Status of EU DEMO Design and R&D Studies

Transcription

1 EFDA Power Plant Physics and Technology Status of EU DEMO Design and R&D Studies Gianfranco Federici and the PPPT Team Associazione Italiana di Scienza e Tecnologia XXI Congresso AIV Catania (Italy) May 16,

2 Outline Background DEMO Outstanding Challenges EU Fusion Roadmap Horizon 2020 and MAG Highlights of some achievements PPPT Future developments Summary 2

3 ASDEX-U EU Path to Net Electricity is based on a DEMOnstration Power Plant Still a divergence of opinions on how to bridge the gaps to FPP However, there are outstanding issues common to any next major facility after ITER, whether a CTF, a Pilot Plant, a DEMO, or else: Power Plant Power exhaust handling (divertor) Operating plasma scenario CD requirements, Coolant for IVCs breeding blanket concept Maintenance scheme plant architecture Structural and PFC materials Only some of these issues can be solved in ITER. DEMO JET ITER JT-60SA DEMOnstration Fusion Power: Reactor performance DEMO Net electricity production Makes its own fuel Availability and reliability operation over a reasonable time span. Materials for extreme environments

4 Net electricity production Several 100 MWs DEMO Top Level Design Requirements Low recirculation power required severe constraints on overall CD efficiency η CD γ CD Tritium self sufficiency DEMO makes its own fuel This places severe restrictions on port space for sensors and H&CD in DEMO. Local TBR ~ PFC armor must be much thinner to achieve TBR > 1 Disruptions and ELMs must be reliably eliminated! A robust solution of all physics and technical issues. interlinks Divertor exhaust loads, PFCs Materials & Techology Current drive systems Adequate availability/reliability operation over a reasonable time span Fast Remote Maintenance schemes to be developed and demonstrated. Simplified designs (with margins) Technologies and materials with established industrial manufacturing feasibility (high TRL!) 4

5 DEMO Physics Issues ITER will address key issues for DEMO beyond present day experiments, the most prominent example being α-heating The EU Program has identified DEMO physics challenges (items not needed for ITER Q=10, but absolutely vital for DEMO) Power exhaust (R DEMO /R ITER = 1.2, but P DEMO /P ITER = 4!) Steady state tokamak operation High density operation Disruptions (W DEMO /W ITER > 5!) Reliable control with minimum sensors and actuators Strong interlinks exist between physics and technology traditionally, physics assumptions have led to requirements for technology (example: divertor target has to stand + 20 MW/m 2 ) not always possible to match need an iteration 5

6 Top Engineering Challenges with potentially large gaps beyond ITER Power exhaust and divertor R&D strategy (M2) Power extraction/ Blanket technology incl. T self-sufficiency (M4) H&CD Systems Efficiency and Reliability (M6) Diagnostics (M1, M6) Reliability of Core Components & RH for high machine availability (M6, M4) Qualification of resilient materials (M3) Safety and licensing (M5) ITER will show scientific/engineering feasibility: Plasma (Confinement/Burn, CD/Steady State, Disruption control, edge control) Plasma Support Systems (Superconducting Magnets, fueling, heating/cd) ITER will not address fusion nuclear technology (most components inside the vacuum vessel are NOT DEMO relevant - not materials, not design) TBM provides very important information, but limited scope 6

7 DEMO Blanket Options There are still many challenging issues to be resolved for the development of a DEMO Blanket. Availability of external tritium supply for continued fusion development beyond ITER s first phase is an issue. T-self-sufficiency is a complex issue that depends on many system physics and technology parameters / conditions. * Vacuum Vessel: 316 SS water self-cooled; # Eurofer limit Coolant Breeder Materials T in / T out Comment HCPB He Li 4 SiO 4 /Be Eurofer 300/~500 o C # ITER TBM HCLL He LiPb Eurofer ~500 o C # ITER TBM WCLL Water LiPb Eurofer? 290 / 325 o C PWR conditions DCLL He (FW), LiPb (BB) LiPb Eurofer 300/~500 o C # Each concept has feasibility/performance issues. A selection now is premature. Still uncertainties/ technical risks. The choices of the materials and the coolant of the breeding blanket have to be made before start of the EDA taking into account maturity of BoP technologies.. 7

8 EUROFER will have narrow window of operation in a Fusion reactor? Upper temperature limit: 550ºC Irradiation at high temperatures (>350ºC ) allows thermal dispersion of damage dislocations Annealing of damage. but RAFM steels lose strength at ~ 550 C.and.Creep rupture becomes an issue at > 600 C There is indication that some EUROFER melts (not mainstream) might have improved embrittlement properties around 300 C may be usable with Water-cooled Lithium Lead blanket; evidence also from Japanese F82H-mod3 melt; candidate for development programme. 8

9 Why a New Fusion Roadmap European Commission proposal for Horizon 2020 ( ), follows the advice of an Independent Panel on Strategic Orientation of the Fusion Programme (Wagner Panel), stating the need of an ambitious yet realistic roadmap to fusion electricity by Central role of ITER 14 MeV neutron sources (IFMIF) for material qualification DEMO as a single step to the commercial power plant DEMO foreseen as a project starting construction in early 2030s Nuclear design requirements: First wall steel conservative value 15 dpa/fpy. Tungsten FW armour conservative value 5dpa/fpy. If Cu alloy substructure for divertor 5 dpa/fpy Phase I DEMO ~ Starter Blanket for 20 dpa/300appm He (steels) 1.33 fpy at 2GW th and 33% availability gives 4 years. Phase II Blanket would be 10 years at 33% availability. Divertor would start with a 2 fpy lifetime as envisaged in EU Power Plant Conceptual Study (2005). 9

10 A Pragmatic Approach Robust technical solutions and well established regimes of operation Inductive burning plasma scenarios may be enough. Adequate risk mitigation strategy for divertor solution Enlarge the basis for breeding blanket choice Targeted effort on enabling technologies (SC, H&CD, RH, pumping etc.) based on ITER R&D Material choices (MAG) Present portfolio of structural and high heat flux material may be enough but a number of high impact risk identified Some development of risk-mitigation materials with more advanced characteristics is essential Realistic approach to material qualification for DEMO with 14MeV n. Safety approach Safety against 'Design Basis Accidents' must be assured by 'passive safety' and 'defence in depth follow ITER experience Demonstrate all the technologies for a commercial FPP Possibility for DEMO to play the CTF role (in parallel with the participation to CTFs possibly built by International Collaborators) 10

11 Breeding Blanket Structural Steels EUROFER (8-9%Cr Reduced Activation Ferritic Martensitic RAFM - Steel) is confirmed by the MAG as Baseline structural steel (TRL4/5) Good overall balance of mechanical properties (strength ductility, fracture toughness, creep resistance, fatigue resistance); broad industrial experience in fabrication; sufficient corrosion resistance to LiPb for interface temp. <~475C; relatively good n-irradiation stability of a bcc material. Needs a risk mitigation programme on blanket advanced steels. MAG identifies two candidates at TRL~3 or above: High temperature FM steels from the non-fusion programme Oxide Dispersion Strengthened (ODS) RAFM or Ferritic Steels (Fusion programme) Note: each back-up already on its year journey of development MAG confirms Baseline for PFCs as Tungsten. Divertor of Early DEMO should be a water-cooled concept. Cu-based alloys (GLIDCOP or CuCrZr) heat sinks. More flexibility of engineering concepts; the Early DEMO coolant has to be fixed at the first phase, and cannot be changed. Advanced PFCs/HHF materials W-fibre reinforced materials most promising avenue for ductilisation of tungsten; W-laminates (currently under irradiation test); Functionally-graded materials (tungsten-copper composites) 11

12 Two DEMO design options are being analysed Provisional DEMO Design Options conservative or early DEMO deliverable in the short to medium term construction to be started in ~ 20 years from now ; based on expected performance of ITER with reasonable improvements in science and technology; eg (but not limited to!), a large, modest power density, long-pulse inductively supported plasma in a conventional plasma scenario. optimistic (advanced?) design, to be delivered on a longer term advanced physics at the upper limit of what may be achieved in ITERph2; Technology/ materials advances which will require validation at higher dpa eg., a higher power density, high non-inductive CD fraction, ss operation The next two slides show some provisional results of a range of design configurations we are exploring 12

13 demo1a_apr_8. Optimised to maximise pulse lenght demo1a_apr_100 Optimised to minimise plasma major radius demo1a_apr_mincost Optimised to minimise constructed cost R=9.0 m; a=2.489 m, A=3.62 R=9.151 m; a=2.597 m, A=3.524 R=9.94 m; a=3.24 m, A=3 P fus =1793MW;t b =1.7hrs; V p =1783 m 3 P fus =2040MW; t b =1.7hrs; V p =1982 m 3 P fus =1800MW; t b =1.7hrs; V p =3390 m 3 I=16.8 MA; <n e,vol >=9.327x10 19 m 3 I=18 MA; <n e,vol >=9.194x10 19 m 3 I=20.4 MA; <n e,vol > =6.686x10 19 m 3 B=6.5 T; B max(cond) = 12 T B=6.432 T; B max(cond) = 12 T B= 4.8 T; B max(cond) = 9.5 T <N WL > =1.1 MW m 2 <N WL > =1.17 MW m 2 <N WL > =0.76 m 3 P div =150.0 MW; P CD =50 MW P div =150.0 MW; P Heat =100 MW (γ=0.4) P div =151 MW; P Heat =50MW current best baseline, minimised R0 (minimising further reduces the pulse length. 50MW heating and current drive. similar to previous with the current drive efficiency halves, and 100MW instead of 50MW. Slightly larger and higher current/lower field, but 50MW H&CD, cost minimised. As for DEMO2, this tends to reduce the magnetic field and push the size up a little to thence recover the 13

14 demo2_12_apr_a35 Optimised to minimise Plasma major radius demo2_12_apr Optimised to minimise Plasma major radius. demo2_12_apr_mincost Optimised to minimise constructed cost R=8.764 m; a=2.504 m, A=3.5 R=8.152 m; a=2.982 m, A=2.734 R=8.405 m; a=2.982 m, A=2.354 P fus = MW; V p = 1765 m 3 P fus = MW; V p = 2363 m 3 P fus = MW; V p = 3550 m 3 I=15.86 MA; <n e,vol >=8.705x10 19 m 3 I=19.85 MA; <n e,vol >=7.683x10 19 m 3 I=23.54MA; <n e,vol >=6.356x10 19 m 3 B=6.8 T; B max(cond) = T B=5 T; B max(cond) = 12 T B=3.9T; B max(cond) = T <N WL > =1.31 MW m 2 <N WL > =1.175 MW m 2 <N WL > =0.94MW m 2 P div =105.0 MW; P CD =132 MW (γ=0.4) P div =101.0 MW; P CD =132 MW (γ=0.4) P div =95 MW; P CD =134 MW (γ=0.4) High A machine, intended to reduce Ip (at the expense of high field). Pdiv limited to ~100 MW. Big TF magnet -- the most expensive option under current PROCESS assumptions. A minimised R machine, allowing A to vary. Higher current than above but P_CD is ~ the same. Lower field and smaller magnets. Somewhat reduced wall loading due to lower aspect ratio/greater wall area. Minimizing constructed cost -- an intermediate size machine which has been pushed to much lower field as the magnets are strong cost drivers. 14

15 Novel Configurations Assessed Snowflake configuration developed Talk of F. Crisanti (ENEA, Frascati), Coil forces much higher than standard configuration (x2-4) Tear drop shape reduces plasma volume, but partly compensated by lower decay index allowing higher elongation Movement of strike points during core plasma perturbations cancels wetted area advantage of flux expansion R.Albanese, R.Ambrosino, M.Mattei Snowflake (SF) Quasi Snowflake (QSF) 15

16 Liquid Metal Divertor Targets? Experiments with liquid Li in a capillary porous system show good plasma compatibility no droplets However, technological issues must be solved before considering CPS a viable candidate Need heat removal concept that does not rely on strong evaporation (jxb forces on flowing LM!) For q> 10 MW/m 2 Sn is the best option Minimise T retention Rely on a concept for a closed LM cycle under steady conditions needed In 2012 Wettability tests for liquid (Li, Sn, Ga) - mesh material in CPS (SS, Mo, W) J.W. Coenen & T. Wegener (FZJ) Talk G. Mazzitelli (ENEA Frascati), Work to continue in 2013, including tests in Magnum PSI 16

17 Alternative Vacuum Pumps Experiments (proof-of-principle tests): 1. liquid metal (mercury) diffusion pump: proved to be an applicable technology for DEMO (continuous operation, T compatible, compact structure, arbitrary size) a jet stage was integrated, covers the full pressure range for burn and dwell 2. liquid metal ring pump: in collaboration with industry pump timely delivered to KIT and recently prepared for an extensive experimental campaign in THESEUS The infrastructure for the liquid ring pump testing was also developed and procured 3. metal foil test module: target was to demonstrate superpermeability simple test performed: thin vanadium foil welded to stainless steel pipe and exposed to H plasma welding, leak-testing and heating of the metal foil was successful, but the foil broke after a short operation. Talk pf C. Day (KIT) on Non-cryogenic pumps for DEMO. 17

18 Total efficiency η CD γ CD of ITER H&CD DEMO should be a point design w/o experimental flexibility. Optimization of H&CD may follow a different route than for ITER. Substantial current drive may be needed in DEMO if steady state operation is envisaged at realistically achievable bootstrap fraction. LHCD ECCD FWCD NBCD γ A/(Wm 2 ) (indep. T e ) (ITER prediction) 0.1 (ITER prediction ) (ITER prediction) Remark very peripheral absorption potential for optimisation (next slides) small exp. Basis off-axis CD not fully understood H. Zohm et al. Assessment of H&CD System Capabilities for DEMO, EPS 2013 Physics is relatively well understood, but efficiencies can be improved w.r.t. ITER by a targeted optimisation, trading off flexibility 18

19 EC Achievements Electron Cyclotron Top-injection for improved CD efficiency (E. Poli et. al. IPP Garching) α=poloidal angle, β=toroidal angle Further studies have shown efficiencies at 250GHz with γcd>0.34. All cases are depending on plasma density and temperature profiles. 19

20 Radiation mapping shows in-vessel absorbed dose rates range: 2 kgy/h after 1 week following shutdown kgy/h after 1 year foll. shutdown The radiation life of dexterous manipulators (i.e. as used in ITER, JET) does not allow them to be used for blanket or divertor replacement The vertical maintenance system concept is the most viable option with 5 blankets segments (3 outboard / 2 inboard) accessed per vertical upper port An In-Vessel Mover (inserted via the lower port following divertor removal) is required to release the bottom of the blanket from the vessel and manipulate the bottom of the blanket during replacement A CAD model and VR simulations of blanket and divertor maintenance have been developed to demonstrate feasibility Remote Maintenance Upper Cask with telescopic crane system for blanket segment handling (casing removed for clarity) Bioshield Upper port TF Coil Lower port Lower Cask with telescopic lift system for divertor cassette handling CAD model used for WP2012 DEMO Remote Maintenance Studies 20

21 Current Situation (May 2013) EFDA shall expire on 31st December 2013 The Contract of Associations (CoA) will be discontinued at this time EFDA will be replaced by a European Fusion Consortium made up of all EU Associations with a lead coordinating association (likely IPP), to provide financial, legal and administrative functions Governance to be provide by the General Assembly of the Consortium A Central Programme Unit is proposed within the Consortium to manage the technical and programmatic aspects of implementing the Roadmap The first phase of the roadmap (CDA phase) will be implemented by a number of projects to be executed in the period 2014 to To support the transition, EFDA has been asked to provide: Work Plan Document (2014 to 2018) > to be approved by Oct Work Programme Document > to be approved by Oct 13 Enabling: Call for participation (CfP) in Work Plan Projects > selection of Project Leaders in early Dec 13 21

22 Proposed PPPT Organisation European Fusion Consortium Power Plant Physics & Technology PM - Program Management Planning & Resource Management Project Control R&D Management Quality Assurance Information Management PPPT Projects Magnets Conventional Design Advanced Design Containment Structures (e.g. VV, Cryostat) Breeding Blanket Conventional Designs Advanced Designs DPI DEMO Physics Integration Systems Code Plasma Scenarios Heat Exhaust Physics Plasma Control Divertor Conventional Design Advanced Designs H&CD Systems Conventional Design Advanced Design Diagnostics & Control Systems SE - Systems Engineering / Design Integration Requirements Management & Analysis Design Configuration Management (CAD) Design Tools / Methods, Codes & Drawing Management Plant System Modelling & Simulation Specialist Coordination (i.e. Safety, RAMI, Materials, Physics, Cost, Manufacturing etc) Technical Review & Coordination Remote Maintenance System Tritium, Fueling and Vacuum Systems Materials Heat Transfer, Balance of Plant Plant and Site Systems Safety Analysis and R&D Socio-Economic Research Central Project Unit Distributed Project Team

23 Summary ITER is the key facility in the roadmap. DEMO design will benefit largely from the experience that is being gained with the ITER construction. A solution for the heat exhaust in DEMO is needed. A dedicated FNS is needed for material development. The R&D to ensure tritium self-sufficiency should be strengthened. Parallel design concept in addition to ITER TBMs. A system engineering approach is needed. Industry must be involved early in the DEMO definition and design. A detailed implementation plan is being prepared. Availability of sufficient resources and an adequate implementing organization are prerequisite to success. Europe should seek all the opportunities for international collaborations. 23