1 G. Federici, DPG, Berlin 12 - March 2012 Fusion Energy Achievements and Challenges Fusion Energy Achievements and Challenges Gianfranco Federici Head

1G. Federici, DPG, Berlin 12 - March 2012

Fusion Energy Fusion Energy Achievements and ChallengesAchievements and Challenges

Gianfranco FedericiHead of EFDA PPPT Department

March 26, 2012Deutsche Physikalische Gesellschaft e.V.

Berlin, Germany


OutlineOutline

• Incentives for developing fusion

• The next frontier ITER and the role of other machines

• Roadmap to fusion energy

• DEMO Main Technical Challenges – Power exhaust

– Power extraction and tritium breeding (blankets)

– Radiation resistant structural materials

• Summary


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• Summary


Pros• Abundant fuel (D + Li)• No greenhouse gases • Safe – no chain reaction, ~1 sec worth of

fuel in device at any one time• Minimal “afterheat”, no nuclear meltdown• Residual radioactivity small; products

immobile and short-lived• Minimal proliferation risks• No seasonal, diurnal or regional variation

Cons• We don’t know

how to do it yet (it is a really hard problem)

• Capital costs will be high, unit size large

Fusion energy can also be used to H, and for desalination

Incentives for Developing Incentives for Developing FusionFusion


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• Summary


The Next FrontierThe Next Frontier ITERITER

• ITER, to be built and operated as an international project, will push research efforts into this new regime of burning plasma science

Understanding the behavior of burning plasmas is a necessary step towards the demonstration of fusion as a source of energy. Q=10

“Burning” plasma = dominantly self-heated by fusion products (e.g., alpha particles) from thermonuclear reactions in the plasma.

D + T → n + α + 17.58 MeV


20112011September 2011September 2011

December 2011December 2011

Polidal Field Coil Building (257m x 49m x

18m h)


February 2012February 2012

Complete deployment of sismic pads

More fotos and video on: //www.iter.org/org/team/odg/comm


JET and ASDEX-UpJET and ASDEX-UpMitigation of ITER operation risksMitigation of ITER operation risks

Top Risks

•Disruption mitigation has limited effectiveness

•H-mode power threshold at high end of uncertainty range

•ELM mitigation schemes has limited effectiveness

•Vertical stability control limited by excessive noise (or failure of in-vessel coils)

•Lack of reliable high power heating during non-active phase of programme

•Acceptable “divertor” performance with W.

•High levels of T retention require more frequent T removal procedures than foreseen

•Incompatibility of core plasma requirements for Q=10 with radiative divertor operation

•Inability to achieve densities near Greenwald value required for Q=10

Source: Lorne Horton (EFDA-Culham)


10

JT-60SAJT-60SAExplore advanced modes of operationExplore advanced modes of operation

Present J T- 60U NCT device

JT-60SAJT-60U

~4m~2.5m

EAST (A=4.25,1 MA)1.7m

1.1m

SST-1 (A=5.5, 0.22 MA)

3.0m

JT-60SA(A≥2.5,Ip=5.5 MA)

6.2m

ITER(A=3.1,15 MA)

KSTAR (A=3.6, 2 MA)1.8m

• JT60-U: Copper Coils (1600 T), Ip=4MA, Vp=80m3

• JT60-SA: SC Coils (400 T), Ip=5.5MA, Vp=135m3

Source: P. Barabaschi, F4E


Wendelstein 7-XWendelstein 7-X

During first campaign:•8MW ECRH and 7 MW NBI•Diagnostics set probably sufficient to conduct the initial program

•Test divertor unit to study operation limits and divertor physics

W7-X: Assembly according to plan

W7-X: Major Milestones

No delays expected: finished mid 2014

Completion:•Steady state divertor•Increase in heating power, ICRH•Diagnostic completion


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• Summary


Roadmap(s) to Fusion Roadmap(s) to Fusion EnergyEnergy

• Different countries face different energy needs and these drive different strategies for fusion development.

• The greater the perceived urgency for fusion energy the greater the willingness to take larger steps and larger risks.

• All ITER parties have a target to demonstrate fusion-driven electricity production by ~2050.

• The roadmaps of China and India, that foresee the largest increase in energy demand in the next decades, are the most ambitious, in terms of both goals and timescale for next steps. – China is considering the construction of a further DT machine. Engineering Design Phase is expected in 2014 and

first plasma in 2025.

In Eu the Roadmap is being revisited. The plan is to launch a vigorous coordinated effort to prepare for a fusion Demonstration Reactor to be built by the beginning of 2030 (EFDA PPPT Dept. is the very first step in this direction).


Define Next-Step (after Define Next-Step (after ITER)ITER)

• Today, there is still a divergence of opinions on how to bridge the gap between ITER and the first FPP.

• EU (and JA): DEMO and IFMIF;

• US: CTF or a Pilot Plant and no dedicated materials test facility. R. Goldston (IAEA TM, June

2011)

• However, there are some common outstanding issues common to any next major facility after ITER, whether a CTF, a Pilot Plant, a DEMO, or else:– Power exhaust handling (divertor) – Reference plasma scenario CD requirements, – Coolant for in-vessel components breeding blanket concept– Maintenance scheme plant architecture– Structural and PFC materials

• Only some of these issues can be solved in ITER.


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• Summary


DEMO Technical ChallengesDEMO Technical Challengeswith potentially large gaps beyond ITERwith potentially large gaps beyond ITER

ITER objectives and design are well established; - not yet the case for DEMO.

TECHNOLOGY

– PFC and Blanket technology including T self-sufficiency

– H&CD Systems – Efficiency and Reliability

– Reliability of Core Components & RH for high machine availability

– Qualification of resilient structural materials

– Safety and licensing

PHYSICS

– Operating scenario: Long pulse/ Steady-state/ High-Beta

– High density operation

– Power exhaust and divertor R&D strategy

– Abnormal events avoidance/ mitigation

– Plasma diagnostics and control


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• Summary


• Power density fusion reactors much smaller than fission reactors• But peak-to-average heat flux at coolant surfaces much higher

PWR

BWR

HTGR

LMFBR

Fusion 3

MW/m2

Equivalent core diameter (m) 3.6 4.6 8.4 2.1 30Core length (m) 3.8 3.8 6.3 0.9 15Aver. core power density (MW/m3)

96 56 9 240 1.2

Peak-to average heat flux at coolant interface

2.8 2.6 12.8

1.43 50

Source table:Abdou (UCLA)

Power Exhaust and DivertorsPower Exhaust and DivertorsVery High Heat FluxesVery High Heat Fluxes


Divertor TechhnologyDivertor Techhnology

• 2000 cycles at 15 MW/m2 on W.

• More recently 300 cycles at 20 MW/m2 (ITER requirements) + 500 pulses at 0.5 MJ/m2 to simulate ELM-like loads

– Longitudinal macro-cracks appeared in all monoblocks.

– some melting of W at monoblock edges

• But no degradation of their power handling capability

Water-coolingITER Technology, W and Cu-Cr-Zr

• 20 MW/m2 possible 15 MW/m2 reliable add neutrons < 10 MW/m2

Source: Riccardi (F4E), Visca (ENEA)

• >1000 cycles at 10 MW/m2. Recently cycles at 12 MW/m2 Thimble is still the most critical component.

Influence of irradiation is unknown

• Design integration and reliability still to be addressed

He-coolingITER Technology, W and Cu-Cr-Zr• 12 MW/m2 possible 10 MW/m2

reliable add neutrons ~5 MW/m2

-Helium-cooled modular divertor (HEMJ)

Norajitra, KIT

Source: Norajitra (KIT)


Advanced Divertors Advanced Divertors magnetic shapingmagnetic shaping

• created by using only 2-3 existing magnetic coils.

• the peak heat load is reduced, because it flares the SOL at the divertor surface.

• Limited impact expected on the high performance and confinement.

has been studied and achieved in TCV and more recently NSTX

“Snowflake divertor”

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.

Issue – in-vessel coil shielding


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• Summary


Tritium Supply and Tritium Supply and BreedingBreeding

Large consumption of tritium during fusion– 55.8 kg/yr per 1000 MW of fusion power

Production and cost– CANDU reactors: 27 kg over 40 years, $30M/kg currently– Other fission reactors: 2-3 kg/yr $84-130M/kg

• Tritium breeding for self-sufficiency– World supply of tritium is

sufficient for 20 years of ITER operation (will need ~17.5 kg, leaving ~5 kg)

– Verified tritium breeding technology, to be tested on ITER, will be required for DEMO and reactors.

We focus on the D-T cycle (easiest):– D + T → n + α + 17.58 MeV

Tritium does not exist in nature! o Decay half-life is 12.3 yearso T must be generated inside the

blanket

The only possibility to breed tritium is through neutron interactions with Li that must be used in some forms


Source L. Boccaccini (KIT)

Helium at 300-500°C @ 8MPa

Li cer.

Be.

CPS: Coolant Purification Sys.TES: Tritium Extraction System

HCPB

HCLL

Principles of HCPB blanket concept: breeding and T Principles of HCPB blanket concept: breeding and T extractionextraction(shown as example)(shown as example)

There are other alternative Blanket Design Concepts

During ITER Research Programme, TBMs will be installed in ITER to investigate breeding.ITER has three ports for blanket testing and 2 TBMs can be installed in each port.

Breeding BlanketsBreeding Blankets


Internal Components Internal Components Reliability/ MaintainabilityReliability/ Maintainability

Large Port ConceptVertical Port

Concept MMS Concept

• Reliability represents a challenge to fusion, particularly for the core components.

• RH strongly impacts machine availability (MTTR, MTBF) and affects in depth the design of many components/interfaces. It is needed from the design outset.

• Proposed design solutions must be fully remotely maintainable.

• Significant amount of time consuming demonstration and R&D often requiring design iteration and changes before we start to build.


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• Summary


Fusion Structural MaterialsFusion Structural MaterialsFusion reactors need high-temperature, radiation resistant

materials

In DEMO demanding operational requirements that are beyond today’s experience (including ITER and fission reactors), e.g., elevated operating temp., long periods of operation, higher irradiation damage and He accumulation, high reliability and availability, etc.

In Fe for 1 MW/m2 and 1 FPY–10 dpa–100 appm He –450 appm H–He/dpa ~ 10 appm/dpa RAFM: currently EUROFER 9%Cr [1W

0.14Ta 0.2V] steels (reference for DEMO)


The IFMIF FacilityThe IFMIF Facilitywill allow qualifying materials under fusion will allow qualifying materials under fusion

spectrumspectrum

EVEDA Phase in progress (as part of the BA with JapanReduced-cost/ reduced performance options are being explored

In DEMO for 1 MW/m2 and 1 FPY–10 dpa (in Fe)–100 appm He –450 appm H–He/dpa ~ 10 appm/dpa Lack-of irradiation facilities with

adequate n-spectrum (14 MeV + He)

Beam Spot(20x5cm2)

High Flux

Low FluxMedium Flux

LiquidLi Jet

DeuteronBeam

• Deuteron beams:– 2 x 125 mA– Ed = 40 MeV

• Neutron production:

1.1 1017 s -1

• Test volumes:– high flux: 0.5 L > 20 dpa/fpy– medium flux: 6 L > 1 dpa/fpy,– low flux: ~8 L 0.1-1 dpa/fpy

• Accelerator driven Li(d,n) source

• 2 x 125mA 40MeV deuteron beams

• Liquid Li target (~15m/s) subject to 10MW 1GW/m2

• Full range of PIE facilities

• Designed to reach ~150dpa within a few years of full power operation

Source: U. Fischer


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• Summary


SummarySummary

• Fusion has a tremendous potential

• ITER must be a success and it will answer open physics questions related to burning plasmas

• There are still several challenges to be overcome for DEMO, especially for the core components (divertor, blanket) and materials.

• Demonstration of fusion electricity by 2050: challenging but possible

• In Europe the roadmap for the exploitation of fusion is being revisited. Expected a tighter coordinated effort with clearer focus and more technology orientation

• W7-X will demonstrate the quality expected from stellarator optimisation

• If we succeed, with fusion, we handover to future generations a clean, safe, sustainable power source.


Thanks for your attention

• Born 28.5.1960, married with two children (16 and 11)

• Degree in Nucleal Engineering, Polytechnic of Milan 1985

• Ph.D. UCLA 1989 (Fusion Eng. and Applied Plasma Physics)

• Post-Doc Fellowship EU Commission, Fusion 1990-92

• NET Team, 1992-93

• ITER Team, 1994-2006: Divertor and plasma interfaces

• EFDA Garching, 2006-2007: Field Coordinator Vessel/ In-Vessel

• F4E Barcelona, 2008 –2010: Senior Advisor to Chief Engineer

• F4E Garching, 2011-today: Head of EFDA Power Plant Physics and Technology Dept.

Who I am!Who I am!

Documents

1 G. Federici, DPG, Berlin 12 - March 2012 Fusion Energy Achievements and Challenges Fusion Energy Achievements and Challenges Gianfranco Federici Head