Radiation damage of graphite in fission and fusion reactor systems

Abstract

There are several compelling reasons to consider graphite use in fusion reactors. The plasma tolerance for impurities is much higher for the low atomic number ( Z=6) carbon materials, so it is useful as a plasma interface material. Its high thermal shock resistance provides unique limiter design options and its low neutron capture cross section combined with very low activation can provide a fusion reactor system with much alleviated maintenance, waste disposal, environmental and safety problems. Its applicability as a neutron reflector applies to fusion as well as fission systems. However, the nature of the radiation damage to the graphite lattice places some fundamental limits on its useful lifetime in fusion as well as fission reactors. This creates design and economic problems primarily. Increasing the crystalline perfection of artificial graphites is suggested as one method of reducing the crystallite damage. The life expectance for the isotropic conventional graphites will in each case depend on the reactor component for which it will be used and on its design considerations. Based on neutron damage and related dimensional changes it is estimated graphite will be tenable to about 3 × 10 22 n/cm 2 (EDN) at 400°C, 0.6 × 10 22 n/cm 2 (EDN) at 1000°C and 1.4 × 10 22 n/cm 2 (EDN) at 1400°C. There are no data above 1400°C on which to speculate. A dose of 2 × 10 22 n/cm 2 may be accumulated in times ranging from as short as a few months in the first wall region of high power density designs to the fusion plant lifetime (30 years) in the neutron reflector region behind the blanket.