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[en] The requirements for a solid moderator are reviewed and the reasons that graphite has become the solid moderator of choice discussed. The manufacture and properties of some currently available near-isotropic and isotropic grades are described. The major features of a graphite moderated reactors are briefly outlined. Displacement damage and the induced structural and dimensional changes in graphite are described. Recent characterization work on nano-carbons and oriented pyrolytic graphites that have shed new light on graphite defect structures are reviewed, and the effect of irradiation temperature on the defect structures is highlighted. Changes in the physical properties of nuclear graphite caused by neutron irradiation are reported. Finally, the importance of irradiation induced creep is presented, along with current models and their deficiencies.
[en] This document reports on initial activities at ORNL aimed at quantitative characterization of porosity development in oxidized graphite specimens using automated image analysis (AIA) techniques. A series of cylindrical shape specimens were machined from nuclear-grade graphite (type PCEA, from GrafTech International). The specimens were oxidized in air to various levels of weight loss (between 5 and 20 %) and at three oxidation temperatures (between 600 and 750 C). The procedure used for specimen preparation and oxidation was based on ASTM D-7542-09. Oxidized specimens were sectioned, resin-mounted and polished for optical microscopy examination. Mosaic pictures of rectangular stripes (25 mm x 0.4 mm) along a diameter of sectioned specimens were recorded. A commercial software (ImagePro) was evaluated for automated analysis of images. Because oxidized zones in graphite are less reflective in visible light than the pristine, unoxidized material, the microstructural changes induced by oxidation can easily be identified and analyzed. Oxidation at low temperatures contributes to development of numerous fine pores (< 100 m2) distributed more or less uniformly over a certain depth (5-6 mm) from the surface of graphite specimens, while causing no apparent external damage to the specimens. In contrast, oxidation at high temperatures causes dimensional changes and substantial surface damage within a narrow band (< 1 mm) near the exposed graphite surface, but leaves the interior of specimens with little or no changes in the pore structure. Based on these results it appears that weakening and degradation of mechanical properties of graphite materials produced by uniform oxidation at low temperatures is related to the massive development of fine pores in the oxidized zone. It was demonstrated that optical microscopy enhanced by AIA techniques allows accurate determination of oxidant penetration depth and of distribution of porosity in oxidized graphite materials.
[en] Nuclear Block Graphite-10 (NBG-10) is a medium-grain, near-isotropic graphite manufactured by SGL Carbon Company at their plant in Chedde, France. The filler coke is coal tar pitch derived (maximum size ∼ 1.6 mm). The graphite is formed by extrusion and doubly impregnated. The formulation and manufacturing process are based on that of the graphite fuel sleeves used for Advanced Gas Cooled Reactor fuel stringers in the United Kingdom. NBG-10 graphite was developed as a candidate core structural material for the Pebble Bed Modular Reactor (PBMR) currently being designed in South Africa, and for prismatic reactor concepts being developed in the USA and Europe. NBG-10 is one of several graphites included in the United States Department of Energy Very High Temperature Reactor program.
[en] Here we report the physical properties of a series specimens machined from pilot scale (∼ 152 mm diameter x ∼305 mm length) grade PCEA recycle billets manufactured by GrafTech. The pilot scale billets were processed with increasing amounts of (unirradiated) graphite (from 20% to 100%) introduced to the formulation with the goal of determining if large fractions of recycle graphite have a deleterious effect on properties. The properties determined include Bulk Density, Electrical Resistivity, Elastic (Young s) Modulus, and Coefficient of Thermal Expansion. Although property variations were observed to be correlated with the recycle fraction, the magnitude of the variations was noted to be small.
[en] The current status of graphite irradiation induced creep strain prediction is reviewed and the major creep models are described. The ability of the models to quantitatively predict the irradiation induced creep strain of graphite is reported. Potential mechanisms of in-crystal creep are reviewed as are mechanisms of pore generation under stress. The case for further experimental work is made and the need for improved creep models across multi-scales is highlighted.
[en] The application of a creep model previously applied to compressive creep data for H-451 irradiated at 900 deg. C (13.7 and 20.8 MPa) has been extended to compressive creep data for H-451 irradiated at 600 deg. C (13.7 and 20.8 MPa). The basis of the creep model is discussed and the experimental data required to evaluate the terms in the creep model are reported and discussed. The model, which corrects the true (crystal) creep strain for the effect of creep on the dimensional change component of the creep specimen, is shown to be a good fit to the data. Creep strain data for H-451 graphite irradiated at 900 deg. C under a tensile stress of 6 MPa are also reported, along with the required experimental data to evaluate the terms in the creep model. The model is shown to inadequately represent the high dose (post volume turn-around) H-451 tensile creep strain data. Reasons for the models limitation are discussed and an approach to a potentially improved graphite irradiation creep model is suggested
[en] The core of a prismatic High Temperature Reactor (HTR) is constructed from an array of nuclear graphite components including replaceable fuel blocks, replaceable and permanent moderator blocks, and core support posts. Similarly, the core of a Pebble Bed Reactor is confined by large graphite blocks which define the (annular) core geometry. In both HTR designs (prismatic and pebble bed) the large graphite components act as neutron moderator and reflector as well as providing mechanical support to the active core. During reactor operation the graphite components of the core are subjected to complex stress states arising from structural loads, thermal gradients, neutron irradiation damage, and seismic events. Structural design of HTR cores requires that the designer have a suitable theory of failure. Both deterministic (e.g. maximum principal stress theory) and probabilistic (e.g., Weibull failure theory) have been considered as candidates. To test candidate theories a multiaxial testing program was conducted at Oak Ridge National Laboratory on grade H-451 graphite, the fuel element and moderator graphite used in the Fort St. Vrain high temperature reactor in the USA. Large test specimens (∼27 cm length) were subjected to combined axial stress (both tension and compression) and internal pressure. A total of 59 specimens were tested at 9 stress ratios in the first and fourth stress quadrants. In a parallel effort a physically based fracture model of graphite was developed. The model used a fracture mechanics based failure criteria and has been shown to predict the tensile failure probability of several graphites of widely ranging texture. Here we report the basis and performance of the fracture model and multiaxial strength data for grade H-451 graphite. Moreover, we report the successful extension of the model to predict the failure envelope for H-451 graphite in the first and fourth multiaxial quadrants. The model's predictions are compared to experimental multiaxial strength data.
[en] Recycle is being considered as a disposal option for irradiated graphite from gas cooled reactors. Thermal annealing was performed on irradiated graphite samples to establish what fraction of in-crystal (displacement damage) and ex-crystal (pore generation) damage could be recovered. The sample dimensions and electrical resistivity were measured after isochronal annealing at 500, 1000, 1500, and 2000 C. Sample dimensions were unaffected by annealing. Some fraction of the resistivity was unannealable at 2000 C, this behavior which was attributed to (a) structural changes resulting from pore generation (also causing dimensions to be unaffected by annealing), or, (b) defect structures in irradiated graphite that persist even at an annealing temperature of 2000 C. The resistivity annealing behavior was different in low irradiation temperature samples, Tirr < ∼400 C and high irradiation temperature samples Tirr > ∼400 C, which indicated defect structures established during irradiation were different in these two temperature regimes. Moreover, the extent of recovery was much greater in the specimens with higher irradiation temperatures, and was small for lower dose low temperature specimens. Evidently, some in-crystal defects that scatter or trap electrons are still present in the crystal lattice after annealing and their tenacity depends upon the irradiation temperature and accumulated neutron dose.
[en] Oxidation behavior of graphite is of practical interest because of extended use of graphite materials in nuclear reactors. High temperature gas-cooled reactors are expected to become the nuclear reactors of the next generation. The most critical factor in their safe operation is an air-ingress accident, in which case the graphite materials in the moderator and reflector would come in contact with oxygen at a high temperature. Many results on graphite oxidation have been obtained from TGA measurements using commercial instruments, with sample sizes of a few hundred milligrams. They have demonstrated that graphite oxidation is in kinetic control regime at low temperatures, but becomes diffusion-limited at high temperatures. These effects are better understood from measurement results with large size samples, on which the shape and structural factors that control diffusion can be more clearly evidenced. An ASTM test for characterization of oxidation resistance of machined carbon and graphite materials is being developed with ORNL participation. The test recommends the use of large machined samples (∼ 20 grams) in a dry air flow system. We will report on recent results and progress in this direction
[en] The core of a prismatic High Temperature Reactor (HTR) is constructed from an array of nuclear graphite components. Similarly, the core of a Pebble Bed HTR is confined by large graphite blocks which define the (annular) core geometry. In both HTR designs the large graphite components act as neutron moderator and reflector as well as providing mechanical support to the active core. During reactor operation the graphite components of the core are subjected to complex stress states. Consequently, core designers need a suitable theory of failure. Both deterministic (e.g., maximum principal stress theory) and probabilistic (e.g., Weibull failure theory) have been considered. To test candidate failure theories a multiaxial testing program was conducted at Oak Ridge National Laboratory on H-451 graphite. Large specimens (∼27 cm length) were subjected to combined axial stress (tension and compression) and internal pressure. A total of 59 specimens were tested at 9 stress ratios in the first and fourth stress quadrants. Here, we report the basis and performance of a microstructurally based graphite fracture model and multiaxial strength data for grade H-451 graphite, along with the application of the model to predict the failure envelope for H-451 graphite in the first and fourth multiaxial quadrants