method and ion irradiation technique, an attempt is made to explore the microstructural basis of graphite irradiation behaviors. Heating-induced relaxation of graphite microstructures is observed, indicating the presence of residual stresses in nuclear graphite. Creased graphite sheets are identified on ion-milled surface and are believed to accommodate a-axis shrinkage of graphite crystallites at the beginning of irradiation, which is consistent with the initial plateau of slow dimensional shrinking. Compared to NBG-17 graphite, NBG-18 is found to have much more calcination cracks which contract during irradiation, resulting in a deeper volume shrinkage of NBG-18. The quinoline insoluble particles containing chaotic structures are shown to severely contract and separate from surrounding structures, leading to voids in between. Both NBG-18 and NBG-17 have abundant quinoline insoluble particles, which might make a contribution to their smaller volume shrinkage than other medium-grain graphite.

Fracture toughness evaluation of a nuclear graphite with non-linear elastic properties by 3D imaging and inverse finite element analysis

Effective small specimen tests are needed to obtain fracture toughness and elastic properties as the limited availability of irradiated graphite restricts the quantity and dimensions of test specimens. Both properties have evaluated simultaneously in a crack propagation test with the double cleavage drilled compression (DCDC) specimen geometry of a fine-grained graphite (SNG742) that has non-linear elastic properties in the unirradiated condition. Three-dimensional displacement fields were obtained by digital volume correlation of in situ laboratory X-ray computed tomographs, and the 3D crack geometry (crack tip position, crack opening displacements and angle) was determined objectively by a wavelet variance method. The tensile softening of the Young modulus was determined by inverse analysis of the strain field using the finite element model updating (FEMU) method. The strain energy release rate of the quasi-static propagating crack was calculated using the contour integral method in a finite element model with the derived non-linear elastic properties and the measured displacements as boundary conditions. The critical strain energy release rate was constant with crack length (118 ± 14 J m−2) and equivalent to a fracture toughness of 1.13 ± 0.07 MPa m1/2.

Investigation on decontamination kinetics of metallic radionuclides in irradiated graphite by chlorination roasting

Irradiated graphite used as moderator and reflector material in reactor core in nuclear industry is an important radioactive waste after reactor decommissioning. Treatment of irradiated graphite waste has been a challenge for IAEA member states. Chlorination roasting was proposed in order to remove the metallic radionuclides in irradiated graphite waste. The promising advantages of the proposed method include that the category of graphite waste can be reduced with near-zero carbon emission, and it can lead to potentially the recycling of irradiated graphite and metallic radionuclides. The decontamination kinetics of metallic radionuclides was evaluated based on a lab-scale tube furnace combined with density functional theory (DFT). 56Fe ions were implanted into nuclear graphite to simulate metallic radionuclides derived by neutron radiation. Decontamination efficiencies of Fe at 900, 1000, 1100, and 1200 °C for the retention time of 210 min were 42.6%, 81.6%, 94.4%, and 96.7%, respectively. Fe could volatilize from graphite in the form of complex metal chloride with retaining a large amount of graphite inert matrix. The decontamination process was described by the second order reaction model. Through DFT evaluation, Fe was chlorinated via Fe → FeCl → FeCl2 reaction pathway. This decontamination process is efficient, cost-effective, and reliable, which provide a valuable contribution to managing irradiated graphite.

Porosity evolution in proton irradiated microfine-grained POCO graphite

This work investigated the porosity evolution of POCO ZXF-5Q graphite that has been irradiated by 340 kW, 120 GeV protons inside NT02 target system in Fermilab's NuMI beamline. This POCO graphite has undergone direct bulk dimensional swelling at low dose irradiation and its local microstructural change is still not well-understood during this process. In this work, the (sub-) micrometre scale porosity from six locations across proton beam fluence and temperature gradients have been studied using focused ion beam-scanning electron microscopy (FIB-SEM) tomography. A deep learning-based tomographic image segmentation technique has been established and implemented for porosity segmentation and quantification. It has been found that there is a decrease in the total volumetric percentage of the porosity at proton beam centre (∼ 8 – 8.4 vol.%), by comparing to un-irradiated POCO (∼ 12 – 13 vol.%) and to beam 2σ and 5σ radii (∼ 12 vol.%). This decrease in porosity volume percentage was found to be caused by the reduction in pores with volumes > 0.1 μm3 induced by material bulk dimensional swelling at proton beam centre area. The porosity reduction in relation to dimensional change and irradiation creep was discussed among with other contributing factors, and further investigations through well-controlled irradiation experiment are still needed.

Temperature dependence of Cs release during oxidation of Cs-loaded graphite

As one of potential treatments for irradiated graphite (i-graphite) waste, the oxidation method has been attracting more concern due to features of high treating efficiency and significant volume reduction. The release of fission products during oxidation is of great significance for further off-gas decontamination procedures. Focusing on the release behavior of volatile cesium (Cs), Cs-loaded graphite was prepared by impregnation in this work followed by definite distribution and chemical speciation analyzing. Results showed that the increasing temperature and humidity during the oxidation were evidenced to promote the release of Cs compounds, and the chemical speciation of released compounds was prominently dependent on oxidizing temperature. When the oxidation temperature was around 700 °C, Cs2O was predominately released with the morphology of spherical particles. When the oxidation temperature was increased to 900 °C, the predominant releasing composition turned to Cs2CO3 with morphologies of both spherical particles and nanorod-like shapes. The different predominant releasing chemical compositions at different temperatures mainly resulted from the kinetics of graphite oxidization and temperature synergistically. These findings will provide theoretical foundations to the environmental evaluation of i-graphite thermal oxidation.

Carbonate of strontium anchored at the edge of graphite

With the continuous growth of the fourth-generation nuclear reactor, the oxidation of irradiated graphite (i-graphite) has become a great concern worldwide in recent years, not only for the far-reaching significance in the treatment of i-graphite waste, but also for the safety evaluation of the reactor. Fission products (FPs) and activation of impurities are generated in i-graphite during reactor operation and inevitably released as radioactive dust when i-graphite is oxidized. The investigation of the binding mechanism between nuclides and graphite dust is vital for further separating procedures of off-gas decontamination. Herein, the mechanism of FP release during the oxidation of High-Temperature Gas-Cooled Reactors’ matrix graphite was explored with Strontium (Sr) as an example. The Sr-loaded matrix graphite was prepared by impregnation and characterized by secondary ion mass spectrometry (SIMS), X-ray powder diffraction (XRD), scanning electron microscope-energy dispersive spectrometer (SEM-EDS), and Raman spectra. The released Sr chemical speciation was extrapolated as SrCO3 through the investigation of the temperature and air flow rate effects on oxidation products. Additionally, the main morphology of the released fragment was identified as SrCO3 clusters anchored at the edge of graphite. The ab initio calculations revealed the mechanism for the anchoring of SrCO3 at the edge of graphite via adsorption. These findings provide a novel understanding of Sr-releasing behavior when i-graphite is oxidized.

Applicability of abrasive waterjet cutting to irradiated graphite decommissioning

Characterization, dismantling and pre-disposal management of irradiated graphite (i-graphite) have an important role in safe decommissioning of several nuclear facilities which used this material as moderator and reflector. In addition to common radiation protection issues, easily volatizing long-lived radionuclides and stored Wigner energy could be released during imprudent retrieval and processing of i-graphite. With this regard, among all cutting technologies, abrasive waterjet (AWJ) can successfully achieve all of the thermo-mechanical and radiation protection objectives. In this work, factorial experiments were designed and systematically conducted to characterize the AWJ processing parameters and the machining capability. Moreover, the limitation of dust production and secondary waste generation has been addressed since they are important aspects for radiation protection and radioactive waste management.The promising results obtained on non-irradiated nuclear graphite blocks demonstrate the applicability of AWJ as a valid technology for optimizing the retrieval, storage, and disposal of such radioactive waste. These activities would benefit from the points of view of safety, management, and costs.

High temperature annealing of irradiated nuclear grade graphite

Previous work has shown that the material properties of nuclear grade graphite are substantially affected by the atomic and microstructural changes that occur during neutron-irradiation. The parameters that play a role in these changes are irradiation dose, temperature, graphite composition and the initial microstructure of the graphite. Understanding the details of how these material changes occur and exactly what the changes consists of will enable the prediction of the material property changes as a function of these variables. Other work has shown these atomic level and microstructural changes can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. In this work, experiments were carried out to investigate how the properties of irradiated graphite recover when heated. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron-irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young's modulus and electrical resistivity between annealing temperatures of 500 °C and 2380 °C. Graphite grades NBG-18, IG-110 and PCEA are considered that were both stressed and unstressed during irradiation.

Generic method to assess transmutation feasibility for nuclear waste treatment and application to irradiated graphite

Graphite-moderated nuclear reactors have already produced more than 250,000 tons of irradiated nuclear graphite, or i-graphite, world-wide. The sustainability of this technology relies on the end-of-life management of its moderator, which is activated into a long-lived, low or intermediate-level nuclear waste, by neutron fluxes, during operating time. In particular, carbon-14 is created. Nuclear transmutation, enabled by laser-driven particle acceleration, has been envisioned as a potential novel treatment scheme for long-lived nuclear waste. By triggering controlled nuclear reactions with energetic particles, long-lived radio-nuclides could be transformed into short-lived or stable isotopes. Such a system could treat the carbon-14 nuclei trapped within the i-graphite matrix, an isotope which is difficult to isolate by other means. This work performs a quantitative preliminary study of this transmutation scheme, in order to assess its feasibility at an industrial scale. The method used can be transposed to assess any transmutation scheme using a beam of particles directly sent on the material to be treated. First, a nuclear interaction channel which transmutes carbon-14 nuclei without creating new long-lived radio-nuclides is identified. It consists in the choice of a type of particle, among which protons, γ photons and neutrons can all be accelerated by laser–matter interaction; and it is completed by the adequate energy at which this particle must be sent on i-graphite. To that end, the nuclear cross-sections of carbon-12, carbon-13 and carbon-14 are reviewed, neglecting other impurities in i-graphite. Then, based on the interaction channel identification, the energy cost of this scheme is estimated. Protons between 1 and 5MeV make it possible to transmute carbon-14 without creating any new long-lived activity. However, our result show that, even in this favourable reaction channel, the transmutation energy cost is too high for an i-graphite transmutation scheme to be industrially feasible.

Eco-Friendly Synthesis of Wrinkle-Free Ultra-Flat Multi-Layer Graphene by Femtosecond Laser Irradiation of Graphite Under Ambient Conditions

Here, we report a single-step, high-yield technique for producing planar sheets of wrinkle-free multi-layer graphene (MLG) under ambient conditions by ablation of graphite powder dispersed in ethanol using a femtosecond laser. The parameters for the synthesis of MLG have been optimized and established. The number of layers and quality of MLG produced at several laser fluences (39.4 J/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , 98.72 J/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , 197.45 J/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , 296.16 J/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> ) was established based on detailed characterization using field emission scanning electron microscopy (FESEM), high-resolution transmission electron microscopy (HRTEM) with selected area electron diffraction (SAED), atomic force microscopy (AFM), Raman spectroscopy, X-ray photoelectron spectroscopy (XPS), and Fourier transform infrared spectroscopy (FTIR). Wrinkle-free, ultra-flat crystalline MLG sheets with 2-12 layers were produced in each case with an average lateral dimension of ∼1-3 μm. The associated oxy- functional groups were characterized and quantified using FTIR and XPS. The highest percentage of sp <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> C atoms and C/O ratio for MLG was achievedat the lowest laser fluence. The higher the laser fluence higher the defects produced in the MLG sheets. The technique produces zero chemical waste, has a very high yield of ∼ 95%, and is promising for bulk-scale production of MLG.

Estimating the dimensional change behavior of nuclear graphite with heavy-ion irradiation-induced bending

Two promising reactor Generation IV designs employ nuclear graphite as their moderator and reflector. To this end, new graphite grades are being developed, which will require their suitability for reactor design to be established. Understanding graphite irradiation behavior, especially dimensional change, is critical when selecting a particular nuclear graphite grade. This is usually established by fast neutron irradiation experiments in a material test reactor (MTR). However, due to the limited accessibility to the MTR irradiation facility, cost, and the time required for such experiments to be carried out, the development of new nuclear graphite grades is restrictive. Here, a novel approach is proposed to enable suitable grades to be selected from a number of candidates before an MTR program is implemented by estimating the irradiation-induced dimensional change behaviors of several nuclear graphite grades based on the bending of graphite foils under ion irradiation. The method is capable of efficiently distinguishing the most dimensionally stable graphite grades in the laboratory without the need to irradiate many samples in an MTR program.

Microstructural changes in nuclear graphite induced by thermal annealing

Irradiation-induced property change in nuclear graphite is particularly important when considering the in-service lifetimes of graphite components. In other works, it has been shown that annealing of irradiated graphite above the irradiation temperature may heal atomic-level defects and thus reverse some amount of physical property change. In this work, virgin nuclear graphite IG-110 was annealed at 2500 °C to observe any microstructural changes due solely to high-temperature thermal annealing. The results shown in this study suggest that fullerene-like defects, which are not observed in as-prepared nuclear graphites, will arise due to thermal annealing near graphitization temperature. Consequently, such defects may directly contribute to non -recoverable physical property change which has previously been observed in irradiated nuclear graphites. • Non-recoverable physical property changes to nuclear graphite. • High-temperature thermal annealing of nuclear graphite. • Microstructural changes induced solely by high-temperature thermal annealing. • Fullerene-like defects characterized via transmission electron microscopy.

Experimental Investigation on Compressive Strength, Ultrasonic Characteristic and Cracks Distribution of Granite Rock Irradiated by a Moving Laser Beam

Efficient fracturing is the key issue for the exploitation of geothermal energy in a Hot Dry Rock reservoir. By using the laser irradiation cracking method, this study investigates the changes in uniaxial compressive strength, ultrasonic characteristics and crack distributions of granite specimens by applying a laser beam under various irradiation conditions, including different powers, diameters and moving speeds of the laser beam. The results indicate that the uniaxial compressive strength is considerably dependent on the power, diameter and moving speed of the laser beam. The ultrasonic-wave velocity and amplitude of the first wave both increase with a decreased laser power, increased diameter or moving speed of the laser beam. The wave form of irradiated graphite is flattened by laser irradiation comparing with that of the original specimen without laser irradiation. The crack angle and the ratio of the cracked area at both ends are also related to the irradiation parameters. The interior cracks are observed to be well-developed around the bottom of the grooving kerf generated by the laser beam. The results indicate that laser irradiation is a new economical and practical method that can efficiently fracture graphite.