Neutron-irradiated Graphite Research Articles

The structure evolution in neutron-Irradiated nuclear graphite and post-annealing

The neutron irradiated microstructure and performance of nuclear graphite as a neutron moderator and structural materials are directly related to the safety and longevity of reactors. Nuclear-grade isotropic graphite IG-110U and ETP-10 were irradiated at 200 °C with low-dose (1.38 × 1023 n/m2, 0.02dpa) and high-dose (1.92 × 1024 n/m2, 0.25dpa) neutron and high-temperature annealed at high temperatures up to 1400 °C, with the whole process analyzed using X-ray diffractometer and CARBONX program. After neutron irradiation of both doses, the c-axis lattice constant <c> became significantly larger, while the a-axis lattice constant <a> only changed a little. Although the crystallite size along c-axis Lc and the number of internal atom layers M both increased significantly, the crystallite size along a-axis La changed slightly for the low-dose samples. However, all three parameters decreased significantly after high-dose irradiation. During the subsequent high-temperature annealing, there were "recovery" points where both the microscopic parameters and macroscopic dimensions were restored to the initial unirradiated value and kept mostly constant. The movement and recombination of interstitial atoms, vacancies and their clusters played a major role in this annealing process. However, the crystallite sizes of two graphites irradiated with high-dose neutrons did not reached their initial values even with annealing at 1400 °C, mainly due to the unrecoverable defects, such as interstitial loops, pores, and microcracks.

Defect structure classification of neutron-irradiated graphite using supervised machine learning

Molecular dynamics simulations were performed to predict the behavior of graphite atoms under neutron irradiation using large-scale atomic/molecular massively parallel simulator (LAMMPS) package with adaptive intermolecular reactive empirical bond order (AIREBOM) potential. Defect structures of graphite were compared with results from previous studies by means of density functional theory (DFT) calculations. The quantitative relation between primary knock-on atom (PKA) energy and irradiation damage on graphite was calculated.and the effect of PKA direction on the amount of defects is estimated by counting displaced atoms. Defects are classified into four groups: structural defects, energy defects, vacancies, and near-defect structures, where a structural defect is further subdivided into six types by decision tree method which is one of the supervised machine learning techniques.

The Structure Evolution in Neutron-Irradiated Nuclear Graphite and Post-Annealing

The Structure Evolution in Neutron-Irradiated Nuclear Graphite and Post-Annealing

Calculation of Temperature Conditions of a Plasma Sputtering Cell for Decontamination of Nuclear Power Plant Constructions

We are developing a technology for the “dry” deactivation of nuclear power plant (NPP) constructions by ion sputtering of surfaces with microsized radioactive contaminants. Our technology is implemented by plasma discharge in an inert gas medium with mass transfer of sputtered material and its deposition in the diffusion mode on the anode substrate. In our technology, unlike traditional radiochemical methods, radionuclides do not transfer into liquid radioactive waste (LRW), but condense in a compact solid form, which makes it possible to use them. Since our technology can be applied both during decommissioning of NPPs (including deactivation of neutron-irradiated nuclear graphite) and during routine operation and scheduled repairs of nuclear reactors, it is possible to extract the necessary isotope concentrate in the required quantities using an intense neutron flux in the NPP. The design of a plasma sputtering cell to remove radionuclides from the irradiated graphite and the NPP construction surfaces involves ignition of a direct current plasma discharge in an inert gas (for example, argon) at a pressure P ~ 0.1–1 atm and control of the temperature conditions of the sputtering material deposition. In this paper, the temperatures of the anode (tantalum collector) and the cathode (sputtered graphite) have been calculated depending on the input power to the argon plasma discharge. Data on the temperatures of the cathode and anode (collector) surfaces make it possible to control the elemental composition of the sputtered atoms and to form nanoscale layers of radionuclide concentrate on collector for use in radiation medicine and new betavoltaic batteries. Thus, the technology is important not only for the deactivation of NPPs but also for the formation of nanoscale layers of beta-active materials.

Assessment of the fracture toughness of neutron-irradiated nuclear graphite by 3D analysis of the crack displacement field

Digital volume correlation of in situ synchrotron X-ray computed tomographs has been used to measure the three-dimensional displacement fields around quasi-static propagating cracks in neutron irradiated and unirradiated graphite in specimens of the double cleavage drilled compression geometry. The crack tip location and crack opening were extracted from the displacement fields using a phase congruency edge detection method as cracks were propagated over ∼5 mm. The cracks propagated in mode I, maintaining a constant crack opening angle that was ∼50% smaller for the irradiated graphite. 3D finite element simulations, using the measured full field displacements as boundary conditions, obtained the critical elastic strain energy release rate for crack propagation by calculation of the domain contour J-integral. When the non-linear properties of unirradiated graphite were considered, the strain energy release rate for propagation was constant (180 ± 22 Jm-2) with increasing crack length. The irradiated graphite (fluence of 19.7 × 1020 neutrons cm−2 or 2.6 dpa, 4% weight loss by radiolytic oxidation) had linear elastic properties, and the strain energy release rate for propagation increased linearly from 118 ± 12Jm-2 to 485 ± 75 Jm-2 with crack length.

Defect topology and annihilation by cooperative movement of atoms in neutron-irradiated graphite

Graphite has been used as neutron moderator or reflector in many nuclear reactors. The irradiation of graphite in a nuclear reactor results in a complex population of defects. Heating of the irradiated graphite at high temperatures results in annihilation of the defects with release of an unusually large energy, called the Wigner energy. From various experiments on highly irradiated graphite samples from CIRUS reactor at Trombay and ab-initio simulations, we have for the first time identified various 2-, 3- and 4-coordinated topological structures in defected graphite, and provided microscopic mechanism of defect annihilation on heating and release of the Wigner energy. The annihilation process involves cascading cooperative movement of atoms in two steps involving an intermediate structure. Our work provides new insights in understanding of the defect topologies and annihilation in graphite which is of considerable importance to wider areas of graphitic materials including graphene and carbon nanotubes.

Local Fluctuations of Transformation to “Amorphous Diamond” from C60 Fullerenes and Neutron-irradiated Graphite under Shock-compression

Local Fluctuations of Transformation to “Amorphous Diamond” from C60 Fullerenes and Neutron-irradiated Graphite under Shock-compression

Vacancy-driven variations in the phonon density of states of fast neutron irradiated nuclear graphite

Research examining radiation damage and its effects in graphite began in the 1940’s during the development of moderated nuclear reactors. Interest in this topic is expanding because of emerging applications associated with fullerenes and carbon nanostructures as well as its long-standing use as a fission moderator. In this work, we report the measurements of the full phonon density of states of irradiated nuclear graphite using inelastic neutron scattering experiments at room temperature. The samples were previously exposed to different levels of neutron doses and irradiation temperatures. The phonon density of states of perfect and defected graphite supercells including different configurations of vacancies and interstitials were calculated using first-principles direct method. Despite high neutron irradiation doses, the scattering measurements show that the induced damage is localized, and the layered structure is preserved. A comparison of the measured phonon densities of states of irradiated samples with those calculated of defected supercells indicates that the main changes observed are mainly attributed to formation of vacancies.

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥800 °C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

Calculation of temperature conditions of a plasma sputtering cell for decontamination of nuclear power plant constructions

We are developing a technology for the “dry” deactivation of nuclear power plant constructions by ion sputtering of surfaces with micro-sized radioactive contaminants, our technology is implemented by plasma discharge in an inert gas medium with mass transfer of sputtered material and its deposition in diffusion mode on the anode substrate. Unlike traditional radiochemical methods, in our technology, radionuclides do not transfer into liquid radioactive waste, but condense in a compact solid form, which makes possible to use them. Весаuse our technology can be applied both during decommissioning nuclear power plants (including deactivation of neutron-irradiated nuclear graphite), and during routine operation and scheduled repairs of nuclear reactors, so it is possible to extract the necessary isotope concentrate in the required quantities, due to intense neutron flux in nuclear power plant. The design of plasma sputtering cell to remove radionuclides from the irradiated graphite and NPP construction surfaces involves ignition of a direct current plasma discharge in an inert gas (for example, argon) at the pressure of P ~ 0.1–1 atm and control of the temperature conditions of the sputtering material deposition. In this paper temperatures of the anode (tantalum collector) and the cathode (sputtered graphite) were calculated depending on the input power to the argon plasma discharge. Data on temperatures of cathode and anode (collector) surfaces make it possible to control the elemental composition of the sputtered atoms and to form nanoscale layers of radionuclide concentrate on collector for using in radiation medicine and new beta-voltaic batteries. Thus, the technology is important not only for the deactivation of nuclear power plants, but also for the formation of nanoscale layers of beta-active materials.

Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite

ABSTRACT The dimensional changes and thermal conductivity with the annealing of fine-grained isotropic graphite IG-110U and ETP-10 irradiated to 0.02 and 0.25 dpa (1.38 x 1023 and 1.92 x 1024 n/m2, E > 1MeV) at a design temperature of <200°C were studied. The irradiated graphite exhibited a small volume expansion and large degradation in thermal conductivity. Post-irradiation annealing experiments were carried out on dimensional changes and thermal conductivity up to 1700°C, and the results were analyzed in terms of changes in the defect concentration of graphite crystals. The rapid recovery of thermal conductivity observed below 200°C in the graphite irradiated to 0.02 dpa was attributed to the annihilation of Frenkel defects, whereas the recovery observed in both dimension and thermal conductivity above 200°C in the graphite irradiated to 0.02 dpa and 0.25 dpa was caused by the annihilation of small interstitial clusters of 4 ± 2 atoms. The role of large clusters of interstitials and vacancies in the changes to smaller dimension than pre-irradiation at high annealing temperatures are discussed. The temperature dependence of stored energy release was estimated from the changes in defect concentration calculated from the recovery of thermal conductivity.

Investigation of kinetic recovery process in low dose neutron-irradiated nuclear graphite by thermal annealing

ABSTRACT To investigate the kinetic recovery process of low dose neutron-irradiated graphite, nuclear-grade isotropic graphite IG-110U and ETP-10 were neutron irradiated using the JMTR up to 1.38 × 1023 n/m2 (En > 1 MeV) at ~473 K. In-situ measurement of macroscopic length was conducted during the isothermal and isochronal annealing process from room temperature up to 1673 K. From room temperature to 773 K for IG-110U, and to 1023 K for ETP-10, macroscopic lengths, lattice parameters, and unit cell volumes of both specimens recovered to their pre-irradiation values, and this recovery process subdivided into two stages. The activation energies of macroscopic volume recovery at 523–673 K and 673–773 K were determined to be ~0.22 eV and ~0.57 eV for IG-110U, respectively; ~0.13 eV and ~2.59 eV at 523–923 K and 923–1023 K for ETP-10, respectively. The migration of not only single interstitials but also interstitials dissociated from submicroscopic interstitial groups along basal planes followed by vacancy-interstitial recombination play a dominant role in the first stage. The second stage is suggested to proceed via the motion of carbon groups along basal planes for IG-110U, and migration of single interstitials along the c-axis for ETP-10. During 773 K or 1023 K up to 1673 K, macroscopic length continuously shrank with decreasing shrinking rate, even with a turnaround to swell at 1173 K for IG-110U.

14C Exposure from Disposal of Radioactive Waste Compared to14C Exposure from Cosmogenic Origin

ABSTRACTThe potential14C (carbon-14, radiocarbon) flux from disposal of14C containing waste into air is compared with the natural14C emanation rate from soil in order to put the14C hazard potential from disposal of this waste in perspective with the14C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The14C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural14C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the14C release rate from this graphite. Moreover, ion exchange—with ingressing anionic species that have a higher affinity than contained anionic14C—is proposed as a release mechanism for these resins.

Evaluation of the primary displacement damage in the neutron irradiated RBMK-1500 graphite

Numerical simulations of the neutron induced primary displacement damage in the RBMK-1500 reactor graphite were performed using the Monte Carlo technique based on the Norgett-Robinson-Torrens (NRT) model. The sequence of the displacement evaluation consists of the reactor core modelling using MCNP6 and GEANT4 codes and output parameters, such as recoil ion and neutron flux distribution used by SRIM2013 and NJOY2016 codes, respectively, to evaluate displacements per atom. The comparisons were made between these codes in order to evaluate the accuracy of each approach. The neutron flux in the spent RBMK-1500 graphite was sufficiently correctly represented by both MCNP6 and GEANT4 models. The comparison between SRIM and GEANT4 codes shows that GEANT4 predicts fewer vacancies by more than 16% than SRIM in the ion energy range from 1 keV to 1 MeV. The neutron energy required to produce recoils due to scattering is around 80 eV. Below this energy displacements occur mostly due to neutron capture reactions. The average displacement rate was found to be around 0.51 displacements per atom per full power year, which corresponds to the 1.36·1014 n/cm2s neutron flux.

Quenchable compressed graphite synthesized from neutron-irradiated highly oriented pyrolytic graphite in high pressure treatment at 1500 °C

The super hard material of “compressed graphite” (CG) has been reported to be formed under compression of graphite at room temperature. However, it returns to graphite under decompression. Neutron-irradiated graphite, on the other hand, is a unique material for the synthesis of a new carbon phase, as reported by the formation of an amorphous diamond by shock compression. Here, we investigate the change of structure of highly oriented pyrolytic graphite (HOPG) irradiated with neutrons to a fluence of 1.4 × 1024 n/m2 under static pressure. The neutron-irradiated HOPG sample was compressed to 15 GPa at room temperature and then the temperature was increased up to 1500 °C. X-ray diffraction, high-resolution transmission electron microscopy on the recovered sample clearly showed the formation of a significant amount of quenchable-CG with ordinary graphite. Formation of hexagonal and cubic diamonds was also confirmed. The effect of irradiation-induced defects on the synthesis of quenchable-CG under high pressure and high temperature treatment was discussed.

Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR

We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13C nuclear spin-lattice relaxation rate by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of below about 10 K can well be described by a thermally activated form, , yielding a singular Zeeman energy of () meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments.

An understanding of lattice strain, defects and disorder in nuclear graphite

In this study, microstructural parameters, such as lattice dimension, micro-strain and dislocation density, of different neutron-irradiated graphite grades have been evaluated using the diffraction profiles of X-ray diffraction (XRD) and the scattering profiles of Raman spectroscopy. Using Generation-IV candidate graphite samples (grade PCEA, GrafTech), subjected to neutron irradiation at 900 °C to 6.6 and 10.2 dpa, and graphite samples of similar grain size and microstructure taken from the core of the British Experimental Pile Zero reactor, which have been irradiated at 100–120 °C to 1.60 dpa, an investigation is presented on the effect of irradiation dose and temperature on the aforementioned microstructural parameters. Using two complementary techniques in Raman spectroscopy and XRD, which produced agreeable results, the average lateral crystallite size of the two graphites tested was found to decrease with increasing exposure to fast-neutron irradiation or at lower irradiation temperatures. Conversely, dislocation densities and micro-strains were found to increase following the same changes in irradiation conditions. Supporting evidence for the microstructural information obtained is provided by direct observations made using high-resolution transmission electron microscopy. These images demonstrate the presence of irradiation-induced prismatic edge dislocations as well as features that indicate the presence of basal dislocations. They also provide supporting evidence for the progressive deterioration of the graphitic planes via damage mechanisms as proposed in the literature.

Micro to nanostructural observations in neutron irradiated nuclear graphites PCEA and PCIB

The neutron irradiation-induced structural changes in nuclear grade graphites PCEA and PCIB were investigated using scanning electron microscopy (SEM), X-ray diffraction (XRD), Raman spectroscopy, transmission electron microscopy (TEM), selected area electron diffraction (SAED) and electron energy loss spectroscopy (EELS). The graphite samples were irradiated at the Advanced Test Reactor at the Idaho National Laboratory. Received doses ranged from 1.5 to 6.8 displacements per atom and irradiation temperatures varied between 350 °C and 670 °C.XRD and Raman measurements provided evidence for irradiation induced crystallite fragmentation, with crystallite sizes reduced by 39–55%. Analysis of TEM images was used to quantify fringe length, tortuosity, and relative misorientation of planes, and indicated that neutron irradiation induced basal plane fragmentation and curvature. EELS was used to quantify the proportion of sp2 bonding and specimen density; a slight reduction in planar-sp2 content (due to the buckling basal planes and the introduction of non-six-membered rings) agreed with the observations from TEM.

Precision dilatometer analysis of neutron-irradiated nuclear graphite recovery process up to 1673 K

ABSTRACTSpecimens of two kinds of isotropic nuclear graphite, IG-110U and ETP-10, were neutron-irradiated at fluence of 1.92 × 1024 n/m2 (E > 1.0 MeV) at 473 K. The recoveries of the macroscopic lengths of these specimens during isothermal and isochronal annealing at temperatures of up to 1673 K were investigated in a step-wise manner by using a precision dilatometer. The macroscopic lengths after isochronal annealing for 6 h at each temperature decreased gradually as the temperature was increased to 1673 K. The recovery trends of the c-axis and a-axis lattice parameters differed from one another, and from the macroscopic length recovery trends. For the IG-110U specimen, the activation energies (Ea) of macroscopic volume recovery corresponding to annealing at 523–773, 773–923, 923–1073, and 1073–1173 K were found to be 0.15, 0.34, 0.73, and 2.59 eV, respectively. For the ETP-10 specimen, the Ea corresponding to 523–923, 923–1223, and 1223–1373 K were determined to be 0.15, 0.46, and 2.19 eV, respectively. These results indicate that both graphite specimens underwent three or four stages of macroscopic length recovery between 523 K and the annealing temperatures at which their initial lengths were recovered. It is suggested that during the first stage recovery proceeded via the migration of single interstitials along the basal plane and the resulting V-I recombination. In the middle stages, recovery occurred due to the migration of interstitial groups such as C2 along the basal plane, while in the last stage, it proceeded via through-layer migration of interstitials or migration of single vacancies.

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ∼400 °C up to 9.3 × 1025 n/m2 (E > 0.1 MeV). Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free conditions. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significant increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3–40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.