International Meeting “MFE Roadmapping in the ITER Era” PPPL, 7 th -10 th September 2011 (this work was supported by UK EPSRC and Euratom) CCFE is the

International Meeting “MFE Roadmapping in the ITER Era”PPPL, 7th-10th September 2011

(this work was supported by UK EPSRC and Euratom)CCFE is the fusion research arm of the United Kingdom Atomic Energy Authority

Technical Challenges on the path to DEMO

Derek StorkEuratom-CCFE Fusion Association,

Culham Science Centre,Abingdon, OX14 3DB, UK



This is a personal opinion…………

… intended as a contribution to a debate … How can we gain back time on Fusion’s Development Roadmap?

Technical Challenges on the path to DEMO - D Stork invited talk PPPL MFE Roadmapping meeting September 2011

Outline• Comments on Fusion Roadmap ‘standard model’

– DEMO mission and critical path• Dividing resources for a “DEMO programme” into:

– Baseline– Optimisation programme– Strategic Risk Reduction programme

• Categorisation of DEMO Technical Challenges– Baseline, Optimisation or Strategic Risk Reduction?– Characteristic of challenge– Motivate definition of the ‘DEMO Stage’– Use in refining the Accompanying Programme to ITER (and perhaps

ITER “Phase II”?)• Baseline DEMO Technical Challenges & programme elements• ‘DEMO Optimisation’ Technical Challenges & programme

elements• ‘DEMO Strategic Risk Reduction’ programme elements


Fusion Roadmap(still a Fast Track ?)

ITER

IFMIF (Materials testing)

DEMO Power Plants

JET + Other m/c

Concept improvements

Technology Programme

Satellite Tokamak

?

?

DEMO

Need to define the ‘DEMO stage’ mission

and facilities


Roadmap: DEMO’s mission

• The main DEMO reactor is the ‘last research machine’ before the First of a Kind Fusion Power Plant (FPP).

• In Europe DEMO’s mission has been quoted as:– completion of nuclear lifetime testing of in-vessel components;– demonstration of tritium self-sufficiency – integration of full breeding

blankets into full tritium fuel-cycle plant;– demonstration of efficient and low-turnaround remote maintenance

and replacement of the key tokamak systems (divertor and blanket); – demonstration of fusion’s environmental (low activation materials)

and safety (safe operation; acceptable licensing/safety case) credits;– supply of nett electricity to the grid;– demonstration of high level of reliability and availability;– supply of economically competitive electricity.

supply of nett electricity (several 100 MWs) at intermittent times

Demonstration of high levels of reliability and availability at end of programme allow economic assessment of a fusion power plant

Urgent to have early DEMO implementation – ‘existence proof’ important for those outside Fusion! – de-emphasise the economics.

EFDA DEMO

Ad Hoc Group 2010

Zohm, FST 2010


Roadmap: DEMO’s critical items (I)

DEMO concept based on PPCS Model C) – KIT, CEA (2007)

System and Detailed design and validation of

Load Assembly in-vessel items requires

full nuclear qualification of structural materials.

finalised/qualified Blanket concept

finalised/qualified Divertor concept

Moreover System and Detailed design of Balance-of-Plant and Remote Maintenance requires finalised/qualified Blanket and Divertor concepts

Clearly these items are the ‘critical three’ for DEMO

Load Assembly remains the core of any DEMO machine


Source – 2008 Ann Report of the Association FzK/Euratom – EFDA/06-1454 study E Magnini et al

Roadmap:DEMO’s critical items (II) For the Blanket, the Divertor and the nuclear degradation of their structural and PF materials Engineers must have validated data and engineering rules for final design against anticipated Load conditions and degradation of properties.

Everything is inter-dependent in a complex way. Detailed examples abound

…Use active cooling?…or..… use ODS?…or…

Eurofer embrittlement

Creep issues for Eurofer


Roadmap: DEMO critical paths (I)

• No timetabled path to producing at divertor with ~50 MW.m-2 handling .

• Other key items for DEMO detailed design are on critical paths with relatively fixed long lead duration:– nuclear qualification of

structural materials to ~ 4-6 MW.a.m-2 14 MeV neutron flux;

– output of post-experimentation testing of TBMs from ITER DT operation.

Requires IFMIF tests

ITER schedule gives this as ~2029


Roadmap: DEMO critical paths (II) IFMIF

No accepted timetable for IFMIF, but we take ‘Critical’ (middle of the road) timescale from EU Road-mapping presentation §

At 20 dpa/fpy 40 dpa is reached in 2030

§ Moslang, Baluc, Diegele, Fischer et al., CCE-Fu Workshop on EU Fusion Roadmap – Garching April 2011

Conclusion: crucial data from Blankets and Nuclear Materials is not available until ~ 2029-30: marks the point at which ‘system design’ can start


Dividing the DEMO stage programme areas

• ‘DEMO Baseline programme’ – aim to realise the DEMO machine at the earliest possible time;– Handle the ‘DEMO-phase’ major risks.

• ‘DEMO optimisation programme’ – Concentrates on developing technologies/techniques which

• make cost-of-electricity more attractive;• Improve reliability of plant• can be ‘feasibly’ incorporated relatively late into DEMO-phase.

• ‘DEMO strategic risk management programme’– handles the ‘long-term’ programme technical risks– develops technologies/systems which will ‘future proof’ a Fusion Economy

All potential programme expenditure (Technology Facilities, Satellite Machines, Development Lines) should fit into one of these programmes.

Use this categorisation to determine where the contributions of eg. ITER can best be utilised, and avoid duplication of effort.

Reso

urces


Baselining DEMO choices…carry out preparation/evaluation phase with the basic philosophies of…

• Using ‘Systems engineering’ and ‘Systems code’ approach• Not introducing further competing critical paths –

Baseline should be conservative.

• Aiming for a reliable and available product – maximum use of Industry at this stage in Baseline programme R&D.

• Maximising use of synergy/common cause with other technology development fields – Generation IV fission materials, High Temperature superconductors etc.-- both for inclusion and exclusion from the baseline!

….In an EU context – ‘Preparation Phase’ would run to mid-2010s to allow evaluation of the Baseline options, establish Core Design Team(s) and complete BA activities.


‘Baseline’ will be an evolving entitycan’t fix everything on ‘Day 1’ but all have to be firm in time

System Reviews

Re-baseline(n – off)

Final baseline decision

Overall SDR

Baseline Programme

Optimisation Programme


Baseline programme should concentrate on: showstoppers to operation – where ‘existence

proof is needed’ the novel elements of the DEMO stage – where

full-scale integrated tests will occur for the first time

the core of the Load Assembly key drivers for Machine Integrity and

Availability decisive tests to enable focussing selection from

competing solutions to core needs.

Baseline Programme Elements: characteristics

eg, Divertor

eg, Blanket and Ancillary systems

eg, Structural Materials, Remote Handling

eg, H&CD systems

For timely delivery, and technical and politic

al cohesion –

Baseline MUST take maximum advantage of ITER and

existing faciliti

es – only bring in new faciliti

es where the need

is absolute.


Baseline:Structural Materials (I)• DEMO mission + Systems approach with ‘Conservative Baseline’ filter should

produce an operating concept to give Baseline structural material. Conservative approach favours existing Reduced Activation Ferritic-Martensitic (RAFM) steels eg. Eurofer.

– Improving engineering database for RAFM steels exists • Engineering parameters• Radiation effects• Joining techniques

but:– RAFM steels have known narrow temperature operating window

• must be ~ > 350ºC to avoid radiation embrittlement• Must be ~ <550ºC to avoid loss of strength/creep rupture issues

– and He-induced swelling at high dpa values

• If, more developmental HT steels eg. Oxide Dispersion Strengthened (ODS) alloys are to be in the Baseline, they must pass clear, simple criteria for basic properties (eg. ductility at room temperature) by an early date.

• Baseline risk-mitigation is needed for known Eurofer shortcomings, eg:– characterise as far as possible ahead of IFMIF tests;– minimise by design choices;– seek common solutions from Fission developments


Baseline:Structural Materials (II)

Eurofer risk-mitigation by design/system choices

0

5

10

15

0 5 10 15 20 25

Blanket lifetime fluence (MW.year/m2)

Co

st

of

ele

ctr

icit

y (

ce

nts

/kW

h)

Baseline slightly higher Cost-of Electricity (CoE) target for the first DEMO – system studies (PROCESS) show decreasing gains above ~ 60 dpa. -- Availability (dependent on Blanket Lifetime) can be sustained by increasing machine size.

Ward & Dudarev : IAEA FEC 2008

70

75

80

85

7 8 9 10 11

Major Radius (m)

Pla

nt

Ava

ilab

ility

(%

)

Design for (examples!):

-- high temperature operation (no change in DBTT)or-- high temperature annealing cycles (ductility restored)

Gaganidze: J Nucl Matls & IAEA FEC 2008


Baseline:Structural Materials (III) Risk-mitigation by characterisation of Eurofer ahead of IFMIF

Transmutation in pure Fe – realistic reactor FW spectrum simulated

– reactor PPCS Model B simulation

Use ‘isotopic tailoring’ to produce He-in lattice

• Simulated by (example!) 10 B- doped RAFM steel He significantly increases brittleness at ~ > 400 appm (lattice results) – in conjunction with ~ 17 dpa

Simulate by Ion beam bombardment – He2+ (Caution! – representative of bulk??)• Surface He bombardment shows enhanced embrittlement at ~ 10 – 100 appm/dpa

Lattice helium: He ‘eyes’ – 5.8 103 appm

Surface helium:He2+ beam103 appm/ ~ 100 dpa

[Gilbert & Sublet: Nucl Fusion 2011]

[Materna-Morris et al: IAEA FEC 2008]

[Jitsukawa et al: IAEA FEC? 2008]

Also – ‘early IFMIF’?

– based on energy extrapolation of de-rated EVEDA current results ?

– solid (phase 1?) C target? Aim for ~3-5 dpa/fpy by early 2020s? ~100M€ ?


Baseline:Structural Materials (IV) Eurofer risk-mitigation seeking common solutions with Fission steels

“Future steels development options will be based on evolutionary (ingot metallurgy/classical precipitation) and revolutionary (nanoscale ODS) approaches” – S Zinkle – 23rd SOFT

Eurofer

ODS-Eurofer

ODS Ferritics

• Generation IV Fission programme needs high-temperature steels

• If we aim for ~50 dpa for early DEMO baseline then we overlap requirements of many GenIV concepts (Zinkle, Diegele)

• Industry is much more at home with classical metallurgy

• Common developments to obtain RA versions of Fission ‘3rd and 4th generation’ FM steels? (research melts have >105 hours at ~ 620ºC)


Baseline: EU Blankets (I)

PPCS Model

Structural material

Blanket concept (breeder/coolant)

Divertor (coolant)

A Eurofer WCLL(LiPb/water)

W/Cu/water

AB Eurofer HCLL(LiPb/He)

W/He

B Eurofer HCPB(Li4SiO4/Be/He)

W/He

C Eurofer/ODS DCLL(LiPb/He/LiPb)

W/He

D ODS/SiC SCLL(LiPb/LiPb)

W/LiPb

EU Power Plant Conceptual Studies (PPCS-2005) featured blanket concepts

These Blanket concepts chosen for EU ITER TBM. EU has a pair for ‘baseline’ concepts – one eventually to ‘optimisation’?.

Programmatically, ITER programme pays for this DEMO development


Baseline: EU Blankets (II)

Figure 1 : He Series flow configuration of HCLL blanket module.

PbLi inlet

PbLi outlet

He inlet

He outlet

He Feeding FW + SP

He Collector FW + SP & He feeding CP

PbLi feeding

PbLi collector

FW

CP

SP

HCPB HCLL

Concepts differ in Balance of Plant Tritium extraction details

… but at least the Helium balance of plant will have similar issues

Common advantage potential high thermal efficiency from helium cooling.

Common disadvantages: high helium pumping power; lack of developed helium Balance of Plant (compared to PWR ‘off the shelf’ BoP for water-cooled blanket)

Boccacini - Invited Talk; 26th SOFT, Porto, 2010 + 2 Orals; + 14 Posters


Baseline:EU Blankets (III)• ‘Leading concept’ choice needed eventually to progress DEMO conceptual

design of BoP- facility-based R&D to decide this. • ‘Systems Engineering’ based choice - focus on merits & drawbacks:

– Corrosion of Eurofer piping by the Lithium-lead coolant - lack of data at blanket flows (few mm/sec to avoid MHD) modelling shows acceptable corrosion rates, ~ 20 m.yr-1 for the low flow rates;

– fabrication of the Li- ceramic pebbles without the cracking currently seen, and also without key impurities which delay hands-on recycling (eg. Pt-193 in the Lithium-ortho-silicate pebbles);

– higher radiation damage in the solid breeder - embrittlement of the beryllium pebbles and occurrence of high swelling above 550°C,- compromising high temperature ops;

– tritium release from beryllium pebbles poor until temperatures (~750°C) -- too high for known steels ( inadequate tritium recovery and high in-situ tritium inventory);. Alternative beryllide alloys (eg. Be12Ti) with more acceptable tritium release are in development, [Japan-EU BA ] currently no pebble-based solution.

• liquid breeder, with low radiation damage issues appears advantageous, but needs more highly enriched fuel (90% 6Li cf. 40% for HCPB) to achieve similar tritium breeding ratios (HCLL TBR, = 1.12; HCPB TBR = 1.15).

• HCPB might eventually be regarded as an ‘optimisation programme’ item?


Baseline: EU Blankets (IV):Determine Remote Handling concept

‘Multi Module Segment’

concept

MMS now the favoured EU concept - applies to HCPB and HCLL advantages : - pipe re-welding (He production limit) located in a low neutron flux region - improved manifold design - decreases He pressure drop; - limits EM loads on module attachments in case of disruption


Divertor power loading is a show-stopper issue for a DEMO reactor – ~ 50 MW.m-2 ‘unshielded’

As a solid, only Tungsten can satisfy criteria of melt-damage resistance, high thermal conductivity and low-erosion under plasma bombardment.

Tungsten validation:

JET ILW should further validate Tungsten as Divertor PFC material (2013-4)

JET DT experiments to validate low-tritium retention in tungsten (seen eg. in D+ plasma streams on Pilot PSI) (2015)

ITER ‘Phase II’ should run a full Tungsten divertor – water-cooled ~ 15 MW.m-2

Baseline: Divertor (I)

ITER Divertor monobloc ~ 15 MW.m-2

W

This programme is important, but not sufficient to test a DEMO concept


Baseline: Divertor (II)• Divertor baseline will be actively cooled – even if clever

concepts can give (see later) order of magnitude relief.• Chosen concept needs a full-power density test:

- on a HHF facility but also - in a tokamak environment.

• This mission alone is important enough to justify a ‘DEMO stage satellite’ machine – as part of the DEMO Stage baseline programme.

• DEMO Divertor satellite would require:– Long pulse (cf. R) capability for ‘steady-state’ plasmas;– heating, current drive, fuelling, plasma (& ELM??) control

systems to enable high radiation fraction plasmas at high for testing the Divertor.

– High pressure, high temperature Helium or water coolant loop (or both!)

DEMO Divertor s

atellite is

urgent –

could be a conversion/upgrade of existin

g or planned m

achine


Baseline: Divertor (III) Physics Scenarios

System studies, eg. PROCESS code in Fusion Power Plan studies show:

CoE depends more heavily on operational and engineering parameters than on physics variables:

3.04.04.05.06.0 11

)1

( CoEGWNeth NPA

Thermodynamic efficiency

Physics - high , high density

Availability

Net electrical power

D J Ward, CCFE EFDA-RP-RE-5.0[2004]

Thus technology development is more important than physics development at the DEMO Stage.

However the physics

determines if the scenario is basically feasible/attractive

scenario interacts with the technology as a key selection criterion (via the Divertor and the H&CD)


Choosing Baseline Physics Scenarios :- better avoid too much science fiction

PPCS plasma cross-sections (& ITER for comparison)

PPCS A PPCS B PPCS C PPCS D

Ip (MA) 30.5 28.0 20.1 14.1

Pfus (GW) 5.0 3.6 3.4 2.5

R (m) 9.55 8.6 7.5 6.1

BT @ R (T) 7.0 6.9 6.0 5.6

Energy confinement enhancement

20% above ITER 30% above ITER

20% above ITER

Density Limit 20% above ITER 50% above ITER

N (thermal pressure) 2.8 2.7 3.4 3.7

PCD (MW) 246 270 112 71

Q 20 13.5 30 35

Bootstrap current fraction

0.45 0.43 0.63 0.76

• PPCS invoked -- high density operation – significantly above Greenwald and -- enhanced energy confinement to achieve high and high fusion yield

D Maisonnier et al. NF47 (2007) 1547

EU PPCS - 2005

High gives high Bootstrap current reduces external CD


Choosing Baseline Physics Scenarios (II): – site your machine conservatively.- better find out the scaling laws in DEMO-like plasmas

T,lim =Nx(I/aB)

stable

ITER Q=10; PPCS Mod A/B PPCS Mod C

DEMO Plasmas will be in a novel regime. [D J Ward EPS 2010]

Choice between High density and High temperature both with high radiation fraction - impurity-driven radiation for Divertor power reduction and/or - synchrotron radiation

In such plasmas, for a baseline, we need to know asap: - confinement scaling laws? - high- stability limits? - Confinement of high Fcontent at high th –relevant populations!

Baseline DEMO programme role for present/approved tokamaks? (JT-60SA/JET)

‘Advanced’ DEMOs are not sited conservatively eg. - N


Baseline: Divertor (III):High-temperature Helium-cooled divertor?

-Helium-cooled

modular divertor (HEMJ)

DEMO He-cooled Tungsten-armoured concept (KIT)

Thimbles tested at 12 MW.m-2 ≤ 200 cycles

• Conservative baseline for DEMO would favour Helium-cooled divertor, as foreseen in PPCS DEMO Model B ‘Only’ ~ 65% radiation required for this design

• Tungsten ductile operating window ~ 750°C (set by DBTT) and ~ 1200°C (set by recrystallisation)

W

W-La2O3 W-26%Re

830ºC

1200ºC

System simplification if Divertor and Blanket coolants are the same. For EU urgent to evaluate seriously He-cooled divertor development potential


Baseline H&CD system(s) • Conservative (N< 3) DEMO Studies call for very high

installed Current Drive powers 240 – 270 MW for PPCS Models A/B and AB..

• PPCS assumptions were:– ‘wall plug efficiency’ wp=0.6– ‘current drive efficiency’ as 1.5 MeV NNBICD=0.45 (1020 A.W-1.m-

2)

• On today’s H&CD technology/operational achieved status required H&CD grid demand would be much higher

• For all real systems:– wall plug efficiency is much less than 0.6– non-NNBI systems have lower CD. – but latest ECCD expts?

• In all ‘near-term’ designs H&CD systems dominate the power balance (circulating power, nett power to grid) and contribute plant complexity.

• A serious and focussed development programme is needed.


Baseline H&CD system(s) (II)• DEMO Baseline should have a minimum number of separate systems.

– Each system chosen should have maximal separate task capability (avoid ‘one system per task’ mindset!)

– Systems justified on the basis of elaborate feedback loops should be critically examined (are the required diagnostics forming the feedback loop really likely to be on a reactor?

– Baseline choice would emphasise those systems which couple easily and flexibly to a range of plasma configurations (NNBI, ECRH)

• Initial phase of evaluation should establish for each system:– R&D status – does a source exist?

– does a launching system exist? – will the ITER programme, by the end of Phase I, prove the source and launch concept? – what are the R&D needs for developing and optimising the system for DEMO? Can they be handled on ITER?

– Physics status – does the database for this system show it can generate relevant high-performance plasmas on its own? – does the database for CD efficiency exist? – will it exist post- ITER Phase I? – what are the urgent needs for demonstration(s) on tokamaks other than ITER?


Baseline H&CD (III):assumptions vs reality - ITER System efficiencies

HV and Source power = 58.2 MW

Auxiliary power (Ion Dump, water cooling pump, Cryo)

= 4.4 MW

NB to plasma = 18.8 MW

Accelerated beam

=40.8 MW

TR ~ 18.8/40.8 ~ 0.46

‘source’ ~ 40.8/(58.2+4.4) ~ 0.66

For NBI WP ~ 0.66x0.46 ~ 0.30 – half the PPCS assumed value!

Source – courtesy R S Hemsworth - ITER

Neutralised Beam = 23.2 MW

For ECRH WP ~ 0.52 – but CD ~ 0.15 – 1/3 the assumed PPCS value

Implies DEMO CD powers of ~ 490 MW – 920MW required! Motivator for a Pulsed DEMO baseline? Do we need steady-state ?


Reactor Power Flow with ‘realistic’ H&CD (representative - figures in MW)

Conceptual 4.5GW (fusion); 1.35GW (Electrical) reactor - similar to PPCS Model A – helium cooled – H&CD systems 33% efficient

-- Neutron power multiplication in blanket -- divertor takes all charged particle (conducted ) power

1351

Blanket

Divertor

4140

R

1114-R

Thermal power

5587

Generator

=0.42

2346

Neutron power

power + Aux power

3600

1114

Bulk Plasma

Radiation (R)

Helium coolant pump

333

350

Plasma

Fusion Power 4500

‘Current Drive’ ‘Heating’

94 119

285 359

Recirculating power

994

Ele

ctri

city

Gri

d

4140

Heating and Current Drive

644

214

~ 1GW circulating

power!!


Baseline H&CD (IV): pulsed or steady-state DEMO?

Fixed pulse length – 8 hrs

Motivation for a Pulsed DEMO concept depends on Current Drive scenario and technology prospect:

‘Advanced’ tokamak – hence mainly intrinsic CD? Can external CD overall efficiency be raised?

Major engineering issue for Pulsed Machine would be fatigue life – for pulse length of 8 hours – then loading of > 30000 cycles during 30 year life. For discussion see David Ward’s talk.

Pulsed DEMO would inevitably be bigger – larger solenoid required for flux swing – predict coe increase by ~ 20% -- but some H&CD power alleviates machine size/ flux needs.

D J Ward (CCFE) –PROCESS

–EFDA Study 2008


Baseline: Diagnostics & Control Baseline DEMO diagnostics will be limited by:

limited views, blanket module maintenance simplification;

required radiation hardness for systems (especially windows).

systems engineering-based simplification & conservative approach

need to reach high-level of reliability – favours limited, simple systems

Thus number of active feedback control loops will be limited on a reactor.

Multi- diagnostic actuator loops will fall foul of ‘one H&CD system’ philosophy

transfer of some concepts to reactor (eg. in-vessel coils complex & uncertain

DIIID

JT-60SA would be an appropriate machine on which to test baseline strategies

Baseline should be framed using ‘sparse control’ concepts as developed in other fields.


Baseline for other systems• Pumping

– Major issue ITER batch-regeneration cryopumps don’t scale to DEMO– Re-assessment of the existing technological alternatives and choice of

most promising continuous regeneration technique (eg. snail cryopumps?) then vigorous development programme.

• Magnet Technology– to enable database gathering from ITER - baseline should be Low

Temperature Superconductors (Nb3Sn and NbTi) as in ITER – (HTSC development will be handled by other technology programmes).

• Safety and Licensing issues– ITER experience has to be taken for the baseline regulatory rules.

• Remote Handling - determined by Blanket and Divertor concepts.

• Balance of Plant– Blanket choices drive EU towards Helium circulation systems;– should capitalise on Generation IV fission systems developing these but

to minimise risk Helium BoP development be part of Baseline.


DEMO Optimisation programme

• The Optimisation Programme should run concurrently with the Baseline Programme, taking a fraction of the resources and aiming to realise results by the time the critical path detailed design decision is made (2028-2029).

• Optimisation Programme content depends on Baseline Choices! – pulsed vs steady-state;– Baseline H&CD ‘set’ (or single system) – for the

baseline system, optimisation in Baseline Prog!;– Eurofer alone or +RAFM ODS;– HCPB or HCLL; etc.


Optimisation of DEMO H&CD: NBI challenges

Wall plug efficiency of 50-60% requires simultaneously: Improvement in neutralization efficiency from 58% (gas) to ~95%

Development of photoneutralizer and/or Energy recovery of the dumped ion beam

Improvement in transmission from 75% to 95% Reduction of beam divergence Removal of halo Increased current density

Choice of Materials – potential show-stopping issues In the Drift Duct liner --Copper and CuCrZr eliminated due to irradiation damage GlidCop is a possible replacement but untested in HHF and HV applications Beamline structural material within 4m of First-wall has same issues as FW.

Achieving reliability requires simultaneously: Demonstrating HV holding of >1MV at 10-50A current

Present status: 750keV/221mA & 500keV/20A (few seconds) at JAEA Breakdown follows clump theory but degraded for large grids

Replacement or control of caesiumAlternative proving elusive; understanding role for improved management


EU Roadmap to DEMO NB

ops exp SPIDER

ops exp MITICA

ops exp ITER

ITERprogramme

maintainabilityoperational

requirements reliability

neutralisation

high voltage

injector design

materials

integration

transmission

energy recoverysource

Power/ voltage

efficiency

geometry

define type of DEMO

define NB role

EFDA work

programme

DEMO design process

E Surrey – EFDA PPP&T meeting – Garching – March 2011


Optimisation of DEMO H&CD: ICRF

ITER Physics Basis [NF

39(1999)2512

Experimental current drive efficiency in agreement with theory

Technical status of ICRF sources/transmission at ITER frequencies shows advantages:

commercial (tetrode) sources/generators 60-70% efficient; transmission line relatively standard – developed for ITER ~95% efficient launcher (ITER development) ~ 95% efficient

However:

maximum RF power coupled into H-mode is still ≤12MW (and data is from 1989-1990!)

RF coupling to shaped plasmas with H-mode/ ELMy edge is problematical and not proven by large experimental database.

FW current drive efficiency is low (scales to ~0.15 at Te(0) ~ 20 keV) needs experimental proof

If ICRF is to be retained in the baseline, high power (>20MW) systems need to prove generation and sustainment (CD) of high performance plasmas.


Optimising ICRF: High frequency FW?

[Van Eester & Lerche EFDA CCIC-08 ]

DEMO simulation at 250 MHz

Ntor=90

Ntor=30

Develop HF system for FWCD off-axis?

Promising in theory but: - ITER will not test the system (needs modified launcher for high k⁄⁄ - new source development) - what would then be the purpose of ITER ICRF?


Near-Term EU FWCD Assessment

• Performance estimate.•SWOT Analysis• Sensitivity studies.• Strawman design(s)• R&D programme• Cost, manpower, timescale

Option Recommendation

Option Assessment: Impact on DEMO; RAMI analysis; cost; R&D requirements.

Systems DesignDEMO ICD Requirement

Assume ITER SOL

Assess Coupling

Assess

Confirm , k// Options

R Koch – EFDA PPP&T Meeting – Garching March 2011

Comments: -community should then review ITER ICRF - please include strong large Tokamak- based programme!

Arcing inside in-tokamak structure? Materials and dielectrics to withstand FW 14MeV flux?


Optimisation of ECRH systems ECRH has great advantages over NBI of small

‘nuclear island’ extension (capital cost reduction) ECRH has few coupling problems, but still not employed as

the dominant heating system in a high performance plasma context.

Key experiment for development would be 100% ECR-heated plasmas with high performance (20+ MW on JET or raise ECRH power on JT-60SA?)

Also for ECRH, experimental investigation of current driveefficiency merits special programme.

ECRH

NNBI


Optimisation of ECRH systems(II) Investigate reduction in complexity of in-tokamak launch by

developing frequency tuneable EC H&CD - allows for antennas with fixed launching angle and removal of the remote/front steering mirror concepts fundamentally different development branches of EC H&CD components.

Develop the broadband synthetic diamond window options and improve diamond window reliability?

Gyrotron efficiency (ITER prototype at JAEA) is now ~55% and transmission is ~95%. Improvements in gyrotron efficiency could come by improving the electron gun performance and making use of multi-stage depressed collectors the gyrotron 55% to > 70% possible?

RAMI analysis especially Materials assessment of radiation hardness of in-port launcher system – identification of issues. M Thumm – EFDA PPP&T meeting – Garching – March 2011


DEMO Divertor Optimisation :‘advanced divertors’ – magnetic shaping, not technology

‘Super-X’ is one concept where magnetic geometry could handle extremely high Divertor loads

• SOL taken to large major radius – natural flux expension;

• SOL passes through low PF region - connection length is increased – further spread of power – - volume to enable power radiation before striking target.

Concept to be tested on MAST-Upgrade. If successful could be incorporated into Divertor satellite and DEMO

Issues – in-vessel coil shieldingEFDA evaluation beginning

Super-X divertor concept

Kotschenreuther, Valanju, U Texas


Other DEMO Optimisation programme elements? • Other likely programme lines:

– Development of high purity versions of Eurofer, and low activation versions of conventional high temperature FM steels.

– Development of ODS Ferritic steels (if not in baseline!).– Resolution of issues relating to ‘second string’ Helium-cooled

blanket concept.– Other solutions to Divertor problem by magnetic concept

(‘snowflake’?) rather than engineering.

• Possible programme lines:– Divertor technology back-ups (water-cooled as back –up form

helium or vice-versa!); Liquid lithium divertor?– (If baseline is pulsed) Fusion-relevant energy storage systems.


DEMO Strategic Risk Reduction

An assessment of strategic risks to a DEMO programme is urgent (elements of this are proposed in the EFDA 2012 PPP&T programme)

Two elements stand out for this presenter:Component Test FacilityHigh Temperature superconductors, as a guard

against future Helium shortages[associated to this – helium leak reduction programme.]


DEMO Strategic risk reduction (I):A Component Testing programme?

Critical paths mean DEMO is unlikely to be ready for commissioning before the late 2030s.

Consequently it is high priority to ensure an efficient DEMO programme – high availability for Blanket Testing. Should get to ‘plateau’ region of radiation embrittlement (~> 6

MW.a.m-2 or 60 dpa) as soon as possible. This is 3 full power years. At 30% availability takes 10 years.

DEMO requires to breed tritium, relying for high availability operation on some of the components under test;

DEMO is a large and complex machine. Mean-Time-To-Replace (MTTR) test components will thus be large – leading to possible significant delays in a test programme.

IFMIF does not test components. As a strategic risk reduction exercise, the goals of a

Components testing programme and the feasibility of a pre-DEMO Components Test Facility (CTF) should be examined.

[EU Fusion Facilities Review -2008; UK 20 year Fusion Review 2009]


CTF• To be useful a CTF must:

– produce long periods of steady-state plasma burn to achieve the required integrated neutron yield –• with a fusion spectrum;• in a tokamak environment with accompanying stress fields;

– be compact and tritium efficient enough not to depend on tritium breeding;

– accommodate fully functional test components on the scale of ~ 1 m2 (relevant scale for component issues);

– deploy significant area, over 10 m2, to test several scaled components in parallel(e.g. blanket modules);

– be able to test prototype components up to some level before the serious start of a DEMO programme.


Compact CTF design – testing capability

ST-CTF (Culham, EU)

Compact Spherical TokamakCompact Spherical TokamakFusion power ~ 36 MWFusion power ~ 36 MW

Neutron wall- load ~ 1.0 MW.mNeutron wall- load ~ 1.0 MW.m-2-2

PPDIVDIV ~ 30 MW.m ~ 30 MW.m-2-2

2009 version has Super-X concept2009 version has Super-X conceptTritium consumption ~ 1.8 kg/fpyTritium consumption ~ 1.8 kg/fpy

Tritium bred ~ 47% of usageTritium bred ~ 47% of usage Tritium would be available from Candu programme for both ITER and a CTF.

MAST-Upgrade will test ST physics

Testing to 20 dpa (2 MW.a.m-2) at 1MW.m-2 and 33% availability takes ~ 6 years.

Does CTF have a consistent DEMO-stage mission??


A tentative time-lineCTF looks late unless we move fast

2010--15 2016--20 2021--25 2026--30 2031--35 2036--40

60 dpa IrradSite

Design + Constuct

Commission

Design ConstructOperate

Pre-concept + Baseline select

Baseline Concept

+ R&D

Concept DR(Baseline Blanket )

Baseline Scheme+R&D

IFMIF

ITER TBM data

Detailed DEMODesign

Construct DEMO

DEMO DivertorSatellite

CTF ?

20 dpa


DEMO Strategic Risk reduction (II):High-Temperature superconducting magnets

• High-temperature superconductors as a replacement for ~ 4K technology lead to:– Very modest power savings ( ~ 20 MW out of 570 MW BoP power for

‘Model B HCLL reactor goes to cryoplant);– simplification of cryogenic plant & shields etc.

very much ‘Generation II’ issues – other industries will develop HTSC and we cannot match their huge research budgets.

• …but strategically, high-T superconductors are needed in Fusion Technology because of the Helium resource problem.

• Terrestrial Helium presently comes from Natural Gas exploration – finite resources (~100 years)

• Huge reserves in atmosphere ~ 4 109 tonnes – enough for ~ 107

ITER cryosystems• ….. air separation of helium will be expensive to develop (can we

afford it in our baseline???)…… Problem will hit ‘roll-out’ of Fusion Economy (CCFE/Cambridge/Linde modelling).

• Long-term Fusion should unlink itself from Helium where possible, (and strategically needs to develop leak-tight He systems!)


Superconductors – helium free?

A

B

‘A’ -- field at the ITER TF conductor surface ‘B’ -- field at the ITER PF conductor surface

YBCO-type HTS can get SC performance above Liquid Neon temperatures – developments are clearly needed.

Neon can be air-separated routinely

(~ 4* concn of atmospheric Helium)

HTSC ‘Roebel’ cable - 1.3m length


Summary and Conclusions• To optimise the ‘DEMO Stage’ schedule we propose alignment into:

– Baseline;– Optimisation/Short term risk reduction;– Strategic Risk Reduction

• DEMO Mission clarification + Systems Engineering approach should guide the Baseline design selection

the existing critical path is through IFMIF structural materials qualification and ITER TBM results - conservative choices and de-risking should be used to avoid making the critical path situation more complex!

• DEMO electrical grid supply is maintained but downplayed. Pulsed Operation may be a conservative early choice.

• Maximum use of parallel programmes (Generation IV, HTSC developments) is urged for political and economic reasons.

• A ‘DEMO Divertor satellite’ is identified as a baseline facility.• The key optimisation issue is the Improvement of H&CD efficiency.• A Components Test Facility should be examined as a strategic risk

reduction programme element.• High temperature superconducting magnets should not be in the

baseline, but are needed in the longer run to reduce reliance on increasingly scarce Helium resources.



Appendix slides


Blanket choices affect complete DEMO design

• Energy use of secondary circuits (and hence nett plant efficiency) eg. high pumping power required for:– MHD-induced pressure drops for Liquid-metal designs;– high-flow, high pressure Helium cooling ( ~400 MW pump power!).

• Character of ‘Balance of Plant’:– Water-cooled blanket PWR - like primary circuits piggy-back

on Fission-plant engineering;– High-pressure He cooling primary circuits – may be developed by

Generation IV fission – or needs dedicated Fusion development ?• In-vessel operational safety/availability:

– hazards of interaction between coolant and blanket material (eg. H2O – Li ceramics or H2O – beryllium);

– hazards from corrosion by coolant (Li molten salts, liquid LiPb);– rupture of high pressure coolant (water raises steam – rupture to

vessel?; He ruptures module – regenerates cryopump?).• Minimisation of blanket change duration drives aspects of

in-vessel design (pipe connections, supports.


Accelerator challenge

‘5 dpa’ route


Decay of Blanket structure RAFM steel:irradiation period 5 years

EUROFEREUR

Ref [2]


EUROFER (data in wt%) Element Case 1

(specification, without impurities)

Case 2 (real material)

Case 3 (achievable

material) Al 0.008 0.0001 As 0.02 0.001 B 0.001 0.0001 C 0.11 0.11 0.11 Ca 0.0002 0.0001 Ce 0.003 0.0001 Co 0.005 0.001 Cr 9.0 9.0 9.0 Cu 0.0037 0.001 Fe bal bal bal Hf 0.0001 0.0001 K 0.0002 0.0001 Mn 0.40 0.40 0.40 Mo 0.0012 0.0001 N 0.03 0.03 0.001 Nb 0.001 0.00001 Nd 0.0002 0.0001 Ni 0.005 0.001 O 0.01 0.001 P 0.005 0.001 Re 0.0001 0.0001 Ru 0.001 0.001 S 0.003 0.001 Sb 0.01 0.001 Si 0.05 0.05 0.05 Sn 0.003 0.001 Ta 0.07 0.07 0.07 Ti 0.01 0.01 0.01 V 0.20 0.20 0.20 W 1.1 1.1 1.1 Zr 0.0001 0.0001

Real materials have trace impuritiesEurofer Chemical composition(wt%):

1.Pure - ideal2.Real - present day3.Achievable

Fusion Reactor Materials environmental basis: Manufacturability

Reference Eurofer


For EUROFER to achieve Reference composition Nb impurity needs to be further decreased by two orders of magnitudes to 0.00001% (~0.1 ppm)

EUROFER Blanket Material• replace every 5 years;• Pfus = 3 GW;• Neutron Wall Load = 2.3 MW.m-2

for 5 years

Fusion Reactor Materials environmental basis: Manufacturability – effect on waste

Ref [11] : P Batistoni et al.

Hands-on recycling level


H&CD efficiency for DEMO:assumptions vs reality (II) - ECRH system efficiency: ITER System

PPS to gyrotron

TR

PLaunch

PRF

gyro =PRF/PPS

TR ~ 600/630 ~ 0.95

Acc

el P

S

~72

kv

Dec

el

PS

~

28kv

PRF=1MW

Beam current ~38AJapanese Gyrotrongyro= 55%

For ECRH WP ~ 0.55x0.95 ~ 0.52

See eg. Kasugai et al., and refs therein; IAEA FEC Geneva 2008


Fusion economy – helium demand

3500 GWe (~30% of market)

Cai Zhiming – Univ of Cambridge; Richard Clarke - CCFE

Saving by using HTSC sustainable economy

Fusion roll-out with LTSC+He-cooled Blanket and Divertor

Aggressive Leak-learning and recycling scenario

Documents

International Meeting “MFE Roadmapping in the ITER Era” PPPL, 7 th -10 th September 2011 (this work was supported by UK EPSRC and Euratom) CCFE is the