To assess the safety for disposal of radioactive waste cement composite, the leaching of Cs-137, Co-60, Mn-54 and Sr-85 from a waste composite into a surounding fluid has been studied [1, 2]. Leaching tests were carried out in accordance with a method recommended by IAEA |2, 3|. Determination of retardation factors, KF, and coefficients of distribution, kd, using a simplified mathematical model for analysing the migration of radionuclides, has been developed. |4, 7, 8|

Results presented in this paper are examples of results obtained in a 10 year mortar and concrete testing project, which will influence the design of the engineered trenche system for future central Yugoslavian radioactive waste storage center. |7, 8|.

Transmission computerized tomography (TCT) was used for nondestructive characterization of nuclear waste packages (HLW, ILW). Ready conditioned waste packages may be tested to reveal density fluctuations, cracks, fluids etc. Emission computerized tomography (ECT) combined with TCT gives the activity distribution corrected for density variations within the waste packages.

The effects of accelerated gamma and alpha irradiation experiments on cement grouts alone and grouts containing simulated combustible plutonium contaminated material (PCM) at a number of dose rates and under various environmental and confinement conditions are considered. Cracking and spallation can occur over a wide range of water/cement ratios at a high dose rate of 2.8 Gys-1 for grouts based on blast furnace slag compositions. Grouts based on pulverised fuel ash do not show any signs of spallationand cracking. One possible reason for this is the accumulation of radiolytic gases to high pressures in pores in the cement. It has been demonstrated that under some circumstances, at the Laboratory scale, PCM waste form can swell sufficiently to rupture its container and may disintegrate.

The leaching of solid inorganic waste from loaded concrete or cement by Incoming water can be described in terms of a “steady-state” outward diffusion of the saturated solution, formed inside the pores. The derived equations permit the prediction of long-term leaching behaviour. Radiotracer experiments enahle the determination of the parameters involved.

The influence of γ-irradiation on the properties of two types of bituminous reprocessing waste has been studied. Standard test methods were adapted for use with radioactive samples. Measurements of penetration, gas generation, thermal stability and leaching of irradiated and non-irradiated samples have been compared.

The Swedish repository for low and intermediate level waste- SFR was ready for operation in early 1988. The license to start operation for the low level waste part was given in the spring of the same year. A very detailed safety analysis of the repository has been made by the SWEDISH NUCLEAR FUEL AND WASTE MANAGEMENT CO.- SKB. The main issues of this work are presented in this paper.

Requirements on radioactive waste of negligible heat generation with respect to transuranic elements, derived from the safety analysis of the operational phase of the Konrad repository, are described and resulting activity limitations are discussed. For assessing the long term safety, the calculation of individual doses in connection with the very slow water movements and, additionally, toxicity comparisons to other waste and to the natural toxicity at the site are discussed. Some information is given on the procedure of taking into account the activity limitations in an optimal way when planning emplacement campaigns.

In the frame of the European R&D program on “Management and Storage of Radioactive Waste” assessments for HLW repositories in different geological formations have been performed with realistic data. Part of this action was the discussion of altered evolution scenarios for a repository in a salt dome.

For the considered HLW repository a combined altered evolution scenario with brine intrusion via the main anhydrite vein and brine intrusion from undetected brine pockets has been considered. For this scenario limited intrusions of brine at very early times are assumed, which inundate boreholes and parts of the repository. Radionuclides are leached out of the waste forms and contaminate the brine. At later times brine intrusion via the main anhydrite is assumed to take place. Under certain conditions contact between both amounts of brine is possible, and contaminated brine reaches the overburden. Here, the radionuclides are swept away with the groundwater flow.

Decisive for the transport of radionuclides from the repository to the geosphere are several interacting effects (e.g. convergence, permeability, pressure) which determine the movement of brine through the drifts and seal-ings. For the parameters involved in the modelling of physical and chemical effects as realistic data as available (best estimate) were used. Local sensitivity studies show that with the best estimate values for some parameters the combined intrusion scenario is just at the transition point between release and no release.

Results are obtained showing time histories for the decrease of the voids in the repository, release rates for relevant nuclides injected into the overburden, and for dose rates. Dependences on the model parameters are discussed and most relevant parameters are identified.

Performance assessment studies have been undertaken on the geological disposal of high-level waste in a clay layer in the framework of the CEC project PAGIS. The methodology applied consists of two consecutive steps : a scenario and a consequence analysis. The scenario analysis has indicated that scenarios of normal evolution, of human intrusion, of climatic change, of secondary glaciation effects and of faulting should be evaluated. For the consequence analysis as well deterministic “best estimate” as stochastic calculations, including uncertainty, risk and sensitivity analyses, have been elaborated.

The calculations performed show that most radionuclides decay to negligible levels within the first fewjneters of the clay barrier. Just a few radionuclides, 99Tc, 135Cs and 237Np with its daughter nuclides 233U and 229Th can eventually reach the biosphere. The maximum dose rates arising from the geological disposal of HLW, as evaluated by the “best-estimate” approach are about 10−11 Sv/y for river pathways. If the sinking of a water well into the 150 m deep aquifer layer in the vicinity of the repository is considered together with a climatic change, the maximum calculated dose rate rises to a value of 3×10−7 Sv/y. The maximum dose rates evaluated by stochastic calculations are about one order of magnitude higher due to the considerable uncertainties in the model parameters. In the case of the Boom clay the estimated consequences of a fault scenario are of the same order of magnitude as the results obtained for the normal evolution scenario. The maximum risk is estimated from the results obtained through stochastic calculations to be about 5×10−8 per year. The sensitivity analysis has shown that the effective thickness of the clay layer, the retention factors of Tc, Cs and Np, and the Darcy velocity in the aquifer are parameters which strongly influence the calculated dose rates.

CONDIMENT is the source code used in MELODIE, the overall French computer code for risk assessment of nuclear waste repositories in geological formations. This code models the diffusion-convection of elements released by a waste package, up to a distance of a few meters from the package. Two versions have been developed simultaneously:

- CONDIMENT2 deals with the case of a single element. This version is more specifically designed for vitrified high-level wastes. The boundary conditions are furnished by studies on aqueous corrosion of French nuclear glass R7T7.

- CONDIMENT3, deals with two ions that are liable to precipitate. This version is more specifically designed for wastes immobilized in cement.

CONDIMENT3 is verified in a configuration for which an analytical solution exists.

This national safety study of a repository in the Atlantic Ocean uses as much input from German research work as possible. It is based on a computer code IMPONADOR which can be run on a personal computer. It also introduces a new way of presenting information about radionuclide confinement to individual compartments to facilitate the understanding of release processes. The work is part of the international cooperation within the Seabed Working Group of the OECD/NEA.

Recent advances in complementing the capabilities of our safety assessment code development allow for an explicit investigation of the consequences of parameter uncertainties by Monte Carlo simulation of the disposal system. The implemented methods aire described and applied to impact assessment of a brine intrusion scenario on a high level waste repository in a salt dome. Results display the strong effect of parameter uncertainties on the analysis results. With help of several sensitivity estimators input parameters are ranked according to their influence on output uncertainty.

A series of calculations of radionuclide release was performed with the AREST and SYVAC-Vault models (SVM) in order to assess concurrance. Specifically, the effects of precipitation and decay chain in-growth on the predicted release of nuclides from waste packages containing spent nuclear fuel were compared between each code. The results for maximum release rates generally agreed within a factor of 10. The differences in results can be explained based on the differences in geometry and boundary conditions between the two codes. Both codes showed nearly identical enhancement factors in release rates of uranium-series nuclides (U-238, U-234, Th-230, Ra-226) arising from the effect of decay-chain in-growth. Calculated enhancement factors in release rates for precipitation of a new uranium-bearing solid within the waste package were also in good agreement between AREST and SVM.

In order to perform the safety evaluation of nuclear waste repositories, a global model, called MELODIE, is currently being developed at the CEA/IPSN, in collaboration with other CEA teams and ENSMP (Ecole Nationale Supérieure des Mines de Paris). The version now in operation allows the assessment of the radiological consequences due to a repository located in a granitic formation for a period of several hundred thousands of years.

The calculations are based on models which represent the following physical and chemical phenomena:

– release of the radionuclides from the waste form and through engineered barriers,

– radionuclide transport through the geosphere,

– radionuclide behaviour in the biosphere.

Three separate models have been developed for each of these subjects; they are integrated in the code through a modular flexible dataprocessing structure.

In addition, a sensitivity and uncertainty analysis algorithm has been implemented into the code, which allows the evaluation of the influence of the parameter values on the result and of its uncertainty.

After a quite general description of MELODIE, we intend to present what has been done with it, mainly the calculations performed in the PAGIS (CCE) exercise: global dose calculations and ranking of the most important parameters through the sensitivity analysis.

We also note the studies performed with the geosphere module of MELODIE (METIS), especially the participation to the HYDROCOIN (0ECD/NEA) exercise. In addition, the main future development axes of MELODIE are outlined.

A joint US/FRG nuclear waste repository simulation experiment was performed at the Asse Salt Mine in the Federal Republic of Germany (FRG). The High Level Waste (HLW) disposal in boreholes was simulated by the simultaneous emplacement of electrical heaters and cobalt-60 sources at four individual test sites located in a special underground test room at the 800 m-level.

In order to resolve the issues of rock mass/waste package interaction the temperature field, brine migration into the heater boreholes, borehole gas pressure and composition, and rock mass stresses and displacements were monitored during the test. In order to validate computer code predictions the acquired data were compared to calculational results. Corrosion specimens remained in the heater boreholes during the course of the experiment and were afterwards examined.

A first generation of underground migration experiments is described, consisting of labelled clay cores emplaced in boreholes drilled in the Boom clay formation. The boreholes are sealed by natural convergence of the clay and porewater percolates through the labelled clay cores which are consolidated in situ. After monitoring of the radioactive tracers in the percolating porewater, the experiment is retrieved from the borehole and the tracer profile is measured in the clay cores. With the exception of accelerated porewater flow, due to the existence of a high hydraulic head around the underground gallery, the experimental conditions are close to those expected in the far-field of a closed repository for radioactive waste. The main advantage of this approach is the availability of real porewater during relatively long-term experiments. Results are reported for the experiments performed with europium and strontium tracers.

High-level nuclear wastes are heat generating wastes. Heat will be transferred to the surrounding salt formation. This heating of the host rock will result in an increased temperature and in stress changes.

From 1983 through 1985 two underground tests were conducted in the Asse Salt Mine (Federal Republic of Germany) in which among others thermally induced stress changes were investigated.

In the Temperature Test 6 which was conducted at the 750 m-level only electrical heaters were used to simulate the heat generating high-level waste. Three hydraulic stress monitoring stations were arranged in the un-heated area at a distance of about 40 meters to the central heater. Measurements using AWIO flat jacks were also carried out in the heated region of the Temperature Test Field 6 (TVF 6).

The joint US/FRG Brine Migration Test (BMT) - a nuclear waste repository simulation experiment - was performed at the 800 m-level and used radioactive sources and electrical heaters to impose the heat load on the host rock. Stress change measurements during this experiment were performed using hydraulic pressure cells and straingaged stressmeters.

During 1985 an experimental area for the first in-situ test disposal of high-level radioactive sources was created at the 800 m-level within the Asse anticline structure. Thirty high-level radioactive canisters will be emplaced in six down-boreholes located in two test galleries in the experimental area. There is one additional borehole in each gallery equipped only with electrical heaters. Hydraulic pressure cells, AWID flat jacks, and straingaged stressmeters were installed in the HAW-test field for stress change measurements.

The paper gives a quantitative description of the chemical reactions, the dissolution and precipitation processes which take place when water or brines come into contact with the main mineral assemblages of the German Zechstein salt formation, rock salt with Ca-sulfates, and the potash beds hartsalz and carnallitite. The results of two large scale in situ dissolution tests are presented. These results have been successfully reproduced by computer modelling using the geochemical code EQ3/6.

The Project Gewähr 1985 (PG'85) safety analysis considered the practicality of HLW disposal in the crystalline bedrock of Northern Switzerland. More recently, an alternative disposal option in the argillaceous sediments overlying the crystalline basement has been examined. This paper examines some aspects of the performance of the near-field engineered barriers in an argillaceous environment, with specific comparison to the PG'85 crystalline case. Two phenomena are considered in particular detail: oxidation of the host rock during the operational phase and the related failure mode of the steel tunnel liner.

In situ waste package performance experiments involving simulated (non-radioactive) defense high-level waste (DHLW) containers have been in progress since late 1984 at the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve full-size, simulated DHLW containers of several metals and designs emplaced in the WIPP bedded rock salt. These test containers are surrounded by granular backfill (packing) materials, have in many cases been intentionally injected with brines, and are heavily instrumented. A majority of the test packages also contain nonradioactive DHLW borosilicate glass waste form, either within the container and/or outside of it. The primary purpose of these WIPP simulated DHLW experiments is to evaluate the in situ durability and performance of all waste package engineered barrier materials, and to perform package concept validation testing.

Twelve of the test DHLW containers, emplaced in WIPP test Room B, have been in heated operation since 1985 and had a maximum surface temperature of about 190°C. These containers were recently retrieved, after about 3 years of heated exposure, for detailed posttest laboratory analyses of: general corrosion and metallurgical degradation, waste form and backfill materials alterations, and other rock salt-brine-barrier materials near-field interactions with the “repository” geochemical environment. Test canisters and overpacks made of ASTM Grade-12 titanium showed essentially no visible degradation in either the base metal or welds; cast mild steel A216/WCA over-packs have suffered some uniform corrosion. Significant degradation of the removed instruments and associated test apparatus has been found: pieces of stainless steel (both 304L and 316) apparatus have undergone extensive stress-corrosion cracking failure and non-uniform attack; Inconel 600-sheathed instruments have undergone both extensive uniform and localized (pitting) attack. Granular backfill materials have been significantly compacted by creep closure to about a density of 2 kg/m. Laboratory analyses are still in progress. Further details on these materials results plus instrumentation data and other in situ WIPP waste package test observations are discussed.