Editorial: Advances in Nuclear Fusion Energy and Cross-Cutting Technologies

1. Introduction

The current energy model, which is based largely on the use of fossil fuels and has a direct influence on global warming and climate change, presents serious problems of unsustainability in the long term. This, together with the growing energy demand (estimated to grow by up to 40% in the next 25 years) has led for years to the search for new energies, among which nuclear fusion stands out for its great potential. Fusion energy has been one of the greatest scientific challenges of humanity since the 1950s. It is a safe, sustainable and massive energy source, based on an inexhaustible and distributed fuel, suitable to complete the global energy mix. Record fusion energy production using magnetic confinement has recently been demonstrated in the European JET tokamak, and a positive plasma energy balance has been obtained in the NIF inertial fusion device. In the Chinese tokamak EAST, a stable plasma has been maintained up to 1000 s, and in the stellarator W7-X, plasma energy in excess of 1 GJ has been achieved. The ITER project, currently the world’s largest investment project under construction (over EUR 30 billion), where countries from all over the world (Europe, Japan, Korea, USA, Russia, China, India…) are joining forces to demonstrate the viability of this energy source, is about to complete the assembly of the tokamak. Thanks to these achievements, interest in fusion has grown enormously and industries and private investors have joined the effort to succeed in it.

In the changing scenario regarding the development of fusion devices, breeding blankets (BBs) will play a key role in the coming decades and will be the number one priority of the new European roadmap to fusion. The breeding blanket is the key system in a fusion reactor since it must perform a number of essential functions; among them, the most important is to generate tritium (T) through neutron reactions with the lithium (Li) target of which the BB is made.

The reaction used currently to demonstrate fusion is that of deuterium (D) and tritium (T), both isotopes of hydrogen. When they fuse, they generate a helium atom, a neutron and the energy that will later be used to transform it into electricity. While deuterium is abundant in seawater, any large fusion device must generate its own tritium fuel, since tritium is not abundant in nature due to its radioactivity and relatively short half-life (12.32 years). Therefore, for fusion to become a reality, it is essential to develop technologies that allow the T needed for fusion to be produced. A limited amount of T can be generated in CANDU fission reactors to start up the fusion plant, but this is an extremely costly process. Therefore, in a fusion reactor tritium must be generated in sufficient quantity by capturing fusion neutrons in lithium-bearing materials (in solid or liquid form) with a margin (provided by the neutron multiplier, Be or Pb in most concepts) necessary to compensate different losses. Hence, the BB must be designed to allow an efficient extraction of tritium and minimize losses.

According to the current European Roadmap, the reactor DEMO is expected to act as a component test facility for breeding blankets. While operating with a near full-coverage BB, to be installed from day one to achieve tritium self-sufficiency, extract thermal energy and convert it into electricity, it will also be used in a limited number of segments for the testing and validation of more advanced, high-performance BB concepts with the potential to be implemented in a future First-of-a-Kind (FoaK) fusion power plant (FPP).

In 2019, the Helium-Cooled Pebble Bed (HCPB) and the Water-Cooled Lithium Lead (WCLL) BB concepts have been selected for the EUROfusion DEMO tokamak project as the most mature and technically sound to be used as a “driver blanket” for DEMO. Meanwhile, the Dual-Coolant Lithium Lead (DCLL) BB has been identified as a potentially more attractive “advanced blanket” for a future fusion power plant of both tokamak-type (the dominant magnetic confinement concept adopted for example in JET, ITER and DEMO) and potentially stellarator-type (concept adopted for example in TJII and W7X).

This Special Issue of Energies, entitled “Advances in Nuclear Fusion Energy and Cross-Cutting Technologies”, belonging to Section “A: Sustainable Energy”, contains five papers that highlight recent advances in fusion technologies in Europe especially related with tritium breeding technology (breeding blankets and auxiliary systems). Contributions from both leading researchers and emerging investigators were solicited.

2. Published Papers Highlights

The conditions of temperature, pressure, magnetic fields, corrosion and neutron damage that will occur in combination in breeding blankets are not comparable to any other environment. Being able to test such components under equivalent integrated conditions is critical. However, to this day, it is still an unresolved challenge. Therefore, until it is possible to obtain experimental data under real fusion conditions, computer simulation is given the responsibility of evaluating the viability of designs. For this, this compendium of papers that deal with the simulation and various analyses of different BB concepts is particularly relevant:

1. The HCPB line in Europe has been led by the Karlsruhe Institute of Technology (formerly known as KfK and FZK) in Germany since the 1980s. The HCPB blanket concept uses pressurized helium gas (at 8 MPa) as a coolant, lithium ceramic as a tritium breeder, beryllium-containing material as a neutron multiplier material, the reduced-activation ferritic martensitic steel Eurofer97 as a structural material, and tungsten as armor on the first wall (FW). The tritium bred in the HCPB BB will be extracted through a purge gas system.

In their article, G. Zhou et al. [1] present a description of the European DEMO HCPB breeding blanket’s design and its evolution. The design activities and performance analyses are summarized, showing that the HCPB BB is a high tritium-breeding and robust blanket candidate for the European DEMO. Challenges, solutions and technology R&D activities to maturate the HCPB BB are also presented with constructive criticism. In addition, three alternative concepts of interest are explored as the CO₂ Cooled Pebble Bed (CCPB) concept, the Helium-Cooled Molten Lead Ceramic Breeder (MLCB) concept and the Water-Cooled Lead Ceramic Breeder (WLCB) concept, which aim to combine the advantages of both the WCLL and HCPB concepts.

2. An intense research effort has been conducted since 2014 throughout the EU, under the umbrella of the EUROfusion consortium and led by the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), to reach a sound WCLL BB design capable of matching all the design requirements propagated from the plant level to the level of the BB system. In particular, the WCLL BB must withstand thermal and mechanical loads it undergoes under both nominal and accidental scenarios, fulfilling the structural design criteria prescribed by the RCC-MRx code, currently the reference code for the DEMO BB’s structural design. In this way, the WCLL BB can operate in any loading conditions without incurring the condition of incipient structural crisis, which may jeopardize its integrity.

In their work, A. Gioé et al. [2] present a coupled thermofluid dynamic and structural assessment. In particular, attention has been paid to the top cap region of the WCLL BB’s central outboard blanket segment as its design has resulted in being particularly challenging and significantly different from the design of the other regions. The obtained results were compared with those gathered from the analogous study performed adopting a pure FEM approach. The results show that, globally, the pure FEM approach allows more conservative results to be obtained than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible to adopt the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalize the most promising.

3. The design of safety systems is a key part of the conceptual design phase of the EU DEMO tokamak fusion reactor including the vacuum vessel pressure suppression system, which will be responsible for the mitigation of accidents such as an in-vessel Loss-of-Coolant Accident (LOCA). The LOCA is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyze the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localized quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked.

One of the possible initiating events of this type of accident is a break in the first wall, cooled by a high-pressure coolant, which, when released inside the vacuum vessel, would increase its pressure above the operating limit of 2 bar.

In their article, M. Spró et al. [3] develop a 3D CFD transient model to simulate the evolution of the transient when the released coolant is pressurized water. The model represents an extension of a model previously developed for a helium-cooled FW, taking into account the additional complications deriving from the flash boiling that occurs when pressurized water is released inside the vacuum vessel. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident’s evolution in the water case is less violent than in the helium case.

4. The blanket region, responsible for tritium breeding, is characterized by high tritium concentrations, high temperatures and metallic surfaces with large heat transfer capacity in which tritium can permeate. Therefore, the problem of tritium permeation and the resulting tritium content in the primary coolant are of great relevance for DEMO.

In the article by V. Narcisi et al. [4], the tritium permeation rate from the blanket into the coolant is assessed for the pre-conceptual design of the Water-Cooled Lead–Lithium variant, and possible mitigation strategies are suggested. As this Special Issue is also devoted to cross-cutting technologies, the work started with a literature review of experience with CANDU in the management and treatment of tritium, focusing on the maximum tritium concentration target in the primary coolant and on the water leak rates, which are relevant for tritium emissions and the dose to workers and the public. Afterwards, a preliminary assessment of the maximum tritium concentration target in the DEMO primary coolant was performed and different strategies (off-line, on-line and hybrid) for the water coolant purification system coupled with the DEMO operating scenario were analyzed. The hybrid solution results the most viable strategy if a tritium concentration lower than 1 Ci kg⁻¹ must be ensured. However, this choice implies a non-negligible amount of tritiated water to be stored in dedicated areas. Further assessments and safety evaluations are thus encouraged to check the full viability of this strategy.

5. In the roadmap towards a commercial fusion power plant, the European efforts are mainly focused on magnetic confinement devices of the tokamak type, more advanced worldwide under the technology and engineering aspects—with several big projects in sight (ITER, JT-60SA, DTT, etc.). Regarding stellarators, substantial progress has been made in understanding their physical aspects, especially thanks to the operation of the W7-X stellarator. To also bring stellarator engineering to maturity, a specific program has been defined in Europe to develop the HELIAS (HELical-axis Advanced Stellarator)-type stellarator as an alternative to the European DEMO tokamak main line. For that, and to exploit the large experience in BB design for the DEMO tokamak, the Spanish Research Centre for Energy, Environment and Technology (CIEMAT) is leading the development of a Dual-Coolant Lithium Lead breeding blanket for the HELIAS device.

In their article, D. Sosa et al. [5] explore DCLL BB-specific design solutions to cope with the challenges associated with stellarators. Previous MHD analyses, concluded in a BB quasi-toroidal segmentation, would nonetheless complicate the traditional remote handling of the BB through ports. Such controversy has motivated the search for a brand-new solution: the use of a fully detached first wall based on liquid metal Capillary Porous Systems (CPS) to switch the maintenance problem to small FW panels, instead of entire BB segments, that could be manageable through ports. This strategy could allow a reduction in the damage produced inside the BB, increasing the machine’s availability while maintaining the tritium-breeding performance required for a BB. The preliminary studies demonstrate that the use of FW CPS would drastically reduce the radiation damage received by the blanket by 29% along with a small decrease in the Tritium Breeding Ratio (TBR) around 4%. Such an outcome has shown the existence of margins for further development of the FW CPS concept for HELIAS, as they have been not found, at least to date, to be significant showstoppers for the use of this technological solution.