Understanding the corrosion of spent nuclear fuel (SNF) and the subsequent mobilization of released radionuclides, particularly under oxidizing conditions, is one of the key issues in evaluating the long-term performance of a nuclear waste repository. However, the large amounts of iron in the metal waste package may create locally reducing conditions that would lower corrosion rates for the SNF, as well as reduce the solubility of some key radionuclides, e.g., Tc and Np. In order to investigate the interactions among SNF-waste package-fluids, four smallscale (∼1:40 by length) waste package mockups were constructed using metals similar to those proposed for use in waste packages at the proposed repository at Yucca Mountain. Each mockup experiment differed with respect to water input, exposure to the atmosphere, and temperature. Simulated Yucca Mountain process water (YMPW) was injected into three of the mockups at a rate of 200 μL per day for five days a week using a calibrated needle syringe. The YMPW was prepared by equilibrating 50 mg/L silica as sodium metasilicate with air, and adding enough HCl to lower the pH to 7.6 in contact with an excess of powdered calcite.

X-ray powder diffraction and scanning electron microscopy confirm that, where corrosion occurred, the dominant corrosion product in all cases was magnetite. In the high temperature (60°C) experiment, hematite and a fibrous, Fe-O-Cl phase were also identified. The Fe(II)/Fe(III) ratios measured in the corrosion products using a wet chemistry technique indicate extremely low oxygen fugacities (10-36 bar). Experiments are in progress in which 0.1g powdered UO2 was included in the mock-up in order to investigate the relative kinetics of Fe and U oxidation and to identify the U corrosion products formed under these conditions.

A comprehensive matrix of 60 tests was designed to explore the effect of calcium chloride vs. sodium chloride and the ratio R of nitrate concentration over chloride concentration on the repassivation potential of Alloy 22. Tests were conducted using the cyclic potentiodynamic polarization (CPP) technique at 75°C and at 90°C. Results show that at a ratio R of 0.18 and higher nitrate was able to inhibit the crevice corrosion in Alloy 22 induced by chloride. Current results fail to show in a consistent way a different effect on the repassivation potential of Alloy 22 for calcium chloride solutions than for sodium chloride solutions

An aging assessment of the OPG waste resin storage system predicted the potential for premature failure of the carbon steel resin liners. Consequently, resin liners made of 316L stainless steel with a minimum content of 2.5% molybdenum were selected to replace the carbon steel liners. The 2.5% Mo 316L stainless steel was specified to enhance pitting resistance in the spent resin environment. With the additional Mo, one would expect that a brief electrochemical corrosion test will reveal the superiority of such alloy over conventional 316L steel. This study reports a contrary experience

The Grimsel Test Site owned and operated by Nagra is located in the Swiss Alps (www.grimsel.com). The Sixth Phase of investigations was started in 2003 with a ten-year planning horizon. With the investigations and projects of Phase VI the focus has shifted more towards projects assessing perturbation effects of repository implementation and projects evaluating and demonstrating engineering and operational aspects of the repository system. More than 17 international partners participate in the various projects, which form the basic organisational “elements” of Phase VI. Scientific and engineering interaction among the different projects is ensured via an annual meeting and several experimental team meetings throughout the year. On-going projects include: evaluation of full-scale engineered systems under simulated heat production and long-term natural saturation (NF-Pro/FEBEX), gas migration through engineered barrier systems (GMT, finished this year), emplacement of a shotcrete low-pH plug (ESDRED/Module IV), testing and evaluation of standard monitoring techniques (TEM).

Numerous in-situ experiments with inactive tracers and radionuclides were successfully carried out over the past few years at the Grimsel Test Site (GTS). For the GTS Phase VI, three major projects have been initiated to simulate the long-term behaviour of contamination plumes in the repository near-field and the surrounding host rock:

•The CFM (Colloid Formation and Migration) project, which focuses on colloid generation and migration from a bentonite source doped with radionuclides

•The LCS (Long-Term Cement Studies) project, which aims at improving the understanding of low-pH cement interaction effects in water conducting features

•The LTD (Long-Term Diffusion) project, which aims at in-situ verification of long-term diffusion concepts for radionuclides

As Phase VI approaches its mid-term point, what are the next steps planned? The accomplishments assessed to date and the opportunities with the on-going projects as well as new projects – currently under discussion – are presented herein

The molecular signature of dissolved organic matter (DOM) in groundwaters can be used as a tool when investigating the palaeohydrological response of groundwater systems in relation to changes in recharge environment, and also for examining groundwater compartmentalisation, mixing and transport at underground repositories for radioactive waste. The DOM in groundwaters from two compartmentalised bodies of groundwater of distinctly different origin within the Äspö Underground Research Laboratory (URL), Sweden and in Baltic seawater has been isolated using tangential flow ultrafiltration (TUF) and diafiltration. Recoveries of DOM ranged from 34.7 to 0.1 mg/L with substantial differences in the concentrations of the groundwaters collected only 120 m apart. Analysis by infrared spectroscopy (IR) and pyrolysisgas chromatography-mass spectrometry (Py-GC-MS) of the isolated DOM revealed that the groundwaters contained abundant alkylphenols which may represent heavily decomposed proteins or lignins originating from biopolymers contained within soils. The difference in the distribution and relative abundance of major pyrolysis products groups such as alkyphenols confirmed that the groundwater and Baltic seawater DOM samples were chemically distinct indicating minimal infiltration of marine groundwater derived by recharge from the Baltic or earlier Littorina Sea within the two compartmentalised groundwater bodies.

An integrated numerical model for groundwater flow and radionuclide migration analyses in a water-saturated HLW repository with a multiple-canister configuration is developed by incorporating the heterogeneity of fractured host rock based on the previous multiple-canister model (MCFT2D [1, 2]). The current model incorporates i) heterogeneity of the fractured host rock represented stochastically by discrete fractures, ii) disposal-pit vertical emplacement concept, iii) representation of the waste package consisting of a waste canister and a bentonite-filled buffer, and iv) a user-determined repository configuration of multiple canisters using the repository parameters such as disposal tunnel spacing, waste package pitch, tunnel diameter, the number of tunnels in a repository, and the number of canisters in a tunnel. The current model can facilitate investigations into the effects of heterogeneous fractured host rock on water flow and nuclide migration for the different configurations of multiple canisters, as well as optimization of the repository design parameters in terms of release of nuclides from the repository.

Reducing water ingress into the Shaft at Dounreay is essential for the success of future intermediate level waste (ILW) recovery using the dry retrieval method. The reduction is being realised by forming an engineered barrier of ultrafine cementitious grout injected into the fractured rock surrounding the Shaft. Grout penetration of 6m in <50μm fractures is being reliably achieved, with a pattern of repeated injections ultimately reducing rock mass permeability by up to three orders of magnitude.

An extensive field trials period, involving over 200 grout mix designs and the construction of a full scale demonstration barrier, has yielded several new field techniques that improve the quality and reliability of cementitious grout injection for engineered barriers.

In particular, a new method has been developed for tracking in real-time the spread of ultrafine cementitious grout through fractured rock and relating the injection characteristics to barrier design. Fieldwork by the multi-disciplinary international team included developing the injection and real-time monitoring techniques, pre- and post injection hydro-geological testing to quantify the magnitude and extent of changes in rock mass permeability, and correlation of grout spread with injection parameters to inform the main works grouting programme.

Diffusion coefficients of iodine (I), cesium (Cs) and strontium (Sr) in solid cement blocks of ordinary portland cement (OPC) and low-heat portland cement containing 30 wt% fly ash (FAC) are measured by in-diffusion experiment. By using “laser ablation inductively coupled plasma mass spectrometry”, the penetration profiles of trace elements by diffusion were obtained quantitatively in cement solids. The apparent diffusion coefficients of the near-surface of the cement samples were estimated to be about 10-13 m2 s-1for I, Cs and Sr. Lower diffusion coefficients were observed for FAC than those obtained for OPC.

The ubiquity of iron oxide minerals and their ability to retain metals on their surface can represent an important retardation factor to the mobility of radionuclides. In a deep repository for the spent nuclear fuel, the intrusion of the groundwater might produce the anoxic corrosion of the iron, with magnetite as one of the end-products. In this study, as expected considering the strontium speciation in solution, strontium is sorbed onto magnetite at alkaline pH values while at acidic pH the sorption is negligible. Magnetite is able to sorb more than the 50% of the strontium from a 8·10-6 mol·dm-3 solution at the pH range representative of most groundwater (7-9). A surface complexation model has been applied to the experimental data, allowing to explain the results using the Diffuse Layer Model (DLM) and considering the formation of the inner-sphere complex >FeOHSr2+ (with a calculated logK=2.7±0.3). Considering these data, the magnetite capacity to retain strontium and other radionuclides is discussed

The influence of hypochlorite, chlorite and chlorate in the UO2 dissolution rate has been studied experimentally using a continuous flow-through reactor. Uranium concentration in each outflow solution was measured as a function of time and dissolution rates were determined once the steady-state was reached. The results obtained show that the influence of the hypochlorite anion concentration on the UO2 dissolution rate can be expressed by the following empirical equation

rdiss = 10-8.7±0.1•[ClO-]0.28±0.04

The dissolution rates obtained in this work were higher than those previously determined in presence of either oxygen or hydrogen peroxide using the same experimental methodology.

In contrast, neither chlorate nor chlorite had any significant effect on the UO2 dissolution rates under the experimental conditions of this work.

The long-term prediction of the equilibrium concentration of α-isosaccharinic acid (α-ISA) in cement pore water is a crucial step in the assessment of the role of cellulose in the safety of a cementitious repository. The aim of the present contribution is to summarise recent efforts in identifying the most important processes leading to the formation or degradation of α-ISA and in predicting its most likely concentrations in cement pore water. The issues considered are the kinetics involved in the formation of α-ISA, reactions of α-ISA with dissolved or solid compounds that may lead to limitations of its pore water concentrations and the chemical stability of α-ISA in a heterogeneous alkaline environment. Some new results are presented showing that α-ISA is degraded to low-molecular weight organic compounds in the presence of oxygen, whereas such processes occur only to a minor extent under anaerobic conditions. It is concluded that the processes involved in the degradation of cellulose under alkaline conditions are not sufficiently understood to explain fully the observed concentrations of α-ISA in long-term experiments.

Globally, the nuclear industry has a large number of legacy wastes that are stored in ponds, silos and tanks that are nearing the end of their design lifetime and hence said wastes need processing. In the UK there are significant quantities of radioactive sludge that have arisen from the corrosion of early Magnox fuel cans which have been stored underwater. As part of the present aggressive clean-up programme these materials will be retrieved, separated, processed and immobilised as dry waste forms for long-term storage. It is envisaged that hydraulic retrieval will be used for these ILW sludges resulting in some activity being released from the sludge phase to the process liquors challenging downstream ion exchange effluent treatment plants.

In order to understand this challenge, experiments have been conducted on sludge in ILW storage ponds and during sludge transfer operations to study the activity released from said sludges. In particular the solubility, adsorption behaviour of Sr-90 is discussed and how this and other aspects of the sludge chemistry impact upon the ion exchange effluent treatment process. The novel methodologies employed to obtain this data is also discussed.

Radionuclide retardation in mica gneiss, unaltered, moderately and strongly altered tonalite was studied by a thin section, batch, in-diffusion and column methods. Objectives were to examine retention processes in different scales and understand the influence of the rock matrix heterogeneity. Attempts were made for a more detailed interpretation of experiments using migration models used in performance assessments adapted for interpreting the laboratory scale experiments. Batch experiments were explained adequately using matrix diffusion-sorption model, instantaneous kinetic sorption model or model in which both mechanisms were taken into account. A numerical code FTRANS was able to interpret in-diffusion of calcium into the saturated porous matrix. Elution curves of calcium for the moderately and strongly altered tonalite fracture columns were explained adequately using FTRANS code and parameters obtained from in-diffusion calculations. Kd-values for intact rock obtained from fracture column experiments were lower than Kd-values for crushed rock indicating that batch experiments overestimate the retardation of radionuclides. Higher sorption and fair dependence on fraction size was obtained for altered tonalites due to the composition of alteration minerals and large specific surface areas.

In Belgium, the Boom Clay formation is considered to be the reference formation for HLW disposal R&D. Assessments to date have shown that the host clay layer is a very efficient barrier for the containment of the disposed radionuclides. Due to absence of significant water movement), diffusion - the dominant transport mechanism, combined with generally high retardation of radionuclides, leads to extremely slow radionuclide migration. However, trivalent lanthanides and actinides form easily complexes with the fulvic and humic acids which occur in Boom Clay and in its interstitial water. Colloidal transport may possibly result in enhanced radionuclide mobility, therefore the mechanisms of colloidal transport must be better understood. Numerical modeling of colloidal facilitated radionuclide transport is regarded an important means for evaluating its importance for long-term safety.

The paper presents results from modeling experimental data obtained in the framework of the EC TRANCOM-II project, and addresses the migration behavior of relevant radionuclides in a reducing clay environment, with special emphasis on the role of the Natural Organic Matter (NOM) [1]. Percolation type experiments, using stable 14C-labelled NOM, have been interpreted by means of the numerical code HYDRUS-1D [2]. Tracer solution collected at regular intervals was used for inverse modeling with the HYDRUS-1D numerical code to identify the most likely migration processes and the associated parameters. Typical colloid transport submodels tested included kinetically controlled attachment/detachment and kinetically controlled straining and liberation.

Interpretation of trace metal mobility in geological environments is often hampered by conflicting data from alternative experimental protocols and the lack of detailed mineralogical characterization of the host medium. To illustrate this issue, the release of uranium, thorium and the rare earth elements (REE) was investigated in polished rock slab samples from the U-Th deposit at the Palmottu Natural Analogue study site (SW Finland) by means of leaching experiments. The samples were sequentially leached with artificial groundwater of moderately high carbonate content at pH8, and nitric acid solutions at pH5 and pH3. The mineralogy and composition of the U, Th and REE mineral phases was studied using SEM-WDS and EDAX methods before and after each leaching step. In parallel, leaching was carried out on crushed material of the same samples and the leachates analysed by ICP-MS.

The most notable U minerals are uraninite, uranophane and two secondary U-Pb phases. Thorium occurs predominantly in monazite and at lower concentrations in uraninite. Accessory thorite is also present, which together with monazite contains most of the REE. Differential leaching of the elements was noted across all phases on the timescale of the experiments. Uraninite is partly dissolved at pH3. The main secondary uranium phase, uranophane, was stable in moderately acidic solution, but easily dissolved in the artificial groundwater and at pH3. Some release of REE was observed although the main REE-bearing phase, monazite, showed no evidence of degradation.

This study provides insights in the preferential release of radionuclides in granitic bedrock. An understanding of these processes is essential when assessing the safety of a spent fuel repository. Once released from the primary waste form U is expected to precipitate as secondary phases within micro fractures, as observed at Palmottu and numerous other deposits.

This paper presents results from deterministic calculations of radionuclide migration in a deep geological environment. The concept assumes single canister with spent nuclear fuel situated in bentonite (near-field) and surrounded by crystalline host rock (far-field). The results are presented in the form of release rates from the near-field and far-field, which contains also a part of advective release from the system of two single fractures. Additionally, radiological risk is evaluated in the form of dose rates for a water drinking scenario for an exposed population group. The analyses have been done by using the methodologies and computer codes used at the Institute for Energy EC-JRC in the Netherlands and the Mining Institute KSC RAS in Russia.

A previous radiological assessment of the UK Low Level Waste Repository (LLWR) has considered how the prevailing reducing chemical conditions in disposal trenches, may limit uranium release through the extreme low solubility of U(IV) solids. This study considers the additional effects that the physical and chemical nature of the uranium wastes may have on the release of uranium. Fluoride process residues produced by refining of uranium metal comprise the majority of the legacy inventory. Based on historic records and descriptions of the uranium wastes a conceptual model has been developed which bounds the release rate of uranium present as inclusions and dissolved in the solid residues by the dissolution rate of a magnesium fluoride matrix. The model is represented in a 3-dimensional groundwater flow and geochemical model. Initial findings indicate that the model correctly represents the range of fluoride and uranium concentrations that are measured in leachate from the LLWR trenches. Incorporation of this model in future safety assessments, together with a reduction in the derived inventory of uranium, is likely to result in a significant lowering of the peak groundwater dose to acceptable levels, even in the case that the site re-oxidizes. The study builds confidence in the inherent safety features that are provided by the sparingly soluble uranium waste residues and the reducing chemical conditions of the LLWR trenches.

Spent nuclear fuel from the Swedish energy programme will be stored in an underground repository situated in saturated fractured rock at a depth of approximately 500 m. This paper describes numerical simulations of radionuclide migration in the near-field (consisting of a canister filled with spent fuel and an engineered system backfilled with swelling clays) for the recently completed safety assessment SR-Can [1] using a Matlab / Simulink code. Handling of input data for the models from the site descriptive programme from on-going investigations at two candidate sites and the numerical modelling concept are discussed.

Complementary indicators have been used in developing a safety case in order to avoid uncertainties in the biosphere modeling used to estimate conventional dose or risk. For example, radionuclide fluxes can be used to evaluate the effectiveness of barrier performance. However, it is difficult to define relevant yardsticks for comparison, because the fluxes of naturally occurring radionuclides due to geological processes vary considerably depending on time and location. This paper discusses the relevance of alternative yardsticks for assessing modeled radionuclide fluxes by selecting yardsticks calculated from fluxes of natural radionuclides at the groundwater discharge point from the geosphere to an aquifer; these are then compared with fluxes of repository-derived radionuclides at the same point. Such yardsticks avoid surface geological processes that may also contribute to natural fluxes, allowing comparison at a suitable, common evaluation point that avoids dependence on site-specific conditions. The effectiveness and robustness of barrier performance is demonstrated using the developed yardsticks and the sensitivity of the analysis to groundwater flux is illustrated.

Unsaturated vegetated soils are an important component in performance assessment models used to quantify risks associated with deep engineered repositories for underground radioactive waste disposal. Therefore, experimental studies, funded by Nirex over nearly 20 years, have been undertaken at Imperial College to study the transfer of radionuclides (Cl-36, I-129, Tc-99) from contaminated groundwater into crops. In parallel to this has been a modelling programme to aid interpretation of the experimental data, obtain parameter values characterising transport in soil and plant uptake and provide new representations of near-surface processes for performance assessment. A particular challenge in achieving these objectives is that the scale of the experimental work (typically ≤ 1m) is much smaller than that required in performance assessment. In this paper, a new methodology is developed for upscaling model results obtained at the experimental scale for use in catchment scale models. The method is based on characterising soil heterogeneity using soil texture. This has the advantage of allowing hydrological and radionuclide transport parameters to be correlated in a consistent manner. An initial investigation into the calculation of effective (i.e. upscaled) hydrological and transport parameters has been undertaken and shows the results to be potentially highly (and non-linearly) sensitive to soil properties. Consequently, they have important implications for future site characterisation programmes supporting a proposed underground waste repository.