The safety case for a deep geological repository in Switzerland: engineered barriers for high-level waste disposal

1 Introduction

In the Swiss disposal concept spent fuel (SF) and vitrified high-level waste from reprocessing (RP-HLW) are to be disposed of in the high-level waste (HLW) section of the deep geological repository (Smith, 2026). The HLW near field is a multicomponent system of engineered barriers surrounded by the geological barrier. The HLW multi-barrier system consists of the following elements according to Nagra’s current provisional design: the SF and RP-HLW matrices (Guillemot et al., 2026), the HLW disposal canisters, the buffer and the sealing system.

An impermeable canister for the disposal of HLW is a key component of all deep geological repository programmes. The canister is the only barrier within the disposal system that provides complete confinement of radionuclides. Consequently, the focus is often on the length of time during which the barrier function can be guaranteed. Canister lifetime requirements can vary significantly between the different national programs (Table 1). They typically depend on the waste inventory, the disposal concept, the characteristics of the near field and the geological environment, as well as relevant national regulations.

TABLE 1. Canister and buffer concepts proposed by various waste management organisations (WMO) (Muñoz et al., 2024).

The design of disposal canisters that can remain structurally intact for millennia as well as the demonstration of their lifetime remains a scientific and engineering challenge. Canisters will undergo a time-dependent degradation due to corrosion within the buffer surrounding the container. Studies of the corrosion behaviour of candidate canister materials for the disposal of HLW are ongoing since several decades on a national (Hesketh et al., 2023; Binns et al., 2023; Kursten et al., 2023; Koho et al., 2023; Diler et al., 2023) as well as on an international (Muñoz et al., 2024) level. The various approaches employed for modelling canister corrosion and predicting canister lifetime have been recently reviewed (King et al., 2024a; King et al., 2024b). A buffer material surrounding and protecting the disposal canisters is a common component in multi-barrier systems for the disposal of HLW. As can be seen in Table 1, most countries plan to use bentonite as buffer due to its beneficial properties, such as its inherently low hydraulic conductivity at full saturation in combination with its self-sealing ability. Furthermore, the high clay content of bentonite has a marked radionuclide sorption capacity thus contributing to radionuclide immobilisation. Of central importance is the provision of favourable environmental conditions supporting the longevity of the canister by providing a long-lasting, mechanically homogeneous, and chemically favourable interface, while also inhibiting microbial activity (Nagra, 2024a). Alternatively, cement is proposed as buffer in the Belgian concept, and, in the French case, no buffer is used at all.

The emplacement drifts, which in Nagra’s concept are constructed at about 900 m depth in Opalinus Clay, are supported by a cement-based tunnel liner system providing operational safety during the construction and emplacement phase. Various assessments have demonstrated that the mutual interactions between the bentonite and the liner, canister and bentonite as well as between the liner and Opalinus Clay are of limited extent and do not compromise long-term safety (Nagra, 2024b). During the emplacement process, HLW emplacement drifts are continuously backfilled and closed off with a sealing element comprised of pre-compacted bentonite blocks.

In the scope of this paper, the functionality and performance of the components of the HLW emplacement drifts in Nagra’s concept are illustrated and arguments provided as to why mutual interactions between the components do not compromise their functionality. Emphasis is placed on how this system contributes to the relevant safety functions such as (S2) complete containment of radionuclides for a period of time (S3) immobilisation, retention and slow release of radionuclides and (S4) the compatibility of the components of the repository system (Smith, 2026). Based on the system described here, spent fuel and RP-HLW can be safely disposed of in the HLW section of Nagra’s combined repository.

2 Description and evolution of the HLW near field

2.1 Operational phase

During the construction phase of the HLW repository section, according to the provisional concept considered for the general licence application, 900 m long dead-end HLW drifts will be excavated with a tunnel boring machine oriented in the direction of the maximum principal stress. They will have an excavation diameter in the order of 3.5–3.7 m. Prefabricated single-shell segmental liners consisting of pre-cast concrete tubbing elements will be installed during tunnelling. The liner is designed to provide stable conditions during construction and operation. The annular gap between the liner and the rock will be filled completely with a highly fluid grout (Nagra, 2024b). The internal diameter of the liner is about 2.80 m. The choice of excavation and lining method allows the degree of tunnel convergence to be limited and, as a consequence, the extent of the excavation disturbed zone (EDZ) will also be limited (Nagra, 2024c). The HLW emplacement drifts are placed at an appropriate distance from each other (about 40 m according to the current provisional design), such that thermal, hydraulic or chemical interactions between drifts are avoided. This also prevents the EDZs of adjacent drifts from overlapping.

The disposal canisters will be transported underground inside a protective shielding overpack and will be emplaced co-axially in the HLW emplacement drifts at intervals of about 3 m. Each canister will be placed on a pedestal of highly compacted MX-80 bentonite blocks (water content 18%, dry density 1.8 g cm^-3), and the remaining void spaces backfilled with highly compacted granular MX-80 bentonite (water content 5%, dry density >1.45 g cm^-3) (Figure 1). Together, the pedestal and granular bentonite are termed the bentonite buffer. Once all the disposal canisters have been emplaced and the drift has been backfilled, a V1-HLW seal (see Section 5) will be emplaced to close each drift.

2.2 Post-closure phase

The early period after emplacement is marked by the development of a thermal pulse extending radially around the HLW repository section originating from the radiogenic heat released by the waste (Guillemot et al., 2026). The temperature reaches a maximum of ∼ 137 °C at the surface of the disposal canisters (Nagra, 2024d). In the current concept, the maximum heat power of SF canisters is 1,500 W at the time of emplacement. The thermal pulse results in a pressure increase caused by different thermal expansion of porewater and the rock surrounding the HLW disposal area that dissipates after about 2,000 years (Nagra, 2024d). Due to the thermal pulse, the mineralogy of the concrete liner is also altered, and chemical interactions of limited extent take place with the bentonite buffer and Opalinus Clay (Nagra, 2024b). In the period up to around 1,000 years after closure, the temperature at the canister surface will decrease to ∼ 75 °C (Nagra, 2024d). The temperature will continue to decrease gradually and reach 50 °C after approximately 10,000 years.

In the early thermal phase, the thermal gradient combined with the high suction of the bentonite surrounding the canisters will lead to an increase of relative humidity and water saturation within the HLW emplacement drift. After resaturation of the EDZ and recovery to static formation pressure, the self-sealing capacity of the Opalinus Clay will cause a distinct reduction in the transmissivity of the EDZ fractures, and the EDZ will eventually approach an effective hydraulic conductivity that is no more than 1 – 2 orders of magnitude above the conductivity of the intact rock (Wüst et al., 2026).

The saturation of the bentonite leading to swelling has the effect that the bentonite is homogenised and the porosity redistributed. Furthermore, the bentonite density is sufficiently high to prevent, when saturated, microbial activity due to the low water activity, lack of physical space, and high swelling pressure. As porewater from the surrounding rock enters the backfilled emplacement drift, equilibration with bentonite minerals will occur, which will influence the species dissolved in the porewater. The bentonite porewater composition results from the reaction between the OPA porewater and MX-80 bentonite and is a relatively benign near-neutral solution (Nagra, 2024b). The saturated bentonite together with the Opalinus Clay create a low-permeable barrier surrounding the HLW canisters, where mass transport is dominated by diffusion (Nagra, 2024d).

Within HLW emplacement drifts, redox conditions will quickly become anoxic as several processes (diffusion of dissolved O₂ towards the Opalinus Clay, aerobic respiration, corrosion and mineral oxidation) consume the entrapped O₂. The production of radiolytic oxidants from gamma radiation interacting with water in the bentonite will also stop by the time the nearfield becomes saturated as the radiation field will have dropped to very low levels, leading to anoxic conditions which will prevail indefinitely (Nagra, 2024e).

In addition to the chemical evolution of the near field described above, the mechanical loads on the canister also evolve. Upon bentonite saturation, the swelling pressure and the porewater pressure will lead to normal stresses acting on the canister, in addition to the temporary thermally induced pressure. Furthermore, in situ stresses in the host rock are also present and their magnitude and anisotropy is larger than those arising from the near field. As design basis for the canisters and as input for the assessment of their lifetime we conservatively assume the maximum deviatoric stress state on the canister, arising from the minimum horizontal and vertical lithostatic stresses, i.e. 15.2 and 21.6 MPa respectively.

When the canister breaches, porewater ingress is possible. Once water comes in direct contact with SF, a small fraction of the total radionuclides inventory will be instantly released, while the vast majority will occur by congruent dissolution at a very slow rate (Guillemot et al., 2026). In the case of RP-HLW, upon contact of water at the glass surface a slow dissolution of the radionuclides evenly distributed within the glass matrix will initiate (Guillemot et al., 2026). Radionuclides from both waste types will be dissolved into the bentonite porewater, at concentrations that are limited by their corresponding solubility limits. From the HLW nearfield, the radionuclides which are not sorbed by the bentonite, will diffuse outwards in the Opalinus Clay (Nagra, 2024f).

3 Disposal canisters

Before emplacement, HLW will be encapsulated into long-lived disposal canisters. The primary safety function of the disposal canisters is the complete containment of radionuclides for a certain period of time. However, they also contribute to the immobilisation, retention, and slow release of radionuclides, as well as the compatibility between the various barriers, since they are designed not to be detrimental to the performance of the other engineered barriers and host rock. According to Swiss regulations (ENSI, 2020), the HLW disposal canisters must guarantee complete containment of the radionuclides for at least 1,000 years. This implies that they will not breach under the mechanical stresses and despite any degradation or corrosion phenomena occurring in the repository during this period. Nevertheless, the analyses presented hereafter demonstrate that the containment function provided by the current provisional canister design will remain for longer than 10,000 years.

3.1 Canister design and manufacturing

The SF canisters are cylindrical with an outer diameter of 110 cm and a length of approximately 5 m (Figure 2). They can accommodate 12 BWR SF assemblies or 4 PWR SF assemblies in an appropriately designed internal basket. The RP-HLW canisters have an outer diameter of 72 cm and a length of approximately 4.3 m, accommodating 3 RP-HLW flasks (Figure 2). The wall thickness of both canister types is 140 mm. Several possible canister materials have been and continue to be evaluated. The current provisional design assumes that the HLW disposal canisters made of low carbon steel. A low strength, bespoke alloy has been proposed, selected to enhance weldability and to minimise the effects of hydrogen on the material properties.

Each canister is made up of three main pieces: two lids, one at each end, and a cylindrical body. All pieces will be manufactured by forging. For operational safety, a gas-tight inner lid is also present. The bottom lid will be welded to the cylindrical body prior to waste emplacement and then stress relieved in a global furnace. After the HLW is placed in a disposal canister, the top lid will be welded in place, sealing the canister. Welding will be performed remotely using electron beam welding, followed by local post-weld heat treatment to relieve stresses. By combining thermal and mechanical modelling of the weld process and the post-weld heat treatment (PWHT), it is possible to calculate the stress field developed due to the weld heat cycle, followed by the stress relaxation during the heat treatment. Following this approach peak values of the axial stress of 82.4 and 100.1 MPa were calculated for SF and RP-HLW canisters respectively (Nagra, 2024e). After completion of the closure weld, quality assurance procedures consisting of ultrasonic and eddy current inspection will be employed to limit the number and size of weld flaws. For these inspection procedures, the detection limits are 4 × 15 mm for sub-surface flaws and 1 × 10 mm for surface-breaking flaws (Nagra, 2024e).

3.2 Canister degradation

Canister exposure to the repository environment will result in slow, long-term corrosion which will ultimately determine its lifetime under the expected external loads. The nature of the corrosion processes that will impact the canister will change over time as the environmental conditions within the repository evolve. Corrosion will not only reduce the load-bearing wall thickness of the canister but may also degrade the material properties due to the absorption of hydrogen produced during the long-term anaerobic phase.

King et al. (2024a), King et al. (2024b) have recently reviewed the various approaches employed for predicting the lifetimes of disposal canisters. A common approach is to assess the mechanical and corrosion performance of the canister separately. Thus, the canister is designed to remain structurally sound for the expected service loads and a minimum necessary wall thickness is defined. The remaining wall thickness is then calculated based on the required canister lifetime and the rates of the different corrosion mechanisms expected to occur under repository conditions. This approach was followed in two recent safety cases involving carbon steel disposal canisters, in France (Andra, 2022) and in Japan (NUMO, 2021). An alternative approach, which is followed here, is to consider the effects of both mechanical and corrosion degradation processes simultaneously. Such an approach is preferable for a single-shell carbon steel canister design as interactions between mechanical and corrosion processes are expected in the repository. In both approaches, a series of reasoned arguments are developed to justify the exclusion of corrosion processes not expected to occur.

For the disposal canisters described above, emplaced in a deep geological repository with the environmental conditions evolving as described in Section 2, the following corrosion processes are not considered possible or have a negligible impact on canister lifetime, and are excluded from the lifetime assessment:

Oxic corrosion: The atmospheric O₂ initially trapped in the emplaced bentonite buffer is consumed within a matter of days or weeks following backfilling of an emplacement drift. During that time there is insufficient adsorbed water on the canister surface to support aqueous corrosion processes. Even if oxic corrosion occurs and the initial O₂ inventory is consumed exclusively by corrosion, a wall loss of 80 μm for an SF canister, or 150 μm for a smaller RP-HLW canister, would occur. These amounts are insignificant compared with the wall thickness of 140 mm.

Radiation-induced corrosion: A recent review of experimental evidence (King, 2024) indicated that, within the range of experimental variability, there is no significant enhancement of the rate of uniform corrosion of carbon steel in bulk solution for dose rates below 10 Gy·h^-1. This critical dose rate for the effect of irradiation is consistent with the findings from a recent parametric study of the effects of dose rate and total absorbed dose on the corrosion of carbon steel (Bevas et al., 2024). In compacted bentonite, there is no significant effect of irradiation in any of the reported studies at dose rates of up to 3,500 Gy·h^-1. Since the calculated maximum dose rate at the outer surface of the canister is 0.2 Gy·h^-1, which is orders of magnitude lower than the critical dose rate, irradiation-influenced corrosion is excluded from the assessment, since no effect on the corrosion rate is expected.

Localised corrosion: Localised corrosion results from the spatial separation of the anodic and cathodic reactions. In aqueous systems, the supply of H₂O to the corroding surface is relatively rapid and it is difficult to create spatial gradients in the concentration of H₂O. Furthermore, the potential difference that results from the spatial separation of the anodic dissolution of Fe and the cathodic reduction of H₂O is small, so the electrochemical driving force for localised corrosion under anaerobic conditions is limited. Thus, localised corrosion under anaerobic conditions is unlikely and would, in any case, cause limited damage. However, even if localised attack does initiate, both laboratory and field measurements indicate that the degree of localisation decreases with increasing depth of (average) corrosion (JNC, 2000), implying that localised corrosion will gradually become uniform before the structural integrity of the canister is compromised.

Stress corrosion cracking (SCC): Of the various environments in which SCC of carbon steel has been reported, only one is similar to what is expected in the repository. Near-neutral pH SCC is associated with a dilute bicarbonate solution (in the range of 0.001–0.01 mol·L-1) with a pH in the range of 5.5–7.5 and potentials corresponding to the corrosion potential in anaerobic environments (King, 2010a; King, 2010b). In terms of the mechanical loading requirements, a cyclic load is a pre-requisite, which will be absent on the canister under repository conditions. Stress corrosion cracking is thus excluded from consideration for the assessment of canister lifetimes because of the absence of a suitable chemical environment in the repository and the absence of cyclic loading found necessary to sustain crack growth in environments like the one expected in the repository.

Microbially influenced corrosion: Near field microbial activity and the possibility of biofilm formation are excluded because of the inhibiting effect of compacted bentonite on microbial activity (see Section 4). Microbial activity in the EDZ can produce aggressive metabolic by-products such as sulphide, but the flux of corrosive species to the canister is predicted to be much lower than the corresponding rate of anaerobic corrosion due to the reduction of H₂O. Therefore, no additional corrosion due to the remotely produced sulphide is expected since the sulphide that would eventually reach the canister surface would react with corrosion products rather than consume metal.

The following corrosion processes are expected to occur, and are expected to have an important contribution to canister degradation, and thus their effect is quantified and included in the lifetime assessment:

Anaerobic uniform corrosion: Like other ferrous alloys, carbon steel will corrode in O₂-free H₂O accompanied by the evolution of H₂. The mechanism for the anaerobic corrosion of carbon steel leads to the formation of magnetite in pure water, and to the formation of additional carbonate-containing products in compacted bentonite (Féron et al., 2009; JNC, 2000; King, 2008). Magnetite surface films exhibit a duplex structure, with a thin inner barrier layer (of nanometre dimensions) and an outer porous layer formed by a dissolution-precipitation process (King, 2008). The corrosion reaction in the repository is under anodic control, governed by the rate of ion transport across the thin barrier film. The outer porous layer thickens over time but remains porous. In the presence of compacted bentonite, the corrosion product film grows into the clay layer, creating a zone of altered material immediately adjacent to the corroding surface (Féron et al., 2009). Since the properties of the inner barrier film determine the corrosion rate, periodic spalling of the outer porous film would not impact the corrosion behaviour, and there is no mechanism by which the corrosion rate could increase over time. Therefore, the use of a constant corrosion rate based on empirical measurements is a valid approach to estimating the long-term corrosion behaviour of the canister. Indeed, since the corrosion rate generally continues to decrease with time, due to thickening of the barrier layer over extended periods (Hesketh et al., 2023), it can be argued that the use of a corrosion rate determined from an experiment of finite duration (see Figure 3) is conservative as the long-term rate would be expected to be lower. A literature review on reported corrosion rates for carbon steel under repository-relevant conditions (anaerobic, saturated compacted bentonite, duration of at least 4 years) indicates that the expected rate lies between 0.026 and 0.7 µm/a, with the best estimate being 0.3 µm/a (Diomidis et al., 2023). The maximum rate of 0.7 µm/a was conservatively used in canister lifetime assessments. On the other hand, after breaching of the canister, the corrosion of the inner surface and basket occurs at a rate of between 0.0003 and 7.5 µm/a, with the best estimate being 1 µm/a, due to the different environment. The corrosion of canister internals is relevant to safety due to the resulting hydrogen gas generation (Papafotiou et al., 2026).

Hydrogen induced cracking: Hydrogen is generated at the external surface of carbon steel canisters as a result of the corrosion process that occur in anaerobic conditions, where a H₂ partial pressure of up to ∼ 11 MPa is expected to be present (Nagra, 2024d). Hydrogen will begin to accumulate both within the steel itself and inside the canister as atomic H diffuses through the canister wall and desorbs as H₂ on the inner surface of the canister wall (Turnbull, 2009). The absorption of hydrogen into the low strength carbon steel matrix structure can lead to a reduction in tensile ductility and fracture toughness. The susceptibility to hydrogen-induced cracking in carbon steel increases at increasing strength level. Hydrogen-induced cracking is included in the assessment of the canister integrity through the specification of a reduced fracture toughness. The value of fracture toughness is given by the minimum value of 55 MPa√m identified by a literature review on the effects of hydrogen on low strength steel nominally similar to the canister material (Nagra, 2024e) and also cited in ASME (2019). This value is considered conservative primarily because it has been determined at room temperature, and fracture toughness increases with temperature (Turnbull, 2009).

3.3 Structural integrity

Based on a conservative maximum corrosion rate of 0.7 µm/a, after 10,000 years in the repository, the canister is expected to have lost 7 mm wall thickness due to anaerobic uniform corrosion, while the fracture toughness of the material would be 55 MPa√m due to the effect of hydrogen (see Section 3.2). Furthermore, residual stresses would remain at the weld, as well as flaws embedded deeper than 7 mm, as described in Section 3.1. Such a degraded canister would be subjected to the mechanical loads described in Section 2.2. To evaluate whether such a canister is breached or not, it is necessary to assess whether the canister has undergone either gross plastic collapse, or fracture (brittle or ductile) or plastic collapse at the weld due to the presence of weld defects.

Failure by gross plastic collapse was assessed and shown not occur since the von Mises stresses in the canister wall, as calculated by finite element analyses (FEA), were significantly lower than the minimum yield strength of the selected canister material (220 MPa). The assessment of the second mode of failure, that could occur due to the presence of weld flaws, requires the employment of fracture mechanics. In this approach an engineering critical assessment (ECA) methodology was used, based on the BS 7910 standard (BSI, 2019). The maximum principal stress profiles along radial paths on the two long symmetry planes through the weld centreline were extracted from the FEA. The linearised stresses were used in the ECA calculations, together with the residual stresses after PWHT, and the expected flaws. The 1 × 10 mm surface breaking flaws that could have escaped detection would have been corroded away, and thus only flaws embedded deeper than 7 mm would remain after 10,000 years. The assessment of such flaws, expressed in failure assessment diagrams (FAD), is shown in Figure 4. In the FAD, the plastic collapse ratio (Lr) shows how close the structure is to yielding, while the fracture ratio (Kr) shows how close the component is to brittle fracture. The presence of all assessment points in the safe area inside the failure curve (shown as a black line), and the large distance between the assessment points and the curve, indicate a component that has not breeched with a significant safety margin.

4 Buffer

In Nagra’s disposal concept the buffer provides a stable environment around the HLW disposal canisters that limits corrosion, including the inhibition of microbial activity, and ensures that the mechanical loads on the canisters are evenly distributed. Furthermore, it contributes to the immobilisation, retention, and slow release of radionuclides by forming a physical and chemical barrier around the HLW disposal canisters. In the provisional design of the repository, the physical barrier is provided by selecting a buffer material with a low permeability. The chemical barrier is provided by selecting a buffer material with a mineralogy and equilibrium porewater that promotes the sorption and immobilisation by precipitation of many of the radionuclides that may be released after breaching of the canisters. Furthermore, the buffer needs to have chemical and mineralogical compatibility with adjacent components, including the Opalinus Clay host rock. Finally, the buffer shall ensure sufficient gas and heat transport to its surroundings.

4.1 Physical and chemical characteristics

Bentonites are mainly composed of smectites and in particular montmorillonite, a dioctahedral 2:1 clay mineral of the smectite group (Borchardt, 1989), made of 2 tetrahedral and 1 intercalated octahedral sheets (TOT) alternating with layers of charge compensating interlayer cations (e.g., Na⁺, K⁺, Mg²⁺, Ca²⁺, etc.). In Wyoming MX-80 bentonite, Nagra’s reference buffer material, the dominant interlayer cation is Na⁺. Bentonite’s cation exchange capacity determines the equilibrium composition and characteristics of the porewater (Nagra, 2024b).

Due to its high suction potential, compacted and confined bentonite will swell upon saturation. Pure bentonite exhibits a multi-modal pore-size distribution, comprising of macropores (inter-grain) and micropores (intra-aggregate and inter-aggregate pores). During saturation of compacted bentonites, the microporosity is slightly or even strongly reduced, depending on the saturation history. At the same time, micro- and meso-porosity increases slightly during hydration at the expense of the macropores (Romero and Gonzàlez-Blanco, 2017; Martin et al., 2023). An extensive review of the hydraulic and swelling parameters of bentonite has been compiled by Dixon et al. (2023). Swelling will also help to further homogenize any density variabilities within the bentonite buffer. Because bentonite is a clay-rich material, it has good geochemical compatibility with the Opalinus Clay host rock and has a high sorption capacity for many radionuclides (Miron et al., 2024). The swelling pressure developed upon saturation provides stability to the support of the emplacement drift and limits tunnel convergence after the tunnel support degrades. Its swelling capacity also allows any gas pathways that are formed to reseal. The compacted saturated bentonite buffer, with its high swelling pressure and nanoporous structure, is also a hostile environment for microbes; at high enough densities (≥1.45 g/cm³), microbial activity is inhibited (Stroes-Gascoyne, 2011; Masurat et al., 2010). The lack of microbial activity precludes the incidence of microbially influence corrosion (see Section 3.2). Furthermore, the bentonite aids in the rapid transition to anoxic conditions (Giroud et al., 2018), thus leading to a low corrosion rate of the carbon steel canister (see Section 3.2 above; Diomidis et al., 2023).

4.2 Buffer performance

4.2.1 Thermal interactions

The post-closure performance of the buffer relies mostly on the hydraulic conductivity and swelling pressure of the bentonite, which are dependent mainly on the dry density of MX-80 bentonite (Figure 5). The buffer will experience considerable temperature variation due to the decay heat of the HLW in the disposal canister. However, this thermal load is transient, and the timescale is in the order of a few thousand years (Nagra, 2024b). The initial thermal peak will not impact the safety relevant properties of the bentonite under the anticipated conditions, as experimental evidence have shown that in a dry state the bentonite can accommodate elevated temperatures up to 150 °C over extended periods of time (Martikainen et al., 2024). Furthermore, it has been estimated that the amounts of smectite potentially converted to illite during the elevated temperature period would not exceed ∼ 13–14% of the initial bentonite, leaving 86%–87% of the initial montmorillonite intact. This conservative estimate accounts for a small K+ gradient from Opalinus Clay to bentonite, which is supplied by diffusion to the bentonite over 10,000 years. These data are corroborated by experiments conducted in the laboratory and in underground research facilities (Bradbury et al., 2014).

During the early post-emplacement period, a saturation front will develop due to the steep pressure gradient into high-suction bentonite material which drives vapour diffusion and/or liquid flow, resulting in early saturated conditions near the drift perimeter. Non-isothermal conditions will affect the further saturation behaviour of the bentonite buffer. In particular, porewater in the buffer will be evaporated and transported away from the canister by vapour diffusion into cooler regions, where the vapour will condense, creating a counter flow of liquid water towards the canister. Swelling of the outer parts of the bentonite buffer may result in additional compaction and increased density close to the canister which would partially homogenise over time.

4.2.2 Corrosion interactions

During anoxic corrosion of the disposal canisters, Fe(II) as a corrosion product is expected to either precipitate at the interface (see Section 3.2) or diffuse into the bentonite buffer. The amount of bentonite that could be converted to Fe-silicates as a result of reaction with Fe(II) has been estimated, based on experiments, to be less than ∼3% (upper bound of 17%) over a period of one million years (Nagra, 2024b). On the other hand, a common outcome of large-scale in-situ experiments is the usually limited extent of Fe mobilisation and clay alteration induced by steel corrosion processes, as only a tiny fraction of the clay thickness is affected by significant chemical/physical modifications (Nagra, 2024b). Thus, the original swelling and sorbing properties of the smectites in bentonite are not significantly altered by the uptake of Fe(II) at repository-relevant temperatures (Wersin and Kober, 2017; Hadi et al., 2017) and the radionuclide retention and rheological properties of the bentonite buffer are not compromised.

4.2.3 Interactions with the liner

In general, the differences between the Opalinus Clay porewater and bentonite porewater are relatively small and the high buffering capacity and high cation exchange capacity of the bentonite will regulate the porewater composition for an extended period. In the outermost part of the bentonite buffer, chemical interaction with the concrete liner around the emplacement drifts locally affects the porewater composition. Over time, however, once the liner has degraded, the porewater composition in the buffer will become increasingly similar to that of the Opalinus Clay due to diffusive exchange with the surrounding porewater (Nagra, 2024b).

Considering both the temperature pulse and the mineral/porewater evolution across the Opalinus Clay/liner/bentonite interfaces, the altered thickness at the external side of the liner is not expected to exceed about 10–20 cm (Nagra, 2024b). In both the bentonite and Opalinus Clay domains, the total amount of dissolved clay minerals is very small at the end-time of the simulations (several thousands of years) (Ma et al., 2024). Precipitation of secondary minerals causes a reduction in porosity in both the bentonite and Opalinus Clay at the respective interfaces to cementitious liner materials. As this is a very local phenomenon in the millimetre to centimetre range, it does not affect the performance of the bentonite (Nagra, 2024b).

4.3 Radionuclide retention and slow release

Radionuclide concentration levels released into the near-field after canister breaching will be in general very low, as spent fuel and glass dissolution rates are, under the expected conditions, very low (Guillemot et al., 2026). The concentration levels of elements within the canister as well as within the bentonite buffer are controlled by the solubility limitation of the in-canister water and bentonite porewater composition. Derived maximum solubility levels reach maximums in the micromolar range (Hummel et al., 2023), while most elements are in the nano-molar range. For instance, within a spent fuel canister, U concentration levels for different calculated porewater endmember compositions results in concentration levels in the 0.03–0.9 nM and for Pu in the range of 60 nM (Nagra, 2024b). Once radionuclides are released into the near-field, the bentonite ensures very low diffusivities (Miron et al., 2024) and a strong sorption capacity for almost all solutes (Glaus et al., 2024), thereby retarding the release of radionuclides. Retardation will allow for substantial decay of short-lived and many long-lived radionuclides within the buffer, thus strongly attenuating their release rates into the Opalinus Clay host rock.

Furthermore, gas generation due to the corrosion of steel canisters is slow enough to allow the gas to diffuse through the bentonite buffer without causing irreversible damage (Diomidis et al., 2023, Nagra, 2024c, Nagra, 2024d). Pre-existing or induced discontinuities (e.g., cracks, flow channels along grain boundaries) tend to self-seal when the bentonite is fully saturated (Seiphoori, 2015; Leupin et al., 2021).

5 Sealing element

A sealing-system has been developed for the HLW drifts, for the HLW repository section and for the repository as a whole (Smith, 2026). Each HLW emplacement drift will be closed with a V1-HLW seal (Figure 6), comprised of a sealing element made of compacted bentonite blocks, stabilised by concrete abutments on either side. The abutments serve to absorb the mechanical stresses resulting from the swelling of the bentonite of the sealing element and the buffer. The transition layer comprises of inert calcareous aggregates and serves to physically separate the concrete elements from the clay material to limit chemical interaction. The main functionality of the sealing element is to limit water flow from the repository side along the repository components towards the portal side. As such, the seals contribute to slow and diffusion-dominated transport of radionuclides. The bentonite in the seal design also promotes radionuclide sorption. This means that the seals contribute to the retention of radionuclides in the closure system. By limiting the porewater flow from the entrance of the HLW emplacement drifts, the seals also prevent the bentonite buffer from becoming saturated with water coming from the entrance of the drift, promoting radial and regular saturation of the bentonite buffer by porewater from the host rock around the drifts. The use of bentonite for the sealing elements minimises chemical interactions with the surrounding clay rock. The compacted bentonite has a dry density greater than 1.45 g/cm³ to ensure both sufficient swelling pressure and that the intrinsic permeability at complete saturation meets its design target of 10⁻¹⁹ to 10⁻¹⁸ m² (Martin et al., 2023). The way in which the seals contribute to the limited radionuclide release from the HLW emplacement drifts is presented in Li (2026), while their performance is presented in Papafotiou et al. (2026).

6 Conclusion

The deep geological disposal of spent nuclear fuel and vitrified high-level waste is a scientific, engineering and societal challenge that affects both current and future generations. Consequently, regulatory requirements and scrutiny are focused both on the design and the long-term safety assessment of relevant facilities. In this manuscript we describe the provisional design of the engineered barriers foreseen in the provisional repository concept for the disposal of HLW in Switzerland and present the scientific basis on which the evolution of their properties and the assessment of their performance are founded.

An important component of HLW disposal, both from the operational and long-term safety perspective is the canister. It is designed to provide complete containment of radionuclides for a certain period of time, but also to allow efficient operations and handling. The proposed thick walled canister is made from a low carbon steel, selected for weldability, minimal hydrogen effects and predictable corrosion performance. Based on the understanding of the evolution of the immediate environment, it is possible to assess which corrosion mechanisms may contribute to canister degradation and assess their impact over time. This allows the description of the state of the canister at any point in the future supported by experimental evidence, data from the literature and modelling. This includes a reduced wall thickness due to corrosion and a reduced fracture toughness due to hydrogen embrittlement. An engineering critical assessment approach based on a recognised methodology, taking into account applied and residual stresses, as well as postulated defects, demonstrates that the regulatory lifetime requirement is satisfied with a significant safety margin. The proposed canister design will remain unbreached for at least 10,000 years.

The buffer protects the canister chemically and mechanically by reducing corrosion rates and distributing lads homogeneously. Further, the buffer contributes to the immobilisation, retention, and slow release of radionuclides by forming a physical and chemical barrier around the HLW disposal canisters. Bentonite, which is selected as provisional buffer material, has a suitable mineralogy and ensures suitable porewater chemistry that promote the sorption and immobilisation by precipitation of many of the radionuclides that may be released after breaching of the canisters. Solubility limits within the bentonite porewater are in the nanomolar to micromolar range for most considered radionuclides. Radionuclide release from the near-field will be highly restricted due to the swelling capacity of the bentonite, where hydraulic conductivity will be strongly decreased, leading to a diffusive regime. These properties also make bentonite a suitable material for the sealing elements used to seal the HLW emplacement drifts. Assessment of interaction processes between the bentonite and the rock support or the canister as well as studies on the thermal effect have indicated that the bentonite will fulfil its expected safety function for the period of consideration.

Therefore, the properties of components of the HLW near-field are understood well enough to allow an adequate description of their evolution over long time-scales, thus allowing to demonstrate their barrier function and long-term performance.

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Footnotes

Edited by: Sajid Iqbal, Korea Advanced Institute of Science and Technology (KAIST), Republic of Korea

Reviewed by: Yifeng Wang, Sandia National Laboratories (DOE), United States

Leonardo Marchiori, University of Beira Interior, Portugal