Neutron-irradiated Graphite Research Articles

Accumulation of thermal resistance in neutron irradiated graphite materials

A nuclear graphite, H451, and two high thermal conductivity graphite composites have been irradiated in the temperature range of 310–710 °C in the high flux isotope reactor and their thermal conductivities monitored in situ. Data were measured continuously up to a fast neutron dose of approximately 1 × 10 25 n/m 2 ( E > 0.1 MeV). Data are interpreted in terms of the added thermal resistance and materials compared on this basis. Following this analysis it is shown that for the three materials studied, which have significantly different initial thermal conductivity values, the accumulation of thermal resistance is greater for the materials with lower initial thermal conductivity. Given that vacancies dominate phonon scattering at these irradiation temperatures and dose levels, these data clearly indicate that materials of higher perfection have a slower rate of stable vacancy accumulation during irradiation.

Hydrogen absorption into neutron-irradiated graphite and estimation of the trapping effect

Bulk hydrogen retention and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various conditions. Two kinds of hydrogen trapping sites may exist and be additionally produced during irradiation: interstitial cluster loop edge sites (trap 1) and carbon dangling bonds at edge surfaces of crystallites (trap 2). Neutron irradiation preferably creates trap 2 sites at lower fluences and trap 1 sites at a higher fluence. Trap 2 tends to be annealed out at high temperatures, although trap 1 is hardly decreased even at 1873 K. The activation energy of hydrogen diffusion is found to be increased from 1.04 to 1.60 eV by neutron irradiation.

Radiation-Induced Amorphization

### History and applications The earliest description of a change in mineral property caused by radiation-damage was by Jons Jacob Berzelius in 1814. Berzelius, a Swedish physician and mineral chemist, was certainly one of the most eminent scientists of his century, discovering the elements cerium, selenium and thorium. He discovered that some U- and Th-bearing minerals glowed on moderate heating, releasing, sometimes violently, large amounts of energy. This “pyrognomic” behavior was first observed in gadolinite, (REE)2FeBe2Si2O10, as it released stored energy. Today, this same type of catastrophic energy release is a significant concern in reactor safety where rapid energy release from neutron-irradiated graphite can lead to a rise in temperature and the rupture of fuel elements. In 1893, Brogger defined “metamikte” in an encyclopedia entry for amorphous materials. Metamict minerals were believed to have been previously crystalline, as judged by well-formed crystal faces, but had a characteristic glass-like, conchoidal fracture and were optically isotropic. Prior to the discovery of radioactivity by Becquerel in 1896, the metamict state was not recognized as a radiation-induced transformation. Hamberg (1914) was the first to suggest that metamictization is a radiation-induced, periodic-to-aperiodic transformation caused by α-particles that originate from the constituent radionuclides in the uranium and thorium decay-series. A detailed summary of the history of studies of radiation effects in minerals can be found in Ewing (1994). In the 1950s, interest in metamict minerals was revived. Adolf Pabst rescued the term “metamict” from obscurity in his presidential address to the Mineralogical Society of America in 1951 (Pabst 1952). However, mineralogists paid limited attention to nuclear effects in minerals or the general importance of radiation effects in modifying the properties of materials. By the early 1940s, E.P. Wigner already had anticipated that the intense neutron flux in the Hanford Production reactors …

Stable Vacancy Clusters in Neutron-Irradiated Graphite: Evidence for Aggregations with a Magic Number

By combining positron-annihilation experiments and first-principles calculations, abundant planar ${V}_{6}$ rings are identified in heavily neutron-irradiated graphite. The calculations show that the ${V}_{6}$ ring is particularly stable. Annealing experiments exhibit that the ${V}_{6}$ ring survives up to $1500\ifmmode^\circ\else\textdegree\fi{}\mathrm{C}$. These results suggest vacancies aggregating with the magic number in graphite.

Damaging Process of Graphite – A New Model and its Impact on Degradation of Materials Performance

The most widely accepted model for development of defect structure in neutron irradiated graphite has been such that following the first production of a pair of an interstitial and vacancy, di-interstitials and vacancies are formed and their subsequent growth would result in the production of an interstitial plane or loop in-between the basal planes and vacancy clusters, respectively, which could cause the loss of thermal conductivity and dimensional change.Recently we have claimed that the formation of vacancy clusters and growth of the interstitial planes are not necessarily a unique interpretation of the damaging process. Instead, the damaging process is described by orientational disordering within the basal planes, i.e. fragmentation into small crystallites and rotation of their crystalline axes, change of stacking order and elongation of the interplanar spacing. The orientational disordering within the basal planes proceeds coordinately over a few layers with their layered correlation maintained. This process accompanies changes in bonding nature producing 5 member- and 7 member-atomic rings as appeared in fullerenes. This is so to speak “self-restoring or reconstruction” to maintain resonance bonds as strict as possible without the formation of dangling bonds.This paper reviews irradiation effects in graphite such as increase of hydrogen retention, loss of thermal conductivity and dimensional change on the bases of our new model, taking account of the changes of the bonding nature in irradiated graphite.

Change in physical properties of high density isotropic graphites irradiated in the “JOYO” fast reactor

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor “JOYO” to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

In situ AES study of damaging process of graphite surface under energetic ion bombardment

In order to investigate the behavior of hydrogen implanted in graphite and the role of hydrogen both in damaging and annealing processes, we have performed an in situ AES analysis of graphite under hydrogen or inert gas (He or Ar) ion bombardment as well as under annealing. The results are summarized as follows. Damaging processes introduced by either H, He or Ar ion bombardment are found to be quite similar to each other, depending only on the lattice displacement effect, and both the electronic excitation effect and the chemical effect of implanted hydrogen are hardly seen in C-KLL AES spectra as well as laser Raman spectra. Considering the large enhancement of hydrogen retention in neutron irradiated graphites, it is very likely that hydrogen is trapped at preexisting damage sites, though it is hardly possible to distinguish a CH bond and a CC single bond in the present AES analysis. A certain difference does appear in the annealing between hydrogen implanted graphite and a helium implanted one. These results make us to conclude that hydrogen is trapped with a chemical bond to carbon atoms even at higher temperatures.

Radiation damage of graphite: degradation of material parameters and defect structures

On the basis of our previous work, we have discussed the relation between damage structure and the degradation of the material parameters of neutron-irradiated graphites. The defects produced in a basal plane and/or in between the basal planes (in-plane or two-dimensional defects), which are appreciable in the early stage of irradiation, seem to play a critically important role both in the reduction in the thermal conductivity and increase in the hydrogen retention. They also cause a dimensional change through increase in lattice spacing between the basal planes of graphite. Although the in-plane defects are rather easily annealed out, there seems to be no way to avoid their production under neutron irradiation, particularly at low temperatures. After heavy irradiation, the defects grow into three-dimensional clusters, probably accompanying some sp 2-to-sp 3 transition. They play an important role in volume expansion and result in complete loss of the layered structure of graphite (amorphization), which is very difficult to anneal. Considering the annealing behaviors of the thermal conductivity, lattice constant and electrical resistivity, we propose a new model based on the sp 2-to-sp 3 transition that can explain the observed effect for both damage and annealing processes without any contradiction.

The effect of neutron irradiation on the trapping of tritium in carbon-based materials

Carbon-based materials are considered for protection of plasma-facing components in the next step fusion device. To investigate the effects of neutron damage on the tritium behaviour an experimental study on the tritium retention of various neutron-irradiated graphites and carbon/carbon fibre composites was started. The irradiation dose of the specimens ranges from 10 −3 to 3.5 dpa g and the irradiation temperature from 390 to 1500°C. A comparison of tritium retention in pre- and jpostirradiated carbon-based materials as a function of the sample temperature is reported in this paper and the results are discussed. The first results indicate that the retention of tritium is higher in irradiated graphite than in unirradiated graphite and depends largely on the density and microstructure. The retention is also influenced by the tritium loading temperature. Graphite of type S 1611, irradiated at 400 and 600°C up to a damage of 0.1 dpa g, retained about two times more tritium than the unirradiated material.

Trapping and detrapping of hydrogen in carbon-based materials exposed to hydrogen gas

Measurements of hydrogen solubility have been performed for several unirradiated and neutron-irradiated graphite (and CFC) samples at temperatures between 700 and 1050°C under a ⋍ 10 kPa hydrogen atmosphere. The hydrogen dissolution process has been studied and it is discussed here. The values of hydrogen solubility vary substantially among the samples by up to a factor of about 16. A strong correlation has been observed between the values of hydrogen solubility and the degrees of graphitization determined by the X-ray diffraction technique. The relation can be extended even for the neutron-irradiated samples. Hydrogen dissolution into graphite can be explained by the trapping of hydrogen at defect sites (e.g. dangling carbon bonds) considering an equilibrium reaction between hydrogen molecules and the trapping sites. The migration of hydrogen in graphite is speculated to result from a sequence of detrapping and retrapping events with high-energy activation processes.

A study of thermal neutron irradiated graphite by positron annihilation

A study of thermal neutron irradiated graphite by positron annihilation

On the mechanism of dimensional change of neutron irradiated graphite

In order to investigate the mechanism of large volume expansion of graphite due to neutron irradiation we have made in situ observation of damage structure of graphite with a high resolution transmission electron microscope (HRTEM). Both bending and randomizing the orientation of broken basal planes are proposed to be the origin for the large volume expansion of the irradiated graphite. It is also found that graphite easily loses its lattice ordering in the basal planes as demonstrated by halo rings in the (101̄0) diffraction pattern, while hardly losing its layered structure, i.e., (002) spots are retained.

Hydrogen solubility and diffusivity in neutron-irradiated graphite

Hydrogen solubility and diffusivity in neutron-irradiated graphite

Hydrogen solubility and diffusivity in neutron-irradiated graphite

Hydrogen solubility measurements and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various fluences up to 5.4×10 24 n/m 2 . The absorption of hydrogen could be expressed as a diffusion-controlled process. The rate constant of hydrogen absorption was different from that for desorption. This difference may be ascribed to the effects of trapping sites in graphite. After neutron irradiation at 1.9×10 24 n/m 2 (~ 0.2 dpa), the hydrogen solubility was 20–50 times larger than that of unirradiated samples. The increase of hydrogen solubility was saturated above the damage level of ~ 0.3 dpa. The diffusivity of hydrogen was decreased by neutron irradiation up to 1.9×10 24 n/m 2 , and then increased above this fluence. This behavior can be ascribed to the production of the trapping sites for hydrogen, and the elongation of the distance between the basal planes by neutron irradiation.

Interstitial string model for defective graphites

The carbon atoms in defective graphites are arranged mainly in layers, with a spacing that is sample dependent and larger than that in high-perfection graphites. The generally accepted explanation for this variation is that the larger separations in more disordered graphite arise from poorer mutual orientation of the graphite layers (“turbostratic disorder”). Computer models of interstitials in supercells show that strings of interstitials along the hexagonal axis are unusually stable and might better explain the spacings observed and other properties, certainly for neutron-irradiated graphite, but possibly also for other graphites.

Reflection and Photoemission Studies of Neutron-Irradiated Graphite

Neutron-irradiated graphites were studied by reflectivity and photoemission (UPS, ARUPS, XPS) measurements. The π-band reflectivity peak of graphite, located at 5 eV, changed significantly and a small absorption band ascribed to vacancies produced by neutron bombardment was found to grow around 3 eV. Modification of the valence band by neutron irradiation was studied by ARUPS. The π-valence band shifts to lower binding energy towards the Fermi level and its band width becomes smaller. These results were also confirmed by the optical joint density of states obtained from K-K analysis of the reflectivity.

Tritium behavior in neutron irradiated lithium graphite intercalation compound (LiC 6)

Tritium release behavior from neutron irradiated LiC 6 was investigated by heating up to 1000 °C. Released chemical species were found to be mainly molecular tritiated hydrogen (HT or T 2), accompanied by a small portion of CH 3T and HTO. Regarding tritium release kinetics, two peaks in tritium release were observed at 200–300 °C and 500–600 °C. The latter could be analyzed by diffusion-controlling kinetics, while the former by both diffusion-controlling and 1st order reaction kinetics. Judging from the values of diffusion coefficients and of activation energy, it is concluded that the tritium species behave with some interaction to lithium ions in LiC 6.

Degradations of thermal shock resistance and fracture toughness of neutron irradiated reactor graphite.

This paper deals with the irradiation effects on the thermal shock resistance Δ = σtk/Eα (σt : tensile strength, k: thermal conductivity, E: Young's modulus, α: thermal expansivity) and the thermal shock fracture toughness ∇=KIck/Eα (KIc : fracture toughness value of mode I) in addition to other mechanical properties such as the diametral compressive strength and fracture toughness of two kinds of near isotropic graphite neutron irradiated at 750∼1,000°C to a dose of (1.1 ∼ 1.5)1021 n/cm2 (>29 fJ). One of the graphite specimen is an isostatically molded graphite IG-11 with fine grain petroleum coke and the other is a binderless molded graphite HCB-18 which was prepared using mesophase pitch carbon with very fine grain size. These measurements are carried out by means of disk testing method developed by us. Results show that both the thermal shock resistance and the thermal shock fracture toughness of the two kinds of graphite after irradiation decrease considerably in contrast with increasing trends of th...

Effect of oxidation on physical properties of neutron irradiated nuclear grade graphite

Changes in thermal conductivity, electrical resistivity and Young's modulus due to thermal oxidation in air were studied for the neutron irradiated nuclear grade graphite IG-11. Samples were irradiated and then oxidized in air at 450°C up to the maximum weight loss of 27%. Neutron irradiation caused the increases of the rate of oxidation in air, electrical resistivity and Young's modulus, and the decrease of thermal conductivity as well. Subsequent oxidation led to the results of the increase of electrical resistivity, and the decreases of Young's modulus and thermal conductivity. An analytical expression was given to the present experimental results, and the tendencies of the changes in the properties of neutron irradiated nuclear grade graphite due to the oxidation were clarified to be the same as those of unirradiated samples.

Fine structure of Wigner energy release spectrum in neutron irradiated graphite

Stored energy release spectra have been measured using the linear-rise method for pyrolytic graphite samples, neutron-irradiated to a fluence of 4 × 10 17 n/cm 2 at about 80°C. Seven heating rates were chosen in the range of 1°C/min to 100°C/min. Total stored energy release is (8±1)J/g. The spectra are composed of three peaks. The reactions of three peaks follow first-order kinetics, and their respective frequency factors and average activation energies are 2.2 × 10 12 s −1 and 1.34 eV; 8.5 × 10 12 s −1 and 1.50 eV; and 1.5 ×10 14 s −1 and 1.78 eV. The initial activation energy spectrum is approximated by the Gaussian distribution. The fraction of each peak varies with the heating rate. The mechanism of the Wigner energy release is discussed on the basis of these kinetic results.