Progress in Fusion Materials Research - The FIRE … · Progress in Fusion Materials Research Steve Zinkle Materials Science & Technology Div. IEEE/NPSS 23rd Symposium on Fusion Engineering

Progress in FusionMaterials Research

Steve ZinkleMaterials Science & Technology Div.

IEEE/NPSS 23rd Symposium onFusion EngineeringMay 31-June 5, 2009San Diego, California

2 Managed by UT-Battellefor the U.S. Department of Energy

Outline• Examples of structural materials design data• Overview of key temperature regimes for radiation damage

– Amorphization, point defect swelling, void swelling– Upper and lower operating temperature limits

• Materials design strategy and prospects for developing newhigh performance structural materials for fusion energy


The Launching of Fusion Energy Bridges theDevelopment of Modern Materials Science

1940 1950 1960 1970 1980 1990 2000 2010

CP-1

Shippingport

Development of Mat. Sci.as an academicdisciplineDevelopment of

Fracture mechanics

Initiation offusionmaterials R&D

Nuclear >10%US electricity

ITERconstruction

Tokamak era

•Fusion energy systems should take maximum advantage of currentand emerging materials and computational science tools

ENIACENIAC1 Gflops achieved;high performancecomputing centersestablished

1 Tflops 1 Pflops

ZETA torroidalpinch


Development of structural materials for applicationsinvolving public safety is historically a long process• “When you hear something about a new material, write it

down because it will be the best thing you’ll ever hear aboutit” (Jim Williams, paraphrasing Bob Sprague of General Electric)

• Aerospace structural materials– Over 50 years to develop TiAl intermetallics from initial studies in

1950s– Design cycle times have been reduced to 3-5 years, but development

and qualification of new materials still requires >7 years• Qualification time dominated by creep and fatigue testing

• Structural materials for nuclear reactors– Qualification requires all of the mechanical property testing on

unirradiated material, plus neutron irradiation and testing ofirradiated material• Sequential approach would lead to unacceptably long qualification times


History of improvement in temperature capability of Ni-base superalloys• Historical rate of improvement is ~5oC/year

Y. Koizumi et al., Proc. Int. Gas Turbines Conf., 2003, paper TS-119


Design tensile strengths for Type 316 stainless steel

A.A.F. Tavassoli


H. McCoy, ORNL

1350 K, 17.2 MPaHeat 1614,45 µm

Heat 530870,44 µm

Heat 530870,44 µm

Large variability in thermal creep behavior for threeheats of nominally identical Nb-1Zr

• In addition to grain size, these results show that other microstructuralinhomogeneities can also affect the thermal creep behavior of Nb-1Zr


Development of Texture in Annealed Nb-1Zr• Texture pattern in recrystallized Nb-1Zr is strongly dependent on

annealing conditions

FHR: fast heating rate (>1000˚C/min)SFR: slow heating rate (10˚C/min)

Radiation Damage can Produce Large Changes in Structural Materials

• Radiation hardening and embrittlement (<0.4 TM,>0.1 dpa)

• Phase instabilities from radiation-inducedprecipitation (0.3-0.6 TM, >10 dpa)

• Irradiation creep (<0.45 TM, >10 dpa)

• Volumetric swelling from void formation (0.3-0.6TM, >10 dpa)

• High temperature He embrittlement (>0.5 TM,>10 dpa)

0

200

400

600

800

1000

1200

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

USJF82Hss2

Engi

neer

ing

Stre

ss, M

Pa

Engineering Strain, mm/mm

Representative USDOE/JAERI F82H Data:200-600°C, 3-34 dpa

Unirradiated YS

200°C/10 dpa250°C/3 dpa

300°C/8 dpa

400°C/10 dpa400°C/34 dpa

500°C/8 dpa

500°C/34 dpa

600°C/8 dpa

Tirr=Ttest

100 nm

50 nm50 nm

100 nm100 nmafter S.J. Zinkle, Phys. Plasmas 12 (2005) 058101

mechanicalproperty experiments

PASSANS TEM

APT

THDS

TEM, in-situ TEM

fracture testing

tensile testingFinite

elementmacroscopicdeformation,integrated systems

Thermo-dynamics,Kinetics

Rate theory1-D clusterevolutionequations

B.D. Wirth, UC-Berkeley

Radiation damage is inherently multiscale with interactingphenomena ranging from ps to decades and nm to m


Multidisciplinary Fusion Materials Research has Demonstrated the Equivalency ofDisplacement Damage Produced by Fission and Fusion Neutrons

Fission

(0.1 - 3 MeV)

A critical unanswered question is the effectof higher transmutant H and He

production in the fusion spectrum

5 nmPeak damage state iniron cascades at 100K

50 keV PKA(ave. fusion)�

10 keV PKA(ave. fission)

MD computer simulations predict comparablesubcascades and defect production for fission, fusion

Similar defect clusters produced by fissionand fusion neutrons as observed by TEM

Fusion(14 MeV)

Similar hardening behavior confirms the equivalency

50

100

150

200

250

300

350

0.0001 0.001 0.01 0.1 1 10 100

Neutron Radiation Hardening of Copper

0.2%

Yie

ld S

tren

gth

(MPa

)

Damage Level (dpa)

Tirr=30 - 200˚C

14 MeV neutrons: filled symbolsfission neutrons: open symbols

• 3 distinct swelling regimes areobserved in irradiated Al2O3

0

2

4

6

8

10

0 200 400 600 800 1000 1200 1400 1600

Keilholz; 1971 neutrons 20 dpa

Zinkle; 1998 Ar ions 10 dpa

Vo

lum

e I

nc

rea

se

( %

)

Irradiation Temperature ( K )

• Activation Energies: Al vacancy; 1.8-2.1 eV O vacancy; 1.8-2 eV Al, O interstitial; 0.2-0.8 eV

Clinard, 1982Al2O3 Swelling

Amorphous Point defect swelling Void swelling

Stage IIIStage I

0.1

1

10

0 200 400 600 800 1000 1200 1400 1600

Snead 2006Snead 2006 Blackstone 1971

Price 1973

Price 1969

Price 1973,#2Price 1973Senor 2003Snead, Unpub.

Swel

ling

(%)

Irradiation Temperature (°C)

0.5

0.3

20

23

5

0.2

0.7

7

Saturatable RegimeAmorphousRegime Non-Saturatable Regime

8.5 dpa

5 dpa

1.75 dpa

Irradiation-induced swelling in SiC

Stage I Stage III


Low tensile ductility in FCC and BCC metals after irradiation atlow temperature is due to formation of nanoscale defect clusters

Outstanding questions to be resolved include: Can the defect cluster formation bemodified by appropriate use of nanoscale2nd phase features or solute additions? Can the poor ductility of the irradiatedmaterials be mitigated by altering thepredominant deformation mode? (e.g.,twinning vs. dislocation glide)

Cu, Cu, TTirrirr=90=90˚̊C, 0.5 dpaC, 0.5 dpa

0

2

4

6

8

10

0.0001 0.001 0.01 0.1 1 10

Fabritsiev et al (1996)Heinisch et al (1992)Singh et al. (1995)

Unifo

rm E

long

atio

n (%

)

Damage Level (dpa)

Unirradiatedelongations

Tirr=36-90˚C

Tensile ductility of neutron-irradiated Cu alloys


Radiation hardening in V-4Cr-4Ti

0

200

400

600

800

1000

0 100 200 300 400 500 600 700

Effect of Dose and Irradiation Temperature onthe Yield Strength of V-(4-5%)Cr-(4-5%)Ti Alloys

Yiel

d St

reng

th (

MPa

)

Irradiation Temperature (˚C)

Ttest~Tirr

unirradiated

0.1 dpa

0.5 dpa 5-50 dpa

V-4Cr-4Ti

0.3 TM-2

0

2

4

6

8

10

12

0 100 200 300 400 500 600 700Un

iform

Elo

ngat

ion

(%)

Test Temperature (˚C)

0.5 dpa

0.1 dpa

Test Temperature isIrradiation Temperature

High hardening and loss of uniform elongation occursfor irradiation and test temperatures <0.3 TM


Comparison of the uniform elongation of neutronirradiated vanadium and molybdenum alloys

• Radiation-induced flow localization occurs for irradiationtemperatures below 0.3 TM

0

1

2

3

4

5

6

7

8

300 400 500 600 700 800 900 1000 1100

V-4Cr-4TiMo-5Re

Unifo

rm E

long

atio

n (%

)

Irradiation Temperature (K)

Ttest~Tirr

V-4Cr-4Ti Mo-5Re

0

1

2

3

4

5

6

7

8

0.15 0.2 0.25 0.3 0.35 0.4 0.45

V-4Cr-4TiMo-5Re

Unifo

rm E

long

atio

n (%

)

Homologous Irradiation Temperature (Tirr/TM)

Ttest~Tirr

V-4Cr-4Ti

Mo-5Re

Corresponding critical temperature for W alloys is ~830oC


Can we break the shackles that limit conventionalstructural materials to ~300oC temperature window?

Structural Material Operating Temperature Windows: 10-50 dpa

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

Thermal creep

Radiationembrittlement

ηCarnot=1-Treject/Thigh


The Overarching Goals for Fusion Power Systems Narrow theChoices and Place Significant Demands for Performance ofStructural Materials• Safety

• Minimization of Rad.Waste

• Economically Competitive– High thermal efficiency (high

temperatures)– Acceptable lifetime– Reliability

Fe-9Cr steels: builds upon 9Cr-1Mo industrial experience and materials database(9-12 Cr ODS steels are a higher temperature future option)V-4Cr-4Ti: Higher temperature capability, targeted for Li self-cooled blanket designsSiC/SiC: High risk, high performance option (early in its development path)W alloys: High performance option for PFCs (early in its development path)

Piet et al., Fusion Tech. 17 (1990) 636Based on


50040030020010000

2

4

6

8

10

12

TIME (h)

CREE

P ST

RAIN

(%

)

347HFG SS(foil from tubing)

creep rupture750°C 100 MPa

standard 347 SS(ORNL foil)

ORNL mod 2347 SS

Modern Materials Science Applied to “Old” Alloys YieldsDramatic Improvements in Performance and New Applications

Alloy composition is changed to produce amicrostructure of fine, stable precipitates

Low chromium steels for chemistry andenergy industry applications New cast austenitic steels for automotive and

heavy vehicle exhaust components

New austenitic stainless steels formicroturbine recuperators, and

Solar Turbines’ new Mercury50 Gas-Turbine Engine

0

25

50

75

100

125

150

0 200 400 600 800 1000 1200 1400

0.2%

Yie

ld S

tren

gth

(ksi)

Temperature (°F)

Current Alloy

New Alloy

Old New Solar Turbine Corp. Microturbine

High Performance Diesel Engine

10000800060004000200000

1

2

3

4

5

6

time (h)

cree

p st

rain

(%

)

CAST CF8C ALLOYSCREEP RUPTURE

850°C 35 MPaCommercialcomponent

CAT/ORNLmodifiedht 17577

Commercial Bainitic Steel ORNL “Super” Bainitic Steel

Similar Scientific Microstructural Design Produces:


Technology Transfer of CF8C-Plus Cast Stainless Steel• MetalTek International, Stainless Foundry & Engineering, and

Wollaston Alloys received trial licenses in 2005 (18 monthsafter project start)

• Over 350,000 lb of CF8C-Plus steel have been successfullycast to date

– Now used on all heavy-duty truck diesel engines made byCaterpillar (since Jan. 2007)

– Solar Turbines (end-cover, casings), Siemens-Westinghouse(large section tests for turbine casings), ORNL, and a globalpetrochemical company (tubes/piping).

– Stainless Foundry has cast CF8C-Plus exhaust components forWaukesha Engine Dresser NG engines

6,700 lb CF8C-Plus end-cover cast byMetalTek for Solar Turbines Mercury 50gas turbine

80 lb CF8C-Plusexhaust componentcast by StainlessFoundry forWaukesha NGreciprocating engine

SiMo cast iron

CF8C-Plussteel


Development of New Alumina-Forming,Creep Resistant Austenitic Stainless Steel

800oC in air, 72hY. Yamamoto, M.P. Brady et al., Science 316 (2007) 433


Modified chemistry and thermomechanical treatment procedure for new 9Crferritic/martensitic steels produces high strength

• Tensile strength and ductility arecomparable to high-strength experimentalODS steel

• Good toughness and high temperaturestrength are also produced indispersion-strengthed Fe-9.5Cr-3Co-1Ni-0.6Mo-0.3Ti-0.07C steel due to highnumber density of nano-size TiCprecipitates

Klueh and Buck, J. Nucl. Mater. 283-287 (2000)

R.L. Klueh et al., J. Nucl. Mat. 367-370(2007) 48; Scripta Mat. 53 (2005) 275

Modified thermomechanical treatment of 9Cr-1Mo steel introduces fine-scale precipitates


Large Increase in Rupture Life of Modified 9Cr-1Mo at650°C

•Thermo-mechanical treatment (TMT) of modified 9Cr-1Mo produced steel with over an order-of-magnitudeincrease in rupture life

R.L. Klueh et al., J. Nucl. Mat. 367-370(2007) 48; Scripta Mat. 53 (2005) 275


Nanostructuring Achieves Good FractureToughness and High-Strength Properties

UT-BATTELLE

Grain boundary nucleation of NC

OAK RIDGE NATIONAL LABORATORYU.S. DEPARTMENT OF ENERGY

• Nano-size grain size with very high grain boundary interfacial area• High number density of NC in-matrix with λ = 10-15 nm• High number density of NC decorating grain boundaries

BF TEM Fe M Jump Ratio

NFA 14YWT Developed at ORNL


Theory Has Shown That Vacancies Play a PivotalRole in the Formation and Stability of Nanoclusters

in ODS steel

The presence of Ti and vacancyhas a drastic effect in loweringthe binding energy of oxygen inFe (C. L. Fu and M. Krcmar).


14YWT Shows High Strength and Some Ductilityto -196ºC

0

200

400

600

800

10001200

1400

1600

1800

2000

2200

2400

0 5 10 15 20 25

14YWT-SM6

-196ºC-150ºC-100ºC25ºC300ºC500ºC600ºC700ºC

Stress (MPa)

Strain (%)

0

200

400

600

800

1000

1200

1400

1600

1800

0 100 200 300 400 500 600 700 800

14YWT-SM6_Tensile Data 2:28:36 PM 3/6/2006

YS - 14YWT-SM6UTS - 14YWT-SM6YS - Eurofer 97 ODSUTS - Eurofer 97 ODSYS - 14WT-SM3UTS - 14WT-SM3YS - Eurofer 97UTS - Eurofer 97

Stress (MPa)

Test Temperarture (ºC)

ODS Eurofer 97 supplied by L. Rainer

• At -196ºC (ε = 10-3s-1)- mixed mode dimple

rupture-cleavage failure- reduction in area = 43%- σf = 3.0 GPa- T.E. = ~7%

UT-BATTELLEOAK RIDGE NATIONAL LABORATORYU.S. DEPARTMENT OF ENERGY


High Fracture Toughness Achieved in 14YWT

-200 -100 0 100 200 300 400 500 6000

25

50

75

100

125

150

175

200 14YWT [KJc(1T)] 14YWT [KJIc] 12YWT [KJc(1T)] 12YWT [KJIc]

Frac

ture

Tou

ghne

ss [

MPa√m

]

Temperature [°C]

Unirradiated Fracture Toughness

-250 -200 -150 -100 -50 0 50 100 150 200 250 30025

50

75

100

125

150

175

200

14YWT [KJc(1T)] (unirradiated) 14YWT [KJIc] (unirradiated) 14YWT [KJc(1T)] (Tirr= 300oC) 14YWT [KJIc] (Tirr= 300oC)

Frac

ture

Tou

ghne

ss [M

Pa√ m]

Temperature [°C]

Irradiated 14YWT Fracture Toughness

• L-T Orientation• Pre-cracked: crack length to width (a/w) ratio of 0.5• Tested using the unloading compliance method (ASTM 1820-06)• KJc for brittle cleavage calculated from critical J-integral at fracture, adjusted to 1-T reference specimen KJc(1T) • KJIc for ductile deformation behavior calculated from critical J-integral at onset of stable crack growth

• The fracture toughness of 14YWT ismuch better than that of 12YWT

• The DBTT is shifted from ~75 ºC for12YWT to -150 ºC for 14YWT

OAK RIDGE NATIONAL LABORATORYU.S. DEPARTMENT OF ENERGY UT-BATTELLE

• Neutron irradiation to 1.5 dpa at300ºC does not degrade the fracturetoughness

• More testing is required though…


Solid State Consolidation of “Low-Cost” TitaniumPowders

• Potential reduced cost of plate, sheet, andnear net shapes by 50 to 90%

• Scrap reduced from 50% (conventionalmaterial) to less than 5% for new low costtitanium

• Demonstrated and exceeded “wrought”material performance parameters

Solid State Consolidated Plate, Bar, Sheet, and NetShape Components Produced from New “Low

Cost” Titanium Powders

W.H. Peter, C.A. Blue, et al.

Cost Break Down to Produce 1” Thick TitaniumPlate Using Kroll – VAR Melted Ti

A.D. Hartman et al. ,JOM September 1998. pp. 16-19


Development of low-cost, high-performance cast stainless steelfor ITER structures

ITER and shield module




•The US contribution to the ITER projectinvolves many complex stainless steelcomponents.•Casting methods may considerablyreduce time and cost associated withfabrication.•However, greater strength is neededfor the cast material to be accepted foruse in ITER.





•A science-based approachinvolving modeling,advanced analyticaltechniques and industrialexperience has been usedto develop cast steels withimproved properties.





Impact testing at 200-300°C.Samples do not break.







•Improved steels exhibit greater strength andperformance than typical cast grades andmeet ITER criteria.







•Improved steels exhibit greater strength andperformance than typical cast grades andmeet ITER criteria.

Tensile tests at 25°C



Recent research suggests high-strength steels that retainhigh-toughness are achievable

1st and 2ndgeneration steels(HT9, 2 1/4Cr-1Mo, etc)

Ultra high strength steels(nanocomposited ODS,Aermet, etc.)

•Generally obtained by producing high density of nanoscale precipitates and eliminationof coarse particles that serve as stress concentrator points


Potential Alloy Development Options for Tungsten

• Solute additions that beneficially modify the elastic constants(Poisson ratio effect) in order to improve ductility and fracturetoughness

• Nanoscale engineering of internal boundaries– Coherent twin boundaries for strengthening (instead of grain

boundaries); cf. K. Lu et al., Science 324 (2009) 349

• Create a high density of nanoscale precipitates for radiationresistance


SiC/SiC Composites Development (fusion program)Reference Chemical Vapor Infiltrated (CVI) Composites for Irradiation Studies

• Hi-Nicalon™ Type-S or Tyranno™-SA3 / PyC(50–150nmt) / CVI-SiC composites have beenselected as the reference materials

• Extensive engineering data generation for irradiated properties (including statistical strength) isplanned (prior studies utilized simple qualitative screening tests)

0

50

100

150

200

250

300

0.0 0.1 0.2 0.3 0.4

Tensile Strain (%)

Tens

ile S

tres

s (M

Pa) .

ORNL Tyranno-SA / PyC / CVI-SiC

tPyC = 25nm

tPyC = 250nm

0

1

0.1 1 10 100Neutron Dose [dpa-SiC]

S uIr

rad./S

uUni

rrad

.

3

2

2

7

67

7

4

5

5

5

5

2

2

3

1

1

1

888

82

2

7710 9

6

1: 500C, HFIR2: 400C, HFIR 3: 200-500C, HFIR4: 300-500C, JMTR5: 430-500C, EBR-II

Hi-Nicalon Type-S/PyC/FCVI-SiCHi-Nicalon/PyC/FCVI-SiCNicalon/PyC/FCVI-SiCTyranno-SA/PyC/FCVI-SiCMonolithic CVD-SiC

6: 300C, HFIR7: 800C, HFIR8: 800C, JMTR9: 740C, HFIR10: 630, 1020C, ETR

3

Bend strengths of irradiated “3rd generation”composites show no degradation up to 10 dpa

Future work: advanced matrix infiltration R&D; joining;hermetic coatings; SiC/graphite composites, etc.

“3rd generation” composites

earlycomposites

L.L. Snead, Y. Katoh, T. Nozawa et al.

Current Status of Current Status of SiC/SiC SiC/SiC CompositesComposites

Tensile strength<unirradiated>

(500MPa/FS)

Tensile modulus(300GPa/FS)

Density(3.0Mg/m3/FS)

Thermal conductivity<unirradiated>

(20W/m-K/FS)

Max. Temperature <unirradiated>

(1500℃/FS)

Tensile strength <irradiated>

(500MPa/FS)

Max. Temperature<irradiated>

(1500℃/FS)

Thermal conductivity <irradiated>

(20W/m-K/FS)

TAUROARIES-ATD-SSTRJUPITER-CVI (1999)CREST-NITE (2001)

Estimation fromflexural strength

A. Kohyama

Fusion materials research must rely heavily on modeling due toinaccessibility of fusion-relevant operating regime

1015

1016

1017

1018

1019

1020

1021

105 106 107 108 109

MagneticInertial

Con

finem

ent

Qua

lity,

neτ

(m-3

s)

Ion Temperature (K)

1955

1960

19651970-75

1975-80

1980s1990s

Breakeven

Ignition

• Extrapolation from currently available parameter space to fusion regime is muchlarger for fusion materials than for plasma physics program

• An intense neutron source such as IFMIF is needed to develop and qualify fusionstructural materials

Plasma physics experimental achievements


Conclusions

• Existing structural materials are not ideal for advanced nuclearenergy systems due to limited operating temperature windows– May produce technically viable design, but not with desired optimal

economic attractiveness

• Substantial improvement in the performance of structural materialscan be achieved in a timely manner with a science-basedapproach

• Design of nanoscale features in structural materials confersimproved mechanical strength and radiation resistance– Such nanoscale alloy tailoring is vital for development of radiation-

resistant structural materials for fusion energy systems– Experimental validation will ultimately require testing in appropriate

fusion-relevant facilities

Documents

Progress in Fusion Materials Research - The FIRE … · Progress in Fusion Materials Research Steve Zinkle Materials Science & Technology Div. IEEE/NPSS 23rd Symposium on Fusion Engineering