Experiments have been conducted to determine an upper limit to the dissolution rate of UO2 under reducing conditions at great depths in granitic bedrock. The aqueous phase was a dilute, synthetic groundwater that simulates fresh groundwater conditions, so-called modified Allard groundwater. Four dissolution tests of long duration were performed. Test duration ranged from 52 to 140 days. Reducing conditions were produced with metallic iron. Average dissolution rates for the UO2 samples under the test conditions were preliminarily calculated based on the total recovery of U at the end of the tests. The dissolution rates ranged from 0.12 to 0.14 ppm/yr for the sum of the total release from the third and fourth cycles of testing (269 days). There was no indication of an effect of alpha radiolysis on the dissolution rate results for UO2 samples with 0, 5, and 10% 233U.

Silver exchanged zeolites A, X and Y were used to occlude silver iodide (AgI) at 400°C. Heating to 900°C in a hot isostatic press caused decomposition of the zeolite materials and the formation of more dense phases. Silver zeolites A and X both formed sodalite, seen as a potential 129I waste form, while silver zeolite Y formed an x-ray amorphous phase containing AgI. Silver zeolite A produced the best potential waste form, a monolithic sodalite with negligible porosity.

The model tested in this paper couples glass dissolution rate and geochemistry in the surrounding environment, thus leading to the comprehensive description of the interactions between the glass and the minerals and their transformations. Leaching of glass is simulated at 90°C and S/V=80 cm-1 in pure water or with iron from the canister and overpack or with argillite. The glass dissolution is described with an affinity law with respect to a nontronite-like phase, which saturation state depends on Si, Al, Fe, Na and Ca activities. The simulations results allow to reproduce both the decrease of the initial alteration rate and the so-called residual regime while the composition of the alteration layers are explained in terms of precipitated phases. In presence of metallic iron, the corrosion does not affect much glass dissolution rate if silica sorption is neglected. On the contrary, in presence of argillite, the alteration is enhanced by the dissolution of primary clay minerals and the precipitation of feldspar, K-zeolite and clay minerals.

In this paper, the proposed Yucca Mountain Repository is viewed from a corrosion perspective. A major component of the long-term strategy for safe disposal of nuclear waste at the Yucca Mountain Repository is first to completely isolate the radionuclides in the waste packages for long times and to greatly retard the egress and transport of radionuclides from penetrated packages. Therefore, long-lived waste packages are important. The corrosion resistance of the waste package outer canister is reviewed, and a framework for the analysis of localized corrosion processes is presented. An overview is presented of the Materials Performance targeted thrust of the U.S. Department of Energy/Office of Civilian Radioactive Waste Management's Office of Science and Technology and International. The thrust program strives for increased scientific understanding, enhanced process models and advanced technologies for corrosion control.

Ethanol/bentonite slurry is an alternative to the conventional cement-based material that may cause chemical disturbances to the host rock and the buffer material. In this study we carried out an experiment in which the ethanol/bentonite slurry was injected into a planar synthetic fracture at a constant rate. The pressure build-up due to the increase in viscosity of the bentonite ethanol slurry was observed and, eventually, consolidation of the slurry led to termination of the injection. A mathematical model was also developed to simulate the advection, dispersion and matrix dispersion of ethanol that contributed to the increase in viscosity of the ethanol/bentonite slurry, together with a Stoke's flow of the slurry before consolidation. Furthermore, applicability of the ethanol/bentonite slurry as a grouting material in a range of geological settings is discussed based on the result of a case study using the same model.

The underground formation that is currently being considered in Belgium for the permanent disposal of high-level radioactive waste and spent fuel is a 30-million-year-old argillaceous sediment (Boom Clay layer). This layer is located in the northeast of Belgium and extending under the Mol-Dessel nuclear site at a depth between 180 and 280 meter.

Within the concept for geological disposal (multibarrier system), the metallic container is the primary engineered barrier. Its main goal is to contain the radioactive waste and to prevent the groundwater from coming into contact with the wasteform by acting as a tight barrier. The corrosion resistance of container materials is an important aspect in ensuring the tightness of the metallic container and therefore plays an important role in the safe disposal of HLW. The metallic container has to provide a high integrity, i.e. no through-the-wall corrosion should occur, at least for the duration of the thermal phase (500 years for vitrified HLW and 2000 years for spent fuel).

An extensive corrosion evaluation programme, sponsored by the national authorities and the European Commission, was started in Belgium in the mid 1980's. The main objective was to evaluate the long-term corrosion performance of a broad range of candidate container materials. In addition, the influence of several parameters, such as temperature, oxygen content, groundwater composition (chloride, sulphate and thiosulphate), γ-radiation, … were investigated. The experimental approach consisted of in situ experiments (performed in the underground research facility, HADES), electrochemical experiments, immersion experiments and large scale demonstration tests (OPHELIE, PRACLAY). Degradation modes considered included general corrosion, localised corrosion (pitting) and stress corrosion cracking.

This paper gives an overview of the more relevant experimental results, gathered over the past 25 years, of the Belgian programme in the field of container corrosion.

Mechanical properties of YSZ doped with Np oxide were studied to investigate the sufficiency to be a waste form for immobilisation of highly concentrated TRU. The study was conducted focusing on Vickers (HV) and Knoop (HK) hardness, Young's modulus (E) and fracture toughness (KIC). The results showed that YSZ is harder and more resistant to elastic deformation and crack development than such waste forms as borosilicate glass and synroc. The effects of porosity and Np content on HV, HK, E and KIC are also discussed.

In order to study americium incorporation into calcite, CaCO3, under conditions of crystal growth, two samples of single crystal Am-doped calcite were synthesized and studied by cathodoluminescence (CL) spectroscopy in comparison with undoped and Eu-doped artificial calcite. Americium contents in calcite crystals were (in kBq/g): 1) 6.9; 2) 1.9(E+4). The CL emission of undoped and Am-Eu-doped calcitesamples was characterized by three broad bands at 2.03; 2.47 and 2.96 eV. Weak CL lines related to typical transitions 5D07F1,2,4 of Eu3+ and Am3+ions were observed at 1.68; 1.99, 2.06 eV and 1.60; 1.98 eV, respectively. Degrading of calcite structure under irradiation has been studied using CL emission of high power electron beam.

The Swedish KBS-3 copper-cast iron canisters for geological disposal are in an advanced stage of its development. This reports deals with the cast iron insert that provides the mechanical strength of the canister. The ductility of manufactured inserts often fall below the initial requirements [1]. An issue then was what consequences low tensile properties and scatter may have on the long-term failure probability of the canisters. This paper describes an approach to assess the failure probabilities for given actual material data and how to derive acceptance criteria for manufactured canisters. A statistical test programme was implemented using three inserts to determine the tensile, compression and fracture properties. Tested specimens were investigated by micro-structural analysis to determine the microstructure and categorize and size defects. It was found that the material scatter and low ductility was caused by many defect types, but with slag defects as the most important ones. These data were then used to compute defect distributions and as direct input to FE-calculations of KBS-3 canisters. A large number of FE- analyses were performed at the maximum design load (44 MPa) covering distributions of material parameters and geometrical features of the canisters. The computed probabilities were very low for both failure modes even for material data with poor ductility.

The sorption of thorium and americium has been measured onto crushed samples of freshordinary Portland cement (OPC), degraded OPC (DOPC) and green tuff under a range of aqueous conditions as part of research into the disposal of TRU waste in Japan. Sorption onto OPC was measured from deionised water,3 mol dm−3 sodium nitrate solution and simulated seawater, and onto DOPC from demineralised water; all solutions were pre-equilibrated with the relevant cement. Sorption onto tuff was measured from pre-equilibrated deionised water, 3 mol dm−3 sodium nitrate solution, simulated seawater, OPC leachate and 1 mol dm−3 ammonium hydroxide solution. RD values were determined from the amount of thorium or americium remaining in solution after centrifugation, 0.45 μm filtration and 10,000 nominal molecular weight cut-off (NMWCO) filtration. Centrifugation gave lower RD values than filtration. MeanRD values for sorption onto OPC and DOPC were in the range 4 × 104 to ≥1 × 106 cm3g−1 for thorium, and 3 ×103 to ≥1 × 105 cm3g−1 for americium, after filtration through 10,000 NMWCO filters. Mean RD values for thorium sorption onto tuff increased from2 × 103 cm3g−1 in tuff-equilibrated deionised water, to ≥ 4 × 106 cm3g−1 from ammonium hydroxide solution (10,000 NMWCO-filtration). A similar trend was seen for10,000 NMWCO-iltered americium samples where mean RD values increased from 4 × 103 cm3g−1 in deionised water to 1 × 105 cm3g−1 in ammonium hydroxide solution.

Potentially corrosive brines can form during post-closure by deliquescence of salt minerals in dust deposited on the surface of waste packages at Yucca Mountain during operations and the pre-closure ventilation period. Although thermodynamic modeling and experimental studies of brine deliquescence indicates that brines are likely to form, they will be nitrate-rich and noncorrosive. Processes that modify the brines following deliquescence are beneficial with respect to inhibition of corrosion. For example, acid degassing (HCl, HNO3) could dry out brines, but kinetic limitations are likely to limit the effect to increasing their passivity by raising the pH and increasing the NO3/Cl ratio.

Predicted dust quantities and maximum brine volumes on the waste package surface are small, and physical isolation of salt minerals in the dust may inhibit formation of eutectic brines and decrease brine volumes. If brines do contact the WP surface, small droplet volumes and layer thicknesses do not support development of diffusive gradients necessary for formation on separate anodic-cathodic zones required for localized corrosion. Finally, should localized corrosion initiate, corrosion product buildup will stifle corrosion, by limiting oxygen access to the metal surface, by capillary retention of brine in corrosion product porosity, or by consumption of brine components (Cl−).

The leaching behavior of nuclear wastes embedded in bitumen has been extensively studied for a decade by many teams of researchers. The main conclusion is that both water uptake and salts release obey to a diffusive regime. The most preponderant parameters of the Bituminized Waste Products (BWPs) leaching are, on the one hand, the nature and the content of the salts that are embedded in the bitumen, and on the other hand, the microstructure of bitumen and the external constraint applied to this confining material.

The numerical code COLONBO, designed by the CEA, allows to predict the behavior of the BWPs leaching in the case of salts embedded in soft bitumen and in free swelling conditions (no mechanical constraints during the leaching). This paper intends to achieve two goals: (i) to propose an analytical resolution of the system of equations implemented in COLONBO by using accurate simplification. The validation of this approach relies on the comparison with the predicted solutions of COLONBO and, when available, with the experimental results. (ii) To extend this simplified modeling to the case of restricted swelling conditions. The point is that the validation of this approach would allow us to propose a unified system of equations, allowing to simulate the BWPs leaching in whatever conditions and improving consequently the field of validity of the predictions for the behavior of bituminized waste products.

The Eurobitum bituminised waste produced by the former EUROCHEMIC reprocessing plant in Mol-Dessel (Belgium) contains a lot of salts, mainly NaNO3. This NaNO3 may possibly affect significantly the evolution of a radwaste repository through redox and ion exchange reactions with Boom Clay, which in Belgium is studied as a reference host formation. Yet, the leaching behaviour of NaNO3 from the Eurobitum and its migration in the Boom Clay is rather unclear. Therefore, some scoping numerical calculations have been performed. In a first phase, we calculated the water flux towards the disposal gallery to assess the timescale of the saturation process of the disposal galleries. In a second phase, the leaching rate of NaNO3 out of the drums is compared with its diffusive removal through the Boom Clay.

The water flux to the gallery was calculated at approximately 100 ml/day per metre gallery, based on the natural pressure gradient observed around the HADES Underground Research Laboratory. Different approaches were elaborated to find a NaNO3 source term for the diffusion calculations. For various ways of characterising the nitrate release, the numerical calculations generally show that the NaNO3 concentrations in the first decimetres of Boom Clay will not be much higher than 1 M. Moreover, a sensitivity analysis showed that if the nitrate is released from the bituminised waste product within 10 000 years, the Boom Clay controls the concentration profiles; if the release is slower, the source term has a more pronounced effect on the near field concentrations.

In the geological disposal of radioactive wastes, cement used for the repository construction alters the condition of groundwater to highly alkaline of pH about 13. Such alkaline groundwater around the repository would alter the surfaces of the rocks in the flow paths to the amorphous phase. Once the surface is altered, it takes a geological period (so much long time) to restore the surface to its former condition. This study examined the sorption behavior of europium (Eu-152 (tracer), Eu-151 (5×10-5 M, carrier) onto some silica minerals with polymeric silicic acid. Polymeric silicic acid also affects the alteration of the solid surface.

The results showed that the kinds of silica minerals strongly affect the sorption behavior in the range of 5<pH<8. The main difference was due to the degree of crystallization of the solid phase, the specific surface area and the concentration of polymeric silicic acid. Since the amorphous silica or polymeric silicic acid takes a loose structure, the isoelectric point is high compared to the crystal ones such as opal-CT and cristobalite, which in turn decreases the sorption. The crystalline silica with polymeric silicic acid remarkably decreased the sorption of europium. This suggested that the surface of solid phase was altered by polymeric silicic acid.

In the frame of an integrated in situ test on the alteration of the SON68 reference glass in realistic disposal conditions, the leaching of Np, Pu, and Am from α-doped SON68 glass samples and their migration in three clay-based backfill materials at high density was measured. The addition of powdered glass frit to a Ca-bentonite decreased the glass alteration by two orders of magnitude, and resulted in a concomitant decrease of the radionuclide leaching in comparison with dried Boom Clay, which is the more agressive backfill. All actinides were selectively retained in the alteration layer, but the retention degree was lower for Np than for Pu and Am. All backfill materials proved to be efficient barriers against radionuclide migration. For nearly all radionuclides and backfill materials, the migration profile showed a contribution of colloidal transport.

Sorption processes have been largely studied and modelled by considering single mineral surfaces (iron oxides, clay etc). In the case of a complex surface such as claystone, it seems important to be able to predict the behaviour of the radioactive elements in contact with the natural site. The modelling part becomes very complex if the whole minerals composing the geological media have to be taken into account. In that way, this study aims at testing various mineral mixtures, to identify the most reactive phase controlling the adsorption processes, in order to get a “simplified” system for a model calculation. Various mineral assemblages (goethite, pyrite, calcite and montmorillonite) were studied here, in contact with an anionic specie (selenium). Experiments were carried out using single, binary, ternary or quaternary solid mixtures, in order to evaluate the major reactive mineral phase. A surface complexation model was then considered to fit the experimental data. Stoichiometries and constants were extracted using FITEQL by fitting experimental results obtained on single solid in NaNO3, and used in a predictive calculation with PHREEQC for mixtures in synthetic grounwater. By doing this, we wanted to verify if the assumption that surface reactivity of single minerals keeps the same in a different environment can be used for predictive calculations.

Results obtained have shown that the presence of calcite and pyrite does not influence the total amount of selenium adsorbed. Then a simplified model was proposed, taking into account only the major phase responsible for the sorption of Se.

United Kingdom Nirex Limited develops and advises on safe, environmentally sound and publicly acceptable options for the long-term management of radioactive waste. One option Nirex has developed is a phased geological repository concept for intermediate level waste and some low level wastes that makes use of a combination of engineered and natural barriers. Physical containment of radionuclides would be achieved by immobilisation and packaging of wastes (mostly) in stainless steel containers.

Existing models of the migration of dissolved radionuclides from packaged wastes suggest that radionuclide release is determined largely by the rate of diffusion through the encapsulation grout used to immobilise the waste. The use of such models requires diffusion coefficient data for radionuclides in waste encapsulation grouts. This paper describes a programme of through-diffusion experiments, and modelling interpretation, aimed at deriving diffusion coefficients for some radionuclides in two types of encapsulation grout.

An intrinsic diffusion coefficient of HTO of around 1×10−13 to 2×10−13 m2s−1 was determined for a 3:1 mix of blast furnace slag to ordinary Portland cement, compared to around 4×10−13 to 5×10−13 m2s−1 for a 3:1 mix of pulverised fuel ash to ordinary Portland cement. These values are lower than that assumed for a non-sorbing radionuclide in an earlier modelling exercise. Porosity values around 0.3 were obtained in each case. For 36Cl as chloride, the experiments showed no significant breakthrough over the experimental timescale of about one year, suggesting an intrinsic diffusion coefficient below 5×10−13 m2s−1. One possibility is that chlorine-containing solids are precipitating in the cement. An intrinsic diffusion coefficient for 137Cs in the 3:1 mix of pulverised fuel ash to ordinary Portland cement of 4×10−15 m2s−1 was estimated, significantly lower than that determined for HTO.

The results from three of the sixteen experiments could not be fitted with a simple diffusion model, and for a further five experiments there was some doubt as to whether simple diffusion behaviour had been observed. It is suggested that this may have been due to cracks in the grouts that were sufficiently large to affect the diffusion properties of the grouts, although none was visible to the naked eye. Cracking of the waste encapsulation grouts could provide a mechanism for enhanced migration of radionuclides from waste packages, compared with diffusion in a homogeneous porous medium alone.

This paper presents a study on gas migration behavior in a bentonite specimen with the aid of X-ray computer tomography (CT) scan data. The laboratory experiment was carried out to clarify gas migration behavior through saturated, compacted bentonite. X-ray CT was used to estimate the spatial distribution of gas and water saturation during gas migration test in the bentonite. For the gas migration test, the controlled flow rate of gas injection was adopted for pre-compacted samples of Kunigel V1 bentonite using helium gas, which is safer than hydrogen gas.A specimen was isotropically consolidated and saturated by synthetic seawater, simultaneously, by applying a backpressure. This was followed by injecting the gas using a syringe pump. Inlet and outlet gas fluxes were monitored. This test exhibited a significant threshold pressure for breakthrough, somewhat larger than the sum of the swelling pressure and the backpressure.

The procedure of the X-ray CT measurement is as follows; i) measurement of the initial condition (saturated condition) of the compacted bentonite, ii) measurement of the gas injection condition as a function of time. The digital data obtained from the X-ray CT usually includes some noise. The stacking method can reduce the noise in CT values and enables to identify the gas migration area. The results indicate that gas is transported through preferential pathways in compacted bentonite, and is not homogenous.

A model has been developed to predict the stress corrosion cracking behaviour (SCC) of used-fuel copper containers in a deep geologic repository. By comparison of the predicted evolution of the environment with criteria for the SCC of pure copper, it is concluded that copper containers in a deep geologic repository will not be susceptible to SCC. In particular, the requirement for the simultaneous presence of a thermodynamically stable Cu2O/CuO surface film and a significant concentration of SCC agents is not satisfied at any time during the 106 year simulation period.

Carbon steel has been chosen in the Czech disposal concept of spent fuel disposal in a granite host rock as a reference material for disposal canisters. On the basis of the results of performance assessment studies, it could be decided whether this material is suitable or whether a more corrosion resistant, and also more expensive, material should substitute it. A number of papers have convincingly shown that iron transfer constraints contribute to a significant decrease in corrosion rate, but no study hasso far been devoted to modeling this process. In this paper the effects of initial corrosion rate, corrosion product solubility and porosity and other repository parameters on the transfer of iron to the host rock are modeled using a numerical transport computer code. It was found that the critical parameter for iron transfer is the solubility of corrosion products, considerably affecting the steady state corrosion rate. The initial corrosion rate of carbon steel and the sorption properties of bentonite primarily affect the time needed to achieve a steady state of corrosion. The results of the calculations strongly suggest that the constraints on iron transfer from the canister surface will govern the corrosion rate of carbon steel canisters, whose lifetime, owing to this effect, can stretch to millions of years.