Most national high-level waste (HLW) disposal programs actually reflect, or are based on, concepts which were developed during the '70s or early '80s. Although suitable for demonstration of concept feasibility, designs of the engineered barrier system (EBS) do not take into account the tremendous developments in system understanding and materials technology over the last two decades, the practicality (and cost) of their quality assurance and implementation on an industrial scale and the transparency of the demonstration of the safety case. In many ways, due to the increased significance of popular acceptance over the last decade, the last point may be of particular relevance.

This paper reviews the work already carried out on “2nd generation” concepts and extends this to identify the key attributes of an ideal design for the specific case of disposal of vitrified HLW from reprocessing in a “wet” host rock (either crystalline or sedimentary). Based on the concept developed, key R&D requirements are identified.

The $108 million Maralinga Rehabilitation Project to clean-up the former British atomic test site in South Australia, the largest remediation of this type, was completed in 2000. The rehabilitation was agreed by the Commonwealth and South Australian Governments and the Maralinga Tjarujta traditional owners, following methods identified by technical experts. The rehabilitation permits unrestricted access to about 90% of the former restricted area, but excludes full time occupation of the remaining 120 square kilometers.

The clean-up has largely been directed at remediating plutonium contamination caused by minor trials. The clean-up involved two components, the removal of surface soil from the more contaminated areas, and the treatment of contaminated debris pits. Over 350,000 cubic meters of soil and debris was removed over about 2 square kilometers of land, and was buried in 10-15 meter deep trenches under a capping of clean soil of at least 5 meters. Treatment of contaminated pit debris at various sites has involved in situ vitrification, a process which involves application of an electric current to heat and melt debris encasing the contaminated material in a vitric/ceramic block, and exhumation of the contaminated debris and at depths of at least 5 meters.

The independent regulator to the project, the Australian Radiation Protection and Nuclear Safety Agency, has indicated that the clean-up of the test sites has met the radiological criteria laid down in the original plan. Discussions are progressing with the South Australian Government and the Maralinga Tjarujta traditional owners as to the long-term management of the site, and the possible hand back of the land from the Commonwealth to South Australia for addition to the Maralinga Tjarutja freehold lands.

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the long- term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.

The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.

The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a program for site investigations, for formulating functional requirements on the repository's barriers, and for prioritization of research.

The SR-97 disposal canister consists of an inner load bearing cast iron component and an outer corrosion barrier of copper. The mechanical strength of the insert must be such that it withstands, with sufficient safety margins, the combined forces of hydrostatic pressure at the disposal depth (7 MPa) and the bentonite swelling pressure (7 MPa). It must also withstand an increased hydrostatic pressure during a future glaciation. This pressure increase may be as high as 30 MPa. For this load situation, no extra safety margins will be required. Uneven pressure may occur during the water saturation of the buffer and to some extent some uneven pressure may also persist after water saturation. Calculations have been performed for such design basis uneven loads and the present canister design has been shown to withstand them. Under the oxygen-free conditions at the disposal depth, copper will be thermodynamically stable. Only dissolved sulfide will be able to corrode the copper. The supply of sulfide is expected to be low and, consequently, the service life of the canister will be very long. Calculations indicate that no known corrosion processes will lead to canister failure during a shorter time than at least 1 million years.

This paper describes the modelling of the hydromechanical evolution of a nuclear waste canister with an initial defect. The waste canister in the Swedish concept consists of a corrosion- resistant copper shell that covers a cast iron insert that provides mechanical strength.

If there is an initial penetrating defect in the copper shell, water will be able to contact the insert. Ingress of water is a prerequisite for corrosion of the iron insert, which in turn gives rise to production of hydrogen gas. Water is also a prerequisite for corrosion of the fuel assembly's metal parts, fuel dissolution and radionuclide transport. Due to the fact that the corrosion processes both consume water and generate hydrogen, strong couplings exist between the chemical and the hydraulic evolution, which must therefore be described in a coupled context. Certain aspects of the mechanical evolution must also be dealt with in parallel with the hydrochemical processes.

A discussion of the evaluation of the source term in the SR 97 safety assessment of a deep repository for spent nuclear fuel is presented. Since the majority of the radionuclides are embedded in the uranium dioxide fuel matrix, they will be released only after the alteration/dissolution of the matrix. Therefore a description of the process of alteration/dissolution of the spent fuel matrix is needed in a safety assessment.

Under normal repository conditions, i.e. reducing environment and neutral to alkaline pH, uranium dioxide has a very low solubility in water. If solubility is assumed to be the limiting factor, the dissolution of the fuel matrix will proceed very slowly due to the low water exchange in the defective canister. On this basis, a solubility-limited model for the release of the radionuclides from the fuel may be formulated.

The reducing conditions can be upset by the radioactivity of the spent fuel, which generates oxidizing products through water radiolysis. This causes the oxidative alteration/dissolution of the UO2(s) matrix. A model for fuel matrix conversion resulting from radiolytic oxidative dissolution is discussed, as well as parameter variations and the associated uncertainties.

In a repository the spent fuel will come in contact with groundwater after the copper canister has breached. Large amounts of hydrogen are then produced through the anoxic corrosion of the cast iron insert. Recent data on spent fuel leaching in presence of repository relevant hydrogen pressures and the implications on the actual and future spent fuel dissolution modeling will also be discussed.

This study treats radionuclide transport calculations for a canister defect scenario in the safety assessment SR 97, which concerns a deep repository for spent nuclear fuel of the KBS-3 type in Sweden. The aims of the calculations are to:

Quantitatively describe the radionuclide transport.

Show the impact of uncertainty in input data and show which parameters govern the calculated release rates.

Compare three different real sites in Sweden (Aberg, Beberg and Ceberg) with each other and with dose limits given in Swedish regulations (none of the sites is considered in the on-going localization process). Only briefly described in this paper.

Illustrate the impact of the different barriers in the system.

Deterministic calculations illustrate the radionuclide transport for reasonable conditions. Uncertainty cases show the influence of the uncertainty for data related to different parts of the repository system by systematically giving them pessimistic values while all others are reasonable. Simplified probabilistic calculations have also been performed.

The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. In the far field the most important uncertainties affecting release and retention are connected to permeability and connectivity of the fractures in the rock. The dose rate in the biosphere is essentially controlled by the possibilities of dilution.

The calculated maximum doses for the hypothetical repositories are well below the dose limits, and hence they meet the acceptance criteria for a deep repository for spent fuel.

The Waste Isolation Pilot Plant (WIPP) is located at a depth of 655 m in bedded salt at a site about 40 km east of Carlsbad, New Mexico.The 1996 U. S. Department of Energy (DOE) application for certification by the U. S. Environmental Protection Agency (EPA) included results of a performance assessment (PA) for the planned repository.After the EPA certified the facility in May, 1998, emplacement of contact handled transuranic waste (CHTRU) began in March, 1999.The WIPP facility must undergo a recertification by EPA every 5 years to demonstrate compliance with disposal regulations.Performance assessment is expected to be a key part of the recertification process.The PA will include probabilistic calculations to predict the release of actinides to the accessible environment over a 10,000 year period using a variety of plausible scenarios.The 1996 PA used a model for Pu(IV) solubility based on Th data and the assumption of analogous behavior for all actinides in the (IV) oxidation state.That model did not allow for mobility between the various possible oxidation states of Pu.The possible effects of increases in solubility of Pu through complexation with organic ligands was also not included in the PA because it was argued by DOE that such complexation would be insignificant. Subsequent data from tests using actual TRU wastes show strong evidence that the importance of organic ligands and the potential for multiple oxidation states must be carefully considered in estimating the expected solution concentrations of Pu in WIPP brines.This paper reviews the present state of knowledge of the behavior of Pu-containing wastes in contact with brines similar to those expected to be important for the WIPP repository, recent advances in understanding of the nature of Pu-oxide phases, and analytical methods that have high sensitivity for Pu speciation. A conceptual model for Pu in TRU wastes is presented and a series of steps - including experiments and calculations - that could led to an improved basis for the PA calculations to be done during the WIPP recertification process is outlined.

Engineered barrier systems are designed to reduce the near field actinide concentrations in case of water penetration into a repository. In this paper, the influence of buffer materials, such as MgO/CaO and clays, on the solubilities of Am, Np, Pu, and U is studied. The analysis is performed for low level cemented waste forms in a rock salt formation in contact with MgCl2 saturated salt brine (Q-brine).

The evolution of the geochemical milieu by cement corrosion is calculated using reaction path modeling supported by the code EQ3/6. The influence of different buffer materials is analyzed with respect to their impact on the solution chemistry and corresponding actinide concentrations.

The potential technical, scientific, safety, security, environmental and economic advantages of implementation of deep geologic repositories for the disposal of nuclear waste within an international framework were outlined in papers from the IAEA and from the Pangea organization over the past three years. Of increasing importance are the security and non-proliferation benefits which could result from having closely supervised centralized repositories for declared surplus fissile materials resulting from disarmament activities in nuclear weapon states.

The Pangea concept is for repositories sited on technical and safety grounds independent of national boundaries. The sites would be in geologic formations situated in flat, stable, arid regions, where the convincing demonstration of long-term safety should be easier to achieve. Pangea identified several potential host regions of the world which warrant further investigation of their suitability for an international repository. As yet a formal feasibility study has been initiated only in Australia. This was allocated first priority due to the vastness of the region with potentially suitable geology, also because of its stable governmental system, its strong environmental record and non-proliferation credentials. Australia also has a strong anti-nuclear movement, no nuclear power plants, but an important uranium mining industry.

Hence progress to date in Australia has been mixed. Political resistance has culminated in legislation in Western Australia prohibiting the construction and operation of a repository for the disposal of imported nuclear waste, unless approved by both houses of the WA Parliament. Within the scientific, technical and business communities, there has been much greater readiness to objectively weigh up the pros and cons of hosting an international disposal facility. The public reaction has been mixed, with some individuals and communities strongly opposing the concept whilst others are prepared to seek information before judging the case on its merits.

Meanwhile, on the world stage, the generic concept of international disposal facilities continues to be elaborated - including defining better the roles which such facilities could play in management and disposition of surplus weapons material. In addition contacts are being strengthened with potential users of such facilities and actions being taken to intensify Pangea's activities in other countries identified as having potential host regions for a high-isolation repository of the type proposed by Pangea.

Pangea Resources International AG, Switzerland, has put forward the concept of an international geological repository in an Australian desert area, for final disposal of certain of the world's high-level radioactive wastes and surplus fissile materials. Already publicly contentious in Australia, the concept raises unusual issues in safety and environmental assessment. It is timely, therefore, to consider why and how an initial case for such a repository might be made, in comparison with national options for high-level waste disposal that it might displace.

There are several stages ahead of development of a formal safety case for an international repository. The first, and almost certainly the most challenging, is winning of public and political acceptance in the host country that such a repository is the ‘best’ solution to a global problem.

We consider the basic need to identify and encompass the concerns of widely disparate stakeholders — industry, governments, national and international regulators, environmental interest groups and the public at large — so that public and political debates can be informed effectively. Many key issues will require comparison both of risks arising from very different operations, and of dissimilar prospective safety performances of complete disposal systems over periods spanning thousands of generations. Nevertheless, we conclude that the validity of such a comparative assessment could be assured by consistent application of a judicious blend of assessment techniques across the alternatives. We also conclude that its usefulness as a vehicle for public discussion would be enhanced by careful attention to public concerns, and by transparently independent review by scientific, technical, sociological and ethical specialists.

A study of heat propagation in a very deep borehole radioactive waste repository for Synroc is presented. The deep borehole modeled for the high level wastes is hypothetically sited at approximately 4 km below ground level in a granite host rock. The high level waste containers are placed in the lower section of the borehole to within 1.5 km of the surface and backfilled. A geothermal gradient of 0.03°C/m is assumed. Transient temperature field calculations are carried out for both 20 wt% and 10 wt% high level radioactive waste (HLW) bearing Synroc, for cooling periods between reactor discharge and geological disposal varying from 5 years to 50 years. The temperatures in the borehole repository were found to be very sensitive to the surface cooling period as well as the waste loading. The initial heat input of the waste may be controlled by cooling in an interim storage until the total heat output is reduced to a suitably low level. Therefore, surface cooling of at least 15 years for 20 wt% HLW containing Synroc is necessary for the given deep bore hole configuration to avoid the centerline temperatures that might affect durability of Synroc if ground water obtained access to the waste form. On the other hand the corresponding minimum surface cooling period for 10 wt% HLW containing Synroc is predicted to be about 5 years to produce similar temperature profiles in the borehole to those of 20 wt% HLW bearing Synroc with a 15-year cooling period before disposal. A variation in the borehole diameter has also been performed to elucidate the sensitivity of the temperature field of the repository.

Over the past 20 years, radioactive waste microbiology has included a variety of analyses of disposal options for High Level (HLW) and Low/Intermediate Level (L/ILW) waste. Projects have covered a wide range of areas including: fundamental microbiology of geological formations; microbial tolerance to extreme conditions; biodegradation of repository materials and the interaction of radionuclides with microbes and their by-products. The work has included field sampling, laboratory studies, natural analogue investigations and development of mathematical models. Progress in the topic is considered by evaluating each part of the repository system in terms of microbial presence, possible microbial influence on the performance of the engineered and natural barriers, and how such influences can be quantified. From this perspective, priorities for future work in this area are identified.

A model developed for description of waste glass corrosion has been applied to assess the radionuclide release from real radioactive (intermediate level) vitrified material over extended storage periods. Field data generated during the long-term testing of the prototype waste glass packages were mathematically processed and the derived parameters used in model calculations. Regardless of the corrosive saturated conditions of the wet near-surface repository, the fairly high safety of trench disposal has been demonstrated for borosilicate glass containing real NPP- operational waste.

From the point of view of clear demonstration of long-term safety, some of the long-lived intermediate-level wastes (ILW) are probably the most challenging of all waste types. Repository concepts for such wastes are reviewed with respect to optimization of operational procedures, post-closure performance, ease of making the safety case and costs. In particular, the advantages of designs incorporating a hydraulic cage and partitioning of wastes with different properties are discussed.

Knowledge of the permeability of hydraulically-tight rock at great depth is crucially important for the design and/or assessment of facilities associated with underground disposal of radioactive nuclear wastes. This paper presents a recently developed laboratory permeability test system capable of testing low permeability rocks either by using air as a permeant or by the transient-pulse method under high confining pressure conditions that simulate ground pressures at depths. The new system was used to test Shirahama sandstone and Inada granite, which are two types of rock widely available in Japan. To investigate the effects of heterogeneity on rock permeability, specimens cored parallel to and perpendicular to bedding for sandstone, and specimens cored in the direction perpendicular to Rift Plane, Grain Plane and Hardway Plane for granite, were used. The results of this study showed that: 1) gas permeabilities of a dried rock specimen tested by air permeation are almost the same values as water permeabilities of the same saturated rock specimen tested by the transient-pulse method; 2) the intrinsic permeabilities of Shirahama sandstone and Inada granite range from about 8.33E-16 to 7.38E-17 m2 and from 1.86E-17 to 6.94E-20 m2, respectively. They decrease monotonously with increase in effective confining pressure (defined as the difference between the confining and pore pressures), while the rate of decrease diminishes at higher confining pressures. The reduction in permeabilities is due to the closure of microcracks that control fluid flow at low confining pressures; and 3) Inada granite is a heterogeneous and isotropic material. Its hydraulic heterogeneity is more significant in Rift Plane than in Hardway and Grain Planes.

The concept of deep geological disposal was developed in order to permanently remove radioactive wastes from the human environment. Retrieval of wastes was reckoned to be a scenario of such low probability that little effort was initially devoted to its study. In recent years, however, there has been an increasingly active debate on whether fully passive (and safe) systems provide a sufficient level of practicability of retrievability. This debate is linked directly to practical, technical matters, such as the design of the facility, the operating procedures and also the institutional programs throughout the lifetime of a repository. But there are also philosophical issues involved in addition to these purely technical issues.

For retrievability, the questions are: How easy does retrieval have to be in the different stages of repository development? What is the rationale for requiring retrievability? What technical measures and methods are feasible? Should specific features facilitating retrievability be introduced into the repository design? How do such measures impact on other aspects of system performance and on other issues (such as safeguards)? This paper addresses the questions raised and presents a subjective set of conclusions, as a means to stimulate further debate on this topical issue.

Diffusion and sorption in the rock matrix are important retardation mechanisms in radionuclide transport in fractured media. The surface area in contact with the flowing water so called flow-wetted surface controls the interaction with the rock matrix. The flow-wetted surface, FWS, may be determined from the frequency of open fractures intersecting a borehole. The choice of the packer distance used in these hydraulic measurements is however crucial. If the packer distance is large, several open fractures may be found in the same packer interval. Analytical solutions and numerical calculations are used in order to address the determination of the flow-wetted surface from borehole data. The study is focused in two important aspects of the fieldwork, namely the distance used between the packers and b) the number of measurements that are required in order to obtain a accurate determination of the flow-wetted surface. A large volume of hydraulic data in Sweden obtained by using quite small packer distance is revised. The most usual packer distance was 3 meters. However, for some boreholes or sections of them even smaller packer distances were used. From these data, the flow-wetted surface was determined at several sites in Sweden. In some sites, the used packer distance was too large that no accurate determination of the flow-wetted surface could be done.

Scenarios are used to deal with uncertainty on the evolution of complex systems. The scenario selection methodologies also have to provide a basis for reflexion and traceability of the reasoning. In the case of a simple concept landfill, a methodology that fulfils these requirements has been developed. It distinguishes on the one hand the landfill evolution, and on the other hand, the human behavior evolution. It is based on three steps: (i) characterization of the landfill itself, in order to define situations leading to releases of radionuclides, (ii) characterization of the human behavior in order to define situations leading to intrusions, and (iii) characterization of the interfaces between the two subsystems, in order to identify situations which could lead to a radiological risk for man. The method has been applied to performance assessment for landfill of uranium mill tailings. It has led to one reference scenario, and five altered scenarios taking into account specific landfill evolutions such as loss of covering or loss of integrity of the dike, and human intrusions such as housing on the site of landfill or excavation of the site.

The long-term radiation effects on materials in the near-field of a nuclear waste repository have been evaluated using accelerated laboratory experiments with energetic electron or ion beam irradiation. The materials studied include: zeolites, layered silicates (smectite clay and mica), as well as crystalline silicotitanate (CST) which is an important ion exchange material for the chemical separation of high-level liquid radioactive wastes.

In situ transmission electron microscopy (TEM) during irradiation by energetic electrons and ions has shown that all of the studied materials are susceptible to irradiation-induced amorphization. At room temperature, complete amorphization was observed after ionizing doses of 1010 ∼ 1012 Gy or displacement doses on the order of 0.1 dpa (equivalent to doses received in 400-1,000 years for a high-loading nuclear waste form). Amorphization may be preceded or accompanied by dehydration, layer spacing reduction and gas bubble formation. In the case of zeolites, CST and some layered silicates, radiation effects are significantly enhanced at higher temperatures. Our experiments have shown that amorphization or even partial amorphization will cause a dramatic reduction in ion exchange and sorption/desorption capacities for radionuclides, such as Cs and Sr. Because the near-field or chemical processing materials (e.g. zeolites or CST) will receive a substantial radiation dose after they have incorporated radionuclides, our results suggest that radiation effects may, in some cases, retard the release of sorbed or ion-exchanged radionuclides.