Design and manufacturing of WEST Baffle

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Design and manufacturing of WEST Baffle Tristan Batal a,∗ , Mehdi Firdaouss a , Marianne Richou a , Fabien Ferlay a , Sébastien Larroque a , Laurent Gargiulo a , Timothée Dupas b , Jean-Marc Verger a , Louis Doceul a , Franck Samaille a , Jérôme Bucalossi a , Michael Salami c a b c

CEA, IRFM, F-13108 Saint-Paul-Lez-Durance, France SODITECH Ingénierie SA, 1 bis allée des gabians, 06150 Cannes la Bocca, France AVANTIS groupe, 12 route de Saint Mathieu, 06130 Grasse, France

h i g h l i g h t s • • • •

Disruption’s torque in the PFC was simulated thanks to ANSYS. The ANSYS thermal results comply with WEST project requirements. The cycling analysis complies with WEST project requirements. 316L components comply with A level RCC-MRx criteria.

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Article history: Received 19 September 2014 Received in revised form 23 January 2015 Accepted 27 January 2015 Available online xxx Keywords: WEST Baffle Plasma facing components X-point configuration Actively-cooled

a b s t r a c t The Tore Supra tokamak is being transformed in an X-point divertor fusion device in the frame of the WEST project, launched in support to the Iter tungsten divertor strategy. The WEST Baffle aims to evacuate particles by creating a toroidal pumping throat. It also protects the lower divertor cooling pipes and the passive stabilization plate from heat fluxes. The Baffle is made of actively cooled plasma facing components (PFCs), with underneath a passive stabilization plate and a support beam. The Baffle design is presented in a first part. In a second time the different ANSYS simulations are described: electromagnetic, thermal and mechanical results are presented. The design choices are justified for the different results obtained thanks to the simulation. In a final part, some aspects of the PFC manufacturing are presented. © 2015 Elsevier B.V. All rights reserved.

1. Introduction Tore supra is a tokamak using superconducting magnets. It was designed to carry out long plasma operation, so it uses actively cooled plasma facing component. The WEST projects aims to transform Tore Supra from a limiter configuration to an X-point configuration in order to validate the technology of a full tungsten ITER-like actively cooled divertor [1]. The WEST Baffle, made of plasma facing components (PFCs), aims to evacuate impurities and particles form the plasma. It also protects the lower divertor cooling pipes underneath. Its design and structural analysis was done taking into account electromagnetic, thermal and mechanical loads as well as

∗ Corresponding author. Tel.: +33 442253717. E-mail address: [email protected] (T. Batal).

assembling and disassembling constrains. The different simulations showed that the actual design can handle the different loads. 2. WEST Baffle The WEST Baffle is made of several components. The support is a 316L stainless steel circular beam on which a soft copper circular stabilization plate is fixed thanks to 288 M8 screws. The 144 actively cooled plasma facing components are made of CuCrZr while the incidence surface is coated with a 15 ␮m tungsten layer. They have an angular size of 2.5◦ and are located on the top of stabilization plate. The PFCs are fixed on the support beam with 288 M10 screws going through the stabilization plate. The PFCs have partial slit in the poloidal direction to reduce eddy current in case of disruptions. To avoid crack ignition, the end of the split is drilled to reduce the stress.

http://dx.doi.org/10.1016/j.fusengdes.2015.01.053 0920-3796/© 2015 Elsevier B.V. All rights reserved.

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Fig. 2. Torque in the PFC during disruptions.

Fig. 1. Baffle structure.

The WEST tokamak will be used to test different divertor configurations. Because of space constrains, a Baffle sector has to be removed each time when a lower divertor target sector is changed. Therefore, 12 PFCs are fixed on 30◦ sectors by means of inner and outer partial ring elements – both made of 316L stainless steel – 24 M10 screws and 24 M8 screws. Also, hexagon socket head cap screws are used to fix PFC on the support beam, to facilitate assembling operations. Their head is hidden from the plasma in counter bores. All the screws used are made of stainless steel BUMAX 109 (equivalent to class 10.9 bolts) or BUMAX 88 (equivalent to class 8.8 bolts). The cooling pipes for water supply are located at the back of the Baffle, under the outer partial ring. There are 12 hydraulics headers, i.e. one per sector. The material of the pipes is 316L stainless steel (Fig. 1). 3. Electromagnetic analysis During disruption, the magnetic field inside the torus can vary rapidly. This will induce eddy current in the PFC and in the stabilization plate. The stabilization plate has a simple geometry so the forces created by those eddy current can be estimated easily by calculating its electrical resistance. However, the geometry of the PFC is more complex because it has a partial slit in the poloidal direction to reduce the eddy current during disruptions. During a disruption, the vertical component of the variation of the magnetic field can be up to 40 T/s. Because of the toroidal magnetic field a radial torque is induced in the PFC (Fig. 2). The plate was divided into three parts and three different torques were calculated thanks to ANSYS software. The total torque is about 1500 N m for a split covering 75% of the PFC length. There was no split in the first versions of the design, and the radial torque was about three times higher (Figs. 3 and 4).

Fig. 3. ANSYS simulation of eddy current during disruption in the PFC, A/m2 .

4. Thermal analysis Different scenarios had to be studied in order to validate the thermal characteristics of the PFC: • Normal operation (30 s, 60 s and 1000 s plasma) • Vertical displacement event (VDE, 100 ms) • Thermal quench (TQ, 0.5 ms) followed by current quench (CQ, 4 ms) The design criterion for the PFC was a maximum temperature of 450 ◦ C, which is the limit temperature of CuCrZr to avoid creep. In this analysis, the VDE, the thermal and current quench occurs at the very end of normal operation plasma, in order to be conservative. Different local particle heat fluxes were taken into account to determinate the temperature of the PFC during normal operation [2]:

Fig. 4. Radial torques in the PFC during disruption for a split covering 75% of the PFC.

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Fig. 8. Temperature of the WEST Baffle screws in normal operation, ◦ C. Fig. 5. Plasma particles deposition heat flux.

Fig. 6. Cooling channels of the PFC.

• Plasma particles deposition, up to 1.5 MW/m2 (Fig. 5) • Suprathermal electrons up to 0.35 MW/m2 • Ripple losses up to 1.27 MW/m2 The power to extract was calculated for the most conservative local heat loads in toroidal positions in order to pre-design the cooling system of the PFC. It was calculated that each PFC had to be able to extract 32.5 kW of thermal power in steady state. In every scenario, the ripple losses have the most significant impact. According to this the cooling channels were designed to optimize the temperature of the PFC and its fixing system (Fig. 6). The pressure inside the channels is 33 bar, the entrance temperature of the fluid is 70 ◦ C and the fluid velocity is 6 m/s. The heat transfer coefficients, the exit temperature of the fluid and the limit temperature and limit heat flux at the surface of the channels were determined by in-house CEA MATLAB program [3]. Thanks to this data the temperature of the PFC was simulated for every scenario. The limit temperature of the PFC and the surface channel were never exceeded. The temperature of the screws was also studied. As the screws are in counter bores, only the heat load due to ripple losses can impact them, because the trajectory of these particles is quite vertical. The limit temperature of BUMAX 109 screws is 550 ◦ C. The maximum temperature observed was 526 ◦ C in the most conservative case (Figs. 7 and 8).

Fig. 9. Boundary conditions of the thermo-mechanical analysis.

5. Mechanical and thermomechanical analysis A thermo-plastic analysis has been made for the PFC and the stabilization plate. The allowable number of cycles is defined by the WEST scientific program [1]: • 30000 plasmas (normal operation) • 3000 VDE A level RCCM criteria were used for all the other components. It prevents excessive deformation, elastic and elastoplastic instability, time dependant fracture, progressive deformation and fatigue. This requires two analyses, an elastic one and a thermo-elastic one. The following criteria are applicable for the linearized stresses extracted from finite element elastic analyses. • • • •

Primary membrane stress (Pm) < Sm Primary bending stress (Pm + Pb) < 1.5 Sm Local stress (PL + Pb) < 1.5 Sm Secondary stress (Pm + Pb + Q) < 3 Sm

Two scenarios were studied for the mechanical analysis: normal operation and VDE. The most conservative thermal scenarios were used in both cases. For every scenario, elastic, thermo-elastic and thermo-plastic analyses were done, in order to check the number of cycles for the PFC and the stabilization plate and the A level RCCM criteria for the other components. The screws were checked with AFNOR standard (0.85% of yield stress for preload). The different boundary conditions for the analysis in normal operation are (Fig. 9):

Fig. 7. Temperature of the WEST Baffle in normal operation, ◦ C.

• Bolt preload • Radial displacement of 2.25 mm for support beam due to temperature rise

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Fig. 10. Temperature of the WEST Baffle feet, ◦ C. Fig. 12. Stress in the support beam, normal operation, MPa.

Fig. 11. Thermo-mechanical simulation of the WEST Baffle cooling pipes, toroidal displacement, mm.

• Toroidal displacement of 0.4 mm for cooling pipes, due to differential expansion As the support beam is circular, it tends to extend when its temperature rises. The support beam is fixed on a component called “feet”, which also support the WEST lower divertor. Because of the rigidity of these “feet” the support beam cannot extend freely. So a simple model was first done to estimate the radial dilatation of the support beam, which was then used as a boundary condition for the mechanical simulation of the WEST Baffle. A similar model was run for the cooling pipes, because their temperature are lower than the suport beam’s one, which induces differential expansion (Figs. 10 and 11). For the VDE simulation, different combinations of these boundary conditions were added [4]:

Fig. 13. Total deformation in the PFC, normal operation.

threads in the beam tapped holes and the weald bead in the pipes are not modeled in this analysis. The results can be seen in Table 1. The number of cycles was also fulfilled in normal operation and VDE for the PFC and the stabilization plate (Fig. 13). The maximum total deformation was: • 0.165% (30,000 cycles) and 0.33% (3000 cycles) for the stabilization plate • 0.24% (30,000 cycles) and 0.36% (3000 cycles), for the PFC

• Halo forces pushing on front half part on the PFC, due to the plasma, −4.75 kN/PFC (VDE) • PFC torques, previously simulated, 492 N m in the front part and 500 N m for each other part of the PFC (VDE) • Stabilization plate forces, vertical (5.770 kN) and radial (3.125 kN), due to current flowing along the stabilization plate (VDE)

6. Manufacturing of the WEST Baffle

The level A RCC-MRx criterion was fulfilled for the support beam, the cooling pipes, the two partial rings and the cooling pipes (Fig. 12). The stress in those components was compared with 1.5 Sm, as bending effect is widely dominant compared to shear effect (membrane stress is low). For the thermo-mechanical analysis, the criterion used was 3 Sm. Some overstressed areas on the beam’s tapped holes and cooling pipes junction due to boundary conditions were found. However it is not relevant, because the

The PFCs of the west baffle are made of CuCrZr. They are manufactured with larger CuCrZr plate. The channels in the Baffle are deep drilled and sealed with e-beam welded caps (Fig. 14). The cooling pipes of the Baffle are made of 316L stainless steel, so it is impossible to weld it on the PFC, which are made of CuCrZr. Multimaterial transitions of CuCrZr/stainless-steel, made by explosion assembly, are used on each PFC to allow connection of the stainless steel hydraulic headers by welding (Figure 15).

The total deformation of these two components was always below those values, except for some singularities. The adjusted load in the screws was also checked and the preload was never overloaded.

Table 1 Stresses in the support beam, cooling pipes, inner and outer partial ring.

Support beam Cooling pipes Inner partial ring Outer partial ring

1.5 Sm (MPa)

Pm + Pb max (MPa)

3 Sm (MPa)

Pm + Pb + Q max (MPa)

191 191 191 191

221 (not relevant) 201 (not relevant) 190 180

381 381 381 381

333 336 374 176

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Fig. 14. Channel plug at the back of the PFC.

Fig. 15. Test for electron welding of the channels plug.

7. Conclusion The actual WEST Baffle design is robust enough to cope with different electromagnetic, thermal and mechanical loads. This design will facilitate disassembling and assembling operations, and complies with WEST project requirements. Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to fully validate such a technology (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this second major upgrade, Tore-Supra would be able to address the problematic of long plasma discharges with a metallic divertor.

References [1] J. Bucalossi, M. Missirlian, P. Moreau, F. Samaille, E. Tsitrone, D. van Houtte, et al., The WEST Project: Testing ITER Divertor High Heat Flux Component Technology in a Steady State Tokamak Environment, ISFNT, 2013. [2] M. Firdaouss, T. Batal, J. Bucalossi, P. Languille, E. Nardon, M. Richou, Heat Flux Depositions on the WEST Divertor and First Wall Components, 2015 (these proceedings). [3] J. Schlosser, J. Boscary, Finite elements calculations for plasma facing components, in: Proceedings of Specialist Workshop on High Heat Flux Component Cooling, Grenoble, 1993. [4] S. Larroque, C. Portafaix, A. Saille, L. Doceul, J. Bucalossi, F. Samaille, et al., The WEST project mechanical analysis of the divertor structure according to the nuclear construction code, Fusion Eng. Des. (2014), http://dx.doi.org/10.1016/j.fusengdes.2014.02.073.

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