Technischer Bericht NTB 13-04

Longterm degradation of organic polymers under L/ILW-deep repository conditions

Summary

Technical Report NTB 13-03

Redox properties of iron-bearing clays and MX-80 bentonite – Electrochemical and spectroscopic characterization

Summary

The characterization of the redox properties of Fe-bearing minerals in the presence and absence of dissolved Fe2+ is of major relevance for the assessment of redox reactions in natural and engineered environments such as radioactive waste repositories. In this study, we developed an electrochemical approach based on the use of soluble organic electron transfer mediators, which enabled us to quantify the redox properties of Fe-bearing clay minerals, MX-80 bentonite and combinations of clay minerals, Fe oxides and dissolved Fe2+. Using mediated electrochemical oxidation and reduction, we quantified the electron accepting and donating capacities of ferruginous smectite SWa-1, Wyoming montmorillonite SWy-2 and MX-80 bentonite at pH 7.5. All structural Fe in clay minerals was redox-active in contrast to that present in other, not further defined phases of MX-80. The materials investigated were redox-active over a very wide range of Eh-values, that is the Fe2+/Fetotal ratio of the minerals changed from 0 to 100 % between +600 and -600 mV (vs. SHE). Redox properties were highly path-dependent due to structural changes of the minerals as revealed from the study of native and redox-cycled clay minerals after repeated reduction and re-oxidation cycles. Irreversible alteration of the mineral structure, however, was less obvious for materials with lower total Fe content such as MX-80 bentonite and SWy-2. Systems containing native montmorillonites (SWy-2 or MX-80), goethite and dissolved Fe2+ were also able to buffer the reduction potential E_H between 0 and -300 mV. Regardless of their Fe oxidation state, Fe-bearing minerals are redox-active over a wide potential range and therefore very relevant as redox buffers deter­mining the fate of redox-active radionuclides and metals in waste repositories.

Technical Report NTB 13-02

An Assessment of the Impact of the Long Term Evolution of Engineered Structures on the Safety-Relevant Functions of the Bentonite Buffer in a HLW Repository

Summary

Technischer Bericht NTB 13-01

Standortunabhängige Betrachtungen zur Sicherheit und zum Schutz des Grundwassers – Grundlagen zur Beurteilung der grundsätzlichen Bewilligungsfähigkeit einer Oberflächenanlage für ein geologisches Tiefenlager

Summary

Technical Report NTB 12-07

Geochemical Synthesis for the Effingen Member in Boreholes at Oftringen, Gösgen and Küttigen

Summary

Technical Report NTB 12-06

Canister Design Concepts for Disposal of Spent Fuel and High Level Waste

Summary

As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1'000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs.

The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements (minimum 1'000 year lifetime without breaching) and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters, with a BWR fuel canister selected as a reference, and HLW canisters, based on two vitrified HLW containers per canister, were then developed. Corrosion, degradation, failure mechanisms, tolerance to fabrication flaws and other structural performance issues have been investigated in the context of proposed weld designs, inspection and manufacturing. Different ways of achieving compliance with the canister requirements have been explored.

The resulting design concepts provide a description of geometric shape, dimensions, material, welding and fabrication and inspection options for the canisters, including the internal supporting structures. The corrosion, material degradation, structural performance, weld design, inspection and manufacturing issues of each option are discussed, and it is shown how the proposed designs meet the canister requirements. One of the benefits of the study is that the design development and analysis of performance led to the identification of the specific work needed to develop detailed design concepts and prototype canisters. The most important aspects are developing the weld design, weld method and approach to post-weld stress relief, as well as the determination of the maximum allowable defect size and safety factor in relation to crack growth probability.

Technical Report NTB 12-05

Comparison of Sorption Measurements on Argillaceous Rocks and Bentonite with Predictions Using the SGT-E2 Approach to Derive Sorption Data Bases

Summary

In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository.

A detailed procedure was developed for deriving SDBs for argillaceous rocks (and bentonite) based on sorption edge measurements on illite (and montmorillonite), the hypothesis that 2:1 clay minerals are the dominant sorbents and a series of so called conversion factors which take into account the different radionuclide speciations in the different porewaters.

Since this methodology for generating SDBs is relatively new, a validation and demonstration of the robustness and reliability of the sorption values derived was required. This report describes an extensive piece of work in which blind predictions of sorption values were compared with measured ones.

Sorption isotherms were measured for the following metal ions Cs(I), Co(II), Ni(II), Eu(III), Th(IV) and U(VI) in a range of realistic porewater chemistries for a range of host rock mineralogies. In the end 53 isotherm data sets were measured. For each of these isotherms a prediction was made of the sorption at trace concentrations using the SDB derivation methodology. A comparison between measured and predicted values for each case was then made.

This validation study shows that the methodology used for the derivation of the sorption data bases for argillaceous rocks and bentonite produces reliable sorption values.

Technical Report NTB 12-04

Sorption Data Bases for Argillaceous Rocks and Bentonite for the Provisional Safety Analyses for SGT-E2

Summary

In Stage 1 of the Sectoral Plan for Deep Geological Repositories, four rock types have been identified as being suitable host rocks for a radioactive waste repository, namely, Opalinus Clay for a high-level (HLW) and a low- and intermediate-level (L/ILW) repository, and 'Brauner Dogger', Effingen Member and Helvetic Marls for a L/ILW repository. Sorption data bases (SDBs) for all of these host rocks are required for the provisional safety analyses, including all of the bounding porewater and mineralogical composition combinations. In addition, SDBs are needed for the rock formations lying below Opalinus Clay (lower confining units) and for the bentonite backfill in the HLW repository.

In some previous work Bradbury et al. (2010) have described a methodology for developing sorption data bases for argillaceous rocks and compacted bentonite. The main factors influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers) and the water chemistry which determines the radionuclide species in the aqueous phase. The source sorption data were taken predominantly from measurements on illite (or montmorillonite in the case of bentonite) and converted to the defined conditions in each system considered using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab → Field conversion factor was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. This methodology to develop sorption data bases has been applied to the selected host rocks, lower confining units and compacted bentonite taking into account the mineralogical and porewater composition ranges defined.

Confidence in the validity and correctness of this methodology has been built up through additional studies: (i) sorption values obtained in the manner described above have been compared with those in already existing SDBs for Opalinus Clay and bentonite used in Project Opalinus Clay (Entsorgungsnachweis; see Bradbury & Baeyens 2010), (ii) blind sorption model predictions of isotherms on MX-80 bentonite and Opalinus Clay in realistic groundwater compositions have been compared with measured isotherms (Bradbury & Baeyens 2011), and finally, (iii) blind predictions made using the above methodology have been compared with recent sorption measurements on argillaceous rocks (Baeyens et al. 2014). In all cases the results obtained in the different comparative approaches have been consistent with the predictions made using the methodology described.

In a few cases the mineralogy of the rock type was too poor in clay minerals to apply this approach. SDBs were nevertheless developed but based on a methodology in which calcite was the main sorbing phase. The procedure is fully described in the present report.

Further, a methodology for the derivation of SDBs for a host rock altered by hyperalkaline solutions from a cement based repository is described and the resulting SDBs for Effingen Member and Helvetic Marls are given.

Technical Report NTB 12-03

Effective Diffusion Coefficients and Porosity Values for Argillaceous Rocks and Bentonite: Measured and Estimated Values for the Provisional Safety Analyses for SGT-E2

Summary

In Stage 2 of the Sectoral Plan for Deep Geological Repositories so-called provisional safety analyses have to be performed. Among other input data, geochemical parameters to describe the transport and retardation of radionuclides in the argillaceous rocks considered and in compacted bentonite are required. In the present report, a comprehensive set of diffusion parameters for all clay host rocks, confining units and compacted bentonite is derived.

Diffusion of tritiated water (HTO), ³⁶Cl^- and ²²Na⁺ was studied on samples from Effingen Member, 'Brauner Dogger', Opalinus Clay and Helvetic Marls using the through-diffusion technique described in detail by Van Loon & Soler (2004). Earlier measurements on Opalinus Clay are also shortly summarised in this report. The diffusion measurements gave values for effective diffusion coefficients and diffusion accessible porosities. The general observed trend ^NaD_e > ^HTOD_e > ^ClD_e is in agreement with the expected behaviour of the three species in clay materials (Glaus et al. 2010), i.e. ion exchanging cations show an enhanced mobility due to surface diffusion effects (Gimmi & Kosakowski 2011) and anions are slowed down due to anion exclusion, i.e. due to the negatively charged clay surfaces, anionic species are repelled from these surfaces resulting in an accessible porosity that is smaller than the total porosity as measured with HTO (Van Loon et al. 2007).

The effect of porewater composition on the diffusion of HTO, ³⁶Cl^- and ²²Na⁺ in Opalinus Clay was investigated. For ionic strength (IS) values between 0.17 M and 1 M (0.17 M ≤ IS ≤ 1.07 M), no significant effect on the effective diffusion coefficient could be observed. In the case of ³⁶Cl^-, no effect on the accessible porosity was observed. The anion diffusion accessible porosity equals 50 – 60 % of the total porosity, independent on the ionic strength of the porewater. For the other host rocks no experimental data are available. However, it is assumed that also for the other host rocks, no significant effect can be expected. Further investigations will be performed.

The diffusion parameters were compared with other data taken from the literature and measured on a series of sedimentary rocks such as chalk, clay and limestone rocks. All data could be described by one single modified version of Archie's relation (extended Archie's relation). For values of the porosity greater than ca. 0.1, the classical Archie relation was valid. For porosity values smaller than 0.1, the data deviated from the classical Archie relation, i.e. the decrease of De with porosity, was less fast. This phenomenon can be explained by additional changes of tortuosity with porosity values. At high porosity values, which can be found in low density rocks, the microfabric of the clay rock is of a house-of-cards type. With increasing density, the randomly oriented clay platelets become more and more oriented in a specific direction perpendicular to the direction of compaction. As soon as the platelets are more or less horizontally oriented, further decrease of the porosity has no longer an effect on the orientation and consequently on the tortuosity. The rock bulk dry density threshold value is ca. 2500 kg m^-3, representing a porosity value of 0.1.

The extended version of Archie's relation (e-Archie) is the basis for a procedure for estimating effective diffusion coefficients to be used in safety analyses. Important input parameters are the diffusion coefficient of the radionuclides in free bulk water and the transport relevant porosity. Although each radionuclide has its own free water diffusion coefficient, radionuclides were subdivided into two main groups with a free water diffusion coefficient of (20.0 ± 2.5) × 10¹⁰ m² s^-1 and (7.5 ± 2.5) × 10^-10 m² s^-1. The porosities to be used were given by Nagra and were derived mainly from drilling core samples of the host rocks. The total porosity was based on measurements of the bulk and grain density of the rocks. The values for anion accessible porosities were based on the observation that for most clay rocks ca. 50 % of the total porosity is accessible to anions.

In case of cations undergoing ion exchange, correction for surface diffusion was made using the method developed by Gimmi & Kosakowski (2011). A correction factor CF was calculated using the surface mobility of the cation and the sorption values (K_d) as calculated by Baeyens et al. (2014). The reference values of the effective diffusion coefficients, and their upper and lower bounding values, were multiplied with these correction factors. For most clay rocks investigated, the correction factors used ranged from 1 to at most 30, depending on the sorption value of the cation. The largest values were calculated for Cs⁺, which is known to be affected by surface enhanced diffusion in argillaceous rocks (Appelo et al. 2010, Melkior et al. 2005, Melkior et al. 2007). In case of Helvetic Marls, the correction factors were between 30 and 400. This is caused by the much larger capacity ratio (κ) values, which are directly proportional to the reciprocal transport porosity of the clay rocks.

As the depth of the host rocks varies notably in the siting regions, it is not possible to define a unique temperature value for the considered formations. Instead, a temperature range was defined for each potential host rock. Effects of temperature on diffusion were evaluated using the Arrhenius equation with an average activation energy for diffusion of 22.9 kJ mol^-1 (Van Loon et al. 2005a).
For each host rock, a table with effective diffusion coefficients was compiled. The tables contain a reference value at 25 °C that was calculated for the reference porosity using the master e-Archie's curve. Upper and lower values were estimated by combining the upper e-Archie's curve with the upper value for porosity, and the lower e-Archie's curve with the lower value for porosity. Only the effect for the maximum temperature boundary was considered. To this end, a combined uncertainty on porosity and temperature was estimated by error propagation. The calculated uncertainty was added to the reference value at 25 °C.

Technical Report NTB 12-01

The Long Term Geochemical Evolution of the Nearfield of the HLW Repository

Summary

The work presented in this report focuses on the spatial and temporal evolution of the nearfield of the high level radioactive waste repository situated in the Opalinus Clay formation.

The major components of the nearfield of such a repository are spent fuel, vitrified high-level waste, canisters (assumed for the purposes of the present report to be made of carbon steel), compacted bentonite and a concrete liner. Over the one million year time period considered in safety analysis, these components will chemically interact with one another and potentially change their retention characteristics.

As a starting reference point the "initial" (unreacted) states of the Opalinus Clay, bentonite, concrete liner (mineralogies and water chemistries) and the canister are briefly described.

The main processes considered to influence the evolution of the repository in time and space, and which often operate over different time scales, are: interactions of the concrete tunnel liner with compacted bentonite and Opalinus Clay, temperature gradients caused by the heat generating high level waste, mineralogical changes to the compacted bentonite through interactions with the corrosion products of the iron canisters, and finally, the dissolution of the spent fuel and vitrified high-level waste. The consequences of these processes (as a function of time) on the long term barrier performance of the nearfield have been estimated, particularly with respect to radionuclide solubilities and the sorption, diffusion and swelling characteristics of the bentonite.

The main conclusions drawn are as follows: The alteration depth into the bentonite due to the interaction with the concrete liner (assumed to be 15 cm thick) is likely to be much less than 13 cm over a one million year time scale, with the main reaction products being clays (illite), hydroxides, carbonates, calcium silicate hydrates, and aluminosilicates. The swelling pressure and the sorption capacity of the bentonite in this region will be reduced, but not to zero.

Experimental findings and modelling studies indicate that any alterations due to the post closure temperature transients will not change the swelling and retention properties of the outer (more than) half of the bentonite which can be relied upon to fulfil its buffer function fully.

The dissolution of spent fuel and vitrified high-level waste are not expected to have any detrimental effects on the sorption properties or the swelling capacity of the bentonite. However, the potential influence of the release of boron from the vitrified high-level waste on complexation with highly charged radionuclides needs to be addressed.

Current studies indicate that the Fe²⁺ released by the corrosion of the steel canisters can lead to the alteration of montmorillonite at temperatures below 100 °C to form Fe-rich smectites or non-swelling clays and chlorites. Fe-rich smectites have similar properties to the Na-montmorillonite they replace. Hence, the near-field barrier function will not be significantly influenced with respect to sorption and swelling. However, if non-swelling clay minerals or chlorites are formed, then noticeable changes are expected, reducing the bentonite swelling capacity and sorption properties.

The release of iron from canister corrosion is a slow process. Estimates based on the corrosion rate indicate that the canister corrosion and the following conversion of montmorillonite needs between 100'000 and 200'000 years. The situation described here represents a "worst case" scenario. Other iron phases like magnetite or siderite are stable in this environment and decrease the availability of Fe²⁺ for Montmorillonite transformation. This suggests that significant quantities of bentonite will still be available up to one million years after closure of the repository.

Technischer Bericht NTB 11-01

Vorschläge zur Platzierung der Standortareale für die Oberflächenanlage der geologischen Tiefenlager sowie zu deren Erschliessung

Summary

In line with the provisions of the nuclear energy legislation, the sites for deep geological disposal of Swiss radioactive waste are selected in a three-stage Sectoral Plan process (Sectoral Plan for Deep Geological Repositories). The disposal sites are specified in Stage 3 of the selection process with the granting of a general licence in accordance with the Nuclear Energy Act.

The first stage of the process was completed on 30th November 2011, with the decision of the Federal Council to incorporate the six geological siting regions proposed by the National Cooperative for the Disposal of Radioactive Waste (Nagra) into the Sectoral Plan for Deep Geological Repositories for further evaluation in Stage 2. The decision also specifies the planning perimeters within which the surface facilities and shaft locations for the repositories will be constructed.

In the second stage of the process, at least two geological siting regions each will be specified for the repository for low- and intermediate-level waste (L/ILW) and for the high-level waste (HLW) repository and these will undergo detailed geological investigation in Stage 3. For each of these potential siting regions, at least one location for the surface facility and a corridor for the access infrastructure will also be specified.

Nagra is responsible, at the beginning of Stage 2, for submitting proposals for potential locations for the surface facilities and their access infrastructure to the Federal Office of Energy (SFOE); these are then considered by the regional participation bodies in the siting regions. The present report and its appendices volume document these proposals. In Stage 2, under the lead of the SFOE, socio-economic-ecological studies will also be carried out to investigate the impact of a repository project on the environment, economy and society. The present reports also contain the input data to be provided by Nagra for the generic (site-independent) part of these impact studies.

A meaningful discussion of the proposals made by Nagra (cf. appendices volume) requires an understanding of the functioning of the components of the repository, supported by relevant basic data. This general report, which is independent of the proposed siting regions, provides an overview of the facilities and their functioning for both the L/ILW and the HLW repositories, the operating procedures and the impacts associated with construction and operation. The report

The appendices volume contains the proposals made by Nagra for the siting areas for the surface facilities within the six potential planning perimeters and how these will be developed and accessed. Following a general overview and brief presentation of the siting regions, the individual proposals for the locating the siting areas for the surface facilities are presented in fact-sheets.

These fact-sheets consist of presentations of the geographic situation together with a brief description of the siting areas using a uniform list of criteria and indicators. These criteria and indicators are also applied for the selection and characterisation of the siting areas, with the following overarching objectives:

Technischer Bericht NTB 10-01

Beurteilung der geologischen Unterlagen für die provisorischen Sicherheitsanalysen in SGT Etappe 2 - Klärung der Notwendigkeit ergänzender geologischer Untersuchungen

Summary

The procedure for selecting repository sites for all categories of radioactive waste in Switzerland is defined in the conceptual part of the Sectoral Plan for Deep Geological Repositories (SFOE 2008). The Plan foresees a selection of sites in three stages, the third of which leads to a General Licence Application procedure that defines both the sites and the main features of the repositories. In Stage 1, Nagra proposed geological siting regions based on criteria relating to safety and engineering feasibility. These siting proposals were subsequently evaluated and approved by the responsible federal authorities in their statements. The Swiss Government (the Federal Council) is expected to decide on the siting proposals in 2011 based on these statements. After submittal of the siting proposals for Stage 1, Nagra has started preparations for Stage 2. Working together with the siting regions and the affected Cantons in the context of a participatory process, the objective of Stage 2 is to prepare proposals for the location of the surface facilities within the planning perimeters defined by the Federal Council in its decision on Stage 1 and to identify potential sites. Nagra also has to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. Based on this, and taking into account the results of the socio-economic-ecological impact studies, Nagra then has to propose at least two sites for each repository type to be carried through to Stage 3. The proposed sites will then be investigated in more detail in Stage 3 to ensure that the selection of the sites for the General Licence Applications is well founded.

In order to realise the objectives of the upcoming Stage 2, the state of knowledge of the geological conditions at the sites has to be sufficient to perform the provisional safety analyses. Therefore, in preparation for Stage 2, the conceptual part of the Sectoral Plan requires Nagra to clarify with ENSI, at an early stage, the need for additional investigations aimed at providing input for the provisional safety analyses. The purpose of the present report is to document Nagra's technical-scientific assessment of this need so that ENSI can subsequently evaluate Nagra's assessment. The focus is on evaluating the geological information based on processes and parameters that are relevant for safety and engineering feasibility.

In evaluating the state of knowledge the key question is whether additional information (e.g. from future investigations) could lead to a different decision regarding the selection of the sites to be carried through to Stage 3. In order to set priorities based on safety, as required for Stage 2, it is necessary to use the characteristic dose intervals determined with the help of dose calculations, an evaluation of engineering feasibility and the results of a qualitative assessment. Accordingly, the state of knowledge is evaluated using test calculations to determine dose intervals, an assessment of the engineering feasibility and an evaluation of the qualitative assessment. The test calculations are used to derive dose curves for a wide spectrum of calculation cases for the different repository types in the geological siting regions under consideration. Besides the expected evolution (reference case), the calculations also cover existing uncertainties in the relevant processes and parameters. This allows the characteristic dose intervals required by the Sectoral Plan to be determined for the different siting regions for each repository type; these dose intervals are evaluated with respect to their unambiguousness concerning two aspects required by the Sectoral Plan: safety-related suitability and safety-related equivalence of the siting regions. For the purpose of evaluating safety, the state of knowledge is considered to be sufficient if clear, unambiguous statements can be made despite the fact that the parameter ranges are selected generously to allow for existing uncertainties and if these statements will not change if the uncertainties and associated parameter ranges are reduced through future investigations. Whether the information used to assess engineering feasibility is sufficient is also considered, along with an evaluation of the significance of uncertainties in the relevant processes and parameters. The analysis of the state of knowledge shows that, taking into account the investigations performed by Nagra since 2008 and those planned for the near future, the available information is sufficient for the provisional safety analyses and the safety-based comparison of the siting regions. For all siting regions, concrete statements can be made regarding suitability and equivalence of sites from the point of view of safety despite the deliberate selection of wide parameter ranges to take into account uncertainties; engineering feasibility has also been demonstrated. This means that, besides the work already completed or planned by Nagra, no additional investigations will be necessary for the provisional safety analyses to be performed in Stage 2 of the Sectoral Plan.

The report also describes the work already initiated and planned by Nagra for Stage 2 of the Sectoral Plan. It is envisaged that this work will contribute to reducing the uncertainties mentioned in this report, thereby at least partially reducing the corresponding parameter bandwidths. This work relates to the geometry (including structures) of the host rocks and the effective containment zones and to information on deposits of raw materials and on state parameters in the different siting regions, the properties of the host rocks and the effective containment zones (including sorption measurements), hydrogeological conditions and long-term evolution. The work carried out since submission of the documentation for Stage 1 of the Sectoral Plan and the initiated and planned activities comprise supplementary field investigations (participation in investigations in new boreholes drilled by third parties, seismics, mapping), analyses, laboratory programmes (including investigations on cores from new boreholes) and other studies. In addition to these activities that are focused on geology, investigations will be carried out on a wide range of topics to provide input for the provisional safety analyses, the evaluation of engineering feasibility and the safety-based comparison. In particular, this includes studies on repository layout and gas formation and release.

The state of knowledge will be reviewed once again by the authorities in Stage 2 when Nagra submits the provisional safety analyses and the safety-based comparison of the sites. The review by the authorities will contribute to the decision on the sites for which field investigations will be conducted in Stage 3 with the aim of obtaining more detailed information for identifying the sites for the General Licence Applications for the L/ILW and HLW repositories. If necessary, field work can be carried out at more than two sites for each repository type in Stage 3; this is compatible with the requirements set out in the conceptual part of the Sectoral Plan.

Technical Report NTB 09-08

Physico-Chemical Characterisation Data and Sorption Measurements of Cs, Ni, Eu, Th, U, Cl, I and Se on MX-80 Bentonite

Summary

This report describes the work carried out in LES on MX-80 bentonite in support of Swiss radioactive waste performance assessment studies. With particular regard to Stage 2 of the Sectoral Plan for deep geological disposal, it was considered to be important to bring together in one document the information and results that have accrued over the years from both “in house” studies and associated relevant literature data. The report gives a brief overview of the physico-chemical characteristics and porewater chemistry determined for MX-80 bentonite followed by the results of an extensive experimental sorption programme on the uptake of Cs(I), Ni(II), Eu(III), Th(IV), U(VI), Cl(-I), I(-I) and Se(IV) on the same material. Sorption values are also given for K(I), Ca(II) and Sr(II) which were deduced from porewater chemistry modelling studies.

Technical Report NTB 09-07

Comparison of the reference Opalinus Clay and MX-80 bentonite sorption data bases used in the Entsorgungsnachweis with sorption data bases predicted from sorption measurements on illite and montmorillonite

Summary

In Stage 2 of the Sectoral Plan for 'Deep Geological Disposal' preliminary safety analyses will be carried out for potential sites identified within regions previously selected as being suitable for constructing HLW and L/ILW radioactive waste repositories. The rock formations in question are Opalinus Clay (HLW, L/ILW) and 'Brauner Dogger', Effingen Member and Helvetic Marl (L/ILW). Sorption data bases for all of these host rocks are required to perform the planned preliminary safety analyses.

In a previous report (Bradbury et al. 2010), a methodology was described for developing sorption data bases for argillaceous rocks, so called Generic Rock Sorption Data Bases. In Bradbury et al. (2010) it was argued that the main factor influencing sorption on argillaceous rocks is the phyllosilicate mineral content, particular the 2:1 clay mineral content (illite/smectite/illite-smectite mixed layers). The second important influence is the water chemistry which determines the radionuclide species in the aqueous phase. Primarily sorption measurements on illite were used and these data were converted to the defined conditions in the argillaceous rock by using a series of so called conversion factors to take into account differences in mineralogy, in pH and in radionuclide speciation. Finally, a Lab→Field conversion factor (CF_Lab→Field) was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions. The intention in Stage 2 of the Sectoral Plan is to use this methodology to develop sorption data bases (SDB) for the host rocks under consideration (Opalinus Clay, 'Brauner Dogger', Effingen Member and Helvetic Marl).

Since this methodology for generating SDBs is relatively new, a direct means of creating confidence in its application and verifying its applicability was required. Hence, it was decided to compare and contrast the sorption values obtained in the manner described above with those in an already existing SDB for Opalinus Clay used in the Entsorgungsnachweis (Nagra 2002).

In order to test the procedure further, a second such study was undertaken with MX-80 bentonite. A SDB for MX-80 bentonite was derived from measurements made on montmorillonite using the same methodology and compared with the values used in the Entsorgungsnachweis (Nagra 2002).

The main conclusion from detailed comparisons made for both Opalinus Clay and MX-80 bentonite is that very strong evidence has been provided to demonstrate that the same basic approach as used here can be applied with confidence to other argillaceous rock systems for which direct sorption measurement data may be sparse or missing.

Technical Report NTB 09-06

The Nagra Research, Development and Demonstration (RD&D) Plan for the Disposal of Radioactive Waste in Switzerland

Summary

Nagra's mission is to develop safe geological repositories in Switzerland for all radioactive wastes arising in Switzerland. Two types of repositories are foreseen, one for low and intermediate level waste (L/ILW) and one for spent fuel, vitrified high level waste and longlived ILW (SF/HLW/ILW). Repository implementation involves a stepwise process that takes several decades, thus a comprehensive planning base for the scientific and engineering work is needed, which is presented in the RD&D (research, development and demonstration) plan. The main objective of the RD&D Plan is to establish the purpose, scope, nature and timing of various future RD&D activities, starting from the various requirements and planning assumptions.

Chapter 1 presents the overall objectives of the report and a brief history of the steps leading to the present situation. The planning of work is conditioned by the Federal Government decision in 2006 that Nagra had successfully shown in Project Entsorgungsnachweis (disposal feasibility) that safe disposal of SF/HLW/ILW in Switzerland is technically feasible. Earlier studies and safety authority reviews had already in 1988 lead to the Federal Government decision on the overall feasibility of safe disposal of L/ILW. Following the Project Entsorgungsnachweis decision, the Federal Government initiated the Sectoral Plan for Geological Repositories, which elaborates the siting process. The Sectoral Plan provides a framework within which specific objectives must be met for selecting suitable sites for disposal of both L/ILW and SF/HLW/ILW for which general licence applications are to be made. The overall planning is based on the status as of Nagra’s submission of proposals for Stage 1 of the Sectoral Plan, which should result in the selection of geologically suitable siting regions.

Chapter 2 presents the overall planning premises for implementation of repositories for L/ILW and SF/HLW/ILW including the assumed schedule, the waste types and quantities and the safety strategy for the repositories. The time plan is presented, which includes the Sectoral Plan and general licence procedure, construction and operation of Underground Research Laboratories (URLs) at the sites, construction licence procedure and the operating licence procedure. It is expected that emplacement of L/ILW could begin in about 2035, whereas SF/HLW emplacement would begin about 2050. Waste emplacement is followed by a monitoring period (planning assumption of 50 years), at the end of which an application would be made for closure. The various waste types and quantities produced by nuclear power plants and medicine, industry and research are summarised and the safety concepts for the two repositories are illustrated.

Chapter 3 discusses the RD&D planning process and methodology and the various categories of requirements that dictate the nature and timing of the programme and the planned RD&D. These include legal, regulatory and policy requirements; waste producer requirements; authorities’ recommendations; public expectations; and technology and safety requirements. Together these frame the issues that must be addressed in the programme and establish the nature and timing of the various elements of technical and scientific work.

Chapter 4 summarizes the RD&D issues for SF/HLW identified by Nagra in Project Entsorgungsnachweis and the recommendations made in the formal reviews of the project that were requested by the Federal Government. The broad areas of work required for development of a L/ILW repository are also identified. The status of development of the various categories of technology required for repository implementation is discussed in the context of worldwide progress in the various technology development areas.

Chapter 5 elaborates the strategic requirements for developing the two types of repositories. The definition of the waste types, their properties and the legal and regulatory requirements for disposal set the framework for the timing of implementation as well as for the repository concepts needed to safely dispose of the waste. In the repository concepts long-term safety must be provided by multiple passive safety barriers with a balanced contribution from the engineered and geological systems. The repository design concepts include: i) the main facility, where wastes will be disposed of and which will be backfilled and sealed in due time after waste emplacement; ii) the test zones, where site-specific data for the safety-relevant properties of the host rock are acquired to confirm the safety and technical feasibility; iii) the pilot facility, where the behaviour of waste, backfill material and host rock is monitored until the end of the monitoring period and in which data is collected to confirm safety with a view to closure. Host rocks with favourable properties must be selected within stable large-scale geologic-tectonic situations, which ensure a significant contribution of the geological barrier to the safety functions.

The Sectoral Plan process is discussed in some detail in order to explain the requirements on Nagra’s RD&D programme for each of the stages of site selection up to the general licence application. For Stage 1, the selection of geologically suitable regions, the steps are outlined that led to the proposal of the six geological siting regions for the L/ILW repository (Südranden, Zürcher Weinland, Nördlich Lägeren, Bözberg, Jura-Südfuss, Wellenberg) and the three geological siting regions for the HLW repository (Zürcher Weinland, Nördlich Lägeren, Bözberg). Stage 2 requires the selection of at least two sites for L/ILW and HLW repositories, which is followed by Stage 3, with the selection of one site for each repository, which would provide the basis for the general licence application. The specificity of the requirements of the Sectoral Plan combined with the results of the prior work programme lead to a rather clear definition of the RD&D activities and the main expected reports for each of the stages up to the general licence application.

For the subsequent stages of repository implementation, i.e. construction and operation of a rock laboratory at the chosen site, construction of the repository and operation of the repository until final closure, the broad nature of the work is defined. This permits the level of maturity of the science and technology for each stage to be identified, such that the RD&D studies are appropriately timed and resources are effectively managed.

Chapter 6 gives an overview of the RD&D work to be done in the next 5 to 10 years, i.e. the time frame up to the general licence application, including the objectives, status and principal focus in the various areas, including:

Technical Report NTB 09-05

Critical Review of Welding Technology for Canisters for Disposal of Spent Fuel and High Level Waste

Summary

Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of radioactive waste produced in Switzerland. As part of Nagra's long term disposal strategy, plans must be developed for two repositories, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. Within the next 10 years, Nagra plans to apply for general licences for these repositories. In the application, documentation will be required showing that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel, one of the preferred materials under consideration. The information in this report will be used in conjunction with a material selection report already produced. This report is intended as a preliminary step to provide input to developing design concepts for the SF and HLW canisters.

Objective

The objective of this report was to carry out a critical review of all available welding technologies for the application of sealing carbon steel canisters for SF and HLW to be disposed of in a repository. The review discusses the following key variables:

Work Carried Out

The review of potential welding technologies began with a feasibility review carried out by TWI experts in the relevant processes. Certain feasibility criteria were used to rule out processes clearly not suitable for the application. The next stage was to carry out research in the form of a literature review. This encompassed all remaining processes and was focused on looking for previous applications of the processes for the material and thickness suggested, and also safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was reviewed individually by a TWI engineer with expertise and experience in the process. This information was used at a meeting to weigh up the advantages and disadvantages of each process and decide on which are likely candidate processes. These candidate processes were then reviewed further to include the likely metallurgical effects and potential inspection techniques.

Conclusions

Technical Report NTB 09-04

A Review of the Possible Effects of Hydrogen on Lifetime of Carbon Steel Nuclear Waste Canisters

Summary

In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1'000 years but demonstration of a minimum lifetime of 10'000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl^-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability.

To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the context of the waste canisters it is concluded that the likelihood of cracking due to absorbed hydrogen is remote. Recommendations are given for selection of steel composition, processing and welding procedures to eliminate effectively the probability of cracking for the assumed operational conditions for the disposal canister.

Technical Report NTB 09-03

Sorption Data Bases for Generic Swiss Argillaceous Rock Systems

Summary

In Switzerland the site selection procedure for both HLW and L/ILW repositories is specified by the Swiss Federal Office of Energy in the Sectoral Plan for Deep Geological Repositories. In the forthcoming stage 2 of this plan, potential sites will be identified within regions previously selected based on the presence of suitable host rocks, namely Opalinus Clay, "Brauner Dogger", Effingen Member and Helvetic Marl. Preliminary safety analyses are an integral part of this procedure, and require, amongst other information, the radionuclide sorption properties of the host rock. This report describes a methodology to develop a Generic Rock Sorption Data Base (GR-SDB) for argillaceous rocks. The method will be used to compile specific SDBs for the above mentioned host rocks.

Arguments are presented that the main factor influencing sorption on argillaceous rocks is the phyllosilicate mineral content. These minerals are particularly effective at binding metals to their surfaces by cation exchange and surface complexation. Generally, the magnitude of sorption is directly correlated with the phyllosilicate content (2:1 type clays: illite/smectite/illite-smectite mixed layers), and this parameter best reflects the sorption potential of a given mineral assembly. Consequently, sorption measurements on illite were preferably used as source data for the GR-SDB.

The second component influencing radionuclide sorption is the porewater chemistry. In the present report, generic water compositions were extracted from the analytical ranges of deep ground waters in various sedimentary formations in Switzerland. In order to cover the range of ionic strength (I) and pH values of Swiss ground waters in argillaceous rocks, five types of generic water compositions were defined, combining low, intermediate and high values of ionic strength and pH.

The GR-SDB for in situ conditions was derived using conversion factors (CF). As the name implies, these factors were used to convert the (predominantly) illite sorption values into sorption values valid for the defined generic conditions with regard to mineralogy and porewater composition. Conversion factors were used to adapt sorption values to mineralogy (CF_min), to pH value (CF_pH) and to radionuclide speciation (CF_spec). Finally, a Lab→Field conversion factor (CF_Lab→Field) was applied to adapt sorption data measured in dispersed systems (batch experiments) to intact rock under in-situ conditions.

Calcareous rock is used in safety analyses as being representative of a clay rock which has lost most of its favorable sorption properties due to near-field effects such as alteration by an alkaline plume and subsequent processes. It is assumed that calcareous rocks do not contain any significant quantities of phyllosilicates and that only uptake data on calcite are relevant. Sorption data on calcite are extremely sparse and the uptake mechanisms are not fully understood. However, when the existing sorption data (log R_d values) are plotted against the ionic radii of the respective metals, an acceptable linear correlation between these two quantities is found. This so-called linear free energy relationship is used to complement the sparse experimental data in the SDB for calcareous systems.

Technical Report NTB 09-02

A Review of Materials and Corrosion Issues Regarding Canisters for Disposal of Spent Fuel and High-level Waste in Opalinus Clay

Summary

The project "Entsorgungsnachweis" presented by Nagra to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. Nagra proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that Nagra had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated.

To deal with these concerns Nagra convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to Nagra the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by Nagra on the long term performance and safety of SF/HLW canisters in the Swiss repository concept.

For the assessment of the corrosion behavior of canisters, the CMRB distinguished four phases during which the corrosive environment is expected to gradually change from aerobic dry to anoxic wet conditions. Possible damage mechanisms of steel were identified for each phase and critically examined, including effects due to radiation, solid reaction products, microbial activity and the occurrence of stress assisted failures. The expected performance of other canister materials was also considered. The CMRB concludes that Nagra presents a convincing case that using steel canisters surrounded by bentonite as part of a multi-barrier system using Opalinus clay as the geological barrier is a viable concept for the safe disposal of SF/HLW under the assumption that the maximum acceptable hydrogen production rates given by Nagra can be confirmed in future. A few issues related to the long term performance of steel canisters need to be further elaborated and clarified by Nagra, but the CMRB found no major issue that would invalidate the use of steel canisters as part of the Nagra multi-barrier concept. The CMRB deems that the research program pursued by Nagra is carefully managed, effective and credible. Within the planning horizon for implementation of a repository for SF/HLW in Switzerland, the time table for canister development presented by Nagra is realistic. While vigorously pursuing the evaluation of the evolution of the near field environment and its effect on the corrosion of steel, Nagra should from now on initiate a comprehensive program on the evaluation of technological solutions for fabrication, welding, surface finishing and stress mitigation of thickwalled steel canisters.

Technical Report NTB 09-01

Topics and processes dealt with in the IP FUNMIG and their treatment in the Safety Case of geologic repositories for radioactive waste

Summary

The scope of the FUNMIG Integrated Project (IP) was to improve the knowledge base on biogeochemical processes in the geosphere which are relevant for the safety of radioactive waste repositories. An important part of this project involved the interaction between data producers (research) and data users (European radioactive waste management organisations). The aim thereof was to foster the benefits of the research work for performance assessment (PA), and in a broader sense, for the safety case of radioactive waste repositories. For this purpose, an adapted procedure was elaborated. Thus, relevant features, events and processes (FEPs) for the three host rock types clay, crystalline and salt were taken from internationally accepted catalogues and mapped on each of the 108 research tasks by a standardised procedure. The main outcome thereof was a host-rock specific tool (Task Evaluation Table) in which the relevance and benefits of the research results were evaluated both from the PA and research perspective. Virtually all generated data within FUNMIG are related to the safety-relevant FEP groups "transport mechanisms" and "retardation".

In a general sense, much of the work within FUNMIG helped to support and to increase confidence in the simplified PA transport and retardation models used for calculating radionuclide (RN) transport through the host rock. Some of the studies on retardation processes (e.g. coupled sorption-redox processes at the mineral-water interface) yielded valuable data for all three rock types dealt within the IP. However, most of the studies provided improved insight to host-rock specific features and processes, whereby the majority of this work was dedicated to clay-rich and crystalline host rocks. For both host rock types, FUNMIG has significantly contributed to improving understanding on a conceptual level, both by providing new experimental data at different spatial scales and by developing new modelling approaches.

Selected highlights with regard to FUNMIG's achievements include: For argillaceous host rocks, the systematic effort of investigating and comparing diffusion and sorption processes at different scales in different clay rocks with a variety of methods has substantially increased the knowledge basis for future safety cases. For crystalline host rocks, valuable data on the generation, transport and filtration of clay colloids from the nearfield and their impact on radionuclide transport under realistic conditions have been obtained. The results from studies on organic colloids and on biofilms including their interaction with radionuclides have been shown to be of interest for future safety cases of salt-host rocks. Among the main research issues from a PA perspective to be addressed in the future, we note the following: (i) the question of irreversibility of radionuclide sorption to colloids in crystalline fractures, (ii) a comprehensive model for cation and anion diffusion in clays for different scales and (iii) the applicability of mechanistic retardation models for strongly sorbing radionuclides in intact clay and crystalline host rocks.

An important lesson learnt from the interaction between research and PA is that it would be desirable to conduct an analogue evaluation procedure for the proposed task before the start of the research work. In this regard, the procedures developed within FUNMIG are a useful tool for planning future Integrated Projects.

Technical Report NTB 08-12

Corrosion of carbon steel under anaerobic conditions in a repository for SF and HLW in Opalinus Clay

Summary

Nagra is considering carbon steel as one of the canister material options for the disposal of highlevel waste and spent fuel in a deep geological repository in Opalinus Clay. Following a brief period of aerobic conditions, the canister will be exposed to an anaerobic environment for much of its service life. Knowledge of the rate of anaerobic corrosion is important not only for estimating the canister lifetime but also for determining the rate of hydrogen generation.

This report describes a critical review of the anaerobic corrosion behaviour of carbon steel under environmental conditions similar to those expected in the repository. The aims of the report are:

The review is based on selected studies from various national nuclear waste management programs, supplemented where appropriate with studies from other applications and with evidence from archaeological analogues.

The corrosion rate of carbon steel decreases with time because of the formation of a protective surface film. There are differences in behaviour in bulk solution and in the presence of compacted bentonite. In bulk solution, the corrosion rate decreases to an apparent steady-state rate after a period of approximately six months, with a long-term rate of the order of 0.1 μm⋅yr^-1. The surface film comprises a duplex structure, with a magnetite outer layer and a spinel-type inner layer. In compacted clay systems the rate of decrease in corrosion rate is slower, with steadystate not being reached after several years of exposure. There is a significant body of evidence from apparently well-conducted experiments that indicate an anaerobic corrosion rate of the order of 1 – 2 μm⋅yr^-1 in systems containing compacted clay and the protective films tend to be carbonate-based rather than magnetite-based.

There is no evidence in the literature that the use of a constant long-term corrosion rate for safety assessment purposes is not justified. Factors that are important in determining the structure and properties of the corrosion product film are reviewed, including the effects of the aerobicanaerobic transition on the film composition and structure, possible spalling of protective films, and the effect of the accumulation of corrosion products on the corrosion rate of the underlying steel.

Technical Report NTB 08-10

Chemical reactivity of alpha-isosaccharinic acid in heterogeneous alkaline systems

Summary

Cellulose degradation under alkaline conditions is of relevance for the mobility of many radionuclides in the near-field of a cementitious repository for radioactive waste, because metal-binding degradation products may be formed. Among these α-isosaccharinic acid (α-ISA) is the strongest complexant. The prediction of the equilibrium concentration of α-ISA in cement pore water is therefore an important step in the assessment of the influence of cellulose degradation products on the speciation of radionuclides in such environments.

The present report focuses on possible chemical transformation reactions of α-ISA in heterogeneous alkaline model systems containing either Ca(OH)₂ or crushed hardened cement paste. The transformation reactions were monitored by measuring the concentration of α-ISA by high performance anion exchange chromatography and the formation of reaction products by high performance ion exclusion chromatography. The overall loss of organic species from solution was monitored by measuring the concentration of non-purgeable organic carbon. The reactions were examined in diluted and compacted suspensions, either at 25 °C or 90 °C, and under anaerobic atmospheres obtained by various methods. It was found that α-ISA was transformed under all conditions tested to some extent. Reaction products, such as glycolate, formate, lactate and acetate, all compounds with less complexing strength than α-ISA, were detected. The amount of reaction products identified by the chromatographic technique applied was ~50 % of the amount of α-ISA reacted. Sorption of α-ISA to Ca(OH)₂ contributed only to a minor extent to the loss of α-ISA from the solution phase.

As the most important conclusion of the present work it was demonstrated that the presence of oxidising agents had a distinctive influence on the turnover of α-ISA. Under aerobic conditions α-ISA was quantitatively converted to reaction products, whereas under strict anaerobic conditions, only small amounts of α-ISA were transformed. It can be hypothesised that, under these conditions, either traces of oxygen remaining bound to Ca(OH)₂ or unidentified impurities in Ca(OH)₂ were responsible for the reactions observed. The involvement of microbially mediated processes can be excluded, because the reactions proceeded in a similar qualitative manner, however faster, at 90 °C than at room temperature.

The possible chemical degradation of α-ISA to organic compounds with less complexation capabilities under anaerobic repository conditions is therefore not supported by the experimental findings of the present study.

Technical Report NTB 08-07

Effects of post-disposal gas generation in a repository for low- and intermediate-level waste sited in the Opalinus Clay of Northern Switzerland

Summary

Within the framework of Stage 1 of the "Sectoral Plan for Deep Geological Repositories" Nagra has proposed Opalinus Clay as a possible host rock for a repository for low- and intermediate-level waste (L/ILW). Opalinus Clay is characterised by a low permeability and is, therefore, an excellent barrier against radionuclide transport. Because significant amounts of gas are generated in a repository for L/ILW a demonstration is required that despite the low gas permeability of the Opalinus Clay the gas can escape without compromising long-term safety. The present study provides a comprehensive assessment of the question how gas generation and transport in a L/ILW repository affects system behaviour. For the purpose of the present study a geological repository for L/ILW in the Opalinus Clay of Northern Switzerland with a depth of about 300 – 400 m below the surface is assumed. The report provides relevant information regarding the layout and the operation of the L/ILW repository as well as a brief survey of the waste inventories and the expected amounts of gas generated. Furthermore the state of geoscientific understanding of gas transport processes in the underground structures of the repository and in the surrounding host rock is presented and the impact of gas generation on the isolation capacity of the repository is considered. The modelling activities described in the present report started in 2005 and were completed by the end of 2007. The results of the model calculations were used to optimise the layout of the L/ILW repository with respect to the effects of gas generation and transport. Specifically a design option was studied in which, by an appropriate choice of backfill and sealing materials, the gas can escape along the access ramp into the overlying rock formations without creating undue gas overpressures.  

The estimates of the gas generation rates for the L/ILW repository are based on a waste inventory accounting for the existing nuclear power plants, with an assumed operation period of 50 years, and for wastes from medicine, industry and research with a collection period up to the year 2050. This inventory includes a total mass of approximately 40'000 tons of steel and other metals and about 2'200 tons of organic matter. Complete corrosion / degradation of all gas-generating materials yields a gas volume of approximately 20 to 30 million cubic meters (STP). The highest gas generation rates are expected in the early post-closure period up to several hundreds of years, followed by a steady decline. The expected total duration of the gas generation phase is in the order of 200'000 years. 

The total volume of voids in the backfilled repository is in the order of 58'000 m³ for the assumed waste inventory. If the total amount of corrosion and degradation gases were enclosed hermetically in this void volume, a high gas pressure would result. In the real system, however, at least a part of the gas will be released through the host rock, resulting in much lower pressures. In order to keep the gas pressure low even in the case of a very low host rock permeability and / or an increased gas production, specially designed backfill and sealing materials could be used such as high porosity mortars as backfill materials for the emplacement caverns and sand/bentonite mixtures with a bentonite content of 20 – 30 % for backfilling other underground structures and for the seals ("engineered gas transport system" – EGTS). The EGTS is aimed at increasing the gas transport capacity of the backfilled underground structures without compromising the radionuclide retention capacity of the engineered barrier system. Sand/bentonite mixtures with a low bentonite content exhibit a low permeability for water and a relatively high permeability for gas due to their (micro)structure. 

The development of gas overpressures in the backfilled emplacement caverns is unavoidable due to the large amount of corrosion and degradation gases. Numerical simulations show that, for the expected gas generation rate, the planned repository layout and a typical gas permeability of the host rock, the gas pressure in the emplacement caverns remains below the threshold pressure for the onset of pathway dilation (approximately 6.5 MPa for the assumed site conditions). For such conditions, no additional design measures are needed to mitigate gas impacts. For the case of conservative gas generation rates, or the case of a very low gas permeability of the rock (k_OPA ≤ 10^-21 m²), the gas pressure could rise above the critical threshold pressure for the onset of pathway dilation. Consequently, the use of appropriate backfill and sealing materials that ensure a release of a part of the gas along the access ramp would be a suitable design measure to limit gas pressure. Calculations indicate that such an approach could limit pressures in the emplacement caverns so that even in the case of a very low permeability host rock overpressures above hydrostatic pressure would remain within a range of 3 – 4 MPa. 

As a result of the elevated gas pressures in the emplacement caverns, pore water containing dissolved radionuclides will be displaced into the geosphere. The gas pressure build-up as an additional driving force for mass transport also tends to increase the path length for radionuclide transport in the host rock, an effect which is further enhanced by the anisotropy of the intrinsic rock permeability. The displaced water is widely spread over the footprint area of the repository towards the adjacent rock formations above and below the host rock. The numerical simulations indicate specific water fluxes in the host rock of up to 10^-11 m/s in the very early gas generation phase (< 1'000 years after repository closure). The fluxes decline steadily with time until the regime of diffusion-dominated transport is reached in the late times of the gas production phase (specific water flux typically < 10^-13 m/s after several 10'000s of years). A comparison of these results with those from safety calculations using a wide range of specific water fluxes leads to the conclusion that pore water displacement caused by elevated gas pressures will not compromise the long-term safety of a L/ILW repository in Opalinus Clay. 

Technischer Bericht NTB 08-06

Modellhaftes Inventar für radioaktive Materialien MIRAM 08

Summary

Based on the description of the "Model Inventory for Radioactive Materials MIRAM 08” (status 31.12.2007), this report documents the inventory of already existing and future arisings of radioactive wastes and materials in Switzerland. 

The objective of MIRAM 08 is to quantify and comprehensively characterise all waste that already exists and will arise in the future. This will allow well founded data to be provided as input to safety analyses and planning of the facilities and operation of deep geological repositories. 

The report comprises: 

Technischer Bericht NTB 08-05

Vorschlag geologischer Standortgebiete für das SMA- und das HAA-Lager.
Begründung der Abfallzuteilung, der Barrierensysteme und der Anforderungen an die Geologie.
Bericht zur Sicherheit und technischen Machbarkeit

Summary

According to the conceptual part of the Sectoral Plan for Deep Geological Repositories (BFE 2008), the first step in the site selection process requires the waste producers to submit proposals for geological siting regions for repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). The proposals prepared by Nagra on behalf of the waste producers for stage 1 of the Sectoral Plan procedure are justified and documented in Nagra (2008b).

The Sectoral Plan states that the siting proposals have to be prepared in five steps. In the first step, the waste inventory, including reserves for future developments in the nuclear power programme, is allocated to the L/ILW and HLW repositories. Based on this allocation, the barrier and safety concepts for the two repositories are defined in the second step. Quantitative and qualitative requirements on the geology are then derived with a view to evaluating the geological siting possibilities. This applies to the time period under consideration, the space required by the repository, the properties of the host rock (depth, thickness, lateral extent, hydraulic conductivity), the long-term stability of the geological situation, the reliability of geological information and engineering suitability.

Steps 3 to 5 cover the evaluation of the geological siting possibilities. In the third step, the large-scale geological-tectonic situation is assessed and large-scale areas for further consideration are identified. The fourth step involves selecting the preferred host rock formations within these large-scale areas. In the fifth step the rock configurations, i.e. the spatial arrangement of the preferred host rocks within the large-scale areas under consideration, are evaluated and geological siting regions are identified.

This report complements Nagra’s report documenting the proposals for geological siting regions (Nagra 2008b). It explains the allocation of the waste to the L/ILW and HLW repositories, defines the barrier and safety concepts and justifies the repository-specific requirements on the geology (steps 1 and 2). These requirements are used in the stepwise narrowing-down procedure to select the geological siting regions. The report does not take the form of a conventional safety analysis in that, in particular, it does not contain a safety assessment of concrete geological repositories at specific sites in specific host rocks. Instead, the fundamental requirements are derived using guiding safety considerations and experience. Site-specific safety analyses will be part of the subsequent stages 2 and 3 of the Sectoral Plan process.

The key results are presented in the following.

Waste allocation 

According to the Federal Office of Energy (BFE 2008), the waste properties to be considered for waste allocation, repository design and identification of geological siting regions are inventory, half-lives and the activity and radiotoxicity of the safety-relevant radionuclides and their evolution with time. Waste volumes, material properties and their potential influences on the host rock, heat production, content of potentially gas-producing components (metals, organics) and content of complexants also have to be considered. The description of the waste properties forms the starting-point for the waste allocation. The waste is divided into the categories high-level waste (HLW), alpha-toxic waste (ATW) and low- and intermediate-level waste (L/ILW), as specified in the Nuclear Energy Ordinance (2004). The description shows clearly that, with respect to all properties, HLW differs significantly from ATW and L/ILW. For this reason, as reflected in disposal concepts to date, HLW is disposed of in a separate repository with a specifically designed barrier system.

The ATW and L/ILW differ in terms of specific radiotoxicity, specific activity and specific heat production, in terms of both absolute values and evolution with time. However, many of their other properties are very similar, particularly the material inventory. 

In principle, a combined repository for all ATW and L/ILW would be conceivable. Experience shows that such a facility constructed in a suitable host rock in a favourable geological setting has the potential to fulfil the safety requirements specified by the authorities. On the other hand, experience also shows that calculated doses are dominated by just a few of the ATW and L/ILW waste types. If these dominant waste types could be disposed of elsewhere, the requirements on the geology could be reduced while the level of safety would remain the same; this has the effect of increasing the possibility of finding suitable siting regions. For these reasons, the existing concept (a HLW repository with a facility for long-lived intermediate-level waste (ILW) and a L/ILW repository) has been maintained, with the aim of allocating the dosedominating ATW and L/ILW to the ILW facility. This allows to reduce the safety-related requirements on the geology for the L/ILW repository. Based mainly on generic dose calculations, the ATW and L/ILW are allocated to the two repository types as they differ only slightly with respect to material inventory and gas generation rates and both the HLW repository (including the ILW facility) and the L/ILW repository are designed in such a way that repository-induced influences (originating from the waste) do not significantly affect longterm safety.

The proposal made by Nagra includes two variants, characterised by minimum requirements on the large-scale hydraulic conductivity of the host rock for the L/ILW repository of 10^-10 m/s and 10^-9 m/s respectively. As expected, the volume of waste allocated to the L/ILW repository is somewhat smaller for the 10^-9 m/s variant than for the 10^-10 m/s variant. Based on an assessment of the geological possibilities within Switzerland, which shows that there are sufficient suitable host rocks and effective containment rock zones with a large-scale hydraulic conductivity of 10^-10 m/s or better, this allocation variant is termed the reference allocation and the variant with 10^-9 m/s is termed the alternative allocation. In both cases, all the ATW is allocated to the HLW repository (ILW facility). For the reference allocation, somewhat less than 1% of the volume of L/ILW is also allocated to the HLW repository (ILW facility); in the case of the alternative allocation it is somewhat less than 10%.

Barrier and safety concept

The barrier concept describes the functions of the different engineered and geological barriers of the deep repository, based on a system of staged, passive safety barriers consisting of the waste matrix, disposal container, backfilling of the underground disposal chambers, backfilling and sealing of the underground structures, the host rock and any confining rock units and the overall geological situation.

The safety concept shows how the different engineered and geological barriers contribute to system safety and what safety functions they perform. The safety functions provide the physical separation of the waste from the human environment and ensure the required long-term stability, containment of the radionuclides, delayed release of nuclides and nuclide retention in the near-field and the geosphere, thus ensuring low release rates.

In the selected safety concept, both the engineered and the geological barriers (host rock, any confining units and their geological situation) contribute significantly to the barrier function of the overall system. In line with the requirements set out by the authorities, a system in which the containment and retention of radionuclides rely on the engineered barriers alone will not come into consideration.

The concept also describes the contribution to safety of the different components of the barrier system. For both repository types, by far the largest proportion of radiotoxicity will decay within the engineered barriers due to immobilisation of the radionuclides and radioactive decay. There is a further decrease due to decay during transport through the host rock, meaning that the proportion of radionuclides leaving the engineered and natural barriers represents only a tiny fraction of the original radiotoxicity. This means that the resulting doses are well below the protection objective specified by the regulatory authority.

Requirements on the geology

Specifying the requirements on the geology is carried out in two steps. In the first step, indicators are defined that adequately encompass the criteria set out in the Sectoral Plan and are used to identify geological siting regions. In the second step, the requirements and evaluation scales for the indicators are defined.

The starting-point for defining the indicators are the safety functions discussed above together with a set of overarching principles relating to reliable implementation of the geological repositories and the reliability of geological information. A set of indicators with associated requirements and evaluation scales are then defined, the application of which in the narrowing down process results in geological siting regions where repositories can be constructed that fulfil the safety functions and principles and ensure sufficient safety.

The requirements and evaluation scales for the indicators are defined using radionuclide release calculations, model calculations of the behaviour of individual barriers or properties, measured and empirical values and qualitative information.

Based on generic safety considerations and experience derived from earlier system and safety analyses, the following features are considered to be particularly important for the site evaluation:

Technischer Bericht NTB 08-04

Vorschlag geologischer Standortgebiete für das SMA- und das HAA-Lager. Geologische Grundlagen

Summary

Beilagen als einzelne PDFs 

On 2nd April 2008, the Federal Council approved the conceptual part of the Sectoral Plan for Deep Geological Repositories. The Plan, which was prepared by the Federal Office of Energy (SFOE), sets out the details of the site selection procedure for geological repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW).

The Sectoral Plan specifies that selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages. Stage 1 ends with the definition of geological siting regions within which the repository projects will be elaborated in more detail in the later stages of the Sectoral Plan (stages 2 and 3). The final step will be the identification of sites for construction of the repositories and granting of the required general licences. As a first step, the Sectoral Plan calls for the waste producers to submit proposals for geological siting regions based on criteria and requirements set out in the Sectoral Plan.

The proposals for geological siting regions are documented in a report (Nagra 2008a) prepared by Nagra on behalf of the waste producers for stage 1 of the Sectoral Plan process. The report sets out the procedures and arguments used in identifying the geological siting regions for the L/ILW and HLW repositories proposed in stage 1 and follows the criteria and requirements set out in the Sectoral Plan. It is supplemented inter alia by the current report, which presents and illustrates the geoscientific background leading to the results of the site selection process in greater detail than in Nagra (2008a).

The geoscientific background is based on the one hand on an evaluation of the geological investigations previously carried out by Nagra on deep geological disposal of HLW and L/ILW in Switzerland (investigation programmes in the crystalline basement and Opalinus Clay in Northern Switzerland, investigations of L/ILW sites in the Alps, research in rock laboratories in crystalline rock and clay); on the other hand, new geoscientific studies have also been carried out in connection with the site selection process. These include:

The results of the geological studies that are relevant for stage 1 of the Sectoral Plan are documented in around 50 Nagra reference reports, several publications in technical journals and in numerous reports on rock laboratory projects (particularly Mont Terri).
Formulation of the siting proposals in accordance with the Sectoral Plan is conducted in five steps:

The first two steps are documented in detail in a separate report (Nagra 2008b). Steps 3 to 5 cover the evaluation of the geological siting possibilities, for which the geological knowledge base and key geoscientific aspects are discussed in the present report:

The procedure leads to the following geological siting regions:

For the L/ILW repository:

For the HLW repository:

In three of the geological siting regions (Zürcher Weinland (Cantons Zürich, Thurgau), North of Lägeren (Cantons Zürich, Aargau) and Bözberg (Canton Aargau)), the possibility exists in principle of siting the L/ILW and HLW repositories together as a so-called 'combined repository'.

The evaluation conducted in accordance with the conceptual part of the Sectoral Plan gives the following results: For the L/ILW repository, the geological siting regions Southern Schaffhausen, Zürcher Weinland and Bözberg are considered very suitable and North of Lägeren, Jura-Südfuss and Wellenberg are considered suitable. For the HLW repository, the siting regions Zürcher Weinland and Bözberg are considered very suitable and North of Lägeren is considered suitable.

The work done by Nagra to narrow down and evaluate the siting options from the point of view of geology and safety will be supplemented by a survey of land use planning performed by the authorities. The authorities and the Federal Council may choose to include additional non-technical-scientific criteria in the decision-making process.

The evaluation and assessment of the geological siting possibilities discussed in this report use all available geological information that is relevant in the context of deep geological disposal. The most recent scientific literature has been included and the geological database has been expanded (e.g. through participating in investigations being carried out by third parties or by acquiring third-party data). Where meaningful, data have also been re-analysed (e.g. seismic measurements, hydraulic tests). The technical-scientific knowledge base varies from region to region, but still allows priorities for the next steps to be set with confidence and well founded proposals to be made for geological siting regions to be carried through to the next stages of the Sectoral Plan.

The siting proposals will be evaluated by the federal authorities and, following a public hearing phase, the decision of the Federal Council on the geological siting regions is expected in around 2½ years. This will be followed by stage 2 (identification of at least 2 sites each for L/ILW and HLW within the geological siting regions defined in stage 1) and stage 3 with the general licence procedure. The siting decision for the geological repositories for L/ILW and HLW as part of the general licence is expected in around 10 years. The general licence is granted by the Federal Council but must be approved by Parliament and is subject to an optional national referendum.

Beilagen:

 

Technischer Bericht NTB 08-03

Vorschlag geologischer Standortgebiete für das SMA- und das HAA-Lager.
Darlegung der Anforderungen, des Vorgehens und der Ergebnisse

Summary

Important steps in the process of managing radioactive waste have already been implemented in Switzerland. These include the handling and packaging of the waste, waste characterisation and documentation of waste inventories and interim storage along with associated transport. In terms of preparing for deep geological disposal, the necessary scientific and technical work is well advanced and the feasibility of constructing geological repositories that provide the required long-term safety has been successfully demonstrated for all waste types arising in Switzerland; these feasibility demonstrations have been approved by the Federal Council. Sufficient knowledge is available to allow the next steps in the selection of repository sites to be defined. The legal framework is also in place and organisational measures have been provided that will allow the tasks to be performed in the coming years to be implemented efficiently. The conceptual part of the Sectoral Plan for Deep Geological Repositories that was approved by the Federal Council on 2nd April 2008 plays a major role, as it regulates the details of the site selection process to be conducted over the next years. The Sectoral Plan specifies that selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages. Stage 1 ends with the definition of geological siting regions within which the repository projects will be elaborated in more detail in the later stages of the Sectoral Plan (stages 2 and 3). The final step will be the identification of sites for construction of the repositories and granting of the required general licences.

As a first step, the Sectoral Plan calls for the waste producers to submit proposals for geological siting regions for the repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). This report documents and justifies the siting proposals prepared by Nagra on behalf of the waste producers for stage 1 of the Sectoral Plan process. 

Formulation of these proposals in accordance with the Sectoral Plan is conducted in five steps: 

Technischer Bericht NTB 08-02

Bericht zum Umgang mit den Empfehlungen in den Gutachten und Stellungnahmen zum Entsorgungsnachweis

Summary

In the process of evaluating the Opalinus Clay project demonstrating the feasibility of disposing of spent fuel, vitrified high-level waste and long-lived intermediate-level waste in Switzerland, the authorities and their experts made numerous recommendations regarding future procedures and activities to be implemented in the disposal programme. Nagra analysed these reviews and expert opinions and took the recommendations into consideration when preparing its future work programme. In 2006, the Federal Council decided that the feasibility of disposing of these waste categories had been demonstrated successfully, but called on the waste producers to prepare a report that systematically addresses the open questions and recommendations and shows how these will be dealt with in a timely and technically appropriate manner. In the present report, Nagra fulfils this requirement on behalf of the waste producers. The report sets out the recommendations made by the authorities and explains how they will be handled. In many cases, the work required has already begun; in other cases plans are already in place. To facilitate the discussion in the report, the recommendations and the responses of Nagra in each case are divided into topical areas. The main part of the report provides a summary discussion, which also addresses time-related aspects of implementing the recommendations. An appendix to the report provides a detailed overview, in the form of a table, of all the recommendations and the associated responses of Nagra.

Technischer Bericht NTB 08-01

Entsorgungsprogramm 2008 der Entsorgungspflichtigen

Summary

Important steps in the management of radioactive waste have already been implemented in Switzerland and there is now wide experience in carrying out the associated activities. These include the handling and packaging of waste, waste characterisation and compiling of inventories and interim storage and the associated waste transport. In terms of preparing for deep geological disposal, the necessary scientific and technical work is well advanced and the feasibility of implementing geological repositories that provide the required long-term safety has been successfully demonstrated for all wastes arising in Switzerland; these feasibility demonstrations have also been approved by the Federal Council. Sufficient knowledge is available to allow the next steps in the selection of repository sites to be performed. The legal framework and organisational measures are also in place that will allow the prescribed steps to be performed in the coming years to be implemented efficiently. The conceptual part of the Sectoral Plan for Deep Geological Repositories that was approved by the Federal Council on 2nd April 2008 plays a major role, as it regulates the details of the site selection process to be conducted over the next years.

This report documents the waste management programme prepared by the waste producers, as required by the legislation (Nuclear Energy Act (KEG 2003), Art. 32 and Nuclear Energy Ordinance (KEV 2004), Art. 52). The report was prepared by Nagra on behalf of the waste producers and covers all aspects as required by law. The following areas are addressed: 

• Origin, type and volumes of radioactive waste: The origin, types and volumes of radioactive waste to be disposed of in Switzerland are known. The reference case for the waste management programme assumes operation of the existing nuclear power plants for a period of 50 years and a collection period up to around 2050 for radioactive waste from medicine, industry and research (the time by which, in the reference case, emplacement in the repository of the low- and intermediate-level waste from the power plants will have been completed). Sufficient flexibility in dealing with future developments has to be retained. For this reason, the types and volumes of radioactive waste that would arise in the case of extension of the operating lifetime of the existing power plants and the collection period for waste from medicine, industry and research by 10 years are included for planning purposes. Also considered are the wastes to be expected in the case of an additional production of 5 GW_e for a period of 60 years by way of replacing the existing power plants and the stepwise expiry of the electricity supply contracts with France, while assuming a moderate increase in electricity consumption.
  
  The resulting wastes are conditioned, characterised and inventoried on an ongoing basis. Before conditioning of a waste stream begins, the proposed conditioning procedure is evaluated by Nagra in terms of the suitability for disposal of the resulting waste packages. This is a prerequisite for clearance by the authorities of routine conditioning procedures. Conditioned waste will also be evaluated when preparing the safety reports in support of the programme milestones and it is possible that some conditioning procedures will be modified in the light of new understanding. Besides information on waste that already exists, a model inventory of waste that will arise in the future has also been compiled. This provides a reliable basis for planning and implementing geological repositories and managing available interim storage capacity.

• Geological repositories, including design concepts: The Swiss waste management concept assumes two deep geological repositories, one for low- and intermediate-level waste (L/ILW repository) and one for spent fuel, vitrified high-level waste from reprocessing and long-lived intermediate-level waste (HLW repository). These two repositories could be implemented at different sites or at the same site if the geological situation is suitable. For the latter possibility of a combined repository, the disposal chambers of the two facilities could be constructed in the same or in different geological formations. Against the background of the legal and regulatory framework, the conceptual requirements and assumptions to be taken into consideration for the different repositories are defined and corresponding projects have been developed. The proposed design concepts are based on the requirement in the nuclear energy legislation that long-term safety is to be assured by a system of multiple passive safety barriers. For repository implementation, a range of design alternatives exist for individual components that allow consideration of the site-specific situation. In order to ensure optimum configuration of the repository installations, there must be sufficient flexibility to allow information and experience arising in the future (results of site explorations, improvement of knowledge through research and development) to be taken into account. Planning must also account for waste arising as a result of future developments in the areas of nuclear energy and application of radioactive materials in medicine, industry and research. With this in mind, the possibility of increasing the disposal capacity of the repositories has to be considered when planning the facilities.

Technical Report NTB 07-01

Grimsel Test Site
Investigation Phase IV
Borehole Sealing

Summary

Within the context of the Phase IV (1994 - 1996) research and development activities at the Grimsel Test Site (GTS), Nagra developed, in collaboration with the Agence nationale pour la gestion des déchets radioactifs (Andra), an investigation project for the sealing of boreholes drilled from underground. The project had the following goals: 

Technical Report NTB 05-03

Grimsel Test Site
Investigation Phase VI
Pore Space Geometry Project
Characterisation of Pore Space Geometry by 14C-MMA Impregnation

Summary

In Finland high-level radioactive waste is planned to be disposed of in a deep geological repository within a crystalline host rock. The potential role of the geosphere as a safety barrier in repository performance assessment is well established. However uncertainties in both transport pathway definition and pore space characterisation of crystalline rock still exist and the repository safety evaluation today requires going from laboratory and surface-based field work to the underground repository level. Little is known about the changes to rock transport properties during sampling and decompression. Recent investigations using resin impregnation of the rock matrix at the Grimsel Test Site imply that non-conservative errors in calculated transport properties derived from laboratory data may reach factors of two to three.

Due to the potentially great significance of pore space characterisation to safety analysis calculations, it was decided to study the rock matrix characteristics in situ using methylmethacrylate (MMA) resin labelled with ¹⁴C. During the last decade, the poly-methylmethacrylate (PMMA) method has been developed for characterising the porosity of low permeable granitic rocks in the laboratory. Impregnation with ¹⁴C-labelled methylmethacrylate (¹⁴C-MMA) and autoradiography allows investigation of the spatial distribution of accessible porosity at the centimetre scale. Quantitative measurements of total or mineral-specific local porosities have been obtained using image analysis tools. Electron microscopy examinations and mercury porosimetry measurements have provided detailed information on pore and fissure apertures.

The objective of this work was to develop an in situ application of the PMMA impregnation technique. The changes in rock porosity due to stress relaxation when overcoring the samples from the bedrock for the laboratory studies were examined. The concept behind the work was to inject ¹⁴C-MMA from a central borehole at a depth of around 1 metre from the tunnel wall within a 20 cm injection interval. It was assumed that six additional smaller diameter radial boreholes would permit increased drying of the rock matrix and that intrusion of MMA would be observed from the observation boreholes during the injection of the resin. The two main differences between in situ and laboratory PMMA impregnations were: 1) drying in situ was done by ventilating warm air whereas in the laboratory, samples were dried by heating in a vacuum; 2) polymerisation in situ was through auto polymerisation whereas in the laboratory polymerisation was achieved by irradiation of the samples.

It was found that air ventilation drying around the injection borehole was not effective enough to dry the rock matrix when ¹⁴C-MMA impregnation was applied in situ. However the penetration of ¹⁴C-MMA into Grimsel granodiorite in situ was successful. Autopolymerisation of the resin reduced the amount of impregnation, but the thermal polymerisation succeeded well. The MMA intruded to depths of 2-5 cm from the injection borehole; with maximum penetration along the foliation of the rock texture. The amount of PMMA showed clearly a decreasing trend from injection borehole surface to a depth of 5 cm in the rock matrix.

Technischer Bericht NTB 05-02

«Geologische Tiefenlagerung der abgebrannten Brennelemente, der hochaktiven und langlebigen mittelaktiven Abfälle
Darstellung und Beurteilung der aus sicherheitstechnisch- geologischer Sicht möglichen Wirtgesteine und Gebiete

Summary

In order to demonstrate the basic feasibility of the safe disposal of spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (ILW) in a deep geological repository, Nagra has submitted Project Entsorgungsnachweis (the term translates into English as “demonstration of disposal feasibility”) at the end of the year 2002. This feasibility study is based on the Opalinus Clay host rock option in the Zürcher Weinland area in northern Switzerland. The choice of both Opalinus Clay and the Zürcher Weinland is the result of a procedure focussed on safety, in which other possible options were also evaluated but set aside by Nagra in a stepwise manner. The Swiss safety authorities have critically observed all phases of the evaluation procedure, provided comments to Nagra’s corresponding reports and have given their approval to important decisions.

The present report was put together by Nagra in response to a request by the Federal Department of Environment, Transport, Energy and Communications as a part of the decision basis for the further steps after Project Entsorgungsnachweis. The report shows which possibilities for the disposal of SF, HLW and ILW exist in Switzerland and summarises the current state of general academic and applied geoscientific research as well as the projectspecific knowledge base that has been developed by Nagra over the past 30 years.

The descriptions and assessments of the potential host rocks and areas are based on attributes that take into account experience gained both in Switzerland and abroad and are in agreement with international practice. The assessments put the highest priority on long-term safety and are restricted to the corresponding geological features. These assessments lead to the following results:

1. Due to the required long-term stability and the relative simplicity of structure the preferred geologic-tectonic region is the Molasse Basin and the north-eastern Tabular Jura.
2. Within this region different host rocks and areas can in principle ensure the safety of a deep geological repository for SF/HLW/ILW if the engineered barriers are adapted to the geological conditions.
3. An assessment of the differences between potential host rocks leads to the conclusion that the Opalinus Clay has advantages from the geological point of view over other host rocks (crystalline basement, claystones of the Lower Freshwater Molasse).
4. Opalinus Clay (the host rock preferred by Nagra) can be found in other areas besides the Zürcher Weinland which might, in principle, be appropriate for siting a SF/HLW/ILW repository. These are the areas of Nördlich Lägeren, Bözberg and Jurasüdfuss.

An assessment of potential siting areas in view of the preparation of a General Licence application will also have to consider land use planning and socio-economic aspects. This will be carried out in the next step according to the Sectoral Plan for Geological Disposal (Sachplan Geologische Tiefenlager) under the guidance of the relevant Swiss authorities. The Sectoral Strategy (Konzept) of the Plan, which defines both the relevant criteria and the procedure, is currently being developed by the Federal Office of Energy.

Technical Report NTB 05-01

Grimsel Test Site
Investigation Phase V
HPF-Experiment:
Modelling Report

Summary

Technical Report NTB 04-09

A Report of the Spent Fuel Stability (SFS) Project of the 5th Euratom Framework Program:
Spent Fuel Evolution under Disposal Conditions – Synthesis of Results from the EU Spent Fuel Stability (SFS) Project

Summary

Over the period 2002-2004, a large number of European organisations cooperated on the EU Project SFS – Spent Fuel Stability under disposal conditions. The objective of the SFS Project was to develop a reliable and robust model for the spent fuel source term which can be used in performance assessment exercises by the waste management agencies responsible for assessing the feasibility and safety of potential geological disposal systems for spent fuel, whatever the countries and disposal system designs may be.

A new model for short-term release of fission products (Instant Release Fraction or IRF) was developed based on the anticipated fission product release from various fuel microstructures (gap, rim, grain boundaries) and the potential solid-state diffusion of fission products prior to canister breaching. For the oxide matrix of the spent fuel, a Matrix Alteration Model (MAM) was developed, which is linked to the production of oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order to (i) upgrade the current radiolytic kinetic scheme, (ii) determine the relationship between fuel alteration rate and alpha activity by performing experiments on alpha-doped samples of UO₂ and (iii) experimentally test the potential inhibition effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed that was calibrated using experiments on inactive UO₂ samples, although the hydrogen effects remain to be incorporated completely into the model. The integrated model combining the IRF and MAM was used to illustrate long-term performance of representative spent fuel disposed of in granite, salt and clay host rock environments.

The findings of the SFS Project have significantly enhanced the understanding of phenomena that may affect radionuclide release from spent fuel under disposal conditions and have helped to more clearly identify areas in which uncertainties should be reduced through future research.

Technical Report NTB 04-08

A report of the Spent Fuel Stability (SFS) Project of the 5th Euratom Framework Program:
Estimates of the Instant Release Fraction for UO₂ and MOX Fuel
at t = 0.

Summary

Technical Report NTB 04-07

Matrix Diffusion for Performance Assessment – Experimental Evidence, Modelling Assumptions and Open Issues

Summary

In this report a comprehensive overview on the matrix diffusion of solutes in fractured crystalline rocks is presented.

Some examples from observations in crystalline bedrock are used to illustrate that matrix diffusion indeed acts on various length scales. Fickian diffusion is discussed in detail followed by some considerations on rock porosity. Due to the fact that the dual-porosity medium model is a very common and versatile method for describing solute transport in fractured porous media, the transport equations and the fundamental assumptions, approximations and simplifications are discussed in detail. There is a variety of geometrical aspects, processes and events which could influence matrix diffusion. The most important of these, such as, e.g., the effect of the flow-wetted fracture surface, channelling and the limited extent of the porous rock for matrix diffusion etc., are addressed. In a further section open issues and unresolved problems related to matrix diffusion are mentioned. Since matrix diffusion is one of the key retarding processes in geosphere transport of dissolved radionuclide species, matrix diffusion was consequently taken into account in past performance assessments of radioactive waste repositories in crystalline host rocks. Some issues regarding matrix diffusion are site-specific while others are independent of the specific situation of a planned repository for radioactive wastes. Eight different performance assessments from Finland, Sweden and Switzerland were considered with the aim of finding out how matrix diffusion was addressed, and whether a consistent picture emerges regarding the varying methodology of the different radioactive waste organisations. In the final section of the report some conclusions are drawn and an outlook is given. An extensive bibliography provides the reader with the key papers and reports related to matrix diffusion.

Technical Report NTB 04-06

Effects of Post-disposal Gas Generation in a Repository for Spent Fuel, High-level Waste and Long-lived Intermediate Level Waste Sited in Opalinus Clay

Summary

The present study provides a comprehensive treatment of the issue of gas generation in and transport from a repository for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (ILW), sited in the Opalinus Clay of the Zürcher Weinland in northern Switzerland (Project Entsorgungsnachweis). It thus provides a synthesis of all information available on the fate of gas, including the original sources of data and the arguments and discussions documented in the Geosynthesis Report (Nagra 2002a), the Safety Report (Nagra 2002c) and the Models, Codes and Data Report (Nagra 2002d).

The issue of how gas generation in and transport from waste repositories may influence disposal system performance has been under study for many years, both at Nagra and internationally. Quantitative analysis is required for a number of reasons: Some of the gases produced may be radioactive (e.g. those containing ¹⁴C), thus their release from the repository to the biosphere may represent a radiological hazard. Gas accumulation may eventually result in pressures high enough to damage the engineered barrier system or host rock, or influence groundwater movement. This could potentially affect the transport of dissolved radionuclides. Moreover, in order to establish confidence in the safety case, all significant processes associated with system performance and safety should be addressed appropriately.

The report consists of three main parts:

First, a synthesis of basic information on the host rock and on details of repository construction and wastes is given, focusing on the design and materials aspects that influence gas production and transport from waste emplacement tunnels and projected gas production rates in the repository. The nature and quantities of gas-generating materials and the processes leading to gas production are discussed in detail. The total generated gas is estimated to be 4 × 10⁷ m³ (SATP) for SF/HLW and 5 × 10⁵ m³ (SATP) for ILW, the latter including the contribution from Zircaloy and degradation of organic matter. Based on a steel corrosion rate of 1 μm a^-1, complete corrosion of the steel canisters would take about 2 × 10⁵ years, at which time gas production would cease. In the second part of the report, the gas transport characteristics of the engineered barrier system and the geosphere are discussed. The relevant gas transport processes are, i) advective-diffusive transport of gas dissolved in the porewater, ii) visco-capillary two-phase flow, and iii) dilatancycontrolled gas flow. It is shown that gas transport in macroscopic tensile fractures can be ruled out under the expected repository conditions. The gas-relevant properties are discussed for all EBS materials, including bentonite, sand/bentonite mixtures, cementitious mortar and concrete, as well as for the undisturbed host rock, EDZ, steeply dipping faults, confining units and regional aquifer at the site. Finally, the conceptual understanding of gas transport pathways from the emplacement tunnels either through the tunnel system or through the host rock, and further through overlying formations into the biosphere, is presented. The third part discusses the effects of gas on system performance, based on the available information on gas generation, gas transport properties and gas patheways provided in the previous parts of the report. Simplified model calculations based on a mass balance approach for the gas generated within the repository are performed. In a first step, the pressure evolution and gas migration in the SF/HLW/ILW repository are calculated. It is shown that the generated gas is transported through the host rock and the tunnel system and slowly accumulates in the overlying Wedelsandstein formation, from where it is gradually released through diffusion to the Malm aquifer. In a second step, this model is applied to the transport and release of potentially volatile ¹⁴C, assumed to be carried with the principal non-radioactive gasses, hydrogen and methane. These calculations suggest that doses in such cases would be well below the regulatory dose limit. The possible displacement of some porewater from the near field into the surrounding rock and/or through the ramp/shaft system due to gas pressure build-up is evaluated in an additional assessment model. In the framework of various assessment cases (base case, parameter variations and "what if?" case) it is shown that the expected doses for such cases are well below the regulatory dose limit.

Technical Report NTB 04-05

Modelling of Tracer Profiles in Pore Water of Argillaceous Rocks in the Benken Borehole: Stable Water Isotopes, Chloride and Chlorine Isotopes

Summary

Technical Report NTB 04-04

Comparison of ORIGEN2.1 with Selected Computer Codes

Summary

Codes used to calculate radionuclide inventories have been compared with the aim of testing the predictive capability of the ORIGEN2.1 code. A suite of safety relevant radionuclides were examined and differences in the output of the codes were used as an indication of uncertainty in the ORIGEN2.1 output.

It was found that the activities of most of the safety relevant nuclides can be reliably predicted by ORIGEN2.1 calculations. However, for 9 out of the 64 radionuclides considered in this work, ORIGEN2.1 predicted activities that cannot be considered reliable. The origin of the discrepancies between the codes for these 9 nuclides will be further investigated by Nagra.

It was also shown that the agreements and the deviations noted for the safety relevant nuclides do not change significantly for a variation of the key parameters burnup, enrichment and power level.

The calculations give fair estimates of appropriate error bars that should be assigned to all of the obtained activities within ORIGEN2.1 calculations.

Technical Report NTB 04-03

Nuclide Transport and Diffusion for Vein and Fracture Flow

Summary

Modelling radionuclide transport through crystalline rock is usually based on a small water flow in a system of narrow fractures. This flow is denoted as fracture flow. In our model, it implies planar water-conducting channels and adjacent zones of dominant matrix diffusion. According to the constitution of the rock, it can be necessary to consider additionally a vein flow being characterized by cylindrical water-conducting channels and adjacent zones of dominant matrix diffusion. Transport calculations, based on a dual porosity concept, were performed for vein as well as for fracture flow. An extensive discussion of the results provides an overview on important parameter dependencies and on the major vein flow effects. Formulae for quick estimates are given to guide quantitative interpretation of break-through curves. The discussion of analytical results for nuclide diffusion from a planar and from a cylindrical boundary backs up the comments on matrix diffusion.

The following effects of vein flow onto the break-through curves are illustrative examples of useful findings:

Technical Report NTB 04-02

Experimental and Modelling Investigations on Na-Illite: Acid-Base Behaviour and the Sorption of Strontium, Nickel, Europium and Uranyl

Summary

In an extensive study the physico-chemical, protolysis and sorption characteristics of Sr(II), Ni(II), Eu(III) and U(VI) have been measured on illite and modelled over a wide range of pH, sorbate and NaClO₄ concentrations.

Samples of Illite du Puy, collected in the region of Le Puy-en-Velay, France, were carefully conditioned to the Na-form and physico-chemically characterised. Potentiometric titrations on suspensions of the Na-illite were carried out using a batch back titration technique in 0.01, 0.1 and 0.5 M NaClO₄ background electrolytes from pH~2 to ~12 in an inert atmosphere glove box. The supernatant solutions from each titration experiment in each series were analysed for K, Mg, Ca, Sr, Si, Al, Fe and Mn.

Sorption edges (solid/liquid distribution ratios versus pH at trace sorbate concentrations and constant ionic strength) were determined for Sr, Ni, Eu and U on Na-illite as a function of NaClO₄ concentration under anoxic conditions (CO₂ ≤ 2 ppm, O₂ ≤ 2 ppm.). Sorption isotherms for the same set of radionuclides under similar conditions were measured for Na-illite suspensions in 0.1 M NaClO₄ at fixed pH values.

The titration data were modelled in terms of the protolysis of two amphoteric edge sites (=S^W1OH and =S^W2OH) without an electrostatic term. The protonation/deprotonation constants and site capacities obtained from the titration measurements were then fixed. The sorption edge and isotherm data were modelled with strong (=S^SOH) and weak (=S^W1OH) surface complexation sites, assumed to have the same protolysis constants, again without electrostatic terms. Uptake by cation exchange was included in all of the calculations. This sorption model, the 2 site protolysis non electrostatic surface complexation and cation exchange model, had been developed previously for montmorillonite and was successful in describing the sorption characteristics of Sr, Ni, Eu and U on Na-illite over a wide range of conditions. Cation exchange capacity, strong and weak site capacities and protolysis constants for Na-illite are given, together with surface complexation constants and selectivity coefficients for Sr, Ni, Eu and U.

At 0.01 M NaClO₄ and pH < 8 the sorption of Sr, Ni, Eu and U was dominated by a cation exchange mechanism. The strong dependency of sorption on pH observed under these conditions arose from the competitive effects of Ca and Al on the uptake of the sorbate. Selectivity coefficients for Ca and Al with respect to Na were deduced from these measurements.

Technical Report NTB 04-01

Grimsel Test Site
Investigation Phase V
Modelling the Transport of Solutes and Colloids in a Water-Conducting Shear Zone in the Grimsel Test Site

Summary

This report describes modelling of the transport of solutes and colloids in an experimental system comprising an artificial dipole flow field in a water-conducting shear zone at Nagra's Grimsel Test Site (GTS) in the central Swiss Alps. The modelling work forms part of the Colloid and Radionuclide Retardation Project (CRR), which includes a series of field transport experiments and a supporting laboratory programme, as well as modelling studies. Four independent groups representing different organisations or research institutes have conducted the modelling, with each group employing its own modelling approach or approaches. Only the work conducted at the Paul Scherrer Institute (PSI) is described in the present report.

Bentonite, which is widely considered as a potential backfill material for a range of radioactive and chemotoxic wastes, could conceivably provide a source of colloids that could then influence the transport of radionuclides released from a geological repository for radioactive waste. The main objective of CRR is to enhance understanding of the in situ retardation of radionuclides in the presence of bentonite colloids, in a system analogous to the near-field/geosphere interface of a geological repository.

The field transport experiments were carried out by injecting various cocktails of tracers, some of which included bentonite colloids, into the injection borehole of the dipole and measuring the resulting breakthrough curves. Modelling work was carried out in order to assist in the planning of the main experimental runs and to contribute to the interpretation of the results. Three model variants are used in the present study, namely a 1-D advection-dispersion model, similar to that developed in support of the earlier GTS Migration Experiment (MI), a 2-D advection-dispersion model, and a non-Fickian dispersion model: the CTRW (continuous time random walk) model. The 1-D and 2-D models treat dispersion as a diffusion-like process that obeys Fick's laws. They also include the retardation mechanisms of matrix diffusion of solutes and solute sorption on matrix pore surfaces. Colloids are excluded from matrix pores in all the model variants. The CTRW model allows a more general treatment of dispersion, but does not currently include matrix diffusion, and so was only applied to the transport of colloids.

The modelling of preliminary tests carried out in advance of the main CRR experimental runs showed that the 1-D and 2-D advection-dispersion models with matrix diffusion provide similarly good fits for tracers conveyed as aqueous species, using reasonable and consistent sets of parameter values. They were less successful at modelling colloid breakthrough, and various explanations for this have been considered. Of these, the occurrence of non-Fickian dispersion is considered the most likely. The CTRW model, which allows for non-Fickian dispersion, indeed provides an adequate fit in the case of colloids with a consistent set of parameters.

On the basis of the modelling of the preliminary tests, predictions of the breakthrough of Am, Pu, Np, U and Cs, both with and without the addition of bentonite colloids to the injection cocktail, were made for the main experimental runs in advance of the experiments being carried out. The experimental measurements confirm the model assumption that at least part of the injected inventories of Am, Cs, Pu and Th migrates in association with bentonite colloids. Furthermore, discrepancies between predictions and measurements that Am, Pu and Th are transported in colloidal form, even when no bentonite colloids are added to the injection cocktail. The addition of bentonite colloids, however, increases the recovery of these tracers. The characterisation of colloids in the injection cocktails, which was not available at the time that the model predictions were made, enables improved agreement to be obtained between model calculations and measured breakthrough curves.

The CRR experiment and the present modelling study have a number of limitations. For example, there is the possibility that non-Fickian dispersion affects the transport of solutes as well as colloids. It is not, however, possible to discriminate between the impact of this non- Fickian dispersion and matrix diffusion effects by modelling the breakthrough curves. If non- Fickian dispersion of solutes takes place, then this has implications for the derivation of parameter values for safety assessment (and especially sorption coefficients) from field tracer transport experiments. In particular, values derived using advection-dispersion models with matrix diffusion and with dispersion modelled using Fick's laws need to be viewed with caution.

The modelling approaches used in the present study may not be directly applicable to safety assessment problems and the direct implications of the results of this study for safety assessment are limited. It can, however, be said that the study has demonstrated the high degree of mobility of bentonite and other colloids in a system that is at least in some ways comparable to those of interest in safety assessment, and has shown that bentonite colloids can at least potentially affect the transport of some safety relevant radionuclides over longer temporal and spatial scales than those addressed here. 

Technical Report NTB 03-13

Grimsel Test Site
Investigation Phase V
Effective Field Parameter FEP

Summary

Prediction modelling of groundwater flow and solute transport is an indispensable prerequisite for any long-term safety analysis concerning radioactive waste disposal in deep geological formations. This requires extensive geological and hydraulic information, a powerful numerical model, and the definition of effective parameter sets for the model.

The objectives of the EFP-project (Effective Field Parameter) are to examine methods for characterising the rock mass and for developing a structural model, to check the numerical model for calculation of the temporal and spatial distribution of tracer concentration on a large-scale, and to validate the developed model using in-situ data from large-scale tracer experiments at the Grimsel Test Site (GTS).

The main working programs of EFP include:

Technical Report NTB 03-12

Sorption Data Bases for Opalinus Clay Influenced
by a High pH Plume

Summary

The interaction of groundwater with the large quantities of cement/concrete used in the construction and backfilling of emplacement tunnels containing long-lived intermediate level radioactive waste may give rise to the release of a pulse of hyperalkaline fluid (pH plume) into the surrounding rock. Since the pH of this plume could remain in excess of 12.5 for tens of thousands of years, many minerals in a sedimentary host rock would be unstable leading to dissolution reactions, secondary mineral precipitation and changes in groundwater chemistry.

An Opalinus clay formation in the Zürcher Weinland, is under consideration by Nagra as a potential host rock for a repository of spent fuel (SF), vitrified high-level waste (HLW) from reprocessing of spent fuel and long-lived intermediate-level radioactive waste (ILW). The purpose of this report is to assess the effects of the interactions between a pH plume and Opalinus clay on the sorption properties of the formation and to provide appropriate sorption data bases.

Technical Report NTB 03-11

Grimsel Test Site
Investigation Phase V
GAM – Gas Migration Experiments in a Heterogeneous Shear
Zone of the Grimsel Test Site

Summary

This report documents the scientific investigations carried out as part of the GAM project between June 1997 and April 2001 at the Grimsel Test Site (GTS) within the framework of Investigation Phase V (1997 – 2001). Four radioactive waste management organisations participated in the GAM experiment, namely ANDRA, ENRESA, NAGRA and Sandia National Laboratories (SNL) for the US Department of Energy (DOE). The experiment team consisted of the delegates of the participating organisations, research groups from the Swiss Federal Institute of Technology / Zurich and from the Technical University of Catalonia / Barcelona and, last but not least, several contractor teams.

Essential aims of the GAM investigation programme were the development and testing of laboratory and field equipment for tracer experiments. Innovative laboratory technologies were applied, such as Laser Scanning Confocal Microscopy and X-ray tomography, flow visualisation in artifical fractures, nuclear magnetic resonance measurements and neutron radiography. Furthermore, a new technique was tested for the recovery of well preserved core samples from the GAM shear zone. Novelties in field testing comprised the use of an on-line counter for the particle tracer tests and a georadar survey of gas and brine injection tests with a high frequency borehole antenna.

The development of upscaling methodologies and the derivation of effective parameters for single- and two-phase flow models was another issue of interest. The investigations comprised theoretical studies on solute transport in non-uniform flow fields and assessment of the impact of the microstructure on solute and gas transport. Closely related to these theoretical studies was the numerical interpretation of the combined solute and gas tracer tests, which revealed the great potential of such data sets with regard to model discrimination.

As a final step in the synthesis task of the GAM project, a model abstraction process was established, aimed at integrating the descriptive studies on various scales with the hydraulic investigations to produce a consistent conceptual model of flow and transport processes in the heterogeneous shear zone.

Technical Report NTB 03-10

Time-dependent Flow and Transport Calculations for Project Opalinus Clay (Entsorgungsnachweis)

Summary

This report describes two specific assessment cases used in the safety assessment for a proposed deep geological repository for spent fuel, high level waste and long-lived intermediate-level waste, sited in the Opalinus Clay of the Zürcher Weinland in northern Switzerland (Project Entsorgungsnachweis, NAGRA, 2002d).

In this study the influence of time dependent flow processes on the radionuclide transport in the geosphere is investigated. In the Opalinus Clay diffusion dominates the transport of radionuclides, but processes exist that can locally increase the importance of the advective transport for some time. Two important cases were investigated:

(1) glaciation-induced flow due to an additional overburden in the form of an ice shield of up to 400 m thickness and (2) fluid flow driven by tunnel convergence.

For the calculations the code FRAC3DVS (Therrien & Sudicky, 1996) was used. FRAC3DVS solves the three-dimensional flow and transport equation in porous and fractured media.

For the case of glaciation-induced flow (1) a two-dimensional reference model without glaciations was calculated. During the glaciations the geosphere release-rates are up to a factor of about 1.7 higher compared to the reference model. The influence of glaciations on the transport of cations or neutral species is less than for anions, since the importance of the advective transport for anions is higher due to the lower accessible porosity for anions. The increase in the release rates during glaciations is lower for sorbing compared to non-sorbing radionuclides. The influence of the tunnel convergence (2) on the transport of radionuclides in the geosphere is very small. Due to the higher source term the geosphere release rates are slightly higher if tunnel convergence is considered.

In addition to the two assessment cases this report investigates the applicability of the one-dimensional approximation for modelling transport through the Opalinus Clay. For the reference case of the safety assessment the model chain STMAN-PICNIC-TAME is used. In order to evaluate radionuclide release and transport, the geometry of the repository near-field/geosphere system is simplified and the Opalinus Clay is treated as a one-dimensional layer. In this study the code FRAC3DVS is used to assess the effects of the simplifications by calculating a two-dimensional model which includes both the Opalinus Clay and the SF / HLW bentonite annulus.

The one-dimensional approximation gives results similar to the geometrically more realistic FRAC3DVS model. Discrepancies introduced by the one-dimensional approximation are shown to be small and the results are always conservative compared with the FRAC3DVS calculations. This modelling exercise thus gives strong support for the applicability of the one-dimensional approximation.

Technical Report NTB 03-09

A Generic Procedure for the Assessment of the Effect of Concrete Admixtures on the Retention Behaviour of Cement for Radionuclides: Concept and Case Studies

Summary

Concrete admixtures are unavoidable components of cements used for the conditioning of radioactive waste and of concretes used in the construction of a cementitious underground repository for the geological disposal of radioactive waste.

Almost nothing is known about the influence of concrete admixtures on the sorption of radionuclides on cement, which is one of the main retardation mechanisms responsible for the isolation of radionuclides from the environment. In this work a screening procedure is proposed in the form of a checklist, through which it can be decided whether or not a specific concrete admixture may have adverse effects on the sorption of radionuclides by cement. The screening procedure is intended to be generic and applicable to any concrete admixture that may be used in the future. It focuses on the following potential main effects of concrete admixtures or their transformation products possibly formed under highly alkaline conditions: (i) interaction between radionuclides and concrete admixture in solution (complexation) and (ii) competition for surface sites between concrete admixtures and radionuclides or other strong complexants sorbing on cement. The screening procedure is set up in a hierarchical manner. The degree of complexity of the procedure increases with increasing depth of the investigations. The effort for the assessment of a specific concrete admixture under investigation can be kept thereby to a reasonable minimum.

In parallel to the development of a suitable experimental protocol, a few selected concrete admixtures, i.e. sulfonated naphthalene-formaldehyde condensates, lignosulfonates and a plasticiser used at the Paul Scherrer Institut for waste conditioning, were subjected to this screening procedure. The effect of these concrete admixtures on the sorption properties of Ni(II), Eu(III) and Th(IV) on cement was investigated using crushed hardened cement paste (HCP) and cement pastes prepared in the presence of these concrete admixtures. Although some of the concrete admixtures investigated were shown to be strong complexants, no adverse effect on the sorption of the radionuclides tested could be observed under realistic conditions, i.e. at representative HCP to pore water ratios and representative concrete admixture to cement ratios. With the exception of lignosulfonate, this finding could be explained by the sorption of the concrete admixtures onto the HCP, which could be modelled within an order of magnitude using a Langmuir isotherm. The pore water concentrations of the concrete admixtures tested were thereby reduced to levels at which the formation of radionuclide complexes was no longer of importance. Further, the sorption model suggests that the HCP surface does not become saturated with the concrete admixtures tested. The sorption on HCP of the radionuclides tested and isosaccharinic acid, a strong complexant produced in cement conditioned wastes containing cellulose, was found to be unaffected by the amount of concrete admixtures sorbed on the HCP under the experimental conditions investigated. Chemical transformation of the concrete admixtures studied was not investigated in detail. An indication for a chemical transformation with a possible impact on radionuclides sorption was only found in the case of lignosulfonates using the methods applied in this study.

As a result of the experiments carried out within the framework of this study, it can be concluded that the proposed screening procedure is well suited for a broad assessment of the effect of concrete admixtures on the retention behaviour of HCP for radionuclides. However, more substance specific knowledge may be needed for a comprehensive assessment, if, in a specific case, it is not possible to break down the complexity of the system, cement – concrete admixture – radionuclide retention, in the manner proposed in this work.

Kopie von Technical Report NTB 03-08

Cellulose Degradation at Alkaline Conditions: Long-Term Experiments at Elevated Temperatures

Summary

The degradation of pure cellulose (Aldrich cellulose) and cotton cellulose at the conditions of an artificial cement pore water (pH 13.3) has been measured at 60 °C and 90 °C for reaction times between 1 and 2 years. The purpose of the experiments is to establish a reliable relationship between the reaction rate constant for the alkaline hydrolysis of cellulose (mid-chain scission), which is a slow reaction, and temperature. The reaction products formed in solution are analysed for the presence of the two diastereomers of isosaccharinic acid using high performance anion exchange chromatography combined with pulsed amperometric detection (HPAEC-PAD), other low-molecular weight aliphatic carboxylic acids using high performance ion exclusion chromatography (HPIEC) and for total organic carbon. The remaining cellulose solids are analysed for dry weight and degree of polymerisation. The degree of cellulose degradation as a function of reaction time is calculated based on total organic carbon and on the dry weight of the cellulose remaining.

The degradation of cellulose observed as a function of time can be divided in three reaction phases observed in the experiments: (i) an initial fast reaction phase taking a couple of days, (ii) a slow further reaction taking ~100 days and (iii) a complete stopping of cellulose degradation levelling-off at ~60 % of cellulose degraded. The experimental findings are unexpected in several respects: (i) The degree of cellulose degradation as a function of reaction time is almost identical for the experiments carried out at 60 ° C and 90 °C, and (ii) the degree of cellulose degradation as a function of reaction time is almost identical for both pure cellulose and cotton cellulose. It can be concluded that the reaction behaviour of the materials tested cannot be explained within the classical frame of a combination of the fast endwise clipping of monomeric glucose units (peeling-off process) and the slow alkaline hydrolysis at the temperatures tested here. It may be hypothesised that the alkaline hydrolysis has even not been observed in the experiments. However, if this is true, cellulose degradation proceeded via another unknown type of reaction. Mass balances for carbon show that the large majority of reaction products found in solution can be explained by formation of isosaccharinic acids and other low-molecular weight carboxylic acids.

With respect to long-term predictions for cellulose degradation at room temperature it can be concluded that the kinetic parameters for alkaline hydrolysis as proposed in the work of PAVASARS (Linköping Studies in Arts and Science, 196, Linköping University, Sweden, 1999) are too large and that complete cellulose degradation at these temperatures occurs only within time scales larger than hundreds of years. However, it is not possible from the experimental evidences, to corroborate the validity of a linear extrapolation (“Arrhenius equation”) of the reaction rates measured at temperatures between ~140 and 190 °C to room temperature, from which it was previously concluded that complete cellulose degradation would take time spans of the order of millions of years.

An interesting observation in the present experiments is the chemical instability of ­isosaccharinic acid at 90 ° C, which has been hypothetically interpreted as a fragmentation induced by the sorption of α-isosaccharinic acid on Ca(OH)₂. Carbon mass balances show that -isosaccharinic acid is thereby transformed to other low­molecular weight carboxylic acids. Such a reaction would be an interesting long-term perspective for performance assessment of the disposal of cellulose-containing radioactive waste, in that it may reduce the concentration of organic compounds strongly complexing radionuclides.

Technical Report NTB 03-07

Diffusion of HTO, 36Cl-, 125I- and 22Na+ in Opalinus Clay: Effect of confining pressure, sample orientation, sample depth and temperature

Summary

The Opalinus Clay (OPA) formation in the Zürcher Weinland is a potential host rock for a repository for spent fuel, vitrified high-level waste and long-lived intermediate-level waste in Switzerland. Owing to its small hydraulic conductivity (10^-14 - 10^-13 m·s^-1), it is expected that transport of solutes will be dominated by diffusion. This study addresses the diffusion of tritiated water (HTO), ³⁶Cl^-, ¹²⁵I^-­ and ²²Na⁺ through Opalinus Clay samples. The samples were collected in the Mont Terri Underground Rock Laboratory, where the OPA formation is located at a depth between -200 and -300 m below the surface, and in the deep borehole in Benken (Zürcher Weinland), where the OPA layer is located at a depth between -539 and -652 m.

Effective diffusion coefficients (D_e), rock capacity factors (α) and diffusion-accessible porosities (ε) were measured using the through-diffusion technique. Transport (diffusion) was measured both perpendicular and parallel to the bedding. Special cells that allowed the application of an axial confining pressure were designed. The pressures applied ranged from 1 to 5 MPa for Mont Terri samples and between 4 and 15 MPa for Benken samples, the upper values representing the in-situ confining pressure at both locations. The test solutions used in the experiments were synthetic Opalinus Clay pore water, which has Na and Cl as main components (Mont Terri: I = 0.39 M; Benken: I = 0.20 M).

Pressure only had a small effect on the value of the effective diffusion coefficients. In the case of Mont Terri samples, increasing the pressure from 1 to 5 MPa resulted in a decrease of the effective diffusion coefficient of 20% for HTO, 27% for ³⁶Cl^-, 29% for ¹²⁵I^-­­ and 17 % for ²²Na⁺ In the case of Benken samples, increasing the pressure from 4 to 15 MPa resulted in a decrease of D_e of 17% for HTO, 22% for ³⁶Cl^-, 32% for ¹²⁵I^-­ and 17 % for ²²Na⁺. Moreover, the effective diffusion coefficients for ³⁶Cl^- are smaller than for HTO, which is consistent with an effect arising from anion exclusion. This ion exclusion effect is smaller in samples from Mont Terri than in samples from Benken, which can be explained by the higher ionic strength of the Mont Terri water used in the experiments. The diffusion of ²²Na⁺ is similar to that of HTO in the case of Mont Terri OPA. For Benken OPA, the D_e value of ²²Na⁺ is a factor of 2 higher than that of HTO. This last observation cannot be explained so far but is comparable to experimental data from ANDRA (1999) on Callovo-Oxfordian claystones from the Meuse/Haute Marne site.

¹²⁵I^-­ is retarded with respect to ³⁶Cl^-. This is caused by a weak sorption of ¹²⁵I^-­ on the Opalinus Clay. The distribution coefficients, calculated from the rock capacity factor under the assumption that the diffusion-accessible porosity of ¹²⁵I^- is the same as for ³⁶Cl^-, range between 0.01 and 0.02 cm³·g^-1. The effective diffusion coefficients of ¹²⁵I^-­ are comparable with those of ³⁶Cl^-.

Out-diffusion data of HTO, ³⁶Cl^- and ²²Na⁺ are in good agreement with the through­diffusion data. In the case of ¹²⁵I^-­ the agreement is less. The flux calculated with D_e and α derived from through-diffusion measurements is smaller than the observed flux. This indicates that other (unknown) processes are taking place.

The diffusion coefficients measured in this study on Mont Terri samples are in good agreement with recent measurements of three other laboratories, within the framework of a laboratory comparison exercise. The values of the diffusion-accessible porosities, however, show a larger degree of scatter, indicating that through-diffusion is not the method of choice for obtaining reliable porosity values.

Diffusion parallel to the bedding is higher than diffusion perpendicular to the bedding. The effective diffusion coefficient for diffusion parallel to bedding is a factor of 4 – 6 larger than for diffusion perpendicular to the bedding. This is due to the layered structure of the Opalinus Clay, resulting in a smaller tortuosity factor for diffusion along the bedding planes. The observed effect was similar for HTO, ³⁶Cl^- and ²²Na⁺. This anisotropy is more pronounced for the Opalinus Clay from Benken than for Mont Terri, indicating that the clay platelets are more preferentially oriented in the case of Benken OPA.

The temperature dependence of diffusion of HTO in OPA is of an Arrhenius type. The activation energy (22 kJ·mol^-1), however, is larger than for diffusion in bulk water (18 kJ·mol^-1). This indicates that confined water in the narrow pores of the Opalinus Clay has partly a different structure.

Diffusion measurements with HTO on OPA samples from different depths showed that the effective diffusion coefficients for diffusion perpendicular to the bedding decrease with increasing depth. The difference between the top and the bottom of the OPA layer, however, is not more than a factor of 1.5. For diffusion parallel to the bedding, no difference between top and bottom could be observed. It can be concluded that the Opalinus Clay layer is very homogeneous with respect to its diffusion properties.

The effective diffusion coefficient measured for the HTO in OPA is in good agreement with values measured in other sedimentary rocks and can be related to the porosity using Archie’s Law with exponent m=2.5.

Technical Report NTB 03-06

Project Opalinus Clay: Integrated Approach for the Development
of Geochemical Databases Used for Safety Assessment

Summary

Chemical retention plays a central role in the Swiss repository concept for spent fuel/highl-level waste (SF/HLW) and intermediate level waste (ILW). Chemical retention is taken into account in the safety assessment calculations by applying the concept of solubility limits and K_d values for the safety-relevant nuclides. The necessary data were compiled in five geochemical databases, the derivation of which is described in detail in the corresponding reports (Berner 2002a; 2003; Bradbury & Baeyens 2003a and b; Wieland & Van Loon 2002).

The elaboration of the geochemical databases (GDBs) was done by a team of scientists from the Paul Scherrer Institute and the Safety Assessment Group at Nagra in a two years' effort based on many years of extensive scientific investigations. An integrated approach was applied, which was based on the principles of chemical thermodynamics, sound experimental sorption and diffusion data and expert judgement. A consistent procedure and a number of quality assurance measures contributed to obtaining high quality retention data. A strong emphasis was on the derivation of transparent and traceable "best estimate" data and associated uncertainties.

The applied methodology can be separated into three parts. The first part consisted of elaborating the geochemical foundations, which included a detailed update of the Nagra/PSI thermodynamic database, derivation of the geochemical in-situ boundary conditions (e.g. pH and Eh) in the different compartments and generation of experimental sorption data in clay and cement systems. The second part involved the derivation of scientifically sound retention data, i.e. the solubility limits and sorption values for safety-relevant radionuclides under relevant repository conditions. The last part in the overall procedure involved the critical evaluation and, where necessary, adaptation of the values for their use in safety assessment calculations. This also included comparison with recent databases from other countries and evaluation of natural analogue data.

The solubility limit databases for the canister-bentonite and cementitious environments show reasonable agreement with other databases from recent safety assessments in spite of the differences in the underlying thermodynamic data and assumed geochemical conditions.

Our approach for deriving K_d values and apparent diffusion coefficients in the bentonite and the clay host rock differs from those applied in most other assessments. Whereas the latter are principally based on diffusion measurements in compacted clays, we systematically derived K_d values from well-controlled batch experiments and adapted these to the compacted in-situ conditions. Nevertheless, the proposed K_d and D_a values agree fairly well with those used in the other assessments, except for tetravalent species. This agreement is further supported by comparison of sorption values derived in batch systems with those obtained from Japanese diffusion measurements on Kunigel bentonite. This result strongly suggests that compaction and swelling do not have a strong effect on the retention properties of the clay.

In general, higher sorption and lower D_a values for tetravalent metals are used in our assessment for the bentonite near field and clay host rock relative to other assessment studies. This discrepancy is particularly large in the case of the redox-sensitive Tc(IV), U(IV) and Np(IV) species. The reasons for this are not clear, but might arise from (i) the more conservative treatment of uncertainties in other assessments, (ii) poorly constrained redox conditions in diffusion experiments, or (iii) extrapolation of batch sorption data to compacted in-situ conditions. The last point, however, is rather unlikely from the indications pointed out above. In case of the cementitious near field the difference in K_d values between our and other assessments is less pronounced than for the clay cases.

A major identified uncertainty is related to the porewater composition in the clay compartments. This was accounted for in the derivation of GDBs by including a large range of pH/pCO₂ and Eh in the uncertainty treatment of solubility limits and K_d values. Furthermore, important conceptual model uncertainties identified in the present work include the effect of carbonate on solubility and sorption of tetravalent metals, the redox chemistry of Pu, the nature and crystallinity of solubility-controlling phases, kinetics of redox-sensitive radionuclides and physico-chemical properties of water in compacted clays. Since the derived databases are based on a reference temperature of 25 °C, further uncertainty arises from the slightly higher temperature (≈50 °C) predicted for the conditions relevant to safety assessment. From a preliminary assessment we expect, however, the temperature uncertainty to be of less relevance relative to those mentioned above.

The conceptual model uncertainties in the retention data were accounted for in the geochemical databases by the conservative approach for deriving the uncertainty range and for the estimation of the pessimistic values. In addition, the high uncertainty related to the transport behaviour of redox-sensitive RN was implicitly considered by analysing the effects of an oxidising near field as a "what-if?" case. 

Technical Report NTB 03-03

Grimsel Test Site
Investigation Phase V
The CRR Final Project Report Series III: Results of the Supporting Modelling Programme

Summary

The Colloid and Radionuclide Retardation Experiment (CRR) is dedicated to improving the understanding of the in situ retardation of safety-relevant actinides and fission products associated with bentonite colloids in the vicinity of the Engineered Barrier System (EBS)/host rock interface. In addition to a series of in situ dipole experiments that were carried out at the Grimsel Test Site (GTS), the project partners, namely ANDRA (F), ENRESA (E), FZK-INE (D), JNC (J), USDoE/Sandia (USA) and Nagra (CH), funded an extensive programme of laboratory and modelling investigations. The aims of CRR were: examination of the in situ migration of bentonite colloids in fractured rocks, investigation of the interactions between safety-relevant radionuclides and bentonite colloids in the laboratory and in situ and, in addition, the testing of the applicability of numerical codes for representing colloid-mediated radionuclide transport.

The present report is one of three final project reports that summarise the findings of the CRR project. In addition to this modelling report, the series includes reports on the field and laboratory work. This report summarises and discusses the results of the modelling investigations that were carried out by four teams (Enviros, FZK-INE, JNC and PSI) working largely independently with the aim of developing understanding of the structures and processes affecting tracer transport in the in situ field tests. 

The modelling teams had access to the same set of field observations and supporting laboratory work and the models developed by the teams had many similarities. All of the models considered that radionuclide tracers could be transported either in solution or in association with colloids and all considered:

The most significant difference was probably the treatment of the interaction between solutes and colloids; the assumptions employed included equilibrium sorption, non-equilibrium sorption with first-order kinetics and irreversible sorption of radioactive tracers on colloids. A similar range of assumptions has been used in modelling colloid-facilitated radionuclide transport in recent repository safety assessments. There were also significant differences in the treatment of hydrodynamic dispersion, including its treatment as a diffusion-like process described by Fick's Laws and as a non-Fickian process, as well as a model that explicitly modelled the dispersion arising from a network of multiple orthogonal fractures.

The Enviros and PSI teams carried out some predictive modelling in advance of the main experimental runs in order to test their approaches and to aid in the planning of the experiments. Most of the modelling was, however, carried out after the main runs and involved a degree of inverse modelling. The success of some modelling approaches (and, equally importantly, the difficulties of others) in predicting or reproducing the experimental results enabled a number of conclusions to be drawn.

The CRR experiment and the modelling work discussed in this report indicate that, in the main experimental run #32 with bentonite colloids added to the injection cocktail:

Some radionuclides, including Am, Pu and Th, were also transported in colloidal form even when no colloids were added to the injection cocktail (run #31). The addition of bentonite colloids, however, increased recovery of these tracers. The role of colloids in the transport of Np and U in the two main runs was not unambiguously determined. Laboratory experiments, however, demonstrated that colloidal species are of minor relevance for Np(V) and U(VI).

Regarding processes:

Technical Report NTB 03-02

Grimsel Test Site
Investigation Phase V
The CRR Final Project Report Series II: Supporting Laboratory Experiments with Radionuclides and Bentonite Colloids

Summary

A series of laboratory experiments was carried out in support of the in situ programme and of the modelling studies of the 'Colloid and Radionuclide Retardation Project' (CRR) at Nagra’s Grimsel Test Site (GTS). The aim of those experiments was to study the geochemical interaction of radionuclides in the system Grimsel groundwater – granodiorite / fracture filling material – bentonite and to provide the necessary data for the final planning of the in situ migration experiments. The laboratory studies showed that:

Technical Report NTB 03-01

Grimsel Test Site
Investigation Phase V
The CRR final project report series I: Description of the Field Phase – Methodologies and Raw Data

Summary

The Colloid and Radionuclide Retardation Experiment (CRR) is dedicated to improve the understanding of the in situ retardation of colloid-associated, safety-relevant actinides and fission products in the vicinity of the Engineered Barrier System (EBS)/host rock interface. In addition to a series of in situ dipole experiments that were carried out at the Grimsel Test Site (GTS), the project partners, namely ANDRA (F), ENRESA (E), FZK-INE (D), JNC (J), USDOE/Sandia (USA) and Nagra (CH), funded an extensive programme of laboratory and modelling investigations. The aims of CRR were: examination of the in situ migration of bentonite colloids in fractured rocks, investigation of the interactions between safety relevant radionuclides and bentonite colloids in the laboratory and in situ and, in addition, testing of the applicability of numerical codes for representing colloid-mediated radionuclide transport.

The present report is the first of a quadruplet of final project reports that summarise the findings of the CRR project. In addition to this field report, the series includes laboratory and modelling reports along with a final, synthesis report. This report summarises and discusses the results of the field investigations that were carried out in 2001 and 2002 as part of the overall CRR project.

The overall concept behind CRR is based on the fact that, in most high-level radioactive waste repository designs, the waste is packed in massive metal canisters which are surrounded by a large volume of bentonite clay (collectively known as the Engineered Barrier System, or EBS). The canisters will slowly degrade and eventually fail, releasing some radionuclides, most of which are expected to be retained and to decay within the bentonite. However, it is conceivable that erosion of the bentonite at the EBS/host rock interface will produce bentonite colloids and that a limited amount of radionuclides escaping the EBS may become associated with these colloids and migrate through water conducting features in the geosphere towards the biosphere.

The central part of the CRR project was a series of dipole tracer tests that were carried out in a well-defined shear zone, in which dipole flow fields of 2.2 and 5 m length were generated. Preliminary tracer tests were performed with uranine, followed by tests with bentonite colloids and homologue elements for the tri- and tetravalent actinides (Tb for Am, Hf and Th for Pu, respectively). The tests culminated in the injection of the final tracer cocktails containing different isotopes of Am, Np, Pu, U, Tc, Th, Cs, Sr and I in the absence and presence of bentonite colloids.

The field installations consisted of several on-line measurement devices such as a downhole uranine detection device for the determination of the tracer input functions, a High Purity Germanium (HPGe) detector for γ-spectrometric measurements as well as a Laser Induced Breakdown Detector (LIBD) and a Photon Correlation Spectroscopy (PCS) apparatus for on-site colloid detection. The analytical techniques that were used off-site consisted of α-/γ-spectrometry and ICP-MS (Inductively Coupled Plasma-Mass Spectrometry) measurements for radionuclide detection as well as of Single Particle Counting (SPC) for the determination of the different colloid size classes. The interaction of the strongly sorbing tri- and tetravalent actinides with the equipment was avoided by producing as many as possible of those parts of the in situ equipment that were in direct contact with the tracers in PEEK (an inert plastic).

The natural colloid background of the groundwater in the experimental shear zone showed an average colloid diameter around 200 nm and a stable colloid concentration around 5 μgL^-1. Increased colloid concentrations observed temporarily at the beginning of the experiments were most likely due to mechanical stress induced by pressure pulses generated during installation of the test setup. The four different colloid detection techniques, namely LIBD, ICP-MS1 , PCS and SPC, produced internally consistent breakthrough data of the injected bentonite colloids. The bentonite colloids arrived slightly earlier than did the conservative dye uranine and the recovery was about 90%. Filtration effects varied depending on the colloid size and measurement technique employed and, as such, require further investigation.

Homologue pre-tests proved to be very useful for the prediction of the in situ behaviour of triand tetravalent actinides. In the absence of bentonite colloids, a clearly lower recovery was found for the homologues than when injected together with bentonite colloids and the peak maxima of the homologue breakthrough were slightly shifted to earlier arrival times compared to that of the uranine.

The tracer cocktail composition for the final tracer injections covered the entire range of oxidation states from -I to VI and was decided based on the results of laboratory experiments, the kinetics of redox reactions and practical constraints on the in situ use of these elements. The preparation of an injection cocktail which contains tri- and tetravalent actinides proved to be problematic, as shown by the presence of a variable colloidal fraction for Am, Pu and Th, even in the absence of bentonite colloids. However, the injection cocktail, which included bentonite colloids, showed high colloid association and long term stability for the tri- and tetravalent actinides with the bentonite colloids, indicating that a significant proportion of the radionuclides were associated with the added bentonite colloids.

In the first run (without bentonite colloids), the tri- and tetravalent actinides Am, Th and Pu displayed lower recovery, less tailing and a peak time which was about 10 minutes earlier than U, Np and I (which is assumed to behave in a conservative fashion), indicating that a fraction of these actinides was transported in a colloidal state. With regard to the varying colloid content in the injection cocktail, the source of these colloids cannot yet be uniquely defined (homogeneous- or heterogeneous radiocolloids) and artifacts, for example, during cocktail preparation, cannot yet be ruled out completely.

With the addition of bentonite colloids, an increased recovery of Am, Pu and Th compared to the first run was observed. The shape of the breakthrough curves did not change significantly as the peak in the first experiment was also affected by a colloidal fraction. Only about 1% of the Cs was colloidally transported which implies that 90% of the initially colloid bound Cs in the injection cocktail (10% Cs in the injection cocktail was in colloidal form) desorbed during migration.

Finally it should be noted that the field experiments constitute only a part of the overall CRR project and interpretation and transfer of these data needs to be carried out taking into account the results of the laboratory experiments along with the effects of site groundwater chemistry, very short test duration and other technical constraints.

Technischer Bericht NTB 02-24

SMA/WLB:
Bohrlochversiegelung/-verfüllung
SB4a/s schräg

Summary

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 nonsorbing nuclides are barely or not at all held back in the sealed borehole. The impact of the borehole is comparable to flowpaths along cataclastic zones. It is due to the high retaining capacity of the nearfield 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.

Technical Report NTB 02-23

Project Opalinus Clay: FEP Management for Safety Assessment – Demonstration of disposal feasibility for spent fuel, vitrified high-level waste and long-lived intermediate-level waste (Entsorgungsnachweis)

Summary

Technical Report NTB 02-22

Project Opalinus Clay: Radionuclide Concentration Limits in the Cementitious Near-Field of an ILW Repository

Summary

The disposal feasibility study currently performed by Nagra includes a succession of quantitative models, aiming at describing the fate of radionuclides potentially escaping from the repository system. In this chain of models the present report provides the so called "solubility limits" (maximum expected concentrations) for safety relevant radionuclides from ILW wastes, disposed of in a chemically reducing, cementitious environment.

From a chemical point of view, the pore waters of hydrated cement matrices provide an exceptional environment. Compared with usual ground waters exhibiting pH-values of around 8, cement pore waters are strongly alkaline with pH-values from 12.5 to 13.5 and contain nearly no carbonate and only little sulfate. Oxides and hydroxides mainly determine solubility and speciation of the elements.

Solubility and speciation calculations in cementitious pore waters were performed using the very recently updated Nagra/PSI Chemical Thermodynamic Data Base (TDB) for the majority of the 36 elements addressed as potentially relevant. Wherever possible, maximum concentrations compiled in this report were based on geochemical calculations. In order to ensure full traceability, all thermodynamic data not included in the TDB are explicitly specified in the document. For similar reasons the compilation of results (Table 1) clearly distinguishes between calculated and recommended items. The heading "CALCULATED" lists maximum concentrations based on data fully documented in the TDB; results under the heading "RECOMMENDED" include data from other sources.

The pH sensitivity of the results was examined by performing calculations at pH 13.4, in accordance with the pH of non-altered cement pore water. Solubility increases predominantly for elements that tend to form anionic hydroxide complexes (Sn, Pd, Zr, Ni, Eu, Cd, Mo, Co). Oxidizing conditions around +350 mV might be expected in the environment of nitrate-containing wastes. In this case, significant solubility increases were calculated for U, Np, Pu, Se and Ag.

Special attention is allocated to the uncertainties of the evaluated maximum concentrations, expressed as upper-and lower limits. The conceptual steps to determine these uncertainties are explained in the section 3. Due to lack of data, it was not always possible to assess uncertainties in a manner consistent with that used to assess the solubility limits. For some elements, uncertainties had to be derived from less sharply defined data or even with the help of estimates. Such less rigorous approaches are justified by the fact that in performance assessments particularly the upper limits are as important as are the maximum concentrations themselves. However, appropriate information was available to define an upper limit for nearly all of the relevant nuclides.

Technical Report NTB 02-21

Glass Dissolution Parameters: Update for Entsorgungsnachweis

Summary

This document provides updated long-term corrosion rates for borosilicate glasses used in Switzerland as a matrix for high-level radioactive waste. The new rates are based on long-term leaching experiments conducted at PSI and are corroborated by recent investigations. The asymptotic rates have been determined through weighted linear regressions of the normalised mass losses, directly calculated from B and Li concentrations in the leaching solutions. Special attention was given to the determination of the analytical uncertainty of the mass losses. The sensitivity of the corrosion rates to analytical uncertainties and to other criteria (e.g. the choice of data points for the regressions) was also studied. A major finding was that the uncertainty of the corrosion rate mainly depends on the uncertainty of the specific glass surface area. The reference rates proposed for safety assessment calculations are 1.5 mg m^-2 d^-1 for BNFL glasses and 0.2 mg m^-2 d^-1 for COGEMA glasses.

The relevance of the proposed corrosion rates for repository conditions is shown based on the analysis of processes and parameters currently known to affect the long-term kinetics of silicate glasses. Specifically, recent studies indicate that potentially detrimental effects, notably the removal of silica from solution through adsorption on clay minerals, are transitory and will not affect the long-term corrosion rate of the Swiss reference glasses. Iron corrosion products are also known to bind silica, but present data are not sufficient to quantify their influence on the long-term rate.

Technical Report NTB 02-20

Cementitious Near-Field Sorption Data Base for Performance Assessment of an ILW Repository in Opalinus Clay

Summary

The present report describes a cement sorption data base (SDB) for the safety-relevant radionuclides to be disposed of in the planned Swiss repository for long-lived intermediate-level radioactive wastes (ILW). This report is an update on earlier SDBs, which were compiled for the cementitious near field of a repository for low-and intermediate-level radioactive wastes (L/ILW) by BRADBURY & SAROTT (1995) and BRADBURY & VAN LOON (1998).

The radionuclide inventories are determined by the waste streams to be disposed of in the ILW repository. A list of the safety-relevant radionuclides was provided based on the currently available information on ILW inventories. The compositions of the cement porewaters in the near fields of the L/ILW and ILW repositories, which had been calculated using well-established codes for modelling cement degradation, were compared to identify any differences in the near-field conditions and to assess their influence on radionuclide sorption.

Sorption values were selected based on the previously reported SDBs for the near field of the L/ILW repository. Sorption values were revised if new information and/or data were available which allowed changes to or re-appraisals of the data to be made. The sorption values recommended in this report were either selected on the basis of data from in-house experimental studies or from literature data.

For some key radioelements, i.e., Cs(l), Sr(II), Ni (II), Eu(III), Th(IV) and Sn(IV), new data were available from in-house measurements. These elements had been selected for experimental studies due to their relevance to safety assessment and/or their importance as appropriate chemical analogues.

Degradation products of bitumen and cellulose, concrete admixtures and cement-derived near-field colloids were taken into account as the main potential perturbations, which could reduce radionuclide sorption in the near field. Possible impacts of the perturbing factors on radionuclide mobility were considered and quantified in terms of sorption reduction factors.

Technical Report NTB 02-19

Far-Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

Summary

An Opalinus Clay formation in the Zücher Weinland is under consideration by Nagra as a potential location for a high-level and long-lived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations.

In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/PCO₂. In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases.

The basis for the sorption data bases is "in-house" sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zücher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation.

Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (R_d) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature R_d values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting R_d values were then further modified using so-called Lab – Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions.

Finally, it is important to have some appreciation of the uncertainties associated with the selected R_d values in the sorption data bases. An attempt has been made in the report to move away from the unsatisfactory "realistic/conservative" terminology and to develop a procedure for estimating uncertainty in a transparent and traceable manner.

Technical Report NTB 02-18

Near-Field Sorption Data Bases for Compacted MX-80 Bentonite for Performance Assessment of a High-Level Radioactive Waste Repository in Opalinus Clay Host Rock

Summary

Bentonites of various types and compacted forms are being investigated in many countries as backfill materials in high-level radioactive waste disposal concepts. Nagra is currently considering an Opalinus clay (OPA) formation in the Zürcher Weinland as a potential location for a high-level radioactive waste repository. A compacted MX-80 bentonite is foreseen as a potential backfill material. Performance assessment studies will be performed for this site and one of the requirements for such an assessment are sorption data bases (SDB) for the bentonite near-field. The purpose of this report is to describe the procedures used to develop the SDB.

One of the pre-requisites for developing a SDB is a water chemistry for the compacted bentonite porewater. For a number of reasons mentioned in the report, and discussed in more detail elsewhere, this is not a straightforward task. There are considerable uncertainties associated with the major ion concentrations and in particular with the system pH and Eh. The MX-80 SDB was developed for a reference bentonite porewater (pH = 7.25) which was calculated using the reference OPA porewater. In addition, two further SDBs are presented for porewaters calculated at pH value of 6.9 and 7.9 corresponding to lower and upper bound values calculated for the range of groundwater compositions anticipated for the OPA host rock.

"In house" sorption isotherm data were measured for Cs(I), Ni(II), Eu(III), Th(IV), Se(lV) and I(-I) on the "as received" MX-80 material equilibrated with a simulated porewater composition. Complementary "in house" sorption edge and isotherm measurements on conditioned Na/Ca montmorillonites were also available for many of these radionuclides. These data formed the core of the SDB. Nevertheless, some of the required sorption data still had to be obtained from the open literature. An important part of this report is concerned with describing selection procedures and the modifications applied to the chosen values so that they are compatible with the reference mineralogy and porewater chemistries.

The SDB comprises of distribution ratios (R_d) obtained from batch sorption type measurements made on dispersed systems. It is not intrinsically evident that these values are valid for compacted systems as required in the performance assessment. Arguments justifying this Lab to Field transformation are presented in a separate report and the main conclusions are summarised here.

Finally, an attempt is made to assess the uncertainties associated with the selected distribution ratios in the SDB.

Nagra is considering a scenario where oxidising conditions exist in the near-field of the compacted bentonite surrounding spent fuel. In such a case the MX-80 porewater is considered to have the same composition as that in the reference case (pH = 7.25), but with a redox potential (Eh) of +635 mV. Tc, Se, U, Np and Pu have been identified as the only safety relevant radionuclides which will have redox states different from those in the reference reducing case. Sorption values for the above radionuclides are presented in the Appendix.

Technical Report NTB 02-17

A Comparison of Apparent Diffusion Coefficients Measured in Compacted Kunigel V1 Bentonite with those Calculated from Batch Sorption Measurements and De (HTO) Data: A Case Study for Cs(I), Ni(II), Sm(III), Am(III), Zr(IV) and Np(V)

Summary

Recently, a bentonite sorption data base, comprising values taken from batch sorption data, was developed for a performance assessment study for high­level 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(ll), Sm(III), Am(lll), 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 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.

Technical Report NTB 02-16

NAGRA/PSI Chemical Thermodynamic Data Base 01/01

Summary

The Nagra/PSI Chemical Thermodynamic Data Base has been updated from version 05/92 to 01/01 to support an ongoing safety assessment of a planned Swiss repository for high-level radioactive waste. Database version 05/92 distinguished between “core data” and “supplemental data”. Core data are for elements commonly found as major solutes in natural waters. These data are well established and have not been changed to any significant degree in this update. Supplemental data comprise actinides and fission products, Mn, Fe, Si and Al. Our update from version 05/92 to 01/01 involved major revisions for most of the supplemental data. Altogether, more than 70% of our database contents have been updated.

Data for U, Np, Pu, Am and Tc recommended by the internationally recognised NEA TDB project were considered in our update. Our reasons for not accepting several NEA recommendations are documented in detail. Thermodynamic data for Th, Sn, Eu, Pd, Al, and solubility and metal complexation of sulphides and silicates were extensively reviewed. Data for Zr, Ni and Se were examined less rigorously as these elements are currently being reviewed in phase II of the NEA TDB project.

Our experiences from this two year team effort can be summarised as follows. (1) Detailed in-house reviews and critical appraisal of NEA recommendations greatly improved the chemical consistency and quality of the selected data, as shown e.g. by comparison of complexation constants for M(III) and M(IV) oxidation states of actinides and fission products. (2) On the other hand, we could discern major gaps in the data, especially missing carbonate complexes. (3) In some systems, e.g. ThO₂ – H₂O and UO₂ – H₂O, experimental data could not be described by a unique set of thermodynamic constants. There, a pragmatic approach based on solubility data was chosen to provide data for application to performance assessment.

Technical Report NTB 02-15

Diffusion of Tritiated Water (HTO) and 22Na+-ions through Non-degraded Hardened Cement Pastes – II. Modelling Results

Summary

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 high­porous 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 tracer­deposition mechanisms playing a crucial role in diffusion experiments using cementitious materials.

Technical Report NTB 02-14

Stability and Mobility of Colloids in Opalinus Clay

Summary

The Opalinus Clay formation in northern Switzerland is currently under evaluation for its suitability as a host rock for a spent fuel, vitrified high-level waste and long-lived intermediate level waste repository. The Swiss concept of geological disposal of radioactive waste is based on a multi-barrier system. In the near field, spent fuel and vitrified high-level radioactive waste are contained in massive steel canisters surrounded by a dense bentonite clay barrier. The long-lived intermediate-level waste, on the other hand, is surrounded by cementitious materials in separate emplacement tunnels. In the far field, the host rock formation is expected to act as an effective barrier for radionuclide migration. In this report, we discuss the potential role of mobile colloidal particles in facilitating radionuclide transport through the Opalinus Clay formation (far field). Since rock fractures resulting in preferential flow paths are not expected in Opalinus Clay, we limit the discussion to convective-diffusive transport through the rock matrix.

Numerous reports in the literature have suggested that mobile colloidal particles in subsurface porous media may serve as carriers for strongly sorbing contaminants, such as many radionuclides, and thereby facilitate contaminant migration. The objective of this report is to (i) discuss the potential composition of colloidal particles in Opalinus Clay, (ii) evaluate the colloidal stability of colloids in Opalinus Clay under consideration of the pore water chemistry, (iii) discuss the potential mobility of colloidal particles in Opalinus Clay under various assumptions concerning flow conditions and colloid-matrix interactions.

The most relevant types of colloids in Opalinus Clay were inferred from the rock composition and from reference pore water chemistry. These include clay minerals, quartz, calcite, iron oxides, and organic matter. An evaluation of published data on surface charge and colloidal stability of these types of colloidal particles suggests that they would readily aggregate in Opalinus Clay pore water, which has a high ionic strength (~0.1-0.3 M) and neutral to slightly alkaline pH (~6.8-8.2).

Based on the texture of Opalinus Clay and hydrologic parameters, simple diffusion, advection, and colloid filtration calculations were conducted. Colloidal mineral particles larger than ∼ 60 nm are expected to be rather immobile, since they settle by gravitational forces. Also, the mesoporous structure of Opalinus Clay is likely to limit their mobility by physical straining. Colloidal particles smaller than ∼ 60 nm are expected to be removed very effectively from the pore water by diffusion and deposition to immobile matrix surfaces. Maximum travel distances (99.99 % removal) of colloidal particles, as calculated by filtration theory, are therefore below 1 m even under worst-case assumptions. In summary, the mobility of colloids in Opalinus Clay is expected to be extremely low due to the following factors: (i) high ionic strength of pore water and resulting low colloidal stability, (ii) mesoporous structure of Opalinus Clay and resulting filtration by physical straining, (iii) extremely low hydraulic conductivity and advective pore water velocity in Opalinus Clay, (iv) much slower diffusion of colloidal particles compared with dissolved radionuclide species.

In conclusion, Opalinus Clay contains various types of colloidal size particles, but their colloidal stability and mobility is expected to be extremely low due to both chemical and physical factors. Thus, colloid-facilitated transport of radionuclides in Opalinus Clay is unlikely, as long as fracture flow does not occur. If further hydrogeologic studies reveal the relevance of fracture flow, the role of colloids needs to be re-evaluated in this con text. The composition of Opalinus Clay organic carbon and dissolved organic carbon and its potential role in radionuclide migration through Opalinus Clay requires additional research.

Technical Report NTB 02-13

Redox Conditions in the Near Field of a Repository for SF/HLW and ILW in Opalinus Clay

Summary

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 highlevel 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.

SF/HLW

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 H₂(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. 

ILW

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. 

Technical Report NTB 02-12

Application of the Nagra / PSI TDB 01/01: Solubility of Th, U,
Np and Pu

Summary

If a true thermodynamic equilibrium with a well-known solid has established, chemical equilibrium thermodynamics allows estimation of the maximum concentration of a given radionuclide in a specified pore fluid of an underground repository. In the course of the review process for the Nagra/PSI Chemical Thermodynamic Data Base 01/01 we identified important cases of insufficient chemical knowledge leading to gaps in the database. First, experimental data for the systems ThO₂ – H₂O and UO₂ – H₂O cannot be interpreted by a unique set of thermodynamic constants. There we chose a pragmatic approach by including parameters in the database that are not thermodynamic constants in a strict sense, but that reproduced relevant experimental observations. Second, potentially important thermodynamic constants are missing because of insufficient experimental data. Estimations of these missing constants led to problem specific database extensions. Especially constants for ternary hydroxide-carbonate complexes of tetravalent actinides have been estimated by the “backdoor approach”, i.e. by adjusting thermodynamic constants to maximum feasible values that are still consistent with all available experimental solubility data. We conclude that gaps in the Nagra/PSI TDB 01/01 concerning Th, U, Np and Pu data do not lead to grossly wrong estimates of their respective solubility limits in the case of a bentonite pore water defined for the planned Swiss high-level waste repository. Attempts to improve our chemical knowledge should concentrate on the key parameters identified in this study.

Technical Report NTB 02-11

Canister Options for the Disposal of Spent Fuel

Summary

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.

Technical Report NTB 02-10

Project Opalinus Clay
Radionuclide concentration limits in the near-field of a repository for spent fuel and vitrified high-level waste

Summary

The disposal feasibility study currently performed by Nagra includes a succession of quantitative models, aiming at describing the fate of radionuclides potentially escaping from the repository system. In this chain of models the present report provides the so called "solubility limits" (maximum expected concentrations) for safety relevant radionuclides from SF/HLW wastes, disposed of in a reducing clay (Opalinus Clay, bentonite) environment.

Solubility and speciation calculations in bentonite pore waters were performed using the very recently updated Nagra/PSI Chemical Thermodynamic Data Base (TDB) for the majority of the 37 elements addressed as potentially relevant. Particularly for the most relevant actinides, the straightforward applications with this updated TDB yielded results in contradiction to chemical analogy considerations. This was a consequence of incomplete data and called for problem specific TDB extensions, which were evaluated in a separate study. However, a summary of these problem specific extensions is provided in section 4.1.

The results presented in this report solely depend on geochemical model calculations. Thus, it is of utmost importance that the underlying data and assumptions are made clear to the reader. In order to ensure traceability, all thermodynamic data not included in the Nagra/PSI TDB are explicitly specified in the report, in order to provide complete documentation for quality assurance and for comprehensibility.

In order to clearly distinguish between results derived from data carefully reviewed in the Nagra/PSI TDB and those calculated from "other" data, the summary of expected maximum concentrations provided in Table 1 includes two columns. The heading CALCULATED provides maximum concentrations based on data fully documented in the updated TDB, whereas maximum concentrations, which include additional problem specific data and/or data from other sources, are given under the heading RECOMMENDED.

The present study also pays specific attention to the uncertainties of evaluated maximum concentrations and represents them as lower- and upper limits. The conceptual steps for deriving uncertainties are briefly outlined in section 3. Mainly due to lack of data/knowledge, it was not always possible to assess uncertainties in a manner consistent with that used to assess maximum concentrations. In a number of cases it was necessary to rely on less well traceable information or even on estimates and/or "expert judgement" to provide uncertainties. This less rigorous approach is justified by the fact that uncertainties (particularly the upper limits) are deemed as important as the maximum concentrations themselves. Although (upper) limits are specified wherever adequate information was available, no uncertainties at all could be derived for a few elements.

A specific class of variabilities arises from major uncertainties in the underlying chemical system, particularly from lack of knowledge of the partial pressure of CO₂. This class of uncertainties is visualised as "solubility" vs. "pCO₂" diagrams when appropriate. Limits were derived either from maximum values of thermodynamic uncertainties or from chemical system variabilities.

Another major uncertainty concerns the definition of redox conditions in the underlying chemical system. Although not really expected to occur, an oxidising environment instead of reducing conditions could establish in the vicinity of the disposed of wastes. In the sense of a "what-if" study, section 5 provides model calculations for the redox sensitive elements Pu, Np, U, Tc, Se and Sb, performed under oxidising conditions.

Technical Report NTB 02-09

Assessment of Porewater Chemistry in the Betonite Backfill for the Swiss SF/HLW Repository

Summary

The porewater chemistry in the bentonite backfill will strongly affect the mobility of radionuclides via sorption and solubility equilibria. The aim of this study was to derive a reference bentonite porewater composition for the Swiss high-level waste repository and to estimate compositional uncertainties. Special emphasis was put on the evaluation of acid-base buffering mechanisms and the related variables pH and pCO₂.

Recent experimental data at high solid/water ratios are analysed by means of a classical aqueous solution - mineral equilibrium model using two distinct sets of constants for sorption reactions. An optimised thermodynamic model is then applied to repository conditions. A sensitivity analysis is then performed to identify critical geochemical parameters and to quantify their effect on porewater chemistry. The evolution of porewater chemistry in time is studied with the help of two alternative models, one based on water-exchange cycles, the other relying on diffusion. For the derivation of the reference porewater, a redox model based on the equilibrium between magnetite and dissolved Fe²⁺ is integrated in the clay/water reaction model. Uncertainties related to pH, Eh, and major anion concentrations (Cl^-, SO₄ ^2-, CO₃ ^2-) are evaluated. Finally, available trace element data are presented.

Two limiting bentonite porewater compositions with pCO₂ fixed at 10^-3.5 and 10^-1.5 bar have been modelled, which define the pH and Eh boundaries specified below. A third, intermediate composition, calculated with pCO₂ fixed at 10^-2.2 bar, is considered to reflect the most probable repository conditions, and is thus defined as reference bentonite porewater. The resulting three waters are Na-(Ca-Mg-)-Cl-(SO₄) dominated and have a ionic strength of about 0.3 M.

The main results of this modelling study can be summarised as follows:

Technical Report NTB 02-08

The Uptake of Eu(III) and Th(IV) by Calcite under Hyperalkaline Conditions: The Influence of Gluconic and Isosaccharinic Acid

Summary

Calcite is an important component of Valanginian marl, a potential host rock for a low and intermediate level radioactive waste (UILW) repository in Switzerland. This mineral also forms an important component of the disturbed zone around a repository, as it remains largely unaffected by the hyperalkaline waters migrating out of the cementitious repository.

The sorption behaviour of Eu(lIl) and Th(IV) on Merck calcite in an artificial cement pore water (ACW) at pH 13.3 has been studied in batch-type sorption experiments. In addition, the effect of α-isosaccharinic acid (ISA) and gluconic acid (GLU) on the sorption of these two cations has been investigated.

In the absence of ISA and GLU, a strong interaction of Eu(lIl) and Th(IV) with Merck calcite was observed. Eu(lll) and Th(IV) sorption kinetics were fast and the isotherms indicated a linear adsorption behaviour over the experimentally accessible concentration range. In the case of Eu(III), a decrease of the R_d value with increasing solid to liquid (S:L) ratio was observed indicating that, along with adsorption, other processes might influence the immobilisation of this cation by Merck calcite under ACW conditions. In the case of Th(IV), however, changes in the S:L ratio had no effect on the sorption behaviour.

High ISA and GLU concentrations in solution significantly affected the sorption of both Eu(lIl) and Th(IV): R_d values for Eu(lIl) decreased significantly at ISA concentrations higher than 10^-5 M and at GLU concentrations higher than 10^-7 M. The sorption of Th(IV) was reduced at ISA concentrations above 2·10^-5 M and at GLU concentrations above 10^-6 M.

The effects of ISA and GLU on the immobilisation of Eu(lIl) and Th(IV) were interpreted in terms of complex formation in solution. In the case of Eu(lIl) the metaIligand complexes were found to have a 1:1 stoichiometry. Complexation constants derived for the aqueous Eu(III)-ISA' and Eu(III)-GLU complexes were determined to be logß⁰_EuISA = -31.1±0.2 and logß⁰_EuGLU = -28.7±0.1.

In the case of Th(IV) it was assumed that a Th(IV) - ISA - Ca complex and a Th(IV) GLU - Ca complex were formed both having a 1:2:1 stoichiometry. The complexation constants for these two complexes were determined to be logß⁰_ThISA = -5.0±0.3 and logß⁰_ThGLU = -2.14±0.01 .

Assuming that the concentrations of ISA and GLU in the pore water of the disturbed zone are similar to the maximum concentrations estimated for the cement pore water in the near-field of the repository, i.e. 10^-5 M ISA and 10^-7 M GLU, then the formation of aqueous complexes with ISA or GLU would not significantly affect Eu(lIl) and Th(IV) sorption on calcite. 

Technical Report NTB 02-07

Partitioning of Radionuclides in Swiss Power Reactor Fuels

Summary

The potential preferential release of some fission and activation products from spent nuclear fuel into porewater after canister breaching in a deep repository in Switzerland is discussed. Data from studies of fission gas release from UO₂ and mixed oxide (MOX) fuels that are representative of fuel from Swiss nuclear power reactors are used to estimate the average fission gas release (FGR) for spent fuel. The evaluations are performed for average burnup UO₂ and MOX fuel (48 GWd/tIHM) as well as for higher burnup fuels (MOX 65 and UO₂ 55, 65 and 75 GWd/tIHM). For the case of UO₂ fuel, the estimates include an analysis of the fraction of FGR present in the rim region, which is particularly important at higher burnup. Information from a variety of leaching studies on LWR fuel are then reviewed and compared to FGR as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction or IRF) upon breaching of the fuel cladding. For higher burnup fuels, where leaching data are largely unavailable, IRF values are based on extrapolations of FGR behaviour observed at lower burnup and on an assessment of the impact of fuel restructuring at higher burnup on FGR. The IRF thus includes releases from the fuel/cladding gap and grain boundaries, as well as from the rim region. The expected release rates of radionuclides from Zircaloy cladding are also discussed, based on review of low temperature corrosion and radionuclide release data relevant to repository conditions.

Technical Report NTB 02-06

Project Opalinus Clay:
Models, Codes and Data for Safety Assessment
Demonstration of disposal feasibility for spent fuel, vitrified high-level waste and long-lived intermediate-level waste (Entsorgungsnachweis)

Summary

The present report is designed to provide readers with the necessary information to perform, if they so wish, independent checks of the results of the evaluation of the "assessment cases" described in the Safety Report (Nagra 2002c) for Project Entsorgungsnachweis. It also describes the conceptual models and corresponding codes for the near field, geosphere and biosphere that were used in the safety assessment to evaluate the assessment cases, including the reasons why they are considered adequate for their intended purposes, and the operational elements and procedures that were used to manage the required large number of calculations. Models and codes not described in this report are the supporting models used to derive parameter values for the near field, geosphere and biosphere codes, and to support model assumptions. These include, for example, groundwater flow models, mechanistic models of sorption, temperature evolution models, waste dissolution models, etc., and are described in the Project Entsorgungsnachweis reference reports.

An assessment case is a specific set of assumptions regarding the broad evolution of the repository and its environment, the conceptualisation of individual features, events and processes (FEPs) relevant to the fate of radionuclides within the disposal system and the parameters used to describe these FEPs. In the safety assessment, a broad range of assessment cases is analysed in order to illustrate the impact of various detrimental FEPs and uncertainties on the level of safety provided by the disposal system. The assessment cases are defined, the underlying reasoning documented and the results of their analysis presented in the Safety Report. In the interests of transparency, the Safety Report presents these descriptions in a mainly qualitative fashion, without exhaustively documenting all relevant formulae and data. The present report complements the Safety Report with a comprehensive description of models, codes and data and thus provides traceability within the safety assessment. The two reports together satisfy the assessment principle (see Chapter 2 of the Safety Report) that the development of the safety case and its results should be documented in manner that provides both transparency and traceability.

In the safety assessment for Project Entsorgungsnachweis, assessment cases are divided into a number of groups, according to the issues or uncertainties that they address. The main part of the present report focuses, in turn, on groups of assessment cases that explore:

Overviews are given of the conceptual models underlying the assessment cases and the various assumptions and simplifications that are made in order to arrive at sets of mathematical equations and input parameters so that each case can be evaluated using corresponding computer codes, with references to appendices that give detailed descriptions of the codes and the equations that they solve, as well as tables containing the data used and the data source. In the appendices, descriptions are given of the capabilities of each code in terms of the phenomena (or "Super-FEPs") that they address. This is important within the FEP management procedure to allow an evaluation of the suitability of a given code to address a given Super-FEP (Nagra 2002d). It is shown that the codes used are sufficiently versatile to evaluate all the required assessment cases. There are a few safety-relevant phenomena that the codes are not "qualified" to evaluate, but, in all cases, either significant effects can be ruled out by supplementary studies (e.g. for criticality), effects are intrinsically favourable to safety and can be conservatively neglected (e.g. transport resistances in the SF / HLW near field), or parameters (e.g. canister breaching time) can be chosen to ensure that calculations err on the side of pessimism.

In addition to the models and codes used to analyse the assessment cases, simplified insight models are used to examine particular aspects of system performance and sensitivity to key system properties and model assumption. The insight models are described in Chapter 9.

The deterministic evaluation of assessment cases is complemented by probabilistic calculations to build further system understanding and, in particular, to indicate the performance of the system for parameter combinations not analysed by deterministic calculations. The computational tool used to sample input parameters from probability density functions (PDFs) and the PDFs themselves are described in Appendices 2 and 3, respectively.

Except for a few cases, no justifications are given in the present report for the model assumptions made and no final or intermediate results are presented, except where this is considered to help understanding. For a full presentation of results and a justification of assumptions, the reader is referred to the Safety Report and the reference reports.

It is, however, noted that all assessment cases are evaluated using, to some extent, pessimistic or conservative conceptual assumptions, parameters and model simplifications, resulting in calculated doses that should be regarded as upper bounds for, rather than predictions of, expected doses. To assess the degree of bias introduced by such assumptions, the modelling approaches used to evaluate each assessment case are examined systematically in Chapter 10. It first summarises the various assumptions and simplifications made in order to arrive at sets of mathematical equations and input parameters so that each case can be quantitatively evaluated, and then it assesses these assumptions and simplifications in terms of their degree of realism, pessimism or conservatism. This information is then used as a tool to systematically assess the bias that has been introduced into the models ("bias audit").

Technical Report NTB 02-05

Project Opalinus Clay
Safety Report
Demonstration of disposal feasibility for spent fuel, vitrified high-level waste and long-lived intermediate-level waste (Entsorgungsnachweis)

Summary

12 A4 pages (see above for download)

Technischer Bericht NTB 02-03

Projekt Opalinuston
Synthese der geowissenschaftlichen Untersuchungsergebnisse – Entsorgungsnachweis für abgebrannte Brennelemente, verglaste hochaktive sowie langlebige mittelaktive Abfälle

Summary

7 A4 pages (see above for download)

Technischer Bericht NTB 02-02

Projekt Opalinuston
Konzept für die Anlage und den Betrieb eines geologischen Tiefenlagers – Entsorgungsnachweis für abgebrannte Brennelemente, verglaste hochaktive sowie langlebige mittelaktive Abfälle

Summary

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 (Nagra 1980) 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, longterm geological disposal" as formulated in the most recent requirements of the authorities (EKRA 2000, KEG 2001)

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 (Nagra 2002a) and a safety assessment report (Nagra 2002b). 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:

Technical Report NTB 01-08

Porewater chemistry in compacted re-saturated MX-80 bentonite:
Physico-chemical characterisation and geochemical modelling

Summary

Bentonites of various types are being investigated in many countries as backfill materials in high-level radioactive waste disposal concepts. Being able to understand the chemistry of the porewater in compacted bentonite, and the factors which influence it, is critical to the synthesis of sorption data bases and to predicting radionuclide solubilities, and hence to repository safety studies. However, quantification of the water chemistry in compacted bentonite is difficult because reliable samples for chemical analysis cannot be obtained even by squeezing at exceedingly high pressures.

In this report concepts are developed which are somewhat different from those used in previously published works on bentonite porewater. Considerations of the swelling properties of montmorillonite led to the proposition that there were, generally speaking, three types of water associated with re-saturated compacted bentonite. The water defined as the porewater is only a small fraction of the total. The porewater volume present in re-saturated bentonites having different initial dry densities was quantified using CI^- "through diffusion" data.

Highly compacted bentonite is considered to function as an efficient semi-permeable membrane so that re-saturation involves predominantly the movement of water molecules and not solute molecules. This implies that the composition of the external saturating aqueous phase is a second order effect. Consequently CI^- concentrations in the porewater could be calculated from the deduced porewater volume values and the measured CI^- inventory.

The pH of the porewater of a compacted bentonite is an extremely important parameter because of its influence on radionuclide solubility and sorption. Arguments are presented in support of the thesis that the initial pH is fixed by the high buffering capacity afforded by the amphoteric ≡SOH sites. The pH of the porewater depends directly on the speciation of these sites i.e. the proportions of sites present as ≡SOH, ≡SOH₂⁺ and ≡SO^-. In the report it is explained how this speciation is determined by the preparation process in the "as received" powder.

As a consequence of the high cation exchange capacity of montmorillonite, the large mass of montmorillonite in relation to the small porewater volumes in a highly compacted re-saturated bentonite, the major ion composition in the porewater will be controlled by the montmorillonite and the other solid phases present and will be very strongly buffered.

The above considerations are used in conjunction with detailed physico-chemical characterisation studies on MX-80 (Appendix) to calculate initial porewater compositions in compacted bentonites.

For the MX-80 material specified, the porewaters calculated for initial dry densities between 1200 and 1600 kg m^-3 had relatively high ionic strengths (0.3 to 0.33 M), similar cation concentrations and a pH equal to 8.0. The porewaters changed from being Na₂SO₄ rich at 1200 kg m^-3 to a NaCI/ Na₂SO₄ type water at 1600 kg m^-3.

Technical Report NTB 01-07

Water-extractable Organic Matter from Opalinus Clay: Effect on Sorption and Speciation of Ni(II), Eu(III) and Th(IV)

Summary

The purpose of the present study is to characterise water-extractable organic matter from Opalinus clay (OPA) with respect to possible complexing properties. OPA samples obtained from the Mont Terri rock laboratory and from the Benken site were used. The effect of organic matter extracted from these samples on the sorption of Ni(II), Eu(lIl) and Th(IV), representing bi-to tetravalent elements, on a commercial cation exchange resin was studied at pH ~8. The solid-liquid distribution coefficients were compared to those measured for synthetic waters, which are similar in composition to the aqueous extracts, but contain no organic matter. Within the range of estimated uncertainties, no difference in sorption is observed for most of cases. Only for a few extracts, slight reduction of sorption (less than a factor of 5) of Eu(lll) and Ni(lI) is found. Test experiments using small-molecular weight ligands and Aldrich humic acid show that the sensitivity of the ion exchange method is adequate at the specific conditions of the OPA extracts. The results of accompanying fluorescent spectroscopy experiments do not show any influence of the extracts on Cm(lll) speciation, which is dominated by carbonate complexes. This suggests that the reduction of sorption partly observed in the ion exchange experiments is not caused by the formation of complexes between the radionuclides and the organic matter in the extracts, but rather due to a possible underestimation of minor systematic uncertainties, such as unknown differences in the chemical composition between the extracts and the synthetic waters.

From these findings and from UV-VIS spectroscopic characterisation of the organic matter in the extracts, it can be concluded that only a small fraction of the organic matter may be present as humic or fulvic acids. The largest part of the organic matter are most probably either small-molecular weight molecules or macromolecules with a very low content in ligand sites.

The OPA samples tested are representative for the different facies in the OPA formations of Mont Terri and Benken. Therefore, the conclusions drawn for the influence of water-extractable organic matter on the sorption and speciation of radionuclides can safely be applied to any location within the OPA formation at Benken. The similarity of OPA formations at Mont Terri and Benken with respect to chemical behaviour is once more corroborated by this study. The latter location is considered as a potential site for the disposal of high-level and long-lived intermediate-level radioactive waste in Switzerland.

Technischer Bericht NTB 01-06

Optimierungsstudie für ausgewählte Abfalltypen aus Medizin, Industrie und Forschung

Summary

Für diesen NTB existiert keine Zusammenfassung

Technical Report NTB 01-05

Indications for self-sealing of a cementitious repository for lowand intermediate-level waste

Summary

Repositories for low and intermediate level nuclear waste contain large amounts of cementitious material. As a consequence of the interaction with formation waters, the cement will be degraded forming secondary minerals. The amount of precipitating secondary minerals depends on the chemical composition of the formation water. Furthermore, in the vicinity of the repository the hydraulic conditions and the parameters describing mass (radionuclide) transport will change with time during the cement degradation phase. As a result, porosity changes due to mineral and cement reactions will influence permeability and diffusivity: formation water rich in CO₂ will lead to calcite precipitation in the water conducting zones surrounding the cementitious waste repository and, therefore, will have an impact on the radionuclide release from the cementitious repository into the host rock environment.

Laboratory column experiments showed concurrent porosity and permeability changes during degradation of porous cement discs. However, very different quantitative results have been observed when CO₂-rich or pure water were used. The sequentially coupled flow, transport and chemical reaction code, MCOTAC, is used to include such observations in the modelling. A porosity-permeability and a porosity-diffusivity relation are used for describing cement degradation and related secondary mineral precipitation. For these complex coupled processes one-dimensional modelling has reached its limits of applicability. Therefore, two-dimensional model calculations are used to predict the temporal evolution of transport parameters for radionuclides within a “small scale” near-field of a cementitious waste repository. Mineral reactions influence hydraulic and transport parameters within such a near-field, causing reduced solute transport in the vicinity of the repository due to porosity and permeability changes at the rock-repository-interface. Also, the transport of radionuclides from the repository may be drastically reduced by porosity and permeability decreases. This is especially important for those radionuclides which show little or no sorption, since only the transport parameters (water flow velocity, dispersion and diffusion) will influence their migration behaviour. Within the “small scale” porous medium approach, coupling of chemical reactions and hydrodynamic parameters indicates a self sealing barrier at the host rock-repository interface for several scenarios. This barrier might persist for very long times and effectively contain radionuclides within the engineered repository system.

Technical Report NTB 01-04

Calculations of the Temperature Evolution of a Repository for Spent Fuel, Vitrified High-Level Waste and Intermediate Level Waste in Opalinus Clay

Summary

Thermal evolution is an important aspect of the performance of a repository for nuclear waste, because elevated temperatures affect many processes influencing the behaviour of engineered barrier systems and the host rock. The thermal evolution of a repository for spent fuel, high-level waste (HLW) and intermediate level waste (ILW) has been evaluated. The repository is assumed to be located at a depth of 650 m in the Opalinus Clay formation in northern Switzerland. The disposal system consists of spent fuel and HLW, encapsulated within steel canisters that are emplaced within horizontal tunnels, with the space between the canisters and the surrounding rock backfilled with bentonite. Waste emplacement tunnels are spaced 40 m apart. Concrete containers of ILW are placed in separate emplacement tunnels, with the void space around the containers filled with cementitious mortar.

The initial heat output of the waste and the rate of its decrease with time due to radioactive decay depends on the nature of the waste. For the case of spent fuel, some canisters contain only UO₂ assemblies, while others contain both UO₂ and MOX (mixed oxide) fuel assemblies. The initial heat output of both canister types is restricted to 1500 W/canister, but the time-dependent heat output of the UO₂/MOX canisters declines more slowly, as a result of significantly higher Pu content. For HLW canisters, the initial heat output is ~700 W and it decreases more rapidly than is the case for spent fuel, because of the low actinide content of HLW.

The temperature evolution of the engineered barrier system and surrounding rock is simulated using a finite-element model, using reference data for the thermal properties of spent fuel, HLW, bentonite backfill and Opalinus Clay. The results show that the surface temperatures of both the HLW and spent fuel canisters reach a maximum value of ~150°C within a few years after emplacement, although the HLW canister temperature decreases much more rapidly than for the case of spent fuel canisters. The temperatures within the bentonite depend strongly on the assumptions regarding its water content. For the expected low water inflow rates in a repository in low permeability Opalinus Clay, the bentonite has a low thermal conductivity and the temperature at the point midway between spent fuel canisters and the bentonite/host rock interface will reach a maximum of ~110°C. For HLW canisters, the mid-point of the bentonite will not exceed 100°C. Host rock temperatures remain below ~90°C at all times. In the unlikely event of a more rapid rate of inflow of water from the Opalinus Clay, the bentonite thermal conductivity is significantly greater and temperatures at the canister/bentonite interface and throughout the bentonite are markedly reduced.

For the case of ILW tunnels, a boundary element model is used to calculate temperature evolution, taking into consideration the hydration heat of the concrete mortar and radioactive decay. It is found that the maximum temperature reached within the engineered barrier system is ~50°C, about 12°C more than the ambient temperature in the Opalinus Clay at a depth of 650 m.

Technical Report NTB 01-03

Contaminant Transport in Fracture Networks with Heterogeneous Rock Matrices: The PICNIC Code

Summary

In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flowpaths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling.

To make use of these investigations the "PICNIC project" was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCHMD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales.

The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flowpath and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely different timescales. To account for one-dimensional matrix diffusion into homogeneous planar or cylindrical rock layers, analytical relations in the Laplace domain are used. To deal with one-dimensional or two-dimensional matrix diffusion into heterogeneous rock matrices, a finite-element method is embedded. The capability of the code for handling two­dimensional matrix diffusion is – to our knowledge – unique in fracture network modelling.

To ensure the reliability of the code, which merges methods from graph theory, Laplace transformation, finite-element methods, analytical and algebraic transformations and a convolution to calculate complex radionuclide transport processes over a large and diverse application range, implementation of the code and careful verification have been alternated for iterative improvement and especially the elimination of bugs. The internal mathematical structure of PICNIC forms the basis of the verification strategy.

The code is verified in a series of seven steps with increasing complexity of the rock matrix. Calculations for single nuclides and nuclide decay chains are carefully tested and analysed for radionuclide transport in single legs, in pathways and in networks. Different sources and boundary conditions are considered. Quantitative estimates of the accuracy of the code are derived from comparisons with analytical solutions, cross-comparisons with other codes and different types of self-consistency tests, including extended testing of different refinements of the embedded finite­element method for different rock matrix geometries. The geosphere barrier efficiency is a good single indicator of the code accuracy. Application ranges with reduced accuracy of the code are also considered.

For one-dimensional matrix diffusion into homogeneous and heterogeneous rock matrices, cross­comparisons with other codes are performed. For two-dimensional matrix diffusion, however, no code for cross-comparison is available. Consequently, the verification for these geometries relies chiefly on the verification for one-dimensional matrix diffusion, on qualitative estimates and on different self-consistency tests. The steady-state release for a single nuclide is additionally verified quantitatively.

PICNIC has been verified as far as possible at present to allow application with confidence in performance assessment and in modelling of transport experiments. It is shown that structural geological information on small-scale heterogeneity can be entered easily into PICNIC. It is explained, e.g. that considering two-dimensional matrix diffusion into the layer of altered wallrock adjacent to open channels in the cataclastic zone can strongly increase the performance of the geosphere for migrating radionuclides, depending however on the properties of the nuclides and the rock. Considering the effects of matrix diffusion into a second rock layer can also be highly beneficial.

Technical Report NTB 01-02

Experimental studies on the inventory of cement-derived colloids in the pore water of a cementitious backfill material

Summary

The potential rôle of near-field colloids for the colloid-facilitated migration of radionuclides has stimulated investigations concerning the generation and presence of colloids in the near-field of a repository for low- and intermediate level waste (L/ILW). The highly gas permeable mortar (Nagra designation: mortar M1) is currently favoured as backfill material for the engineered barrier of the planned Swiss L/ILW repository. The cementitious backfill is considered to be a chemical environment with some potential for colloid generation.

In a series of batch-style laboratory experiments the physico-chemical processes controlling the inventory of colloids in cement pore water of the backfill were assessed for chemical conditions prevailing in the initial stage of the cement degradation. In these experiments, backfill mortar M1 or quartz, respectively, which may be used as aggregate material for the backfill, were immersed in artificial cement pore water (a NaOH/KOH rich cement fluid). Colloid concentrations in the cement pore water were recorded as a function of time for different experimental settings. The results indicate that a colloid-colloid interaction process (coagulation) controlled the colloid inventory. The mass concentration of dispersed colloids was found to be typically lower than 0.02 ppm in undisturbed batch systems. An upper-bound value was estimated to be 0.1 ppm taking into account uncertainties on the measurements.

To assess the potential for colloid generation in a dynamic system, colloid concentrations were determined in the pore water of a column filled with backfill mortar. The chemical conditions established in the mortar column corresponded to conditions observed in the second stage of the cement degradation (a Ca(OH)₂-controlled cement system). In this dynamic system, the upper-bound value for the colloid mass concentration was estimated to be 0.1 ppm.

Implications for radionuclide mobility were deduced taking into account the experimental results of the study. Predictions of the colloidal effect on radionuclide mobility were based on the assumption that colloids dispersed in the pore water of the backfill material may reduce radionuclide sorption (R_d values) on cement. This effect was described in terms of a sorption reduction factor. The distribution ratios (R_c) of radionuclides between the cement pore water and the colloidal phase as well as the colloid mass concentration (m_c) are the two important colloidal parameters affecting sorption reduction. No significant sorption reduction is expected for weakly and moderately sorbing radionuclides (R_d ≤ 1 m³ kg^-1) up to a colloid concentration of 1 ppm. Moreover, no significant sorption reduction is anticipated for strongly sorbing radionuclides (R_d > 1 m³ kg^-1) below colloid concentrations of 0.1 ppm. This value is considered to be representative for the backfill material. At higher colloid concentrations, however, sorption reduction may occur in case of the strongly sorbing radionuclides. Nevertheless, due to an extremely strong uptake of these elements by cement in the absence of colloids, the effective sorption values in the presence of colloids are predicted to be still high.

Technical Report NTB 01-01

Model Radioactive Waste Inventory for Reprocessing Waste and Spent Fuel

Summary

This report describes a model inventory concerning spent fuel (SF), high level vitrified (“glass”) waste from reprocessing (HLW) and long lived intermediate (ILW) wastes. The inventory describes the conditioned and packaged SF, HLW and ILW that are expected to be produced by the 5 operational NPPs (3.2 GW(e) over 60 years (192 GWa(e)), along with a more general 300 GWa(e) scenario) arising from both reprocessed and unreprocessed spent fuel, including mixed oxide (MOX), obtained by the recycling of Uranium and Plutonium obtained from reprocessing. In addition, ILW from other sources are considered to be included in the overall uncertainties and not modelled in any detail.

The waste arisings are described by 17 waste sorts, each defined using an average waste package description. Radionuclide activities, other radiological characteristics, material content, specific properties and predicted maximum values of the various attributes of the packages are provided in a data base of the waste sorts. This data is presented and discussed, along with a short description of the origin of the raw waste.

The summation of the waste volumes, activities and materials gives an overview of the repository waste contents. Uncertainties associated with the repository inventory have been analysed. The results provide an indication of the possible additional waste volume arisings to be considered for repository planning.