Nuclear Graphite Research Articles

Molten salt synthesis and antioxidation property of SiC‐coated mesocarbon microbeads

AbstractMesocarbon microbead (MCMB) is a prospective candidate as the raw material for nuclear graphite. However, the poor resistance to high‐temperature oxidation limits its application. Herein, a dense SiC coating was prepared by molten salt synthesis on the surface of MCMB to improve its antioxidation performance. The effect of molten salt synthesis reaction time on the phase composition, microstructure, and antioxidation performance of the SiC‐coated MCMB particles was investigated. A theoretical model was established to explain the SiC coating growth rule well, in conformity with the carbon vacancy diffusion mechanism in SiC coating. The SiC coating synthesized for 7 h with the thickness of .385 µm remarkably promoted the high‐temperature antioxidation property of MCMB. The kinetics analysis indicated that the SiC coating obstructed the oxygen diffusion effectively during the oxidation process. The as‐fabricated SiC‐coated MCMBs with good oxidation resistance show great promise for application in nuclear industry and other antioxidative fields.

Catalyzed oxidation of IG-110 nuclear graphite by simulated fission products Ag and Pd nanoparticles

To evaluate the stability of nuclear materials in high temperature gas reactors under air ingress conditions, catalytic oxidation of IG-110 graphite by two simulated fission products, metallic Pd and Ag, was studied in oxidative atmosphere and at temperatures up to 1000 °C using an integrated furnace, mass spectroscopy and infrared spectroscopy system. Transmission electron microscopy and X-ray diffraction studies show that Pd and Ag nanoparticles were successfully introduced onto powdery IG-110 graphite through an impregnation and subsequent heat-treatment process. The combined mass spectroscopy and infrared spectroscopy methods allow simultaneous analysis of two gaseous products, CO and CO2, and separate measurements of activation energy for their formation reactions. It was found that the introduction of Pd or Ag to IG-110 graphite substantially catalyzed the oxidation of graphite, characteristic of decreased onset temperatures for the oxidation of graphite. Moreover, the catalytic effects by Pd and Ag are considerably different based on measured concentration ratios of CO2 to CO as a function of oxidation temperatures. Ag makes the graphite oxidation commence at approximately 400 °C with CO2 being the dominant product. In contrast, Pd significantly increases the concentration ratio of CO2 to CO at temperatures higher than approximately 690 °C, although it decreases the onset temperature for the oxidation reaction to around 525 °C. To understand the catalytic difference, the mechanism of the graphite oxidation is discussed based on the changes of surface oxygen species on Ag and Pd.

Investigation On Oxidation Behavior of Nuclear Graphite IG-110 At Elevated Temperature

IG-110 is one of nuclear graphite types for high temperature gas cooled reactor. One of the issues of the graphite in the reactor system is the interaction with air or/and water which can reduce the integrity of the material. In this study, the sample of IG-110 has been tested at high temperature in an air environment. The oxidation test was done using a magnetic suspension balance to analyze the changes in weight with direct monitoring. The test was carried out at a temperature of 630°C. Morphological and microstructural analysis were observed and tested using scanning electron microscope (SEM) and X-ray diffractometer (XRD). The oxidation rate of the sample was relatively low. Rougher sheets shape, larger pores, and slight cracks were observed for the tested sample which indicate the traces of oxidation of the samples with no any critical damage occurred in the sample. Lattice parameters, density, crystal size, and d-space of the sample after oxidation test slightly change, which show the reversible characteristic of the sample from the temperature viewpoint.

Effects of tensile and compressive stresses on damage evolution law of nuclear graphite

Nuclear graphite, a quasi-brittle material containing microdefects, is essential for constructing the neutron moderators, reflectors, and cores in very-high-temperature gas-cooled reactors. Such reactors are prone to damage during loading, resulting in the failure of nuclear graphite materials or structures and possibly affecting the integrity of the entire core structure. Therefore, the damage evolution law of nuclear graphite should be studied while assessing the safety of the core. In addition, considering the different mechanical characteristics of quasi-brittle materials under tensile and compressive stresses as well as the complex stress states of quasi-brittle nuclear graphite structures in the core, the effects of stress states on the damage evolution law of nuclear graphite should be unveiled. These aspects are difficult to evaluate owing to the complexity of damage characterisation and measurement throughout the test loading of nuclear graphite under different stress states. Therefore, we propose damage measurement tests of nuclear graphite under unidirectional tensile and compressive stress states based on digital image correlation to obtain damage evolution laws under different stress states. By analysing the effects of different stress states on the damage evolution laws of nuclear graphite, we found that the damage properties under unidirectional tensile and compressive stresses show variations and that damage more likely occurs under tensile stress.

Topography changes and microstructural evolution of nuclear graphite (IG-110) induced by Xe26+ IRRADIATION

Topography changes and microstructural evolution of nuclear graphite (IG-110) induced by Xe26+ IRRADIATION

Topography changes and microstructural evolution of nuclear graphite (IG-110) induced by Xe26+ irradiation

AbstractThe microstructure of nuclear graphite, a key material in nuclear reactors, is affected by the high-flux irradiation. The damage to the graphite by irradiation is important for reactor safety. To understand the damage of nuclear graphite by irradiation, IG-110 nuclear graphite, a representative nuclear graphite, was chosen to investigate the change in morphology and microstructure caused by 7 MeV Xe26+ irradiation with peak damage doses of 0.1-5.0 displacements per atom (dpa) for samples of a size of 40.0 mm×40.0 mm×2.0 mm. The topography and microstructure of IG-110 were characterized by SEM, AFM, grazing incidence XRD, Raman spectroscopy and nano-indentation. Results indicate that after 7 MeV Xe26+ irradiation at a dose of 0. 11 dpa, a ridge-like structure appears on the surface of the IG-110 graphite, mainly in the binder region, and the surface roughness increases slightly. With a further increase of the irradiation dose, the ridge-like structure also appears in the filler region. At a dose of 0.55 dpa, pore shrinkage increases accompanied by pore closure, and the surface roughness also increases. The changes in topography and microstructure caused by irradiation are attributed to the expansion of graphite along the c-axis direction. With increasing irradiation dose the defect density and the degree of in-plane disorder in the graphene sheets increases, while the modulus and hardness of the graphite first increase and then decrease. Their increase is caused by dislocation pinning and closure of fine pores, while their decrease is attributed to an increase in porosity and the generation of an amorphous structure.

Microstructural characterization of the CGB graphite grade from the molten salt reactor experiment

In the 1960s, the Molten Salt Reactor Experiment demonstrated the feasibility of molten salt reactors for civil applications. The Molten Salt Reactor Experiment used a nuclear graphite grade known as CGB in the core as the fast neutron moderator. For application in this purpose, CGB graphite underwent additional impregnation steps that are not typically used in conventional nuclear graphites. These additional impregnations significantly lowered the permeability and reduced the size of pore openings in CGB graphite, for the purpose of inhibiting molten salt ingression into the graphite. Although several publications report the mechanical properties and irradiation response of CGB graphite, little information has been published about the microstructure or sealant used for this graphite grade. Here the results of multiple advanced microscopy techniques are presented with the objective to investigate the unique microstructure and sealing technology of legacy material from the Molten Salt Reactor Experiment so it can be potentially readopted for modern reactors. The microstructural characterization shows that some pores were fully sealed by the impregnation whereas other regions contained foam-like materials that partially filled the void content. This manuscript also investigates and documents the nature and possible source of the sealant material. The pore sealing methods used for CGB are discussed and compared with other common technologies used to reduce the impregnation of molten salts into graphite components. Knowledge of CGB graphite's microstructure can enable development of new pore infiltration technologies to improve in operando behavior advancing the next generation of molten salt reactors.

An analysis of the effects of graphitization temperature on the irradiation dimensional change behavior of graphite

Graphitizations at temperatures greater than 2700 °C are generally preferable for nuclear graphite manufacture in order to obtain higher order of graphitization and lattice density. In 1993, Brocklehurst and Kelly published data showing graphitization temperature played a significant role on the dimensional change behavior (DCB) of pile grade A (PGA) graphite, irradiated at 600 °C with fast neutrons. These data were analyzed using a recently developed regression analysis method to generate derived figure of merits (DFOM). DFOMs are useful tools, which can be used to enable professional engineering judgement (PEJ). The resulting DFOMs reinforce the general premise of ASTM International nuclear graphite specifications, which require graphitization to be performed at a temperature of at least equal to 2700 °C.

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.

Application of ion implantation as a tool to study neutron induced morphological changes in HOPG and RBMK-1500 reactor graphite

Based on the RBMK-1500 (rus.: peaктop бoльшoй мoщнocти кaнaльный, PБMК; reaktor bolshoy moshchnosti kanalnyy, “high-power channel-type reactor”) reactor irradiation conditions, ion implantation was used as a tool to study neutron induced morphological changes in both highly oriented pyrolytic graphite (HOPG) and nuclear grade RBMK graphite. Graphite samples were implanted with 180 keV 14N+ ions at the fluences of 1.0 × 1016 ions/cm−2 and 2.5 × 1016 ions/cm−2. To study temperature effects on both ion migration and structural disorder level in the graphite matrix, the implantation procedures were carried out under different temperature conditions – room temperature (RT) and 500 °C. Subsequently, the distribution profiles of the implanted 14N+ ions were obtained by using secondary ion mass spectroscopy (SIMS) technique, while the microstructural properties of the graphite were evaluated by Raman spectroscopy. Implantation induced primary displacement damage profiles and implanted ion profiles were evaluated theoretically by using GEANT4 10.6 and SRIM-2013 codes. The profiles of implanted nitrogen obtained by SIMS technique were found to be in good agreement with theoretical ones. The surfaces of HOPG samples displayed macroscopic damage in form of fractures after 14N+ ion implantation at 500 °C; however, in case of RBMK graphite, the sample surfaces remained without visually observable changes. The Raman spectra showed an increase of sp3-related content and formation of amorphous carbon in both HOPG and RBMK graphite samples; the latter exhibits higher structural disorder at microscopic level and stronger amorphization than HOPG.

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.

Benchmark of Neutron Thermalization in Graphite Using a Pulsed Slowing-Down-Time Experiment

A benchmark has been developed using a pulsed slowing-down-time experiment to isolate the thermalization process in graphite. The experiment was conducted at the Oak Ridge Electron Linear Accelerator facility at Oak Ridge National Laboratory, and it measured the time spectrum of neutrons leaking from a graphite pile during slowing down and thermalization within graphite. Simulations of the benchmark experiment were performed using the MCNP6.1 Monte Carlo code and the ENDF/B-VII.1 and ENDF/B-VIII.0 cross-section databases. The benchmark provides a time spectrum (i.e., time-dependent counts in a detector) that allows for validation of the graphite thermal scattering libraries (TSLs). The impact on the simulations using a suite of graphite TSLs was compared with the experimental results. Given the density of nuclear graphite, the TSL corresponding to graphite with 30% porosity, as implemented in ENDF/B-VIII.0, was found to most accurately represent the measured time spectrum corresponding to the thermal energy range with an average deviation of ±1.7%.

An investigation of fracture behaviors of NBG-18 nuclear graphite using in situ mechanical testing coupled with micro-CT

An investigation of fracture behaviors of NBG-18 nuclear graphite using in situ mechanical testing coupled with micro-CT

Microstructure and defect evolution of nuclear graphite under temperature-dependent ion irradiation

Nuclear graphite plays a vital role in GEN-IV reactors and is expected to function in the harsh environments such as high temperature and high-dose irradiation. The ion irradiation behaviour of ET-10 nuclear graphite in comparison with highly orientated pyrolytic graphite (HOPG) was reported in this work, to explore the combined effects of ion irradiation and temperature. Ion irradiation experiments were carried out using 190 keV H+ and 1.8 MeV Ar+ at different doses. SRIM simulation, scanning electron microscopy (SEM) and Raman spectroscopy were employed to characterize the microstructural changes and defect evolution of graphite specimens under different irradiation conditions. The vein structure in HOPG graphite and the vertical scale-like structure in ET-10 graphite caused by ion irradiation were observed. The effects of temperature on ion irradiation could be divided into two stages of recovering and saturation according to mutual analysis of SEM and Raman. It was also found that the recovery degree of the half-width at half maximum (HWHM) for the G band was greater than that of ID/IG, probably because the deformed or ruptured graphite layers were difficult to recover at rising temperatures and/or could even be enhanced. HWHM(G), ID/IG, and La was comprehensively discussed, giving an overall perspective of defect evolution and crystallite size changes under temperature-dependant ion irradiation.

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Physical, electrical, thermal properties

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Physical, electrical, thermal properties

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.

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Mechanical properties

Programs at Oak Ridge National Laboratory are investigating the effects of neutron irradiation in the High Flux Isotope Reactor (HFIR) on properties of fine-grain isotropic nuclear graphites. Specimens were irradiated in the HFIR flux trap, which requires using specimens that are smaller than the sizes recommended by ASTM standards. This work has investigated whether subsized specimens can be used to reliably predict strength behavior. The results show that (1) tensile strength was loosely coupled to specimen volume, (2) the compression specimens had a slightly higher strength for larger volumes, (3) the uniaxial flexural strength was nearly independent of specimen effective area for one orientation whereas the other orientation showed a slight strength–size scaling Weibull behavior, and (4) equibiaxial flexural strength followed traditional strength–size scaling Weibull behavior. These results suggest that fine-grain graphite is more similar to ceramics than to medium- and large-grained graphite with respect to both the strength-limiting flaws (Griffith theory) and the effect of machining on the strength behavior. This work shows that using specimens that are smaller than ASTM recommendations are suitable for investigating the strength changes caused by neutron irradiation because the primary interest is the relative change in the strength distribution parameters rather than their absolute values.

Catalyzed oxidation of nuclear graphite by simulated fission products Sr, Eu, and I

The influence of three fission products Sr, Eu, and I on the oxidation of IG-110 nuclear graphite was studied in the temperature range of 400 to 1000 °C. Sr and Eu were introduced as chlorides, and I was introduced as NaI. The temperature dependence of both CO2 and CO production during the graphite oxidation measured with mass spectroscopy and infrared spectrometry shows that the introduction of these three compounds to graphite significantly decreases the onset temperature for the oxidation of graphite. Among the three compounds, NaI is the most active towards the oxidation reaction, characterized by a significant decrease of the onset temperature from approximately 650 to 400 °C before and after its introduction to graphite. Separate measurements of CO2 and CO concentration at varying temperatures enable the calculation of the activation energy for the formation of CO2 and CO. The activation energies for the oxidation of pure and fission product-impregnated graphite samples decrease in the following order: standard IG-110 graphite, EuCl3-impregnated IG-110, SrCl2-impregnated IG-110, and NaI-impregnated IG-110. This trend indicates that the three compounds catalyze the oxidation of graphite at temperatures relevant to the operation of high-temperature gas-cooled reactors. Furthermore, it is found that the three compounds can also affect the molar ratio of reaction products CO2 and CO, and the rates of the graphite oxidation. At temperatures higher than about 850 °C, the impregnated samples exhibit lower CO2: CO ratios than the pure graphite. Different from EuCl3 and NaI, the introduction of SrCl2 decreases the graphite oxidation rates at temperatures higher than about 770 °C. Their catalytic mechanism can be understood based on a redox cycle of the intermediate active species, promoting the dissociation of molecular oxygen and transfer to the carbon.

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.

A multiscale modelling of polycrystalline nuclear graphite under effect of microcracks and incomplete-graphitized regions

A multiscale modelling of polycrystalline nuclear graphite under effect of microcracks and incomplete-graphitized regions