ASIPP. G. -N. Luo. Division of Fusion Reactor Materials Science & Technology Institute of Plasma Physics, CAS, Hefei, Anhui, China

Transcription

1 Plasma Facing Components (PFCs) for Magnetic Confined Fusion Devices G. -N. Luo Division of Fusion Reactor Materials Science & Technology Institute of Plasma Physics, CAS, Hefei, Anhui, China Joint ICTP/CAS/IAEA School & Workshop on PMI in Fusion Devices, July 18-22, 2016, Hefei, China

2 Outline Fusion reactor What are PFCs PFCs for EAST PFCs for ITER PFCs for DEMO 2016/7/18 Plasma Facing Components 2

3 Outline Fusion reactor 2016/7/18 Plasma Facing Components 3

4 Roadmap for Magnetic Fusion China Nuclear facilities, CMIF/IFMIF, etc. for materials & tritium CFETR/DEMO Fusion Power Plant ITER 2030 s 20?? EAST HL-2M 2025 Current devices Non-nuclear facilities for engineering design/test 2016/7/18 Plasma Facing Components 4

5 China Fusion Engineering Testing Reactor (CFETR) Cryostat SC Magnets Blanket Diagnostics CD & H Port plug Divertor Vacuum vessel CASK And tritium plant, power supplies, vacuum, cooling, cryogenics, remote handling, hot lab for maintenance, etc., subsystems! 2016/7/18 Plasma Facing Components 5

6 Outline What are PFCs? 2016/7/18 Plasma Facing Components 6

7 What are PFCs? PFCs are components facing directly the high T plasmas, protecting vacuum chamber and other in-vessel components PFCs comprise large-area first wall and divertor surfaces, as well as smaller but vital systems (startup limiter, antennas) PFCs must withstand intense plasma heat & particle fluxes, and function with high n wall loading & bulk nuclear heating Key requirements Low surface erosion from sputtering and plasma transients Low plasma contamination & Long surface/structural lifetimes Minimum tritium retention or permeation & nuclear activation Strong enough resistance to electromagnetically induced loads 2016/7/18 Plasma Facing Components 7

8 Major PFCs FW & divertor Particle flux ~10 20 /m 2 s Heat flux 0.5~2MW/m 2 Neutron wall loading 0.5~1.5MW/m 2 Area m 2 Blanket system First wall Divertor system Particle flux /m 2 s Heat flux 10~20MW/m 2 (SSO) Area Whole: ~100m 2, Strike point: ~0.1-1m /7/18 Plasma Facing Components 8

9 Harsh environments 5.7m Extreme T & gradients ~10 8 K ~10 5 K ~10 3 K ~3 K Core Plasma Edge Plasma FW/Div SC Coils 2016/7/18 Plasma Facing Components 9

10 Harsh environments Magnetic field High field side : 10-18T Core space: 6-10T Low field side : 4-6T Variation with space : several T/m Variation with time : several T/s /7/18 Plasma Facing Components 10

11 Key issues of PFCs Choice of materials and coolants Joining of PFM and heat sink material Capability of power handling (HHF testing) Reliability and integrity of PFCs in operation Effect of neutron irradiation on the joints Quality control of mass production The PFCs will be crucial both for achievable plasma performance & machine availability! 2016/7/18 Plasma Facing Components 11

12 Materials for PFCs PFM carbon based (C & CFC), beryllium, refractory (W & Mo) Advantage Disadvantage C/CFC Be W (Mo) 1. Low Z material 2. Compatible with plasma 3. No melting 1. High sputtering 2. Chemical erosion 3. High codeposition 4. High retention 1. Low Z material 2. Medium retention 3. Oxygen uptaker 1. Low melting point 2. High sputtering 3. Toxicity 1. Lowest sputtering 2. Highest melting point 3. High thermal conductivity 4. Low retention 1. High Z material 2. Brittle material 3. H/He effects Heat sink CuCrZr alloy and RAFM steels (possible ODS-strengthened) Advantage Disadvantage CuCrZr 1. High thermal conductivity 2. Good mechanical properties 1. Degradation of mechanical properties after annealing 2. Radiation hardening 3. Neutron activation RAFM 1. Low neutron activation 2. Good mechanical properties 1. Low thermal conductivity 2. Difficult in joining (especially for the ODS ones) 2016/7/18 Plasma Facing Components 12

13 Cooling for PFCs Inertial cooling power removal via heat capacity & radiation Active cooling heat removal by pressurized water or helium Coolant Inertial Water Helium 1. Easy to 1. High capability of 1. Medium capability manufacture, power handing of power handing Advantage inspect and install (20MW/m 2 ) 2. Good performance (10MW/m 2 ) 2. Feasibility of high 2. No leakage risk of steady state temperature wall heat flux operation 1. Low capability 1. Difficult to join 1. Very difficult to Disadvantage of power handling PFM and heat sink 2. High temperature manufacture and inspect gradient of PFCs 2. Low reliability 2016/7/18 Plasma Facing Components 13

14 Water cooled PFCs flat plate or Be or C HIP / VHP VPS/CVD/PVD or pure Ti or filler interlayer VHP/ Brazing Advantage: High heat transfer capability Easy to manufacture and inspect Withstand heat loads of ~5MW/m 2 Disadvantage: Singularity of thermal stress Peeling-off in case of interface failure Flat plate PFC w/ round cooling tube Flat plate PFC w/ hypervapotron cooling structure 2016/7/18 Plasma Facing Components 14

15 Water cooled PFCs monoblock CFC or W CuCrZr tube Advantage: Lower thermal stress than flat plate Intrinsic safety in case of failure (no peeling) Withstand heat loads of 10-20MW/m 2 Disadvantage: Difficult to manufacture and inspect Very expensive AMC (CFC) Brazing (CFC) Direct Cu casting Cu interlayer Brazing HIP & HRP (diffusion bonding) 2016/7/18 Plasma Facing Components 15

16 Helium cooled PFCs design P. Norajitra, et al, JNM (2009) High wall temperature to reduce hydrogen retention and recycling Pressurized helium jetting flow to reinforce heat transfer capability Difficult to manufacture and inspect M.S. Tillack, et al, FED 86 (2011) /7/18 Plasma Facing Components 16

17 1E-6 m/m Thermal stress analyses Thermal expansion coeff. at C Thermal stress distri. at 20MW/m 2 Cu alloy Be CFC W Large difference of thermal expansion btw PFM and heat sink (HS) Debonding of PFM/HS joints Leaking of HS and other joints F. Crescenzi, et al, FED 89 (2014) /7/18 Plasma Facing Components 17

18 Electromagnetic (EM) loads Plasma instabilities such as disruptions and vertical displacement events (VDEs), give rise to severe EM transients, and thus strong forces on PFCs and other in-vessel components. These can be calculated via combining the plasma simulation code like DINA for time variation of plasma and halo currents, with the finite element code like ANSYS-EM. Sunil Pak, et al, FED 88 (2013) /7/18 Plasma Facing Components 18

19 PFCs for EAST 2016/7/18 Plasma Facing Components 19

20 PFMC evolution in EAST W W W Mo C Full C PFC Mo-FW + C-Div Mo Full SS PFC 1 st plasma Mo C C C W&C-Div+Mo-FW 2018~2019 / Full W PFC 2016/7/18 Plasma Facing Components 20

21 Design of W/Cu divertor EAST goal: long-pulse high performance plasma operation Conceptual design Engineering design 2016/7/18 Plasma Facing Components 21

22 Design of W/Cu PFCs 80 Cassette Bodies Monoblocks 720 Monoblock PFUs 240 Flat type PFUs Dual chamfering for EAST 2016/7/18 Plasma Facing Components 22

23 Flowchart of manufacturing Cu,Cr&Zr melting Ingot drawing CuCrZr tubes NDT NDT HIP W block HIP W/Cu monoblock W powder sintering W compact rolling W plate NDT Pure Cu W bar HIP W/Cu slice Cu,Cr&Zr melting Ingot rolling CuCrZr plate NDT HIP NDT Machining Raw mater. PFU manuf. W/Cu mono-block PFUs Baffle Dome upper Dome lower panel NDT+leak check NDT EBW of endcup+pfus+trans-plate+baffle+legs+in/outlets EBW of two halves+legs+in/outlets Target DOME He leak check at 1.5MPa/180 Assembly to CB and in-vessel installation Assembly 2016/7/18 Plasma Facing Components 23

24 ITER-like monoblock W/Cu PFC W/Cu monoblocks prepared employing Hot Isostatic Pressing (HIP) technology (900 0 C, 100MPa) W/Cu PFUs manufactured successfully by HIP technology (600 0 C, 100MPa), properties of CuCrZr after HIP satisfy the requirement US-NDT results: Bondings between monoblocks/ofc/cucrzr excellent Annealing behavior of CuCrZr tube 2016/7/18 Plasma Facing Components 24

25 Flat-plate W/Cu PFCs Casting + HIP: The interface of W/Cu were joined by casting (1200⁰C), and then the interface of Cu/CuCrZr was bonded by HIP at lower temperature of 500~600⁰C. NDT results: bondings between W tiles/ofc/cucrzr plate excellent BAFFLE Plate 2016/7/18 Plasma Facing Components 25

26 Parts joining: E-beam welding Joining between monoblock units with endbox/manifold and flat baffle Joining of two-half parts for dome structural design 1.2 mm 3.1 mm CuCrZr Supporting legs and inlet/outlet cooling tubes joined to CuCrZr heat sink EB: 60kV, max 100mA; Seam: > 3mm deep, ~1mm wide at surface 2016/7/18 Plasma Facing Components 26

27 E-beam HHF testing W/Cu monoblock PFU: survived 1000 cycles of heat load of 10MW/m 2, cooling water of 4m/s, 20⁰C, 15s/15s on/off cycles W/Cu flat type mock-up: 1000 cycles, 5MW/m 2, 4m/s, 20⁰C 2016/7/18 Plasma Facing Components 27

28 US-NDT for W/Cu PFCs US-NDT for monoblock PFU Single probe: scanning the inner surface The defects of Φ1mm in the interface of W/Cu and Cu/CuCuZr was detected clearly using this set-up W/Cu mono-blocks and 720 PFUs tested More than W/Cu slices and 240 flat PFUs have been tested by this set-up with detection limit of Φ1mm NDT for flat type PFU 2016/7/18 Plasma Facing Components 28

29 Helium leak detection for PFCs Evaluation of welding quality and reliability for EAST OVT, IVT and DOME components During 180⁰C for 20 min under 1.5 MPa (He inside) Background vacuum: < 5.4x10-3 Pa; leak level: 2x10-11 Pa.m 3.s -1 Acceptance criteria: 1x10-10 Pa.m 3.s components (80 OVT, 80 IVT, 80 DOME) were tested 2016/7/18 Plasma Facing Components 29

30 W(upper) + C (lower) divertors + Mo-FW W Mo C 2016/7/18 Plasma Facing Components 30

31 H port dedicated to PWI research on EAST Material and Plasma Evaluation System (MAPES) - Maximum sample weight: 25kg - Sample holder moving velocity: 1-15mm/s - Maximum sample diam.: 500mm - Can insert into LCFS - Sample water-cooling & heating - Gas puffing system - Diagnostics: Langmuir probes, thermocouples, spectroscopy, 2016/7/18 Plasma Facing Components 311

32 Multiple diagnostics for PWI research on EAST Divertor detachment High resolution spectrometer Divertor gas puff system attached detached Edge Langmuir probes Multichannel optical system Postmortem analysis International collaboration Laser Induced Breakdown /Ablation Spectroscopy (LIBS/LIAS) 2016/7/18 Plasma Facing Components 322

33 W monoblock shaping # Tolerance of W PFC surface - Toroidal: 2 mm - Neighbor: 0.5 mm Monoblock shaping - dual 1mm 1mm chamfering B t Outer divertor W monoblock geometry optimization for steady condition X.H. Chen et al, FED /7/18 Plasma Facing Components 33

34 Effective W Sputtering Yields in EAST W erosion yield governed by C ion bombardment Better W erosion yield suppression by Li coating compared to Si one, lower than those in ASDEX Upgrade, but still higher than that in JET 2016/7/18 Plasma Facing Components 34

35 ELM Dominated W Erosion in H Mode Discharge W atom flux / (m 2 s ) EAST #: DN / 2.3T W atom (intra-elm) W atom (inter-elm) Js (intra-elm) Js (inter-elm) Js (A / cm 2 ) photons/(s cm 2 sr) EAST Shot: WI 400.9nm CII 426.7nm Si II 413.1nm R-R osp (cm) W erosion during ELM accounts for about 70% of total W erosion amounts in an ELM cycle on average. Intra-ELM W influx arises from proximity of outer strike point and Inter-ELM W flux peaks slightly away from outer strike point. A cm -2 MW m Js Peak q Time ( s ) 2016/7/18 Plasma Facing Components 35

36 PFCs for ITER 2016/7/18 Plasma Facing Components 36

37 ITER PFCs PFCs in ITER are mainly divided into two regions based on their different functions: Shielding Blanket (First Wall) Divertor All of PFCs are designed as multimodules, which can be installed and uninstalled by remote handing, considering future radioactive case. ITER is a Nuclear Facility INB-174 (in France). (First Wall) 2016/7/18 Plasma Facing Components 37

38 FW(Blanket) Major functions: Absorbing radiation and particle heat fluxes from the plasma Contribute to shielding to reduce heat and neutron loads in the vacuum vessel and ex-vessel components Provide limiting surfaces that define the plasma boundary during start-up and shutdown Provide passage for the plasma diagnostics Procurement: First wall (NHF) EUDA ~50% First wall (EHF) RFDA ~40%, CNDA ~10% Shield block CNDA ~50%, KODA ~50% Blanket modules RFDA ~40% Materials FW: flat Be/Cu/SS; Shield block: SS FW Block FW Block 2016/7/18 Plasma Facing Components 38

39 FW/Blanket B. Bigot, Progress towards fusion at ITER, ICFRM-17, Aachen, Germany, Oct /7/18 Plasma Facing Components 39

40 Divertor Major functions: To minimize the impurity content of the plasma To absorb radiation and particle heat fluxes from the plasma while allowing neutral particles to be exhausted to the vacuum system Provide passage for the plasma diagnostics Provide shielding to VV and external components Procurement: Outer Vertical Target JADA Inner Vertical Target EUDA Dome RFDA Cassette and Integration EUDA B. Bigot, Progress towards fusion at ITER, ICFRM-17, Aachen, Germany, Oct /7/18 Plasma Facing Components 40

41 Divertor Vertical Target Plasma-Facing Unit Monoblock with swirl tape at high heat flux handing area W armour/ofcu/cucrzr-ig Tube to tube joint CuCrZr-IG/ Alloy625/ 316L pipe Dome Plasma-Facing Unit W/Cu flat tile with CuCrZr/316L(N)-IG hypervapotron coolant channel 2016/7/18 Plasma Facing Components 41

42 Divertor 2016/7/18 Plasma Facing Components 42

43 Divertor 2016/7/18 Plasma Facing Components 43

44 Divertor 2016/7/18 Plasma Facing Components 44

45 High Heat Flux Testing ITER Divertor Test Facility (IDTF) Location: Efremov Institute, St- Petersburg, Russia Purpose: Performance and series tests of Divertor PFCs Operating principle: HHFT of component s surface by e-beam Maximum beam power: 800 kw Maximum accelerating voltage: 60kV Vacuum level: 10-5 mbar Advanced system of diagnostics R.A. Pitts, et al., 18 th ITPA DivSOL Topical Meeting, Hefei, China, March /7/18 Plasma Facing Components 45

46 HHFT on OVT PFUs Straight W part of PFUs subjected to the tests: 20 monoblocks of 12mm (axial) x 28mm(poloidal) x 8 mm (thickness at the top of the tube) 5 x 4 monoblocks Heat loads: 10s on and 10s off Target of Thermal hydraulic parameters: Inlet: 3.9 MPa; 11 m/s; 70 ⁰C to be used in ITER Window type mask 2016/7/18 Plasma Facing Components 46

47 Temperature, [ 0 C] HHF test results Average surface temperature Normal (IR) (13 tile) Hottest (IR) (15 tile) Pyrometer (13 tile) 20MW/m 2 No significant increase of T surf observed for 5000 cycles testing at 10MW/m 2 No sudden crack openings at the W/OFC-OFC/Cu joints during whole testing Number of cycle Successfully withstand 5000 cycles at 10MW/m cycles at 20MW/m /7/18 Plasma Facing Components 47

48 PFCs for DEMO 2016/7/18 Plasma Facing Components 48

49 DEMO PFCs DEMO presents a much larger PMI challenge than ITER! Higher heat flux, longer pulses, higher duty factor 4x ITER s heat flux 5000x longer pulses 5x higher even short-term average duty factor Erosion, dust production, tritium retention and component lifetime issues are much more challenging due to DEMO s mission DEMO must show practical solutions that allow for continuous operation for at least 2 full-power years between PFC change outs ITER plans to change out divertors after ~ 0.08 full-power years at much lower power Many solutions used on ITER are not Demo-relevant Moderate fraction of radiated power Intermittent dust collection and tritium clean-up 2016/7/18 Plasma Facing Components 49

50 Helium cooled finger concept 10 MW/m 2 P. Norajitra, IHHFC, San Diego, CA, USA, Dec 10-12, /7/18 Plasma Facing Components 50

51 htc [W/m²K] Jet Impingement HtC, W/Km2 Single-jet Heat transfer coefficient (Low Re Number) k-e Suga's cubic k-w v' Spalart-Allmaras Multi-jet R, mm Distance from Jet Center [mm] 2016/7/18 Plasma Facing Components 51

52 Helium cooled T-tube concept Heat flux ~10MW/m 2 The max helium jet velocity ~230m/s Maximum heat transfer coefficient ~ W/m 2 K The maximum temperature of the W armor~1782⁰c M.S. Tillack, et al, FED 86 (2011) /7/18 Plasma Facing Components 52

53 Water cooled monoblock concept Marianne Richou, et al., FED 89 (2011) W is considered as the armour material with T operating window ⁰C Eurofer as heat sink material ( ⁰C) Cooling water: 325⁰C, 20m/s, 15.5MPa 2016/7/18 Plasma Facing Components 53

54 Summary PFCs are crucial both for achievable plasma performance & machine availability! EAST has achieved a full tungsten upper divertor, maybe full W-PFC in 2-3 years. ITER is finalizing its W-divertor design, and also validating the PFCs prototype technologies. Fusion society has started conceptual activities for DEMO PFCs, still long long way to go! 2016/7/18 Plasma Facing Components 54

55 Thanks! 2016/7/18 Plasma Facing Components 55