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DEMO / FPP Main Sub-systems

Key R&D Issues for DEMO an analysis based on functional requirements JET, 2 February 2009 J.Pamela, EFDA, with contributions from A.Bécoulet, CEA, D.Borba, EFDA, J-L.Boutard, EFDA, L.Horton, IPP, D.Maisonnier, EC. MAINTAINANCE. Remote. Handling. System. Cryostat. Poloidal Field Coil.

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DEMO / FPP Main Sub-systems

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  1. Key R&D Issues for DEMOan analysis based on functional requirementsJET, 2 February 2009J.Pamela, EFDA, with contributions from A.Bécoulet, CEA, D.Borba, EFDA, J-L.Boutard, EFDA, L.Horton, IPP, D.Maisonnier, EC

  2. MAINTAINANCE Remote Handling System Cryostat Poloidal Field Coil Toroidal Field Coil Breeding Blanket Power Conversion System Supply Electric Power D + T + ashes Heating and to the Grid Pumping Current Drive D + T T Systems Isotope T Separation ELECTRICITY PRODUCTION He LONG PULSE / C.W. He FUEL CYCLE Li D D Re-circulating electric power to the H&CD system PeHCD DEMO / FPP Main Sub-systems

  3. DEMO Key step before the first generation of FPP’s: => qualify all components and processes required for the first FPP => prepare forenvironmentally friendly power production (reduced/low activation materials) => prepare for cw operation Reactor grade functional requirements are likely to constitute key drivers in DEMO design and R&D: • Closed Fuel cycle • Efficiency of electricity production • Maintainability • Reliability DEMO R&D needs (beyond the well known breeding blanket and materials issues) are analysed here in light of these functional requirements

  4. Only efficiency and availability (reliability and maintainability) addressed in this talk

  5. 1- Efficiency of electricity production Cost of electricity (CoE) studied in the frame of the Power Plant Conceptual Studies (courtesy D. Ward, UKAEA): r= discount rate F = learning factor A = plant availability ηth = thermodynamic efficiency Pe = unit size βN = normalised plasma pressure N = multiplier of the density compared to the density limit Overall current drive efficiency ηCD not explicit in formula. Non-linear impact on overall efficiency through re-circulating power

  6. Recirculated power and plant efficiency Example: steady state tokamak power plant (D.Ward, EFDA Task TW6-TRP-002) • R=7.5m, a=2.5m, B=5.86T, I=19.4MA • hybrid mode of plasma operation (advanced H-mode): H=1.3 and bN=3.6 • high current drive efficiency 0.38 1020 A/Wm2 (yielding about 0.05 A/W at n=1.03 1020 m-3) that assumes 2MeV Neutral Beams (technology well beyond anything available or under development today) • bootstrap current fraction 0.54 => 9 MA current remain to be driven • 180 MW of H&CD power need to be coupled to the plasma • high temperature He cooled blanket and divertor assumed => 50% thermodynamic efficiency

  7. Recirculated power and plant efficiency Example (ctd): 2.4 GW fusion power converted into 1.56 GW of gross electric power (assuming 1.18 energy multiplication factor in the blanket) 1/3 of gross electric power need to be recirculated within this plant to • pump He (~200 MW), • power the H&CD systems (~300 MW with veryoptimisticηCD) and • operate the balance of plant (~60 MW) MAIN PENALTIES: HELIUM COOLING AND H&CD

  8. Key parameters driving cost of electricity Not addressed in this talk and not relevant to DEMO: r and F and Pe Parameters addressed in this talk: • βN and N => tokamak physics R&D • hCD => tokamak physics and H&CD technology R&D • hth => high coolant temperature => R&D on materials and cooling techniques • Magnets => potential gain on efficiency (balance of plant power) and on capital cost with High Temperature Supraconductors (HTS) • A => Maintainability and Reliability driven R&D

  9. High βN and N => tokamak physics requirements (1/8) Rationale: Fusion power strong function of - density (~n2) - pressure High βN also provides high bootstrap current fraction which reduces requirements on current drive

  10. With additional power: Steep edge pressure gradient: H-mode scenario Separatrix distance from axis H-mode pedestal X-point Divertor High βN and N => tokamak physics requirements (2/8) Standard tokamak operation: the H-mode scenario pressure L-mode distance from axis

  11. With additional power: Steep edge pressure gradient: H-mode scenario H-mode pedestal High βN and N => tokamak physics requirements (3/8) Standard tokamak operation: the H-mode scenario • Advantages: • Natural plasma state in the divertor tokamak configuration with additional heating power. • Minimal control requirements. • Disadvantages: • Limited plasma pressure achievable bN ~ 2-3. • - Instabilities associated to large edge pressure gradient (“ELMs”) pressure L-mode distance from axis

  12. High βN and N => tokamak physics requirements (4/8) Advanced Tokamak modes of operation with internal transport barrier H-mode plasma with an ITB • Advantages: • Improved core confinement and better stability properties leading to higher plasma pressure achievable bN ~ 4, as required for high DEMO efficiency. • Challenges: • Operation in steady state with advanced control techniques: • Plasma Profiles • Plasma Instabilities Layer of reduced transport H-mode pressure L-mode distance from centre ITB reviewed in R. Wolf PPCF 2003, Connor Nuc. Fus 2004, Challis PPCF 2004

  13. AUG DIII-D JT-60U JET 4 unstable 3 bN Advanced H-modes (Hybrids) 2 Standard H-mode 1 2 4 6 Pressure peaking: p0/<p> High βN and N => tokamak physics requirements (5/8) Advanced Tokamak mode of operation with internal transport barrier Operation in steady state requires advanced control techniques of the plasma profiles (pressure) This may constitue a major challenge in a burning DT plasma

  14. bN=3.3 5 1.0 0.8 0.6 current [MA/m2] 3 0.4 2 0.2 1 Normalised radius High βN and N => tokamak physics requirements (6/8) Advanced Tokamak mode of operation with internal transport barrier - Operation in steady state requires advanced control techniques of the plasma profiles (current) Field line twist in a tokamak: q q=2 q=1 Significant off-axis current drive is required to maintain qmin>2, in addition to the off-axis current produced naturally by the pressure gradient

  15. UNSTABLE Coils STABLE WITH CONTROL STABLE High βN and N => tokamak physics requirements (7/8) Advanced Tokamak mode of operation with internal transport barrier Operation at high pressure (βN> 4) requires qmin > 2 and active control of plasma instabilities.

  16. High βN and N => tokamak physics requirements (8/8) Operation at high density (n/nG > 1) - The empirical Greenwald density limit describes well the experimental results. • DEMO designs rely on operation close or above the empirical Greenwald density limit n/nG = 1.2 to 1.5

  17. Current drive efficiency hCD (1/7) Three steps1- conversion of electric power into power launched in plasma: conversion efficiency hconv2- coupling of launched power to plasma: coupling efficiency hcoupl3- current drive efficiency (current driven per power unit coupled to plasma eCD)Penet = Pegross - PeHCD – PeBoP = Pegross - ICD/hconvhcoupleCD – PeBoP

  18. H&CD Conversion and Coupling Current drive efficiency hCD (2/7) hconv = Plaunched / (Psource+Paux ) Paux Auxiliary PS, Cooling, pumping, cryogenics, … Coupling source transmitter Plaunched Psource Pcoupled Cooling, pumping hcoupl = Pcoupled / Plaunched Plost

  19. Current drive efficiency hCD => tokamak physics and technology requirements (3/7) Wall-plug to coupled power efficiency CONVERSION COUPLING (Technology) (Physics) CD efficiency (DEMO-like plasmas) PHYSICS high NNBI high Low (20-30%) low-medium ICRH medium Medium (40-50%) LHCD medium high Medium (30-40%) high ECRH low-medium Low(20-30%)

  20. Current drive efficiency hCD (4/7) ‘Wall plug’ efficiency of JET NB heating systems • Neutral Beam power to plasma: 25MW • Total power consumption: 116MW • HV beam acceleration : 99MW; Regulation/switching losses on HVPS: 12.6MW; Ion source: 2MW; Cooling: 0.6MW; He refrigerator: 0.65MW; Other: 0.5MW • Overall conversion efficiency: 25MW/116MW = 22% • Main losses / main opportunities for improvement: • neutralisation efficiency (~ 0.35 for positive ions) and • geometric beam transmission (~ 0.7, -> 0.78) Courtesy TTC. Jones, UKAEA

  21. Current drive efficiency hCD (5/7) Neutralisation efficiency for NBI improvement with negative ion beams Foreseen efficiency range of “advanced” neutralisation methods applicable to negative ions (photo-det. or metal vapour), Very demanding / remain to be developed JT60U ITER AUG JET D2 gas neutralier

  22. Current drive efficiency hCD (6/7) ‘Wall plug’ efficiency of Tore Supra ECRH system • Power to plasma (full system): 2.4 MW • Total power consumption: 10.7MW • Overall conversion efficiency: 2.4 MW/10.7 MW = 22% • Main losses: Gyrotron, tetrode series regulator, HV power supply • Opportunities for improvement: • Gyrotrons with depressed collector • ( e.g. 170 GHz 1 MW tube developed by Japan for ITER reach 55% efficiency with depressed collector; but… still insufficient => multi-stage collectors ?) • solid state HVPS Courtesy R. Magne, CEA

  23. There is a serious H&CD efficiency issue:wall-plug-to-coupled-power efficiency of present systems in the range 20-40%, while 60-70% are required for a c.w. DEMO

  24. Thermodynamic efficiency hth Optimisation of temperature of coolant (1/7)MATERIALS • MATERIAL DEVELOPMENT IS FUSION SPECIFIC • A NUMBER OF ISSUES STEMS DIRECTLY FROM THE PROPERTIES OF THE 14 MeV NEUTRONS AND THE REQUIREMENT FOR AN ENVIRONMENTALLY ACCEPTABLE ENERGY SOURCE: He and H production (14 MeV DT fusion neutrons) Strong requirements in terms of structural properties Heat sink & Armour Materials Reduced / Low activation OTHER REQUIREMENTS ON MATERIALS ARE MORE RELATED TO REACTOR GRADE FUNCTIONAL REQUIREMENTS, IN PARTICULAR HIGH TEMPERATURE OPERATION

  25. Thermodynamic efficiency hth Optimisation of temperature of coolant (2/7) • High Thermodynamic Efficiency: • Towards High Temp. of Coolant (He-cooling as reference option) Strong drive for materials R&D • DEMOBLANKET (~80 dpamax, He ,H): reduced activation Ferritic/Martensitic & Ferritic Steels • EUROFER: Tmax ~550 0C • ODS-EUROFER: Tmax ~650 0C • ODS Ferritic Steels: Tmax ~750 0C DIVERTOR (40 dpamax, He, H): Refractory W-alloys under assessment Armour materials: W-Y2O3, W-Si-Cr, W-TiC Structural materials: W-Ti, W-Ta W-V Temperature upper limit determined by Creep Strength Temperature upper limit Determined by Recrystallisation

  26. Thermodynamic efficiency hth(3/7) 105 h 102 h 5500C 6500C 7500C 8500C 9500C Upper Temperature Limit for Structural steels Estimated Upper Limit on the Basis of a Creep strength of 100 MPa for ~50,000 Hours

  27. Thermodynamic efficiency hth(4/7) Pressurised Water Low-medium Temperature He Relatively narrow Temperature Window Lower Temperature Limit as well : Radiation Induced Increase of Ductile-Brittle Transition Temp. DBTT Fission Reactor Data Conventional 9% Cr Steels & EUROFER: Significant loss of Fracture toughness at ~300 0C Operation at higher temperature, above 500C might help Fusion Relevant He Production might increase further the Loss of Fracture Toughness: Recent SINQ Results at ~10 dpa, 60-80 appmHe/dpa

  28. Thermodynamic efficiency hth(5/7) W W-26%Re W-La2O3 Upper Temperature Limit for W-alloys (divertor) Severe Plastic Deformation & Alloying: To Improve the Fracture Toughness of W and W-alloys Annealing 1 hour 835 0C Initial Microstructure Non-Affected Loss of Fracture Toughness 1200 0C Unacceptable Recrystallisation Recrystallisation resistance is unacceptably affected New alloys are being developed : W-Ti, W-V for structural applications, W-Y2O3, W-Si-Cr, W-TiC for armour. Behaviour under irradiation to be tested in a second stage R. Pippan et al.

  29. Thermodynamic efficiency hth(6/7) 10 MW/m2 W tile: max. allow temp. 2500°C max. calc. temp. 1711°C DBTT (irr.): 700°C Thimble: max. allow. temp. 1300°C max. calc. temp. 1170°C DBTT (irr.): 600°C ODS-Eurofer: He-out temp. 700°C He-in temp. 600°C DBTT (irr.): 300°C He-cooled Divertor Concept for DEMO Heat Flux Fatigue Testing ~10 MW/m2 ~100 cycles Infra-Red Image during testing P. Norajitra et al., FZK

  30. Thermodynamic efficiency hth(7/7) Optimisation of temperature ofcoolant is a complex issue • In the long term, high temperature cooling might be a must (He>550ºC), • to increase plant efficiency (although the balance between increased complexity and large circulating power required for high T operation might offset the gain) • but maybe also to avoid deterioration of materials strength (DBTT). • It is essential to continue R&D on He-cooled in-vessel components (Materials, welding and joining techniques, He technology). • However, it is not clear whether all He-related technologies will be mature enough for application on DEMO. • Helium is today a by-product of petroleum industry; it is a finite resource on earth. • It is therefore appropriate to consider back-up options, e.g. water-cooled DEMO, which requires specific R&D.

  31. Overall plant efficiency: High Temperature Superconductors (1/2) • PPCS and DEMO studies show magnets contribute to about 1/3 of the cost of plant ! • In addition 4 K cooling increases • plant complexity & cost • recirculated power (about 10 MW for the plant mentioned previously) • …and it impacts reliability. • Nb3Sn used on ITER with 11.8T limit on TF conductor • Prospect to extend use of Nb3Sn to 13.5T limit => increased margins/performance (test on-going on Dipole) • But Nb3Sn requires low T operation 4K (Liquid He) • Potential gain in operating at higher T (reduced recirculated power): • Ultimate target is for HTS working above 80K to allow liquid Nitrogen as coolant instead of liquid He and simplify cryostat

  32. Overall plant efficiency: High Temperature Superconductors (2/2) Potential candidate YBCO for TF coils, operation at > 65K … still a long way to go in terms of conductor performance and development

  33. 2- Availability DEMO shall qualify components and processes in reactor relevant conditions Machine availability is therefore a key parameter. MTBF: Mean Time Between Failure (or Mean Time Between Replacement) MTTR: Mean Time To Repair (or Mean Time To Replace) In DEMO, MTTR for all in-vessel components shall be minimized but will remain long (several months) MTBF should therefore be considerably longer to guarantee a high availability => hence high requirements on Maintainability (minimise MTTR) & Reliability (maximise MTBF)

  34. 3- Maintainability (1/2) Maintenance => significant impact on Design • RH procedure and equipment • RH manual (requirements for components to be handled) • suitable reactor layout (access, transfer of components…) • A key issue: blanket maintenance • ITER blanket segmentation not reactor relevant (too many modules  replacement too long). • No satisfactory solution available. One option considers the transfer of large “banana-like” segments from top vertical ports => R&D will be precisely defined at the end of DEMO Conceptual Studies (which remain to be launched …)

  35. 3- Maintainability (2/2) Local inspection and maintenance may require special systems in addition to fixed units Picture: the AIA (Articulated Inspection Arm) has just completed its first inspection in Tore Supra under high temperature and UHV conditions.

  36. 4- Reliability (1/11) Some of the main issues expected: • Reliability of tokamak operation: avoidance of disruptions and other events • Wall materials life-time: • Plasma compatibility with relevant metallic wall materials => R&D Physics • R&D on plasma facing materials • minimisation of power to divertor targets • Complexity of in-vessel systems • 14 MeV neutron fluxes, up to 80dpa before maintenance! => R&D on materials and component Avoid copper as a heat sink in PFCs and in H&CD components under neutrons (swelling & activation) • Complexity of H&CD systems(significant part of the plant if large CD required !) • Diagnostics and control requirements

  37. 4- Reliability (2/11) Reliable Tokamak Operation Avoidance of Disruptions • “Disruptions”, abrupt termination of the plasma, need to be avoided in DEMO • Why ? • transient loads on plasma facing components (erosion/melting) • forces on vessel and in-vessel components • significant loss of availability (long recovery from disruption) • How ? • disruption avoidance (precursors) and mitigation • operation away from stability limits(avoid too ambitious plasma operation) Infrared view of the vessel during a disruption at JET showing the thermal loads in the plasma facing components

  38. ITER JET 4- Reliability (3/11) Compatibility between plasma operation and relevant wall materials • Carbon will NOT be used as plasma facing material on DEMO • Erosion • T trapping • loss of thermal conductivity under irradiation • Challenging development of plasma scenarios compatible with metallic walls • avoidance of edge localised modes (ELMs) • control of impurities • optimisation of plasma radiation etc. • ASDEX-Upgrade W-experiment • JET ITER-like Wall experiment (Be and W) • ITER W-Divertor foreseen for DT operation

  39. Plasma PRadiated PLOSS Targets 4- Reliability (4/11) Tokamak Operation Operation with high radiation fraction > 90 % Reduction of power loads on Plasma Facing Components will increase their lifetime This drives tokamak operation towards highly radiative modes => Impurity seeding of both divertor and core plasma …radiation levels limited by: dilution of the fuel; impurity accumulation and impurity sputtering of the tungsten target

  40. Plasma PRadiated PLOSS Targets 4- Reliability (5/11) Tokamak Operation and first wall/divertor life time If very high radiation fraction cannot be obtained routinely in a robust plasma scenario, there will be no other choice but reducing plasma performance to keep power loads at reasonable levels

  41. 4- Reliability (6/11) In-vessel components: blanket

  42. 4- Reliability (7/11) Swelling & loss of Conductivity => Avoid Copper or use DS Cu alloys Swelling under irradiation Swelling and Loss Electrical Conductivity under irradiation J. Nucl. Mater. 233-237 (1996) 127-137 J. Nucl. Mater. 191-194 (1992) 386-390

  43. (in addition, impact on waste) Gamma Dose Rate versus Decay Time (1) Activation => Use of Copper will be banned from DEMO and reactors plasma vicinity Irradiation condition : 5 years in the First Wall neutron spectrum • Waste Limits (2) • High Level Waste limit: 2E-2 Sv/h • Medium Level Waste limit: 2E-3 Sv/h • Low Level Waste limit: 1E-4 Sv/h • Hands-on Level limit: 1E-5 Sv/h • UKAEA FUS 509. Handbook of Activation Data Calculated Using EASY-2003 • After R. Lindau et al. Fusion Engineering and Design 75–79 (2005) 989–996

  44. 4- Reliability (8/11)Copper Copper broadly used today for components subject to neutron bombardment:- as heat sink material in plasma facing materials- to fulfill all high (electric/thermal) conductivity requirements in H&CD systems- all NB actively cooled components (source, accelerator, neutraliser, dumps, beam scrappers, duct protections…) - LHCD couplers - ICRH antenna straps and transmission linesReplacing copper (or at least drastically limiting its use in case swelling and loss of conductivity are acceptable) is a major and long-lead R&D challenge which is not sufficiently addressed today

  45. 4- Reliability (9/11) Operation and plasma control No Physics R&D on DEMO Most probably : • a limited number (possibly a single one) of very robust modes of plasma operation with limited control requirements • a single H&CD system • minimised number of diagnostics for operational purposes only • no or very few active diagnostics => Integrated DEMO conceptual studies should define rather early a set of constraints on plasma operation and control (including H&CD and diagnostics) to guide the development of plasma scenarios on present devices and ITER

  46. Reliability (10/11)H&CD systems Improving reliability requires significant R&D to reduce complexity and improve individual sub-systems …for example on Neutral Beams: • Cs consumption in source (Cs-free solution?) • Large number of sub-systems requiring remote handling for maintenance • Specifically demanding maintenance of source • Gas pressure too high (losses, pumping requirements) • Injector conditioning techniques (impact on availability!) • HV breakdown avoidance/mitigation • Find substitute to copper • etc….

  47. Final comments onReliability (11/11) JET statistics on injected NB Power (courtesy TTC.Jones UKAEA) maximum nominal power 25 MW (attention is brought to the fact that the requested NB Power spans a broad spectrum from 5 to over 20 MW) Reliability == More Margins !!

  48. Summary on efficiency and availability driven DEMO R&D needs • (Breeding Blanket not addressed here but of course an essential component) • Materials for in-vessel components: • He production by 14 MeV neutrons and structural or functional requirements drive fusion specific R&D • High T operation (required to optimise CoE) puts severe additional requirements on materials • Limit the use of / find a substitute to copper as heat sink material • Maintenance: a very significant programme of RH will be needed but can be defined only once DEMO Conceptual Design is available

  49. Physics/Tokamak operation • Significant improvement in tokamak operation reliability required (avoidance of disruptions, ELMs etc.) • Minimisation of power loads on plasma facing components (e.g. high radiative fraction OR modest performance) • Compatibility between plasma scenarios and metallic wall materials • A GAP: no Tokamak (operating or foreseen, not even ITER) will provide demonstration of operation at DEMO/reactor relevant first wall temperature ! • Development of plasma scenarios towards high performance (bN, density) but with limited/reasonable current drive and control requirements => A TRADE-OFF BETWEEN PERFORMANCE AND RELIABILITY WILL BE NECESSARY

  50. Diagnostics • Limited in number as a result of an integrated development of plasma operation and control • Robustness: radiation resistant, high temperature, limited maintenance / requires significant and long R&D efforts • Heating and Current Drive systems: The needs for R&D in this area have always been dramatically underestimated and under-financed; mistakes of the past should not be repeated R&D efforts have been/are driven by ITER on beam energy (negative ions) or gyrotron frequency (170 GHy for ITER) but little is done otherwise => Significant simplifications and developments required • Overall efficiency of H&CD systems (from plug to plasma) has to be increased from 20-40% today to 60-70% • Duty cycle (from 10’s of seconds to cw operation; reduced maintenance needs) • very high reliability under nuclear operation

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