1

Breeding Blanket Concepts for Fusion and Materials Requirements

A. R. Raffray1, M. Akiba2, V. Chuyanov3,

L. Giancarli4, S. Malang5

1University of California, San Diego, 458 EBU-II, La Jolla, CA 92093-0417, USA2Blanket Engineering Laboratory, JAERI, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan

3ITER Garching Joint Work Site, Boltzmannstr. 2, Max-Planck-Institut für Plasmaphysik, 85748 Garching, Germany4CEA-Saclay, DEN/CPT, 91191, Gif-sur-Yvette, France.

5Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe, Germany

Plenary Presentation at ICFRM-10

Baden Baden, Germany

October 15-19, 2001

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Outline of Presentation

• Highlight key performance and attractiveness parameters for breeding blanket concepts

• Summarize range of blanket concepts being currently considered – Focus on MFE (although significant IFE/MFE synergies)

– Differentiate between class of concepts (material basis)

– Example description for each class of blanket concepts

– Highlight key material-related issues associated with each class of blanket concepts.

• Evolve an example classification of concepts based on the level of attractiveness and the development risk associated

• Propose a strategy for blanket development and supporting material R&D

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Performance and Attractiveness Factors Guide the Blanket Design Process

• Design and material impact on all these factors

• Require close interaction between design and material communities

• Electrical power production – Neutron energy multiplier

– Power cycle efficiency

• Safety – Short-term activation: possibility of

passive accommodation of off-normal scenarios without major consequences

– Long-term activation: waste disposal requirements

• Availability – Commercial reactors would require high

availability

– Key parameters: reliability, lifetime and downtime of the blanket system

• Design and Fabrication– Simplicity

• Tritium – Need for tritium self-sufficiency from

blanket tritium breeding

– Total tritium inventory in blanket and tritium permeation

• Economics: – Ultimate economic measure: COE

– Affected by all other factors

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500 700 900 1100 1300

Cycle He Maximum Temperature (°C)

E.g. Why High Temp. Material?

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A Number of Different Breeding Blanket Concepts Have Been Considered Recently

• Focus on solid wall concepts but material issues also applicable to thin liquid film concepts

• Blanket concepts can be classified based on the structural and breeder materials– Ceramic Breeder + Ferritic/Martensitic Steel Concepts (water-cooled and He-cooled)

– Pb-17Li + Ferritic/Martensitic Steel Concepts (self-cooled and water-cooled)

– Self-Cooled Lithium + Vanadium Alloy Concepts

– SiCf/SiC Based Concepts (self-cooled Pb-17Li and He-cooled ceramic breeder)

– Other concepts (less studied)• Advanced concepts• FLiBe

• For each class of concept, an example blanket design is described and key material issues discussed

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Ceramic Breeder + Be and Ferritic/Martensitic Steel Concepts Have Been Considered with Water and He as Coolant

• Generally good compatibility among CB, structural material and coolant

• No MHD effects

• CB and Be in form of sintered blocks or pebble beds

• Candidate breeder materials: – Ternary CB (Li4SiO4, Li2TiO3, Li2ZrO3)

– Li2O

Struc. Tmax=550°C

Cool. Tmax=320°C

Cool. P =15 MPa

Cycle Eff. =35%

Energy Multip. =1.3

Lifetime>10MW-a/m2

Example Water-Cooled Concept: SSTR (JAERI)

• Modular design

• Use of reduced activation F82H steel

• Binary Be & CB pebble beds – Maximize keff and T breeding

performance

• PWR-like water conditions

• Possible use of supercritical-pressure water

– 500°C/25 MPa

– Increase cycle efficiency to ~40-45 %

– Results in high stresses - use of advanced FS, such as Dispersion Strengthened FS, might be required

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• Chemical compatibility between Be and water/air (hydrogen production)– Water-cooled concept

– Accidental water/air ingress

– Possible solution to reduce H2 production: Use of Be12Ti, which has better compatibility with water.

• Tritium inventory in blanket and permeation to coolant

• Thermo-mechanical interactions among pebbles and between

pebbles and structure including neutron irradiation effects

• Limits on blanket lifetime due to irradiation damages in ceramic breeder and beryllium

• Fabrication and re-processing of the ceramic breeder– Cost and waste considerations indicate need for re-processing of replaced blanket

modules and re-use of the Li

Key Material Issues of Ceramic Breeder Concepts

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Pb-17Li + Ferritic/Martensitic Steel Concepts Include Self-Cooled and Water-Cooled Options

Example Dual Coolant Concept: FZK DC • Uncouple FW cooling from blanket cooling

– He coolant for more demanding FW cooling (no MHD uncertainties)

– Self-cooled Pb-17Li with SiCf/SiC flow channel insulating inserts for blanket region

• Use of ODS-steels would allow for higher temperature but more demanding welding requirements

– Compromise: ferritic steel structure with ~mm’s ODS layer at higher temperature FW location

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1

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2Pb-17Li

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1

He System

Shield

SiCf/SiCChannelInserts

EUROFER Structure(FW+Grids)

ODS LayersPlatedto the FW

• Pb-17Li is an attractive breeder material – Good tritium breeding capability

– Possibility to replenish 6Li on-line

– Almost inert in air and relatively mild reaction with water

– In general limited extrapolation of blanket technology

• Simplest FMS and Pb-17Li concept is a self-cooled configuration (ARIES-ST and FZK DC concepts)

• Water-cooled option to avoid MHD effects (WCLL concept)

Struc. Tmax=550°C

Pb-17Li Tmax=700°C

He Cool. Tmax/P =480°C/14 MPa

Cycle Eff. =45%

Energy Multip. =1.17

Lifetime =15MW-a/m2

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Key Material Issues for Pb-17Li + Ferritic/Martensitic Steel Concepts

• Performance of this concept is limited by :– Maximum allowable FW temperature

– Structural material compatibility with Pb-17Li (~ 500 C)

– Possible development of advanced ODS steel to allow for higher temperature

• Fabrication/joining

• For water-cooled concept, need to limit tritium permeation from Pb-17Li to water

– Development of permeation barriers

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Self-Cooled Lithium and Vanadium Alloy Concept

Example Li +V Concept: ARIES-RS• Simple box-like structure

• Insulating coating on V alloy to reduce MHD effects – E.g. CaO maintained by adding 0.5% Ca in Li

– Allows for low system pressure

• Multiple flow passes in the blanket provide the capability for FW surface heat flux ~1 MW/m2

• Lithium provides the advantages of:– High tritium breeding capability,

– High thermal conductivity,

– Immunity to irradiation damage– Possibility of unlimited lifetime if 6Li burn-up can be

replenished

• Vanadium alloys offer advantages of:

– Low after heat

– High temperature and high heat flux capability.

– Compatibility with liquid Li

Struc. Tmax=700°C

Cool. Tmax=610°C

Cool. P < 1MPa

Cycle Eff. =46%

Energy Multip. =1.21

Lifetime =15MW-a/m2

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Key Material Issues for Lithium and Vanadium Alloy Concept

• Insulated walls are a must to reduce MHD effects for acceptable pressure drop and heat transfer

MAJOR ISSUE– Need to develop reliable and self-healing insulation coating

– Compatible with low activation requirements, material interfaces and tritium recovery systems in a fusion environment

• Chemical reactivity of Li

• Radiation damage effect on V-alloy from fusion neutron spectrum

• Fabrication method and cost of V-alloy under the goal of minimizing impurities

• Joining of V-alloy to another structural material

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Concepts Utilizing SiCf/SiC Composite as Structural Material

Example Self Cooled Pb-17Li + SiCf/SiC Concept: ARIES-AT• Simple box design geometry• Utilizes 2 cooling passes to uncouple structure temperature from outlet coolant temperature• Reasonable design margins as an indication of reliability

SiCf/SiC offers advantages of:• Safety

- Low afterheat: Possibility of passive accommodation of accident scenarios (E.g. LOCA, LOFA)

- Low long term activation favorably influences waste disposal requirement (Class C or better)

• Performance - High temperature operation: possibility of high cycle

SiCf/SiC has been considered with:

• Self-cooled Pb-17Li (e.g. TAURO and ARIES-AT) • He-cooled CB (ARIES-I, DREAM, A-HCPB, and A-SSTR2)

Struc. Tmax=1000°C

Cool. Tmax=1100°C

Cool. P =1-1.5 MPa

Cycle Eff. =59%

Energy Multip. =1.11

Lifetime(?) ~18MW-a/m2

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Key Material Issues for Fusion Blanket Concepts with SiCf/SiC Composite

• Major issue linked to uncertainty about SiCf/SiC behavior and performance at high temperature and under irradiation

- Thermal conductivity - Maximum allowable operating temperature- Neutron irradiation effect/Lifetime- Allowable stress- Hermeticity (in particular for He-cooled concept)- Fabrication

• Key issues associated with the breeders are similar to those for other concepts utilizing similar breeding materials

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Other Concepts Receiving Some Degree of Attention but Generally Less Studied Include:

Advanced concepts, e.g.• EVOLVE concept with Li evaporation cooling and W alloy

- Make use of high latent heat of evaporation of Li

- Operation at Li saturation temperature of 1200°C and low pressure

- High heat load capable - FW q’’ > 2 MW/m2 and neutron wall load >10 MW/m2

- Key material issue: qualification of W alloy under these conditions

Struc. Tmax=1300°C

Cool. Tmax=1200°C

Cool. P =0.04 MPa

Cycle Eff. >55%

Energy Multip. =1.2

Lifetime(?)

Struc. Tmax=550°C

Cool. Tmax=550°C

Cool. P =0.5 MPa

Cycle Eff. =38%

Lifetime =15MW-a/m2

FLiBe concepts with Ferritic or Other Structural Materials:

• Advantages of FLiBe

- Low chemical activity with air and water

- No MHD effect on pressure drop for self-cooled concepts

- Compatible with most structural materials to high temperature

• Issues

- Material compatibility issues with transmutation products

- Poor heat transfer: Possible MHD effects on heat transfer; need turbulence

- Tritium control

• Limited effort on conceptual design studies for FLiBe concepts in MFE reactors

- e.g. FFHR-2 helical-type reactor, with FLiBe + FS blanket concept

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Substantial Progress Has Been Made in Understanding the Performance and Behavior Of Breeder, Multiplier and

Structural Materials Over the Last 10-15 Years

Example Accomplishment: Tritium inventory predictions for ceramic breeders

• Tritium transport in CB is complex process involving several mechanisms

• Overly conservative T inventory estimate in early studies in the absence

of adequate analytical tools and fundamental property data characterization

• Focused international R&D program including material fabrication and

characterization, laboratory and in-reactor purge experiments, and

fundamental and integrated model development

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Tritium Inventory in Ceramic Breeder (g)

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STARFIRE

BCSS/Li

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O

BCSS/LiAlO

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NET/KFK/Li

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SiO4

ITER-CDA like/Li

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SiO4

ITER-CDA/Li

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O

PROMETHEUS/Li

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O

ARIES-1/Li

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ZrO3

KFK Update/Li

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SiO4

EU BIT/LiAlO2

EU BOT/Li

4

SiO4

ITER-EDA like/Li2

ZrO3

ITER-EDA /Li2

ZrO3

ITER-EDA, 2 /Li2

ZrO3

• Intended as a qualitative illustration

• Specific inventory dependent on parameters which may differ from study to study,

• Magnitude of CB tritium inventory prediction in blankets has decreased dramatically over the years as a direct result of R&D effort

• For such classes of blanket, T inventory in other components and materials (such as in Be) are nowconsidered much more problematic

Chronology of T Inventory Estimates for CB Blankets from Different Studies

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However, Many Issues Still Remain for Each Class of Blanket Concept

• Blanket concepts with the potential for high performance and attractive safety features tend to have higher associated development risk

• Semi-qualitative subjective classification of different blanket concepts

– Attractiveness measure is a subjective assemblage of attractive features, such as cycle efficiency, safety, and lifetime

– Measure of development risk includes degree of extrapolation from current material and component performance and complexity and scale of effort required to validate the concept

– Individual classification could change over a certain range according to the reviewer

– However, the overall classification of choice of structural materials and of breeder materials is unlikely to vary appreciably

Example classification of blanket concepts based

on attractiveness and development risk

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Proposed Blanket R&D Strategy (1)

• The objective of blanket R&D should be to lead at least to the development of a blanket with a minimum acceptable level of performance and attractiveness for a commercial reactor

– Lower-bound blanket performance levels which would still result in an acceptably attractive commercial fusion reactor should be developed

• Blanket concepts with the lowest development risk meeting these performance criteria should be developed and tested

– Typically medium-risk medium-performance concepts• Representing a fall-back position• Provide a reference scale to judge more advanced concepts.

– Can be tested in ITER

• First time operation of full blanket systems in a fusion environment.

• These tests will give important information on various aspects of blanket component functions

• But not sufficient (in particular for lifetime) and other approaches would also be required

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Proposed Blanket R&D Strategy (2)

• In parallel, critical R&D should be done for the more advanced, higher risk but higher pay off concepts

– This R&D is very much material-related

– Promises offered by the performance of advanced concepts provides a challenge to the material R&D community to help the blanket design to achieve this

• Design effort on advanced blanket conceptual design should also be pursued to help guide the material R&D towards high performance material and to provide a vision and a goal for attractive concepts for the future

This requires close interaction and coordination between material and design communities, for example, through:

– Cross meeting participation, organization of focused workshop• e.g. International town meeting on SiCf/SiC at Oak Ridge in Jan. 2000