As part of an investigation into the consequences of adding inorganic metal salts to composite cements, based on blastfurnace slag (BFS) and pulverised fuel ash (PFA), the effect on the hydration behaviour of adding AlCl3 or As2O3 to the mix water of a number of cementitious systems was investigated using isothermal conduction calorimetry (ICC) and x-ray diffraction (XRD). Four cement systems were investigated; ordinary Portland cement (OPC), 3:1 BFS: OPC, 9:1 BFS:OPC and 3:1 PFA:OPC. AlCl3·7H2O and As2O3 were added to the mix water at 0.1% and 1% concentrations. Results from ICC indicated that the addition of As2O3 had very little effect on the setting and early hydration of any of the cements. AlCl3, however, inhibited the setting of the composite cements at the 1% addition level but accelerated the OPC at 0.1%. The expected hydration products were observed, with the additional observation of Friedel's salt in the presence of aluminium chloride additions and vaterite in both the aluminium and arsenic-containing systems after 180 days hydration.

Reactions of zeolite Na-A with AgI, and the sodium, copper and lead forms of zeolites A, LTA, X and Y with NaI, have been examined as possible starting routes to the long term immobilisation of iodine-129. Heating the salts in air, at 500°C, with the sodium forms of the zeolites leads to the formation of occlusion products, where the iodide salt migrates into the zeolite pores. Detailed studies of the Na-A / 5AgI complex indicate it has a uniform distribution of Na, Si, Al, Ag and I, and is thermally stable to ca. 750°C, where there is a substantial weight loss as iodine is released. In situ powder X-ray diffraction studies have been used to monitor the occlusion reaction at 400°C, and show that the occlusion product decomposes to produce a single crystalline phase at 800°C prior to further decomposition at 850°C to a mixture of nepheline and elemental silver.

A screening procedure is proposed for the assessment of the effect of concrete admixtures on the sorption of radionuclides on cement. The screening procedure focuses on a broad and generic assessment and can thus be used as a tool for the assessment of concrete admixtures possibly used in the future. The experimental feasibility of the screening procedure and the relevance of the results obtained are tested using a selection of superplasticisers and set modifiers. Selected results obtained for these admixtures are presented in this contribution.

The effects of organic cleaning agents (totally three) and degradation products from solid technical adsorbents (two) on the sorption of Eu on cement and TiO2 were measured with variation of concentration (0.001–10 % of the water phase for the cleaning agents, alkaline degradation products corresponding to 0.03–8 mM DOC for the adsorbents) and reaction time (up to 420 d). The effects of the cleaning agents Industrikombi and Prefect Citron were minor. The degradation of the ion exchanger (metaacrylic polymer) had no significant effect on the Eu-sorption. The other adsorbent (filter aid, inert fibre; acrylonitrile polymer) was rapidly degrading at high pH, and the degradation products had a significant reducing effect on the sorption of Eu at added DOC-levels in the mM-range. This category of adsorbent materials should not be incorporated in a cement waste matrix or be stored in a cementitious environment together with radionuclides.

In this paper, a simplified model which allows the derivation of analytical relations aiming at describing the diffusion-dissolution-swelling processes taking place in bituminized waste exposed to water or salt solutions is presented. The analytical solutions render the comparison with experimental data and tendencies straightforward. The validity of the simplified approach is demonstrated by comparing analytical and numerical solutions with experimental data.

An experimental approach was developed to quantify the influence of α and γ radiolysis of water on the dissolution kinetics of the spent UO2 fuel matrix. Two types of experiments were carried out: leaching experiments in deaerated media with UO2 fuel pellets doped with alpha emitters (238Pu), and leaching experiments with UO2 fuel pellets (standard or doped with alpha emitters) submitted to γ irradiation. Leaching experiments on UO2 pellets doped with alpha-emitters in deionized water and in deaerated media showed a correlation between the alpha particle flux at the reaction interface and uranium release in solution. The higher the alpha particle flux at the reaction interface, the higher the uranium release rate (mean rates over 30 days at S/V = 3 m−1: 15-year batch, r = 2.5 mg·m−2d−1; 1500-year batch, r = 0.2 mg·m−2d−1; 40 000-year batch, r = 0.04 mg·m−2d−1). These results, obtained at low flux levels that are realistic on a disposal time scale, supplement existing data describing the effect of doping on UO2 alteration at higher flux levels. Leaching experiments with the 1500-year batch under gamma irradiation allowed us to quantify the impact of a gamma irradiation field on the UO2 release. At 650 Gy·h−1 the effects of alpha irradiation appear to be fully masked since the uranium release is 450 times higher than in the presence of the alpha irradiation field alone. The Chemsimul code was also used to model water radiolysis and uranium release into solution.

This study examines the effect of water radiolysis on the dissolution of uranium dioxide. A model is created to describe the system of uranium dioxide fragments in water, and the production and reactions of radiolysis products (using recent kinetic data). The system is evaluated under different conditions using MAKSIMA-CHEMIST. Conditions examined include presence of carbonate in the water and effects of hydrogen. The simulations are compared to experimental results on spent fuel dissolution. Surprisingly, the simulated U(VI)-release agrees within a factor of three with the experimentally found U(VI)-release. The inhibiting effect of hydrogen is clearly demonstrated by the simulations. From the results of the simulations we are also able to conclude that the main inhibiting effect of H2 is the reaction with OH• and not the reduction of U(VI) to U(IV).

Experimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2 surfaces in poising the redox state of solid/water systems. For this purpose, an experimental programme was developed consisting on dissolution experiments with PWR spent fuel fragments in an anoxic environment and by using different solution compositions.

The collected data so far, indicate that production and fate of radiolysis products follow the same trends independently on the solution composition used in the tests. Hydrogen and oxygen concentrations show an initial increase with time until reaching a constant concentration. The trend observed for hydrogen peroxide is a decrease at short contact times to reach again a constant concentration with time. These steady-states indicate an overall balance of the generated radiolytic species.

Modelling work indicates that uranium dissolution is controlled by the oxidation of the spent fuel matrix in 10mM bicarbonate solutions while in the tests carried out at lower or without carbonate concentrations uranium in the aqueous phase is governed by the precipitation of schoepite.

These results are determinant to highlight that reducing conditions are restored in the aqueous phase in relatively short periods of time and at short distances away from the dynamic redox spent fuel/water interface.

Effects of water radiolysis on the dissolution rate of UO2 were studied by irradiating UO2 colloids with an alpha particle beam of a cyclotron. The solution was kept under reducing conditions by applying a current of -10 or -50 nA and an ultrahigh purity argon bubbling. The solution pH was fixed at 6 during experiment. The particle flux was 1.1×1011 α.cm−2.s−1 entering in solution with an energy of 5 – 6 MeV. Despite initial reducing conditions water radiolysis increased the dissolution rate of uranium dioxide due to reaction with oxidizing radiolytic species such as H2O2, O2, OH•, HO2•, ClO3−. The monitoring of the redox potential showed that the solutions became rapidly oxidizing. The dissolution rate values were between 0.01 and 26 mg·m−2·d−1 depending on the ratio of the irradiated UO2 mass to the solution volume. This dependency is caused by hydrogen peroxide consumption at the UO2 surface. The rate of hydrogen peroxide consumption by reaction with colloids corresponds to an equivalent rate of UO2 oxidation/dissolution between 2 and 60 mg.m−2.d−1 indicating that only a small fraction of produced H2O2 is consumed by the dissolution reaction.

The aim of this work is to provide credibility to the radiolytic models developed for performance assessment studies. Nowadays, there is a tendency to consider them useful tools for predicting the behaviour of spent fuel matrix under repository conditions, although they still suffer of limitations due, among others, to the following reasons: limited availability of kinetic data, difficulty to handle heterogeneous systems, lack of model validation.

A kinetics-based model to predict the dissolution of UO2 α-doped pellets under initial anoxic conditions is presented and compared with experimental results previously obtained. The uranium and plutonium concentrations in solution are predicted by considering the presence of an α-radiation field and its influence due to radiolysis of water on the pellet surface oxidation and subsequent dissolution.

The initial parameters required by the model in order to reproduce the pellet alteration process are: system geometry, chemical composition of the leachant, physico-chemical characteristics of the leachate and oxidation conditions of the pellet surface (expressed in terms of U(VI)/U(IV) ratio). The last one is the key parameter in the model for simulating the initial quick dissolution process. The results obtained are compared with experimental data. The agreement between the predictions obtained and the experimental published data is good.

The influence on the matrix oxidation – dissolution process due to the α-radiation field as well as the release of Pu are reproducible by the model. The Pu concentration trends as a function of time are explained in connection with the matrix dissolution process and are controlled by the formation of the secondary phase “Pu(OH)4(s)”.

In 1978, the Swedish Corrosion Institute published a report describing an assessment of the long-term corrosion behaviour of copper canisters for nuclear waste disposal. They concluded that the canisters would remain intact for a period exceeding 100,000 years in a deep underground repository in the Fennoscandian bedrock. In the intervening 25 years a great deal has been learnt about the corrosion of copper in the expected repository environment. In 2001, SKB and Posiva reviewed this new information and re-assessed the predicted canister lifetime [1]. It was concluded that the original prediction made in 1978 of canister lifetimes exceeding 100,000 years remains valid.

Results from an atmospheric exposure of copper at the Äspö Hard Rock Laboratory (HRL) are reported. Corrosion test coupons were exposed under sheltered conditions at 450 meters depth. The exposure was performed in order to gain insight into the likely state of a copper canister surface at the time when water begins to fill a repository for nuclear waste. The conditions for the experiment were selected to be relevant for the conditions predicted for a repository during an initial dry phase. Visual observations show that copper exposed at ambient temperature remained bright and shiny whereas coupons exposed at elevated temperature had a dull greenish appearance.

Two sensors, an on-line corrosion sensor based on the wire-resistance probe design and a reference electrode which are applicable in repository conditions are introduced. Both sensors have been tested in simulated repository conditions in bentonite saturated with highly saline groundwater at 80 °C temperature and 14 MPa pressure up to 150 h exposure. In addition, it has been demonstrated that after an initial transient period of ca. 5–8 h, Cu corrosion in highly saline groundwater filled with bentonite becomes undetectably low probably because of the anoxic conditions reached in the free groundwater. At the same time, the measured corrosion potential of Cu is still in the active corrosion region. Furthermore, redox potentials in the wetted bentonite remain significantly higher than in the free groundwater. This may mean that the decomposition of oxygen contained in the pores of the bentonite is a process occurring in a time scale larger than that of the present experiments.

This paper presents recent work on evaluating localized corrosion and stress corrosion cracking, two corrosion processes that are important to the long-term performance of Alloy 22 (58Ni-22Cr-13Mo-3W-4Fe). This alloy is the material preferred by the U.S. Department of Energy (DOE) for the outer container of the waste package to be used in the proposed high-level radioactive waste repository at Yucca Mountain, Nevada. It was found that both welded and thermally aged materials are more susceptible to localized corrosion in chloride solutions at temperatures above 60 EC than the mill-annealed material. This observation suggests that welding and certain post-welding operations may decrease the estimated life of the waste packages. However, no stress corrosion crack growth was observed in concentrated chloride solutions and simulated, concentrated groundwater at 95 EC when precracked compact tension specimens were tested under both constant and cycling loading.

In Sweden, spent nuclear fuel will be encapsulated in sealed cylindrical canisters, consisting of a ferrous insert and a copper outer container, for disposal in a geologic repository. Ferrous support structures will also be used in the repository, which will be backfilled with bentonite clay. Once any residual oxygen has been consumed, any ferrous material exposed to anoxic groundwater will undergo anaerobic corrosion, liberating hydrogen, forming a magnetite film, and releasing iron ions into the surrounding matrix. In order to characterise these processes the rate of hydrogen generation of steel in bentonite was measured using a barometric gas cell technique. The initial corrosion rates were found to be higher than measured previously in comparable aqueous solutions, but the long-term corrosion rates were similar. Analysis of the bentonite matrix showed that iron produced by corrosion had penetrated into the bentonite matrix, suggesting that ferrous ion exchange had occurred.

The deep geologic repository (DGR) concept is one of the approaches being considered in Canada for long-term management of used CANDU fuel. The Canadian concept entails encapsulating used CANDU fuel bundles in durable used-fuel containers (UFCs) and burying the UFCs in an engineered repository at depth in the Canadian Shield. Currently, an updated concept for a DGR is being prepared by Ontario Power Generation (OPG).

This technical paper describes OPG's studies from 1999 to 2002 that have led to the development of an updated reference UFC design. The updated UFC design holds 324 used CANDU fuel bundles. It consists of an outer corrosion-resistant oxygen-free phosphorous-doped copper vessel and an inner load-bearing steel vessel. The used-fuel bundles are placed within a fuel basket for handling and loading into the UFC. This reference UFC design would be used for future safety assessment and engineering studies.

The SKB project of Sweden has considered copper as the most appropriate metal to be used as container for high activity radioactive waste. However, it is still necessary to carry out some studies that can assure that their chemical, physical, mechanical properties, etc. do not loose stability in a period so long as 100.000 years.

In this work we show using anodic polarization, the corrosion potential (Ecorr) behaviour in different pressurized environment. The copper surface was characterized using optical microscopy. The electrochemical cell was mounted in a high pressure chamber that allowed to work up to 40 atm. The electrolyte solution simulates the deep groundwater, being the composition reported in literature [1].

The experimental results in solutions without bentonite show only slightly changes of the corrosion potential to cathodic values. At pressures of 40 atm, products of corrosion are observed, covering micropitting, that can be induced because of a bigger interaction metal/ ion at these pressures.

In the other case, bentonite presence produces beneficial effects in the resistance of the copper corrosion, since it is observed that the corrosion potential is displaced to more anodics values. This effects could be explained due to that this clay is able to retain ions that are aggressives for copper, such as chloride, sulphide, etc., and it liberates others that don't produce deterioration, such as sodium, that is in concordance with others authors[15].

The corrosion of copper in 1 M NaCl has been investigated at room temperature and at 80°C by an on-line resistance probe, solution analysis of dissolved copper and weight loss measurements. At room temperature, corrosion of the copper as indicated by an increase of the probe resistance has been detected during the first 60–80 h of exposure. After the corrosion potential has reached the immunity region of Cu, a decrease in the resistance of the sensor is observed, probably due to redeposition of Cu from the solution. It can be concluded that the corrosion of copper at room temperature virtually stops after 60–80 h due to the anoxic conditions established in the experiments. The similarity in the trends of the corrosion and redox potential during exposure seems to indicate that Cu(II) contained in the native oxide on Cu and/or formed by homogeneous oxidation of Cu(I) during the initial oxic period of exposure acts as a redox-agent determining the chemical conditions in the closed system.

At the end of 2001, measurements indicated that the resistance between the conductors had fallen in SKB's (Svensk Kärnbränslehantering AB) electrically heated test canisters. Electrochemical impedance spectroscopy (EIS) was employed together with chemical analysis of gas compositions and corrosion products to investigate the causes of the resistance decrease inside the canister. The measured system is complicated and the evaluation not straightforward. However, it seems that the plug connecting the external electrical power supply to the heaters in the canister is a critical component and that a moisture film around the electrical contacts of the plug may explain the resistance decrease. It is suggested to continue to monitor the EIS spectra with time to gain more information about the cause of the resistance drop.

Cementitious materials could be used in an underground repository as seals, bulkheads, grouts, or to provide a smooth surface for vehicles in the rooms and tunnels during repository construction. There is a possibility that an alkaline plume could be released from the cementitious materials in the repository and impact the corrosion behaviour of the canisters. The available literature information on the corrosion and electrochemical behaviour of copper in alkaline environments has been reviewed. Alkaline conditions promote passivation of the canister, whereas the presence of chloride groundwaters supports general dissolution. Localized corrosion due to an alkaline plume is considered to be unlikely because the protectiveness of the passive film (as measured by the difference between the corrosion and pitting potentials) increases with increasing pH.