Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X

Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X M.W. Jakubowski 1, A. Ali 1, P. Drewelow 1, H. Niemann 1, F. Pisano 4, A. Puig Sitjes 1, G. Wurden 3, C. Biedermann 1, B. Cannas 4, D. Chauvin 2, R. König 1, V. Moncada 2, T.T. Ngo 2, A. Rodatos 1, J. Fellinger 1, D. Hathiramani 1, T. Sunn Pedersen 1 and W7-X team 1 Max-Planck-Institut für Plasmaphysik, Greifswald, Germany 2 CEA, Cadarache, France 3 Los Alamos National Laboratory, Los Alamos, USA 4 University of Cagliari, Cagliari, Italy

Wendelstein 7-X design Magnetic field 2.5 T Superconducting coils 70 Cold / total mass 425 t / 700 t Magnetic field energy 600 MJ Plasma volume 30 m 3 Plasma duration 30 minutes Heating power 10 MW Maximum heat load 10 MW/m 2 IPP 1

Wendelstein 7-X design Magnetic field 2.5 T Superconducting coils 70 Cold / total mass 425 t / 700 t Magnetic field energy 600 MJ Plasma volume 30 m 3 Plasma duration 30 minutes Heating power 10 MW Maximum heat load 10 MW/m 2 plasma contour triangular plane divertor units bean plane target plates triangular plan 2

heat flux density [MW/m 2 ] separatrix Island divertor creates 3D power deposition patterns 0 standard iota horizontal target baffle high iota baffle 8 6 4 2 0-10 0 10 20 distance from separatrix [cm] [Y. Feng, EMC3-Eirene, priv. comm.] plasma vessel

Towards ITER/DEMO compatible divertor 2015 / 2016 5 MW 4 MJ 6 sec Inertial cooling limiter 4

Towards ITER/DEMO compatible divertor 2015 / 2016 5 MW 4 MJ 6 sec Inertial cooling 2017 / 2018 10 MW 80 MJ 10 sec Inertial cooling graphite divertor, 10 MW/m2 5

Towards ITER/DEMO compatible divertor 2015 / 2016 5 MW 4 MJ 6 sec 2017 / 2018 10 MW 80 MJ 10 sec 2020 10 MW 18 GJ 30 minutes Inertial cooling Inertial cooling Water-cooling of PFCs Water cooled heat exchanger Inertial cooling 10 MW/m 2 Actively cooled steady-state high heat flux divertor - 10 MW/m 2 - Increase of heating power Replace carbon by tungsten walls IPP 6

Typical scenario of OP1.1 hydrogen discharge 2015 / 2016 5 MW 4 MJ 6 sec [kev] 3 MW of ECRH heating is converted into electron temperatures of ca. 9 kev and diamagnetic energy of ca. 300 kj. Power loads to limiter are of order of 1.5 2 MW/m 2 with transient power loads of up to 4 MW/m 2. 60% of power ends on limiters 7

W7-X limiter, photograph W7-X limiter, heat flux density, #20160310 Thermography of limiter during initial campaign Two strike lines follow geometry of the scrape-off layer. They are separated by a watershed area. 8

W7-X limiter, photograph W7-X limiter, heat flux density, #20160310 Thermography of limiter during initial campaign Two strike lines follow geometry of the scrape-off layer. They are separated by a watershed area. Due to shallow impact angles leading edges are present. Challenge for future campaigns of W7-X and ITER. 9

W7-X limiter, IR image w/o plasma W7-X limiter, heat flux density, #20160310 Thermography of limiter during initial campaign Two strike lines follow geometry of the scrape-off layer. They are separated by a watershed area. beginning of campaign end of campaign erosion redeposition Due to shallow impact angles leading edges are present. Challenge for future campaigns of W7-X and ITER. Material migration makes interpretation of thermographic data more difficult: hot spots surface layers 10

43 m q [MW/m 2 ] 36 m Typical for inboard limiters: power fall off length with near and far SOL 16 14 12 L c = 79 [m] data near far total 10 8 6 4 2 0 0 1 2 3 4 5 6 r eff - r LCFS During OP1.1 W7-X showed typical exponential fall off of the power loads of inboard limiter plasmas with modification due to field lines with different connection length. 11

The W7-X Island Divertor Structure (TDU) 2015 / 2016 5 MW 4 MJ 6 sec 2017 / 2018 10 MW 80 MJ 10 sec baffles pumping gap Target (CFC) Inertially cooled: PFC designed to sustain 10MWm 2 Baffles (Graphite) 9

13 Symmetrisation of divertor power loads The asymmetry of power loads can arise due to several factors: Error fields Misalignment of divertor elements during installation Up-down asymmetries caused by particle drifts Requires automatic image matching of different cameras/diagnostics very accurate spatial calibration/correction of optical distortions. The asymmetries in power loads B 11 /B 0 among 2.7 10-4 the limiters have been observed in OP1.1 Specific plasma scenarios allowing to create a database for different parts of the divertor Standard Correction configuration of asymmetries at j = with: 180 n=0, n=1 correction fields re-alignment of divertor units (W7-X standard case, statistical tilting of entire modules of the magnet system by up to 0.1º) Limiter 3 saw consistently less heating (energy) than Limiter 1. [G.Wurden, et al. NF 2017]

Hot spots detection on carbon PFCs Overheating due to excessive power loads, e.g. baffle structures can cope with steady state heat flux density of up to 0.5 MW/m 2, e.g. due to toroidal currents additional power loads reach areas of lower heat resistance. Localized fast particle losses from the neutral beam heating system (NBI) can also produce the hot spots, the duration of the beam injection is up to 10 s. hot spots Hot spots on the surface of JET divertor [G. Arnoux, et al., Rev. Sci. Instr. 83 (2012) 10D727] Due to carbon erosion and re-deposition thin surface layers are formed on the plasma facing components during plasma operation. Typically they have very a poor thermal connection to the bulk material and therefore heat up very fast and reach much higher temperatures than a virgin surface of the PFCs. May lead to false positives! Hot spot due to delamination during tests of W7-X tiles at GLADIS facility. [A. Rodatos, H. Greuner, GLADIS 14

Towards ITER/DEMO compatible divertor 2015 / 2016 5 MW 4 MJ 6 sec 2017 / 2018 10 MW 80 MJ 10 sec 2020 10 MW 18 GJ 30 minutes Inertial cooling Inertial cooling Water-cooling of PFCs Water cooled heat exchanger Inertial cooling 10 MW/m 2 Actively cooled steady-state high heat flux divertor - 10 MW/m 2 - Increase of heating power Replace carbon by tungsten walls IPP 15

Safety of actively cooled divertor is an important issue Max. surface temperature 1200 C, reaction time ~ 0.5 s q = 10 [MW/m 2 ] CFC Divertor fingers CuCrZr Cooling body Max. interface temperature 475 C to avoid cracks. Consequences of developing a delamination inside one of the tiles: increased thermal resistance reduced performance de-bonding of the tile end of operation Crack developed in the bond

A. Ali, et al., PFMC 2017 normalized surface temperature Detecting defects with temperature changes (I) virgin surface layer 1 q [MW/m 2 ] 0.8 0.6 surface temperature algorithm output 0.4 0.2 intact surface layer delaminated 0 0 5 10 15 time [s] surface layers Test of algorithms to detect surface layers surface layers on W7-AS tiles 17

normalized surface temperature Detecting defects with temperature changes (II) virgin delaminated 1 0.8 0.6 0.4 0.2 intact surface layer delaminated τ allows to get the characteristic time scale for the cooling process independent of the absolute temperature Successful automatic detection & classification of delaminations at test facility 0 0 5 10 15 time [s] cycles t [s] delaminated intact 18 [A. Rodatos, et al. RSI 87 (2016) 023506]

Challenge to observe the whole divertor graphite wall stainless steel standard divertor optical resolution @ 4μm ~ 6 mm baffles 6 mm @ 4μm high iota divertor 19

Thermographic observations at W7-X PFCs at W7-X will be protected from overheating with a system of 10 IR/VIS endoscopes (divertor) and ICRH/ECRH protection systems During steady-state operation (OP2) ca. 50 cameras will monitor W7-X PFCs delivering huge amount of data. For safety/plasma control diagnostics near real-time, automatic analysis required. 20

Safety based on thermographic systems Hot spot detection based on threshold analysis for given ROIs. Critical events reported to central control system. next talk by Aleix Puig Sitjes Automatic strike line detection with contour analysis based on Fourier descriptors Special scenario with modulated power developed to characterize surface layers/divertor structure failures. This will provide more insight into development and thermal properties of surface layers and hopefully allow to minimize fail positives. Create the algorithms automatically identifying features in images. Courtesy: F. Pisano, B.Cannas 21

Outlook: Automatic data processing Large amount of data requires automatic analysis during the discharges as well as in post processing. For physics post-pulse automatic preanalysis, automatic event recognition, data stored in layers with different levels of details. Searchable, flexible, universal database of events connected to real data in archive. Large database allows to establish risk-based analysis of hot spots. Control of strike line aided with NNs. 3D codes too time consuming for real time control necessary Neural Networks (NN) Sensuators: IR cameras, magnetic diagnostics, thermocouples, Actuators: Heating and control coils Parametrization Robustness of NN towards noise in data Compatibility of simulated and experimental data Courtesy: D. Böckenhoff, H.Hölbe 22

Thank you!! 23

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Effects not yet taken into account in the modeling Volume Drifts ^ B recombination example from W7-AS: From: P. Grigull et al., J. Nucl. Materials 313-316 (2003) 1287

PFC Installation Completed in W7-X (4/2017) Start of plasma Operation in September 2017 R. König et al. PFMC 2017 10

Technical set-up of W7-X divertor

Magnetic Island Divertor Stepwise divertor integration OP 1.2a Test divertor unit (TDU) made of uncooled graphite tiles Identical shape to later HHF Completion of first wall protection (uncooled graphite tiles on CuCrZr heat sinks) 8 MW/m 2, 80 MJ OP 1.2b Test of uncooled scraper elements in module 51 and 30, protecting edges of divertor at low iota pumping gap OP 2 High heat flux (HHF) divertor with carbon-fiberreinforced carbon (CFC) tiles Complete water cooled liners for plasma vessel and ports Heat loads of up to 10MW/m 2 Marcin Jakubowski, et al., SOL in initial campaign of W7-X, PSI 2016, Rome CuCrZr + C- tiles Scraper element TDU

First Experimental Campaign on W7-X First campaign on W7-X conducted from Dec 10th, 2015- Mar. 10th, 2016 3.2.2016, 15:21:25.822 (local time) All technical and scientific objectives successfully achieved (cf. [1]) allowed the W7-X Team to safely increase technical limits (e.g. 2 MJ released heating energy to 4 MJ About 20 diagnostics successfuly commissioned and delivered results Some figures from initial campaign t pulse up to 6 s P ECRH up to 4 MW T e up to 9 10 kev Plans for OP1.2 campaign (2017-2018) t pulse up to 60 s P ECRH up to 8 MW T e up to 9 10 kev T i up to 2 kev T i up to 3 kev n e ~ 2 3 x 10 19 m -3 t E ~ 150-200 ms [1] Sunn Pedersen et al., Nucl. Fusion 55 (2015) 126001 n e ~ 2.0 x 10 20 m -3 t E ~ 0.5 s

Different time constants for different PFCs Divertor surface temperature in equilibrium after ca. 3 seconds Source: J. Boscary, KiP Design Review 30

Characterisation of divertor strike line Standard configuration f = f = -12-18 f -6 = 0 24

32 Initial campaign with 5 graphite limiters. 2015 / 2016 5 MW 4 MJ 6 sec Inertial cooling 3D scrape-off layer consists of 3 different flux tubes with different connection length: L c = 36 m, 43 m, 79 m

33 The W7-X Island Divertor Structure Pumping gap plasma contour target plates divertor units triangular plane triangular plane R=5500 Z=0 bean plane Z=0 R=5500 Z=0 R=5500