Pathways for Fusion Energy: Impactof Materials Options

Steven J. Zinkle1,2, Lance L. Snead3, Brian D. Wirth1,2
1University of Tennessee, Knoxville, TN USA
2Oak Ridge National Lab, Oak Ridge, TN USA
3Massachussetts Institute of Technology, Cambridge, MA USA

ANS topical meeting on Technology of Fusion Energy
Philadelphia, PA
August 22-25, 2016
1

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includetran
Historicaltrends
vehiclemass major
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design and construction ofvehicles.
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the progres
designand constructionof vehicles.
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thedurability and higher strength.
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and thedurability
and higher
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yearshastheseenmodernautomobileshift
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variousplastics
polymersand composites)
aluminum,(including
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Evolution of materials used in commercial aircraft

3W.G. Roeseler et al., in 16th Int. Conf. on Composite Materials,
Kyoto, Japan (Japan Soc. for Composite Mater., 2007)

US Fission Reactor Performance
Steady improvement in fuel burnup and capacity factor, butlittle change in thermodynamic efficiency in past 40 years

Gen I reactors

4Gen II reactors

S.J. Zinkle & G.S. Was, Acta Mater., 61 (2013) 735(updated to include most recent reported data) andNRC list of power reactors (2016)

Timeline of structural materials used in light water reactors

5S.J. Zinkle, K.A. Terrani, L.L. Snead,Current Opin. Sol. State & Mater. Sci.,in press (20th anniv. special issue, 2016)

Thermal creep and void swelling in sodium-cooled fast
reactor cladding is problematic for conventional steels

6

S.J. Zinkle, K.A. Terrani, L.L. Snead,Current Opin. Sol. State & Mater. Sci.,in press (20th anniv. special issue, 2016)

Fusion materials R&D and selection will have a major
impact on the ultimate viability of fusion energy• Robust plasma facing materials– plasma compatibility– high heat flux capability– minimal net surface erosion– low tritium retention– moderate lifetime with high reliability– use captured heat for power conversion?

Solid vs. liquidwall approaches• High operating temperatures (thermodynamic efficiency) andhigh reliability/ long lifetimes are likely needed to offset thehigher capital costs for fusion– Requires high performance structural materials• Reduced activation materials for environmental attractiveness– no public evacuation during accident scenarios– no long lived wastes requiring geologic disposition• Blanket materials for fuel regeneration and heat extraction• Magnet and plasma diagnostic/heating materials

7Requirements:Ø Have an economically competitive life-cycle cost of electricity:
Low recirculating power (Increase plasma Q, …); COE has a “hyperbolic”
dependence (∝ 1/x) andHigh power density (Increase Pf ~ β2BT4 , …)
improvements “saturate”
High thermal conversion efficiency;
after certain limit
Less-expensive systems.
Ø Gain Public acceptance by having excellent safety andenvironmental characteristics:
Use low-activation and low toxicity materials and care in design.
Ø Have operational reliability and high availability:
Ease of maintenance, design margins, and extensive R&D.
Ø Acceptable cost of development.

Stay close to present-day

Larger extrapolation from present

Translation of Requirements to GOALS forFusion Power Plants

Bainitic steel development for fusion vacuum vessel• Target: Fusion structural applications in next-step
fusion devices (FNSF or DEMO);― Vacuum vessel (>400-500°C, relatively low dose)― Structural ring, magnet shields• Why 3Cr-3WV(Ta) steels?:― Inexpensive low alloy steel― Improved tensile and creep properties due toformation of carbide-free acicular bainite ferrite(lower bainitic microstructure)― potentially no requirement of PWHT, suitable forlarge volume components• Approach:– Computational thermodynamics for alloy design>100X lower waste disposal
burden vs. 316SS

L. El-Guebaly, Fus. Sci. Tech. 64 (2013) 449

Fig. 2. Class C WDR for the seven candidate steels for
ARIES-CS thick VV with “nominal” and“present” impurities.

~2X• Effect of minor alloying additions on transformation kinetics• Phase equilibrium of strengthening second-phases– Property evaluation of 3Cr-3WV(Ta) steels

452• Production of CCT diagrams• Mechanical property / weldability
9Y. Yamamoto, ORNL

3Cr data: R. Klueh, Met
Mat Trans 28A (1997)335

Effect of Neutron Irradiation on the Ductile to Brittle
Transition Temperature in Ferritic/martensitic Steels
after Boutard et al., C.R. Physique. 9 (2008) 287 and Klueh & Harries (2001)

Need <12%Cr for
acceptable DBTT

0.3 TM

10S.J. Zinkle & N.M. Ghoniem, J. Nucl. Mater. 417 (2011) 2

Effect of Neutron Irradiation on the Ductile toBrittle Transition Temperature of Reduced
Activation 9Cr Ferritic/Martensitic Steels
150

DBTT Eurofer97 ANL 16dpa
DBTT F82H-mod 16dpa
DBTT Eurofer97 WB 16dpa
T0 F82H IEA 4-5dpa
To JLF1 4-5dpa
DBTT F82H IEA 20dpa
DBTT F82H heat2 4-5dpa
DBTT F82H heat2 20dpa

DBTT or T0(oC)

100~100oC variation in DBTT fordifferent heats of 9Cr reducedactivation FM steel- what is the cause of thisvariation?
- Can further reductions beachieved?

500-50-100-150200250300350400450500

550

Irradiation Temperature (oC)
11S.J. Zinkle, A. Möslang, T. Muroga and H. Tanigawa,Nucl. Fusion 53, no.10 (2013) 104024

Can we break the shackles that limit conventional
structural materials to ~300oC temperature window?
Structural Material Operating Temperature Windows: 10-50 dpa
Radiation
embrittlement

Thermal creep

ηCarnot=1-Treject/Thigh
12

Additional considerations such as He embrittlement and chemicalcompatibility may impose further restrictions on operating window
Zinkle and Ghoniem, Fusion Engr. Des. 49-50 (2000) 709
S.J. Zinkle & J.T. Busby, Mater. Today 12 (2009) 12

Creep rupture behavior for conventional 9Cr steels vs.
next-generation steels (designed by computational thermodynamics)>50% increase in σrupt

13

S.J. Zinkle et al., Nucl. Fusion (2016) SpecialIssue on Materials for a Fusion Reactor, in pressCreep rupture behavior for ODS steels vs. conventional
9Cr steels>100% increase in σrupt

14

S.J. Zinkle et al., Nucl. Fusion (2016) SpecialIssue on Materials for a Fusion Reactor, in press

Approaches for radiation resistance 3: High Sink Strength
Example: Volumetric Void Swelling of Irradiated FeCrNi Austenitic Alloys

High sink strengths canbe achieved via
nanoscale precipitation

109 dpa

200 nmFor void swelling
resistance, sink strengths>1015/m2 are generally
sufficient for light water
fission reactors; fusion
reactor irradiation may
require even higher sinkstrengths (>1016/m2?) due
to transmutant Heproduction

Sppt~4πRN
15S.J. Zinkle and L.L. Snead, Ann Rev. Mat. Res., 44 (2014) 241

Effect of Initial Sink Strength on the Radiation Hardeningof Ferritic/martensitic Steels
Currentsteels

Next-generation(TMT, ODS) steels

Dramatic reduction inradiation hardening occurs
when average spacing
between defect clusters(dislocation loops, etc.) is>10x average spacing
between defect sinks

Nloop-1/3 > 10 Stot-1/2
or equivalently,
Stot > 5-10 Srad defects
Sppt~4πRN
16

S.J. Zinkle and L.L. Snead, Ann Rev. Mat. Res., 44 (2014) 241
S.J. Zinkle et al., Nucl. Fusion (2016 special issue) in press

H retention increases dramatically in the presence ofcavity formation
3 to 5x increase in retained hydrogen when cavities arepresent, even with 2-3x reduction in neutron fluence exposure
1700-3700 appm H(rad.-induced cavities present)

Retained H level is ~100x
higher than expected fromSievert’s law solubilities
H solubility at 0.33 atm H pressure (appm)

500-700 appm H(few cavities)

104

1000

Bolt head
1 mm
320oC, 19.5 dpa

Bolt shank
25 mm
343oC, 12.2 dpa

Near threads
55 mm
333oC, 7.5 dpa

0.33 atm H

Measured H conc.
(~10dpa irrad. 316SSwith cavities)

100

Sievert’s Law(defect-free 316SS)

316 SS
10Fe
11

1.2

1.4

1.6

1.8

2

1000/T (K)Baffle-former bolt removed from Tihange-1 (Belgium) pressurized water reactor
17

Type 316 austenitic stainless steel

F.A. Garner et al., J. Nucl. Mater. 356 (2006) 122
S.J. Zinkle, Nucl. Fusion 53 (2013) 104024

2.2

Temperature dependent microstructures in irradiated materials
SFT, Dislocation loops

Bubbles, voids, precipitates, solute segregation

Amorphization

Network dislocations(intermetallics & ceramics)

Grain boundary
helium cavities

250 nm

50 nm
50 nm

100 nm
5 µm

10 nm

10 nm

Stage I
018100 nm

Stage III
0.1

Stage V

0.2
0.3
0.4
Irradiation Temperature (T/TM)

0.5

0.6

Zinkle & Snead, Ann. Rev.
Mater. Res. 44 (2014) 241

Neutron effects can influence D retention in W
• More D retained in damaged W; D distribution relatively uniform in plasma exposed,
neutron irradiated W• D desorption spectra completely different for ion-damaged and neutron irradiatedsamples --- Temperature of plasma (& neutron/ion) exposure also important factor(20MeV W+, 0.5 dpa, RT)(0.025 dpa, 323K)(0.025 dpa, 323K)

H isotope trapping is strongest at irradiation
temperatures where cavities form
19Y. Hatano et al., Nuclear Fusion, vol. 53, p. 073006, July 2013.

PFC

ITER’s current operating plans involve a tungsten
divertor: significant concern about sub-surface
helium bubble formation & surface morphology changes
that will influence core plasma performance & tritium
20retention, and/or tritium-containing dustTritium inventory and control is significant challenge*

Worry about tritium retention in He bubbles
forming below PFC surface*Ref: D. Whyte, BPMIC presentation (2015)Low	  Readiness	  of	  Breeding	  Blankets	  (no	  ideal	  candidate)	  	  • Large	  knowledge	  gaps	  exist	  even	  with	  a	  successful	  ITER	  TBM	  programme	  • Feasibility	  concerns/performance	  uncertainCes	  à	  Selec@on	  now	  is	  premature	  • Strengthen	  blanket	  technology	  R&D.	  Use	  System	  Engineering	  approach.	  

Current designs are based on 1980s-1990s concepts: any room for further innovation?
Concerns
Tested	  in	  ITER	  TBM	  HCPBHCLLWCLL

DCLL☺

Suitability	  for	  Eurofer	  FW	  heat	  ﬂux	  capability	  Safety	  issues	  of	  coolant	  Technology	  readiness	  BoP	  Poten@al	  for	  high	  coolant	  outlet	  temperature	  Coolant	  pumping	  power	  Shielding	  eﬃciency/	  n-­‐streaming	  void	  space	  Ac@va@on	  products	  in	  coolant	  (water)	  Breeding	  eﬃciency	  Tri@um	  extrac@on	  from	  breeder	  Tri@um	  extrac@on	  from	  coolant	  Tri@um	  permea@on	  through	  heat	  exchanger	  22G. Federici et al., 2013.

Significant gaps in bridging materials from ITER to DEMO- virtually no materials systems currently used are reactor viable Magnet	  ITER	  Fusion	  Life@me	   Power	  Neutron	   Reactor	  Fluence	  	   Annual	  (nm-­‐2)	   Neutron	  Fluence	  (nm-­‐2)	  n m-2s-1

5E+22	  Blanket	  7E+15	  Magnet	  1.9E+21	   2.6E+22	  Divertor	  1.1E+19	   1.5E+20	  Vacuum	  Vessel	  3.4E+11	   4.5E+12	  Cryostat	  3.7E+21	  5.1E+14	  3E+17	  3E+16	  4E+15	  5E+14	  6E+13	  7E+12	  8E+11	  9E+10	  3E+10	  ITER	  Test	  Blanket	  Module:	  RAF	  Steel	  Zinkle & Snead, Ann. Rev. Mater. Res., 44 (2014) 241

First	  Wall	  :	  Be-­‐Cu	  alloy-­‐316	  steel	  23Diagnos@c	  Divertor:W	  (or	  CFC)-­‐Cu	  alloy-­‐316	  steel	  Current estimated lifetime for some key diagnostics range
from 1 week to 1 year
• neutron power
density of
0.5 MW / m2 orneutron power of450 MW• The fluence of
0.3 MWa / m2
gives 2 x107 sburn at 450 MW•The lifetime ismainly limited byradiation effects•except where
shown with ().

G. Vayakis, The ITER radiation environment for
diagnostics, N 55 RI 38 04-05-06 W 0.1

From A. K.Ebisawa et al, Plasma
Diagnostics for ITER-FEAT, HTPD 2000

High	  Temperature/Field	  Superconductors:	  Je	  State	  of	  the	  Art	  •
•
•
•

•
•

HTS	  superconductors	  can	  extend	  the	  magneCc	  ﬁeld	  range	  to	  >>	  20	  T.	  
Stress	  will	  be	  a	  major	  concern	  at	  very	  high	  ﬁelds.	  
Quench	  protecCon	  and	  stability	  have	  not	  been	  addressed.	  
HTS	  wires	  and	  tapes	  are	  sCll	  primiCve	  compared	  to	  Nb-­‐Ti	  and	  Nb3Sn	  (typical	  commercial	  batch	  lengths	  for	  YBCO	  are	  currently	  50	  –	  150	  m).	  
MgB2	  is	  not	  promising	  for	  high	  ﬁeld	  applicaCon	  and	  Jc	  is	  too	  low	  at	  >	  20	  K.	  	  Neutron-­‐induced	  degradaCon	  needs	  further	  study	  but	  HTS	  appear	  to	  be	  comparable	  to	  convenConal	  superconductor	  limits	  Plot	  maintained	  at:	  hfp://magnet.fsu.edu/~lee/plot/plot.htm	  	  25Minervini/Lee - Fusion Nuclear Science Pathways
Assessment: Materials Working Group Meeting

Aurora	  CO,	  May	  4,	  2011	  Fusion materials challenges and opportunities• Challenges– Plasma facing components

Increasing opportunities
for leveraging broader
mater. sci. community• Will tungsten work?

– Tritium containment and online extraction/fuelreprocessing– Nonstructural materials lifetime in a DT fusion
environment• Plasma diagnostics (optical fibers, electricalinsulators, etc.)• Plasma heating feedthrough insulators• Next generation magnet systems (insulation, HTC
superconductors?)• Ceramic breeders– Structural materials• Is there a viable option beyond 5 MW-yr/m2? (50 dpa)
26