[en] The existence of a number of historical wastes has prompted the need to develop a disposal strategy for material contaminated with radium-226. This report reviews the pertinent radiological and chemical properties of radium. Chemical factors that determine the mobility of radium in soil/groundwater environments are discussed. The mineralogy of mine tailings and attempts to leach them are reviewed. Radium levels in leached residues are compared to the standards for radium levels, and realistic targets are suggested for leaching methods. Techniques for scrubbing soil, immobilizing radium and treating wastewater containing radium are reviewed. Recommendations are made for a possible leaching strategy for radium-contaminated soil, and for further research to develop an effective disposal strategy

[en] The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO₂ fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO₂ to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

[en] The basic mechanisms of film growth, transformation, and dissolution of phases formed on surfaces are discussed. Film growth can occur via solid-state processes or via substrate (usally metal or alloy) dissolution, followed by local supersaturation and precipitation of an insoluble phase. The phase(s) formed may be metastable and transform to a more stable phase, via either solid-state or dissolution-reprecipitation processes. Film dissolution reactions can also occur via a variety of mechanisms, including: (i) direct chemical dissolution when no oxidation state change occurs; (ii) redox dissolution when the film dissolves via a redox reaction involving a reducing or oxidizing agent in solution; and (iii) autoreduction, where film dissolution is coupled to metal dissolution. Such film-growth and dissolution processes, which often produce complex multilayer films, are common in the nuclear industry. A number of examples are discussed

[en] The use of titanium alloys in two different waste package designs has been reviewed under the, conditions anticipated in the Yucca Mountain nuclear waste repository. In the first design. they are considered as one of three barrier materials incorporated into the waste package design and potentially in galvanic contact with the other two waste package materials, 316L stainless steel and Alloy-22. In the second design the Ti alloy is considered as a drip shield placed over, and not in contact with, a dual wall waste package fabricated from the other two materials. The possible failure processes, crevice corrosion, pitting and hydrogen-induced cracking (HIC) have been reviewed for the candidate titanium alloys (Ti-12, Ti-16 and Ti-7). Both pitting and crevice corrosion are very remote possibilities under these conditions. For Ti-12, a limited amount of crevice corrosion is possible but repassivation will occur before substantial damage is sustained. When Ti is considered as part of the triple wall waste package, hydrogen absorption leading to HIC, within an acidified but passive crevice, is the most likely failure mechanism. When the Ti alloy is utilized in the form of a drip shield then hydrogen absorption under potentially alkaline conditions is the major fear. Both Ti-12 and Ti-16 have been shown capable of tolerating substantial amounts of hydrogen (∼400 μ g·g^-1 for Ti-12, and > 1000 μg.g^-1 for Ti-16) before any effect on the materials fracture toughness is observed. The rate of absorption to a hydrogen content which exceeds these values will be the key feature determining if, or when, the material becomes susceptible to cracking. Once this condition is achieved, whether or not failure occurs will depend on the strength and location of stresses within the structure. For Ti to absorb hydrogen it is inevitably necessary to subject the material to cathodic polarization, either by coupling to a more active material or by the application of galvanic protection. Absorption occurs when the passive TiO₂ is rendered permeable to hydrogen by cathodically inducing redox transformations in the oxide (Ti⁴+ → Ti³⁺). However, the presence of intermetallic particles (Ti₂Ni in Ti-12, Ti_xPd in Ti-7/Ti-16) could allow hydrogen absorption at lower cathodic polarizations, since these particles could act as hydrogen absorption 'windows' in the oxide. This has been shown to occur for Ti-12, but does not appear to occur for Ti-16. On the contrary, there is speculative evidence to suggest that the intermetallics present in Ti-16 inhibit hydrogen absorption. If failure of Ti by hydrogen absorption leading to HIC is to occur then the cathodic titanium must couple to an available anode. For Ti-12, the Ti₂Ni intermetallics are corrodible and their anodic corrosion could couple to hydrogen absorption by the alloy matrix. For Ti-16 the intermetallics are inch and thus do not appear to be a feasible pathway for hydrogen absorption. A second possibility is that in the presence of F^- enhanced passive dissolution could couple to hydrogen absorption. This is a more likely possibility under acidic conditions in a mixed metal crevice than it is for the more alkaline conditions anticipated on the drip shield. The evidence from dental and flue gas scrubber studies suggest this is unlikely, especially with the Pd-containing alloys. The evidence, however, is not totally conclusive. Other anions, SO₄^2- and HCO₃^- and the species, silica, expected to be present in copious amounts in concentrated groundwaters at Yucca Mountain, are very likely to counterbalance any aggressiveness of F^-. In alkaline solutions, F^- does not appear able to enhance the passive corrosion process. The final possibility that could lead to hydrogen absorption by Ti alloys is their coupling to other waste package materials; i.e. Alloy-22 and 316L stainless steel. Clearly, this is only achievable if Ti is incorporated into the triple wall waste package design. An active couple between a Ti alloy and Alloy-22 is very unlikely, since both are passive materials on which active corrosion is extremely difficult to activate. A couple involving 316L and a Ti alloy could be active if there was a loss of passivity on the steel. This remains a possibility until tests prove otherwise. Based on these studies the order of preference of these alloys as materials for waste packages would be Ti-16 ≥ Ti-7 >> Ti-12. However, before a confident recommendation that Ti-16 be chosen over Ti-7, a better understanding of the presently baffling behaviour of the intermetallic particles in this material is required. (author)

[en] The 23 papers presented at this conference review the technical merits, and particularly corrosion performance, of the three main materials used for nuclear fuel waste containers: titanium and its alloys, copper and its alloys, and iron and carbon steels. The specific questions posed to the Workshop were: 1) Can we predict the lifetime of container materials in a variety of vault environments? 2) Is there a limiting range of conditions beyond which a specific material cannot be used? 3) Do we have the necessary corrosion rate data and/or mechanistic models required to make predictions? 4) Can we justify the use of titanium on the basis of propagation rate measurements for crevice corrosion, or do we need to prove initiation cannot occur? 5) Will the pitting of copper be significant? 6) How thick a carbon steel container would be required, and can it be fabricated and stress-relieved? 7) Are radiation fields of any consequence at the dose rates expected?

[en] Titanium and its alloys (Grades-2, -12, -16) are candidate materials for Canadian nuclear waste containers on the basis of their apparent immunity to many localized corrosion processes. This simplifies markedly the effort needed to justify the use of these materials and to develop models to predict the lifetimes of containers. Here we review the pitting, microbially influenced corrosion (MIC), and corrosion under unsaturated conditions, of titanium. For all these processes, the properties of the passive oxide film are paramount in determining the metal's resistance to corrosion. A review of these oxide properties is included and the conditions to which the metal must be exposed if localized corrosion is to occur are defined. Since these conditions cannot be achieved under Canadian waste vault conditions, it can be concluded that pitting and MIC will not occur and that corrosion under unsaturated conditions is extremely unlikely. (author)

[en] Titanium is a reference material for the construction of waste containers in the Canadian Nuclear Fuel Waste Management Program. It has been in industrial service for over 30 a, often in severe corrosion environments, but it is still considered a relatively exotic material with limited operating history. This has arisen because of the aerospace applications of this material and the misconception that the high strength-to-weight ratio dominates the choice of this material. In fact, the advantage of titanium lies in its high reliability and excellent corrosion resistance. It has a proven record in seawater heat exchanger service and a demonstrated excellent reliability even in polluted water. For many reasons it is the technically correct choice of material for marine applications. In this report we review the industrial service history of titanium, particularly in hot saline environments, and demonstrate that it is a viable waste container material, based upon this industrial service history and operating experience. (author)

[en] This report reviews the chemistry of UO₂ dissolution under conditions relevant to the disposal of used nuclear fuel in a geological vault. It provides the chemical understanding necessary for selecting the most appropriate model for estimating UO₂ fuel dissolution rates in a nuclear waste disposal vault. The report briefly describes the solid-state structures of various uranium oxides; discusses the nature and mechanism of UO₂ oxidation and dissolution in groundwaters; summarizes the factors affecting UO₂ dissolution under oxidizing conditions; discusses the impact of various oxidants and water radiolysis on UO₂ oxidation and dissolution; briefly comments on the effects of vault chemistry and secondary phase formation on the dissolution process; discusses the physical properties of UO₂ that may influence the kinetics of dissolution; and describes our approach for developing a kinetic model of UO₂ dissolution under oxidizing conditions

[en] A model to predict the dissolution of UO₂ fuel under both oxidizing and non-oxidizing conditions is presented and compared with other available models for fuel dissolution. Dissolution rates under oxidizing conditions are predicted by extrapolating steady-state electrochemical currents for the anodic dissolution of UO₂ to the corrosion potentials measured in solutions containing various oxidants, including dissolved oxygen, hydrogen peroxide, and the products of the gamma or alpha radiolysis of water. For non-oxidizing conditions, the dissolution rate of UO₂ is not well known. Attempts to measure this rate are fraught with difficulties, and the published values are difficult to rationalize within the framework of our model. Consequently, we briefly reviewed the literature on the dissolution of similar p-type semiconducting oxides and chose to estimate the chemical dissolution rate of UO₂ by analogy to the well-studied oxide NiO. In this manner we have managed to establish a threshold rate below which the rate of oxidative dissolution becomes negligible in comparison with the rate of chemical dissolution. This threshold agrees quite well with that established electrochemically. Using these extrapolated rates we predict that the rate for oxidative dissolution of CANDU (CANada Deuterium Uranium) fuel due to gamma radiolysis will fall below this threshold after ∼ 200 a, a time period that is short in comparison with the anticipated lifetimes of titanium waste containers, which are expected to last for a period greater than ∼ 1200 a. For dissolution due to alpha radiolysis, oxidative rates are uncertain, but could be above this threshold for a period of 500 to 10 000 a for CANDU fuel, and 500 to 30 000 a for pressurized water reactor (PWR) fuel. The uncertainty in these ranges reflects the poor quality and limited number of corrosion potential measurements in the presence of alpha radiolysis

[en] A multibarrier system is being proposed for the safe disposal of Canada's nuclear fuel waste. The container in which the fuel wastes are emplaced is an essential barrier and one that is expected to remain intact for at least 500 a, and possibly 10⁵a. Titanium is one candidate material that has been evaluated as a potential container material. The possible localized corrosion processes which could occur on titanium have been extensively reviewed and only crevice corrosion and hydrogen induced cracking (HIC) merit consideration under Canadian waste vault conditions. HIC is dealt with in a separate report. In this report we discuss the ability to predict titanium-container lifetimes based on current crevice corrosion modeling capabilities. An analysis of available data on the initiation, propagation and repassivation stages of crevice corrosion shows a variety of ways that titanium-container lifetime can be extended beyond the 1000-a lifetime predicted in our previously established container failure model. To justify the use of any material for construction of a nuclear fuel waste container requires a computational model to predict the time to failure. The previously reported container failure function consisted of three components: a statistical evaluation of early container failure; a low temperature hydrogen induced cracking failure mode; and a temperature dependent crevice corrosion failure mode. This model predicted lifetimes >1000 a for 99.9% of the containers in the disposal vault. An independent assessment of corrosion damage based on the empirical relationship between maximum penetration depth due to crevice corrosion and oxygen consumption agreed with the predictions of the container failure function. The data requirement for establishing the crevice corrosion limits of Grades-2, -12 and -16 titanium and for improving the arguments for the long lifetime prediction are presented. An extensive database including critical oxygen concentration for repassivation and penetration depths at different temperatures, and a coupling with mass transport calculations are required to predict the duration of crevice corrosion and the maximum damage expected. Passive-film properties of Grades-12 and -16 are required to determine the protective nature of the film and justify the long-term predictions of crevice corrosion resistance. (Author)