TRIPLE C Waste Container for Increased Long-term Safety of HHGW Disposal in Salt, Clay and Crystalline
1 Introduction
Nuclear facilities for the utilization and handling of nuclear materials have to fulfill general safety goals. With varying importance and priorities the same five main safety goals apply for safety considerations of all nuclear facilities.
– ISOLATION (prevention of release of nuclear material in biosphere)
– SHIELDING (prevention of irradiation with an overdose)
– CONTROL (prevention of criticality)
– PROTECTION (prevention of destruction, misuse, theft, unintentional intrusion…)
– HEAT REMOVAL (prevention of overheating)
Depending on the facility type and intended use a tailored set of appropriate safety measures has to be foreseen to guarantee the fulfillment of safety goals in all phases of operation and over the whole lifecycle of the facility.
The widespread utilization of fissile materials in nuclear reactors (fuel elements) generates unavoidably large amounts of materials with a high hazard potential (high radioactive, partly with very long half-lifes; chemotoxic: heavy metals like plutonium;
fissile: potential for uncontrolled chain reaction or misuse in nuclear explosives, heat generation; extremely high concentrations).
Typical steps in the history of fuel elements from utilization in the reactor till the final disposal are shown in Fig. 1 [2],[3].
Fig. 1 Stationary phases (SP) and transition phases (TP) in HHGW history
During each stationary phase (SP) or temporary phase (TP) the fulfillment of the five safety goals must be guaranteed by a set of appropriate measures, tailored to the special conditions and requirements of the phase. The priority ranking of the safety goals may change from phase to phase.
- Final repository
It is acknowledged worldwide that HHGW must be safely isolated from the biosphere for a time period of 1 Mio years.
Deep geological repositories like mines or deep boreholes are considered as best solutions to isolate the waste permanently and prevent inadvertent human intrusion. But a deep geological repository for HHGW is a challenging new type of a nuclear facility.
German disposal concepts foresee deep geological disposal (mine) with a combination of geological barriers and engineered barriers (EBS). Fig. 2 shows the general scheme.
Fig. 2 HHGW repository – a dynamic nonlinear system
After closure, the repository has to fulfill extreme safety requirements for a long time in a predicted manner without further human interventions.
The final repository is a dynamic nonlinear system. For such a complex system it will difficult – if not hopeless – to find the appropriate equations system and then find precise solutions to determine the leak rate L(t) as a special OUTPUT function. Nevertheless, some general properties can be formulated which lead to some useful conclusions for the design of the repository as a system and its components, especially the waste container.
The system behavior is somewhat predictable, if the inner status remains near a steady state and the deviations from the starting FEP`s conditions and of the system variables are small over time. Due to the system dependence of the initial conditions, it is a fundamental requirement for the repository design, that the initial conditions (inner and outer FEP`s, e.g. functionality of barrier system, waste distribution, subcriticality) will not change for as long as possible.
The early loss of retention capability of the waste package has the consequence that hazardous materials are released from their original location and are permitted to lead a vagabond life. Principally, self-organization and chaotic behavior of the system become possible. In the worst case, conditions for a self-sustaining chain reaction (criticality) are formed with severe consequences for the status of the whole repository and the release of hazardous material in the biosphere.
The effective lifespan of the repository is adjusted to the half-lifes of the long-living radionuclides. Therefore, the time constants of the engineered barrier system (EBS: retention capability of waste container and geotechnical barriers) should be comparable and fit in this time scale too.
It is planned in many countries to use metallic containers as engineered barriers together with a surrounding layer of bentonite. Sweden and Finland want to apply copper canisters (KBS-3), Germany spheroidal graphite iron (Pollux) and the United States stainless steel for example. The Swedish concept of SKB has very often been cited as reference concept, but came under harsh criticism by the decision of the Swedish Environmental Court at the beginning of 2018 [4] and has finally been postponed by 10 years. It is generally known that all metals exhibit a relatively poor corrosion resistance under disposal conditions, especially if very long time periods are considered [5], [6], [7]. (Fig. 3) For good reasons metallic waste container play therefore only a secondary role in existing safety concepts for repositories planned in different types of host rock (salt, clay, crystalline)
But new developments in high-tech ceramics provide a sound scientific-technical basis for the industrial production of ceramic waste containers. But most important, excellent material properties justify the expectance of long-term retention capability [8].
Fig. 3 Time scales and long-term retention capability of waste containers:
Note: Even 100.000 years are only 10% of the nominal repository lifecycle.
This paper describes why and how silicon carbide (SSiC) waste container can play a decisive role for long-term safety by providing a corrosion-resistant initial barrier, diversity and redundancy in all host rock disposal systems.
- Defense-in-depth for HHGW repository
The concept of defense-in-depth is a fundamental element of safety philosophy for nuclear and non-nuclear complex systems, where ultra-high reliability has to be achieved. Defense-in-depth is not a goal, but a tool that is used for the approach to design and operate a nuclear facility that prevents and mitigates accidents with release of radiation or hazardous materials. The key is creating multiple independent and redundant levels of defense to compensate potential failures in designing and manufacturing as well as accidents during lifecycle so that no single level, no matter how robust, is exclusively relied upon [9].
Basic defense-in-depth features concerning waste can be found in the proposed strategy for development of regulations governing disposal of high radioactive waste in the proposed repository at Yucca Mountain [10]. The development of NRC regulations for geologic disposal represented a unique application of the defense-in-depth philosophy to a first-of-a-kind type of facility. The paper underlines the difference between a geologic repository and an operating facility with active safety systems and the potential for active control and intervention. The safety of a closed system over long timeframes is best evaluated through consideration of the relative likelihood of threats to its integrity and performance. Also it is relatively easy to identify multiple diverse barriers that comprise the engineered and geologic systems. The performance of any of this systems and their respective subsystems cannot be considered truly independent or totally redundant.
The general defense-in-depth framework (DiD) for a repository is shown in Fig 4.
Fig. 4 General defense-in-depth framework for a repository
MATRIOSHKA – Principle: 4p geometry of inner shells [3]
The physical barriers (essential barriers B [1], relevant material zones Z) placed between waste and biosphere form a hierarchy of different Levels of Defense in a successive or consecutive manner. If one level fails, the next level is meant to alleviate the failure of the previous level and so on, so that all the levels must fail before significant consequences will occur (Tab.1). In reviewing the international literature there are only general statements with no specific criteria for determining the adequacy of defense-in-depth in waste disposal. But control of single failures alone requires the existence of a redundant system (combination of two geological barriers or a geological barrier together with EBS). Furthermore, the Fail-safe-principle can be fulfilled only by emplacement host rocks with self-acting closure of cracks and rifts by plastic flow (preferably salt, eventually clay) or the combination of bentonite and crystalline.
Tab. 1 Hierarchy of Levels of Defense, barriers B and zones Z to obtain safety goals;
Not essential long-term safety contributions from existing waste (Level 0)
Geotechnical barriers (backfill, closure and sealing of tunnels and shafts) are not included in the scheme. They are important components of the whole repository concept, but after all only repair measures of the host rock resp. of the overlaying rock and are therefore not considered as autonomous barriers. Nevertheless many authors sum up these pseudo-barriers equally together with the geological and engineered barriers to pretend larger safety marges concerning redundancy and diversity
Principles are developed to help guide implementation of defense-in-depth in waste disposal. Generally, defense-in-depth philosophy consists of four principles [11]:
– prevent accident from starting (initiation, prevention)
– stop accident at early stages before they progress to unacceptable
consequences (intervention)
– provide for mitigating the release of the hazard vector (mitigation)
– provide sufficient instrumentation to diagnose.
A repository after closure is a totally passive system (no operation, no maintenance, no surveillance, no monitoring, no diagnosis). In this case not all principles apply to appropriate defense-in-depth measures. With the increasing loss of information on the site and the inner status (lack of diagnosis) of the repository active human measures (intervention) to stop accidents by retrieval or recovery of waste containers
are limited to a very short period (~ 500 years [1]). This underlines the necessity to design the repository with sufficient passive measures for long-term retention.
Practically all existing concepts to achieve the safety goals rely only on the choice of an appropriate host rock and site. But deeper insight in FEP`s of geological sites changed the perception of the relative importance between different levels of defense. To some extent this new position found its reflection in the German “StandAG” [12].
- Safety requirements according to the new German regulations
The regulations on safety requirements for final deposition of high radioactive waste (EndlSiAnV) [1] are part of the new legal provisions which represent the legal base in Germany for the layout and the evaluation of long-term safety. A summary is given in Tab. 2.
Tab. 2 Criteria for the overall performance of a final repository [1]
With this specification of criteria, the frame has been established for questioning the suitability of existing concepts or for developing targeted concepts being in the phase of planning and realization already. Essential barriers are those which mainly ensure the safe enclosure of radionuclides. Essential barriers may be one or some effective rock regions or, if no such effective rock region can be identified, technical and geo-technical barriers. In extreme case, one essential barrier stands for the overall performance of the whole repository.
The repository as a system fails (system failure, accident), if the amount of released radionuclides leads to values, which exceed the maximal permissible radiation dose or the maximal tolerable concentrations of toxic materials in air, water and food.
In a simplified manner the relationships between the inventory, leak rate and the released hazardous material can be written as follows.
The total inventory mass M(t) of radioactive nuclei is given by M(t) = S mi (t) with i= 1…n (nuclide vector). At closure of repository (t=0) the total inventory is M(0) = M0. Provided, M(t) is distributed evenly on N container, than the inventory of one container is MC (t) = M(t)/N.
The leakrate L(t) in Fig. 2 is given by L(t) = dMrel(t) /dt with dMrel(t) the released mass of hazardous material from the repository in biosphere in time interval dt.
Lets assume, that the source term Q(t) of one container is Q(t) = dMC(t)/dt and all N container should have the same source term. Furthermore, the permeability of all geological barriers P(t, x,y,z) is set P = const. for all nuclides over the lifecycle of repository and over the total volume of the emplacement rock, than
Mrel(t) = ~ ~ PN (1)
Equ. (1) provides a direct relationship between Mrel(t) of the whole repository and and the source term Q(t) of a container. The ultimate goal is Mrel = 0 over 1 Mio years respectively Mrel < 10 -4 M0 , taking the values from Tab. 2.
People expect a nuclear facility to function properly – especially a HHGW repository. But they do fail as the example of Asse II has shown. Different failure types influence the safety and reliability of a repository. The following basic ideas for failure assessment follow very close to the definitions and results of the paper of JONES [13].
Failures can be classified as random or systematic. Random failures of technical systems (e.g. EBS) are caused by time and use and occur independently. Non-random (systematic) failures occur because of a poor specification or design of a system or an unexpected interaction with the system`s environment or external stress. Systematic failures have identifiable causes and familiar sources. They are understandable and explainable. Systematic failures can effect all identical components of a system (e.g. waste container) so the systematic failures are the potential common cause failures.
Achieving the deep geological repository as a reliable and safe total system, the first step is to select a highly reliable subsystem (e.g. first geological barrier = emplacement rock, in German terminology: einschlußwirksamer Gebirgsbereich ewG). But often the best possible subsystem at Level 2 of Defense (available rock type and site on territory) has failure rates, that are too high. Than the necessary step is to provide redundant subsystems either on Level 3 (e.g. 2. geological barrier) and/or on Level 1 (EBS, waste container). If an approriate site with conditions that form a set of redundant geological barriers cannot be found on the territory of a country, then the safety goals of the system must be realized by appropriate measures on Level 1 (waste container).
With metallic waste containers it seems to be very difficult – almost impossible – to demonstrate how long-term retention can be achieved. For decades repository concepts focused therefore their safety considerations on Level 2 only, but with unsatisfactory results.
So the time has come to re-consider the contribution of innovative waste container to the long-term safety.
- Safety measures on Level 1
The existing HHGW (the waste material itself and the metallic cladding of spent fuel elements and canned vitrified waste from reprocessing) do not provide a long-term retention barrier on Level 0, not until innovative ceramic-encapsulated fuel elements (accident-tolerant fuel ATF, disposal-preconditioned DPF) will fill this gap in the future. So realistic measures for ISOLATION start on Level 1. New developments in high-tech ceramics provide a sound scientific-technical basis for the industrial production of ceramic leakproof waste container [14].
5.1. ISOLATION
The central part of a TRIPLE C waste container is a silicon carbide (SiC) container For several reasons the special type SSiC (pressure less sintered silicon carbide)
has been chosen.
The choice of SiC as container material is based on different criteria which are listed in the Table 3. An important impetus came from the former activities in Germany concerning the HTR-reactor and the encapsulation of the fuel, so-called TRISO particles, in a very thin shell of SiC [15], [16] with a thickness of 30 µm.
SiC as a chemical compound was detected in stellar matter, meaning, that it is extremely stable, but it rarely exists on earth as natural mineral. Fortunately SiC can be synthesized in any required quantity from the abundantly available raw materials sand (SiO2) and coke (C) by applying electrical energy.
Tab. 3 Selection criteria for SiC as c ontainer material
The corrosion resistance of SSiC against acids and bases justifies research and development to make this material available for the encapsulation of HHGW. Even though today still a front-edge technology, SSiC container can be manufactured for all existing waste forms.( Fig. 5, Tab. 4 )[14].
A cylindrical container is the basic geometry. Diameter D, height H and wall thickness d are adjusted to the waste geometry. The inside surface is coated by a glassy carbon layer (so-called SIAMANT compound). Depending on the length of fuel elements, the container bodies are monolithic or segmented.
Fig. 5 SSiC container (monolythic or segmented) for all waste forms with
laser-engraved identification code and Safeguards seal [14]
For a long time, the hermetic closing of the container as well as the bonding of segments for forming large container bodies was considered as the fundamental drawback for the application of SSiC container. But with the native bonding technology Rapid Sinter Bonding (RSB) [17] a quick and reliable process for a strong and gas tight seam has been developed. By laser engraving each container gets a permanent identification code and a Safeguards seal (making the container a “batch” in Safeguard’s terminology).
Tab. 4 SSiC container dimensions for different waste forms (Fig. 5)
(container monolithic or segmented)
Taking into account the excellent corrosion resistance of SSiC in acidic and basic environments and its extremely high hardness, it is assumed, that the container wall will not be damaged neither by corrosion nor by erosion during the nominal life time tN of the repository: d(t) = d0 for 0 < t < tN .
Furthermore, it is assumed, that the integrity of the waste package is maintained by respective dimensioning of material zone Z2 (e.g. bentonite) and by appropriate emplacement conditions in a stable host rock.
But despite an intact container wall material transport happens by diffusion. (Fig. 6) [18].
Fig. 6 Pathways for material transport through the container wall
The diffusion coefficient D is specific for each container material and for each type of the nuclide. D is a function of temperature T: D = D(T). The temperature dependence of the diffusion coefficient is usually given by the ARRHENIUS equation:
D(T) = D0 exp[ – EA/(RT)] (2)
with EA in kJ/Mole and R = 8.3143 J/Mole.
The data basis for diffusion coefficients of radionuclides in SiC is quite limited momentarily. Existing values have been measured by radiation and heating experiments in the temperature range from 600° to 1200 °C. A standard data set exists for the metallic fission products Cs, Sr and Ag [15],[16].
For the temperature range in a final repository (T < 200 °C) no relevant data could be found. But it seems admissible to use extrapolated values for the presented estimations which make use of many assumptions anyway. The assumptions are always on the conservative side.
The diffusion processes are described elsewhere [19]. The glassy carbon layer as intended protection layer for the inner wall surface (Ag, Pd) exercises due to its special properties a delaying effect on the diffusion process. In this way and together with additional potting material in zone Z1 a “functional barrier” is formed. Its influence on diffusion can be lumped-up to an increased wall thickness (d+r), r having the dimension of a length.
The critical diffusion coefficients Dcrit, which fulfil the leak-proof criteria of Tab. 2 for the given wall thickness d0 together with the functional barrier (characterized by r) are than
Dcrit < (d0+r)² /6tN (break-through-time criterion) (3)
Taking d0 = r = 10-2 m and tN = 106 years the values for Dcrit are in the range from 10-20 …..10-18 m²s-1 (Fig. 7, hatched area)
Fig. 7 Diffusion coefficients of Ag in SiC and in copper for comparison
The results shall be interpreted in the following way.
If a radionuclide i has a diffusion coefficient Di < Dcrit in the temperature range of the repository (T< 200°C) than the SSiC container is considered as leakproof for this nuclide i over the nominal lifecycle of the repository, provided its overall integrity is maintained.
Within the uncertainties of the assumptions, the chance is given that principally, the safety goal ISOLATION for an essential barrier can be fulfilled by intact SSiC container.
But engineered systems fail. So do SSiC containers too, by random and systematic failures. High standards in manufacturing and quality control of SSiC containers can provide a low random failure rate F1, but nevertheless the consequences of container failure can be considerable.
For example, there are N equal waste containers with equally distributed inventory Mc = M0/N. Assuming N =104 container and only one containers fails totally (F1 =1), than the released amount on Level 1 is M1= F1Mc = 10-4 M0, This equals already the maximal permitted release value for the whole repository (Tab. 2). So additional mitigating measures on Level 2 (and on Level 3 if possible) are necessary.
Normally, the design of a repository starts with a careful search and selection of the site and host rock type (Level 2). Focused only on the site search, achieving ultra reliable system performance (e.g. safety goal ISOLATION) is difficult. It requires identifying the even very unlikely failure causes for the given system and then redesigning the system to remove them (in the extreme case: selection of another type or site of host rock becomes necessary as in the case of Asse II). Maybe that the achievable overall failure rate F is still too high. Then the necessary next step is to provide redundant subsystems. Maybe the site has not a second geological barrier (site Gorleben, Germany): Than a redundant subsystem can be provided only on Level 1., because it is impossible (or too expensive) to construct an essential retention barrier on Level 3. Only ceramic waste containers have the potential for an essential barrier providing long-term retention.
But systematic failures can affect the large number of identical waste containers. A common cause failure is a specific type of systematic failure where several failures result from a single shared cause. Two types have to be distinguished: the common event failure and the common mode failure.
A common event failure is given, when multiple failures result from one single external or internal event. The failures are usually simultaneous or nearly so. Common event failures can launch failure sequences (cascade failures). External events include earthquake, tsunami, hurricane, flood (external FEB`s , Fig. 2). Internal events like criticality may develop in a cascade from external events.
A common mode failure is the other specific type of a common cause failure, where several subsystems fail in the same way for the same reason. Common mode failures occur at different times. The common cause could be a design defect (e.g. inadequate, non-corrosion-resistant container material)
5.2 CONTROL
Common event failures are a major concern for redundant systems. The formation of a critical assembly inside the repository is a common event which can have severe consequences for the overall retention capability, in the first line by heat generation with a subsequent destruction of the repository structure. Therefore, measures have to be foreseen which exclude criticality definitely
Outside FEB`s can trigger a sequence of internal processes which support the formation of a critical assembly. A possible, not totally unlikely scenario (Fig. 6) includes several steps: destruction of original fuel cladding (Level 0), destruction of original fuel geometry, relocation of fuel inside container, loss of container integrity, water ingress in repository and finally in the container, material transport processes in the near-field of the container, coupling with other destroyed containers, ultimately self-organization of a sufficient amount of fissile material in a geometry, which generates a self-sustaining fission chain reaction (criticality, effective multiplication factor keff = 1).
The safety goal CONTROL is achieved, if subcriticality keff(t) < 0.95 is always garantued for 0 < t < tN for the overall repository as well as for each subregion (Tab. 2) [1]. Generally, the effective multiplication factor keff is a function of material composition, geometry, temperature T and time t:
keff = f[material(t), geometry(t), T(t)] (4)
Only one spent PWR fuel element of average burn-up contains enough fissile material to start a chain reaction under “improved” geometrical conditions and in the presence of an appropriate moderator.
Several measures can prevent self-organized criticality:
– stabilization of the material geometry inside the container
(a single tall fuel element is not an optimal geometry for criticality)
– prevention of water access
– neutron absorber in the container.
The TRIPLE C concept foresees a special measure which solves these problems simultaneously. The numerous voids in the container between the waste and the container wall resp. between the single rods of o fuel element are filled with a so-called potting compound (Fig. 8).
After loading the waste in the container, the potting compound – being in a floating state – is poured in to fill all voids. In the simplest way it can be dry sand in a mixture with a boron containing component. But the preferred potting compound solidifies after filling. A SiC precursor with a small surplus of carbon and boron as sinter additive is transformed into solid SiC under the influence of radiation from the waste (RISiC: radiation induced SiC). The necessary activation energy for the endothermic SiC reaction comes from the B-10(n,a) neutron capture reaction [14]. The product is a very hard porous material, which stabilizes the inside geometry and prevents relocation of waste, absorbs neutrons, shields the container wall against radiation defects from neutrons, prevents water ingress, improves the heat transfer inside the container and enhances the overall mechanical stability of the SSiC container.
Fig. 8 Principle arrangement of potting compound containing boron (left) and
for demonstration in a 7-rod bundle in an SSiC container (right)
So potting with an appropriate compound forms a combination of several efficient measures to prevent criticality already on Level 1. These measures are backed-up by a leakproof container, a bentonite buffer and a dry emplacement environment.
5.3 PROTECTION
The main generally expressed concern against the application of all kinds of ceramics is their brittleness and the risk of failure under mechanical stress.
The geomechanical aspects of SSiC waste containers have been investigated by the Geomechanical Institute of TU Bergakademie Freiberg which laid the basis for further investigations [20],[21].The known mechanical properties of SiC under static and dynamic load are completed by supplementary laboratory tests. Although strength values for SiC and especially for SSiC are very high, the extreme brittle behavior has to be considered in case of impact and point-like loading. Comprehensive numerical simulations were performed for the most critical potential loadings during transportation to the final position and during the storage in the emplacement position. As criterion for potential damage a static tensile strength of 150 MPa was used. Investigated load cases include free fall of an unprotected/protected container, rock fall on the container and earth pressure up to a depth of 1200 m. The most important conclusions can be summarized as follows:
– Earth pressure, even with high anisotropy of stress, cannot lead to any damage of
the SSiC container, even if no protective cover is used.
– Extreme loading constellations during transport and emplacement can lead to local
peak stresses in the container body, which exceed the 150 MPa criterion. But by
using an appropriate protective cover (overpack, transport container, buffer)
damage can be excluded with high probability.
5.4 HEAT REMOVAL, SHIELDING
In comparison with the other safety goals, HEAT REMOVAL and SHIELDING have a minor priority.
The limited waste inventory (low heat source) together with an improved heat transfer by the potting material and the excellent thermal conductivity of the SSiC container material will avoid hot spots and provide sufficient heat removal to keep the container surface temperature below the maximal permitted value (< 100 °C).
The SSiC container itself together with the potting material cannot provide sufficient radiation shielding. Therefore an appropriate transfer container is required for the transport of the waste package between final conditioning facility and the emplacement position. Once in the final position (several hundert meters below earth surface), the overlaying rock and earth layers protect the biosphere completely from the radiation, emitted by intact waste container.
- TRIPLE C container
The term TRIPLE C stands for a threefold ceramic encapsulation (Fig. 9).
The crucial component is the SSiC container (B2). The voids between waste (here spent fuel with cladding B1) and the container are filled with the potting material (Z1). A shock absorber (SA, e.g. graphite felt) and an overpack (OP) protect the brittle SSiC container. The newly developed carbon concrete is proposed as material for the overpack [22]. The armor of this concrete container consists of woven carbon fibre structures instead of steel, making the whole composite much stronger, lighter and less susceptible to corrosion.
The function of each single layer has been discussed in Chap. 5. The SSiC container is tailored to the dimensions of the waste. This allows the completion of the INITIAL BARRIER /Tab. 1) at an early stage of the back-end history (preferably already at transition SP2/TP2, Fig. 2). It includes the following steps: loading waste in the SSiC container, potting, hermetical closing of container body with lid, laser engraving with ID and Safeguards seal). Either type of host rock nor specific site conditions of the intended repository must be known at this time. Such early “disposal pre-conditioning” can be very helpful for the subsequent waste management (handling, extended storage, transportation).
Fig. 9 TRIPLE C concept for HHGW container: threefold ceramic encapsulation
A schematic representation of a TRIPLE C waste package in the final repository environment [23] is illustrated in Fig. 10. The inner retention barriers, consisting of the ceramic potting compound and the solid SSiC wall, are invariant for all kinds of host rocks, since their predominant function is to keep the source term for spreading of hazardous materials at Q(t) = 0 at Level 1. This requires an undamaged SSiC container for the total lifecycle of 1 Mio years. The interspace between the container and the carbon concrete overpack is filled with a shock absorbing material. The armour of the overpack and the shock absorber can be used together as a fibre bag cargo lifter.[24] The specifications for overpack and buffer can be chosen at a very late time in the waste history, according to the conditions in the emplacement position. The thickness of the carbon concrete overpack must be designed according to the needs for handling and transport protection. The thickness of the embedding bentonite is dependent of the surrounding host rock and the respective load parameters are contributed by geomechanics [25]. As their main function, the bentonite and the overpack have to protect the inner barriers from mechanical damage by the host rock.
This principle of split and shared functions makes the TRIPLE C container flexible and adaptable to all types of host rock [4],[23].
Fig. 10 TRIPLE C – a waste container concept of ceramic layers in MATRIOSHKA geometry
Fig. 11 shows different steps of encapsulation of a hexagonal PWR fuel element (dummy, WWER – 1000).
Fig. 11 Demo -TRIPLE C container: 7-rod-bundle with demo carbon concrete overpack [courtesy 22] (left) and first steps of encapsulating of PWR/BWR spent fuel elements
- TRIPLE C container change paradigm
SSiC properties and high technological standards of container manufacturing and quality control justify the claim that each TRIPLE C container fulfills the requirements of an essential barrier for the container inventory (Tab. 2).
The total inventory M0 is distributed on N container (M0/N) If properly protected from geomechanical loads each container has the potential for a zero-source term Q(t) = 0 over the repository life time. Together the N leakproof container represent an essential barrier for the total inventory. The retention capability of N individually quality controlled TRIPLE C container is estimated to be higher than the retention capability of one big-volume emplacement rock (volume ~ 109 m³). It seems justified to consider Level 1 as the main retention barrier (ISOLATION). The top priority for Level 2 becomes than PROTECTION for Level 1.
Tab 5 New paradigm in repository philosophy: shift of main retention barrier to EBS
A shift of the main retention barrier from geological barrier to engineered barrier is a paradigm change in the basic philosophy for repository concepts. It may change the perception of the repository safety in the public debate too.
- TRIPLE C waste container enhance longterm safety of repositories
TRIPLE C waste container provide redundancy and diversity to each repository concept especially for the measures focussed on ISOLATION and CONTROL (Tab. 6 ).
Tab. 6 Contributions of TRIPLE C container to redundancy and diversity of safety measures
The use of TRIPLE C containers is not limited to a definite emplacement environment [25]. They can become an essential part of all repository concepts in salt, clay or crystalline (Tab.7). A favorite combination could be the following arrangement: SSiC container (B2) with potting (Z1) and carbon concrete overpack (OP) in bentonite buffer (Z2) and salt emplacement rock (B3; steep or flat: plastic behavior of salt fulfils the fail-safe principle by self-sealing) and after all with a leakproof overlay (B4). Taking into account the easy solubility of salt in water, crystalline (B3) with a leakproof second geological barrier (B4, salt or clay) can be a promising alternative.
For many years the Swedish design with KBS-3 copper container has very often been cited as the internationally accepted Reference Concept and has found derivatives in Finland, Japan, Uk, Switzerland and others. But with the decision of the Swedish Environmental Court [4] in the beginning of 2018 it came under harsh criticism and caused moratoria and scrutiny of national programs. Applying the same, above outlined criteria to the existing repository concepts, even to the newly published 10 German RESUS concepts [26] reveal the same fundamental flaws: absence of long-term safety measures on Level 1 (Tab. 7) resulting in lack of redundancy and diversity for the whole repository concept.
Tab. 7 TRIPLE C container – an excellent match for all repository concepts enhancing long-term safety
Summary/Conclusions
Innovative technologies can help to overcome fundamental flaws in repository concepts, having dominated for decades the safety philosophy for final disposal of HHGW. TRIPLE C container can be implemented in each repository concept. The features of tailored ceramic encapsulation following the TRIPLE C concept justify the claim to build confidence in long-term safety on the engineered barrier system (EBS). Not surprisingly, this shift of the main retention barrier from host rock to EBS is a hardly acknowledged new paradigm. Enforced RD&D will be necessary to demonstrate the superiority of this concept. Extended variety in repository site selection and greater public acceptances will be worth the efforts.
The time has come to reconsider the contribution of innovative waste packages to the increased long-term safety of HHGW disposal in salt, clay and crystalline.
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