Irradiation Testing of Matrix Graphite for Spherical HTR-10 Fuel Elements

In order to evaluate if the fuel elements and their components, matrix graphite and coated fuel particles, can meet the design requirement of 10 MW High Temperature Gas-cooled Reactor, an irradiation test with 4 spherical fuel elements, 60 matrix graphite specimens of 5mm×5mm×40mm and 13500 coated fuel particles was performed in Russian IVV-2M reactor from July 2000 to February 2003. The irradiation temperature was 1000℃. The fast neutron fluence of matrix graphite specimens reached 1.3×10²¹cm^-2 (E > 0.1MeV). Post irradiation examination contained the visual inspection, dimension measurement and determining the density, porosity, specific electrical resistance and bending strength. The irradiation results are given in this paper, and show that the matrix graphite for spherical HTR-10 fuel elements made from the domestic raw materials and fabricated by the quasi-isostatic room temperature moulding process is suitable as a structural material for spherical HTR fuel elements. (author)