Since the late 1940s, many journal papers, conference papers, and reports have been published on the change in properties in graphite due to fast neutron damage. Many different units have been used to define graphite damage dose (or fluence). It is impor­tant to understand the basis of these units because historic data are still being used to justify models

used in assessments for component behavior in reac­tors. Indeed, some of these historic data, for example, stored energy and strength, will also be used to sup­port decommissioning safety assessments.

Early estimations of ‘graphite damage’ were based on the activation of metallic foils such as cobalt, cadmium, and nickel. Later, to account for damage in different reactors, equivalent units, such as BEPO or DIDO equivalent dose, were used where the dam­age is referred to damage at a standard position in the BEPO, Calder Hall, or DIDO reactors. The designers of plutonium production reactors preferred to use a more practical unit related to fuel burnup (megawatts per adjacent tonne of uranium, MW/Atu). Research­ers also found that the calculation of a flux unit, based on an integral of energies above a certain value, was relatively invariant to the reactor system and used the unit En > 0.18 MeV and other variants of this.

Today, the favored option is to calculate the flu — ence using a reactor physics code to calculate the displacements per atom (dpa). However, in the field of nuclear graphite technology historic units are still widely used in the literature. For example, reactor operators have access to individual channel burnup which, with the aid of axial ‘form factors,’ can be used to give a measure of average damage along the indi­vidual channel length.

Fortunately, most, but not all, of these units can be related by simple conversion factors. However, care must be taken; for example, the unit of megawatt days per tonne of uranium (MWd t- ) is not necessarily equivalent in different reactor systems.

When assessing the analysis of a particular com­ponent in a reactor, one must be aware that a single detailed calculation of a peak rated component in the
center of the core may have been carried out to give spatial, and maybe temporal, distribution of that component’s fluence (and possibly temperature and weight loss). These profiles may have then been extrapolated to all of the other components in the core using the core axial and radial ‘form factors.’ In doing this, some uncertainty will be introduced and clearly, some checks and balances will be required to check the validity of such an approach.