Technical Challenges on the Path to DEMO and the Strategy of EFDA on the Power Plant Physics and Technology

Transcription

1 Technical Challenges on the Path to DEMO and the Strategy of EFDA on the Power Plant Physics and Technology Gianfranco Federici Head of PPPT Department HTS Fusion Conductor Workshop, KIT Karlsruhe

2 Outline History of MFE Power Plant Studies Main technical challenges on the path to DEMO New EFDA organization/ Power Plant Physics and technology (PPP&T) PPP&T activities under EFDA Organization of initial activities Work Programme 2011 and Outlook 2

3 History of MFE Power Plant Studies International PPCS: mid 2001 mid 2004 HELIAS> SEAFP: US only 3

4 Basic Baselining Considerations EU PPCS gave a range of options 4

5 Physics Long pulse/ Steady state/ High Beta High density operation Power exhaust / divertor plasma performance Abnormal events avoidance/ mitigation Plasma diagnostics and control Reactor System Codes Physics more?? Technical Challenges on the path to DEMO with potentially large gaps beyond ITER Areas with potentially large gaps to DEMO Technology PFCs performance & reliability / maintainability H&CD Systems Efficiency and Reliability Qualification of resilient structural materials Blanket technology including coolants RH schemes for high machine availability T self sufficiency, fuel cycle (T & vacuum technologies) S/ and HT/S magnets Safety and licensing Tokamak Reactor Machine availability and Findings Report of the Ad hoc Group on DEMO Activities, Revision June efficiency P. Batistoni, S. Clement Lorenzo, K. Kurzydlowski, D. Maisonnier, G. Marbach, M. Noe, J. Paméla, D. Stork, J. Sanchez, M.Q. Tran (Chair) and H. Zohm, 5

6 Long pulse/ Steady state/ High Beta Technical Challenges on the Path to DEMO Physics/1 In tokamaks, ss operation means that plasma current is driven by (1) intrinsic ( bootstrap ) current & (2) externally driven current by H&CD systems. (1)implies operation at high N, challenging the stability limits, (2)implies a large burden on the economic efficiency since the external H&CD system has to be powered by the electrical energy generated by the plant itself. A credible tokamak scenario not yet demonstrated. this requires a) that the full current is driven noninductively, but also b) that the radial profiles of current density and pressure are self consistently aligned. High density operation Courtesy of H. Zohm High density in the core required to achieve sufficient fusion power (P fus ~ n 2 ). High divertor plasma density required to dissipate thermal loads in divertor. In tokamaks, the achievable density limited, by the empirical Greenwald limit (n G ~ Ip/ a 2 ), Unfavourable dependence of current drive efficiencies on density and the limitations of wave propagation in high density plasmas. Needed Development: Demonstrate tokamak operation at high density, possibly above n G. 6

7 Power exhaust and divertor performance Power Exhaust may ultimately determine reactor size Needed developments urgent for system code runs Technical Challenges on the Path to DEMO Physics/3 In DEMO, power is much larger Direct core losses: Bremsstrahlung, Synchrotron radiation Edge radiation will need to be pushed to limit to keep the core confinement to remain above H mode threshold However, this is NOT equivalent to high radiation fraction in present experiments, where power to target 0 Integrated modelling of divertor power exhaust such as that performed as part of the ITER Design have proven extremely valuable in guiding the design and leveraging experiments in machines and R&D. Pin down the problem of the heat flux at the divertor strike points. Confirm the degree of realism of the assumption of 10 MW/m 2. 7

8 Snowflake divertor has been studied and achieved in TCV and more recently NSTX Advanced Divertors magnetic shaping Issue in vessel coil shielding created by using only 2 3 existing magnetic coils. the peak heat load is considerably reduced, because it flares the SOL at the divertor surface. Limited impact on the high performance and confinement of the high temperature core plasma, and even reduced the impurity contamination level of the main plasma. V. Soukhanovskii (LLNL) Super X is one concept where magnetic geometry could handle extremely high divertor loads SOL taken to large major radius natural flux expansion; SOL passes through low PF region connection length is increased further spread of power volume to enable power radiation before striking target. 8

9 General considerations about DEMO TF magnets Design criteria and technology of SC magnets have profound implications on performance and cost of a tokamak see ITER. Early system studies to identify clear sensitivity to important machine parameters, e.g., B, R, A Courtesy of P. Barabaschi After discussion with JL Duchateau, PL. Bruzzone and W. Fietz Design improvements recommended for use of LTSC (Nb 3 Sn) for DEMO Review concepts for Nb 3 Sn conductor construction that do not suffer from strand motion and fatigue due to varying magnetic fields and propose a suitable development programme. Assess balance of increased cost of better neutron shielding of the superconductor vs cost/problems of tolerating the nuclear heating and hence reduced temperature margin for a given j and B. Determine realistics steps including time and resources to industrialize HTS materials 9

10 Concerns for Divertor Lifetime Erosion ELMs Disruptions Power Exhaust Present water technology 25MWm 2 possible, 10MWm 2 reliable Add neutrons < 10MWm 2 Present He technology, very good progress but <<10 MWm 2. Courtesy: L. Boccaccini, KIT Power Exhaust may ultimately determine reactor size 10

11 Blanket designs: Blanket concept will determine BoP Courtesy: L. Boccaccini, KIT 11

12 Remote Handling for DEMO far exceeds ITER requirements Much heavier components (blanket segments ~ tonnes). High radiation environment in the machine (much higher than ITER radiation hard detection systems) Much stricter contamination control Higher reliability/availability lower turn round time J. Bonnemason et al 12

13 H&CD Efficiency: present status Three steps 1. conversion of electric power into power launched in plasma: conversion efficiency conv 2. coupling of launched power to plasma: coupling efficiency coupl 3. current drive efficiency (current driven per power unit coupled to plasma CD ) P enet = P egross P ehcd P ebop = P egross I CD / conv coupl CD P ebop Courtesy: A. Becoulet, CEA J. Pamela et al 13

14 Advantages FM steels well established (in fission). Good balance of properties. Well know fabrication technology. Various options for joining (TBM FW/box: techniques demonstrated). Significant data base to start immediate CDA (conceptual design). Issues & Limitations Limited to ~300/ C. Embrittlement at low T and high dose. Concern: Effects of (additional) transmutational helium (#). Evidence that material is suitable for a first DEMO BB (~30 40dpa). To be confirmed. Further potential to be evaluated. (#) Generic issue testing materials in fission reactors of BB ~40 times more transmutational He generated than in MTR Materials: EUROFER (-type) Steel(s) Courtesy of S. Gonzales, EFDA DBTT ( C) T irr = C Results from fission neutrons KLST DBTT (FZK, NRG) ISO-V DBTT (SCK) Dose (dpa) Additional increase above ~500/700 to 1000 appm He

15 DEMO Structural steels: high temperature strength ODS steels Ultimate Tensile Strength R m [MPa] R m FeCr 13, t m :0h, as hipped 13Cr-1W-0.3Ti-0.3Y 2 O 3 : t m :21h Ar, (HIP: 1150 C, 1000bar, 2h), (30 ;1050 C) 13Cr-1W-0.3Ti-0.3Y 2 O 3 : t m :21h H, (HIP: 1150 C, 1000bar, 2h), (30 ;1050 C) 13Cr-1W-0.3Ti-0.3Y 2 O 3 : t m :21h H, (HIP: 1050 C, 1000bar, 2h), (30 ;1050 C) 13Cr-1W-0.3Ti-0.3Y 2 O 3 : t m :21h H, (HIP: 950 C, 1000bar, 2h), (30 ;1050 C) Eurofer 97 ODS-Eurofer (0.3wt Y 2 O 3 ) FZK EURATOM Coutesy R Lindau FzK ASSOCIATION ODS RAFM steels Eurofer Ref [14]: Z Oksiuta and N Baluc DBTT ~20ºC Test Temperature [ C] Conventional ferritic martensitic steels (EUROFER97) are limited to 550ºC : lose mechanical strength and suffer from thermal creep ODS steels have higher strength at high T and better resistance to thermal creep enlarging the T window ~100K at the upper end; but are brittle at room temperature (even un irradiated) (good mechanical) properties sensitively depend on details of manufacturing and treatments > Development needed. 15

16 Three Programmatic objectives Obj. 1 Secure ITER operation Obj. 2 Prepare Generation ITER Obj. 3 Lay Foundations for Fusion Power Plants Report of the CCE FU Working Group on JET and the Accompanying Program Power Plant Physics and Technology established in EFDA Goals of PPPT to begin a coordinated effort in EU (building on efforts done in the past) to quantify key physics/ technology prerequisites for DEMO. to define a set of technical characteristics for DEMO and, subsequently, to carry out the design work necessary to establish its conceptual design; to define future R&D needs and to carry out specific validating R&D work supportive of the conceptual design activities. 16

17 Objectives/ How to work together An integrated design effort with distributed resources we need to work together A B C 3PT F4E BA 3PT WP 2011 contains some element that are of common interest to the DEMO Design Activities conducted by F4E with Japan in the BA and the PPPT activities could contribute largely to fulfil this commitment JAEA 17

18 Proposed Evolution DEMO Baseline Concepts In the initial phase of work PPPT will analyse at least 2 concepts of DEMO A conservative baseline design ie., a DEMO concept deliverable in the short to medium term, based on the expected performance of ITER with reasonable improvements in science and technology; i.e., a large, modest power density, long-pulse inductively supported plasma in a conventional plasma scenario. An optimistic design, i.e, a DEMO concept based around more advanced assumptions which are at the upper limit of what may be achieved during the ITER phase of fusion development, i.e., an advanced higher power density high CD steady state plasma scenario main purposes: give a clear focus to the early DEMO efforts. define, in a bottom up approach the most urgent R&D needs. 18

19 PPPT Work Programme 2011 Activities to be conducted in 2011 are grouped as follows: 1. Fusion System Design Code: Application and Development Evaluation and comparison of fusion system codes. Code application: benchmark with Japan and initial scoping analysis. Proposal for improvement. 2. Assessment of the Power Exhaust/Extraction in a Reactor Divertor edge modelling studies. Assessment of gaps and needs on new DEMO relevant facilities. Optimisation of conventional divertor technologies and analysis of limits. Studies of novel divertor concepts. 3. Engineering/ Design Preparatory Assessments Assessment of DEMO divertor and breeding blanket concepts including the coolants for invessel components. Assessment of candidate remote maintenance schemes and solutions. Analysis of technology/ engineering issues of steady state vs. pulsed tokamaks. Evaluations of status and prospects of high temperature superconducting magnets. Assessment study of H&CD and fuelling and pumping systems. Assessment of engineering materials database for conceptual design activities. 19

20 PPPT 2011 and Outlook PPP&T Department established. PPP&T WP 2011 endorsed by STAC and approved by EFDA SC. Principle of governance of the Implementing Agreement approved. 26 EU Associations expressed interest to participate to the PPPT. Calls of PPPT: SYS: Fusion System Codes; PEX: Divertor Power Exhaust Studies, Materials, Dust & Tritium and SERF have been launched. Call of PPPT on design assessment studies (DAS), will be launched end of May. Start discuss preparation of WP2012. EFDA STAC of 19 September, Approval EFDA SC November. 20

21 Thanks for your attention 21