Justine Pallu

Meltem Avci-Adali

Peter Mackeben

Leila Mohammadnejad

Giorgio Mattana

Vincent Noël

Benoît Piro
email: benoîtpiro.piro@univ-paris-diderot.fr

A DNA Hydrogel Gated Organic Field Effect Transistor

Keywords: Electrolyte-Gated Organic Field Effect Transistor, DNA Hydrogel, Aptamer Highlights

A surface-bound DNA hydrogel was prepared using an original combination of nucleotides isothermal amplification methods.

• The DNA hydrogel can gate an electrolyte-gate organic field effect transistor at ambient conditions.

• The transistor threshold voltage depends on the hydration ratio of the hydrogel.

• Transistor output characteristics can be used to transduce specific biorecognition events between an aptamer sequence constituting the gel and its target, here VEGF.

Introduction

Constant efforts have been made in the development of electronic-based point-of-care (POC) biosensors because they may be robust, possibly produced at low cost (e.g. printable), miniaturized and easy to interface with classical control electronics [1][START_REF] Sokolov | Fabrication of low-cost electronic biosensors[END_REF][START_REF] Torsi | Organic field-effect transistor sensors: a tutorial review[END_REF][START_REF] Magliulo | Printable and flexible electronics: from TFTs to bioelectronic devices[END_REF]. At the same time, organic electronics has proved to be particularly convenient for transducing and amplifying a recognition signal, such as in Organic Field-Effect Transistors (OFETs), which have been used in various formats (Ion-Sensitive OFETs, Enzyme-OFETs...), leading to extremely sensitive devices. Despite these benefits, FETs are still underused in the framework of POC. The main reason is the lack of versatile transduction schemes applicable to a large variety of targets. Indeed, in a conventional FET structure, the field effect is applied between two electrodes through a solid dielectric material, the dielectric constant of which driving the operating voltage of the transistor. Because the field effect is maximal within the dielectric material, the most sensitive part of the device sits within this layer. However, it is obviously pointless to functionalize the dielectric material in conventional FETs because no molecule can penetrate, diffuse and ultimately bind the specific probes therein. This is one of the reasons why the so-called Electrolyte-Gated Organic Field-Effect Transistor (EGOFET) is so attractive. In such devices, the capacitive coupling between the gate electrode and the organic semiconductor is made by an electrolyte instead of a conventional solid dielectric. Gate-source polarization causes the formation of an electric double layer (EDL) at the gate/electrolyte and organic semiconductor (OSC)/electrolyte interfaces. Consequently, mirror charges accumulate within the OSC and form a thin layer of carriers.

The seminal works on EGOFETs were those from Bäcklund et al. [START_REF] Bäcklund | Operating principle of polymer insulator organic thin-film transistors exposed to moisture[END_REF], Panzer and Frisbie [START_REF] Panzer | High carrier density and metallic conductivity in poly(3-hexylthiophene) achieved by electrostatic charge injection[END_REF] or Crispin, Berggren and coworkers [START_REF] Liu | Vertical polyelectrolyte-gated organic field-effect transistors[END_REF][START_REF] Said | Polymer field-effect transistor gated via a poly(styrenesulfonic acid) thin film[END_REF], who first reported these devices using solid electrolytes (polymer electrolytes or polyelectrolytes). Later, Kergoat et al. [START_REF] Kergoat | A water-gate organic field-effect transistor[END_REF] demonstrated that solid electrolytes could be simply replaced by water. For biosensing applications, it has also been shown that EGOFETs are sensitive not only to changes at the OSC/electrolyte interface [START_REF] Piro | Versatile transduction scheme based on electrolyte-gated organic field-effect transistor used as immunoassay readout system[END_REF][START_REF] Palazzo | Detection beyond Debye's length with an electrolyte-gated organic fieldeffect transistor[END_REF] but also to those occurring at the electrolyte/gate interface [START_REF] Berto | EGOFET Peptide Aptasensor for Label-Free Detection of Inflammatory Cytokines in Complex Fluids[END_REF][START_REF] Mulla | Capacitancemodulated transistor detects odorant binding protein chiral interactions[END_REF][START_REF] Casalini | Multiscale sensing of antibody-antigen interactions by organic transistors and single-molecule force spectroscopy[END_REF]. However, these devices cannot transduce molecular events occurring outside the EDLs, the thickness of which depends on the ionic concentration (λD ≈ 0.304/√ , with λD in nm and I the ionic strength, in mol L -1 ). In buffered media mimicking physiological conditions, λD is < 1 nm. This is an obvious drawback because most biomolecules, including oligonucleotides and DNA aptamers, have dimensions which exceed 1 nm. An approach to solve this problem is to use materials able to switch their overall 3D structure upon target binding. In that sense, hydrogels have attracted much attention due to their potentialities in a large range of applications such as biosensors, actuators and microelectromechanical systems [START_REF]Hydrogel sensors and actuators: engineering and technology[END_REF][START_REF] Kottapalli | Nanofibril scaffold assisted MEMS artificial hydrogel neuromasts for enhanced sensitivity flow sensing[END_REF]. For example, it has been shown that stimuli-responsive hydrogels (SRH) are able to change their 3D structure in response to external stimuli like pH, temperature, or ionic strength changes [START_REF] Xiong | Responsive DNA-based hydrogels and their applications[END_REF]. Capture molecules of interest can be inserted into these SRH by non-covalent or covalent interactions to obtain bio-functional hydrogels in which the ligand binding or the followed reaction (e.g., proton release by enzymes) could be responsible for specific changes such as swelling of the hydrogel and changes in its dielectric constant [START_REF] Xiong | Responsive DNA-based hydrogels and their applications[END_REF][START_REF] Matsumoto | Chemical-to-Electrical-Signal Transduction Synchronized with Smart Gel Volume Phase Transition[END_REF]. Besides, seminal studies have already shown that DNA hydrogels (hydrogels crosslinked by complementary DNA sequences) were able to perform morphological structure changes after target binding. By using an aptamer as crosslinking component, the target binding leads to a large gel volume increase that can be transduced mainly by optical [START_REF] Hu | A Shape Memory Acrylamide/DNA Hydrogel Exhibiting Switchable Dual pH-Responsiveness[END_REF][START_REF] Bilgic | Oligonucleotide Immobilization and Hybridization on Aldehyde-Functionalized Poly(2-hydroxyethyl methacrylate) Brushes[END_REF], but also by electrical methods [START_REF] Kahn | Integration of Switchable DNA-Based Hydrogels with Surfaces by the Hybridization Chain Reaction[END_REF].

In this work, we developed a pure DNA hydrogel grown onto a gold wire. We demonstrated that our transistors can be efficiently gated by this DNA hydrogel when put in-between the gate and the OSC, whether dipped into a liquid electrolyte or alone. As a first proof-of-concept, a vascular endothelial growth factor (VEGF)-specific DNA aptamer was incorporated into the DNA hydrogel and the dependence of the drain current of the transistor upon binding with the target VEGF was compared to the current response upon addition of VEGF on a DNA hydrogel that did not contain the specific aptamer.

Experimental

Materials and equipments

The fabrication procedures for the lithographied transistors and the gate microelectrodes are described in the Supplementary file (SI). Poly(N-alkyldiketopyrrolopyrrole dithienylthieno[3,2-b]thiophene) (DPP-DTT) was purchased from Ossila (England), with Mw = 280 ± 10 kDa and PDI = 3.8 ± 0.1.

Phosphate buffer saline (PBS), dichlorobenzene 98%, chlorobenzene -anhydrous, 99.8%, isopropanol and all other reagents and solvents were purchased from Sigma Aldrich and used without further purification. Aqueous solutions were made with MilliQ water.

Generation of DNA hydrogel on gold wires

All DNA oligonucleotide sequences and procedures for rolling circle amplification (RCA) are given in Supplementary file. To attach the DNA hydrogel to gold wires, the surface was cleaned for 30 min with 100% ethanol. Subsequently, the gold surface was incubated in 5 mM Tris, 1 M NaClO4, pH 7.4 with 230 µM thiol-modified primer 3 for 2 h at room temperature (RT). The RCA mixture and 25 µM of primer 2 with or without VEGF aptamer conjugation and thiol-modified primer 3 were added to the gold wires in a reaction tube. The multi-primed chain amplification (MCA) was performed for 20 h at 37°C.

Electrical characterizations

Gate electrodes were glass-sealed Au wires (∅ = 1 mm). For experiments performed in a liquid electrolyte, 15 µL of PBS were put directly on the wafer over the semiconducting channel, without covering the electrical contacts, and the DNA hydrogel-modified gate was dipped into the PBS droplet. For experiments without liquid electrolyte, the DNA hydrogel supported on the Au gate was directly put in contact with the OSC. The electrical characteristics were recorded using a Keithley 4200 Semiconductor Characterization System. Output curves were acquired by sweeping the drain-source voltage (VDS) between 0 and -0.5 V while the gate-source voltage (VGS) was kept at a constant value (-0.5, -0.6, -0.7, -0.8 and -0.9 V for each VDS sweep). Transfer characteristics were acquired by keeping VDS at a constant value (-0.4 V) and sweeping VGS between + 0.2 and -0.9 V.

Results and discussion

Preparation of the hydrogel onto the Au gate electrode

Immobilization of the DNA hydrogel on the Au gate was done by introducing thiolated primers able to chemisorb on the Au gate surface. After incubation with thiolated primer 3, the gate was transferred into the RCA reaction mixture and the DNA gelation process was initiated by addition of primer 2 with or without VEGF aptamer and thiolated primer 3. This process led to the generation of a DNA hydrogel with a thickness of 1 mm on the Au wire, over a length of 2 mm (see Fig. SI.1).

DNA hydrogel without functionalization

DNA hydrogel without liquid electrolyte

First experiments were performed with a DNA hydrogel without specific binding properties, i.e. without the aptamer sequence added to its structure (MCA was performed with primer 2 only). The DNA hydrogel-modified gate was first swollen in PBS for a few minutes then dried with a flush of argon and placed directly onto the semiconductor channel, without surrounding electrolyte. Transfer and output curves of the DNA hydrogel-gated transistor are shown on Fig. 1. The output curves clearly show a field effect for VGS < -0.6 V; the ION/IOFF is 5×10 2 (off state at VGS = -0.4 V, on state at VGS = -0.9 V). From the plot of , we estimated the threshold voltage Vth = -0.7 V, shifted of ca. -400 mV compared to the reference system (bare Au gate, no hydrogel, PBS as electrolyte; see Fig.

SI.2).

In this configuration, however, the hydrogel is stable for no more than a few minutes, then dries and shrinks, which hampers analytical applications.

DNA hydrogel in liquid electrolyte

In this part of the work, experiments were performed with a drop of PBS deposited on the transistor, into which the hydrogel was placed, as it is depicted in the inset of Fig. 1b. In this configuration, the output curves clearly show a field effect for less negative VGS than for the solid-state EGOFET described above, and the output curves present less hysteresis. The ION/IOFF is also higher, reaching 10 3

(off state at = -0.3 V, on state at VGS = -0.9 V). From the plot of , we can estimate the threshold voltage Vth = -0.5 V, less negative of 200 mV compared to the configuration without PBS.

The results presented in sections 3.2.1 and 3.2.2 demonstrate that a DNA hydrogel can be used successfully to gate an EGOFET. The threshold voltage shift between transistor with and without liquid electrolyte, as well as the decrease in drain current is attributed to the fact that the gate capacitance is decreased in the absence of liquid. In PBS, the hydrogel is in its most swollen state, with massive penetration of water and ions inside its structure. This causes a higher dielectric permittivity (associated to a higher capacitance) compared with the case where the hydrogel, out of a liquid, is less swollen and therefore of lower capacitance.

The following section describes the utilization of such DNA-gated EGOFETs for the recognition of a specific target.

[Insert Figure 1 here; single column] for the hydrogel alone and (•) for the hydrogel dipped in PBS (inset: same y-scale for both curves).

Gate electrode diameter = 1 mm; hydrogel thickness = 1 mm; hydrogel length = 2 mm.

Aptamer-containing DNA hydrogel in liquid electrolyte

The DNA hydrogel composition was modified by adding an aptamer sequence able to provide a recognition site for a specific target. This strategy has been reviewed in [START_REF] Guan | Development of hydrogels for regenerative engineering[END_REF]. Therefore, the MCA was performed with primer 2 with covalently conjugated VEGF aptamer at its 5´-end. The presence of the VEGF aptamer and its accessibility was demonstrated using a FAM-labeled antidote (a

complementary
The long response time (several tens of minutes) is attributed to the diffusion time of VEGF into the 1 mm hydrogel. Current change upon VEGF binding could be due to change in the total hydrogel capacitance as well as a shift of transistor threshold voltage. A new approach of growing ultra-thin DNA hydrogels directly from the gate electrode is being currently developed and will permit deciphering the transduction mechanism. The corresponding results will be presented in a forthcoming paper.

Conclusion

The fully original result that we demonstrated in this short communication is that it is possible to use a DNA hydrogel to gate an EGOFET, with or without a surrounding liquid electrolyte. More easily than for conventional polyelectrolytes, a molecular recognition moiety can be added within this DNA hydrogel, in the form of an aptamer sequence, which paves the way for a large variety of targets due to the infinite possibilities offered by aptamers. We are designing an experiment to be able to measure the gate capacitance as a function of time, with or without aptamer, with or without VEGF, in order to decipher which factor is responsible for the transduction; we plan to describe the results in a forthcoming full-length publication.

Despite the advantages brought by this approach, the slow target diffusion within the hydrogel hampers practical applications. We are now developing approaches to grow ultra-thin SRH layers, able to quickly switch their morphology in the presence of traces of a given target, as we reported recently using surface-initiated atom transfer radical polymerization (SI-ATRP) [START_REF] Fillaud | Switchable Hydrogel-Gated Organic Field-Effect Transistors[END_REF], where the initiator is covalently immobilized on the electrode surface and the SHG is grown from it.

Figure 1 .

 1 Figure 1. Output curves ID = f(VDS) acquired at VGS from -0.5 V to -0.9 V for an Au gate modified by a DNA hydrogel (a) in standalone (air) and (b) in liquid electrolyte (PBS). Inset: Schematic view of the corresponding gate/hydrogel arrangement. (c) Transfer curves ID = f(VGS) obtained at VDS = -0.4 V ( ■ )

  sequence to the aptamer) (Fig. SI.3A). DNA hydrogels with immobilized VEGF aptamers demonstrated significantly increased VEGF binding compared to the DNA hydrogels without VEGF aptamer (Fig. SI.3B): after incubation with VEGF, approximately 175 ± 22 pg VEGF werebound to a 50 µL DNA hydrogel (whose weight is about 50 mg). See Sections 1.4 and 1.5 of the Supporting Materials for details. It is expected that the molecular recognition cause a thorough reorganization of the hydrogel, and therefore a change in its overall capacitance, which should generate a change in the electrical characteristics of the transistor. For the aptamer-containing hydrogel, we measured strictly no difference in electrical characteristics compared to the aptamer-free hydrogel (results not shown). Using a gate modified with a VEGF aptamer-containing hydrogel, we recorded the drain current vs time upon addition of VEGF at a concentration of ca. 2.5 pg µL -1 in the electrolyte surrounding the hydrogel. These measurements were also performed with a gate modified with a pure DNA hydrogel, as a negative control. The results are presented in Fig.2. No significant current change was observed upon addition of VEGF to the DNA hydrogel, which did not contain the VEFG aptamer (curve a; 5% change in amplitude, attributed to the normal drift of the device). In contrast, the addition of VEGF to the aptamer-containing hydrogel led to a significant current change (curve b, ca. 30±5% change in amplitude: n = 20). The [Insert Figure 2 here; single column]

Figure 2 .

 2 Figure 2. Variation of the drain current ID = f(t), for addition of 40 pg VEGF in the electrolyte surrounding the hydrogel, for (a) a hydrogel without aptamer, and (b) a VEGF aptamer-containing hydrogel. VGS = -0.50 V and VDS = -0.40 V. Gate diameter = 1 mm. Hydrogel thickness = 1 mm; Hydrogel length = 2 mm. The liquid electrolyte was contained into a PDMS well of 200 µL which strongly reduced its evaporation during the experiment. n = 20 hydrogel-modified gates were studied. 75% presented similar characteristics, as shown, while 25% shortly detached (partially or totally) from the gate.

Acknowledgments

This work was funded by ANR (Agence Nationale de la Recherche) and CGI (Commissariat à l'Investissement d'Avenir), Labex SEAM (Science and Engineering for Advanced Materials and devices), grand numbers ANR 11 LABX 086, ANR 11 IDEX 05 02.

Conflicts of Interest.

The authors declare no conflict of interest. The funders had no role in the design of the study, in the collection, analyses, or interpretation of data, in the writing of the manuscript or in the decision to publish the results.


Francois Nunio
email: francois.nunio@cea.fr

Alexandre Torre

Louis Zani

Mechanical analysis of the European DEMO central solenoid pre-load structure and coils

Keywords: Central solenoid, DEMO, Pre-load structure, Coil, Winding pack

The pre-conceptual design of the European DEMO fusion tokamak is currently being developed under the coordination of the EUROfusion Consortium. This paper reports the mechanical analysis of the Central Solenoid (CS), which comes right after the phase of definition of the winding pack proposed by the CEA. The first step consists in a draft-sizing of the tie-plates pre-compression structure. Then a global Finite element Analysis (FEA) of the CS, considering a sector-symmetry model and smeared winding pack properties is built-up in order to check this pre-load structure, and to identify the most critical loading scenario for the conductor. Finally, an axisymmetric local FEA of the CS winding pack is performed in order to study in detail the stress level in the conductor jacket and in the insulations layers. This pre-conceptual analysis procedure allowed to assess the acceptability of stress for the CEA winding pack proposal, and will be operated during the future magnetic system design iterations.

Introduction

In the EU-DEMO 2015 reference design, the Central Solenoid (CS) coil assembly is a stack of 5 electrically independent module coils, which mainly provide the flux variation during the plasma initiation, and in addition the capability to sustain a wide range of scenarios through plasma shaping. For some scenarios, two adjacent coils may be fed by currents in opposite directions, resulting in significant repulsive forces within the CS system. It is thus necessary to introduce a vertical Pre Load Structure (PLS) in this system in order to keep the modules in contact and maintain their axes aligned along the scenario.

More specifically, the vertical preload is driven by the need of preventing separation of the modules when they are fed with reverse current, avoiding tensile stress between modules. Also, in case of non-axisymmetric disruption of the plasma, any lateral motion or twisting of the CS module must be prevented by the action of the only effect of the friction.

Consequently, a sufficient amount of vertical preload force must be applied during the CS coil assembly, and such that at least 25% of their interface surface is kept in contact with each other [START_REF] Libeyre | An Optimized Central Solenoid for ITER[END_REF].

In this structural pre-conceptual study, the section 3 presents a global Finite Element Model (FEM) of the overall CS structure, which is built in order to check the preload stress and contact status under various plasma scenario. Additionally, in section 4, the stress state at the level of the conductors jacket and insulation of the most loaded module, is analyzed by implementing a local FEM of a set of 3 pancakes.

Inputs and Design approach

The models are built-up on the basis of the 2015 CAD configuration, considering the 2016 CEA optimized Winding Pack (WP) proposal [START_REF] Torre | EU-DEMO TF and CS Magnet Systems Design and Analyses Performed at CEA[END_REF]. The driving dimensions are summarized in table 1.

The layout and coil numbering are given in figure 1. This optimized WP configuration allows to increase the gap between Toroidal Field (TF) and CS coils in order to accommodate pre-compression tie-plates. In this regard, the coil external and internal radius are modified, and consequently the CS and PF currents defined in the magnetic scenarios given by [START_REF] Ambrosino | Reference Flat Top Equilibria for DEMO with AspectRatio 3[END_REF] are increased in order to keep the same maximal magnetic field than that of the original design (see table 2). This choice is conservative since the magnetic flux is increased. In order to compute the Lorentz forces acting on the coil modules, and locally on the winding pack, a global electromagnetic analysis is performed using the Cast3M software [4]. The force density is then integrated over each module to study the balance of loads, the CS system being considered to be supported from its lower plate to the TF coil. The maximal repulsion occurs for the Start Of Flat-top (SOF) case, between module CS1 and CS2L. The maximal compression occurs for the End Of Flat-top (EOF) case (between CS1 and CS2L), and an unbalance of 186 MN needs to be counteracted by the PLS in order to avoid the separation of the lower module from the bottom supporting plate.

The PLS is usually based on tie-rods located at the inner and outer diameters of the coil. Flat Tie-Plates (TP) can be used in order to save space in between the CS and the TF coils. Before going to a detailed global FEM of the CS system, a first draft of the PLS must be established, on the basis of the following criteria:

1. the preload must be larger than the unbalance load between modules 2. the tie-plates elongation under preload must be larger than the addition of : a. the coil axial shrinkage induced by the electromagnetic load b. the differential thermal contraction between coil and tieplates during cool-down 3. the stress in tie-plates must be lower than 2/3 of the yield strength of the material.

The fulfilment of the criteria #2 requires to minimize the TP section in order to maximize the spring effect, whereas criteria #3 requires to maximize the TP section in order to minimize the stress. A trade-off is then necessary between these two criteria, which are antagonistic.

The use of 316L type steel [5] for both conductor jacket and TP does not allow a good trade-off. In this case, the allowable stress is limited to 159MPa at Room Temperature (RT), and the preload decreases significantly during the cooling-down.

The use of a Nitronic 50 type steel (which has a higher yield strength) is favorable to the fulfilment of criterion #3. In this case, the allowable stress is 266MPa at RT [6], but the loss of preload during the coolingdown is further increased, since this type of material has a lower thermal contraction than that of 316L.

If a JK2LB type steel is used in conjunction for the conductor jacket (very low thermal contraction of 2.1 mm/m [START_REF] Sgobba | Physical Properties of a High-Strength Austenitic Stainless Steel for the Precompression Structure of the ITER Central Solenoid[END_REF]), the preload will increase during the coolingdown and make it possible to meet criterion #2 as well.

Under these conditions, a preliminary design of the pre-compression system is established on the basis of analytical calculations, considering 10 identical TP at the internal and external radii. The dimensions of these plates are 60mm by 275mm (see figure 2).

Global FEM of the CS coil assembly

In order to allow a detailed analysis of the precompression system, taking into account precisely the rigidity and thermal contraction of the winding, a global finite element model is built using an Ansys® FEA [8]. This allows also to consider the coil bending effects which are not captured by a mono-dimensional analytical model. A sector of 18° is considered, taking into account the symmetries of geometry (see figure 2). The displacement on the nodes of the lateral symmetry planes are constrained in the azimuthal direction. The material properties are given in table 3. The homogenized material properties of the coil winding are considered. These are computed separately, by implementing a local model of the winding at the scale of the conductor, a considering micro-mechanical analysis on a representative unit cell of the WP. The evaluations are carried out for 316L and JK2LB jackets, with an insulating layer in G11. The wires bundle is not considered. This local FEM leads to orthotropic properties that represent globally the coil behavior (see table 4). Contact with friction (µ=0,2) is considered between modules and inter-module plates made of G11. The tie-plate dimensioning is analyzed considering several material options for tie-plates and conductor jackets. All the results are expanded in table 5. Two criteria are taken into account to assess the acceptance of the design:

• the axial stress at the mid-plane of the tie-plates should remain lower than the material allowable (RT is the most unfavorable situation)

• the gap between coil modules must be closed on at least 25% of the coil's section.

The first case A1 considers a 316L stainless steel for the tie plates and conductor jackets made of 316LN with a pre-load of 225 MN at RT. In this case, the axial stress in the tie-plates (250 MPa) is much larger than the allowable, and some gaps between CS modules remains open under electromagnetic loads.

The second case B1 considers a Nitronic 50 stainless steel for the tie plates and conductor jackets made of 316LN with a pre-load of 246 MN at RT. In this case, the axial stress in the tie-plates (273 MPa) is at the limit of the allowable, but some gaps between CS modules still remains open under EMAG loads. Increasing the tieplate section, allowing to increase the pre-load, does not lead to a situation where the gaps are kept closed.

The situation is better if a JK2LB stainless steel is considered for the jackets (case C1 to C6). For the C1 case, the pre-load (150 MN at RT) is too small to achieve the closing of the gaps. For a pre-load of 225 MN, the axial stress in the TP is close to the allowable. Increasing the pre-load at RT up to 240 MN allows to achieve the closure of the gaps up to 25% of the coil's section, but the axial stress in the TP is slightly exceeded (case C4). For a preload exceeding 240 MN, the axial stress in the tie-plate largely exceeds the allowable (380x2/3=253 MPa @ RT [6]). The results of the global model are expanded for the C4 case, which is considered as the reference situation for this design study (see figure 3). In terms of displacement, the coil is compressed axially by about 5 mm under PL and reaches 49 mm after cool-down. The axial compaction is maximal under Pre-Magnetisation (PM) scenario; it totals up to 74 mm.

Under EOF scenario, the axial compaction is 69 mm. The radial expansion is maximal under PM scenario and reaches 6 mm.

The contact condition is fulfilled for the 3 magnetic situations. The contact is closed in the inter-module regions, and partly opening in the end-plate region, but the contact is closed over an area which is superior to 25% of the coil's section. It has been checked that the coefficient of friction has a limited influence on the results. Among the scenarios available at the time of this study [START_REF] Ambrosino | Reference Flat Top Equilibria for DEMO with AspectRatio 3[END_REF], the PM situation is the most demanding in terms of "smeared" stress intensity (385 MPa). This stress state requires to be analysed in detail by a local model of the winding of the coil, which is reported in the next section. [mm] 7.44 7.17 5,22 (≈25%) (≈25%) The electromagnetic loads are transferred from the electromagnetic analysis (see section 2) as body force density, and redistributed on cable (equivalent pressure is considered on jacket outermost inner edge, as illustrated in figure 4). The nodal displacements of the upper edge of the model are coupled in the vertical direction in order to avoid local effects. The premagnetization situation is the most critical. The peak stress reaches 1010 MPa, on the second turn of the pancake (counted from inner radius).

(>25%) (>25%) (>25%) Gap between modules [PL+CD+SOF] [mm] 0 0 0 0 0 0 0 0 Gap between modules [PL+CD+EOF] [mm] 1.65 1.51 (≈25%) 0 0 0 0 0

Local FEM of the most loaded pancake

In order to assess the strength of the conductor jacket, the stress must be analyzed across a path which cross the structure in the most loaded region. Here, 6 paths are considered (see figure 5). The Tresca stress is linearized along these path and decomposed in the membrane and membrane+bending components, which can be compared to material allowable. This stress linearization process leads to a maximal primary membrane stress of 653 MPa under the PM situation (paths 2 and 5). The acceptability of this stress state requires a material which yield strength is superior to 980 MPa @ 4K. This condition is at the limit of the allowance of 316LN, but is satisfied by JK2LB, which has a yield strength of 1128 MPa @ 4K, and this is all the more so since the fatigue conditions have not been analyzed. This situation fits perfectly with the need for such a material for the PLS, and corresponds to the ITER design.

The shear and compressive components of the stress in the turn insulation at the interface with the jacket has been analyzed. The ITER insulation shear criteria [START_REF]ITER Magnet Structural Design Criteria -Part 2: Magnet Winding (ITER_D_2ES43V)[END_REF] is fulfilled on the lateral sides of the jackets, and in the inter-pancakes insulation layers. But tension stress arise in a direction perpendicular to the interface in the corner radius region, at a level of 58 MPa. This situation tends to violate the ITER criteria, which forbids tension stress in this direction, unless a test or analysis can prove that cracks or de-bonding relieve the stresses [START_REF]ITER Magnet Structural Design Criteria -Part 2: Magnet Winding (ITER_D_2ES43V)[END_REF]. It has been checked that lowering the outer corner radius of the jacket (ignoring the associated manufacturing issues) can reduce the stress in the jacket, but the tension stress is amplified in the insulation layers.

Conclusion

A pre-dimensioning analysis allowed to define a layout for the pre-compression tie-plates of the preload system of the CS, which served as a basis for the global FE analysis. This analysis allowed to verify that a design based on N50 tie-plates, JK2LB jackets in the conductors and a 240 MN room temperature pre-load allows to maintain the compaction of the CS during the cooling-down and the energization of the magnet. Under nominal pre-load, the maximal axial stress in the tieplate is 267 MPa @ RT. The maximal radial expansion is lower than 6 mm in the case of the pre-magnetization scenario.

At the level of the winding pack of the coil, the local stress linearization revealed a primary membrane stress of 653 MPa, requesting a minimal yield stress of 980 MPa at 4K for the conductor jacket. This situation can be satisfied if JK2LB is used, thus completing the need to use this material for the proper operation of the axial preload system. Tensile stress across insulation layers exist in the region of the corner radius of the conductor jacket, and is difficult to avoid. Further test or analysis should be operated in order to prove that cracks or debonding relieve the stresses. This procedure will be repeated to design and evaluate the mechanical structure of the CS, as part of the future configuration iterations of EU-DEMO.

Fig. 2 .

 2 Fig. 2. Global FEM of the CS coil assembly.

  Five load-steps are analyzed: the preload [PL], the situation after cooling down at 4K [PL+CD], and three operating conditions under electromagnetic load [PL+CD+PRE], [PL+CD+SOF], [PL+CD+EOF]. The displacement is locked after first load-step. Lorentz forces are transferred from the electromagnetic analysis (see section 2) as body loads. Gravity is taken into account, but has very low impact on stress levels.

Fig. 3 .

 3 Fig. 3. Contact status / coil radial expansion (C4) [mm].

Fig. 4 .

 4 Fig. 4. Local FEM of the CS1 module (central pancakes)

Fig. 5 .

 5 Fig. 5. Tresca stress in the conductor jacket.

Table 1 .

 1 CS Winding Pack parameters.

	Parameter		unit	value
	Conductor Size (insulated)	mm	61.0
	Turn Insulation		mm	1.0
	Cable size		mm	33.4
	Jacket thickness		mm	12.8
	Inter-pancakes	insulation
	thickness		mm	1.0
	Ground insulation thickness	mm	8.0
	Number of turns		-	13
	Number of pancakes in CS3U/CS2U -	44
	Number of pancakes in CS1	-	90
	Number of pancakes in CS2L/CS3L -	44
	Total number of pancakes	-	266
	External radius of CS winding	mm	3200
	Internal radius of CS winding	mm	2391
	Height of CS3U/CS2U winding	mm	2789
	Height of CS1 winding		mm	5687
	Height of CS2L/CS3L winding	mm	2789
	Central channel external diameter	mm	10
	Central channel spiral thickness	mm	1

Fig. 1. Magnets layout (length unit=meter).

Table 2 .

 2 CS and PF reference scenario currents set for three defined moments.

			current [MAt]
	Coil	Premag	SOF	EOF
	TF	14.28	14.28	14.28
	CS3U	30.73	9.85	-0.25
	CS2U	30.73	6.13	-30.71
	CS1	62.78	-9.66	-62.78
	CS2L	30.73	7.67	-30.46
	CS3L	28.54	12.3	-21.99
	P1	12.38	12.28	-7.80
	P2	4.63	-7.44	-8.00
	P3	-3.41	1.01	-0.30
	P4	4.34	-8.82	-9.23
	P5	-3.20	-7.35	-7.75
	P6	19.20	18.31	7.70
	PLASMA	0.00	19.60	19.60

Table 3 .

 3 Material properties considered at 4K

	316L (isotropic)
	elastic modulus	E	205. GPa
	Poisson's ratio	NU	0.29
	int. thermal contraction	ALP	-3.0 mm/m
	JK2LB (isotropic)
	elastic modulus	E	205. GPa
	Poisson's ratio	NU	0.29
	int. thermal contraction	ALP	-2.1 mm/m
	Nitronic 50 (isotropic)
	elastic modulus	E	195. GPa
	Poisson's ratio	NU	0.29
	int. thermal contraction	ALP	-2.8 mm/m
	G11 (transverse isotropic in the plane of wrapping 1-2)
		E1	20. GPa
	elastic modulus	E2	20. GPa
		E3	12. Gpa
	Poisson's ratio	NU12 NU13	0.17 0.33

Table 5 .

 5 Summary of the global mechanical analysis (values in red / orange correspond to unfavorable / unacceptable configuration).

			A1	B1	C1	C2	C3	C4	C5	C6
	Jacket Material		316LN	316LN	JK2LB	JK2LB	JK2LB	JK2LB	JK2LB	JK2LB
	Tie-Plate Material		316L	N50	N50	N50	N50	N50	N50	N50
	Tie-Plate allowable stress [Sm@RT]	[MPa]	159	253	253	253	253	253	253	253
	Tie-Plate allowable stress [Sm@4K]	[MPa]	407	776	776	776	776	776	776	776
	Tie-plate pre-Load @ RT	[MN]	225	246	150	208	225	240	260	275
	Tie-plate pre-Load @ 4K	[MN]	186	180	195	253	270	285	305	320
	Axial stress in TP [PL]	[MPa]	250	273	167	231	250	267	289	306
	Axial stress in TP [PL+CD]	[MPa]	207	200	217	282	300	317	339	356
	Axial stress in TP [PL+CD+PRE]	[MPa]	0	0	0	20	39	55	77	95
	Axial stress in TP [PL+CD+SOF]	[MPa]	202	198	214	278	298	314	337	349
	Axial stress in TP [PL+CD+EOF]	[MPa]	0	0	0	0	82	99	121	138
	Gap between modules [PL+CD+PRE]

Acknowledgments

"This work has been carried out within the framework of the EUROfusion Consortium. The views and opinions expressed herein do not necessarily reflect those of the European Commission."


Didier Perrault
email: didier.perrault@irsn.fr

Nuclear safety aspects on the road towards fusion energy

Keywords: safety, fusion facility, tritium

On the road toward fusion energy, ITER is the first fusion facility which will have enough radioactive inventory to be potentially dangerous for the public and the environment. As such, ITER has a licensed nuclear facility status and its operator, ITER Organization, has to define technical, organizational and human provisions such as to prevent or adequately limit the risks of accidents as well as the disadvantages (exposure to ionizing radiation, environmental releases and waste) that the facility presents.

All the facilities which will be created after ITER, in order to develop fusion energy (DEMO...) then to use it (PROTO…), will clearly also be facilities for which the nuclear safety aspects will have to be taken into account. This paper describes the safety characteristics of fusion facilities, the state of progress of the ITER safety demonstration regarding these characteristics and their possible evolutions for the facilities succeeding ITER on the road to fusion energy.

Introduction

Due to the quantities of radioactive substances used, greater than never before, the ITER facility is the first fusion facility using magnetic confinement to require a construction license. Its construction license was issued by the French government in 2012, following a detailed examination of the provisions adopted by the operator to prevent or adequately mitigate the risks and disadvantages it presents. Assessment of these provisions was performed by the French "Institut de Radioprotection et de Sûreté Nucléaire" (IRSN) at the request of the French regulator "Autorité de Sûreté Nucléaire" (ASN).

All the facilities which will be created after ITER in order to develop fusion energy (DEMO...), then use it to produce electricity (PROTO…), will clearly also be facilities for which nuclear safety aspects will have to be taken into account.

This paper aims to present the main safety characteristics of fusion facilities, the state of progress of the ITER safety demonstration regarding these characteristics and their possible evolutions for the facilities succeeding ITER on the road to fusion energy.

Safety characteristics of fusion facilities 2.1 General context

The general safety concept of fusion facilities is based on the nuclear safety concepts of facilities with radioactive inventories, in particular of current power reactors using fission. However there are specific differences in their implementation due to the physical and technological characteristics of fusion installations.

The quantity of radioactive materials of a nuclear power plant (NPP) fission reactor is quite significant. This quantity for an equivalent fusion reactor is much smaller. Two types of radioactive substances are found in a nuclear fusion facility, the tritium and the activated materials. Their radiotoxicity is altogether lower by several orders of magnitude than in fission reactors.

Therefore, the theoretically possible maximum releases (around 1 kg of tritium for ITER) from a fusion facility would result in doses several orders of magnitude lower than those for hypothetical worst-case scenarios of a NPP fission reactor. However, the doses to the public would be in the order of one Sievert, making its evacuation necessary within a large perimeter around the installation. And thus, to prevent such releases, a fusion facility needs, just like a fission reactor, to implement a safety concept.

While fission facilities, based on operational experience, use proven technologies and qualified materials as well as validated calculation methods for their safety demonstration, there has only been minor operational experience available to design ITER, since it is the firstof-a-kind nuclear fusion facility. Also, the specific technologies required for fusion are out of the scope of fission engineering (very large vacuum vessels, high magnetic fields enfolding large volumes).

While the largest part of the inventory in a NPP fission facility is stored inside the fuel rods, the inventories of radioactive materials within a fusion facility are not concentrated locally but found in several large buildings (the tokamak building, the tritium building, the hot cell building and the radwaste building).

Tritium is mostly found in its gaseous state (T2), diffusing easily through seals and certain materials, but also in tritiated water (HTO), which is often corrosive for the equipment involved. So, tritium, which is fortunately not very radiotoxic, can be easily mobilized in case of an accident and be released into the environment.

As for the activated materials, they are mainly in a solid form and thus difficult to mobilize in case of an accident. The rest of them, being the active dust produced and accumulated in the vacuum vessel and the activated corrosion products accumulated in the cooling loops, can also be mobilized.

State of progress for ITER

Under French regulation, ITER is classified, same as nuclear power plants, in class 1 (out of 3) of the nuclear facilities classification for their importance of risks and disadvantages. The authorized operating domain is limited by a quantity of tritium in the whole facility less than or equal to 4 kg, a quantity of tritium in the vacuum vessel less than or equal to 1 kg and a mass of dust in the vacuum vessel less than or equal to 1,000 kg.

Thanks to the safety provisions implemented in ITER, the release in case of accident would result in doses to the public no less than about 5 millisieverts, making protective measures unnecessary (no sheltering, no evacuation).

Considerations for the future fusion facilities

The future fusion facilities should be self-sufficient in tritium. They should have a significantly longer operating time and a higher fusion power for electric power generation compared to ITER. This could increase the inventory of radioactive materials and thus involve more significant nuclear safety and radiation protection issues than the ITER facility. Nevertheless, the consequences of accidents should not lead in all cases to protective measures for the public, because that is what is more and more expected from the public and requested as well by the safety authorities in the future. The future fusion facilities should benefit from the safety demonstration of ITER and the lessons learned from its operation.

Hazards

Most of the hazards to be considered that may affect a fusion facility are well known. Among these classic hazards, it is necessary to underline the particular importance of explosion, because of the presence of hydrogen isotopes and dusts. In addition, some hazards specific to nuclear fusion facilities are to be considered, such as plasma physics events and magnetic system failures.

State of progress for ITER

The regulator and the IRSN have been involved from the very beginning of the ITER project and the licensing was done without delay on the project.The ITER construction license application was accompanied by a file containing a "preliminary safety report" describing the facility and outlining the provisions to manage hazards. The preliminary designs for the "tokamak building" equipment were consistent at this stage of the creation of a facility but they were incomplete for equipment in the "tritium building", the "hot cell building" and the "radwaste building". The regulator has requested a new, more comprehensive version of preliminary designs for the equipment in these three nuclear buildings. The design studies for the transfer cask system (which is intended to carry out highly radioactive equipment between the tokamak building and the hot cell building) are also requested.

While some equipment, and their associated safety requirements, were classic and well-known, others were unusual. For these, the safety requirements adopted by the operator were considered acceptable but the regulator requested the associated demonstration.

Considerations for the future fusion facilities

The future fusion facilities would have new equipment or new safety requirements due to evolutions of the concept (cooling with helium for example). The involvement of the regulator from the very beginning is advisable.

Confinement of radioactive materials

The confinement of radioactive materials in fusion facilities is, as for other nuclear facilities, relying on a set of physical barriers and dynamic systems.

Depending on the building, there are, mainly 2 static barriers, and sometimes 3 barriers. The static barriers are the usual ones (processes, cells, glove boxes, building) except in the tokamak building. Indeed, the first static barrier is the vacuum vessel and its extensions, and it is a very complex zone due to a large collection of systems (cooling, heating, diagnostic, fueling…) that penetrate the vacuum boundary.

Moreover, this first barrier is surrounded by equipment carrying high energy and they could challenge it. The main energies are the plasma energy, the magnetic energy, the thermal energy from the cooling systems, the energy released in the event of an explosion of hydrogen isotopes or a dust explosion, the cooling energy from the system that supplies liquid helium and the energy from activated materials.

So, this first barrier can be challenged by a full range of off-normal events or combination of off-normal events. Thus, a fully comprehensive set of load combinations have to be taken into account in the design of this first barrier. In addition, the code and standards developed for fission facilities are not adapted to certain materials or loads.

For the events which would lead to lose this first barrier, consequences into the environment are limited, thanks to the second barrier, which consists, in the tokamak building, of the rooms surrounding the tokamak, and the associated detritiation system. To design the different rooms of this building, here again, a full range of offnormal events or combination of off-normal events have to be defined for their design.

The concept of the dynamic system is mainly based on the use of scrubber columns to reduce the release of tritium to the environment. At this scale, the concept of this detritiation system is very new.

State of progress for ITER

The design of ITER containment systems relies on rooms with a rather substantial leaktightness. To be sure that the expected level of leaktightness has been reached, the regulator has requested first, to measure, at commissioning, the leak rate in these premises and then, to ensure that these leaktightness levels are maintained throughout operation.

The ITER detritiation systems clean the air with a 99 % efficiency, and 90 % in case of fire. The regulator has requested to demonstrate these expected efficiencies and to ensure that no accident situations may cause ambient conditions or chemical emissions that could significantly reduce these efficiencies.

The basemat of the tokamak complex is part of the second confinement barrier. The tokamak was initially supported vertically by 18 metallic columns embedded in this basement. This design has been considered by the regulator as not optimal, considering the flexion strains on these metallic columns induced by the cryostat thermal deformations and the needed margins. In 2014, a new design with a concrete crown has been approved.

Considerations for the future fusion facilities

For these facilities, the first barrier of the tokamak will be a little less complex. Indeed, to produce more tritium than the reactor consumes, the surface area covered by the tritium breeding blankets should be as large as possible. As a result, the vacuum vessel should have fewer penetrations.

Ionising radiation exposure

As no entrance is allowed within the tokamak building during plasma operation, the main potential sources of occupational radiation exposure are during maintenance activities. Exposure is minimized by classic shielding and access control but also by choosing reduced activation materials and using remote handling, such as the automated transfer casks.

State of progress for ITER

Operating the ITER facility is characterised by relatively short experimental campaigns separated by long maintenance periods. During design, the optimisation work had adopted individual internal dose targets close to zero, a mean individual dose of 2.5 mSv/year, with a maximum individual dose of 10 mSv over a year and an annual collective dose of 500 person.mSv. The regulator has considered that detailed projected assessments of operational doses will be necessary to demonstrate that these objectives would be reached.

As the automated transfer casks cannot be shielded, there must be no access to the zones concerned by the transfer of vacuum vessel internal components. The regulator has considered that the robust nature of the access control design for these no-go zones would have to be demonstrated.

Considerations for the future fusion facilities

Due to plasma operating times much longer than for ITER and to higher fusion power, the ionising radiation exposure risks for operators of future fusion facilities would be greater. Moreover, activation of tritium breeding blankets could be, depending on the type of blanket, an additional important source of radiations.

On the other hand, maintenance periods should be very short compared to ITER. Future facilities would also use, like ITER, a high level of robotized maintenance and reduced activation materials.

Residual heat removal

Same as in a fission nuclear facility, a residual heat generation in a fusion nuclear facility is not stopped completely after shutdown but will continue and decrease exponentially for the time after. This heat generation is due to the radioactive decay of the activated materials. Cooling systems evacuate this residual heat.

State of progress for ITER

For the ITER facility, there is relatively little residual heat to remove (11 MW at plasma shutdown and 0.6 MW after one day). The internal structures of the cryostat are at very low temperatures. In the event of failure of all the cooling systems, the temperature rise in the tokamak structures would be slow, especially since air in this situation would enter into the cryostat and slow down this rise in temperature. It would take about four months for the temperature of the divertor to, due to the residual heat, become high enough to cause failure of its cooling system and consequently cause water ingress into the vacuum vessel. There could then be a large proportion of dust sufficiently hot to be oxidized by water and produce a significant amount of hydrogen in the vacuum vessel. Then, a hydrogen or dust explosion in the vacuum vessel would release tritium and dust into the environment. However, due this 4-month length of time, the residual power dissipation is not considered as a safety function for ITER.

Considerations for the future fusion facilities

For future fusion facilities, the residual heat would be much greater than for the ITER facility, due to the longer operating times with plasma and a greater fusion power, even if the use of reduced activation materials will reduce this residual heat.

The safety consequences of a total loss of tokamak cooling systems need to be evaluated, in particular regarding the tritium breeding blankets. If the evaluation shows that the impact on safety is significant, the decay heat removal becomes a safety function. It is then necessary to design cooling systems for which the probability of loss is inversely proportional to the seriousness of the safety consequences of such loss.

As for the blanket sectors, due to their expected size in the future fusion facilities, their residual heat could be significant. Again, the safety consequences of a loss of their cooling systems need to be evaluated in order to determine if their residual power evacuation is or is not a safety function during their transfer from the vacuum vessel to the hot cells or when stored in these cells.

Disturbances of plasma

The plasma is subject to various types of instability. Small-scale instabilities lead to turbulences similar to that observed in a fluid. Large-scale instabilities affect the plasma as a whole resulting sometimes in plasma completely losing its magnetic confinement. This sudden plasma termination, called "disruption", can also happen after a loss of plasma control, following an event external to plasma (sudden increase of fueling, of heating…).

When a disruption occurs it causes physical phenomena (thermal shocks, electron beams, eddy currents, etc.) which have definite consequences (electromagnetic loads in the vacuum vessel internal components and in the vessel itself, production of dust from the first wall of the blanket) or lead to a risk of water leaks due to electron beams. The design of the vacuum vessel, which is the first confinement barrier, must take into account the electromagnetic loads that the most severe disruptions can produced. The definition of the most severe disruptions is thus of a great importance. It is done on the basis of theoretical studies and on feedback from existing and past tokamaks and taking into account margins.

State of progress for ITER

The regulator has considered that the most severe plasma malfunctions had to be completed by taking into account all the predictable cases such as loss of plasma control, plasma radial instabilities and efficiency of the fusion power termination system. The relevant studies have been recently received by the regulator.

Considerations for the future fusion facilities

To obtain a good control of plasma stability without any particular limitation in duration, the plasma control system in the future fusion facilities will be much more developed than for the ITER facility. Therefore, its possible types of malfunction will be more numerous and different from those considered for the ITER facility. Furthermore, cases of malfunction of the disruption mitigation systems, in particular the consequences of a spurious trip of these systems, should be taken into account.

Explosion

Dusts are produced during normal operation as a result of plasma-wall interaction and they accumulate in the vacuum vessel. A part of the tritium and deuterium introduced into the vacuum vessel to produce fusion reactions is adsorbed onto the first wall of the vacuum vessel internal components. In the event of water ingress into the vacuum vessel, hydrogen is produced by oxidation of dust by water.

Hydrogen isotopes can be easily ignited by a small ignition source. As for dusts, they can explode if there is a rather high ignition energy. The energy delivered as a result of the hydrogen isotopes combustion might be enough to ignite the dust. Thus, in case of water or air ingress into the vacuum vessel, an explosion of hydrogen or dusts could occur. Even if the quantities of dusts and tritium are limited, the vacuum vessel's design pressure being relatively small, these would affect the confinement of the vacuum vessel or its extensions.

Outside the vacuum vessel, the hydrogen isotopes and dusts are distributed in the rooms in small quantities, not enough to endanger the confinement in case of explosion. However, several rooms could contain significant quantities of hydrogen isotopes or dusts and, for these, explosion hazards have to be considered.

State of progress for ITER

Regarding the explosion hazard due to the presence of hydrogen isotopes and dust into the vacuum vessel, the operator initially considered that:

-in case of air ingress, the combustion of part of the hydrogen isotope inventory would, in the worst situation, lead to localized pressures of 200 kPa, thus not endangering the vacuum vessel integrity ;

-in case of water ingress, hydrogen isotopes and dusts would be transferred into the tank of the vacuum vessel pressure suppression system (VVPSS) and their explosion would not be significant enough to endanger the integrity of this tank.

This was not considered as acceptable by the regulator mainly because, for the air ingress, the mobilization and explosion of dust have not been taken into account and, for the water ingress, the quantity of hydrogen produced by oxidation of the dust has been under estimated.

New studies has been delivered by the operator and evaluated with the modification of the VVPSS. It is now considered that the analysis of the hazard of explosion into the vacuum vessel has been improved. However, due the complexity of the phenomena in question (desorption of tritium, possible localized concentration of hydrogen isotopes, mobilization of dusts), it is already considered insufficient to completely eliminate the risk of an explosion affecting the confinement of the vacuum vessel or the VVPSS.

Considerations for the future fusion facilities

On the basis of current knowledge, the limitations adopted by the ITER operator on the quantities of tritium (1 kg) and dust (1,000 kg) in the vacuum vessel could be reached after one or two years of operation, i.e. after 100-200 hours of operation with plasma. After that, thermal desorption operations would have to be performed to reduce the quantities of adsorbed tritium and cleaning operations to reduce the quantities of dust as well.

For future fusion facilities, where operating times with plasma are much longer, limitations identical to those adopted for the ITER facility would require frequent shutdowns, incompatible with the desired availability rates.

If the designers adopt the same limitations, they will need to put forward provisions to reduce adsorption of tritium on the structures and dust creation, or to reduce the time it takes to detritiate the vacuum vessel and remove dust.

The designers may also examine and adopt other choices regarding these limitations (on the quantities of tritium and dust in the vacuum vessel) and show that, with these new limitations, the consequences of accidents, in particular air ingress into the vacuum vessel followed by an explosion of hydrogen isotopes and dust, would be acceptable, given the design provisions adopted elsewhere.

Magnetic system failures

Due to the huge magnetic fields, the coils are superconducting. In the event of loss of superconductivity (for example in case of a leak of helium from the coils cooling system), the conductor status changes from superconducting to normal conducting. To avoid overheating, the stored magnetic energy is discharged from the coil by a fast discharge circuit. The current discharge involves electromagnetic loads in the vacuum vessel and its internal components.

If an electric arc happens (for example after an uncontrolled loss of superconductivity), a high current can propagate within and around the coil and cause damage.

In case of a short circuit, the induced forces in the coils are greater than forces in normal operation and local deformations may occur.

State of progress for ITER

The design of the fast discharge circuit provides a redundancy and relies on switches of different technology. It has been considered as robust.

The regulator requested that the estimated level of damage and of local deformation be confirmed in case of electric arc or short-circuit. The relevant studies have been received by the regulator.

Considerations for the future fusion facilities

For future fusion facilities, the magnetic energy of the toroidal field coils could be significantly higher than in those of the ITER facility. The design is still to be studied in order to avoid loss of integrity of the vacuum vessel in case of a magnetic system failure.

Aircraft crash

On this matter, international practice is that the safety of nuclear facilities, in particular regarding confinement of their radioactive materials, must be assured in case of a plausible aircraft crash. However, the application of this general principle is different from one country to another.

State of progress for ITER

For ITER facility, the probability of general aviation aircraft crash is higher than 10 -7 /year. In this case, the French regulation requests to design the nuclear buildings to withstand a CESSNA 210 or LEARJET 25 C crash. The ITER tokamak is protected by a cover slab.

Considerations for the future fusion facilities

For some future fusion facilities, designs may plan vertical penetrations to extract blanket sectors or halfsectors. So, for these, a cover slab to protect the tokamak could be difficult to set up. Moreover, the regulation may evolve, the general orientation nowadays being to adapt the provisions according to the consequences rather than to the probability.

Accidents

The identification of potential accidents in a nuclear facility is essential to determine the loads to be taken into account in the design of equipment and buildings, and to evaluate the consequences of these accidents.

It is a particularly difficult task for a nuclear fusion facility because a tokamak is a complex and dense zone in equipment stored large energy sources. As a consequence, this specific configuration implies that:

-many different kind of events can happen: plasma disruption, magnetic discharge, water, air or helium ingress into vacuum vessel or cryostat, explosion of hydrogen isotopes or dusts…. but also fire, earthquake etc.

-an event is likeky to directly or indirectly lead to one or several further events.

Moreover, as for other nuclear facilities, accidents:

-have to be divided into several categories according to their estimated occurrence, including the category of the excluded accidents. The accidents which lead to emergency conditions (widespread fire, etc) must be identified ;

-must be studied, taking into account the most penalizing failure of an important safety component required, independently of the initiating event in question ;

-must consider plausible combinations of initiating events and events more severe than those initially determined (extreme earthquake, extreme flooding, etc.

).

The identification of potential accidents that could happen in other buildings than the tokamak building cannot be overlooked because the quantities of radioactive substances released could be rather important.

State of progress for ITER

10 design-basis accidents and 12 beyond-design-basis accidents were studied by the ITER facility operator. Technical discussions during examination of the preliminary safety report led to the operator modifying certain accident scenarios and studying a few more.

The regulator has considered that the list of accidents leading to emergency conditions needs to be completed and that the excluded accidents (uncontrolled fire in a radioactive substances storage, a sudden drop of lift with a nuclear load or a major exposure of workers to ionizing radiation) have to be justified. Most of the studies have been received by the regulator.

Considerations for the future fusion facilities

For the future fusion facilities, the types of accidents to be considered would almost be identical to those adopted for the ITER facility. However, the likelihoods of occurrence and the consequences could be significantly different, in particular due to the greater quantities of radioactive materials or energy involved. Furthermore, given the increased complexity of such facilities, it is possible that other types of accident may need to be studied (drop of a sector or half-sector during its removal from the vacuum vessel…).

Releases under normal operation

As for all nuclear facilities, for a fusion nuclear facility the release limits requested for gaseous and liquid effluents have to be justified. It has to be demonstrated that these limit values have been evaluated on the basis of the best available techniques, under technically and economically acceptable conditions. And the radiological impact of the releases for the public and the environment has to be estimated.

For a nuclear fusion facility, the radiological impact of the releases during normal operation is largely due to discharge of tritium under gaseous form.

State of progress for ITER

The annual gaseous releases from the ITER facility should not exceed 2.5 g of tritium in years of heavy maintenance (long-term shutdown to change the divertor) and 0.6 g of tritium for other years.

During the years of heavy maintenance, tritium release will be mainly due to the degassing of internal components in the vacuum vessel during the long period when it is open and in the hot cells. The release estimates stated above assume that a thermal desorption operation will be performed before opening the vacuum vessel in order to greatly reduce the quantities of tritium present. For the years without heavy maintenance, two-thirds of the tritium release will come from the Tritium Recuperation System (TRS), whose role is to recover most of the tritium present in the activated dust and tritiated intermediate-level long-lived waste from the vacuum vessel internal components.

As for liquid effluents containing tritium, they will mainly come from very slightly tritiated liquid effluents produced by the air detritiation systems and from the treatment of water from the cooling systems. The tritium concentration in these effluents is too low for them to be treated by the Water Detritiation System (WDS), which recover tritium for reinjection into the fuelling system.

The maximum radiological impact due to radioactive liquid and gaseous effluents is assessed to be around 2.3 µSv/year.

Considerations for the future fusion facilities

Long-term shutdowns may not be needed for future fusion reactors, and this would be positive to reduce the gaseous release. However, on the other hand, there will be a greater quantity of tritium and a larger number of release paths and that is not in favour of limiting these releases. The new release paths are the breeding blanket cooling circuits, the waste detritiation equipment and the transfer operations of in-vessel components between the vacuum vessel and the hot cells.

Moreover, future facilities may contain numerous rooms fitted with air detritiation systems as well as large volumes of cooling water. A higher number of air detritiation systems and a greater volume of water detritiation systems are also not in favour of limiting releases of liquid effluents into the environment.

Waste

As for other nuclear facilities, measures must be taken at the design stage in order to limit the volume and the radiological, chemical and biological toxicity of the waste produced and then, by reusing and reprocessing the waste produced, to reduce the size of the repository dedicated to ultimate waste. Fusion facilities produce radioactive waste, particularly activated and/or contaminated by tritium. The specific features of waste in nuclear fusion facilities are that there is no high-level long-live waste and that part of it will contain significant quantities of tritium.

State of progress for ITER

Part of the waste from the ITER facility will contain quantities of tritium that will be too high for it to be directly sent to waste repositories in France. Thus, it is currently planned that they be stored for approximately 50 years, in the planned INTERMED facility, to reduce those quantities of tritium via radioactive decay.

Considerations for the future fusion facilities

Waste produced by the future fusion facilities is likely to be more tritiated than that produced by the ITER facility. This could bring designers to add specific detritiation equipment (thermal furnace, melting furnace, incinerator, etc.). Furthermore, tritium breeding blankets, depending on the type adopted, could be a source of waste different from that in the ITER facility. In any case, waste management for the future fusion reactors will highly depend on the host country's general policy in the matter (recycling, clearance, waste management routes, etc.).

Conclusion

There are specific differences in the implementation of the nuclear safety concepts between fusion facilities and other nuclear facilities.

The safety demonstration for the ITER facility, the first of its kind, still needs additional studies.

All the facilities which will be created after ITER will raise even more significant nuclear safety and radiation protection issues. If not properly addressed, they could be an obstacle (in terms of delay or additional cost) or a stop (no licensing) on the way to fusion energy. Whereas if perfectly mastered, it may turn out to be benefit for fusion power compared to other choices.