Several agencies in charge of nuclear waste management consider the use of compacted bentonites as engineered barriers for the migration of radionuclides towards the geosphere. The geochemical processes occurring in the clay buffer system are of capital importance for the assessment of a HLW repository performance. Therefore, it is necessary to develop tools able to deal with the complex interactions between water and bentonite, submitted simultaneously to a thermal field and water flow. These tools may be referenced as thermo-hydro-geochemical (THG) modelling codes

A full-scale “In situ” test is currently being conducted as part of the FEBEX project. This test is being carried out at the Grimsel Test Site in Switzerland and reproduces the “AGP-Granite” concept of the Spanish agency ENRESA for HLW repositories. Two heaters, of the same dimensions and weight as the canisters in the reference concept, have been placed in a horizontal drift of 2,28 m diameter and backfilled with a total of 115.7 t of highly compacted bentonite blocks, up to a total test length of 17 m. More than 600 sensors have been installed in the experiment, which runs basically in an automatic mode and is being remotely controlled and monitored from Spain. This paper addresses the engineering, installation and instrumentation aspects of this test. In addition, some preliminary results are shown

Plasma processes become more and more attractive for the treatment of radioactive low/medium level waste. Plants with throughput rates ranging from 20 kilograms to several hundred kilograms per hour are under construction or under evaluation in different countries. One major installation is under construction at the ZWILAG-site in Wurenlingen (Switzerland). ZWILAG (Zwischenlager Wurenlingen AG) is a company being responsible for treatment and intermediate storage of all radioactive wastes occurring in Switzerland. Thermal destruction (pyrolysis) of organic material and melting of the inorganic residuals and inorganic feed stock can be done simultaneously in a plasma heated vessel. The following output materials are obtained: Metal, glassy slag, off-gas. Depending on feed stock and storage requirements these outputs can have different properties. The distribution of the activity is of particular importance. Isotopes like cobalt-60 (as a metal) partition almost quantitatively to the metal phase. Volatile isotopes like cesium-137 partition to some extent to the off-gas, they can however be retained in the slag to a high extent by employing a two stage process. For some applications the activity in the metal can be kept below the activity tolerance limits making the metal suitable for even non nuclear applications. The most important property of the slag is its leaching behaviour. Mechanical properties can be of importance, too. The paper will present examples from different types of radioactive waste: typical low level mixed waste, ion exchange resins and zeolites, mixtures of concrete and metal, contaminated soil, material contaminated with natural activity and others. Composition of the materials, activity distributions, data on leaching behaviour and some mechanical properties are presented

Compacted bentonite is a candidate buffer material in geological disposal of high-level radioactive waste. The transport of radionuclides in compacted bentonite is dominated by diffusion, because of its very low permeability. In this study, we focused on the grain size of clay mineral, which is considered to be closely related to the formation factor in the pore water diffusion model[1,2]. The apparent diffusion coefficients (Da) of HTO and cesium ions in compacted clays were determined using montmorillonite samples with different grain size and dry density, and the effect of the grain size on diffusion behavior was discussed.

Bentonite could be one of the materials used in Engineered Barrier Systems. The aim of this work is to assess its sorbing properties using surface complexation modelling of Cs+, Ni2+ and Yb3+ sorption on single oxides

Hydrothermal experiments on saturated barrier were conducted for 6 months and 4 years in the granite of the STRIPA mine (Sweden). The French clay Fo-Ca 7 (80 % Kaolinite/Smectite mixed-layer) was compacted (25 MPa) and placed around a steel heater in a 20 cm diameter borehole. The temperature was about 170°C at contact with the heater and only 80°C at clay/granite interface, thus giving a thermal gradient around 10°C/cm

Concrete and other cement-based materials are increasingly utilized as major structural components of the disposal facilities for low-level and intermediate-level radioactive waste (LLW/ILW). At the same time. cementitious materials function as engineered barriers against migration of radionuclides and other hazardous compounds [1]. Taking into account the expected operation times of such constructions, the long-term prediction of environmental interactions and stability of concretes is important for the development of reliable facilities for LLW/ILW disposal. Thermodynamic approach has been widely used to promote understanding of chemical phenomena in cementitious systems. Recently, several authors attempted to demonstrate usefulness of computer codes for description of the available experimental data and prediction of cement/water equilibria [1-5]. These calculations were based on the Law-of-Mass Action - Reaction Stoichiometry Algorithm (LMA-RSA), widely applied so far in speciation modeling [6]

As part of the evaluation of the safety of the geologic disposal of HLW glass, a new in situ test on α-active glass will be started up in the underground research facility HADES (SCK•CEN, Mol, Belgium), situated in the Boom Clay formation, at a depth of 223 m. Besides a demonstration of the safe disposal of HLW glass containers in Boom Clay formation, the overall objective of this CORALUS test (CORrosion of Active gLass in Underground Storage conditions) is to study the performance of active HLW glass specimens in direct contact with different types of backfill materials, under conditions as representative as possible for those expected to prevail in a geologic repository in the Boom Clay formation. In order to meet this objective, the CORALUS test will integrate not only the experience of previous in situ tests performed at SCK•CEN and other research institutes, but also the expertise of the cooperating institutes. The paper presents the motivation for this new in situ test, its objectives, and the current status of its design

In the middle of 80-s the Moscow SIA “Radon” initiates a program on long term tests of vitrified intermediate level radioactive waste from nuclear power plants with different types'of reactors and with different composition of waste. Glass blocks were prepared and disposed for tests in experimental shallow ground repository and on open testing area. Borosilicate glass as matrix material is investigated in order to obtain data on real behaviour of glass [1]. The waste glass was poured in carbon steel rectangular containers therefore containers also were subjected to aqueous and atmospheric corrosion. Exposure conditions at testing site are varying depending on season. They can be described only as average parameters with some uncertainties.

The durability of concretes under irradiation must be investigated in connection with research on radioactive waste conditioning (matrices, canisters) and disposal (engineered barriers). The weak element of the concrete, the water present in the pores, or in the hydrated compounds, is the source of radiolysis, the chief cause of radiation damage. The emission of associated dihydrogen, or radiolytic breakdown by internal overpressure, are two objective risks for which the many factors involved and the heterogeneous nature of the material make any long-term forecast difficult.

Acidic salts of polyvalent metal antimonates(M-antimonates(M-Sb); M=Cerium, Tin, Titanium and Silicon) have been synthesized under different conditions( 1,2). The products were characterized by powdered X-ray diffraction patterns, Differential Thermal Analysis(DTA) and infrared spectrometry. Chemical stability of the product materials were studied by mixing 50 mg of the prepared materials with 50 ml of the desired medium(H20, HNO3 and HCI) with determination for the concentration of the metalions dissolved(Sb, Ce, Sn, Ti and Si) which showed high chemical stabilities in water and acid media up to I OM HNO 3 and 6M HCI.The data of the distribution coefficients showed that the selectivity decrease in the order; Eu3+> Co2+> Sr2+> Cs+ for cerium(IV) antimonate(CeSb), tin(IV) antimonate(SnSb) and silicon(IV) antimonate(SiSb), while the selectivity decrease in the order UO22+>Th4+ for titanium(IV) antimonate(TiSb). Besides, the effect of EDTA, citric acid, ferric nitrate and sodium nitrate as interfering ions in solution on the distribution coefficient of the studied cations for CeSb and SnSb were studied under certain conditions. The results showed that, the Kj values of CS+, Co2+ and Eu3+ are decreases with increasing the concentrations of ferric and sodium nitrates in the feed solutions for both materials and also Kd values are effected to some extent in case of EDTA and citric acid. Based on the obtained results, practical separation experiments on column were performed(3).

The following figures represents the break-through curves for CeSb and SnSb columns for some radionuclides 60Co (10-3M ) as example. From the above studies we conclude that, CeSb and SnSb have been successfully used for the quantitative separation of cesium which suggests its use in the treatment of radioactive liquid waste.

It is the aim of the present poster to show two approaches for assessing the influence of the alkaline degradation of cellulose on the sorption of tri- and tetravalent radionuclides. The two approaches differ in their degree of conservativity, one is a conservative, the other a realistic model case. The result of both examinations is expressed in terms of sorption reduction factors, valid for a situation where cellulose degradation products are influencing the sorption of radionuclides on cement.

In Korea Depleted Uranium(DU) is used in manufacturing a metallic nuclear fuel for the Korean Multi-Purpose Research Reactor(KMRR). In the manufacturing processes it produces DU chips and scraps as a waste material which composed of U-Ti, U-Zr, U-Mo and U-Si intermetallic compound. In this study Air Controlled Oxidizer(ACO) has been developed which facilitates DU to be converted into U308, the most stable form of uranium. Since DU chips oxidize rapidly and their heat of oxidation is very high(4.199kJ/g, U3O8), the inside temperature of the oxidizer is likely going up rapidly. Therefore the oxidizer must be able to be cooled properly or temperature increasement of the oxidizer must be under control. Kang et al.[1] reported for the oxidation of U-0.75wt%/o Ti chips in air that U308 was detected at the temperature above 350°C. And they[2] also reported that the maximum heat generation per unit time during oxidation was as follows:

where Q was the maximum heat generation per unit time

WDU was the weight of DU loaded in ACO

dw/dt was the reaction rate

Vair was the flow rate of input air

R was the universal gas constant

and T was the absolute temperature.

From eq.(1) the maximum heat generation per unit time during oxidation is only function of the weight of DU loaded in ACO and the oxidation rate which is dependent on the oxidation temperature or the flow rate of input air.

The ACO consists of an air flow meter, an air heater, an oxidation chamber with inner heater(capacity 7.5kW), an ash collection tank, a fly ash collector, a pressure gauge, a safety valve, and a soaking tank. The air flow meter is used to control the flow rate of input air below theoretical air requirement limit for the complete oxidation of DU. The inner heater is used to heat the inside of the oxidizer to an optimum oxidation temperature. The ash collection tank is used to collect uranium oxide powder after completion of oxidation. The fly ash collector is used both to collect flying ashes and to condense vaporized uranium oxide. Also, in ACO, DU chips are not ignited directly in order to prevent rapid temperature increasement. The oxidation environment only is achieved by heating the inside of oxidizer.

To find effect of the oxidation temperature on the temperature of the oxidation chamber during treatment of DU, we conduct the experiment by changing heating rates of inner heater, 3, 4, 5 and 6kW, respectively. We conduct experiments for 120 minute with 2/min input air. However, it turned out that the complete oxidation is reached within 60 minute. After complete oxidation the weight gains of the DU chips is from 4.5 to 5.0 wt0/o and the DU chips are pulverized and they are converted to U 08 and 979 Mat. Res. Soc. Symp. Proc. Vol. 506 1998 Materials Research Society During the oxidation, maximum temperature increases to 470, 497.5, 572.5 and 6771C for heating rates 3, 4, 5 and 6kW, respectively. As the temperatures of the oxidation chamber outside surface are not exceed 1501C, however, DU chips are treated safely. In each experiment, weight before and after oxidation, the oxide forms of the product and the maximum temperature in the oxidation chamber during oxidation are shown in table 1. The maximum temperature profiles of the chamber inside and surface for time and heating rates are shown in Fig. I and 2, respectively.

Lysimeters are widely used to investigate the release of contaminants from waste or soil samples and for waste degradation studies [1-2]. This paper presents results from the first stage of a comprehensive, long-term lysimeter programme currently being undertaken by BNFL [3] to investigate the release of radionuclides from LLW in lysimeters simulating unsaturated trench disposal conditions. Data generated from these experiments will be used to support mathematical modelling of near field disposal systems.

The three lysimeters used were small-scale, gas-tight columns made from PVC tubing. Two sizes were utilised: the 70-litre lysimeter A was filled with 11.5 kg of waste and the 30-1itre lysimeters B and C contained 5.05 kg and 4.45 kg respectively. Average waste density was 0.17 te/m3. The LLW was taken from R&D labs at the Sellafield site and included materials such as rubber gloves, plastic bottles and bags, paper, card and tissues. Lysimeter B had a higher cellulose content (85%) compared to A and C (50%). After loading, the lysimeters were sealed and fitted with a gas vent to prevent a build-up of internal gas pressure. The lysimeters were operated in the conventional single-pass mode. The irrigation rate of 761 mm per annum was based on the annual rainfall typical for a temperate climate minus potential evaporation. The leachant used was rainwater, which was passed through plastic drums containing rock fragments, soil and gravel to simulate the effect of water passing through an engineered cap. Lysimeter A was operated for 21 months, B and C for 39 months. Prior to the start of irrigation, 30 litres of leachant was passed through lysimeter C to determine the effect of pre-washing the waste.

The input leachant and output leachate were routinely monitored and averaged results are presented in Table I where the range represents two standard deviations from the mean. Typical pH profiles with time exhibited a rapid fall to around 5, followed after 4-5 months by an erratic period of increase before settling in the region 6.3 - 6.4. All the lysimeters exhibited broadly similar redox potential trends: erratic variations but values generally below 250 mV. The major ions except nitrate and sulphate exhibited typical washout behaviour: after an initial peak, concentration levels were similar to the input leachant values. Leachate nitrate levels were below the limit of detection possibly due to a combination of nitrate utilisation by microbes for growth and denitrification bacteria. Leachate sulphate levels also showed a marked decline for all the lysimeters after a few months. This change may be due to Sulphate Reducing Bacteria. From the start of gas sampling in month 2, CO2levels were high (>20%) and 02 levels were low (<5%), and this pattern was maintained throughout the experiment. Significant methane generation commenced first in lysimeter C (after 5 months), then A (10 months) and finally B (22 months). H2levels significantly higher than background were also observed up to a peak of 0.18%. The more rapid onset of methanogenesis in lysimeter C has been attributed to the fact that the waste was prewashed prior to the start of routine irrigation. These observed pH, redox and gas composition patterns are consistent with other studies [1] and may be ascribed to the degradation of cellulose in the waste in a series of linked processes: hydrolysis, aerobic respiration, anaerobic fermentation, acetogenesis and methanogenesis.

Radionuclide release plots showed a bimodal pattern which is commonly observed for a range of lysimeter designs and radionuclides including the actmides [2]. There is an imtial peak followed by a long tail when radionuclide concentrations in the leachate are relatively constant over a long time period. The initial peak may be attributed to washout of soluble ionic species from easily accessible sources of contamination. Long term release will be controlled by a combination of physical and chemical factors including solubility; sorption; radionuclide distribution in the waste and its accessibility; water flow rates and flow paths. As source terms for mathematical modelling, the release data are expressed in the form of a dimensionless release coefficient (Rc) equal to the concentration of the radionuclide in the leachate divided by the initial concentration of the radionuclide in the waste. The Rc values presented in Table II are based on the long term release values after the initial peak has subsided. Release coefficients calculated from leachate samples taken from the disposal trenches at the Drigg site are also quoted for comparison.

For each radionuclide, the results show good agreement between the different lysimeters and the site data despite differences in size, waste composition and mode of operation. These prototype experiments did not yield any measurable quantities of actinides in the leachate due to the very low concentrations of these elements in typical Sellafield LLW. Subsequent lysimeter experiments have been conducted with artificially spiked waste samples and these results will be reported separately.

Prediction of pathways of radionuclides, heavy metals and other harmful components isolated in cementitious matrices is considered critical in geochemical studies related to hazardous waste disposal [1-3]. Hence, rigorous multi-phase multi-aggregate thermodynamic models with appropriate kinetic and metastability constraints can be very helpful in solving this difficult problem. In a companion contribution [4], we demonstrate that usage of the Gibbs energy minimization (GEM) algorithms permits direct calculation of solid solution – aqueous solution (SSAS) equilibria which adequately describe the solubility data. The suggested "core" thermodynamic dataset can be extended and tuned on the basis of solubility, mineralogical and petrographical studies of fresh, aged and doped cements.

LiNO3 addition to cement was examined to prevent hydrogen gas generation from metallic aluminum in dry active wastes during waste solidification and under circumstances of underground water penetration into the land disposal site. And its reaction mechanism was identified by some chemical analyses. The volume of the hydrogen gas generation with LiNO3 addition was 10% as much as that without LiNO3, by formation of the insoluble Li-Al preservation film on aluminum. After the land disposal of waste forms, LiNO3 would be expected to be effective to prevent aluminum corrosion as long as the cement forms keep their alkaline character, and when it became ineffective, the circumstances around the waste forms can be made less corrosive for aluminum. The aluminum corrosion at that time would be as much as 10% of that without LiNO3 addition.

Swiss legislation requires the safe final disposal of radioactive wastes. Final disposal is defined by the safety authority guidelines as follows: ‘After a repository has been sealed, no further measures shall be necessary to ensure safety. The repository must be designed in such a way that it can be sealed within a few years.’ The time for repository sealing is not specified.

In Switzerland, the Wellenberg site was selected for the LIILW repository after an extensive site evaluation process and, in June 1994, the application for the general licence was submitted to the Swiss Federal Government. In addition to the federal licence, a mining concession has to be granted by the canton in a public referendum. The decisive cantonal vote in June 95 was negative (52%) and led to a temporary set-back in the project.

Among the reasons for this negative result was the proposed repository emplacement concept. It was planned to successively backfill the voids between the emplaced disposal containers in the caverns layer by layer. Retrieval of the waste would therefore involve high expenditure and effort. The concept therefore implies closing part of the repository at a very early stage. The outcome of the referendum reflected an unwillingness on the part of the population either to accept this situation or to make a decision today on final disposal. In many discussions, a stepwise implementation of the repository was requested, not only from the point of view of technical feasibility but also to allow political and legal aspects to be taken into consideration. A final decision - particularly regarding closure of the repository - could then be taken in the future, after giving the coming generation an opportunity to think over our proposed solution.

To meet this requirement, GNW/Nagra studied modified repository emplacement concepts within the framework of Swiss legislation. In the proposed new project, concrete containers are emplaced in the caverns without any backfill, allowing easy retrieval until the decision is taken to close the facility. This decision may be taken only after completion of container emplacement, allowing for further monitoring of the filled repository caverns. A decision on how to proceed is currently being prepared by a federal working group with representatives from the canton, the safety authorities and GNW/Nagra.

Studies have shown that, with an appropriate design, drainage and ventilation system, the required stability of the modified repository cavern can be achieved also under seismic loading and that the requirements of operational safety can be met. The potential long opening times of the repository caverns show some minor effects on certain repository components, such as e.g. slight carbonatisation of the container backfill. Overall, it can nevertheless be demonstrated that the long-term safety is not adversely affected.

NAGRA (National Cooperative for the Disposal of Radioactive Waste) has been investigating the feasibility for siting a low- and intermediate-level radioactive waste (L/ILW) repository in subsurface geologic environments. The design of the L/ILW repository consists of a horizontally accessed cavern system located beneath Wellenberg in the low-permeability unit of Valanginian marl. One of the concerns for the safety analysis is the effect of gas generation (primarily hydrogen) from anaerobic corrosion and degradation of waste material. The development of a free gas phase and the concomitant pressure buildup can result in the displacement of contaminated pore waters from the repository. Moreover, excess buildup of pressures in the repository can impact the structural integrity of the engineered barrier and of the surrounding host rock. For the investigation of the gas-related phenomena in a proposed L/ILW repository, a process model was developed to simulate the different performance periods of the repository which include: (1) construction/operation period of the repository resulting in desaturation of the host rock caused by pressure decline and ventilation in the cavern, and (2) post-closure period after backfilling and sealing of the caverns resulting in resaturation of the host rock and of the cavern, and in the release of waste-generated gas. Period (1) takes into account degassing of dissolved gas (methane) in the formation water due to the pressure decline around the cavern, and period (2) considers dissolution of free gas caused by the pressure increase associated with the resaturation and with the gas production in the cavern.

Preparing for safe disposal of the LMFR primary circuit equipment and sodium reprocessing for storage and burial with minimum volume of radioactive wastes resulted in testing of radionuclide sorption trapping, distillation and rinsing with water under vacuum used for safe sodium removal and decontamination of equipment.

In Russia the spent sealed radiation sources are disposed of into shallow ground borehole repositories. Design and status of the repositories as well as the basis for the safe disposal of the sources with metal matrices are presented.