18.1.2 Waste segregation and characterisation

The operational safety case for a disposal facility will not be so very different from that of a waste or spent fuel store. The radionuclides of greatest concern are relatively short-lived gamma emitters (e. g. cobalt-60), which may be detected with a gamma spectroscope or even with a simple hand-held counter. The situation for the long-term (i. e. post-closure) safety of a disposal facility is quite different. Here, the radionuclides with the greatest influence are invariably long lived and often difficult to measure. Typically, they include fission products such as technetium-99, iodine-129 and the actinides. In these cases (with the exception of some actinides) sophisticated laboratory tests are needed to obtain specific activity values. This can be made very much more difficult if the waste has been processed in some way (e. g. mixed with concrete) and, even more so, if different waste streams have been mixed in an uncontrolled way. Consequently, it is essential that waste streams should be kept separate and that characterisation should be performed before any processing is done. Such precautions should allow the radionuclide content of waste streams — and thus waste packages — to be adequately specified so that compliance with waste acceptance criteria can be demonstrated.

Where the radionuclide content of a waste stream is variable, it is usual to assume that, while the total activity may change, the relative proportions of the various radionuclides do not. In this case it may be possible to establish a correlation between difficult-to-measure and easy-to-measure radionuclides. Typically, caesium-137 is used as an indicator for other fission products (e. g. iodine-129) and cobalt-60 is used for neutron activation products such as chlorine-36 and nickel-63.6 There are, however, difficulties with this approach for same radionuclides.7

For spent nuclear fuel, vitrified fission products and decommissioning wastes, the radionuclide content of the material is most often derived by calculation using standard codes and neutron cross sections.