1

Fusion Nuclear Technology:  Principles and Challenges

Mohamed Abdou (web: http://www.fusion.ucla.edu/abdou/)Distinguished Professor of Engineering and Applied Science

Director, Center for Energy Science and Technology (CESTAR)(http://www.cestar.seas.ucla.edu/)

Director, Fusion Science and Technology Center (http://www.fusion.ucla.edu/)University of California, Los Angeles (UCLA)

Seminar at Grenoble, France March 23, 2007

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Incentives for Developing FusionIncentives for Developing Fusion

Fusion powers the Sun and the stars

It is now within reach for use on Earth

In the fusion process lighter elements are “fused” together, making heavier elements and producing prodigious amounts of energy

Fusion offers very attractive features:

Sustainable energy source

(for DT cycle; provided that Breeding Blankets are successfully developed)

No emission of Greenhouse or other polluting gases

No risk of a severe accident

No long-lived radioactive waste

Fusion energy can be used to produce electricity and hydrogen, and for desalination

3

ITER

• The World is about to construct the next step in fusion development, a device called ITER

• ITER will demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes

• ITER will produce 500 MW of fusion power

• Cost, including R&D, is 15 billion dollars

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ITER Design - Main Features

Divertor

Central Solenoid

Outer Intercoil Structure

Toroidal Field Coil

Poloidal Field Coil

Machine Gravity Supports

Blanket Module

Vacuum Vessel

Cryostat

Port Plug (IC Heating)

Torus Cryopump

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ITER is a collaborative effort among Europe, Japan, US, Russia, China and South Korea

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Fusion Nuclear Technology (FNT)

FNT Components from the edge of the Plasma to TF Coils (Reactor “Core”)

1. Blanket Components

2. Plasma Interactive and High Heat Flux Components

3. Vacuum Vessel & Shield Components

4. Tritium Processing Systems

5. Instrumentation and Control Systems

6. Remote Maintenance Components

7. Heat Transport and Power Conversion Systems

a. divertor, limiter

b. rf antennas, launchers, wave guides, etc.

Other Components affected by the Nuclear Environment

Fusion Power & Fuel Cycle Technology

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The Deuterium-Tritium (D-T) Cycle

• World Program is focused on the D-T cycle (easiest to ignite):

D + T → n + α + 17.58 MeV

• The fusion energy (17.58 MeV per reaction) appears as Kinetic Energy of neutrons (14.06 MeV) and alphas (3.52 MeV)

• Tritium does not exist in nature! Decay half-life is 12.3 years

(Tritium must be generated inside the fusion system to have a sustainable fuel cycle)

• The only possibility to adequately breed tritium is through neutron interactions with lithium– Lithium, in some form, must be used in the fusion system

8

Tritium Breeding

Natural lithium contains 7.42% 6Li and 92.58% 7Li.

The 7Li(n;n’)t reaction is a threshold reaction and requires an incident neutron energy in excess of 2.8 MeV.0.01

0.1

1

10

100

1000

1 10 100 1000 104 105 106 107

Li-6(n,alpha)t and Li-7(n,n,alpha)t Cross-Section

Li-6(n,a) tLi-7(n,na)t

Neutron Energy (eV)

MeVntnLi

MeVtnLi

47.2

78.47

6

6Li (n,) t

7Li (n;n’) t

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Blanket (including first wall)Blanket Functions:

A. Power Extraction– Convert kinetic energy of neutrons and secondary gamma rays into heat

– Absorb plasma radiation on the first wall

– Extract the heat (at high temperature, for energy conversion)

B. Tritium Breeding– Tritium breeding, extraction, and control

– Must have lithium in some form for tritium breeding

C. Physical Boundary for the Plasma– Physical boundary surrounding the plasma, inside the vacuum vessel

– Provide access for plasma heating, fueling

– Must be compatible with plasma operation

– Innovative blanket concepts can improve plasma stability and confinement

D. Radiation Shielding of the Vacuum Vessel

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Blanket Concepts(many concepts proposed worldwide)

A. Solid Breeder Concepts– Always separately cooled

– Solid Breeder: Lithium Ceramic (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3)

– Coolant: Helium or Water

B. Liquid Breeder ConceptsLiquid breeder can be:

a) Liquid metal (high conductivity, low Pr): Li, or 83Pb 17Li

b) Molten salt (low conductivity, high Pr): Flibe (LiF)n · (BeF2), Flinabe (LiF-BeF2-NaF)

B.1. Self-Cooled

– Liquid breeder is circulated at high enough speed to also serve as coolant

B.2. Separately Cooled

– A separate coolant is used (e.g., helium)

– The breeder is circulated only at low speed for tritium extraction

B.3. Dual Coolant

– FW and structure are cooled with separate coolant (He)

– Breeding zone is self-cooled

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A Helium-Cooled Li-Ceramic Breeder Concept: Example

Material Functions•Beryllium (pebble bed) for neutron multiplication

•Ceramic breeder (Li4SiO4, Li2TiO3, Li2O, etc.) for tritium breeding

•Helium purge (low pressure) to remove tritium through the “interconnected porosity” in ceramic breeder

•High pressure Helium cooling in structure (ferritic steel)Several configurations exist (e.g. wall parallel or “head on” breeder/Be arrangements)

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phase I ARIES-CS HCCBBreeder Unit to be inserted into the space between the grid plates

EU HCPB DEMO Module

First Wall(RAFS, F82H)

Tritium BreederLi2TiO3, Li2O (<2mm)

Neutron MultiplierBe, Be12Ti (<2mm)

JA HCSB Module

ODS +FS

Examples of solid breeder blankets that utilize immobile solid lithium Examples of solid breeder blankets that utilize immobile solid lithium ceramic breeder and Be multiplier for tritium self-sufficiencyceramic breeder and Be multiplier for tritium self-sufficiency

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Solid Breeder Blanket Issues

Tritium self-sufficiency Breeder/Multiplier/structure interactive effects under

nuclear heating and irradiation Tritium inventory, recovery and control; development

of tritium permeation barriers Effective thermal conductivity, interface thermal

conductance, thermal control Allowable operating temperature window for breeder Failure modes, effects, and rates Mass transfer Temperature limits for structural materials and coolants Mechanical loads caused by major plasma disruption Response to off-normal conditions

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Liquid Breeders

• Many liquid breeder concepts exist, all of which have key feasibility issues. Selection can not prudently be made before additional R&D results become available.

• Type of Liquid Breeder: Two different classes of materials with markedly different issues.

a) Liquid Metal: Li, 83Pb 17Li

High conductivity, low Pr number

Dominant issues: MHD, chemical reactivity for Li, tritium permeation for LiPb

b) Molten Salt: Flibe (LiF)n · (BeF2), Flinabe (LiF-BeF2-NaF)

Low conductivity, high Pr number

Dominant Issues: Melting point, chemistry, tritium control

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Liquid Breeder Blanket Concepts

1. Self-Cooled– Liquid breeder circulated at high speed to serve as coolant

– Concepts: Li/V, Flibe/advanced ferritic, flinabe/FS

2. Separately Cooled– A separate coolant, typically helium, is used. The breeder is

circulated at low speed for tritium extraction.

– Concepts: LiPb/He/FS, Li/He/FS

3. Dual Coolant– First Wall (highest heat flux region) and structure are cooled

with a separate coolant (helium). The idea is to keep the temperature of the structure (ferritic steel) below 550ºC, and the interface temperature below 480ºC.

– The liquid breeder is self-cooled; i.e., in the breeder region, the liquid serves as breeder and coolant. The temperature of the breeder can be kept higher than the structure temperature through design, leading to higher thermal efficiency.

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EU – The Helium-Cooled Lead Lithium (HCLL) DEMO Blanket Concept

Module box (container & surface heat flux extraction)

Breeder cooling unit (heat extraction from PbLi)

Stiffening structure (resistance to accidental in-box pressurization i.e He leakage) He collector system

(back)

[0.5-1.5] mm/s[18-54]

mm/s

HCLL PbLi flow scheme

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Li/Vanadium Blanket Concept

Breeding Zone

(Li flow)

Primary Shield

Secondary Shield

Reflector

Li

Li

Li

Secondary shield(B4C)

Primary shield

(Tenelon)

Reflector

Lithium

Lithium

Vanadium structure

Vanadium structure

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Pathway Toward Higher Temperature Through Innovative Designs with Current Structural Material (Ferritic Steel):

Dual Coolant Lead-Lithium (DCLL) FW/Blanket Concept

First wall and ferritic steel structure cooled with helium

Breeding zone is self-cooled Structure and Breeding zone are

separated by SiCf/SiC composite flow channel inserts (FCIs) that Provide thermal insulation to

decouple PbLi bulk flow temperature from ferritic steel wall

Provide electrical insulation to reduce MHD pressure drop in the flowing breeding zone

Self-cooled Pb-17Li

Breeding Zone

He-cooled steel

structure

SiC FCI

DCLL Typical Unit Cell

Pb-17Li exit temperature can be significantly higher than the operating temperature of the steel structure High Efficiency

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US DCLL DEMO Blanket Module

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Flows of electrically conducting coolants will experience complicated

magnetohydrodynamic (MHD) effects

What is magnetohydrodynamics (MHD)?– Motion of a conductor in a magnetic field produces an EMF that can

induce current in the liquid. This must be added to Ohm’s law:

– Any induced current in the liquid results in an additional body force in the liquid that usually opposes the motion. This body force must be included in the Navier-Stokes equation of motion:

– For liquid metal coolant, this body force can have dramatic impact on the flow: e.g. enormous MHD drag, highly distorted velocity profiles, non-uniform flow distribution, modified or suppressed turbulent fluctuations

)( BVEj

BjgVVVV

11

)( 2pt

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Main Issue for Flowing Liquid Metal in Blankets: MHD Pressure Drop

Feasibility issue – Lorentz force resulting from LM motion across the magnetic field generates MHD retarding force that is very high for electrically conducting ducts and complex geometry flow elements

c

wwMHD a

tVBLLJBp

2

p, pressureL, flow lengthJ, current densityB, magnetic inductionV, velocity, conductivity (LM or wall)a,t, duct size, wall thickness

Thin wall MHD pressure drop formula

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Inboard is the critical limiting region for LM blankets– B is very high! 10-12T

– L is fixed to reactor height by poor access

– a is fixed by allowable shielding size

– Tmax is fixed by material limits

LTcLaPm pNWL

wtpaS /

Combining Power balance formula

With Pipe wall stress formula

With thin wall MHD pressure drop formula (previous slide) gives:

Tca

BLPS

p

w

22

NWL

Pipe stress is INDEPENDENT of wall thickness to first orderand highly constrained by reactor size and power!

(Sze, 1992)

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0

50

100

150

200

250

300

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

Inboard Field, B (T)P

ipe W

all

Str

ess,

S (

MP

a)

No pipe stress window for inboard blanket operation for Self-Cooled LM blankets (e.g. bare wall Li/V)

(even with aggressive assumptions) U ~ 0.16 m/sPmax ~ 5-10 MPa

Best Possible DEMO Base Case for bare wall Li/V:

NWL = 2.5 MW/m2

L = 8 m, a = 20 cm

T = 300K

ITER

Pipe stress >200 MPa will result just to remove nuclear heat

Higher stress values will result when one considers the real effects of:– 3D features like flow

distribution and collection manifolds

– First wall cooling likely requiring V ~ 1 m/s

Allowable

Marginal

Unacceptable

ARIES-RS

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-1 -0.8 -0.6 -0.4 -0.2 0 0.2 0.4 0.6 0.8 1

-1

-0.8

-0.6

-0.4

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0.2

0.4

0.6

0.8

1

Large MHD drag results in large MHD pressure drop

• Net JxB body force p = cVB2 where c = (tw w)/(a )

• For high magnetic field and high speed (self-cooled LM concepts in inboard region) the pressure drop is large

• The resulting stresses on the wall exceed the allowable stress for candidate structural materials

• Perfect insulators make the net MHD body force zero

• But insulator coating crack tolerance is very low (~10-7).

– It appears impossible to develop practical insulators under fusion environment conditions with large temperature, stress, and radiation gradients

• Self-healing coatings have been proposed but none has yet been found (research is on-going)

Lines of current enter the low resistance wall – leads to very high induced current and high pressure drop

All current must close in the liquid near the wall – net drag

from jxB force is zero

Conducting walls Insulated walls

-1 -0.8 -0.6 -0.4 -0.2 0 0.2 0.4 0.6 0.8 1

-1

-0.8

-0.6

-0.4

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1

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Issues with the Lithium/Vanadium Concept

• Li/V was the U.S. choice for a long time, because of its perceived simplicity. But negative R&D results and lack of progress on serious feasibility issues have eliminated U.S. interest in this concept as a near-term option.

Conducting walls

Insulating layer

Electric currents lines

Leakage

current Crack

Issues

• Insulator– Insulator coating is required

– Crack tolerance (10-7) appears too low to be achievable in the fusion environment

– “Self-healing” coatings can solve the problem, but none has yet been found (research is ongoing)

• Corrosion at high temperature (coupled to coating development)

– Existing compatibility data are limited to maximum temperature of 550ºC and do not support the BCSS reported corrosion limit of 5m/year at 650ºC

• Tritium recovery and control• Vanadium alloy development is very costly and requires a very long time to

complete

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Proposed US DCLL TBM Cutaway

PbLi Flow Channels

He-cooledFirst Wall

PbLi

He

He

SiC FCI

2 mm gap

US DCLL TBM – Cutaway Views

484 mm

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Flow Channel Inserts are a critical element of the high outlet temperature DCLL

FCIs are roughly box channel shapes made from some material with low electrical and thermal conductivity

– SiC/SiC composites and SiC foams are primary candidate materials

They will slip inside the He Cooled RAFS structure, but not be rigidly attached

They will slip fit over each other, but not be rigidly attached or sealed

FCIs may have a thin slot or holes in one wall to allow better pressure equalization between the PbLi in the main flow and in the gap region

FCIs in front channels, back channels, and access pipes will be subjected to different thermal and pressure conditions; and will likely have different designs and thermal and electrical property optimization

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Example: Interaction between MHD flow and FCI behavior are highly coupled and require fusion environment

PbLi flow is strongly influenced by MHD interaction with plasma confinement field and buoyancy-driven convection driven by spatially non-uniform volumetric nuclear heating

Temperature and thermal stress of SiC FCI are determined by this MHD flow and convective heat transport processes

Deformation and cracking of the FCI depend on FCI temperature and thermal stress coupled with early-life radiation damage effects in ceramics

Cracking and movement of the FCIs will strongly influence MHD flow behavior by opening up new conduction paths that change electric current profiles

Simulation of 2D MHD turbulence in PbLi flow

FCI temperature, stress and deformation

Similarly, coupled phenomena in tritium permeation, corrosion, ceramic breeder

thermomechanics, and many other blanket and material behaviors

Abdou Lecture 4

Key DCLL parameters in three blanket scenarios

Parameter DEMO ITER H-H ITER D-TSurface heat flux, Mw/m2 0.55 0.3 0.3

Neutron wall load, Mw/m2 3.08 - 0.78

PbLi In/Out T, C 500/700 470/~450 360/470

2a x 2b x L, m 0.22x0.22x2 0.066x0.12x1.6 0.066x0.12x1.6

PbLi velocity, m/s 0.06 0.1 0.1

Magnetic field, T 4 4 4

Re 61,000 30,500

Ha 11,640 6350

Gr 3.521012 7.22109

*Velocity, dimensions, and dimensionless parameters are for the poloidal flow. DEMO parameters are for the outboard blanket.

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MHD / Heat Transfer analysis (DEMO)

• Parametric analysis was performed for poloidal flows to access FCI effectiveness as electric/thermal insulator

SiC=1-500 S/m, kSiC=2-20 W/m-K

• Strong effect of SiC on the temperature field exists via changes in the velocity profile

• FCI properties were preliminary identified: SiC~100 S/m, kSiC~2 W/m-K

FS

GA

PF

CI

Pb

-Li

100 S/m

20 S/m

FCI = 5 S/m

Temperature Profile for Model DEMO Case

kFCI = 2 W/m-K

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Molten Salt Blanket Concepts

• Lithium-containing molten salts are used as the coolant for the Molten Salt Reactor Experiment (MSRE)

• Examples of molten salt are:

– Flibe: (LiF)n · (BeF2)

– Flinabe: (LiF-BeF2-NaF)

• The melting point for flibe is high (460ºC for n = 2, 380ºC for n = 1)

• Flinabe has a lower melting point (recent measurement at SNL gives about 300ºC)

• Flibe has low electrical conductivity, low thermal conductivity

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Molten Salt Concepts: Advantages and IssuesAdvantages

• Very low pressure operation

• Very low tritium solubility

• Low MHD interaction

• Relatively inert with air and water

• Pure material compatible with many structural materials

• Relatively low thermal conductivity allows dual coolant concept (high thermal efficiency) without the use of flow-channel inserts

Disadvantages

• High melting temperature

• Need additional Be for tritium breeding

• Transmutation products may cause high corrosion

• Low tritium solubility means high tritium partial pressure (tritium control problem)

• Limited heat removal capability, unless operating at high Re (not an issue for dual-coolant concepts)

Abdou Lecture 4

Summary of MHD issues of liquid blankets

MHD issue Self-cooled LM blanket

Dual-coolant LM blanket

He-cooled LM blanket

MS blanket

MHD pressure drop and flow distribution

Very important Important Inlet manifold ? Not important

Electrical insulation

Critical issue Required for IB blanket

Not needed ? Not needed

MHD turbulence

Cannot be ignored

Important.

2-D turbulence

Not applicable Critical issue (self-cooled)

Buoyancy effects

Cannot be ignored

Important May effect tritium transport

Have not been addressed

MHD effects on heat transfer

Cannot be ignored

Important Not important Very important (self-cooled)

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Issues and R&D on Liquid Metal Breeder Blankets

• Fabrication techniques for SiC Inserts• MHD and thermalhydraulic experiments on SiC flow

channel inserts with Pb-Li alloy• Pb-Li and Helium loop technology and out-of-pile test

facilities• MHD-Computational Fluid Dynamics simulation • Tritium permeation barriers• Corrosion experiments• Test modules design, fabrication with RAFS, preliminary

testing• Instrumentation for nuclear environment

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Fusion environment is unique and complex:multi-component fields with gradients

Neutrons (fluence, spectrum, temporal and spatial gradients)

• Radiation Effects (at relevant temperatures, stresses, loading conditions)

• Bulk Heating• Tritium Production• Activation

Heat Sources (magnitude, gradient)• Bulk (from neutrons and gammas)• Surface

Synergistic Effects• Combined environmental loading conditions• Interactions among physical elements of components

A true fusion environment is ESSENTIAL to Activate mechanisms that cause prototypical coupled phenomena and integrated behavior

Particle Flux (energy and density, gradients)

Magnetic Field (3-component with gradients)

• Steady Field• Time-Varying Field

Mechanical Forces• Normal/Off-Normal

Thermal/Chemical/Mechanical/ Electrical/Magnetic Interactions

Multi-function blanket in multi-component field environment leads to:- Multi-Physics, Multi-Scale Phenomena Rich Science to Study

- Synergistic effects that cannot be anticipated from simulations & separate effects tests. Even some key separate effects in the blanket can not be produced in non-fusion facilities (e.g. volumetric heating with gradients)

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ITER Provides Substantial Hardware Capabilities for Testing of Blanket System

Vacuum Vessel

Bio-shield

A PbLi loop Transporter located in

the Port Cell Area

He pipes to TCWS

2.2 m

TBM System (TBM + T-Extrac, TBM System (TBM + T-Extrac, Heat Transport/Exchange…)Heat Transport/Exchange…)

Equatorial Port Plug Assy.

TBM Assy

Port Frame

ITER has allocated 3 ITER equatorial ports (1.75 x 2.2 m2) for TBM testing

Each port can accommodate only 2 Modules (i.e. 6 TBMs max)

But, 12 modules are proposed by the parties

Aggressive competition for Space

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Fusion Nuclear Technology Issues

1. Tritium Supply & Tritium Self-Sufficiency

2. High Power Density

3. High Temperature

4. MHD for Liquid Breeders / Coolants

5. Tritium Control (Permeation)

6. Reliability / Maintainability / Availability

7. Testing in Fusion Facilities

Challenging

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Tritium Self-Sufficiency

• TBR ≡ Tritium Breeding Ratio = = Rate of tritium production (primarily in the blanket)

= Rate of tritium consumption (burnt in plasma)

Tritium self-sufficiency condition: Λa > Λr

Λr = Required tritium breeding ratio

Λr is 1 + G, where G is the margin required to: a) compensate for losses and radioactive decay between production and use, b) supply inventory for start-up of other fusion systems, and c) provide a hold-up inventory, which accounts for the time delay between production and use as well as reserve storage. Λr is dependent on many system parameters and features such as plasma edge recycling, tritium fractional burn up in the plasma, tritium inventories, doubling time, efficiency/capacity/reliability of the tritium processing system, etc.

Λa = Achievable breeding ratioΛa is a function of FW thickness, amount of structure in the blanket, presence of stabilizing shell materials, PFC coating/tile/materials, material and geometry for divertor, plasma heating, fueling and penetration.

NN /NN

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td = doubling time

Current physics and technology concepts lead to a “narrow window” for attaining Tritium self-sufficiency

td=10 yr

td=5 yr

td=1 yr

“Window” for Tritium self sufficiency

Max achievable TBR ≤ 1.15

Req

uire

d T

BR

Fractional burn-up [%]

Fusion power 1.5GWReserve time 2 daysWaste removal efficiency 0.9(See paper for details)

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The D-T fuel cycle includes many components whose operation parameters and their uncertainties impact the required TBR

Examples of key parameters:

•ß: Tritium fraction burn-up

•Ti: mean T residence time in each component

•Tritium inventory in each component

•Doubling time

•Days of tritium reserves

•Extraction inefficiency in plasma exhaust processing

Fuel Cycle Dynamics

Only for solid breeder or liquid breeder design using separate coolant

Plasma

Plasma Facing

Component

PFCCoolant

Blanket Coolant

processing

Breeder Blanket

Plasma exhaust processing

FW coolant processing

Blanket tritium recovery system

Impurity separation

Impurity processing

Coolant tritium

recovery system

Tritium waste

treatment (TWT)

Water stream and air

processing

Fueling Fuel management

Isotopeseparation

system

Fuel inline storage

Tritium shipment/permanent

storage

waste

Solid waste

Only for liquid breeder as coolant design

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Structural Materials • Key issues include thermal stress capacity, coolant

compatibility, waste disposal, and radiation damage effects

• The 3 leading candidates are ferritic/martensitic steel, V alloys and SiC/SiC (based on safety, waste disposal, and performance considerations)

•The ferritic/martensitic steel is the reference structural material for DEMO

(Commercial alloys (Ti alloys, Ni base superalloys, refractory alloys, etc.) have been shown to be unacceptable for fusion for various technical reasons)

StructuralMaterial

Coolant/Tritium Breeding Material

Li/Li He/PbLi H2O/PbLi He/Li ceramic H2O/Li ceramic FLiBe/FLiBe

Ferritic steel

V alloy

SiC/SiC

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44

45

Fission (PWR)

Fusion structure

Coal

Tritium in fusion

46

Lessons learned:The most challenging problems in FNT

are at the INTERFACES

• Examples:

– MHD insulators

– Thermal insulators

– Corrosion (liquid/structure interface temperature limit)

– Tritium permeation

• Research on these interfaces must be done jointly by blanket and materials researchers

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Need for High Power Density Capability

A. To improve potential attractiveness of fusion power compared to other energy sources (e.g., fission)

B. Some plasma confinement schemes have the potential to produce burning plasmas with high power density

PWR BWR LMFBR ITER-Type

Average core power density (MW/m3) 96 56 240 0.4

– FW/Blanket/Divertor concepts developed in the 1970s and ’80s have limitations on power density capability (wall load and surface heat flux)

– Substantial PROGRESS has been made in this area over the past several years (e.g. in the APEX Study in the US )

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Reliability/Maintainability/Availability is one of the remaining “Grand Challenges” to Fusion Energy Development. Chamber Technology R&D is necessary to meet this Grand Challenge.

timereplacementrate failure/1

)rate failure/1(

Need Low Failure Rate:- Innovative Chamber Technology

Need Short Maintenance Time:- Simple Configuration Confinement- Easier to Maintain Chamber Technology

Need LowFailure Rate

EnergyMultiplication

Need High Temp.Energy Extraction

Need High Power Density/Physics-Technology Partnership

-High-Performance Plasma-Chamber Technology Capabilities

thfusion MPMOiC

COE

Availability&replacement cost

Need High Availability / Simpler Technological and Material Constraints

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Availability =

Current plasma confinement schemes and configurations have:

– Relatively long MTTR (weeks to months)

Required MTBF must be high– Large first wall area

Unit failure rate must be very low MTBF ~ 1/(area • unit failure rate)

MTBFMTBF + MTTR

Reliability requirements are more demandingthan for other non-fusion technologies

50

0

200

400

600

800

MT

BF

per

Bla

nk

et S

egm

ent(

FP

Y)

0

5

10

MT

BF

per

Bla

nk

et S

yste

m(F

PY

)

0 1 2 3

MTTR (Months)

Neede

d (R

)

Expected AC

The reliability requirements on the Blanket/FW (in current confinement concepts that have long MTTR > 1 week) are most challenging and pose critical concerns. These must be seriously addressed as an integral part of the R&D pathway to DEMO. Impact on ITER is predicted to be serious. It is a DRIVER for CTF.

51

Upper statistical confidence level as a function of test time in multiples of MTBF for time terminated reliability tests (Poisson

distribution). Results are given for different numbers of failures.

5.04.54.03.53.02.52.01.51.00.50.00.0

0.2

0.4

0.6

0.8

1.0

Test Time in Multiplies of Mean-Time-Between-Failure (MTBF)

Con

fide

nce

Lev

el

Number of Failures 0

1

2

3

4

TYPICAL TEST SCENARIO

“Reliability Growth”

Example,

To get 80% confidence in achieving a particular value for MTBF, the total test time needed is about 3 MTBF (for case with only one failure occurring during the test).

Reference: M. Abdou et. al., "FINESSE A Study of the Issues, Experiments and Facilities for Fusion Nuclear Technology Research & Development, Chapter 15 (Figure 15.2-2.) Reliability Development Testing Impact on Fusion Reactor Availability", Interim Report, Vol. IV, PPG-821, UCLA,1984. It originated from A. Coppola, "Bayesian Reliability Tests are Practical", RADC-TR-81-106, July 1981.

52

Summary of Critical R&D Issues for Fusion Nuclear Technology

1. D-T fuel cycle tritium self-sufficiency in a practical system depends on many physics and engineering parameters / details: e.g.

fractional burn-up in plasma, tritium inventories, FW thickness, penetrations, passive coils, etc.

2. Tritium extraction and inventory in the solid/liquid breeders under actual operating conditions

3. Thermomechanical loadings and response of blanket and PFC components under normal and off-normal operation

4. Materials interactions and compatibility5. Identification and characterization of failure modes,

effects, and rates in blankets and PFC’s6. Engineering feasibility and reliability of electric (MHD)

insulators and tritium permeation barriers under thermal / mechanical / electrical / magnetic / nuclear loadings with high temperature and stress gradients

7. Tritium permeation, control and inventory in blanket and PFC

8. Lifetime of blanket, PFC, and other FNT components9. Remote maintenance with acceptable machine shutdown

time

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• Initial exploration of coupled phenomena in a fusion environment

• Uncover unexpected synergistic effects, Calibrate non-fusion tests

• Impact of rapid property changes in early life

• Integrated environmental data for model improvement and simulation benchmarking

• Develop experimental techniques and test instrumentation

• Screen and narrow the many material combinations, design choices, and blanket design concepts

• Uncover unexpected synergistic effects coupled to radiation interactions in materials, interfaces, and configurations

• Verify performance beyond beginning of life and until changes in properties become small (changes are substantial up to ~ 1-2 MW · y/m

2)

• Initial data on failure modes & effects

• Establish engineering feasibility of blankets (satisfy basic functions & performance, up to 10 to 20 % of lifetime)

• Select 2 or 3 concepts for further development

• Identify lifetime limiting failure modes and effects based on full environment coupled interactions

• Failure rate data: Develop a data base sufficient to predict mean-time-between-failure with confidence

• Iterative design / test / fail / analyze / improve programs aimed at reliability growth and safety

• Obtain data to predict mean-time-to-replace (MTTR) for both planned outage and random failure

• Develop a data base to predict overall availability of FNT components in DEMO

ITER TBM is the Necessary First Step to enable future Engineering Development

Sub-Modules/Modules Size Tests

~ 0.3 MW-y/m2

Stage I

Fusion “Break-in” & Scientific Exploration

Stage II Stage III

Engineering Feasibility & Performance

Verification

Component Engineering Development &

Reliability Growth

Module Size Tests

Module / Sectors Size Tests

D E M O

Role of ITER TBM

1 - 3 MW-y/m2 > 2 - 4 MW-y/m2

54

Tritium breeding capabilities are needed for the continued operation of successful ITER and fusion development

A Successful ITER will exhaust most of the world supply of tritium

ITER extended performance phase and any future long pulse burning plasma will need tritium breeding technology

TBMs are critical to establishing the knowledge base needed to develop even this first generation of breeding capability 0

5

10

15

20

25

30

1995 2000 2005 2010 2015 2020 2025 2030 2035 2040 2045

Year

Pro

jec

ted

On

tari

o (

OP

G)

Tri

tiu

m I

nv

en

tory

(k

g)

CANDU Supply

w/o Fusion

1000 MW Fusion

10% Avail, TBR 0.0

ITER-FEAT(2004 start)

TBM helps solve the Tritium Supply Issue for fusion development

(at a fraction of the cost of purchasing tritium from fission reactor sources!)

Tritium Consumption in Fusion is HUGE! Unprecedented!

55.8 kg per 1000 MW fusion power per yearProduction & Cost:

CANDU Reactors: 27 kg from over 40 years, $30M/kg (current)

Fission reactors: 2–3 kg/year/reactor,

$84M-$130M/kg (per DOE Inspector General*)

*www.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdf

55

BACKUP SLIDES

56

Will the development of high-temperature structural material lead to more attractive blankets?

Not necessarily (unless we can solve the interface problems)

1. Vanadium alloys (high temperature capability)• V is compatible only with liquid lithium• Flowing liquid Li requires MHD insulators• Tolerable crack fraction is estimated to be very low (< 10-7)—much lower

than can be achieved with real coatings• “Self-healing” coating R&D results are negative for non-isothermic systems• Laminated layer insulators (alternating layers of insulator and metallic

protection layer) were proposed, but V layer needs to be too thin

2. High-temperature advanced ferritic steels (ODS, NFS)• May potentially operate up to ~700ºC (compared to 550ºC for EUROFER

and F82H)• At present, we cannot utilize such advanced high-temperature ferritics

• Li and LiPb interface temperature (Tint) is limited by corrosion to ~500ºC Unless corrosion temperature limit is improved, EUROFER and F82H are satisfactory

• Solid breeder/structure interface cannot be increased much above 400–500ºC (to have adequate temperature window for T-release and TBR)

57

Tritium Control and Management

• Tritium control and management will be one of the most difficult issues for fusion energy development, both from the technical challenge and from the “public acceptance” points of view.

• Experts believe the T-control problem is underestimated (maybe even for ITER!)

• The T-control problem in perspective:– The scale-up from present CANDU experience to ITER and DEMO

is striking: The quantity of tritium to be managed in the ITER fuel cycle is much larger than the quantities typically managed in CANDU (which represents the present-day state of practical knowledge).

– The scale-up from ITER to DEMO is orders of magnitude:The amount of tritium to be managed in a DEMO blanket (production rate ~400 g/day) is several orders of magnitude larger than that expected in ITER, while the allowable T-releases could be comparable.

For more details, see:– W. Farabolini et al, “Tritium Control Modelling in an He-cooled PbLi Blanket…” paper in ISFNT-7 (this conference)– Papers and IEA Reports by Sze, Giancarli, Tanaka, Konys, etc.

58

Why is Tritium Permeation a Problem?

• Most fusion blankets have high tritium partial pressure:LiPb = 0.014 Pa Flibe = 380 PaHe purge gas in solid breeders = 0.6 Pa

• The temperature of the blanket is high (500–700ºC)• Surface area of heat exchanger is high, with thin walls• Tritium is in elementary formThese are perfect conditions for tritium permeation.• The allowable tritium loss rate is very low

(~10 Ci/day), requiring a partial pressure of ~10-9 Pa.Challenging!• Even a tritium permeation barrier with a permeation

reduction factor (PRF) of 100 may be still too far from solving this problem!

59

Pillars of a Fusion Energy System

1. Confined and Controlled Burning Plasma (feasibility)

2. Tritium Fuel Self-Sufficiency (feasibility)

3. Efficient Heat Extraction and Conversion (attractiveness)

4. Safe and Environmentally Advantageous (feasibility/attractiveness)

5. Reliable System Operation (attractiveness)

Yet, No fusion blanket has ever been built or tested!

The Blanket is THE KEY component

60

Structure of TBM collaboration* Structure of TBM collaboration* Each 1/2 of a port is dedicated to testing of one TB concept design (one

module or several connected sub-modules). Each Concept design is tested by a partnership of a TB concept leader +

supporting partners One of the two TB leaders occupying the same port must play a role of the

Port master – responsible for integration of the given port Port Master has main responsibility of integration of 2 concepts in the Port Master has main responsibility of integration of 2 concepts in the

same port frame and preparation of the integrated testing program same port frame and preparation of the integrated testing program (replacement strategy) for the port(replacement strategy) for the port

ITER has only 3 ports Only 6 TBMs can be tested

750

1248

Horizontal

524

1700

Vertical Each Port can hold two vertically or horizontally

oriented half-port size integrated TBMs

List of TBM Design Proposals for day-one (DDDs completed by Parties)

Helium-cooled Lithium-Lead TBM (2 designs) Dual-coolant (He+Lithium-Lead) TBM (2 designs) He-cooled Ceramic Breeder/Beryllium multiplier TBM (4 designs) Water-cooled Ceramic Breeder/Beryllium multiplier TBM (1 design) He-cooled Liquid Lithium TBM (1 design)Self-cooled Liquid Lithium TBM (1 design)

* Initial result of TBWG DH meeting, 18-19 July 2006

61

US advanced Ideas to Solve the TBM testing-space problem in ITER, while improving effectiveness of the tests

KO Submodule

JA Submodule

US Submodule

US experiments can focus on niche areas of US interest and expertise,

While benefiting from world resources and testing results

Different sub-modules can be tailored to address the many different:

– design configurations – material options– operating conditions

(such as flow rates, temperatures, stresses)

– diagnostics and experimental focusThe back plate coolant supply

and collection manifold assembly incorporates all connections to main support systems. A “Lead Party” takes responsibility for back plate and sub-module integration

62

0

5

10

15

20

25

30

2000 2010 2020 2030 2040

Year

Pro

ject

ed O

ntar

io (

OP

G)

Trit

ium

In

ven

tory

(kg

)

CANDU Supply w/o Fusion

Projections for World Tritium Supply Available to Fusion Reveal Serious Problems

World Max. tritium supply is 27 kg

Tritium decays at a rate of 5.47% per year

Tritium Consumption in Fusion is HUGE! Unprecedented!55.8 kg per 1000 MW fusion power per year

Production & Cost:

CANDU Reactors: 27 kg from over 40 years, $30M/kg (current)

Fission reactors: 2–3 kg per year. It takes tens of fission reactors to supply one fusion reactor.

$84M-$130M per kg, per DOE Inspector General*

*DOE Inspector General’s Audit Report, “Modernization of Tritium Requirements Systems”, Report DOE/IG-0632, December 2003, available at www.ig.doe.gov/pdf/ig-0632.pdf

63

Tritium Consumption in ITER

Here is from a summary of the final design report. Link is:http://fusion.gat.com/iter/iter-ga/images/pdfs/cost_estimates.pdf

9.4.3 Fuel CostsThe ITER plant must be operated, taking into account the available tritium externally supplied. The net tritium consumption is 0.4 g/plasma pulse at 500 MW burn with a flat top of 400 s

“The total tritium received on site during the first 10 years of operation, amounts to 6.7 kg.”

“whereas the total consumption of tritium during the plant life time may be up to 16 kg to provide a fluence of 0.3 MWa/m2 in average on the first wall”

“This corresponds, due to tritium decay, to a purchase of about 17.5 kg of tritium. This will be well within, for instance, the available Canadian reserves.”

64

It is important to precisely understand the state-of-the-art in blanket and material research, and the role of ITER

Over the past 30 years the fusion nuclear technology and materials programs have spent much effort on developing theories and models of phenomena and behavior.

– But, these are for idealized conditions based on understanding of single effects. There has never been a single experiment in a fusion environment!

Are they scientifically valid?

ITER will provide the 1st opportunity to test these theories and models in a real fusion environment.

Only the Parties who will do successful, effective TBM experiments in ITER will have

the experimentally-validated scientific basis to embark on the engineering development of tritium breeding blankets

65

Another Look at the DCLL Unit Cell

66

A Simplified DCLL DEMO System

Blanket Module

Tritium Extraction

Pump

Heat Exchanger

Cold Trap,

Chem.Control

450C He350C He

450C 650C

From/To Tritium Processing System

From/To Helium Loops and Brayton Cycle Power Conversion System

Coaxial Feed Pipes

• PbLi Hot leg flows in inner pipe (700C)

• PbLi Cold leg flows in outer annulus (450C)

• Cold leg cools Pipe walls and TX/HX shells

67

Coolant Routing Through HX Coupling Blanket and Divertor to Brayton Cycle

Fusion Thermal Power in Reactor Core 2650 MW

Fusion Thermal Power in Pb-17Li 1420 MW

Fusion Thermal Power in Blkt He 1030 MW

Friction Thermal Power in Blkt He 119 MW

Fusion Thermal Power in Div He 201 MW

Friction Thermal Power in Div He 29 MW

Total Power 2790 MW

Overall Brayton Cycle Efficiency 0.42

Power Parameters for DCLL in ARIES-CSPower Parameters for DCLL in ARIES-CS

Blkt He

Typical Fluid Temperatures in HX for DCLL ARIES-CS

Blkt LiPbBlkt LiPb (711°C)+ Div He (711°C)

Cycle He

~681°C580°C

369°C

441°C

452°C

349°C

T

ZHX

Pb-17Li from

Blanket

Hefrom

Divertor

Hefrom

Blanket

BraytonCycle

He THX,out

He THX,in

Blkt He Tin

Blkt He Tout

(Pth,fus+Pfrict)Blkt,He

(Pth,fus)Blkt,LiPb

LiPb Tin

LiPb Tout

Div HeTin

Div He Tout

(Pth,fus+Pfrict)Div,He

68

Blanket systems are complex and have many integrated functions, materials, and interfaces

Tritium BreederLi2TiO3 (<2mm)

First Wall(RAFS, F82H)

Neutron MultiplierBe, Be12Ti (<2mm)

Surface Heat FluxNeutron Wall Load

[18-54] mm/s

PbLi flow scheme

[0.5-1.5] mm/s

69

TBM is an integral part of ITER Schedule, Safety, and Licensing

Early H-H Phase TBM Testing is mandated by ITER IO and licensing team:

1. Optimize plasma control in the presence of Ferritic Steel TBMs2. Qualify port integration and remote handling procedures3. License ITER with experimental TBMs for D-T operation

ITER Schedule 1st 10 yrs

70

Plasma

Radiation

Neutrons

Coolant for energy conversion

First Wall

Shield

Blanket Vacuum vessel

Magnets

Tritium breeding zone

71

Helium-Cooled Pebble Breeder Concept for EU

FW channel

Breeder unit

Helium-cooled stiffening grid

72

US DCLL Concept, 4

Flow Pi, MPa Pi/P

1. Front channel 0.38410-3 0.13

2. Return channel 0.48510-3 0.16

3. Concentric pipe (internal, uniform B-field)

15.410-3 5.1

4. Concentric pipe (annulus, uniform B-field)

0.0286 9.5

5. Concentric pipe (internal, fringing B-field)

0.0585 19.3

6. Concentric pipe (annulus, fringing B-field)

0.0585 19.3

7. Inlet manifold 0.070-0.140 23.2

8. Outlet manifold 0.070-0.140 23.3

Total 0.302-0.442 100

1 2

4

73

8

Summary of MHD pressure drops for ITER TBM

73

Simplified DCLL Blanket Module Flow Scheme

All structural walls are RAFS actively cooled by He

Cold PbLi flows up the FW (where volumetric heating is strongest), turns, and flows back down the back of the blanket module

SiC FCIs separates and insulates the flowing PbLi from the RAFS walls

FCIs are loosely slip-fit together, and GAPs between FCIs and structure is filled in by nearly stagnant PbLi

The interface temperature between the RAFS structure and gap PbLi is controlled by the He cooling, and kept < 500C.

FW

Hea

t F

lux

an

d N

eutr

on

Wal

l L

oa

d

SiC FCIs

Gap between FCI and Structure(Filled with nearly stagnant PbLi)

PbLi Out (700C)

PbLi In (450C)

Helium-cooled RAFS FW and structure

PbLi (625C)

74

Idea of Coaxial Pipe for PbLi feed lines– use FCI to insulate hot leg from cold

From Reactor ~700C

To Reactor ~450C

Coaxial Pipe Outer WallOuter FCI (For MHD insulation)

Coaxial Pipe Inner Wall (~500C) PbLi Gap (~500C)Inner FCI

• Inner FCI insulates inner hot leg PbLi flow

• Allows outer cold leg PbLi flow to cool Inner pipe wall and PbLi gap to < 500C

• Same principle can be applied for TX and HX outer shells

• Allows use of ordinary RAFS for almost all structure

75

DCLL should be effective in reducing MHD pressure drop to manageable levels

Low velocity due to elimination of the need for FW cooling reduces MHD pressure drop.

Higher outlet temperature due to FCI thermal insulation allows large coolant delta T in breeder zone, resulting in lower mass flow rate requirements and thus lower velocity.

Electrical insulation provided by insert reduces bare wall pressure drop by a factor of 10-100.

1 10 100 1000Electrical conductivity, S/m

0

100

200

300

400

500

(dP

/dx)

0 /

(dP

/dx)

PEHPES

76

ITER Objectives

Programmatic• Demonstrate the scientific and technological feasibility of fusion

energy for peaceful purposes.

Technical• Demonstrate extended burn of DT plasmas, with steady state as the

ultimate goal.• Integrate and test all essential fusion power reactor technologies

and components. • Demonstrate safety and environmental acceptability of fusion.

77JAEA DEMO Design

Cryostat Poloidal Ring Coil

Coil Gap Rib Panel

Blanket

VacuumVessel

Center Solenoid Coil Toroidal Coil

Maint.Port

Plasma

FNT: Components from Edge of Plasma to TFCBlanket / Divertor immediately circumscribe the plasma

78

What can be done about MHD pressure drop?

Lower C– Insulator coatings– Flow channel inserts– Elongated channels with anchor links

or other design solutions Lower V

– Heat transfer enhancement or separate coolant to lower velocity required for first wall/breeder zone cooling

– High temperature difference operation to lower mass flow Lower B,L

– Outboard blanket only (ST) Lower (molten salt)

2VBcLP lc represents a measure of relative conductance of induced current closure paths

Break electrical coupling to thick

load bearing channel walls

Force long current path

79

Blanket Materials1. Tritium Breeding Material (Lithium in some form)

Liquid: Li, LiPb (83Pb 17Li), lithium-containing molten salts

Solid: Li2O, Li4SiO4, Li2TiO3, Li2ZrO3

2. Neutron Multiplier (for most blanket concepts)

Beryllium (Be, Be12Ti)

Lead (in LiPb)

3. Coolant

– Li, LiPb – Molten Salt – Helium – Water

4. Structural Material– Ferritic Steel (accepted worldwide as the reference for DEMO)

– Long-term: Vanadium alloy (compatible only with Li), and SiC/SiC

5. MHD insulators (for concepts with self-cooled liquid metals)

6. Thermal insulators (only in some concepts with dual coolants)

7. Tritium Permeation Barriers (in some concepts)

8. Neutron Attenuators and Reflectors

80

Neutron Multipliers

0.001

0.01

0.1

1

10

106 107

Be-9 (n,2n) and Pb(n,2n) Cross-Sections- JENDL-3.2 Data

Be-9 (n,2n) Pb (n,2n)

Neutron Energy (eV)

• Almost all concepts need a neutron multiplier to achieve adequate tritium breeding.(Possible exceptions: concepts with Li and Li2O)

• Desired characteristics:– Large (n, 2n) cross-section with

low threshold– Small absorption cross-sections

• Candidates:– Beryllium is the best (large n, 2n

with low threshold, low absorption)

– Be12Ti may have the advantage of less tritium retention, less reactivity with steam

– Pb is less effective except in LiPb

– Beryllium results in large energy multiplication, but resources are limited

9Be (n,2n)

Pb (n,2n)

Examples of Neutron MultipliersBeryllium, Lead