The neutrons generated by the plasma have to be used for tritium generation; for this scope, blankets are filled by lithium compounds (in solid or liquid form). Taking into account that for each installed GW of fusion power, about 153 g of tritium are consumed in a full power day, the future fusion power plant cannot depend on external tritium sources, but have to produce entirely the tritium necessary for the operation with an excess (several kilograms) to allow the start of new power plants.

Several blanket concepts have been proposed worldwide to accomplish the described functions using different breeder materials, coolants, structural materials and cooling and tritium extraction scheme. Table 4.1 shows a not-exhaustive list of published concepts. Looking at the ITER Test Blanket Programme [21] and at recent reactor studies in Europe (EU) [17], China [57], Japan [44] and Korea [45], only a few of these concepts are presently considered as possible candidates for a DEMO and a first generation of fusion power reactors. In general, concepts for a DEMO consider the use of a Ferritic martensitic steel at reduced activation levels (RAFM); almost all the DEMO developers are considering in the design this class of materials that can be adopted in a temperature windows of about 300–550°C (maybe also extended to 600–650°C), has the potentiality to withstand irradiation damages of more than 70 dpa, can be recycled after irradiation after 100 years and can have a chance of licencing in 20–30 years for reactor application [26]. EUROFER is the RAFM steel under development in Europe, F82H in Japan, CLAM in China and similar materials are also under development in Korea and India.

A large number of these concepts is based on the solid breeder technology (Helium-Cooled or Water-Cooled Solid Breeder concepts, shortly HCSB and WCSB, as reported in the table). Typical solid breeders that are proposed for these blanket concepts include Li-oxide (Li₂O) and ternary ceramics, like Li-orthosilicate (Li₄SiO₄), metatitanate (Li₂TiO₃), metazirconate (Li₂ZrO₃) or allumitate (LiAlO₃). None of these is able to entirely fulfil a list of ideal requirements for breeder materials that encompasses very quick and easy tritium release, high mechanical resistance, compatibility with other materials (including water in the air moisture for out-of-reactor handling), high Li density, simple fabrication and recycling routes, etc. [30]. At present, the interest of the blanket designer has been focalised mainly around Li₄SiO₄ and Li₂TiO₃, accepting compromises, but giving priority to criteria of low activation, relatively high operational temperatures (up to ∼1000°C), sufficient tritium release capability to allow tritium inventory control (only 200–300 g tritium in the reactor breeder) and the feasibility of manufacturing and recycling routes. Li₂TiO₃ presents better characteristics compared with Li₄SiO₄, in terms of mechanical properties, chemical compatibility with water moisture and other breeding materials (like Be₁₂Ti), and shows, however, a lower Li density that requires a higher ⁶Li enrichments for a comparable neutronic performance. To ensure sufficient tritium production, Be is added in the breeding zone due to its excellent neutron multiplication capability and the possibility of being used in solid form in a similar range of temperatures (at least up to 650–700°C).

Solid breeder blankets (with Be) present a simple way to produce and extract tritium; the solid material fills the blanket box that is formed by structures actively cooled for heat extraction; the breeder itself is not in contact with the coolant, but is purged by an independent low-pressure gas flow in order to extract the T produced. The boxes are typical double chambered components, where the high pressure coolant flows inside small channels (mainly realised in plate channels or tube bundles), while an independent low-pressure flow purge the breeder beds. The two chambers (see Fig. 4.1) are divided by thin walls for a very large surface. This results, however, in the permeation of tritium from the breeding beds to the coolant, which is a critical issue. Beryllium is required in a large amount (>3:1 volume ratio in comparison to the breeder) to ensure a sufficient tritium production due to an efficient neutron multiplication. Solid materials are sensitive to radiation and thermal damage, depleting in Li due to burn-up and cracking by radiation and thermal gradients. For these reasons, breeder and Be are used in these concepts mainly in form of a pebble bed with pebbles of diameters around 1 mm or less; pebble beds have relatively low thermal conductivity in the range of 1–5 W/m/K [47,61], but ensure a thermo-mechanical stability avoiding that properties can further deteriorate due to fragmentation. Usually the beds are arranged in thin layers of few cm. Fragmentation of pebbles also has to be minimised to avoid possible blockages of the purge flow with accumulation of tritium in parts of the blanket.

The chemical element which are typically present in the ceramics and the beryllium materials (eg, Li, Si, Ti, Be, O) will be not strongly activated during the irradiation in the reactor (tritium as the goal of the breeding reaction is not considered here as activation product; tritium issues are considered in Section 4.3). However, the presence of typical impurities in the ore (eg, U in several original Be ore) or contamination in the production processes (eg, Al in ceramic pebbles) can jeopardise these excellent intrinsic properties. Control of impurities is the key to low-activation inventory levels in the reactor and so to allow a recycling of the breeders in reasonable time to reuse it in the reactor [29].

One of the most studied versions of this concept is the Helium-Cooled Pebble Bed (HCPB) developed since the 1990s by KIT [11] and recently under further development in the EU Power Plant Physics and Technology (PPPT) studies [5]. It uses helium at 300–500°C and a pressure of 8 MPa as coolant, EUROFER as structural material and Li₄SiO₄ and Be pebble beds in the breeding zone. The current design is shown in Fig. 4.2. The breeder zone is formed by a sandwich structure of horizontal alternating Be and ceramics beds separated by cooling plates. The ceramics beds are formed by pebbles of the diameter of 0.2–0.6 mm, a packing factor of about 63% and use Li enriched up to ∼50% in ⁶Li; the fabrication process for the pebbles is melting and spry. Beryllium is used in form of pebbles of 1 mm diameters and 63% of packing; reference fabrication process is the Rotating Electrode Method. Helium coolant is distributed by a manifold system arranged in a robust back plate; helium cools first the first wall and then the breeding zone. The beds are purged with a ∼0.1 MPa helium flow that transport the tritium to the extraction and removal system located outside the Vacuum Vessel; an addition of H₂ (0.1%) helps per isotopic exchange the tritium release from the pebbles.

Water-cooled versions are also proposed in combination with solid breeders: in Japan, a WCSB concept is the reference concept for the ITER Test Blanket [16] and for the DEMO [56]. Water is used at pressure water reactor (PWR) conditions 290–335°C at 15.5 MPa. The purge mechanism of the breeder zone is similar to the other solid breeder concepts. Safety concerns in this case are for the possible accidental reaction between Be and steam with hydrogen production; for these concepts, alternatives to pure Be pebble are searched (in DEMO and FPP) in order to reduce the safety impact increasing the temperature margin to the exothermic reaction with steam (eg, use of beryllides; eg, Be₁₂Ti compounds [59]). Hence, in the DEMO version, the beds are arranged in parallel as shown in Fig. 4.3; as ceramic breeder Li₂TiO₃ is proposed, fabricated with wet chemistry and sintering [25]. The beds are realised by mixing Li₂TiO₃ and Be₁₂Ti that ensure better thermal (high thermal conductivities, high operational temperatures for Be) and mechanical properties of the beds. However, the presence of a larger amount of Ti (present now in the neutron multiplier) combined with the lower amount of Li density in Li₂TiO₃ aggravates the neutron performances that can be sufficient only with high ⁶Li enrichments and high packing factors (80% is considered in this study).

Liquid breeder concepts like the Helium-Cooled and the Water-Cooled Lithium Lead, shortly (HCLL and WCLL) concepts use the eutectic liquid metal PbLi (composition 15.8 at % Li with melting point at ∼235°C) as breeder; in these concepts, the cooling function is performed entirely (like in the solid breeder concepts) by an independent fluid, helium or water, respectively. PbLi circulates at low velocity (a few cm/s that corresponds to about 10 inventories of the blanket box moved in a day) so that no significant amount of heat is extracted by it. The PbLi recirculation aims only to transport the tritium outside the reactor for its extraction and for PbLi purification and chemical control. The recirculation of PbLi outside the reactor requires large loops that have to handle several thousands of tons of the liquid metals per hour; a recent assessment for the EU DEMO proposes eight loops located in the Tokamak Complex for a total of more than 5000-t inventory of PbLi and a recirculation capability between 0.6 and 1.6 t/s. A concern related to the use of such a liquid metal is the corrosion of the blanket steel structures, pipes and external components of the loop; a large amount of corrosion products could circulate with the risk of deposition in colder regions (or in magnetic fields acting as traps), causing blockage of flow or damage in moving components. Studies of compatibility with EUROFER and other RAFM steels [31] have identified possible critical corrosion rates in the temperature range above 470°C (depending also from the velocity). Hence, an effective control of corrosion products and a purification system of the PbLi have to be implemented in the design. Furthermore, this large amount of liquid metal will be strongly activated during the irradiation in the reactor. A particular concern is the production of volatile elements like Po, Hg or Ta as inherent components of the activation chain of Pb itself through the Bi formation [50]. Effective control of these end products or of the intermediate Bi formation has to be implemented. In addition, the corrosion products themselves can be a strong additional radiation source, increasing the necessity of purification loops, but aggravating operational safety concerns during maintenance.

This technology was used for the WCLL that was developed in the European Union in the 1990s by CEA [20]; Fig. 4.4 shows the design proposed at that time. The blanket boxes are long RAFM steel structures containing a pool of PbLi (hydrostatic pressure around 1 MPa); the box also contains a bundle of vertical tubes submerged in the liquid metal in which the water coolant flows. Water conditions are compatible with the PWR fission technology as mentioned for the Japanese WCCB. Safety concerns are related to accidental conditions in which the coolant loops breaks inside the liquid metal pool, causing an immediate pressurisation of the box enhanced by chemical exothermic reaction between steam and PbLi [27]. The study of this reaction is one of the key R&D launched in the European DEMO Programme to assess the safety performance of this concept. A helium-cooled version of this concept (the HCLL) that avoids this issue was developed in CEA staring from 2004 [54]. Helium is used at the same conditions as described for the HCPB blanket concept.

Interesting among the concepts reported in Table 4.1 is also the Dual Coolant. Here the steel structure is cooled by helium, while the PbLi breeder is moved at higher velocity in order to extract a large part of the heat produced in it (50–60%). The version presently developed in the EU PPPT studies [52] is a low temperature version in which the PbLi temperatures are kept lower than 500°C to avoid compatibility issues with materials in the liquid metal loop and components. High temperature versions (PbLi up to 700°C) have been proposed in the US ARIES Programme [38] and in the EU PPCS [42] with the aim to increase the thermodynamic efficiency of the concepts as future solution for more advanced power plants. The concepts are based on flow-insert channels able to decouple electrically and thermally the PbLi flow in blanket from the RAFM walls. This allows to reach suitable velocities of the PbLi for heat extraction at reasonable magneto-hydrodynamic (MHD) pressure drops and to increase the bulk temperature of PbLi without endanger steel structures that are kept to lower temperatures (<550°C) by the complementary helium cooling.

The use of a liquid metal breeder, like Li or PbLi, or a molten salt, like FLiBe, that combines the function of breeder and coolant, opens the way to design more advanced blanket concepts, where unique loops provide both heat and tritium transport. This simplifies the layout of the blanket system, but complicates the technology of a loop that has to remove heat with large tritium inventories. Concepts of this kind are included in Table 4.1 but are not presently proposed as DEMO (short-term) solutions. They also require different structural materials; eg, silicon-carbide composite (SiC_f/SiC) as ceramic material can be used to avoid MHD electrical coupling with liquid metals in the Self-Cooled Lithium Lead (SCLL) concept, or for very high temperature applications as in the High Temperature Solid Breeder (HTSB) blanket. Vanadium alloy is selected to contain fluid lithium in the Self-Cooled Lithium Vanadium (SCLV) concept.

The key technologies for the breeding blanket have never been adopted in existing devices, and in ITER, the first burning fusion machine, they also will not be used. Instead, ITER will use a shielding blanket that is not suitable for tritium production and electricity generation. However, breeding blankets will be test in ITER in the form of relatively small objects (mock-ups of six different concepts of less than 1 m³ each located in three dedicated equatorial ports) in the framework of the Test Blanket Programme. This program constitutes an important milestone in the path of the development of a blanket system for a first generation of FPP. Here, mock-ups of leading blanket concepts like the HCSB, WCSB and HCLL (Ref. [4] presents an overview of the EU TBM systems) that are developed for testing in ITER conditions. Comparing (see Table 4.2) these conditions to what is expected in, eg, EU DEMO specifications [17], it is clear which major limitation this test will have; the most important is related to the cumulative neutron fluence that will allow only a few indications of the blanket behaviour under irradiation at the beginning of the life. In fact, irradiation levels at end of life as expected in DEMO will exceed of a factor of 20–50 the irradiation experienced by a TBM and even more for a future FPP. TBM FW will also be receded from the FW ideal line of the shielding blanket; this means that phenomena related to the plasma–wall interaction will be not investigated for the tested blanket component. There was (see also Refs. [2,48]) and is still present a big discussion about chances and limits of the blanket test in ITER. What is mostly agreed is that, taking into account the large differences in the key parameter as shown in Table 4.2, the success of the TBM mission will be dependent on the capability to develop modelling and calculation tools able to correctly design the TBM, to explain the results of the tests and then extrapolate them to the current DEMO design. Modelling and tools in the field of neutronics, MHD (in the case of liquid breeder blankets) or pebble bed thermo-mechanic (for solid breeder blankets), thermo-hydraulics, electromagnetic and T transport have to provide these capabilities combined with sophisticated measurement systems able to operate in an ITER fusion environment [8]. An important feature of the TBM test is the adoption of materials relevant for DEMO, eg, breeder and structural materials (EUROFER in case of the EU Programme). Finally, fabrication technologies validate for the TBMs can be used to build DEMO blanket components. In addition, the return of experience in fields like licencing and design guidelines and standard will produce important knowledge for the blanket development for DEMO (see Ref. [49]). The design itself of the TBM as a mock-up of a DEMO blanket is derived from blanket designs taken for the DEMO development in each country. In the European Union, the TBMs are based on a common blanket architecture of the HCPB and HCLL developed during a DEMO study in 2004 [3]. Presently, the design architecture is in rapidly change [5] to adapt the blanket concept to the specific DEMO 2015 requirements; however, these changes are not touching the major characteristic of both concepts in term of temperature, pressure and velocity of coolants and structural materials or breeders.

As far as the second large in-vessel component, the divertor, is concerned, it is mostly excluded from the two main functions analysed in this section (at least in the more studied DEMO versions). The power deposited on the divertor, however, can be significant (up to 17% in some previous design [36]) and could be utilised as well for power generation, increasing the total efficiency of the fusion plant. The present difficulties in the design of solid targets suitable for exhaust particle fluxes of 10–20 MW/m² make the thermo-hydraulic integration of such a circuit in a power generation system very challenging. The solution proposed for ITER and tentatively adopted as start configuration in many DEMO designs is based on the use of water at a low temperature (100–150°C in ITER). As this design is based on the use of a Cu-alloy (ie, CuZrCr) for the cooling tubes, the temperature range can be only slightly increased for DEMO (up to ∼200°C) without enter in operational temperatures not suitable for this material. Also, at this temperature there are many concerns about the lifetime of this component under neutron irradiation [58]. In comparison, the blankets are designed to work at a coolant temperature higher than 300°C to allow for efficient electrical energy production, so that separate coolant loops for the divertor systems have to be designed. As the heat removed by the divertor coolant loops cannot be integrated into the steam generators of the Rankine system, the only reasonable application is to preheat water coming from the condenser. Some present designs of DEMO are considering discharging almost entirely this heat in the plant heat rejection system avoiding the technical complications due to its integration.

Even if it can be technically possible to put breeding materials in the divertor (eg, in the cassette structure of an ITER-like divertor), this possibility has never been exploited in plant design. The reason is to avoid further complications and additional requirements (eg, T requirements) to a design of a component that has already so many critical issues to be solved. A possibility could be to attach target plates (and to realise exhaust pumping) directly on the blanket using then the cooling loop of it and the dedicated tritium extraction systems. This idea is partially followed in the PPPT DEMO design [17] where the dimensions of the divertor have been reduced drastically, and elements of this component (ie, buffers) are directly integrated in the blanket design.

Fig. 4.15 shows schematically the cycle of tritium in a fusion reactor. A first stored tritium amount (several kilograms) constitutes the starting inventory for the plant (in ‘tritium storage’). It should provide sufficient tritium for the first days as soon as the cycle arrives in a stationary condition. The tritium and deuterium are injected in the plasma. As already mentioned only a part of them will burn, but the rest will abandon the plasma and will be extracted by the pumping systems, reprocessed and again sent to the storage to again feed the injection system. This constitutes the so-called ‘inner fuel cycle’. The ‘outer fuel cycle’ starts with the tritium production in blanket under neutron reaction; this tritium is collected and processed to remove it from the different carriers (eg, purge flow, liquid breeder, coolant). Finally it is delivered to the separation system to be integrated in the inner cycle. Impurities are removed from the processing units and sent to the waste management system.

Tritium inventories can be built in the different stages of the cycle depending on the time constants of the different processes and loops (eg, diffusion processes in the breeding materials) or by trapping mechanisms (eg, in metal). A general issue of an FPP is to minimise the tritium inventory in the different parts of the reactor (in the VV as well in the tritium plant); this is a safety concern related to possible mobilisation of these inventories in case of an accident and its release in the environment. For this reason, the tritium cycles must be optimised to reduce time constants in the tritium processes and then minimise the inventories (see Ref. [12] for more details).

The production of tritium is ensured by the presence of lithium compounds in the blanket that can react with the neutrons produced by the thermonuclear reaction. ⁶Li and ⁷Li produce a tritium atom for each neutron absorbed, and in the case of ⁷Li, a further neutron is released in the reaction and available for further reactions. Taking into account that not all neutrons produced in the fusion reaction will react with Li in the blanket and that other materials will absorb a large amount of them, a surplus of neutrons generated in multiplying reactions is indispensable to maintain a positive tritium balance necessary to attain tritium self-sufficiency. The ratio between the tritium produced in the blankets and the tritium consumed in the fusion reactions in the plasma (that coincides with the number of neutrons emitted in the thermonuclear D-T reaction) is called Tritium Breeding Ratio (TBR). The TBR must be greater than one to compensate the T burnt in the plasma. How large the TBR should actually be during the lifetime of the FPP is dependent on the ‘dynamics’ of the entire fuel cycle for the D-T plant, as shown in Fig. 4.13. It involves time constants and efficiencies of many subsystems and has to take into account possible losses of tritium in the different parts of the process and tritium trapping in the materials (see Refs. [55] and [15] for more details on the tritium balances for ensuring self-sufficiency). In addition several other requirements that have to be accounted for, such as duplication time (time necessary to produce the tritium amount necessary to start a new FPP) and storage capabilities; also the natural decay of tritium (a half-life of 12.3 years in the machine has to be considered).

The TBR is a characteristic of the used blanket system and it is determined essentially by the geometry and material composition of the blankets; it can change during the lifetime of blanket components as effect of the Li burn-up and nuclear transmutations. Theoretically, lithium with a natural isotopic composition is capable to provide sufficient TBR, assuming however pure lithium (ie, Li fluid metal) in combination with blanket structural materials with a very low neutron absorption such as vanadium or SiC. A V-Li blanket concept was earlier studied in the US and Russia but almost abandoned in recent studies (see Table 4.1).

To improve the neutron balance, a neutron multiplier is added to the breeder. Such a multiplier material provides ‘neutron multiplication’ through (n, 2n) reactions. Effective multipliers are, first of all, Be and, second Pb. In this case, ⁷Li is not required to produce tritium and a better neutron economy is achieved by increasing the fraction of ⁶Li beyond the natural atomic abundance of 7.5%. Enrichments of 50% up to 90% in ⁶Li may be necessary to compensate parasitic neutron absorptions in additional materials in the breeding zone. The HCPB design described in Section 4.1 is a good example of this technology. The spectrum is relatively well moderated (the neutron multiplier Be is also a good neutron moderator) so that attention must be paid on any material in the blanket that can absorb neutrons. An example is the EUROFER itself that efficiently absorbs neutron at energies of a few tens of keV and below. As the design of a blanket breeding zone is a compromise, among the tritium breeding capability, the thermo-hydraulic design for heat extraction and structural-mechanical design, the amount of steel must be reduced as much as possible while still achieving the targets of mechanical stability (eg, against pressure, thermal and EM loads).

A control of the TBR level during the life of the reactor can simplify the goal to keep the inventory of tritium as low as possible in the storage system; it could allow reducing the tritium production if the tritium level increases in the system (and the contrary). This control can be realised in a simple way only for liquid breeder blankets. Here, the external circulation of the breeder allows a chemical control of the liquid composition. This includes the possibility to tailor the level of ⁶Li enrichment in the breeder, according to the requirements. A similar possibility doesn't exist in solid breeder concepts where the design should primarily ensure the correct level of TBR necessary for the entire lifetime of the blanket. Other possibilities (like the inclusion of a neutron poison or control roads) are imaginable, but they have never been proposed in the published blanket designs.

The tritium generated inside the breeder has to be efficiently extracted. In solid breeders tritium is primarily formed as atom inside the ceramic grid; it has to reach the grain and pebble external boundary, to recombine as molecule to be extracted by the purge gas. The process of extraction from the solid material is complex and different phenomena concur and delay the tritium release. For design purpose a quantity is used to characterise this process, the so-called Tritium Residence Time (TRT); it can be defined as the time constant associated to the release process from the pebble and can be determined experimentally in reactor from thermal transient during irradiation [32]. It is strongly correlated to the temperature of the material; correlations gained in experiment show an exponential dependence with the temperature. Other dependencies are from material, microstructure and composition and from purge gas chemistry (addition of H₂ or H₂O helps to extract tritium per isotopic exchange [41]). Typical values measured for Li₄SiO₄ are ∼1 day at 350°C and less than 1 h at 600°C in a reference purge flow of 0.1 MPa of He with 0.1% H₂. As noted before, large time constants are responsible of large inventories; inventories of 200–300 g of tritium for an amount of ceramic of about 100 t have been calculated in reactor assessments.

As soon as the tritium is extracted from the solid materials, the helium purge-gas transports the tritium outside of the reactor where it is removed from the He carrier and concentrated before being sent into the fuel cycle. Different processes are under investigation to treat the purge gas, and a selection of the most efficient technologies has to be done. The reference process [14] adopts batch cryogenic technology (with molecular sieves and cold traps operating in alternating adsorption and recovery regimes), but more advanced processes relying on membrane, and catalytic membrane reactor technologies are under evaluation to operate in a continuous regime, with a lower time constant and energy saving.

The extraction for liquid breeders is different; eg, in PbLi, tritium is generated during the presence of the liquid metal in the reactor and then is recirculated outside the reactor. Hence, the PbLi is used as a carrier to transport tritium outside of the reactor for its extraction. During its flow inside the blanket, tritium is generated in atomic form in the liquid; its movement is dictated by the movement of the liquid that flow in a complex MHD environment and by diffusion leaded by concentration gradients. Tritium can be trapped in He bubbles (He is formed in PbLi by side nuclear reaction involving Li, in fact each Li + n reaction produces 1 T and one He atoms). Several processes are under investigation for an efficient tritium extraction; they include Gas Liquid Contactors (GLC), Permeation Against Vacuum (PAV) or Vacuum Sieve Tray (VST) methods. Again, the aim of the current investigation is to provide efficient, safety, reliable and economic processes. The subsequent phase of concentration is simple and in case of PAV practically only T (with low H, D impurities) is extracted and can be delivered directly to the fuel cycle.

In considering technical challenges related to the tritium production, the several safety concerns that are related to these topics must be mentioned. As already mentioned, the limits of inventories have to be strictly controlled. Usually only a few kilograms of tritium are allowed in VV or in the fuel cycle to avoid a potential large release requiring evacuation of population. Tritium can constitute a risk during normal operation due to the possibility of being released in the environment. The limits imposed to ITER and tentatively assumed also for a future power plant are considering not more than 1 g per year released, which means for a 2-GW fusion power plan with an availability of 75% (with a production/consumption of ∼84 kg of T in a years) only a 10⁻⁵ fraction of the tritium handled in the facility. The major possible of tritium escaping from the plant is through the coolant system, eg, in the HCLL, the proximity in blanket of coolant and PbLi loop cannot exclude permeation of tritium in the coolant [13]. This tritium can reach the steam generator and then permeate/leak in the secondary loop. Several protections are designed to avoid it, including Coolant Purification Systems (CPS) acting in the PHTS to control the level of contamination of the coolant (see Fig. 4.16). In addition, permeation barriers or chemical control are proposed to reduce the quantity of tritium moved per permeation in the blanket, in the steam generator, or additional intermediates loops, as already mentioned in Section 4.2.

Safety concern will also include waste management of tritium-contaminated components and materials.