Time-temperature conditions and water chemistry on illitization at the Murakami deposit in central Japan were determined. The extent of the illitization and timetemperature condition estimated were as follows: the duration for conversion from 0 to 80%illite (volumetric ratio) was approximately 3.5Ma in the temperature range from 340 to 100°C. Conversion from 0 to 40% requires approximately 3.0Ma in the temperature range from 240 to 100°C. During 2.0Ma in the temperature range from 160 to 100°C, however, illite was scarcely observed. Water chemistry is estimated from two approaches, namely laboratory experiment and numerical analysis. The former is an interactive experiment between seawater and the tuff of the deposit. The latter is a calculation based on the difference of bulk composition between illitized and non-illitized tuff. The extent of each ionic concentration is inferred to be as follows: K+; 560 to 6400, Mg2+; 800 to 1700, Ca2+; 360 to 2900, Na+; 9400 to 15000 (mg/l).

Molecular water diffusion in natural obsidians and tektite was investigated in vapor hydration tests performed between 75 and 230°C for up to 400 days. Reaction progress was monitored using measurements of the birefringent hydration layer, an alteration feature associated with strain caused by molecular water diffusion in obsidians. The hydration rate constants and temperature dependence of the reaction are strongly correlated with the logarithm of the initial total water content of the glass. These values have been quantified for conditions relevant to the potential Yucca Mountain repository. The low initial total water concentrations of Savannah River Lab nuclear waste glasses produced at the bench-top scale help to minimze the effects of molecular water diffusion in waste glasses. The results of this study indicate that molecular water diffusion does not dominate waste glass reactions under conditions considered in this study. However, it is unknown whether molecular water diffusion will be important under other reaction conditions, especially longer time periods.

The source term for nuclear waste repository performance assessments can be constrained by the solubilities of radioelement-bearing solids and/or the rates of release of radioelements from nuclear waste forms. Both solubility and rate limits for the proposed repository at Yucca Mountain, Nevada, can be assessed using information from the natural analog at PeÑa Blanca, Mexico. Petrographic and field relations indicate that uraninite oxidation and transformation to secondary uranyl silicate minerals have been rapid relative to mass transport of uranium out of the PeÑa Blanca system. The rate limiting process for uranium removal is likely to be advective transport in groundwaters with uranium contents controlled by interactions with uranyl silicate minerals such as uranophane. A maximum limit on the rate of uraninite oxidation at PeÑa Blanca is calculated to be 0.032 tons of UO2 per year using geologic constraints on the amount of oxidation and the available time.

The Koongarra uranium deposit in the Northern Territory (NT) of Australia is being studied to evaluate the processes and mechanisms involved in the hydrological/geochemical alteration of the primary uranium mineralisation and model the formation of the secondary uranium orezone and the dispersion fan. The project includes studies of groundwater flow through the orebody, the chemistry of intersecting groundwaters and the uranium distribution and mineralogy of the orezones. The work has resulted in solid phase and groundwater databases, and hydrological data from field and laboratory measurements. These are being used to test a range of geochemical, hydrological and transport models.

This paper discusses the experimental evaluation of the hygro-thermal parameters governing coupled heat and moisture movement within a moist clay buffer, which is developed as an engineered barrier for the isolation of heat emitting nuclear waste containers in either a borehole environment or in an in-room emplacement scheme.

Field heat experiments with kaolinite/smectite clay surrounding heaters in boreholes were conducted for 0.7 and 4 years with temperatures up to 170-180°C. The short test gave a high degree of water saturation even in the hottest part (> 75 %) and almost no change in physical properties and mineral composition. The long test gave a dry inner zone of claystone, indicating gas formation, and rich precipitation of silica/aluminum compounds and sulphate minerals. Brittleness characterized the hot parts and stiffening occurred also in the colder parts due to precipitation of silica and aluminum.

Alteration of compacted sodium-bentonite (Volclay MX-80) caused by groundwater in simulated repository conditions for high level radioactive waste, was studied in an experiment where bentonite (wrapped by a copper cylinder) was let to react with two types (A,B) of synthetic granitic groundwater that are distinguished by their initial concentration of potassium and chloride. The reaction took place in ambient conditions at a temperature of 75 °C and proceeded during several time intervals up to 36 months.

At the end of each time interval the water was chemically analysed for determination of possible changes in composition. Chemical and mineralogical changes in the bentonite were investigated by using NH4Cl extractions, XRD and microprobe (SEM, EDS) analyses and were studied as a function of the reaction time (months) as well as of the distance (mm) from the contact front with water.

Sodium ions were found to migrate out from the bentonite while being replaced by other cations such as calcium, magnesium and to some extent, particularly during the reaction of the bentonite with water B, by potassium. No clear evidence was found for the fixation of potassium ions in the interlayer position of montmorillonite clay and the transformation to illite. The main mineralogical change in the bentonite was from Na- to Ca-rich montmorillonite. Secondary processes were the dissolution-precipitation of sulphur compounds, dissolution of gypsum and carbonates and the dissolution-precipitation of copper compounds.

In cooperation with the Ministry of Research and Technology we investigated the influence of hydroxylapatite (HAP) added to the backfill material of a radioactive waste disposal located in a salt dome. The time dependence of the sorption and precipitation of radionuclides has been studied in three different praxisrelevant salt brines under variation of temperature and pH-value.

Performing batch experiments uranium and thorium phosphates were precipitated from UO2(NO3)2 and Th(NO3)4 solutions at 75°C, the amounts of the relevant actinide compounds also exceeding that of HAP. The isolated precipitates were investigated by use of X-ray-powder-diffractometry and SEM. For the various uranium systems two mineral phases were obtained. From solution II and also from Q-brine the mineral saléeite Mg(UO2)2 (PO4)2 · 9H2O was formed. From solution III meta-autunite Ca(UO2)2(PO4)2 · 6H2O, a dehydration product of autunite, was precipitated.

The paper discusses the influence of creep deformations of either engineered or natural geological barriers on the long term settlement of waste containers located within them. The paper describes, briefly, the potential application of numerical procedures for the estimation of the creep settlement rigid cylindrical containers embedded in engineered and natural geological media.

A model has been developed which describes the steady state transport of dissolving species of radionuclides from a single canister in a repository for spent nuclear fuel into the passing water in fractures in the surrounding rock. The steady state transport of nuclides is described by a network of transport resistances, coupled together in the same way as an electrical circuit network. With the model a number of calculations are done for various sets of fracture geometry data. The calculations indicate that the resistance network model gives results comparable to those of a complex 3-dimensional numerical model. The present model is very simple and fast.

We have developed a coupled transport and reaction computer program which is adapted to handle sharp reaction fronts. Such fronts develop in the ground and typical examples are redox fronts and dissolution/precipitation fronts. The program is developed to describe the chemical evolution around radioactive waste repositories but it is applicable to other situations as well. To validate the program we compare our results with data from other geochemical equilibrium programs, such as EQ3/6 [1], PHREEQE [2] and transport programs such as CHEQMATE [3]. Natural analogues like the Osamu Utsumi mine at Poços de Caldas in Brazil and the uranium ore body at Cigar Lake in Saskatoon, Canada, are studied using the program.

Maximum aqueous concentrations of six radionuclides (Sr, Np, U, Th, Pu and Tc) near a disposal site with a quartz sand or clay engineered barrier in a granitic environment, were calculated in a previous study by computer modelling. A sensitivity study has been carried out in order to assess the importance of three factors in determining the maximum concentrations likely to be found in solution: the maximum temperature reached, water salinity, and the dissolution constants used. In the circulating fluid, final concentrations for Pu, Tc and Th depend mostly on the final temperature. The solubility of Sr is influenced by the fluid salinity. The final concentration of Np is affected by the temperature of the waste disposal site. Uncertainty in the dissolution constants causes the greatest relative variation in calculated U concentrations.

Adequate representation of fracture-matrix interaction during episodic infiltration events is crucial in making valid hydrological predictions of repository performance at Yucca Mountain. Approximations have been applied to represent fracture-matrix flow interaction, including the Equivalent Continuum Model, which assumes capillary equilibrium between fractures and matrix, and the Fracture-Matrix Model, which accounts for nonequilibrium fracture-matrix flow. We analyze the impact of matrix imbibition on episodic nonequilibrium fracture-matrix flow and transport for the eight major hydrostratigraphic units in the unsaturated zone at Yucca Mountain.

The effective diffusivities of carbon dioxide through Tiva Canyon tuff and the lower lithophysal zone of the Topopah Spring tuff (outcrop sample of the layer above the proposed nuclear repository site layer) were determined using a steady-state method (counter diffusion). The diffusivity of carbon dioxide through the Tiva Canyon and lithophysal zone tuffs increased with temperature and decreased with percent saturation. The following correlation was obtained to estimate the effective diffusivity of carbon dioxide through the Tiva Canyon tuff as a function temperature and percent saturation.

De = 1.2168×10−2 − 3.7713×10−5 T + 9.9510×10−8 T2 − 4.2220×10−5 × (Percent Saturation)

where De is in cm2/sec and T in K. These units hold for all other correlations in this paper.

The effective diffusivity of carbon dioxide through the lower lithophysal zone of the Topopah Spring tuff (layer right above the proposed repository site) also increased with temperature and decreased with percent saturation. For this layer, the following correlation was obtained to estimate the effective diffusivity of carbon dioxide as a function temperature and percent saturation.

De = − 1.119×10−3 + 1.2512×10−5 T + 1.8288×10−9 T2 × 2.0725×10−5 × (Percent Saturation)

In the safety assessment of high level radioactive waste disposal, a significant retardation of nuclides can be expected. When radionuclides released from an underground repository are transported with the moving groundwater along cracks in the rock, the radionuclides will be retarded by not only adsorption on the surface of the cracks but also diffusion into the micropores of rock matrix.

Diffusivities of U(VI) were measured in water-saturated granite that was mined in the middle of Japan. The measured penetration profile of U(VI) was composed of two parts. Those were a steep slope near the surface and a gradual slope in the deeper part. This profile was successfully explained by considering two diffusion paths in granite. One diffusion path was possibly a fissure with a width of a few microns and another was a network of submicron pores. The volume of submicron pores was approximately 80% of the total pore volume in the granite. The orders of magnitude of apparent diffusivities for U(VI) were 10−13-10−12 m2/s through the fissure and 10−15 m2/s through the network of pores. The ratio of geometrical factors in the two paths was also calculated from each apparent diffusivity and was almost unity. This ratio calculated suggests that the pore network in the granite matrix has a similar shape to the fissure, thus the microstructure of the granite seems to be fractal.

Current plans for the disposal of low- and intermediate-level radioactive waste in the U.K. comprise emplacement of steel or concrete containers of waste in a deep underground repository where they will be surrounded by a cementitious backfill. Research is in progress to assess the effectiveness of steel and concrete as barriers to radionuclide release from the repository. A joint experimental and modelling approach is used. Studies of the transport properties of dissolved radionuclides and gases through concrete, and the ability of the concrete to control the pH in the repository are described. Localised and uniform corrosion of carbon and stainless steels are investigated in the high pH conditions of the repository. The research shows that engineered barriers of concrete and steel can provide useful physical containment in the short term and contribute to chemical containment for hundreds of thousands of years.

The Nirex Safety Assessment Research Programme (NSARP) comprises scientific research to support the safety assessment of a deep geological repository for low- and intermediate-level radioactive wastes. The NSARP addresses several areas of research, one of which is a continuing investigation of the behaviour of radioelements in a cementitious environment. This paper draws together recent experimental measurements of solubility and sorption under such conditions, along with the results of interpretive thermochemical and sorption modelling. Under conditions similar to those expected in the repository, the predicted solubility of selenium is about 3xl0−8M. Experimental measurements confirm that at low redox potential the solubility of selenium is substantially less than that previously observed under an air atmosphere, and is closer to the predictions of the thermochemical modelling. The solubility of sodium stannate has been studied after heating solutions at 50, 80 and 200°C. Crystalline tin solids were formed at all temperatures. For solutions heated at 200°C the solids were identified as cassiterite. The solubilities of all of the crystalline solids were 1x10−6M to 1x10−9M which are at least five orders of magnitude lower than that of the original sodium stannate. Distribution ratios (Rd) between 1x103 and 5x103 ml g-1 have been measured for the sorption of tetravalent technetium on to a cementitious material. Preliminary results from an in-diffusion experiment involving americium and an intact cement sample are in good agreement (at 1sx104 ml g-1) with Rd values determined from previously reported batch sorption experiments.

We are presenting applications of a new technique for the study of porosity and diffusivity in rocks employing impregnation with labeled polymethylmethacrylate. We studied samples of different rock types and rock that has undergone various types of alteration. We have focussed special attention on fissures and fissure fillings. Association of porosity with certain mineral phases has been addressed. The spatial distribution of rock porosity exhibits a great diversity. It is a rock-typical parameter. Altered rock is, usually, characterized by inhomogenities in porosity and anisotropy of migration pathways.

One of the options envisaged by ANDRA for some of the engineered barriers in a geological repository in hard rock is the use of a permeable granular material (clay/sand or clay/crushed rock mixtures).

This choice will lead to modification of the flow distribution around the repository borehole and the consequences on the migration behaviour in the near field need to be evaluated.

The aim of this study is to develop a borehole model which describes the release in the borehole from the waste matrix through the engineered barrier. This model can be readily integrated in a near field model of the waste storage area.

A study of near-field radionuclide migration is presented. The study [1] has been performed in the context of the SKB91 study which is a comprehensive performance assessment of disposal of spent fuel. The objective of the present study has been to enable the assessment of which nuclides can be screened out because they decay to insignificant levels already in the near-field of the repository.

A numerical model has been used which describes the transient transport of radionuclides through a small hole in a spent fuel canister imbedded in bentonite clay into a fracture in the rock outside the bentonite. Calculations for more than twenty nuclides, nuclides with both high and low solubility, have been made. The effect of sorption in the bentonite backfill is included. Materials data for bentonite where taken from [2]. The size of the penetration hole was assumed to be constant up to the time when the calculations were terminated, 500 000 years after the deposition. The mass transport rate is controlled by diffusion. The model is three dimensional.

This paper describes the geometry of the modelled system, the assumptions concerning the transport resistances at the boundary conditions, the handling of the source term and obtained release curves.