Examination of microcracks in crystalline rocks has shown that the open microcracks form a spatially continuous network that communicates with larger openings. The open cracks become filled with new minerals that often contain U and LREEs. The crystalline rocks most desirable as a host for a repository have low crack porosity, moderate alteration, and occur in an environment that promotes rapid sealing.

The uptake of cesium on natural fracture surfaces taken from granitic bedrock has been studied as well as the depth penetration into the underlying rock. The fracture minerals usually exhibit good sorptive properties with respect to cesium (with the exception of calcite). The surface coatings do not act as physical diffusion barriers and do not prevent a migration into the underlying rock. A considerable depth penetration into the rock (several mm over a period of three months) was observed. The dominating pathways into the rock matrix were the high-capacity minerals in grains or in microfissures. The calculated diffusivity in the rock matrix was of the order of 10−13 m2/s.

This paper summarizes the creep consideration in the design of nuclear waste storage facilities in salt, describes the non-linear analysis method for evaluating the design adequacy, and presents computational results for the current storage design. The application of rock mechanics instrumentation to assure the appropriateness of the design is discussed.

It also describes the design evolution of such a facility, starting from the conceptual design, through the preliminary design, to the detailed design stage The empirical design method, laboratory tests and numerical analyses, and the underground in situ tests have been incorporated in the design process to assure the stability of the underground openings, retrievability of waste during the operating phase and encapsulation of waste after decommissioning.

In this paper we investigate how the geologic structure of a basin influences the reliability with which predictions of radionuclide transport within a depository horizon can be made. Results from a series of stochastic simulations show that, in some cases, transport predictions are sensitive to perturbations in the hydraulic properties of surrounding layers, indicating megascopic effects on a basin-wide scale exert a significant control on transport at the macroscopic scale. Other cases are identified where transport is not sensitive to perturbations in the hydraulic properties of surrounding layers. A key factor in this response is the distribution of the hydraulic gradient in the basin.

A mathematical model has been developed to study the effect of channelling in a porous medium surrounding a repository on the migration of radionuclides. The porous medium is treated as an infinite medium containing parallel, unconnected channels; each channel is surrounded by a rock matrix. The migration in the channel is controlled by convection and diffusion processes in the direction of the flow, whereas in the rock matrix it is dominated by anisotropic diffusion in the longitudinal and transverse directions. Unsteady state convective-diffusion equations for this system have been solved. At the channel-rock matrix interface the concentration and flux were watched to obtain concentration profiles and discharge rates at downstream locations. By means of a parametric study the effects of channel thickness, diffusion coefficients, and flow velocity on the migration rate are elucidated.

Physico-chemical forms of several radionuclides are being investigated in order to characterize the migrating species in a deep geologic repository. Under ground water bicarbonate and carbonate concentration levels, Am(CO3)2 has been observed to be the most important complex while the formation of Am(OH)(CO3)22− has been indicated, stability constants of Am-complexes have been calculated, and a distribution diagram of Am(III) species as a function of pH is presented. Speciation of Tc has been studied using radiochemical and electromigration methods in oxic and anoxic environments; in effective diffusion coefficient of 1.37 × 10−5 cm2 S−1 for ToO4− has been obtained. Preliminary studies on Np(V) indicate that NpO2+ does not form strong complexes with carbonate and bicarbonate ions at ground water concentrations.

It is generally accepted that waste to be disposed of in a deep geologic repository will be suitably packaged. The design of such packages depends on the criteria which they must meet and the geochemical environment in which they are to be placed. Both the criteria and the environment must be considered in determining the components required to make up the package, the functions those components are to perform, and the materials of which the components are to be made.

On October 6, and 7 1982, after a two-day Parliamentary Debate, the continuation of the French nuclear power program was approved. Installed nuclear capacities, should reach 56000 elMN by 1990, i.e. 28% of energy supplies.

The nuclear waste volumes entailed by such a nuclear electricity program, with the accompanying fuel cycle operations and the other nuclear waste producing activities, are the following: Low and Medium Level waste volumes should reach 800,000 cubic meters by the year 2000 - slightly less than was planned for two years ago, because of reduced power consumption, and therefore capacities forecasts. Alpha-bearing and vitrified Wastes should reach respectively 45,000 cubic meters and 3,000 cubic meters by the year 2000.

These waste quantities now make it urgent to set up a wholly comprehensive industrial scale program for their treatment, conditioning and final disposal, together with the relevant research and development activities.

In the United Kingdom the responsibility for radioactive waste management policy lies with the environmental Ministers. The Department of the Environment is also responsible in England for authorisation of all radioactive waste disposal activities and has its own Radiochemical Inspectorate to ensure that government policy is implemented. An independent Radioactive Waste Management Advisory Committee reports directly to the Secretary of State for the Environment.

Low-level wastes are currently disposed of by shallow burial or to the deep ocean. Sites for the disposal of intermediate-level wastes are being sought as a matter of urgency and heat-emitting, reprocessing wastes will be stored on the surface for at least fifty years. The rationale of this policy is explained.

The current status of radiation effects in solid nuclear waste forms is reviewed. The general nature of radiation effects and the simulation techniques used are briefly discussed. Recently reported data on radiation effects are presented and discussed, and those areas requiring further theoretical or experimental study are identified.

Radiation damage due to a α-decay of actinides will modify the material used to store nuclear wastes. For example, ionization-induced processes can fragment the oxide bonds of a glass thereby creating gas bubbles. Phase decomposition, swelling, and disordering of crystalline phases are other possibilities. The self-irradiation damage could also affect the fracture properties of the storage materials. Spontaneous failure and fragmentation of some brittle materials have been observed during storage. Yet the problem of fracture of irradiated waste materials has received very limited attention.

Ion implantation is a rapid technique for simulating damage induced by α-recoil nuclei in nuclear waste forms. The simulation has been found to be quite good in TEM comparisons with natural α-decay damage in minerals, but leach rate differences have been observed in glass studies and were attributed to dose-rate differences. The similarities between ion implantation and recoil nuclei as a means of producing damage suggest that insights into the long-term behavior of glass waste forms can be obtained by examination of what is known about ion-implantation damage in silicate glasses. This paper briefly reviews these effects and shows that leaching results in certain nuclear waste glasses can be understood as resulting from plastic flow and track overlap. Phase separation is also seen to be a possible consequence of damage-induced compositional changes.

A large number of computer calculations of the radiolysis of ground water outside copper canisters have been carried out within the Swedish Nuclear Fuel safety Project.

At dose rates higher than 5 × 10−2 rad/s the hydrggen concentration is constant, approximately 10−3 mol/dm3, and independent of time and dose rate. Therefore the rate of difiusion out of the system is constant, approximately 7 × 10−4 mol/year. At dose rates lower than 5 × 10−4 rad/s the logarithm of the hydrogen production is proportional to the logarithm of the dose rate. The hydrogen production depends on fuel type, burn-up and canister thickness only to the extent that the dose rate varies with these parameters. The hydrogen production is independent of the volume of water in which the energy is assumed to be absorbed (5 or 15 cm layers outside the surface of the canister).

The hydrogen production depends on the concentration of Fe2+ ions in the water. In the basic calculations the concentration was assumed to be 5 ppm. An increase to 50 ppm increases the total hydrogen production after one million years outside a 1 cm thick canister from 360 to 1000 mol. A decrease to 0.5 ppm decreases the same yield to 80 mol.

238Pu-substituted cubic zirconolite (CaPuTi2 O7) was stored at ambient temperature, 575 K and 875 K until alpha decay doses of 2.4 to 3.6 × 1025/m3 had been accumulated. The ambient temperature material swelled to a saturation value of 5.5 vol%, and the originally crystalline structure was transformed to one with an amorphous matrix and small domains that had retained their crystallinity. At 575 K lesser amounts of swelling (4.1 vol%) and transformation were observed, reflecting concurrent partial recovery. The material held at 875 K remained crystalline, swelled only 0.4 vol%, and exhibited formation of isolated defect clusters.

The process of ion exchange is used extensively throughout numerous industries where water of extremely high purity is required. Government and commercial nuclear power facilities are typical of this, and have applied ion exchange processes for the removal of radioactive ions from aqueous solutions. When water is decontaminated in this fashion the ion exchange medium becomes saturated with the radioactive ions from solutions, and is then stored as radioactive waste in standard burial facilities. During this indefinite storage period, the chemical changes which occur in the ion exchange medium may result in oxygen depletion and gas generation within the storage vessel, and subsequent loss of integrity. It is the object of this work to investigate the early phases of radiolytic oxygen depletion and gas evolution through experimental procedure and observation. The results of this work will then be analyzed with respect to applications for the storage of radioactive ion exchange waste.

Research is under way to characterize the effects of self-irradiation on radwastes which may be generated when organic ion-exchange media are used in water demineralization or decontamination operations at nuclear facilities. External factors affecting the relation between laboratory evaluations and field performance are emphasized. Initial experiments do not yet indicate substantial radiation dose-rate effects on radiolytic gas yields or acid product formation, when (fully swollen) sulfonic acid resins are irradiated in a sealed air environment. At the same time, oxygen gas is removed from the environment of irradiated resins. Interaction between mild steel coupons and acidic species produced in the irradiation induced decomposition of sulfonic acid resin results in irradiation enhanced corrosion. Corrosion rates depend on radiation dose rate, moisture content and resin chemical loading. In some cases, corrosion rates decrease with time, suggesting depletion of acidic species within the resin bed, or a synergistic interaction between resin and corrosion coupon. Implications of these and other results on evaluating field behavior of radwaste containing ion-exchange media are discussed.

Samples of CaPuTi2O7 were prepared by cold pressing and sintering. Plutonium was substituted for zirconium in order to characterize radiation damage effects. The energy stored in a sample which had reached saturation in swelling after storage at ambient temperature was measured with a differential scanning calorimeter. The total energy of 6.6±0.1 cal/g is released over the range 485–715° C. The activation energy of annealing of the damage is 1.22±0.05 eV. The temperature dependence of the rate constant is described by kT= 5.96E4 exp(−1.22/kBT) s−1 where kB and T are the Boltzmann's constant and temperature(K) respectively. A sample stored at 600°C was similarly evaluated and showed no release of stored energy to the precision of the apparatus (±0.1 cal/g). These results are applied to analysis of waste incorporation in Synroc and are correlated with analogous parameters for other materials.

Initial criteria are defined which determine when radionuclide transport in jointed, porous rock can be approximated as occurring in an equivalent porous medium. For systems satisfying those criteria, the retardation factor to be used in the equivalent porous media model also is defined.

A chemical transport model, CHEMTRN, that includes advection, dispersion/diffusion, complexation, sorption, precipitation or dissolution of solids, and the dissociation of water has been written. The transport, mass action and site constraint equations are written in a differential/algebraic form and solved simultaneously. The sorption process is modelled by either ion-exchange or surface complexation. The model has been used to investigate the applicability of a kD model for simulating the transport of chemical species in groundwater systems, to simulate precipitation/dissolution of minerals, and to consider the effect of surface complexation on sorption.

Some low-level radioactive waste represents both a chemical and radiological hazard, and it is desirable that management practices take this into account. This paper describes a waste classification system that could provide guidance on how mixed wastes could be separated for disposal, and the degree of disposal isolation each waste material or shipment needs. An example of the system's use is provided.