Radioactive products may be released from a nuclear waste repository to the biosphere through dissolution in, and transport by ground-water. The fundamental chemical phenomena involved are solubility, complex formation, redox and sorption (e.g. ion-exchange) processes. A first estimate of the behaviour of the waste-repository-ground water system is usually obtained by using a chemical equilibrium model in order to calculate properties such as chemical speciation and solubilities of important radionuclides. For practical reasons the modelling is often made in two different regions, the near field and the far field. In the latter, the main components of the aqueous phase are essentially those of undisturbed groundwater, while in the former substantial changes in composition may occur. The total electrolyte concentration is usually small, at least in the farfield region.

The current strategy for the permanent isolation of nuclear wastes in the United States involves the storage of these wastes within repositories mined in deep geologic formations. In this disposal strategy, the isolation of nuclear wastes relies on a series of natural and engineered barriers to prevent the unacceptable release of radionuclides to the accessible environment. An integral part of the development of a qualified subsurface nuclear waste repository, therefore, is the assessment of the ability of these barriers to adequately prevent or retard the migration of radionuclides to the accessible environment. The Basalt Waste Isolation Project (BWIP) under guidance from the Department of Energy (DOE) is investigating the feasibility of storing nuclear w-stes in the basalts beneath the Hanford Site.

In several countries low permeability crystalline rocks (e.g. granites) are under consideration as potential hosts for radioactive waste repositories. In such formations groundwater flow occurs predominantly in specific fractures rather than being a general porous flow through the entire rock matrix. By considering fractures to be simple parallel plates various authors have demonstrated the potential importance of diffusion into dead-end pores and the rock matrix itself (‘matrix diffusion’) as a mechanism for the retention of migrating radionuclides. Complementing these theoretical studies, several insitu migration experiments are planned in single fissures in crysalline rocks in Sweden, the U.K. and the U.S.A. The ‘parallel plate’ approximation to a single fissure is, however, acknowledged to be a gross simplification of any real case where “flowing” fractures are expected to be either filled or coated with secondary minerals, formed by hydrothermal alteration of fracture surfaces In the evaluation of net radio-nuclide retardation, therefore, the effect of sorption onto such secondary minerals must be carefully considered.

Modeling of radionuclide transport in geological formations has until recently been limited to the treatment of nonfractured porous media. Recently researchers in Sweden, Canada, France, and the USA have become interested in the problems of nuclide transport in geological formations which may contain fractures. Since many conventional porous-medium ion-transport codes are available, the question arises as to when fracture widths and fracture frequency dictate the use of these existing codes.

Much effort is currently centred on the construction of elaborate computer programs to model the release to the biosphere of radionuclides from proposed high level radioactive waste repositories in deep geological formations. Although the lack or poor quality of background data is often emphasised, it is generally considered that by examining a wide range of possible values of the important parameters involved, likely processes are assessed or their omission and possible relevence is acknowledged.

Much effort is currently centred on the construction of elaborate computer programs to model the release to the biosphere of radionuclides from proposed high level radioactive waste repositories in deep geological formations. Although the lack or poor quality of background data is often emphasised, it is generally considered that by examining a wide range of possible values of the important parameters involved, likely processes are assessed or their omission and possible relevence is acknowledged. However, one factor that has been almost totally ignored is the presence of micro-organisms in deep geological formations. The common assumption is that the biosphere is limited to the earth's surface, and soil to a depth of a few metres. Recent research has shown, however, that viable micro-organisms can inhabit deep groundwaters and that the biosphere can extend to depths of at least 5 km. Examples of organisms tolerant to extreme environments are given in Table 1 although ranges are likely to be conservative due to the difficulties of culturing these organisms, which tend to have exotic nutritional requirements, and sampling them in extremely hostile environments.

Since 1976, a German R & D project has been carried out to find an alternative concept for the treatment and disposal of MLW and LLW arisings generated mainly in the planned German reprocessing plant and other nuclear facilities (LWR, fuel fabrication, R & D establishments).

Prior to disposal of cemented medium level waste and low level waste in large caverns 1000 m below ground it is important to know what temperatures will be caused by heat generation in the waste product. During the disposal period the heat produced by hydration of the cement used as the binder as well as by decaying radionuclides in the waste may give rise to inadmissibly high temperatures in the waste product resulting in a deterioration of the produc:t qualities. To avoid such undesirable processes, a maximum permissible product temperature of 90°C was specified for the investigations, i.e. a temperature that is significantly below the boiling point of the water contained in the cemented product.

As already mentioned in the overview paper /1/ the in-situ solidification process requires a preconditioning of MLW/LLW arisings by granulation. The main purpose thereof is the dissipation of the heat of hydration already above ground and to replace in the final concrete like waste form a considerable part (i.e. 60 vol.%) with inert granules.

In the process described in the first paper of this session for containerless final storage of radioactive wastes, a granule-cement suspension (waste grout) is transported by means of a vertical pipe from above ground into the underground cavern.

The geomechanical and mining-technological aspects of the construction of salt caverns as disposal chambers have been investigated during project phase 2, completed by mid 1981. With a view towards the stability analysis of such a cavern, FEM-estimates have been carried out and evaluated. From these it can be derived that

- a rotational ellipsoid would be the most suitable shape

- its dimensions should be 82 m (vertical axis) and 42 m (horizontal axis)

- the distance (safety pillar) between the neighbouring caverns should be 170 m (vertical) and 180 m (horizontal).

For practical engineering purposes the rotational ellipsoid can be modified into a cylinder with conic bottom and top. The numerical model simulated the short term as well as the long term characteristics of the surrounding salt rocks. The short term characteristics were assessed by an elastic approach, the long term characteristics by a rheological model. The input parameters have been determined by means of laboratory tests on ASSE rock salt.

In a second step the characteristics of partially and completely filled caverns were simulated. It was shown clearly that deformation of the salt rock comes to a halt when counteracted by the filling.

Based upon the results of the stability analysis, investigations were made to find out a suitable mining technique for the construction of the cavern. Solution mining and conventional development by means of drilling and blasting have been studied alternatively. Since both methods have their advantages and disadvantages a decision in favour of the one or the other cannot be made until the actual site has been defined.

1. Hr. Gordon's comments are specific to defense waste only. The research priorities center around:

2. 1. Improvements in the technology for immobilizing high-level waste. For waste vitrification, these improvements include lower corrosiveness, less crystal formation, lower volatility, reduction in glass cracking, increased life of glass-melter, nondestructive testing methods, and remote handling techniques.

3. 2. Better understanding of the performance of the waste package components in the repository environment. This requires additional research in the areas of repository characterization, the mechanisms of waste form corrosion, radionuclide migration in geologic media, predictive models for projecting waste system performance.. and synergistic effects of waste system components.

4. 3. Confirmation of waste disposal risk assessments and their establishment as the technical basis for regulatory criteria.

5. 4. Research priorities for transuranic (TRU) waste isolation are focused on the development of processing technologies for immobilization and of improved instrumentation for classification and nondestructive examination.