Characterization of tensile strength and fracture toughness of nuclear graphite NBG-18 using subsize specimens

The mechanical properties of NBG-18 nuclear grade graphite were characterized using small specimen test techniques and statistical treatment on the test results. New fracture strength and toughness test techniques were developed to use subsize cylindrical specimens with glued heads and to reuse their broken halves. Three sets of subsize cylindrical specimens of different sizes were tested to obtain tensile fracture strength and fracture toughness. The mean fracture strength decreased as the specimen size increased. The fracture strength data indicate that in the given diameter range the size effect is not significant and much smaller than that predicted by the Weibull moduli estimated for individual specimen groups of the Weibull distribution. Further, no noticeable size effect existed in the fracture toughness data. The mean values of the fracture toughness datasets were in a narrow range of 1.21-1.26 MPa√m.     

Comparison of oxidation model predictions with gasification data of IG-110, IG-430 and NBG-25 nuclear graphite

A phenomenological oxidation kinetics model of graphite is presented and its results are compared with the reported experimental gasification data for nuclear graphite of IG-110, IG-430 and NBG-25. The model uses four elementary chemical kinetics reactions, employs Gaussian-like distributions of the specific activation energies for adsorption of oxygen and desorption of CO gas, and accounts for the changes in the effective surface areas of free active sites and stable oxide complexes with weight loss. The distributions of the specific activation energies for adsorption and desorption, the values of the pre-exponential rate coefficients for the four elementary chemical reactions and the surface area of free active sites are obtained from the reported measurements using a multi-parameter optimization algorithm. At high temperatures, when gasification is diffusion limited, the model calculates the diffusion velocity of oxygen in the boundary layer using a semi-empirical correlation developed for air flows at Reynolds numbers ranging from 0.001 to 100. The model also accounts for the changes in the external surface area, the oxygen pressure in the bulk gas mixture and the effective diffusion coefficient in the boundary layer with weight loss. The model results of the total gasification rate and weight loss with time in the experiments agree well with the reported measurements for the three types of nuclear graphite investigated, at temperatures from 723 to 1226 K and weight loss fractions up to ～0.86.     

Transient gasification in an NBG-18 coolant channel of a VHTR prismatic fuel element

This paper investigates the transient gasification of NBG-18 nuclear graphite with atmospheric air ingress in a 0.8-m long coolant channel of a prismatic Very High Temperature Reactor fuel element. Analysis varied the initial graphite and air inlet temperature, T_o, from 800 to 1100 K at air inlet Reynolds number, Re_in = 5, 10 and 20. The analysis employs a Generic Interface that couples a multi-species diffusion and flow model to readout tables of the CO and CO₂ production fluxes. These fluxes are functions of the graphite local surface temperature, oxygen partial pressure and graphite weight loss fraction and calculated using a chemical-reactions kinetics model for the gasification of nuclear graphite. The analysis accounts for the heats of formation of CO and CO₂ gases, the heat conduction in the graphite sleeve, and the change in the oxygen partial pressure in the bulk gas flow mixture along the channel. Transient calculations performed up to a weight loss fraction of 0.10 at the entrance of the channel, t₁₀. They include the local graphite surface temperature and composition of bulk gas flow, the local and total graphite weight losses and the local and total production rates of CO and CO₂ gases. The heat released in the exothermic production reactions of these gases increases the local graphite surface temperature, accelerating its gasification. At the end of the calculated gasification transient, t = t₁₀, the graphite weight loss is highest at the channel entrance and decreases rapidly with axial distance into the channel, to its lowest value where oxygen in the bulk gas flow is depleted. Increasing T_o decreases t₁₀ and the total graphite loss, while increasing Re_in decreases t₁₀ but increases graphite loss.     

Microstructural characterization and pore structure analysis of nuclear graphite

Graphite will be used as a structural and moderator material in next-generation nuclear reactors. While the overall nature of the production of nuclear graphite is well understood, the historic nuclear grades of graphite are no longer available. This paper reports the virgin microstructural characteristics of filler particles and macro-scale porosity in virgin nuclear graphite grades of interest to the Next Generation Nuclear Plant program. Optical microscopy was used to characterize filler particle size and shape as well as the arrangement of shrinkage cracks. Computer aided image analysis was applied to optical images to quantitatively determine the variation of pore structure, area, eccentricity, and orientation within and between grades. The overall porosity ranged between ∼14% and 21%. A few large pores constitute the majority of the overall porosity. The distribution of pore area in all grades was roughly logarithmic in nature. The average pore was best fit by an ellipse with aspect ratio of ∼2. An estimated 0.6-0.9% of observed porosity was attributed to shrinkage cracks in the filler particles. Finally, a preferred orientation of the porosity was observed in all grades.     

Decontamination of nuclear graphite

Graphite is used in gas-cooled reactors (e.g. AGR, MAGNOX, HTR) and Russian RMBK reactors as a moderator and reflector. About 250,000 Mg of irradiated graphite (i-graphite) has to be considered as radioactive waste in the next few centuries. Fission products and activation of impurities in the graphite contaminate this graphite during reactor operation. Key nuclides for waste management are Co-60 during decommissioning, if decommissioning is performed immediately after reactor shutdown, and the long living radionuclides ¹⁴C and ³⁶Cl for long-term safety in the case of direct disposal. Most radioisotopes can, in principle, be removed by using the purification methods already applied during the manufacture of nuclear graphite. However, due to the same chemical behaviour as ¹²C, this does not seem to be applicable to ¹⁴C.Contaminated graphite cannot be stored in low-level surface disposal facilities such as Centre de L’Aube, in France, due to the long half-life of ¹⁴C [Millington, D.N., Sneyers, A., Mouliney, M.H., Abram, T., Brücher, H., 2006. Report on the International Regulation as regards HTR/VHTR Waste Management, Deliverable D-BF1.1 of the Raphael Project, EC Contract 516508, Confidential report]. Furthermore, the ¹⁴C activity of the graphite reflectors from the two German HTR reactors (AVR and THTR) would amount to more than 90% of the total ¹⁴C activity licensed for the underground disposal site Konrad in Germany for non-heat-generating radioactive waste [Brennecke, P., October 1993. Anforderungen an endzulagernde radioaktive Abfälle (Vorläufige Endlagerbedingungen, Stand: April 1990 in der Fassung vom Oktober 1993) - Schachtanlage Konrad -, BfS-ET-3/90-REV-2, Salzgitter, p. 51].The burning of nuclear graphite would be an efficient method for volume reduction, but would not be accepted by the public as long as all the ¹⁴C were emitted into the atmosphere in the form of CO₂. The required separation of the ¹⁴C from the off-gas is difficult and not economic because this carbon isotope has the same chemical properties as the ¹²C from the graphite matrix. The solidification of the whole amount of CO₂ would cancel out the volume reduction advantage of burning.Thus, a process is required which benefits from the inhomogeneous distribution of the ¹⁴C in the graphite matrix leading to ¹⁴C-enriched and ¹⁴C-depleted off gas streams (Schmidt, P.C., 1979. Alternativen zur Verminderung der C-14-Emission bei der Wiederaufarbeitung von HTR-Brennelementen, JÜL-1567].Tritium and other radioisotopes, including ³⁶Cl and ¹²⁹I, can be removed from graphite by thermal treatment. Even significant parts of the ¹⁴C inventory can be selectively extracted because most of the ¹⁴C may be adsorbed on the surface of the crystallites in the pore structure and not integrated into the crystal lattice. This has recently been demonstrated in principle by the HTR-N/N1 project. As an accompaniment to thermal treatments, steam reforming is an alternative method for decontaminating graphite from radionuclides. The decontamination rates are even higher in comparison to pure thermal treatment in an inert atmosphere, as was first evidenced by basic experiments in the HTR-N/N1 project.     

Effects of dimensional change strain in nuclear graphite component stress analysis

Shrinkage and thermal stresses are induced into graphite components when they are irradiated in nuclear reactor cores. These stresses have to be taken into account in the reactor design and subsequent safety case assessments. This is usually done using graphite irradiation constitutive models programmed into a finite element code. The models use empirical data for the irradiation induced property and dimensional change, which are obtained from graphite material test reactor programmes. The dimensional change in nuclear graphite is one of the most important strains induced by the irradiation fluence. In this paper the effect of two different numerical methods to calculate the dimensional change strain is examined. Then the effect on the predicted stress using two different empirical models for dimensional change is studied. The solutions show that although the difference between two models is small, there are considerable differences in the stress profile.     

Behavior of graphite in nuclear reactor stacks

The results of a study of radiation effects in graphite and the influence of those effects on the design of graphite stacks for nuclear reactors are discussed in this article. Several of the manifestations of these effects may lead to serious complications in reactor performance. Measures used to avert such complications are considered. A unified approach to the physical nature of the radiation effects in graphite is suggested for a broad range of elevated temperatures. The problem of preventing oxidation of graphite is approached in the light of the high temperatures prevailing in the graphite stacks of reactors of recent design.     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

Change in electrical resistivity of nuclear graphite during compression tests and a model for its deformation and fracture mechanism

The results of the resistivity changes during compression of some nuclear graphites are summarized in order to cast light on the fracture mechanism of the materials; data on pyrolytic graphite and amorphous carbon are also taken into account. It is found that all the graphites investigated show an abrupt increase in resistivity when the applied stress increases to about a half of the fracture stress. Above this stress the non-linearity of the stress-strain curve becomes more pronounced and the formation and growth of optically resolvable cracks occur. A model based on the deformation of cracks and pores on the basal plane is proposed for explaining the change in resistivity, and is supported by measurements of the effect of pre-stressing on the Young's modulus, thermal expansion, mercury porosimetry and Knoop micro-hardness of the material.     

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

Specimens 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 (E_a) 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 E_a 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.     

A statistical interpretation of impact endurance curves for polycrystalline graphite

A model for analysing impact data is described, which assumes that the probability of impact failure at constant impactor energy increases linearly with the number of impacts. It is employed to fit an experimental normalized endurance curve for eight polycrystalline graphites in terms of a probability distribution function for single impact failure. The model is readily extended to treat a series of impacts of varying energy, and it is found to give good agreement with experimental data for failures occurring at fewer that twenty impacts.     

The effect of porosity on the thermal conductivity of nuclear graphite

During service in a high-temperature reactor, graphite will be oxidized by impurities such as water vapour present in the helium coolant. Oxidation will affect the thermal conductivity of the graphite and hence the fuel temperature. This report describes experiments on the effect of oxidation at 1000°C by water vapour of a semi-isotropic moulded graphite. The value of thermal conductivity at room temperature decreases with increasing weight loss, but not linearly, the decreases being most rapid at low weight losses. The percentage change in thermal conductivity is approximately linear with the increase in open porosity.