[en] The description of redox conditions in the near field of a nuclear waste repository is an important but difficult aspect in performance assessment. Redox potentials are affected by both the thermodynamics and kinetics of relevant reactions, some of which are not adequately understood. This leads to considerable uncertainty of redox conditions in the repository environment and often to oversimplified terminology in performance assessment such as ‘reducing’ or ‘oxidising’. In this study we assess redox conditions by a holistic approach that considers all relevant sources of information. We apply this approach to the near field of the two types of repositories foreseen in the Swiss high-level waste programme: the spent fuel and high level waste (SF/HLW) and the intermediate-level waste (ILW) repositories. Although the environments surrounding these two waste streams are quite different, namely bentonite backfill versus cement, the procedures for describing redox conditions are similar. Thus, for both cases we first describe the layout of the repository and the properties of materials present in the near field. Then the duration of the initial oxic phase is estimated with the aid of limiting cases. The major part of this study focuses on the thermodynamic relationships and kinetic processes in the engineered barrier once oxygen has been depleted. Finally, from the combined set of information, reasonable ranges of long-term redox potentials are derived. After a relatively short initial oxic phase (< 100 a) the conditions in the bentonite backfill will become and remain reducing. The redox potentials will be largely influenced by the corrosion of steel, which will produce large amounts of magnetite, on the internal side of the bentonite barrier, and by the reducing conditions of the surrounding Opalinus Clay on the external side. The derived Eh range of -100 mV to -300 mV (SHE) for the anoxic stage mainly reflects the relatively large uncertainties in the pH of the porewater. The calculations suggest that the uncertainty with regard to the nature of the Fe(III)-Fe(II) solid phases is less significant in determining the derived redox potentials. The calculated redox potentials are consistent with recent experimental data on the reduction behaviour of U(VI), Tc(VII) and Se(VI/IV). The possible effect of high hydrogen pressures on redox potentials was not included in the analysis because the experimental data on the reactivity of H_2(g) in bentonite are limited. This also holds for Fe(II)-rich silicate phases which may play a role at the canister - bentonite boundary, although significant effects on the redox potentials are not expected. Further experimental data on these systems would be useful for future performance assessments. Heterogeneous reprocessed waste embedded in a cementitious matrix is grouped into two spatially separated waste types, ILW-1 and ILW-2. These two types consist of very different redox-sensitive materials and are assessed separately. After relatively rapid depletion of residual oxygen the conditions in the ILW-1 repository will remain reducing. The redox potential will be largely influenced by steel corrosion producing thin magnetite-type films on steel surfaces. Based on Fe(III)/Fe(II) equilibria calculations, the derived redox potentials for the reducing stage are estimated to be between -750 and -230 mV (SHE). The redox conditions in ILW-2 are expected to be rather similar; however, they might be more oxidising if high nitrate concentrations persist over long time periods. In this case an upper Eh limit of +350 mV (SHE) is estimated. The uncertainties with regard to the redox potentials in solution are large, mainly because of the lack of unequivocal experimental information on the phases forming during long-term corrosion of steels and the limited knowledge on iron-bearing cement phases. In addition, the importance of microbially-induced organic matter degradation, though considered to be of minor importance, is not adequately understood. If significant degradation occurs, lower redox potentials would be expected. This would also be the case if H₂ produced by the corrosion process were more reactive than commonly assumed. Further experimental work focussing on the steel corrosion under alkaline conditions and the identification of iron-bearing phases in the cement repository is needed to improve the understanding of the relevant redox processes. Also, the behaviour of redox-sensitive elements, such as U, Tc and Np in cementitious environments, should be experimentally investigated

[en] This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a low- and intermediate level waste (L/ILW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects arising principally from the heat generated by the waste and the setting of cement. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement caverns and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock. Deep geological repositories are designed to avoid or mitigate the impact of potentially detrimental repository-induced effects on long-term safety. For the repository under consideration in the present report, an assessment of those repository-induced effects that remain shows that detrimental chemical and mechanical impacts are largely confined to the rock adjacent to the excavations, thermal impacts are minimal and gas effects can be mitigated by appropriate design measures to reduce gas production and provide pathways for gas transport that limit gas pressure build-up (engineered gas transport system, or EGTS). Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The disposal system described in this report provides a system of passive barriers with multiple safety functions. The disposal system has been designed to perform with sufficient safety margin for a range of siting conditions. The barriers include the host rock, its surrounding geological setting, the waste forms, drums and the cementitious backfill. They have a range of attributes that intrinsically favour safety and that avoid or minimise detrimental phenomena and uncertainties or mitigate their effects. Nevertheless, potentially detrimental repository-induced effects remain and in the present report, these are investigated and discussed considering a broad spectrum of parameters, reflecting, among other things, the range of potential siting conditions. The L/ILW emplacement caverns are designed, constructed, operated and finally backfilled in such a way that formation of excavation damaged zones is limited. Specifically this is achieved by restricting the size of the excavations and the depth of the repository, using a low-deformation, controlled construction and excavation method and by the fact that the excavations will be backfilled relatively soon after construction with grain supported mortar. At expected repository depths, the caverns will need to be supported to ensure stability and worker protection; this will prevent rock falls and further extension of the EDZ. Based on the modelling results, it can be concluded that the extent of the EDZ around the L/ILW emplacement caverns will not exceed a thickness of one cavern diameter and that the hydraulic conductance of the EDZ around the emplacement caverns, access tunnels and shafts will not exceed a value of 10^-7 m³/s. Self sealing of the EDZ and low hydraulic gradients along the tunnels will result in negligible radionuclide transport by the EDZ pathway. It is shown that gas pressure build-up is controlled by the gas transport capacity of the pathways between the main repository and the access tunnel forming the so-called EGTS. Results obtained with the sensitivity cases for a repository depth of 500 m below ground level indicate that the EGTS can be designed in such a way that gas pressures which could damage the repository and/or the host rock will not be reached. When designing and assessing the performance of a L/ILW repository, the relevant chemical interactions are taken into account. With the current reference design, it is expected that degradation of the cementitious backfill, the concrete tunnel liner and the corrosion of the steel drums and other supporting structures will lead to some alteration of the cementitious nearfield and Opalinus Clay. These detrimental effects are taken into account in dose calculations and have been found not to have a significant impact on the calculated dose rates. (authors)

[en] This comprehensive brochure published by the Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) discusses the many important steps in the management of radioactive waste that have already been implemented in Switzerland. The handling and packaging of waste, its characterisation and inventorying, as well as its interim storage and transport are examined. The many important steps in Swiss management of radioactive waste already implemented and wide experience gained in carrying out the associated activities are discussed. The legal framework and organisational measures that will allow the selection of repository sites are looked at. The various aspects examined include the origin, type and volume of radioactive wastes, along with concepts and designs for deep geological repositories and the types of waste to be stored therein. Also, an implementation plan for the deep geological repositories, the required capacities and the financing of waste management activities are discussed as is NAGRA’s information concept. Several diagrams and tables illustrate the program

[en] In a deep geological repository, the access structures function as the link between the surface and the installations and structures at the disposal level. In the planned implementation scenarios, at least two access structures will be in operation up to the time of closure of the repository. The radioactive waste will be transported via the main access from the surface to the disposal level during emplacement operations. For the construction and operation of a deep geological repository, additional access structures are required. These auxiliary accesses and the associated surface infrastructure (e.g. shaft head installations) form the subject of this report. To provide as broad and comprehensive a description as possible, seven types of auxiliary access facilities are defined; these are characterised in line with the current status of planning and their functions and impacts are described. During construction, operation and dismantling of auxiliary access facilities, the usual conventional safety measures (inter alia) have to be observed (e.g. groundwater protection, fire prevention, facility security, accident prevention). Regarding the 'Ordinance on Protection against Major Accidents' no large quantities of hazardous materials, i.e. above the corresponding threshold quantities, are to be expected in the auxiliary access facilities. Proper handling and compliance with applicable regulations in all phases will ensure no hazard to humans and the environment. As no handling of radioactive materials is foreseen in the auxiliary access facilities, and because exhaust air and waste water from the controlled zones of a repository will, in principle, be removed via the main access and not the auxiliary accesses, a safety-relevant emission of radioactive substances and transport of contaminated material can be ruled out for the auxiliary access facilities during both normal operation and also in the case of an accident. Based on the information presented in this report, NAGRA assumes that the auxiliary access facilities of a deep geological repository can be realised in compliance with prescribed norms and standards and operated safely. The actual locations of these facilities will be concretised in a later project phase, with the concerns and wishes of the affected region being taken into consideration within certain boundary conditions. (authors)

[en] Canister design concepts for the disposal of spent fuel in repositories in both crystalline and Opalinus Clay host rocks are proposed, based on a review of the functional and performance requirements for such canisters, proposed design criteria and an assessment of repository conditions and their impact on long-term performance of possible canister materials. Two proposed canister lifetime targets of 1000 and 100 000 years are considered, based on experience from a variety of performance assessment studies in a number of countries, including Switzerland. The two canister options proposed and evaluated in detail that could meet the lifetime requirements are a thick-walled (∼ 15 cm) carbon steel canister and a composite canister with a copper external shell and a cast iron insert to provide structural integrity (the proposed SKB/Posiva canister). The cast steel canister is at a conceptual design stage, thus from the manufacturing perspective, only the basic feasibility of fabricating a canister shell has been considered. For an evaluation of the long-term integrity, the structural behaviour of the shell under isotropic loading conditions in the repository has been considered, along with a detailed assessment of the impact of various corrosion mechanisms on canister lifetime. The corrosion evaluation indicates that the short (some decades) aerobic phase of the repository would lead to very limited general and pitting corrosion (approximately 1 cm). Subsequent anaerobic corrosion is expected to occur at a rate of 1 μm a^-1. Evaluation of other mechanisms such as microbial corrosion, stress-corrosion cracking and hydrogen damage indicates that they are not expected to lead to canister breaching, thus a lifetime for a steel canister is expected to be at least 10 000 years, well in excess of the target lifetime of 1000 years. The structural analysis indicates that, for the expected total depth of corrosion of 2 cm, the canister has sufficient strength that structural loads would not lead to breaching within 10 000 years. The corrosion assessment of the copper canister for crystalline and Opalinus Clay repository conditions suggests a lifetime of at least 100 000 years, in line with Swedish and Finnish assessments. (authors)

[en] In this report, the procedure and the results of an inverse modelling study on the through-diffusion of tritiated water (HTO) and ²²Na⁺-ions are presented using high-porous hardened cement pastes with a water/cement ratio of 1.3 in the first stage of the cement degradation. For the analysis two alternative models were applied: 1) a diffusion model where a possible sorption of the tracer was entirely neglected, and 2) a diffusion model with linear sorption. The analysis of the through-diffusion phase allowed extracting values for the effective diffusion coefficient (D_e) and the rock-capacity factor (α). Both models could fit the breakthrough curves equally well, and also mass-balance considerations did not allow to clearly preferring one of the two competing models to the other. But blind-predictions for tracer out-diffusion using the best-fit parameter values deduced from analysing the former through-diffusion phase gave a clear indication that linear sorption had to be included in the diffusion model. The extracted K_d values for HTO are in excellent agreement with values from batch sorption experiments and are of the order of 0.8 · 10^-3 m³/kg. Those for ²²Na⁺ are of the order of 1.0 · 10^-3 m³/kg and are by a factor of two larger than values from batch sorption experiments. The values for the effective diffusion coefficients for HTO are of the order of (2-3) · 10^-10 m²/s, and those for sodium are roughly by a factor of two smaller than values for HTO. On the one hand, the observed tracer uptake could only partially be addressed to isotope exchange; the most obvious process which could account for the remaining part of the uptaken tracer mass is diffusion into a second type of porosity, the dead-end pores. On the other hand, the results and conclusions drawn are encouraging for future investigations; therefore no major deficiency concerning the applied equipment and the modelling methodology could be detected. In the report, however, some suggestions for new and improved experiments are made which could shed light on the tracer deposition mechanisms playing a crucial role in diffusion experiments using cementitious materials. (author)

[en] Recently, a bentonite sorption data base, comprising values taken from batch sorption data, was developed for a performance assessment study for high-level waste and spent fuel (Entsorgungsnachweis). Thus distribution coefficients (K_d) determined on dispersed systems were used to calculate apparent diffusion coefficients (D_a) subsequently applied in diffusive transport calculations for the highly compacted system. Whenever such a procedure is adopted, questions invariably arise as to whether this is conservative or not. On the occasions when K_d values have been extracted from (mainly) indiffusion experiments and compared with those obtained from batch tests, apparent discrepancies have been found. In the majority of cases the batch values are larger, sometimes significantly. Hypotheses from 'surface diffusion' to 'double layer pore constrictivity effects' have been proposed to explain the inconsistencies. However, although such discrepancies have been reported periodically over the past twenty years or so, and have become generally accepted facts of life, there are surprisingly few quantitative studies directly dealing with this issue. Further, two other points are worthy of mention. The first is that a diffusion model (including the associated assumptions) is needed in order to deduce K_d values from diffusion measurements. Thus the sorption values calculated are model dependent. The second is that too little attention has been paid to the potential effects of water chemistry, i.e. a comparison between sorption values is only valid when the water chemistry in the batch tests is the same as, or very close to, the porewater chemistry in the intact material. In practice, this condition is difficult to achieve because of the uncertainties concerning the latter. This report describes a study in which K_d values for Cs(l), Ni(II), Sm(III), Am(III), Zr(IV) and Np(V) were calculated from in-diffusion data published in the open literature for a Na-bentonite (Kunigel V1) at dry densities between 400 and 2000 kg m^-3 . The range of oxidation states of the elements considered provides a good representation of those expected in a radioactive waste repository. A porewater chemistry was calculated for each dry density and used in conjunction with sorption models and/or sorption data from batch measurements to produce blind predictions for K_d values for the compacted Kunigel V1 bentonite. These K_d values combined with effective diffusion coefficients (D_e) for tritiated water (HTO) were used to calculate D_a values as a function of dry density and compared with the corresponding D_a values from diffusion measurements. An important motivation for this study was to see whether discrepancies did in fact exist between calculated and measured D_a values originating from batch and diffusion experiments when 'state of the art' knowledge concerning sorption processes and bentonite porewater chemistry was applied to a specific system. The preliminary conclusion drawn is that, in general, the differences between D_a values calculated from batch K_d measurements and D_e (HTO) values, and those measured in-diffusion tests are not great. However, an important consideration is the bentonite porewater chemistry. (authors)

[en] This report reviews the information base and understanding gained in the course of the borehole sealing / filling of the reconnaissance borehole SB4a/slanted (SB4a/schräg). This slanted reconnaissance borehole was drilled in 1994/95 as part of the investigations at Wellenberg in view of a potential repository site for low and intermediate level radioactive waste. Because this particular borehole penetrated the immediate vicinity of the rock body considered to house the repository it was necessary to seal the borehole according to safety technical criteria stipulating a minimum release of radionuclides. The project had pilot-study character as it was for the very first time that measures like these were implemented in Switzerland. The host rock penetrated by the reconnaissance borehole SB4a/schräg encompasses formations at the bottom of the Drusberg-Decke and the top of the Axen-Decke: Palfris Formation, Vitznau Marls, Interhelvetic Mélange and the Tertiary schists of the Globigerina Marl and the Schimberg Shales. From a hydrogeological point of view, the host rock is considered a fractured medium with an extremely low permeability matrix. For practical purposes the formation water is conducted exclusively by the structural features resulting from brittle deformation (such as cataclastic fault zones) and those ductile features which have been re-activated in a brittle manner. The feasibility of borehole sealing had been proven in principle investigations previous to the project's commencement. The prevention of potential flow paths along the borehole, and between the potential repository and the biosphere was deemed possible in principle. However, the selection and sequential placement of the sealing materials was to be adapted to the site-specific host-rock conditions and the in-situ conditions prevailing in the borehole. The sealing concept for the reconnaissance borehole SB4a/schräg was based on a multiple component system whereby the sealing effect was achieved by the sequential placement of materials with different chemical and physical properties. Cements used in deep boreholes and swelling cements are considered filling materials. Sealing materials are barite and clay pellets. The materials' properties were tested extensively in preliminary laboratory experiments. The sealing concept was developed into a sealing programme which took into account the geological and hydrogeological conditions revealed at the site. As a result, it was possible to seal the reconnaissance borehole SB4a/schräg as planned and without notable complications in a two-week field campaign. An examination of safety considerations resulted in the identification of relevant parameters which affect the performance of the barrier with respect to nuclide migration: formation water flux across the borehole system; length of the sealed/filled borehole section; half life and sorption properties of the nuclide concerned. Long-lived or non-sorbing nuclides are barely or not at all held back in the sealed borehole. The impact of the borehole is comparable to flow paths along cataclastic zones. It is due to the high retaining capacity of the near field and the low release of these nuclides that the corresponding radiation exposure however remains below the protection target of 0.1 mSv/a. Given the results of extensive modelling studies with variable assumptions concerning the material properties and with consideration of the short and long-term safety, the performance of the sealed/filled borehole system SB4a/schräg was concluded to be adequate. Accordingly, the sealed borehole SB4a/schräg may be regarded as an element which equals the host rock in its performance as a nuclide barrier. Meticulous planning, implementation and quality assurance made it possible for the sealing and filling of borehole SB4a/schräg to be realised in a manner which allowed to meet the safety relevant requirements concerning the release of radionuclides. (author)

[en] The management of spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (TRU) principally from reprocessing is based on the concept of deep geological disposal, i.e. long-term isolation of the waste in suitable, deep-lying rock formations. The first project studies carried out by Nagra in this connection date back more than 20 years and looked at the option of disposal in crystalline basement rock and clay. The disposal strategy developed by Nagra over the years ties in closely with the concept of 'monitored, long-term geological disposal' as formulated in the most recent requirements of the authorities. This report forms part of the series produced for the 'Entsorgungsnachweis' Project, which also includes a geological synthesis report on the region of the Zürcher Weinland and a safety assessment report. The purpose of the Project is to demonstrate the feasibility of disposing of SF/HLW/TRU in Northern Switzerland. The aim of this report is to investigate the engineering feasibility of constructing a repository for SF/HLW/TRU in the Opalinus Clay of the Zürcher Weinland and to provide project-specific input for the long-term safety assessment. Therefore, a concept for the facilities and operation of the repository was elaborated. The individual structural elements and components for which the feasibility demonstration was performed are part of a modular system, which is brought together to form a stand-alone project, presented in this report as the Reference Project. This Reference Project is the end-result of the procedure summarised below, which consists of the following steps: - Outlining a general procedure for handling and storage of radioactive waste, including engineered barriers and facility design based on specific boundary conditions and requirements; - Approximate design of transport and handling equipment and specification of the clearance profiles for the different underground structures; - Determining the stress on key drift and tunnel cross-sections and preliminary design of rock support measures; consideration of construction procedures; - Reviewing operational safety, ventilation and consideration of retrievability; definition of the Reference Project drawing on experience from other construction projects; investigation of closure of the facility. In order to test the flexibility of the system, 'what-if' scenarios in the form of possible alternative solutions or alternative measures have been discussed on a case-specific basis. As a result of the work that has been performed, the following conclusion can be drawn: a deep geological repository in the Opalinus Clay of the Zürcher Weinland for spent fuel from the operation of the Swiss nuclear power plants and for vitrified high-level and long-lived intermediate-level waste mainly from reprocessing can be constructed and operated and can be closed within a few years using currently available technology and in accordance with legally prescribed safety standards. Societal requirements relating to monitoring and control, as formulated in the draft of the new Nuclear Energy Law of 2001, are fulfilled. The retrievability of emplaced waste is also assured. Spatial reserves exist and the concept for facilities and operation offers a high degree of flexibility for the continuation of the project. (author)

[en] When looking at the dose rate calculations made as part of the provisional safety assessment in Stage 2 of the Sectoral Plan (Nagra 2014a), it is remarkable that only few radionuclides contribute to the resulting dose rate. These radionuclides include C-14 (instant and congruent release), CI-36, Se-79 and I-129. The reasons for this observation are the limited retardation of these radionuclides in clay- and cement-based environment and/or their unknown chemical speciation. This report gathers information regarding inventory, chemical speciation in the relevant waste sorts and transport processes in clay and/or cement (sorption and diffusion). This is done for each of the following radionuclides: CI-36, Se-79, Ag-108m and I-129. As the research on C-14 speciation and retardation is ongoing, a separate report will deal with the contribution of C-14. Based on existing literature, the report investigates the origins of uncertainties, describes the complexity and nature of the waste in which the radionuclides are formed, evaluates the possible speciation for each element and makes suggestions of how the remaining uncertainties can be reduced in the future. The report comes to the conclusion that uncertainties in the inventory or in the speciation for CI-36, Se-79 and Ag-108m could be further reduced by dedicated studies. For CI-36, verifying the inventory of CI-36 in spent fuel, cladding and stainless steel can decrease the dose contribution. In case of Se-79, thermodynamic and modelling exercises to understand the Se(cr) ⇌ Se(aq) system and the uptake of Se in sulphides can reduce the dose contribution. A similar study on the Ag(cr) ⇌ Ag(aq) system could also elucidate the speciation and solubility of Ag-108m. Finally, this report however estimates the chance of reducing the dose rate contribution of I-129 by doing further studies as very low, as the chemical speciation and the inventory are well-known and have been extensively characterised. (author)