The effects of neutron damage on the retention of tritium in carbon composites

Tritium retention measurements have been performed for two carbon composites both before and after neutron irradiation. These two composites were FMI-222 manufactured by Fiber Materials Incorporated and MKC-1PH by Mitsubishi Kasei. The neutron irradiations were performed in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory at 473K and spanned radiation levels from unirradiated up to 1.0dpa. The high energy traps in the composites were saturated in tritium by heating the composties to 1673K with a pure tritium gas pressure of 13.2Pa. The samples were then outgassed in a separate system at a temperature of 1963 K where tritium analysis was performed using an ionization chamber and liquid scintillation counting. For all irradiation conditions, the retained tritium was less that that measured in earlier studies for different types of graphites. The results suggest carbon composites should be preferred over graphites for use in fussion reactors where both tritium and neutron irradation exist.