Testing needs for the development and qualification of a breeding blanket for DEMO

The purpose of this paper is to critically review the testing and qualification strategy for the DEMO breeding blanket in Europe, identifying the potential risks and weaknesses and recommending, where necessary, the changes required to strengthen or accelerate the programme. Based on the information presented in this paper, a risk mitigation strategy is required to reduce the technological, licencing, and regulatory risks associated with the performance and reliability of the breeding blanket technologies for DEMO. An attractive mitigation option, which was originally proposed more than 30 years ago but not implemented, is to perform testing and qualification of breeding blanket technologies and design concepts in a dedicated nuclear plasma device that serves as a 14 MeV volumetric neutron source.


Introduction
The breeding blanket is one of the most important and novel technical systems of the European DEMO and any Deuterium-Tritium fusion power device to follow ITER.The purpose of this paper is to critically review the testing strategy for its development and qualification.
The reader can refer to the annexes at the end of this paper as listed below for further details on some technical aspects of the design and operation of the breeding blanket.In particular, Original content from this work may be used under the terms of the Creative Commons Attribution 4.0 licence.Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI.
• the functions of the DEMO breeding blanket are described in Annex 7.1.• the breeding blanket design concepts that are being considered for DEMO in Europe are described in Annex 7.2.• the main research and development (R&D) issues to be resolved for the breeding blanket are described in Annex 7.3.• the issues arising from the limited availability of tritium supplies are described in Annex 7.4.• the preliminary results of the recent exploration of feasible design windows for a tokamak-based volumetric neutron source (VNS) are described in Annex 7.5.
Overall, this paper serves to facilitate the understanding of the crucial testing needs for the DEMO breeding blanket in Europe to fulfill its functions from reliability and regulatory perspectives.

Breeding blanket R&D status
The role of the R&D programme is to address and resolve the feasibility and attractiveness issues of the breeding blanket concepts being considered.Feasibility issues are those issues that could reduce or close the available design window or lead to an unacceptable reliability, availability, or lifetime; and/or unacceptable safety impact.Attractiveness issues could instead reduce the system performance and lifetime; increase system cost; and/or have less desirable safety or environmental implications.
The rationale and technical foundation of the breeding blanket R&D programme in Europe are based on technical studies conducted over the past three decades including, but not limited to [2-9].The major elements of this programme can be summarised as follows: • Basis R&D activities with tests in non-nuclear facilities together with computer modelling and simulations.This includes also supporting tests of materials specimens (e.g.structural materials, breeders and multipliers) in fission material reactors.• Testing of dedicated Test Blanket Modules (TBM) in ITER • Qualification testing directly in DEMO.

Role and limitations of R&D in non-fusion facilities
The classification of non-fusion facilities in R&D, together with their respective roles and limitations are summarised as follows in [1]: 'Non-fusion facilities can be classified into (a) non-neutron test stands, (b) fission reactors and (c) point neutron sources.Table 1 summarizes the capabilities of non-fusion facilities for simulation of key conditions for fusion nuclear component experiments.The most important conditions are (1) neutron effects (i.e.radiation damage, tritium and helium production), (2) bulk heating (nuclear heating in a significant volume and associated thermal gradients), (3) non-nuclear conditions (e.g.magnetic field, surface heat flux, particle flux and mechanical forces), (4) conditions for simulating thermal-mechanical -chemical and electrical interactions and (5) conditions for integrated tests and synergistic effects.
An important observation that can be drawn from the table is that non-fusion facilities are not able to simulate partially integrated or fully integrated conditions.Their capabilities are limited mostly to single-effect tests and only few can carry out multiple-effect multiple-interaction experiments.
Table 2 discusses the contribution of non-fusion facilities to address critical issues for the breeding blanket.One important observation to be drawn is that representative information on the characterization of failure modes, effects/rates or tritium self-sufficiency cannot be obtained from testing in non-fusion nuclear facilities.Therefore, the validation of blanket concepts requires a complete qualification process that involves both non-fusion and relevant plasma fusion facilities to cover the full spectrum of topics to be demonstrated.
Non-neutron test stand facilities: they have played an important role in R&D and should continue to be exploited in the future.They are mainly used to screen materials and design concept options and can help reduce the risks and costs of more complex integrated tests in a relevant fusion environment.
Fission reactors: they provide neutrons in a moderate volume and are thus suitable for experiments aimed at investigating many single and multiple effects such as: irradiation of structural or high-heat flux-material specimens, irradiation experiment on the thermomechanical behaviour of a ceramic breeder bed and tritium release or tritium inventory associated with a breeder material.Irradiation experiments in fission reactors have also helped in understanding the tritium breeding and behaviour in ceramic breeder samples [10].However, testing in fission reactors suffers from serious limitations including small test volumes and the lack of certain fusion-related conditions, such as relevant radiation damage parameters, surface heat flux, and magnetic field effects.Another set of problems arise from the difference between the fission and fusion reactor neutron and secondary gamma-ray spectra.These differences lead to difficulties in simulating the magnitude, profile, and time-dependent behaviour of reaction rates such as helium and tritium production, as well as power density and atomic displacements.Despite these limitations, irradiation of material specimens in fission reactor tests have proven to be very useful to measure the degradation of material properties.In addition, they were suitable for some multiple effect tests needed to establish the proof of principle for certain designs Finally, accelerator-based neutron sources produce neutrons in a volume that is much smaller than typical fission reactors.Deuterium-tritium sources: they produce 14 MeV neutrons but are limited by target fabrication and cooling considerations to neutron fluxes that are orders of magnitude lower than that in a fusion reactor with a 1 MW m −2 neutron wall load.Examples of these facilities are the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Agency and the Frascati Neutron Generator (FNG) at ENEA in Italy.FNS and FNG have been used extensively for neutronics experiments aimed at the validation of codes and data and identifying uncertainties in calculated nuclear parameters.However, these facilities have limited neutron intensities (<∼10 13 n s −1 ) and result in very low neutron flux even at small distances from the target (∼3 × 10 14 n/m 2 s @ 5 cm) that is more than five orders of magnitude lower than the flux at the first wall of fusion reactors.The lifetime of the target is generally limited to <100 h resulting in extremely low fluences.The flux profile, spectrum, and gradient in these devices are not prototypical of conditions in fusion blankets.The low flux level cannot yield nuclear heating and reactions at rates that allow engineering tests other than neutronics tests.These facilities can be used for verifying the prediction capability of present codes and databases for tritium production in solid breeders (or liquid breeders) to help assess the tritium self-sufficiency issue and to generate safety factors for design purposes [1]'.
Deuterium-Lithium sources: like IFMIF [11], Internation Fusion Materials Irradiation Facility/DEMO Orientated Neutron Scource (IFMIF/DONES) [12,13] or other similar facilities [14,15] as well as fusion-dedicated material irradiation facilities [16] are very important and will contribute substantially to gaining information about the degradation of material properties under neutron exposure at high neutron fluence that DEMO blankets will experience.This is deemed mandatory for materials qualification and the development of licensing codes.However, in these sources the irradiation volume is very small (less than 0.5 litres in DONES for highflux irradiation experiments) and it will therefore not provide information on design reliability or effectiveness of tritium breeding and extraction.

The role of fusion devices
Since its inception, more than 40 years ago, the European fusion strategy has been articulated around the design, construction, and operation of three main devices: JET, ITER (previously NET), and DEMO.Each of these devices were supported by a strong R&D accompanying programme.
Prior to the start of the ITER Conceptual Design Activities, there was a debate as to whether to consider physics research and technology developments for DEMO separately.This was advocated, for instance, in 1985 and later [1,17,18].Specifically, plasma VNS designs based on mirrors, reversed pinches and tokamaks were proposed to address the fusion nuclear technology R&D aspects.These concepts were all based on a beam-to-target fusion approach where a near Maxwellian background plasma is sustained against energy and particle losses by neutral beam injection, and fusion reactions principally occur between the fast and target ions as the beam thermalises via Coulomb collisions [19][20][21].
These device concepts were small (e.g.major radius R of 2-3 meters), low power (e.g.<100 MW) and low tritium consumption, and the designs emphasized maximum access to the fusion core.
In Europe, though, as in most other countries with significant fusion programmes, the decision was taken not to have a major, dedicated technology testing facility, but to use ITER directly for both physics and technology developments including the need to carry out extensive nuclear component testing relevant to a future fusion reactor.
A fundamental problem with this latter approach is that physics testing requires relatively large fusion power (e.g.300-600 MW in ITER) to have performing plasmas.Physics testing alone requires low neutron fluence (<0.01 MWy m −2 ).On the other hand, nuclear technology testing requires only low fusion power (∼20-50 MW) but needs high fluence (>1 MW y m −2 ) [22].The combination of high power for physics testing and high fluence for nuclear technology testing in a single device leads to very high tritium consumption (for example >110 kg per full power year in DEMO).Such large amounts of tritium can only be provided if the device has its own self-sufficient tritium breeding blanket that operates reliably from the very beginning of operation.

Results of the ITER TBMs programme
The testing of TBMs in ITER is viewed as an essential step to reduce the technical risks and uncertainties associated with the demonstration of power extraction and tritium breeding technologies essential for a DEMO fusion power plant.The licensing, operation, and maintenance of the TBMs and their Ancillary Equipment Units are deemed to be valuable in providing information and support for licensing a breeding blanket that breeds tritium.However, large gaps are expected to remain even with a successfully completed TBM programme.The unique testing conditions of the ITER TBM are: • 'Plasma exposure with typical conditions (e.g.plasma radiation, particle loads, etc.); • Strong and spatially complex magnetic field (∼5 T); • Typical off-normal plasma events such as disruptions, Edge Localised Modes, vertical displacement events, etc.; • Actual fusion neutron flux and energy spectrum; • Prototypical ratio of gamma-ray heating to neutron heating; • Tritium production and nuclear volumetric heating in a large volume with spatial gradients; • Beginning of life radiation damage (up to 1 dpa 2 ) with spatial gradients • Strong confinement of radioactivity, allowing build-up of realistic tritium concentrations.
TBM testing in ITER is defined to provide limited experience and scientific exploration of blanket performance and response in a representative fusion environment.The role of TBM testing is: (i) to provide an initial exploration of coupled phenomena in a complex fusion environment, including: investigating unexpected synergistic effects, calibrating results from non-fusion tests, and providing data for model improvement and simulation benchmarking; (ii) to assess the impact of rapid property changes in early life; and (iii) to develop fusion environment experimental techniques and test instrumentation [23]'.
In addition to the key breeding blanket information from the TBM program, it should be noted that important technology information will be obtained on a large prototypical scale from the ITER base shielding blanket, albeit with a non-DEMO relevant structural material, which would be applicable to DEMO and beyond, including (as an example list and not exhaustive) [24]: The DEMO breeding blanket development programme in Europe is directly linked to the successful execution of the TBM programme.Any delay or reduction of the scope of the TBM programme (i.e.reduction of the neutron fluence) will impact the DEMO programme.Likewise, any substantial change in the DEMO breeding blanket design vis-à-vis the solutions to be tested in ITER (either materials combinations or design configuration) will include additional uncertainties.

Low design maturity and technology readiness
Despite its criticality to the development of fusion power, the maturity of the breeding blanket is still very low [25].No breeding blanket has been built or tested to-date.There remain feasibility concerns and performance uncertainties in all currently explored concepts.Substantial R&D is needed to fill the remaining outstanding gaps and to facilitate a down-selection, which at this stage would be far too premature.Given the importance of the breeding blanket and potential associated regulatory aspects, there is an urgent need to accelerate its development and testing programme.The development and qualification of sound breeding blanket concepts are, without any doubt, on the critical path for deploying fusion reactors.
Like other technologies developed through a sequence of activities, often iteratively, a guide to the maturity level of the DEMO breeding blanket design is obtained using a technical readiness level (TRL).This was derived from the NASA model to assess the readiness of space technologies during development [26].
Figure 1 shows that readiness of the technologies for the most mature DEMO breeding blanket concepts in Europe -Helium Cooled Pebble Bed (HCPB) and Water Cooled Lithium Lead (WCLL) -is still relatively low (3)(4).This readiness level was achieved through research in existing technology facilities and irradiation of specimens in fission reactors.Increasing the readiness further from 4 to 6 would require the successful completion of the full TBM programme, which plans to achieve up to 1-2 dpa of fluence during the extensive nuclear operation phase.If the TBM programme would be descoped to a much more limited end-of-life fluence, there would be implications for the follow-up qualification phase, as discussed below.One can see from figure 1 that, independent of the TBM results from ITER, there is still a substantial gap to demonstrate the performance and qualify a breeding blanket for DEMO to increase the TRL from 6 to 8.

Foreword
The current strategy of the EU Fusion Roadmap is to rely on the results of the ITER TBM programme and to use DEMO itself as a qualifying device to operate from the beginning with a self-sufficient breeding blanket.This also implies the need to operate DEMO as a qualifying device for very long times to confirm its ability to achieve adequately high levels of breeding blanket reliability and machine availability to inform the design of future commercial reactors.In the absence of any further breeding blanket qualifying device, DEMO would also need to qualify the performance and reliability of more advanced breeding blanket concept for commercial applications to be tested in dedicated ports or segments [27].
However, installation of a breeding blanket in DEMO without prior fusion testing is found to result in high risks of not attaining the required tritium self-sufficiency, blanket system reliability and an adequate device availability.In addition, there are risks connected to the licensing of a first-ofa-kind system (such as the breeding blanket).As a matter of fact, it is highly questionable whether, from a regulatory standpoint, a safety authority will issue a license to operate a DEMO plant implementing a large full-scale nuclear system like the breeding blanket which has never been tested and qualified in relevant conditions in advance.As the breeding blanket impacts the safety functions of confinement (tritium control, heat removal), it is generally considered a safety feature subject to a rigorous qualification prior to licensing.
To minimise the remaining technical and regulatory risks for the deployment of fusion power described above there is a need to revise and strengthen the breeding blanket development strategy.
It is useful to look at some of the rationale which was behind nuclear fission energy technology developments over the past four decades.For example, there is a rather striking analogy between the design of the first-wall/breeding blanket system in fusion and fuel assembly in fission reactors.Albeit with different life-limiting phenomena, they both experience very high surface and bulk heat fluxes and are subject to large neutron damage (see Annex 7.3).The development and qualification of fuel rods, requiring long-lead-times (>20 years) mainly for irradiation testing, was needed to bring a fuel design from the initial concept through licensing [29].The qualification of the breeding blanket to validate the performance and to sufficiently reduce the safety and reliability uncertainties would not be shorter.
In light of the above, there are compelling arguments for advocating a change of strategy that re-introduces a nuclear plasma device that serves as a 14 MeV n-source (VNS) for testing and qualification of the breeding blanket to be run in parallel to both ITER operation and the DEMO design process.These device concepts are small (e.g.R 2-3 metres), low power (e.g.<100 MW), low tritium consumption (to operate relying at least in the very initial phase on external tritium supplies and without requiring from the very beginning of operation a fully integrated and tritium self-sufficient breeding blanket) and with an easy access to the fusion core.Some of the approaches investigated in the past had sufficient merit and are being re-evaluated within EUROfusion with the aim to identify the most feasible concepts.Based on the preliminary results of this important assessments, expected by the end of 2024, an ambitious plan for the realisation of this device, to be constructed as early as possible and run in parallel to ITER operation and the DEMO design process, should be prepared and implemented.
It should be emphasised that what is advocated here is not a plasma physics experiment.Energy 'break-even' in any beam-heated reactor was recognised to be attainable with far less stringent plasma performance than in other fusion reactor schemes [19].The so-called Two Energy-Component Tokamak Reactor described in [20,21].With approximate energy break-even and with a high fusion power density even at relatively low plasma temperature, was proposed as potential producer of large neutron quantities even when operating with an overall electrical energy deficit.For this application, the most important system parameter is fusion power density (that is, neutron production rate) rather than power multiplication.Conversely, to increase the energy gain (Q) significantly beyond the break-even level requires considerably better plasma confinement and somewhat higher plasma temperature, and thus considerable improvements in plasma confinement, plasma purity, power exhaust and plasma stability.
This approach differs considerably from recent fusion device proposals made for component testing utilising tokamak plasma configuration based on relatively immature plasma confinement schemes (see for example [30,31]).
It should also be noted that proposals of building dedicated testing facilities to accelerate the deployment of reliable Commercial Advanced Nuclear Energy Technologies are also being made in fission [32].
The recommended approach displayed in table 3 indicates the progression of TRL as a function of incremental nuclear testing requirements/needs schematically in three main categories.It goes without saying that in all the phases discussed below, non-nuclear technology facilities for the breeding blanket R&D will continue to play a pivotal role and further investments should be directed to continue improving the testing range of these facilities.

Irradiation tests for concept definition and feasibility
These are tests in fission test reactors that measure the degradation of material properties in relevant material specimens.These tests encompass testing and analyses performed to screen and consolidate materials combinations, design choices and the material composition for the most promising breeding blanket concepts selected as reference to identify potential behaviour challenges and life-limiting phenomena.A lot of these tests were conducted in the past and more are needed for correcting identified design faults or characterising new types of material composition for more attractive and cheaper blanket solutions.As an example, the design modifications that were recently implemented in some of the reference concepts, e.g.replacement of beryllium with beryllide in the HCPB, the introduction of double tubes in the WCLL or new concepts that could lead to design simplification and cost savings (e.g.WLCB) lead to new irradiation testing needs.

Irradiation tests for design evaluation and improvements
This category encompasses irradiation tests aiming at establishing acceptable blanket performance in a representative/relevant fusion environment to reach TRL 5-6 in table 3.For example, these tests would consist of irradiating and examining a set of lead blanket sub-assemblies at relevant fluence levels (<5 dpa).These types of tests are meant to address: • Initial exploration of coupled phenomena in a fusion environment • Exploration of unexpected synergistic effects, and calibrate non-fusion tests • Assessment of the impact of rapid property changes in early life • Collection of integrated environmental data for model improvement and simulation benchmarking • Correction of design faults • Identification of early life failure mode and rates Currently, TBM tests in ITER are the only one contemplated for this category.TBMs have their own integrated loops and systems for tritium breeding, tritium processing, and heat extraction and are meant to provide important information in the following areas: • 'Validation of tritium breeding predictions through measurement of tritium production rates in the blanket and tritium concentrations in purge and coolant streams; • Validation of neutronics predictive capabilities (neutron flux, secondary gamma-ray flux, spectra for neutrons and gamma rays, nuclear heating rate, ratio of neutron to gamma heating); These tests should allow a better determination of the minimum required tritium breeding ratio and provide information that will help better define practical blanket design parameters (e.g.first wall (FW) thickness, structure content, and coolant conditions) that impact tritium breeding; • Acquisition of data on the dynamic behaviour of tritium recovery processes, tritium control, and tritium flow rates during periods of start-up and short pulses when timedependence is strong and over campaigns of many pulses where tritium oscillatory equilibrium concentrations are reached; • Validation of thermomechanical response and adequacy of the structure (including the thickness of first wall) of strongly heterogeneous, ferromagnetic, thin-wall breeding blankets in a fusion environment including vacuum, surface heat flux, spatially-dependent volumetric heating, and 3-component magnetic field with gradients, both under normal and off-normal operating conditions; • Quantitative characterization of many other phenomena including structural response, mass transfer, etc. of strongly heterogeneous breeding blanket concepts in response to fusion loading conditions including coupled heat transfer, thermo-mechanical, and fluid mechanical/magnetohydrodynamic processes that govern the heat transport [23]'.
It remains questionable whether the maximum neutron fluence that the ITER TBM is foreseen to experience would be sufficient to uncover unexpected synergistic effects coupled to radiation interactions in materials, verify performance beyond the beginning of life and until changes in properties become important, provide initial data on failure modes and rates, and ultimately establish engineering feasibility of breeding blankets.Early identification of such behaviour will allow mitigation by correcting design faults or could even lead to abandonment of a blanket option.
The prompt realisation of a VNS test facility would also minimise the risks and uncertainties of the ITER TBM programme which is expected to deliver relevant results only at the end of the next decade.An important additional point is that, presently, with ITER one can only explore a very limited number of design options of blankets (i.e.materials combination and configurations) that are foreseen to be tested in ITER, some of which may suffer from shortcomings in terms of reliability, commercial attractiveness and cost.
It is important to recognise that independent of the results of the ITER TBM tests, which are uncertain, the RoX and technical information already available from the EU TBM R&D Programme (see table 4) is very important and useful to inform the design, licensing and testing strategy of a VNS [33].

Irradiation tests for breeding blanket component qualification and demonstration
Based on the results of the preceding 'screening' testing phases, this irradiation phase must address the engineering development and reliability of breeding blankets to a level of fluence (>20 dpa) sufficient to design, fabricate and operate reliable breeding blankets in DEMO.This is the primary and indisputable motivation that justifies the realization of a VNS.The technical risks and feasibility of licensing using DEMO itself to carry out this qualification were discussed above.
These tests should lead to: • Identify lifetime limiting failure modes and effects based on full environment coupled interactions • Failure rate data: develop a database sufficient to predict mean-time between failure with confidence • Iterative design/test/fail/analyse/improvement of safetyoriented programmes • Obtain data to predict mean-time-to-replace/repair for both planned outages and random failures • Develop a database to predict the overall availability of breeding blanket components in DEMO • Achieve high confidence in obtaining tritium selfsufficiency in DEMO and qualifying all the necessary auxiliary technologies, including the fuel cycle.
The need to establish an adequate reliability database for the breeding blanket that is one of the most important core fusion nuclear components as it generates tritium and power is certainly going to be required from a regulatory standpoint.This reliability database will also provide the basis for establishing a robust preventive maintenance plan.The knowledge of the failure rates of the main components of the breeding blanket, aiming to demonstrate the maintainability, is a key part of licensing process.Moreover such database will justify the initiating events leading to potential accidental situations as well as the safety features to prevent them.

Additional testing considerations
The definition of the best strategy to be adopted for testing the breeding blanket and other DEMO relevant fusion technologies in a VNS is a complex subject that needs to be further discussed with experts.There is a large amount of old literature available on this subject (see for example [2-5, 17, 34]).A sound testing strategy must consider where possible (i) the available return of experience from the ITER TBM programme and anticipate possible risks arising from a reduction of the scope or delay of this programme, and (ii) existing additional uncertainties, linked to specific breeding blanket designs, whose resolution will require new knowledge through experiments, models, and theory in order to demonstrate the feasibility and increase the readiness of the breeding blanket fuel cycle system.Needed tests would range in complexity, including exploration of individual and interactive phenomena, and fully integrated tests.By addressing the complete array of testing issues, this work helps to define needed engineering research which should prove useful in future fusion program planning.A VNS device would allow testing of specific breeding blankets and in-vessel design elements, together with components at reasonable model size, under realistic conditions.Tests must be properly organised and should include scoping tests, performance verification tests and reliability growth tests.While testing at high n-fluence remain an important goal, one should not forget that identifying possible failure mechanisms that might occur at relatively low-fluence and resolving them it is a paramount goal of a VNS and would provide an invaluable contribution to a systematic development of designs of these components towards lower failure rates and would provide the required database.
The feasibility of VNS to provide extensive nuclear testing campaigns at high machine availability require particular attention to the design and reliability of some of the systems (like the vacuum vessel, the superconducting magnets, the (low energy) neutral beam injectors and the nuclear buildings).Easy access and remote maintainability of the internal components is also an important feature of the design.Some initial considerations in these directions are provided in Annex 7.5.
A VNS device must be designed with robust and wellestablished technologies for the permanent components like the vacuum vessel, magnets and the neutral beam injectors.All the innovation of a VNS device lye in the sacrificial components to be tested.This, together with sufficient nuclear shielding and design margins will enable reaching high cumulative fluence to incrementally test different types of blankets.If 10-20 dpa are currently considered by experts as a n-fluence to confirm the performance and reliability of a DEMO blanket concept (TRL 7-8) and achieving this value could be sufficient to inform the DEMO blanket design, testing in a VNS could continue to higher fluence, while building DEMO, to gather valuable database at higher fluence.Alternatively, this facility could be also used to develop and test and qualify in a second phase more advanced concept of interest for a commercial fusion power plant.

Long term R&D at a VNS
Besides the development and validation work described above, mandatory for the success of DEMO, the VNS facility proposed here would contribute to long term R&D on tritium breeding and blanket technology.In particular, it would provide a strong incentive for additional supporting R&D facilities to be realised and operated.
One additional crucial issue in the present debate of alternative configurations for a commercial fusion reactor is whether the basic device can be made compact.Magnet R&D shows remarkable progress in this direction, demonstrating significant field capability on the scale of large model coils.Other components, including the breeding blanket, lack such demonstration.Recognizing the integrated nature of a fusion reactor, it is clear that progress towards smaller, higher performance, possibly simpler devices can only be achieved by suitable testing and demonstration on the large scale offered by a VNS device.

Conclusions and recommendations
Despite its criticality to the development of fusion power, the maturity of the breeding blanket is still very low, and no breeding blanket has ever been built or tested.Large feasibility concerns and performance uncertainties exist for all concepts.R&D is therefore needed to fill the remaining outstanding gaps and a selection now would be premature.
Data from R&D helps screen blanket concept evaluation towards establishing technologies of tritium self-sufficiency and energy extraction.However, one must recognise that the existing non-fusion testing facilities cannot replace the need for a comprehensive testing programme in fusion facilities.Today, the nuclear qualification of the DEMO breeding blanket relies on input provided by the ITER TBM and on the qualification of this complex system directly in DEMO.
Any delay or reduction of the scope of the ITER TBM programme (i.e.reduction of the testing neutron fluence) will, in the absence of a risk mitigation strategy, impact the scope and timeline of the DEMO breeding blanket programme.Similarly, any considerable change of the DEMO breeding blanket designs vis-à-vis those planned for testing in ITER (either materials combinations or design configurations) will result in additional uncertainties and risks.
DONES remains an important complementary device in the EU Roadmap, which is focused on large dpa in very small structural materials specimens.This is deemed mandatory for materials qualification and the development of licensing codes.However, it will not provide any information on the blanket reliability nor the effectiveness of tritium breeding and extraction.
The risks of constructing and operating a DEMO device requiring tritium breeding self-sufficiency from day 1, without extensive prior fusion testing to prove the ability to attain the required tritium self-sufficiency, reliability and a reasonable device availability are judged to be unacceptably high.It is also highly questionable whether, from a regulatory standpoint, a safety authority will issue a license to operate a DEMO plant implementing a large full-scale nuclear system like the breeding blanket which has not been tested and qualified in relevant conditions in advance.
One of the few available options to overcome this crucial qualification bottleneck is to perform testing and qualification of breeding blanket technologies and design concepts in a nuclear plasma device that serves as a 14 MeV VNS.This device will minimise the risks and uncertainties coming from the ITER TBM and reduce DEMO technological risk by qualifying essential technologies in advance of DEMO.In this case DEMO would no longer be a 'qualification' device, but a real demonstrator (first-of-a-kind fusion power plant).
A suitable testing facility could be a small Deuterium-Tritium driven (Q ∼ 1) device (2-3 m) that produces 30-50 MW of fusion power and does not rely on a tritium producing blanket from day one.The plasma serves as a neutron producer without any needs of high performance, except for long burn and high machine availability.A feasibility assessment of such a device is underway in EUROfusion, benefitting from work done in the past.Based on the outcome of this work, an ambitius plan for the realisation of this device to be constructed as early as possible and run in parallel to both ITER operation and the DEMO design process, should be prepared and implemented.This should include also the definition of additional auxiliary technology small-and medum-scale facilities.

Functions of the breeding blanket in DEMO
The breeding blanket is certainly one of the most important and novel components of DEMO and of any other device to follow ITER.
'The breeding blanket must perform several essential functions: 1. First, it must absorb the largest (∼80%) part of the fusion energy transported by neutrons from the plasma and deposited volumetrically in the surrounding in-vessel structures.The remaining part (∼20%) of the fusion power (fusion alpha particles) with the addition of the auxiliary heating power (∼100 MW) constitutes the so-called 'power exhaust', and is deposited as surface heat onto the PFCs, i.e. first wall (integrated into the front-side of the blanket), limiters and divertors.Taking into account the exothermal heat produced by nuclear reactions (about 1.2-1.3energy multiplication factor depending on the neutron multiplier materials adopted in the breeding blanket), in a reactor of about 2 GW of fusion power, the blanket system has to extract about 1900 MW of nuclear power.Conversion of this energy at adequate thermodynamic efficiencies requires that the coolants are at high temperature and pressure.This has a strong influence on reactor engineering.2. Second, it must breed enough tritium by capturing fusion neutrons in lithium-bearing materials (in solid or liquid form).For example, a 2 GW fusion power DEMO is expected to consume around 112 kg of tritium per full power year (fpy), and this clearly underscores the indispensable requirement for the breeding blanket to produce and enable extraction of the bred tritium to achieve tritium self-sufficiency (i.e. it must produce its own fuel).The implications of the tritium breeding requirements on the design and integration of the tokamak in-vessel components that compete for space usage that is needed for breeding (i.e.divertor, protection limiters, auxiliary heating systems, etc.) are briefly discussed below (see also [35]).In addition, the breeding blanket must be designed to enable efficient extraction of tritium and minimise losses of tritium.Further information on the tritium fuel cycle can be found elsewhere.3. Furthermore, together with the vacuum vessel, the blanket must effectively contribute to shielding various components from nuclear radiation (e.g.superconducting magnets and other equipment outside the reactor) [27]'.
In addition, the blanket must provide: 4. a plasma-facing surface which is designed for an acceptably low influx of impurities to the plasma.5. limiting surfaces that define the plasma boundary during startup and shutdown.6. passage for and accommodate requirements of interfacing systems (including heating systems, fueling systems and plasma diagnostics).
Admittedly, there might be design choices which lessen the importance of these other functions, but they should not be discarded and a credible design story is needed to explain how they are accomplished or not needed.
Figure 2 shows a schematic vertical cross-section of DEMO and the physical interfaces between the breeding blanket/vacuum vessel and the other systems like the superconductings coils: toroidal field (TF) and central solenoid (CS).The tritium breeding performance competes with the shielding performance in space restricted regions such as the mid-section of the inboard region.The utilization of the space on the inner side of the torus represents a crucial design aspect in tokamak design.

Breeding blanket concept types
The key parameters for the breeding blanket are: 'A multitude of elements contribute to the evaluation of attractive breeding blanket concepts.The performance and attractiveness of a breeding blanket concept depend on several parameters, including: (a) Power production for given plant size: power production is proportional to the fusion power, the neutron energy multiplier and the cycle efficiency.The choice of blanket material directly affects the neutron energy multiplier.It also affects the cycle efficiency since material temperature limits directly influence the maximum allowable coolant temperature and, in turn, the power cycle efficiency.(b) Safety: this is a key area which particularly influences public perception and acceptance.Blanket materials with low short-term activation are attractive, in particular, if the corresponding blanket and shield systems provide for passive accommodation of off-normal scenarios such as loss of coolant or flow accudents with limited consequences at short and long terms.Long-term activation of blanket materials influences the acceptance criteria of the final waste disposal (c) Availability: commercial reactors would require high availability and thus minimum planned and unplanned downtime for replacement.Key blanket parameters influencing this are the reliability, lifetime and replacement time of the blanket system.(d) Design and Fabrication: simplicity in the blanket design and fabrication process tends to result in lower capital costs and a more reliable system that should be always preferred.(e) Tritium: tritium issues relate to the need to provide selfsufficiency from blanket tritium breeding and to provide acceptable safety parameters including the control of the total inventory in the blanket system and to limit the possibility of permeation and contamination in ancillary equipment and in the tritium processing system (influenced by permeation of tritium from cooling and purge lines).Blanket material and coolant choices directly influence these issues.(f) Economics: the cost of electricity represents the bottom line for commercial fusion reactors.It is influenced by most of the other parameters discussed above and would be the ultimate economic measure [36]'.
It should be noted that the performance and attractiveness of a blanket are coupled with the design and requirements of other reactor systems.Recent work carried out in Europe clearly highlighted that the design choices for the breeding blanket must consider integration constraints arising from the interfacing systems and the whole DEMO plant.The choice of the breeding blanket coolant provides a clear example of a design issue that pervasively affects the overall design layout of the nuclear plant, and bears a strong impact on design integration, maintenance, and safety.It is also desirable to have the same coolant for both the blanket and divertor, which places additional demands on the design [27].'Breeding blanket systems have been under development since the start of civil fusion investigation in the early 1950's.Many concepts were investigated, ranging from more conservative concepts to higher-risk higher-payoff concepts for future reactors.The major candidate breeding materials consist of liquid breeders, mainly liquid metals, and lithium ceramic breeders.The degree of conservatism in the concept is often linked with the choice of structural material since more advanced concepts generally require operation at high temperatures to provide for high cycle efficiency and power production performance and, thus, a greater degree of extrapolation in structural material properties and technology.The choice of structural material, in turn, influences the choice of breeding material based on the accommodation of key issues such as material compatibility and temperature limits [36]'.
The primary blanket options under consideration as candidates for DEMO in Europe are summarised in table 5 and figure 3. The rationale for selection is presented in [27].

Critical breeding blanket technology issues
A summary of the critical R&D issues to be resolved for the breeding blanket is given in table 6 [1].
It is rather striking to recognise certain analogies between the design of the first-wall/ breeding blanket system in fusion and fuel assembly in fission reactors.Albeit with different lifelimiting phenomena, they both experience very high surface and bulk heat fluxes and are subject to large neutron damage (see figure 4).
Performance and reliability are the main design drivers for the breeding blanket.Due to its function on tritium and power generation, it is generally recognised as impacting the safety functions of tritium control, heat removal, and confinement of radioactive material.Accordingly, the production of tritium thanks to the breeding blanket is generally considered a safety important activity that would require demonstration of the qualification before licensing a future fusion reactor.
One of the main knowledge gaps is the identification of failure modes and quantification of failure rates.It is generally agreed that the reliability growth phase will be one of the most time-consuming stages in Fusion Nuclear Technology development'.Prior studies, e.g [17], have shown that the availability of the blanket system must be higher than 88% to meet a DEMO target availability goal of 50%.Since the time to replace blankets is long (weeks), the Mean-Time-Between Failures must be long enough to achieve a high availability target goal.Current assessments show very poor reliability for the breeding blanket concepts considered to date.
Knowledge of failure modes and rates is necessary for the breeding blanket because of their critical impact on plant availability and safety.There is virtually no data on breeding blanket failure modes and rates under fusion reaction conditions (i.e.under energetic neutron and gamma irradiation, high-temperature gradients, subjected to high magnetic forces, under vacuum, etc).Prudent selection of feasible and attractive designs is extremely difficult without such data.
Work to date has identified some possible failure modes; for example, those listed in table 7.

Scarcity of available tritium resources
The criticality of the issue of tritium availability for operating fusion power plants after ITER has been discussed recently in various papers.Based on the results of a study conducted in 2017 [37] and the forecasts of tritium production in Heavy Water Reactors (HWRs) of Canadian Deuterium Uranium (CANDU) type-reactors in countries where tritium extraction is carried out, or planned to be carried out, worst-case scenarios were identified where it would appear that there is insufficient tritium to satisfy the fusion demand after ITER.
The main conclusions of this study can be summarised as follows [38]: The tritium consumption in fusion systems is huge: 55.8 kg per 1000 MW fusion per full power year.This means that for a device like DEMO (e.g.2000 MW Fusion Power Plant (∼500 MWe)) one would consume: 112 kg yr −1 ; 0.31 kg day −1 ; 0.013 kg hr −1 .
This rate of consumption clearly underscores the indispensable requirement for the breeding blanket to produce and enable extraction of the bred tritium to achieve tritium selfsufficiency (i.e. it must produce its own fuel).
The tritium production in fission reactors is much smaller (and cost is very high).
Estimated cost $30 M kg −1 (current) It should be noted that fission reactor operators do not really want to make tritium because of permeation, safety and licensing concerns.They want to minimise tritium production if possible.
'Clearly there is a need to better understand and monitor the future availability of tritium and understand the impact of limited resources on the timeline of DEMO.However, there is essentially very little that the fusion community can do to exert an effect on the supply side, as tritium is a by-product of the operation of these reactors and not the primary economic incentive.Defence stockpiles of tritium are unlikely ever to be shared, and commercial CANDU operators will not alter their plans just to sell more tritium for the start-up of the first fusion power plants.In the short-term it is recommended to monitor the production of tritium in HWRs and estimate the available supply commercially [27]'.

Preliminary results of the recent exploration of feasible design windows for a tokamak-based VNS
Early examples of VNS design concepts proposed in the past are described in [17,18,39].Despite the usefulness of this early work, some of the shortcomings included: (i) favourable physics and technology assumptions, (ii) divertor power exhaust problem underestimated, (iii) use of copper coils, that required limited shielding in most of the cases; and (iv) lack of thorough nuclear design integration considerations.
Most of the concepts are to be operated in driven mode (i.e.beam-heated reactors) with low Q = 1-3.The plasma serves as a neutron producer without any high performance needs, except for long burn and high machine availability.They were fairly small devices (e.g.R: 1-3 m), with low power (e.g.<100 MW) and low tritium consumption.This was important to operate relying at least in the very initial phase on external tritium supplies and without requiring from the very beginning of operation a fully integrated and tritium self-sufficient breeding blanket.They also featured easy access to the fusion core for repairing and replacing testing components.They normally considered installation of breeding blanket elemements on part of the available equatorial ports and the rest of the ouboard space (in total several tens of m 2 to be tested).Due  to the limited size available in a tokamak, the inboard space of the blanket was solely utilized for shielding.
A study is underway in Europe to determine the feasibility of this device, defining its technical characteristics, and to identify and resolve, if any, potential technical showstoppers.The preliminary results of this work are described here.This is not intended to represent a final design choice but rather a 'proxy' to be used, at the time of writing this paper, to identify and resolve crucial machine performance and design integration problems.
The initial set of requirements/design assumptions adopted are listed in table 8, while the potential main technical issues under investigation are described in table 9.
It should also be noted that a tokamak-based VNS option is presently explored, but alternative plasma configurations like stellarators and mirrors will also be investigated, if deemed attractive.As for the latter, important work is ongoing elsewhere (see [40]).
One of the main findings observed during the initial exploratory system studies is that to keep the neutron wall  • Fusion power: (<50 MW)-to minimize the tritium supply requirement, the fusion power has to be restricted to a reasonably low level (<2 kg-T per full-power year).• Operation mode: beam-target driven mode with low Q value (⩽1), steady/state.• Peak-neutron wall loading: (>0.5 MW m −2 ); neutron fluence: >20-50 dpa.
• Breeding: available equatorial ports and outboard wall >10 m 2 of exposed first wall.• Ability to test several candidate blanket concepts of DEMO, including the possibility to use different breeder, multipliers materials and coolants if feasible.• Adequate machine availability: (tbd).
• Configuration, remote maintenance and other design features must emphasize easy access and rapid replacement of device components and test articles and in/vessel components.• Shielding requirements: protect vessel, protect SC coils, protect man-accessible areas.Space at the inboard is very constraints and might requires advance shielding material combinations.• Confinement requirements: no sheltering, no evacuation of the members of public in case of accident.
• Site power consumption must be kept as low as possible: <200 MW (tbd).
• Well established technologies for the magnets, vacuum vessel and neutral beams, nuclear buildings to minimise R&D and speed up deployment times.• Capital cost should be kept as low as possible.
loading sufficiently high, a VNS requires a large neutral beam injection (tens of MW) in a relatively small volume.Even when assuming an energy confinement time lower than predicted by 'standard' scaling laws (e.g.IPB98(y,2)), the energy content of such plasma is expected to be quite high.As such, all MHD stability limits linked to high plasma pressure, like for example the so-called beta limit, require particular attention.
In its simplest formulation, the beta limit can be expressed as an upper limit on the quantity β N , namely: (where p is the plasma pressure, µ 0 is the vacuum magnetic permeability, a is the plasma minor radius and I p is the plasma current, while the value of the constant is typically around ∼3.5%).In order to keep this quantity low, the main driver is the magnetic field B, which shall be maximised, as apparent from the formula above.Alternatively, one can also maximise the plasma current, however this has some implications: 1. Generally speaking, a large current requires anyway a large magnetic field, in order to maintain the safety factor q at acceptable levels, thus ensuring the stability of the plasma.2. If a steady-state plasma is required, the installed NB power must be able to provide a sufficient current drive to avoid the use of a CS.This may not be possible if the plasma current gets too large, or equivalently a larger NB power should be installed, negatively impacting both power exhaust and costs.3. A large plasma currents typically increases the plasma confinement time as well.For a given NB power, this leads   to an increase in plasma pressure, and subsequently in β N .
According to IPB98(y,2) scaling, the effect of B is on the contrary quite limited.
Some geometrical factors, like low aspect ratio and high elongation, allow the increase of the plasma current without requiring a too high magnetic field.
A cross-section of a tokamak VNS considered at the time of writing this paper is shown in figure 5 together with some preliminary design parameters.
The input from previous work from-and technical discussions with Professor M Abdou (UCLA) are acknowledged.Also, the help of J. Elbez-Uzan (DCT), F. Hernandez (KIT), A. Spagnuolo (DCT), S. d'Amico (DCT), R. Kamendje (EUROfusion) and I. Ricapito (F4E), in revising this paper is kindly acknowledged.Special thanks go to the EUROfusion DEMO Central and experts in the EU Laboratories engaged in a VNS feasibility assessment for providing some initial results of this study.

Figure 1 .
Figure 1.Proposed application of technical readiness levels to breeding blanket development and qualification.Assuming TRL definition norms taken from [28].

Figure 2 .
Figure 2. Schematic showing the radial build at the inboard.

Figure 4 .
Figure 4. Analogies between technical issues of the first-wall/ breeding blanket system in fusion and fuel assembly in fission reactors.
ENGINEERING • n-shielding inboard TF coils and NBI ports: difficult -Mitigation: advanced shield compositions (e.c., WC, W 2 B 5 .etc) • Magnets/Equilibrium: -Mitigation: stabilizing plates, inner coils, PF coil nearer to vessel/ plasma • Neutral beam: Pumping/ regeneration and availability/maintainability challenging -Mitigation: improve regeneration of NBI pumps, NBI pumping advanced concepts • Divertor • Integration/ maintenance in-vessel components: -Mitigation: develop robust solutions for mechanical and hydraulic connections • Consolidate testing strategy, improve (simplify) breeding blanket design -Mitigation: testing strategy in port out/ of ports

Figure 5 .
Figure 5. Cross-section of a tokamak-based VNS with some preliminary design parameters (see disclaimer in the text).

Table 2 .
[1]tribution of non-fusion facilities to resolving critical issues for fusion nuclear technology.Reprinted from[1], Copyright (1995), with permission from Elsevier.
a Partial; substantial contribution when supplemented by fusion test; not meaningful in the absence of fusion tests, as no judgment can be rendered on the resolution of the critical issue.

Table 3 .
[28]osed new strategy for the development and nuclear qualification of the DEMO breeding blanket.Assuming TRL definition norms taken from[28].

Table 5 .
Some design parameters for EU breeding blanket concepts.

Performance of Blanket Components under Normal and Off-Normal Operation: •
Liquid metal MHD effects: relationship of fluid flow, heat transfer and structural response in the presence of magnetic fields, bulk heating, surface heating, and full geometric complexity • Interaction of primary and secondary stresses and deformation • Effect of swelling and creep on stress concentrations • Consequences of plasma disruptions • Sources and consequences of hot Spots

Tritium inventory in the solid breeder under actual operating conditions
• Radiation effects on tritium diffusivity, solubility and trapping • Variability in temperature due to radiation effects and mechanical interactions (gap conductance, cracking, swelling, creep, etc.) 6 Tritium

response and lifetime of the first-wall 9 Remote maintenance with acceptable machine shut-down time 10 Accuracy and Survivability of Instrumentation and Control:
• Accuracy and calibration in the Fusion Environment • Lifetime limits due to radiation effects

Table 7 .
Anticipated Breeding blanket failure modes (also modes for fuel assemblies are also shown for comparison.).Flow induced vibration of the water cooled pipes in the breeding zone.•Corrosion attack of the liquid metal and at the interface between the structural material and the ceramic breeder.Grid-to-rod fretting because of fuel assembly vibration • Primary defect formation at residual moisture in the fuel rod • Degradation of cladding properties at high burnups

Table 8 .
Initial set of requirements/ design assumptions for a tokamak-based VNS.

Table 9 .
Potential technical showstoppers of a VNS.PHYSICS • Equilibrium and vertical stability (VS): challenging (tiny plasma, distant coils) Mitigation: lower aspect ratio A, plasma, stabilizing plates for VS, equilibrium in-vessel coils, optimising elongation (high values are better for equilibrium, but worse for VS) • Beta limit -Mitigation: lower β N -e.g.through high magnetic field, higher current.Target: Normalised beta: β N < 3.5% • Fast particle confinement (NB and alphas) Mitigation: higher current, lower A • Divertor/ power exhaust -Mitigation: optimise divertor geometry and impurity mix for dissipation (keeping Z eff low in the core to ensure a sufficently high slowing-down time), maximise neutral pressure