Superconducting Magnets for Fusion and the ITER Project

Transcription

1 Superconducting Magnets for Fusion and the ITER Project Presented by Joseph V. Minervini Massachusetts Institute of Technology Plasma Science and Fusion Center Cambridge, MA American Nuclear Society Northeast Region Meeting Wellesley, MA April 18, 2002

2 Contents Fusion Magnets ITER Project and the ITER Device ITER Central Solenoid Model Coil 3

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4 Progress in Magnetic Fusion Research Fusion Power 1,000 Megawatts ,000 Kilowatts Computer Chip Memory (Bits) 1,000 Watts Data from Tokamak Experiments Worldwide 1,000 Milliwatts Years 2/98

5 Status of Laboratory Fusion Experiments 100 LAWSON PARAMETER, n i t 20-3 (10 m s) E Q =1 DT Higher Density Magnetic T3 ALCA T10 FT TFR TFTR PLT TFR ALCC ATC Ignitor, CIT, FIRE TFTR DIII JET JET PLT JT-60U JET DIII-D JT60 Ignition ITER JET DIII-D DIII-D JT-60U Legend TFTR Moderate Density Magnetic D-T D-D CENTRAL ION TEMPERATURE, T i (0) (kev) #97GR041.ntau-98 FIRE

6 Water-cooled copper tokamak World s largest fusion device Joint European Torus (JET) Culham, England 5

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8 U.S. Fusion Energy Sciences Budget History and Dates of Major Fusion Program Reviews Fiscal Year Budget (FY 1999 $ in Millions) Foster ERAB NRC ERAB OTA ERAB NRC FPAC FEAC & SEAB SEAB & PCAST FEAC PCAST * FESAC Fiscal Year *Reviews Scheduled for 1999: SEAB, NRC, FESAC, Fusion Summer Study

9 U.S. Fusion Budget Vs. the Price of Crude Oil Fusion Budget 30 $ in Millions (Actual) Crude Oil* Dollars per Barrel Years *In Actual $ s from Energy Information Administration/Annual Energy Review 1998, Table 9.1, Crude Oil Price Summary, Refiners Acquisition Costs, Imported, Nominal, Web site: eia.doe.gov/p...w/monthly.energy/mer9-1 1/20/99

10 Superconducting Tokamaks Existing T-7 (Russia) Triam (Japan) T-15 (Russia) Tore-Supra (France) Planned TPX (US) - canceled ITER (Japan, Russia, European Union, Canada) JT-60 Super Upgrade (Japan) Under Construction SST-1 (India) K-Star (South Korea) HT-7U (People s Republic of China) 7

11 Tore Supra CEA de Cadarache, France The largest superconducting device operating at 1.8 K Cadarache 8

12 The International Fusion Magnet Technology Community is Focused on Superconducting Machines Wendelstein Stellarator is under construction in Germany (NbTi Superconductor in Aluminum CICC) Large Helical Device (LHD) is in Operation in Japan 27.6 mm PF Coil and NbTi Cable-in-Conduit Conductor 9

13 The Modular Strategy for MFE Second Phase Scientific Feasibility Third Phase Burning Demo Engineering Base Fourth Phase Electric Power Feasibility Commercialization Phase Economic Feasibility Three Large Tokamaks International Program JT-60 U Burning D-T JET TFTR Adv. Long Pulse D-D Materials Develop Scientific Foundation Non-Tokamak Configurations Long Pulse Adv. Stellarator Spherical Torus, RFP Spheromak, FRC, MTF Choice of Configuration Advanced DEMO Attractive Commercial Prototype Technology Demonstration (the overall Modular strategy includes IFE) Reduced Technical Risk Streamlined Management Structure Better Product/Lower Overall Cost Increased Technical Flexibility Faster Implementation

14 Superconducting Pulsed LN2 Cooled Pulsed Water Cooled 10

15 Burning Plasma Physics -The Next Frontier The US Fusion Program will perform a Common Assessment of ITER, FIRE and IGNITOR at Snowmass, CO in July 2002 International Thermonuclear Experimental Reactor (ITER) (EU, JA, RF, Canada) Fusion Ignition Research Experiment - (FIRE) (US) IGNITOR (Italy) 11

16 What is ITER? The International Thermonuclear Experimental Reactor (ITER) Project is an international energy research effort whose purpose is to demonstrate for the first time the technological feasibility of fusion energy as a source of electric power. ITER is an outgrowth of discussions at the Presidential levels among the EU, Japan, the former Soviet Union, and the US between 1985 and ITER means the way in Latin. 12

17 ITER Objectives Programmatic Demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. Technical Demonstrate extended burn of DT plasmas, with steady state as an ultimate goal. power amplification Q = 10 during an inductive burn of = 300 s aim at demonstrating steady state operation with Q = 5 average 14 MeV neutron wall load = 0.5 MW/m 2 average lifetime fluence of = 0.3 MWa/m 2 Integrate and test all essential fusion power reactor technologies and components. Demonstrate safety and environmental acceptability of fusion. Strategic A single integrated device answering all feasibility issues needed to define a subsequent demonstration fusion power plant (DEMO), except for material developments to provide low activation and larger 14 MeV neutron resistance at least for in-vessel components. 13

18 ITER EDA History Developed design capable of ignition - large and expensive. ITER Parties (Euratom, Japan, RF, USA) endorsed design but could not afford to build it USA withdraws from project. Remaining Parties search for less ambitious goal - moderate plasma power amplification at about half the cost Goal achieved with new design fully documented and endorsed. (ITER-FEAT) End of EDA and start of negotiations on construction and operation, and of coordinated technical activities in their support. Canada offers site and joins as a full participant. These results were achieved at the expenditure of $660M (1989 values, USA $120M) on R&D and 1950 (USA: 350) professional person years of effort. 14

19 Will The US Rejoin ITER? House Science Committee: "The Administration's request for the Fusion Energy Sciences Program is $257.3 million, far short of the $335 million approved by the House in H.R. 4 [energy legislation]. Fusion's potential to wean the Nation from fossil fuels is tremendous, but much research remains to be done before that potential can be realized. The Committee notes with approval that the Administration is reassessing the potential U.S. role in the International Thermonuclear Experimental Reactor(ITER), which may significantly advance the science by achieving sustained-burning plasma. The Committee believes that U.S. participation in such important international research endeavors deserves serious consideration. 16

20 Will The US Rejoin ITER? Testimony of Ray Orbach, New Director of DOE Office of Science: Orbach expressed enthusiasm for the progress that has been made in fusion research, and hopes additional funding will be forthcoming. He said "our investment has really begun to pay off," and said that a decision would be made by Secretary Abraham within the next few months about whether the United States should rejoin the ITER project. "I personally would be very much in favor of" such participation, Orbach said, quickly adding that the decision would be made by Abraham and ultimately President Bush. Orbach envisions the U.S. as a junior partner in ITER, pointing to our position in the Large Hadron Collider, with a capped U.S. contribution,as a model. 17

21 Proposed Canadian ITER Site Clarington, Ontario 19

22 Roadmap ITER 3 years DEMO/PROTO IFMIF H. Bolt, 4/10/02 EFDA reference case demonstration of the feasibility of fusion power generation

23 ITER Parameters Total fusion power 500 MW (700MW) Q = fusion power/auxiliary heating power =10 (inductive) Average neutron wall loading 0.57 MW/m 2 (0.8 MW/m 2 ) Plasma inductive burn time = 300 s Plasma major radius 6.2 m Plasma minor radius 2.0 m Plasma current (inductive, I p ) 15 MA (17.4 MA) Vertical flux surface/separatrix 1.70/1.85 flux surface/separatrix 0.33/0.49 Safety flux surface 3.0 Toroidal 6.2 m radius 5.3 T Plasma volume 837 m 3 Plasma surface 678 m 2 Installed auxiliary heating/current drive power 73 MW (100 MW) 21

24 Design - Main Features Central Solenoid Blanket Module Outer Intercoil Structure Vacuum Vessel Cryostat Toroidal Field Coil Port Plug (IC Heating) Poloidal Field Coil Divertor Machine Gravity Supports Torus Cryopump 22

25 Design - Magnets and Structures (1) Superconducting. 4 main subsystems: 18 toroidal field (TF) coils produce confining/stabilizing toroidal field; 6 poloidal field (PF) coils position and shape plasma; a central solenoid (CS) coil induces current in the plasma. correction coils (CC) correct error fields due to manufacturing/assembly imperfections, and stabilize the plasma against resistive wall modes. 23

26 ITER Magnets Overall Magnet System Parameters Number of TF coils 18 Magnetic energy in TF coils (GJ) ~ 41 Maximum field in TF coils (T) 11.8 Centering force per TF coil (MN) 403 Vertical force per half TF coil (MN) TF electrical discharge time constant (s) CS peak field (T) 13.5 Total weight of magnet system (t) ~ 9,000 24

27 Design - Tokamak Building Provides a biological shield around cryostat to minimize activation and permit human access. Additional confinement barrier. Allows (with HVAC) contamination spread to be controlled. Provides shielding during remote handling cask transport. Can be seismically isolated. 25

28 Direct Capital Cost Co mp o n e nt s/ S ys t e ms Dire c t Cos t ( kiua*) % o f To t a l Mag n e t Sy st e m s Ve s s e l, Bla n ke t, Div e rt or, Pum p ing & Fu e lling Cryo st a t & Th e rm al Shi e ld As s e m b ly Aux iliar ie s Bu ildi ng s Hea tin g & Cur re nt Drive (7 3 MW) Diagn os ti c s ( s t a rt -up s e t) To t a l Dire c t Capit a l Cos ts *1 kiua = $ M $ M M M 27

29 Lifetime Cost kiua* Co nst ruct io n Cos ts Dire ct ca pit al Man age men t & Su pport R&D During Cons t ruct ion Opera tio n Cos ts ( ave rage pe r ye ar) Pe rma nen t pe rso nn e l Ene rg y Fue l Maint e na nc e/ imp ro vemen t s De co mm is sioning ~ ~3 0 ~ 8 ~ *1 kiua = $ M $ M M M 28

30 R&D - Vacuum Vessel (L-3) View of full-scale sector model of ITER vacuum vessel completed in September 1997 with dimensional accuracy of ± 3 mm 30

31 R&D - Blanket Remote Handling (L-6) 31

32 CS Model Coil It is the world s most powerful pulsed superconducting magnet 13 tesla peak field 640 MJ stored energy Pulse rates up to 2 T/s Used world s largest production of Nb 3 Sn superconductor 28 metric tons Used first large scale production of Incoloy Alloy 908 superalloy (60 metric tons) Testing to it s limits is yielding a large database on operating margins and invaluable experience operating large-scale pulsed magnets Testing carried out under bilateral US-JA Collaboration Agreement. 37

33 Central Solenoid Model Coil US CSMC Team led by MIT in collaboration with LLNL and Lockheed-Martin as Prime Industrial Contractor Designed and fabricated Inner Module and Coil structure Performed supporting R&D on superconductors, conduit materials, joints Additional industry support for Magnet Systems and CSMC design and analysis from SWEC during EDA. 38

34 Typical Large Scale Cable-in-Conduit Conductor (CICC) Comprised of several components IncoloyAlloy 908 Conduit (structural materials) Supercritical helium flows in interstices and central channel (heat transfer, thermodynamics, fluid dynamics) >1000 superconducting wires (superconducting materials, electromagnetics) 40

35 CICC Scale is 10 s of mm while superconductivity is on scale of a few microns Strand (0.81 mm diameter) Sub-element Bundle CICC (50 mm x 50mm) Superconducting Filament (~3 mm diameter) 41

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37 Fabrication of US-CSMC Inner Module Coil winding at Lockheed-Martin, San Diego, CA 44

38 Fabrication of US-CSMC Inner Module at MIT Hingham Facility Coil insulating Hingham Facility Vacuum furnace heat treatment at Wall Colmonoy, Dayton OH Shot peening and termination fabrication 45

39 CSMC is Composed of 3 Coil Modules + + CS Insert Coil (JA) US Inner Coil Module JA Outer Coil Module = Schematic Assembly of CSMC and Support Structure Coil Assembly in Test Facility 46

40 US CSMC Inner Module US Coil Vacuum Vessel 47

41 Overview of Model Coil Test Facility at JAERI, Naka, Japan Coils assembled in the Vacuum Vessel 48

42 Central Solenoid Model Coil 50

43 CSMC Pulsed Performance Exceeded Design Requirements Goal for for CS CS Insert and CSMC T/s to to T Achieved T/s in in CS CS Insert close to to theoretical maximum prediction CSMC 11 b is is the limiting layer quenched at at 10.8 T at at T/s Low sensitivity to to db/dt up up to to 22 T/s Very stable performance of of the CSMC and CS CS Insert 51

44 R&D - TF Model Coil (L-2) The model coil has begun testing in in the TOSKA facility at at FzK, Karlsruhe, Germany The coil under manufacturing at Alstom (France) The TFMC at Toska Facility, FzK 54

45 Summary The US is considering 3 options for a next step Burning Plasma Experiment ITER (superconducting) FIRE (pulsed resistive, LN2 cooled) IGNITOR (pulsed resistive, LN2 cooled) US Fusion Community will assess all 3 at Snowmass, CO in July US Congress and the Administration are considering rejoining ITER Superconducting magnets are required for a reactor relevant fusion device Significant progress has been made in large scale superconducting magnet R&D toward that goal 55

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