status and prospect of sic and fci r&d...2009/08/20 · status and prospect of sic and fci...

Status and Prospect ofSiC and FCI R&D

Prepared for

FNST National MeetingAugust 18-20, 2009, UCLA

by

Y. Katoh (ORNL) on behalf of US Fusion SiC/SiC Program and FCI Team

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

SiC Ceramics and Composites for Fusion Energy

• SiC ceramics and composites– Enables very high temperature operation– Offer exceptional radiation stability– Offer ultimate short-term low-activation and decay heat– Are available and being qualified for nuclear application

• Target applications in fusion system include:– Flow channel insert for DCLL/HCLL blankets, incl. ITER TBM– Active cooling panel for high temperature dual-cooled blanket– Blanket structure for long-term He- or PbLi-cooled blanket– Primary heat exchanger used with high temperature blanket


US Fusion SiC/SiC Program Scope• Program scope encompasses from fundamental

materials science to interaction with technology programs– Primary focus on advancing “fusion materials science”

• Physical processes and mechanisms for environmental (including radiation) effect phenomena intrinsic to SiC

• Constitutive modeling of composite properties and support experiments

– Materials development, design, and characterization relevant to various proposed fusion energy applications

– Interact with fusion technology and fission energy R&D• Fusion technology program and activities

– FNST, Blanket design, neutronics, TBM, etc.• Fusion FCI and fission composite R&D in SBIR/STTR

– Toward development of specific materials and enabling technologies• Growing interactions with programs in support of fission energy R&D


Ceramic Composites for Nuclear Applications

• Opportunities for ceramic composites– HTGR/VHTR (reactor unit components)– ALWR (fuel cladding)– GFR (whole core)– LS-AHTR (fuel cladding and core structures)

• HTGR/VHTR will heavily rely on ceramic composites (C/C and SiC/SiC)

– Current designs assume use of metallic alloys– Recent design studies identified composite needs– U.S. NGNP, RSA PBMR

• R&D focus for VHTR application progressing toward licensing

– Nuclear Composite WG in ASTM C28.07 on Ceramic Matrix Composites

– Ceramic Composite TG proposed in ASME Sec. III Graphite Core Component SC


Fundamental questions related with long range applications of SiC: Examples

• Very high dose irradiation effect– Does the self-stabilizing mechanism work forever?

• Very high temperature irradiation effect– What is the lowest temperature at which cumulative irradiation effect takes

place? How?• Interface design for irradiation stability

– What is interface damage in composite like at very high dose? What makes interface most stable?

• Irradiation creep– Does irradiation creep limits dose-load envelope of operating condition?

• Transmutation effect– What are the influences of various transmutation in fusion spectrum on

key properties?• Joining technology

– What are the promising joining technologies for SiC-SiC and SiC-metals which are promising for use in high radiation field?


Irradiation Effects on Fast Fracture Properties

Irradiation conditions in which retention of unirradiated strength is confirmed

0

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1 10 100

Fluence [DPA]Irr

adia

tion

Tem

pera

ture

[C]

Price 1977Price 1982Jones, 1997Snead, 2002New some, 2007Snead, 2007Kondo, 2008Ozaw a 2007New some 2007Hegeman 2005Hinoki 2002Katoh 2005Katoh unpublsihed

Solid symbols: CompositesOpen symbols: CVD SiC

~20% Strength Reduction

• Single-phase, polycrystalline beta-SiC is exceptionally radiation-stable

– Additional experimental data exist indicating strength retention of CVI SiC matrix to 80 dpa at 500ºC (Jones, 1997).

• The latest experiment indicates ~20% strength reduction at 10 dpa, 1500ºC (Kondo, 2008)

– Sole data ever reported indicating strength reduction of high purity CVD SiC by irradiation

– Likely related with void formation on grain boundaries

Katoh & Snead, 2009, Fusion Sci&Tech 56


2010-10/04800>50JCR 5-820092009200910/0450030JCR 9-12

20092009200910/0480030JCR 1-4PIE endPIE beginExtractionInsertionTemp.(°C)DPACapsules

JAEA-DOE Phase-V High Fluence SiC/SiC Irradiation Experiment

• Intended to address high-dose (>30 dpa) irradiated performance of Hi-Nicalon™ Type-S 2D satin-weave, multilayer interphase, CVI SiC matrix composite. – Both parties agreed to perform irradiation and PIE of JCR-1~12 rabbit capsules

in JAEA-DOE collaboration. – Each rabbit capsule contains two bend bar samples. Irradiation started in 2004. – Planned PIE items include strength, elastic modulus, swelling, and thermal

conductivity. Samples would be available for higher level evaluation (such as TEM, interfacial measurement, etc)

Updated August 2009


0.0

0.1

0.2

0.3

1000 1100 1200 1300 1400 1500 1600

Irradiation Temperature (˚C)

Voi

d S

wel

ling,

Svo

id(%

)

~10dpa

~6dpa

5.2dpa

~2dpa

This work ~2dpa

~6dpa

~10d

paPrice (1973)

7.4dpa

1460˚C

1400˚C

1300˚C1100˚C

30nm

• Void swelling in SiC at very high temperatures were demonstrated with quantitatively reliable data.

• Lowest irradiation temperature for void formation was ~1050˚C.

• However, cavity swelling was negligible below ~1300˚C.

• Above ~1400˚C, the volume fraction of the cavities increases dramatically and fluence dependent swelling becomes notable.

Void Swelling of High Purity SiC Irradiated with Neutron

Kondo et al, 2009, J.Nucl.Mater. 386-388


Rupture discassembly

Aluminum liner

SS housing tube

800ºC Subcapsule

1100ºC Subcapsule

1300ºC Subcapsule

HFIR RB-18J Very High Temperature Irradiation Experiment for Advanced SiC/SiC Composites

• 3 temperature zones; 800C, 1100C, and 1300C from top to bottom• SS sleeves, Inconel couplings, graphite holders, platinum-

sheathed TC’s


HFIR-18J Very High Temperature Irradiation Experiment

• Demonstrated no or very minor detrimental irradiation effect up to ~7 dpa at 800 to 1300ºC on:– In-plane tensile strength– Interlaminar shear strength– Trans-thickness tensile

strength– Fracture toughness

• Confirmed saturation at ~800ºC of– Swelling– Thermal conductivity

degradation• Demonstrated comparable

irradiation stability of NITE SiC/SiC with CVI

Ozawa et al, to be presented at ICFRM-14


β-SiC(Rohm and Hass )･･･Monolithic SiC fabricated by CVD；disk φ3mm，thickness： 1mm

He filled and sealed quartz tubeStagnant He conditions

Specimen Capsule

Solid Breeding Materials (Supplied by TYK ,Japan)･･･Sintered Pellets，3mm disk，Thickness : 1mm

Li1.6TiO3Li1.8TiO3Li1.9TiO3Li2TiO3

Li1.6ZrO3Li1.8ZrO3Li1.9ZrO3Li2ZrO3

Li3.2SiO4Li3.6SiO4Li3.8SiO4Li4SiO4

Li0.8SiO4Li0.9AlO4Li0.95AlO2LiAlO2

④

③

②

①

Specimen capsule for compatibility studies

Tungsten springAlumina Screw

Alumina ball

SiC Solid Breeder

Spacer (Inconel)

Inconel tube

Heating conditions (Unirradiated)800,900,1000C for 10h and 100h

Same materials will be irradiated inHFIR 18J at 800C.

Compatibility Study of SiC and Solid Breeding MaterialsUnirradiated Experiments (Hasegawa et al)


FCI in US DCLL Blanket Module

• FCI is a key feature that:– Makes DCLL concept attractive for DEMO

and power reactors.– Is employed also in EU PPCS Model C

and other conceptual blanket designs.

• Two important FCI functions:– Thermally insulate Pb-Li so that the Pb-Li

temperature can be considerably higher (~700ºC) than the maximum interface temperature for steel structures (470ºC).

– Electrically insulate Pb-Li flow from steel structures.

FSG

AP

FCI

Pb-L

i

100 S/m

20 S/m

σFCI = 5 S/m

Temperature Profile for Model DEMO Case

(Smolentsev)


FCI-related Potential Technical Issues for DCLL Blanket as Summarized in IEA SiC/SiC Workshop, Jan-2009, Daytona Beach (Sharafat and Katoh)

• Material Issues:– Thermal insulation– Electrical insulation– Mechanical integrity– Effect of irradiation, flow condition,

electric and magnetic field, and transient loading events on all of above

• Materials System Issues:– Chemical compatibility– Leak tightness

• Blanket Design Issues:– Numerous design options yet to be

optimized for specific blankets– Non-uniform heating (impact on stress

state, deformation, …) • Operation Issues:

– Startup and shut down (freezing and un-freezing, un-even warm up ,drainage, …)

– Impact of flow ( vibration, mechanical loads, chemical compatibility in flow, …)

FSG

AP

FCI

Pb-L

i

100 S/m

20 S/m

σFCI = 5 S/m

Temperature Profile for Model DEMO Case

(Smolentsev)


FCI Material Property RequirementsFCI Material Property Requirements

Effect of the electrical and thermal conductivity of the FCI on heat losses in

the DEMO blanket

S. Smolentsev et al. / Fusion Engineering and Design 83 (2008) 1788–179114

Effect of SiC electrical conductivity and FCI thickness on MHD pressure drop reduction

factor R in poloidal flow for DEMO.


Thermal Insulation Properties• Constitutive modeling allows prediction of thermal conductivity for specific

architectures or designing composites for target thermal conductivity.

• Thermal conductivity changes in radiation environment can be predicted to a reasonable accuracy with current radiation effect and constitutive composite models.

• Fundamental materials science yet to be established.

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10 W/m-K

5 W/m-K

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2 W/m-K

2D HNLS/CVI High Density

Enhanced thermal insulation may be enabled through engineered porosity- Porous SiC inner layer, e.g. foam- Continuous fiber composites witharchitectural porosity

Katoh & Snead, 2009, Fusion Sci&Tech 56


Electrical Insulation Properties• Constitutive modeling allows prediction of electrical conductivity for specific

architectures or designing composites for target electrical conductivity, with known SiC conductivity.

• “Pure” SiC is an impurity semi-conductor. Electrical resistivity, often determined by nitrogen content, is a strong function of temperature.

• Irradiated SiC is a totally different semi-conductor from unirradiated SiC. Irradiation effect presently has to rely on experimental data.

• Transmutation effect in fusion spectrum is unknown and currently unpredictable.

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0

1000/T[K]

Elec

trica

l Con

duct

ivity

[S/m

]

2D SiC/SiC, Thru-Thickness

CVD SiC High Resistivity Grade

2D SiC/SiC, Over-CoatedThru-Thickness

2D SiC/SiC, In-Plane

Graphite (x1/1000)

800 600 400 200 100 20

Temperature [ºC]

PyCSiC

1/T

log(σ)

Composite

PyCSiC

1/T

log(σ)

Composite

PyC

SiC

1/T

log(σ)Compositew/SiC Overcoat

SiC Overcoat

Katoh et al, 2009, J.Nucl.Mater. 386-388


Mechanical Integrity• SiC/SiC strength does not degrade in radiation environment at temperatures of interest

for FCI application.

• Main mechanical integrity issue is secondary stress imposed by temperature gradient and differential swelling.

• Key unknowns include irradiation creep and performance of micro-cracked FCI.

• Various design efforts may be considered to minimize the general and local secondary stresses and temperature inhomogeneity.

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Dose (DPA)

Irra

dia

tion

Cre

ep B

SR

Rat

io

CVD-SiC (HFIR, Katoh)

CVD-SiC (JMTR, Katoh)

CVD-SiC (ETR, Price, 1977)

3C-SiC (HFIR, Katoh)

400-1030C

640C

1080C

1080C

780C

950C

1120C

<1000ºC

>1000ºC

Katoh et al, 2007, J.Nucl.Mater. 367-370 Shinavski, presented at IEA SiC/SiC WS, Jan-2009


Irradiation Creep Study in Current Program (TITAN)

• Objectives: transient creep of SiC ceramics and composites– Stress exponent– Swelling – creep coupling– Interfacial shear creep– T/He effect– Model composites

• Initial campaign:– Intended to determine n and swelling

coupling – Preferred irradiation conditions:

• 300 - 1200ºC, 0.01 - 1 dpa– 3 stress levels (by thickness

variation)– Reasonable materials variation

preferred:• CVD, single-X’tal, NITE-matrix, fibers

– Both in pre-strained and self-loading BSR configurations

Moly “coffin”

DSL unit

Pre-strained BSR unit

CVD SiC separators/liners


Chemical Compatibility• SiC – PbLi compatibility in static environment has been positively addressed.

• Practical operating environment imposes more complex issues– Flow, temperature gradient– Electrical field/current across interface– Mass transport among other materials

exposed to liquid

Pint

Konishi & Yamamoto


Pint, to be published


Concluding Remarks

• Steady progress is being achieved on SiC/SiC R&D in US fusion materials program– Current main focuses on fundamental/intrinsic properties and

environmental (radiation) effect issues– Provides scientific basis to design activities

• Feasibility, intrinsic limitations, design parameter space• Test blanket modules, FNST facility, DEMO blanket

• Increasing importance of interactions with opportunities in fission energy arena is recognized– Near-term opportunity in HTGR/VHTR

• Primary focus on effort toward licensing

– Other opportunities mid-term to long-term

• For specific applications, materials technology program is needed.

status and prospect of sic and fci r&d...2009/08/20 · status and prospect of sic and fci...

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