Many near-surface disposal facilities for radioactive waste exist throughout the world[31] — some having operated for decades whilst a few are now in post-closure institutional control. Designs vary depending on the type of waste to be accepted and the environmental conditions at the disposal site, especially the potential for water to come into contact with the waste. In broad terms we expect the number of engineered barriers to increase with the hazard presented by the waste and, similarly, as we move from dry to temperate and tropical climates. In all cases the overall aim is to contain and isolate the waste so that radionuclides do not enter the human environment.

With respect to containment, a crucial element is the avoidance of water contact with the waste. In a naturally dry environment there may be no need for any more barriers than those offered by the waste and its package but in wetter climates a repository cap can provide the necessary protection against infiltration of rainwater. Where the groundwater table is close to the surface, facilities are normally created above grade. Otherwise, below grade facilities have a number of advantages including less need for lateral support for the walls, simpler operation and a reduced susceptibility to erosion because of the lower profile (Fig. 18.3). For VLLW in a humid environment, the provision of a cap may be enough to supply the needed level of containment. For LILW, however, it is usual to install additional engineered barriers in the form of concrete, which is used to immobilise the waste within the package, to fill the gaps between waste packages in the waste stack and to provide the engineered structures (walls, base and cap) of the facility.

So far as isolation is concerned, the extensive use of concrete will help to reduce erosion and make the structure more resistant to inadvertent human intrusion. But in near-surface disposal, by far the most important element in this respect is institutional control, which should be considered as an engineered barrier in its own right. While there may be an intention to maintain everlasting institutional control over a facility, this is too bold a claim to be used in a safety case. Based on the longevity of existing human institutions and allowing some margin for conservatism, a commonly claimed institutional control period is

300 years. After that period, the safety case must assume that control will be lost and humans may reoccupy the site. To obtain approval for a proposed site, the safety case will need to demonstrate that this can be done safely. A radionuclide that is often prevalent in LILW is caesium-137. This has a half-life of about 30 years, which means that a 300-year control period will produce a decrease in radioactivity of about 1000 times (210). Let us suppose that humans could safely live on a site where specific activities were at exemption levels. For caesium-137, this is 10 Bq/g.14 Let us further allow that mixing of the waste with uncontaminated material (from the cap for instance) reduces the specific activity of the resulting soil by a factor of ten. In that case, the maximum permissible specific activity of Cs-137 in a waste for disposal would be about 100 000 Bq/g (100 kBq/g). This is of the same order of magnitude as the value allowed for near-surface disposal of caesium-137 in France, which is 330 kBq/g.28

Examples of the silo design (Fig. 18.3) may be found in Sweden and Finland, where they are constructed at a depth of about 100 m. A more recent development is one proposed to be built at Vrbina in Slovenia. This is about 12 m below the surface with a maximum depth of more than 50 m.29 The greater depth of disposal that is available with the silo design provides more protection from human intrusion and may thus enable the disposal of LILW.