Overview of Principles, Concepts, and Key Issues of

Fusion Nuclear Technology

Mohamed AbdouProfessor of Engineering and

Director of Fusion Science and Technology CenterUniversity of California Los Angeles

Seminar Presented to KAERI and KBSI, Korea, April 2004

Overview of Fusion Nuclear Technology (FNT)Overview of Fusion Nuclear Technology (FNT)

Seminar Seminar OutlineOutlinePrinciples, Concepts, and Key Issues (this presentation/handout)

• FNT components and functions

• Tritium Breeding

• Solid Breeders

- Design Concepts and Materials

- Tritium Release, Extraction

- Issues and R&D

• Structural Materials

• Liquid Breeders- Concepts: Materials, Configurations

- Li/V - LiPb Dual Coolant - Molten Salts

- Issues

• Tritium Supply and Need for Blanket Testing in ITER

• Possible Areas for US-Korea Collaboration

Other Presentations / Handouts

- Fabrication Technology

- ITER Test Blanket Module (TBM)

- ITER Testing : Engineering Scaling

Incentives 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

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

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

ARIES-AT

Plasma

Radiation

Neutrons

Coolant for energy conversion

First Wall

Shield

Blanket Vacuum vessel

Magnets

Tritium breeding zone

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

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

• The Vacuum Vessel is outside the Blanket (/Shield). It is in a low-radiation field.

• Vacuum Vessel Development for DEMO should be in good shape from ITER experience.

• The Key Issues are for Blanket / PFC.

• Note that the first wall is an integral part of the blanket (ideas for a separate first wall were discarded in the 1980’s). The term “Blanket” now implicitly includes first wall.

• Since the Blanket is inside of the vacuum vessel, many failures (e.g. coolant leak from module) require immediate shutdown and repair/replacement. Adaptation from ARIES-AT Design

Notes on FNT:

Heat and Radiation Loads on First Wall

• Neutron Wall Load ≡ Pnw

Pnw = Fusion Neutron Power Incident on the First Wall per unit area

= JwEo

Jw = fusion neutron (uncollided) current on the first wall

Eo = Energy per fusion neutron = 14.06 MeV

• Typical Neutron Wall Load ≡ 1-5 MW/m2

At 1 MW/m2: Jw = 4.43 x 1017 n · m-2 · s-1

• Note the neutron flux at the first wall (0-14 MeV) is about an order of magnitude higher than Jw

• Surface heat flux at the first wallThis is the plasma radiation load. It is a fraction of the α-power

qw = 0.25 Pnw · fα

where f is the fraction of the α-power reaching the first wall

(note that the balance, 1 – f, goes to the divertor)

Poloidal Variation of Neutron Wall Load– Neutron wall load has profile along the poloidal direction (due to combination of toroidal and poloidal geometries)

–Peak to average is typically about 1.4

(equatorial plane, outboard, in 0º)

Inboard

outboard

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

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 burnup 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

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

– 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

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 solid breeder or liquid breeder design using separate coolant

Only for liquid breeder as coolant design

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

1.05

1.10

1.15

1.20

1.25

0 0.5 1 1.5 2

Ove

rall

TB

R

Effective Thick. of FS in OB FW (cm)

Required TBR

ARIES-ST7% FS, 7% He,

12% SiC inserts, 74% LiPb

Achievable TBR is Very Sensitive to FW Thickness

TBR drops by up to 15% with 2 cm thick FS FW [L. El-Guebaly, Fusion Engr & Design, 2003]

ITER FW Panel Cross Section

1.0

1.2

1.4

1.6

1.8

2.0

0 5 10 15 20

TB

R

Structure Content (%)

Li Breeder

V

FS

SiC

1.0

1.2

1.4

1.6

1.8

2.0

0 5 10 15 20

TB

R

Structure Content (%)

Li17

Pb83

Breeder

(90% Li-6)

VFS

SiC

TBR is Very Sensitive to Structure Content in Blanket

Impact of structure content on TBR depends on breeder and structural material used

V has the least impact on breeding

Up to 30% reduction in TBR could result from using 20% structure in the blanket

Note: Net TBR is substantially lower (~30-40%) than local TBR

1.2

1.3

1.4

1.5

0 5 10 15 20

TB

R

Structure Content (%)

Li2O Breeder

V

FS

SiC

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

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)

JA Water-Cooled Solid Breeder Blanket

Tritium BreederLi2TiO3, Li2O (<2mm)

First Wall(RAFS, F82H)

Coolant water (25MPa, 280/510oC)

Neutron MultiplierBe, Be12Ti (<2mm)

Surface Heat Flux:1 MW/m2

Neutron Wall Load: 5 MW/m2(1.5×1015n/cm2s)

Optional W coating for FW protection

Helium-Cooled Pebble Breeder Concept for EU

FW channel

Breeder unit

Helium-cooled stiffening grid

Stiffening plate provides the mechanical strength to the structural box

Radial-toroidal plate

Radial-poloidal plate

Grooves for helium coolant

Helium

Cut view

Breeder Unit for EU Helium-Cooled Pebble Bed Concept

Mechanisms of tritium transport (for solid breeders)

Mechanisms of tritium transport

1) Intragranular diffusion

2) Grain boundary diffusion

3) Surface Adsorption/desorption

4) Pore diffusion

5) Purge flow convection

(solid/gas interface where adsorption/desorption occurs)

Li(n, 4He)T

Purge gas composition:

He + 0.1% H2

Tritium release composition:

T2, HT, T2O, HTO

Breeder pebble

Some mathematical formulas

t

desss

desssdess

dttRTEKCtC

tRTEtCKtCtKdtdCtR

0

00

0

'))'(/exp(exp)(

))(/exp()()()(/)(

),(),(2),(

)(),(

2

2

trGr

trC

rr

trCTD

t

trC

Diffusion model:

)/exp()( 0 RTEDTD d

22233

322 /expsin

)1(2)(

6atDnx

a

rn

nrD

Gara

D

GC

n

0),(

0),(

0)0,(

0

rr

trC

taC

rC

First -order tritium release rate estimated:

Generation

rate

Activation energy

Surface concentration (atoms/m2)

Desorption rate constant

Desorption energy

MISTRAL (Model for Investigative Studies of Tritium Release in Lithium Ceramics)- a code developed at

UCLA

Transport mechanisms included:grain diffusiongrain boundary diffusionsadsorption from the bulk and from the pores to the surfacedesorption to the poresdiffusion through the pores

Features• includes details of the

ceramic microstructure • includes coverage

dependence of the activation energy of surface processes (adsorption/ desorption)

To understand and predict tritium release characteristics

Solid phase

Gas

phase

Phenomenological cartoon

“Temperature Window” for Solid Breeders

• The operating temperature of the solid breeder is limited to an acceptable “temperature window”: Tmin– Tmax

– Tmin, lower temperature limit, is based on acceptable tritium transport characteristics (typically bulk diffusion). Tritium diffusion is slow at lower temperatures and leads to unacceptable tritium inventory retained in the solid breeder

– Tmax, maximum temperature limit, to avoid sintering (thermal and radiation-induced sintering) which could inhibit tritium release; also to avoid mass transfer (e.g., LiOT vaporization)

• The limitations on allowable temperature window, combined with the low thermal conductivity, place limits on allowable power density and achievable TBR

Effect of helium purge flow rate on pressure drop and tritium

permeation

single size bedbinary

bed

)2/(2)(

)1(175

03

2

PPAd

NRTLP

bp

f

=Porosity,

= pebble sphericity =1 for spherical pebble

N = moles/s

R = ideal gas constant

T = temperature

f = helium gas viscosity

Ab= gas flow cross-sectional area

P0= inlet pressure

L = flow path

dp = particle diameter

Porosity,

Which solid breeder ceramic is better?

400-1400Up to 900325-925397-795

Min.-Max. Temp. for Tritium Release (°C)

93~9687~89~9880-85 Density (%TD)

0.5-21-45-1550Grain Size (μm)

LessLessLittleHighReactivity w/H20

68-7924-33~ 10- Crush Load (N)

< 0.7-1.77.0Swelling @500 ° C (V/V0%)

0.751.82.44.7Thermal Conductivity @ 500 ° C (W/m/ ° C)

0.50.81.151.25Thermal Expansion @ 500 ° C (L/L0%)

0.9~1.50.7~0.850.2~0.7~1.0Diameter (mm)

0.380.430.510.94Lithium Density (g/cm3)

Li2ZrO3Li2TiO3Li4SiO4 Li2O

Residence time @400 °C (h) 10 2 2 1

Properties are for 100% TD

Pore for tritium release

Higher design margin

Relatively narrow T window

Parameters:Lithium density Tritium residence time Thermal-physical propertiesMechanical propertiesTemperature windowTransmutation nuclides (activation products) ReactivityFabrication

Irradiation effects (e.g, swelling) Notes:• Li2O is highly hygroscopic: 2Li2O + H2O → 2LiOH (ΔH = 128.9 kJ/mole); LiOH is highly corrosive• Li2O has been observed to swell under irradiation• Li2O is the only ceramic that may achieve the desired TBR without a neutron multiplier (but not assured)

Tritium Extraction for Solid Breeder Blankets

Tritium form: HT and HTO

1% H2Absorb impurities from tritium stream at ambient temperature Remove the remaining

impurities and the hydrogen isotopes at a cryogenic temperature molecular sieve bed

When the CMSB is saturated with Q2 (hydrogen isotopes) it is taken off line for regeneration and its companion bed can be put into service. A CMSB is regenerated by warming. The Q2 desorbs and is sent to a Pd/Ag permeator.

The “bleed” stream is sent to a shift catalyst bed where reactions such as steam reforming and water gas shift can be used to move hydrogen isotopes from impurities such as CQ4 and Q2O to the form of Q2

Solid Breeder Concepts: Key Advantages and Disadvantages

Advantages

• Non-mobile breeder permits, in principle, selection of a coolant that avoids problems related to safety, corrosion, MHD

Disadvantages

• Low thermal conductivity, k, of solid breeder ceramics– Intrinsically low even at 100% of theoretical density (~ 1-3 W · m-1 · c-1 for ternary

ceramics)

– k is lower at the 20-40% porosity required for effective tritium release

– Further reduction in k under irradiation

• Low k, combined with the allowable operating “temperature window” for solid breeders, results in:– Limitations on power density, especially behind first wall and next to the neutron

multiplier (limits on wall load and surface heat flux)

– Limits on achievable tritium breeding ratio (beryllium must always be used; still TBR is limited) because of increase in structure-to-breeder ratio

• A number of key issues that are yet to be resolved (all liquid and solid breeder concepts have feasibility issues)

Configurations and Interactions among breeder/Be/coolant/structure are very important and often

represent the most critical feasibility issues. • Configuration (e.g. wall parallel or

“head on” breeder/Be arrangements) affects TBR and performance

• Tritium breeding and release - Max. allowable temp. (radiation-induced sintering in solid breeder inhibits tritium release; mass transfer, e.g. LiOT formation)- Min. allowable Temp. (tritium inventory, tritium diffusion- Temp. window (Tmax-Tmin) limits and ke for breeder determine breeder/structure ratio and TBR

• Thermomechanics interactions of breeder/Be/coolant/structure involve many feasibility issues (cracking of breeder, formation of gaps leading to big reduction in interface conductance and excessive temperatures)

Thermal creep trains of Li2TiO3 pebble bed at different stress levels and temperatures

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

Major R&D Tasks for Solid Breeder Blanket • Solid breeder material development, characterization, and

fabrication

• Multiplier material development, characterization, and fabrication

• Tritium inventory in beryllium; swelling in beryllium irradiated at temperature, including effects of form and porosity

• Breeder and Multiplier Pebble Bed Characterization• Pebble bed thermo-physical and mechanical properties,

thermomechanic interactions • Blanket Thermal Behavior

• Neutronics and tritium breeding • Tritium Permeation and Processing• Nuclear Design and Analysis (Modeling Development)• Advanced In-Situ Tritium Recovery (Fission Tests)• Fusion Test Modules Design Fabrication and Testing

• Material and Structural Response

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

Comparison of fission and fusion structural materials requirements

• Fusion has obtained enormous benefits from pioneering radiation effects research performed for fission reactors– Although the fusion materials environment is very hostile, there is confidence that

satisfactory radiation-resistant reduced activation materials can be developed if a suitable fusion irradiation test facility is available

Fission

(Gen. I)

Fission

(Gen. IV)

Fusion

(Demo)

Structural alloy maximum temperature

<300˚C 600-850˚C

(~1000˚C for GFRs)

550-700˚C

(1000˚C for SiC)

Max dose for core internal structures

~1 dpa ~30-100 dpa ~150 dpa

Max transmutation helium concentration

~0.1 appm ~3-10 appm ~1500 appm

(~10000 appm for SiC)

Fission (PWR)

Fusion structure

Coal

Tritium in fusion

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

Liquid Breeder Blanket Concepts1. 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.

Physical Properties of Molten Natural Li (temperature in degrees Kelvin)Valid for T = 455-1500 K

Melting Temperature: 454 K (181ºC)

Density [1] (kg/m3) = 278.5 - 0.04657 · T + 274.6 (1-T/3500)0.467

Specific heat [1; see also 2]CP (J/kg-K) = 4754 - 0.925 · T + 2.91 x 10-4 · T2

Thermal conductivity [1]Kth (W/m-K) = 22.28 + 0.0500 · T - 1.243 x 10-5 · T2

Electrical resistivity [1]e (nm) = -64.9 + 1.064 · T - 1.035 x 10-3 T2 + 5.33 x 10-7 T3 - 9.23 x 10-12 T4

Surface tension [1] (N/m) = 0.398 - 0.147 x 10-3 · T

Dynamic viscosity [1] note: = where = kinematic viscosity (m2/s) ln (Pa - s) = -4.164 - 0.6374 ln T + 292.1/T

Vapor pressure [1]ln P (Pa) = 26.89 - 18880/T - 0.4942 ln T

References:[1] R.W. Ohse (Ed.) Handbook of Thermodynamic and Transport Properties of Alkali Metals, Intern.

Union of Pure and Applied Chemistry Chemical Data Series No. 30. Oxford: Blackwell Scientific Publ., 1985, pp. 987.

[2] C.B. Alcock, M.W. Chase, V.P. Itkin, J. Phys. Chem. Ref. Data 23 (1994) 385.

Physical Properties of Pb-17Li

Melting Temperature: TM = 507 K (234ºC)

Density [1] (kg/m3) = 10.45 x 103 (1 - 161 x 10-6 T) 508-625 K

Specific heat [1]CP [J/kg-K] = 195 - 9.116 x 10-3 T 508-800 K

Thermal Conductivity [1]Kth (W/m-K) = 1.95 + 0.0195 T 508-625 K

Electrical resistivity [1]e (nW-m) = 10.23 + 0.00426 T 508-933 K

Surface tension [2,3](N/m) =0.52 - 0.11 x 10-3 T 520-1000 K

Dynamic viscosity [1] (Pa - s) = 0.187 x 10-3 exp [1400./T] 521-900 K

Vapor pressure [2-4]P (Pa) = 1.5 x 1010 exp (-22900/T) 550-1000 K

References:[1] B. Schulz, Fusion Eng. Design 14 (1991) 199.[2] H.E.J. Schins, Liquid Metals for Heat Pipes, Properties, Plots and Data Sheets, JRC-Ispra (1967)[3] R.E. Buxbaum, J. Less-Common Metals 97 (1984) 27. [4] H. Feuerstein et al., Fusion Eng. Design 17 (1991) 203.

Physical Properties of Molten Flibe (LiF)n · (BeF2)

Melting temperature [1]TM(K) = 636 K (363ºC) n=0.88 (TM=653 K for n=1)TM(K) = 732 K (459ºC) n=2

Density [2] (kg/m3) =2349 – 0.424 · T n = 1 930-1130 K (kg/m3) =2413 – 0.488 · T n = 2 800-1080 K

Specific heat [3]CP (J/kg-K) ≈ 2380 n=2 600-1200 K ?

Thermal conductivity [3]Kth (W/m-K) = 1.0 n=2 600-1200 K ?

Electrical resistivity [2]e (-m) = 0.960 x 10-4 exp (3982/T) n=1 680-790 Ke (-m) = 3.030 x 10-4 exp (2364/T) n-2 750-920 K

Surface tension [2,4] (N/m) = 0.2978 - 0.12 x 10-3 · T n = 1 830-1070 K (N/m) = 0.2958 - 0.12 x 10-3 · T n = 2 770-1070 K

Dynamic viscosity [2](Pa - s) = 6.27 x 10-6 exp (7780/T) n = 1 680-840 K(Pa - s) = 5.94 x 10-5 exp (4605/T) n = 2 740-860 K

Vapor pressure [3]P (Pa) = 1.5 x 1011 exp (-24200/T) n = 2 770-970 K

References:[1] K.A. Romberger, J. Braunstein, R.E. Thoma, J. Phys. Chem. 76 (1972) 1154. [2] G.J. Janz, Thermodynamic and Transport Properties for Molten Salts: Correlation equations for critically evaluated density, surface tension, electrical

conductance, and viscosity data, J. Phys. Chem. Ref. Data 17, Supplement 2 (1988) 1. [3] S. Cantor et al., Physical Properties of Molten-Salt Reactor Fuel, Coolant and Flush-Salts, ORNL-TM-2316 (August 1968). [4] K. Yajima, H. Moriyama, J. Oishi, Y. Tominaga, J. Phys. Chem. 86 (1982) 4193.

Liquid BreedersSummary of some physical property data

• Some key physical property data for Flinabe are not yet available– (melting temperature measurements for promising compositions are in progress. Measurement at Sandia in early 2004 shows ~ 300ºC)

• Physical property data for Flibe are available from the MSR over a limited temperature range

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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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 wall

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Magnetic Field (T)

Pip

e S

tres

s (M

Pa)

Base Case

Blanket Thickness = .25 m

Flow Length = 4 m

Bulk T rise = 300 K

NWL = 2.5 MW/m2

ITER FW ARIES-RS FW

Lithium Inboard Base Case - NWL = 5 MW/m2 - Blanket Thickness = 0.2 m - Blanket Length = 6.0 m - Coolant Bulk T rise = 200 K

LM-MHD pressure drop window for inboard channels is closed!

Tc

BLS

p

w

22)NWL( (Sze, 1992)

U ~ .2-.3 m/s

So a strategy is needed to reduce MHD pressure drop for liquid metals

• Lower K– Insulator coatings/Laminated walls– Flow channel inserts– Elongated channels with anchor

links or other design solutions

• Lower Velocity: U – Heat transfer enhancement or dual/separate coolant to lower

velocity required for first wall/breeder zone cooling– High temperature difference operation to lower mass flow

• Lower Magnetic field and flow length: B,L – Outboard blanket only, with poloidal segmentation

• Lower electrical conductivity: (molten salt)

2UBKLP lMHD “K” factor represents a measure of relative conductance of induced current closure paths

Main options considered: Break electrical coupling to load bearing walls so pipe walls can be made thick for more strength without also increasing pressure drop!

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

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 wall

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

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

He-Cooled PbLi Flow Scheme

pol

rad

PbLi inlet

PbLi outlet

• PbLi is fed at the top and collected at the back

• Meandering PbLi flows in vertical columns delimited by vertical SPs

• Alternative flow holes at front/back of horizontal SPs[0.5-1.5] mm/s

[18-54] mm/s

Key features of Dual Coolant Lead-Lithium Concept

(One of the concepts considered by the U.S. for ITER TBM)

• Cool the ferritic steel FW and structure with separate coolant – He (also used for FW/blanket preheating and possible tritium control)

– The idea is to keep the structure temperature below 550ºC, the allowable temperature for ferritic steel, and the interface temperature below 480ºC

• Breeding zone is self-cooled PbLi– PbLi can be moving at slow velocity, since the heat generation rate in the breeding

zone is lower than the surface heat flux at the FW– PbLi can be operated at temperatures higher than the structure, for higher

thermodynamic efficiency.– Some type of thermal/MHD insulator, e.g., FCIs, is required, but the requirements

are more relaxed than for “all self-cooled” concepts

• Use flow channel inserts (FCIs), wherever possible to:– Provide electrical insulation to reduce MHD pressure drop– Provide thermal insulation to decouple PbLi bulk flow temperature from wall

temperature– Provide permeation barrier to reduce T permeation into the He system– Provide additional corrosion resistance since only stagnant PbLi is in contact with

the ferritic steel structural walls

Dual Coolant Concept Designs from EU and USA

Cross section of the breeder region unit cell(ARIES)

Flow Channel Insert Properties and Failures are Dominant Issues for PbLi Dual Coolant

Blankets• Electrical and thermal conductivity of the SiC/SiC should be as low

as possible to avoid velocity profiles with side-layer jets and excess heat transfer to the He-cooled structure.

• The inserts have to be compatible with Pb-17Li at temperatures up to 700-800 °C

• Liquid metal must not “soak” into pores of the composite in order to avoid increased electrical conductivity and high tritium retention. In general “sealing layers” are required on all surfaces of the inserts.

– even if the change in conductivity is modest from pressure drop point of view it could also affect flow balance

• There are minimum primary stresses in the inserts. However, secondary stresses caused by temperature gradients must not endanger the integrity under high neutron fluence.

• The insert must be applicable and affordable

Molten Salt Concepts: Advantages and Issues

Advantages

• 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)

Molten Salt Blanket Concepts(One of the concepts considered by U.S. for ITER TBM)

• Lithium-containing molten salts are used as the coolant fot 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

Concepts considered by US for ITER TBM:– Dual coolant (He-cooled ferritic structures, self-cooled molten salt)

– Self-cooled (only with low-melting-point molten salt)

HeliumFlows

HeliumFlows

Poloidal cross-section

Example: Dual-Cooled FLiBe + Be Blanket Concept

Dual Coolant Molten Salt Blanket Concepts• He-cooled First Wall and structure• Self-cooled breeding region with flibe or flinabe• No flow-channel insert needed (because of lower

conductivity)

FLINaBe Out2/3

FLINaBe Out 1/3

FLINaBe In

Self-cooled – FLiNaBe Design ConceptRadial Build and Flow Schematic

Tritium Extraction from Liquid Lithium

Protium Addition Unit 2 cm Dia x 50 cm Lg

Organic Coolant

Li Cooler HX 2 cm x 40 cm

t = 250°C m= 7.6 x 105 g/d

Li Cold Trap for Li(T+H) 3 cm x 80 cm

T = 200°C m(Li) = 770 g/d m(T) = 0.23 g/d m(H) = 95 g/d

Decomposer 1 cm x 5 cm

T = 600°C

Li

(Li + H+ T)

T = 600°

Coolant

Cold Trap for Li 2 cm x 20 cm

Coolant

B

A

A

Protium Storage

0.5 L

m = 95 g/d T= Room Temp.

Lithium m= 7.6 x 105 g/d

Module By-Pass

Protium

(H+T) is returned to Isotope Separation System for Tritium Separation

Li returned to Test Blanket LoopA

B

This two-phase mixture is then sent to a cold trap where the LiQ (Q represents H, D and T) is collected. This process reduces the tritium concentration in the stream to 1 ppm.

Periodically the cold trap will require regeneration. This can be accomplished by heating the cold trap to 600 C. At this temperature the hydrogen isotopes will exert a 10 torr partial pressure which can be pumped away and recovered.

The liquid lithium exits the torus and has protium added to it.

This hydrogen-swamped stream is introduced into a cooler to reduce the temperature to 200 C. This will cause a portion of LiH and LiT to precipitate from the liquid lithium.

Lastly, the liquid lithium and some of the LiQ is sent to a heater before being reintroduced into the torus.

Tritium Removal and Recovery from LiPb

Technique includes the following steps:

•tritium permeation into the NaK-filled gap of the double-walled heat exchanger

•tritium removal from the NaK by precipitation as potassium tritide in a cold trap

•tritium recovery by thermal decomposition of the tritide and pumping off the tritium gasTwo cold traps are operated in parallel: one for tritium removal by

circulating the tritium dissolved in the NaK to the cold trap; the other for tritium recovery. For this purpose the cold trap is decoupled from the circulation loop, drained from NaK, heated up to temperatures of about 380oC and the released tritium gas is pumped off and stored in a getter bed.

Tritium Consumption and Production

Fusion Consumption 55.8 kg per 1000MW fusion power per year

Production & Cost

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

• Fission reactors: few kg per year, $200M/kg!! (projected cost after Canadian tritium is gone) It takes tens of fission reactors to supply one fusion reactor.

Conclusions

• ITER’s extended phase requires tritium breeding.

• Large power DT facilities must breed their own tritium.

World Tritium Supply Would be Exhausted by 2025 if ITER Were to Run at 1000MW at 10% Availability

(OR at 500 MW at 20% availability)

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CANDU Supply

w/o Fusion

ITER-FEAT(2004 start)

1000 MW Fusion,

10% Avail, TBR 0.0

Tritium supply and self-sufficiency are as critical to fusion energy as demonstrating a burning plasma.

– There is no practical external source of tritium for fusion energy development beyond a few months of DT plasma operation in an ITER-like device.

– There is NOT a single experiment yet in the fusion environment to show that the DT fusion fuel cycle is viable.

They are “Go-No Go” Issues for Fusion:

• Early development of tritium breeding blanket is critical to fusion now• Testing breeding blanket modules in ITER is REQUIRED

A fusion test facility allows SIMULTANEOUS testing of integrated (synergistic) effects, multiple effects, and single effects

Testing in a Fusion Facility is the fastest approach to Blanket and Fusion Development to Demo

- Allows understanding through single and multiple effects tests under same conditions- Provides “direct” answer for synergistic effects

Specimen (thousands)

Capsule test (100’s) SubmoduleTest Module

9 cm

2.5 cm

50 cm

10.8 cm

10

0

cm

* Figures are not to scale. Note Dimensions (>100)(>30)

• Also Test Sectors (several)

ITER Provides the First Integrated Experimental Conditions for Fusion Technology Testing

• Simulation of all Environmental Conditions

Neutrons Plasma Particles

Electromagnetics Tritium

Vacuum Synergistic Effects

• Correct Neutron Spectrum (heating profile)

• Large Volume of Test Vehicle

• Large Total Volume, Surface Area of Test Matrix

But ITER Operating Parameters pose a serious challenge to obtaining meaningful blanket testing results. Careful design of ITER test blanket module (TBM) must be based on detailed engineering scaling.

Testing tritium breeding blankets has always been a principal objective of ITER

• “The ITER should serve as a test facility for neutronics, blanket modules, tritium production and advanced plasma technologies. The important objectives will be the extraction of high-grade heat from reactor relevant blanket modules appropriate for generation of electricity.”—The ITER Quadripartite Initiative Committee (QIC), IEA Vienna 18–19 October 1987

• “ITER should test design concepts of tritium breeding blankets relevant to a reactor. The tests foreseen in modules include the demonstration of a breeding capability that would lead to tritium self sufficiency in a reactor, the extraction of high-grade heat and electricity generation.”—SWG1, reaffirmed by ITER Council, IC-7 Records (14–15 December 1994), and stated again in forming the Test Blanket Working Group (TBWG)

What is the ITER Test Blanket Module Program?

• The ITER Test Program is managed by the ITER Test Blanket Working Group (TBWG) with participants from the ITER Central Team and representatives of the Parties

• Breeding Blankets will be tested in ITER, starting on Day One, by inserting Test Blanket Modules (TBM) in specially designed ports

• Each TBM will have its own dedicated systems for tritium recovery and processing, heat extraction, etc. Each TBM will also need new diagnostics for the nuclear-electromagnetic environment

• Each ITER Party is allocated limited space for testing two TBM’s. (No. of Ports reduced to 3. Number of Parties increased to 6)

• ITER’s construction plan includes specifications for TBM’s because of impacts on space, vacuum vessel, remote maintenance, ancillary equipment, safety, availability, etc.

• Initial exploration of performance in a fusion environment

• Calibrate non-fusion tests

• Effects of rapid changes in properties in early life

• Initial check of codes and data

• Develop experimental techniques and test instrumentation

• Narrow material combination and design concepts

• 10-20 test campaigns, each is 1-2 weeks

• Tests for basic functions and phenomena (tritium release / recovery, etc.), interactions of materials, 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)

• Data on initial failure modes and effects

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

• Select 2 or 3 concepts for further development

• Identify failure modes and effects

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

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

• 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

Size of Test Article

Required Fluence (MW-y/m

2)

Stage:

Stages of FNT Testing in Fusion Facilities

Sub-Modules

~ 0.3

I

Fusion “Break-in”

II III

Design Concept & Performance

Verification

Component Engineering Development &

Reliability Growth

1 - 3 > 4 - 6

ModulesModules/ Sectors

D E M O

- These requirements have been extensively studied over the past 20 years, and they have been agreed to internationally (FINESSE, ITER Blanket Testing Working Group, IEA-VNS, etc.)

- Many Journal Papers have been published (>35)- Below is the Table from the IEA-VNS Study Paper (Fusion Technology, Vol. 29, Jan 96)

Parameter Value

Neutron wall loada (MW/m2)

Plasma mode of operation

Minimum COT (periods with 100% availability) (weeks)

Neutron fluence at test module (MW·y/m2)

Stage I: initial fusion break-in

Stage II: concept performance verification (engineering feasibility)

Stage IIIc: component engineering development and reliability growth

Total neutron fluence for test device (MW·y/m2)

Total test area (m2)

Total test volume (m3)

Magnetic field strength (T)

1 to 2

Steady Stateb

1 to 2

0.3

1 to 3

4 to 6c

>6

>10

>5

>4

FNT Requirements for Major Parameters for Testing in Fusion Facilities with Emphasis on Testing Needs to Construct DEMO Blanket

b - If steady state is unattainable, the alternative is long plasma burn with plasma duty cycle >80%a - Prototypical surface heat flux (exposure of first wall to plasma is critical)

c - Note that the fluence is not an accumulated fluence on “the same test article”; rather it is derived from testing “time” on “successive” test articles dictated by “reliability growth” requirements

Key Fusion Environmental Conditions for Testing Fusion Nuclear Components

Neutrons (fluence, spectrum, spatial and temporal gradients)- Radiation Effects

(at relevant temperatures, stresses, loading conditions)- Bulk Heating- Tritium Production- ActivationHeat Sources (magnitude, gradient)- Bulk (from neutrons)- SurfaceParticle Flux (energy and density, gradients)Magnetic Field (3-component with gradients)- Steady Field- Time-Varying FieldMechanical Forces- Normal- Off-NormalThermal/Chemical/Mechanical/Electrical/Magnetic InteractionsSynergistic Effects- Combined environmental loading conditions

- Interactions among physical elements of components

R&D Tasks to be Accomplished Prior to Demo

1) Plasma

2) Plasma Support Systems

3) Fusion Nuclear Technology Components and Materials

4) Systems Integration

- Confinement/Burn- Disruption Control

- Current Drive/Steady State- Edge Control

- Superconducting Magnets - Heating- Fueling

- Materials combination selection and configuration optimization

[Blanket, First Wall, High Performance Divertors, rf Launchers]

- Performance verification and concept validation- Show that the fuel cycle can be closed (tritium self-sufficiency)- Failure modes and effects- Remote maintenance demonstration- Reliability growth- Component lifetime

Where Will These Tasks be Done?!• Burning Plasma Facility (ITER) and other plasma devices will address 1, 2, & much of 4• Fusion Nuclear Technology (FNT) components and materials require dedicated fusion

facility(ies) parallel to ITER (prior to DEMO) in addition to TBM testing in ITER.

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.

Excellent opportunities exist for collaboration between US and Korea on

fusion engineering

• US has extensive experience in fusion blanket systems developed over 30 years

• US has focused blanket R&D on key areas of blanket feasibility

• Korea has strong background in fission and now fusion technology systems

• Korea has strong industrial and manufacturing capabilities

• Collaboration possibilities are numerous, especially on development and deployment of ITER TBMs of joint interest.

Possibilities for US-Korea Collaboration on Helium-Cooled Ceramic Breeder

Blankets• Development and characterisation of

ceramic breeder and beryllium pebbles• Thermo-mechanics of pebble beds• Tritium release characteristics of

ceramic breeders and beryllium• Beryllium behaviour under irradiation• Helium cooling technology• Prototypical mock-up testing in out-of-

pile facility• In-pile testing of sub-modules• Development of instrumentation

Utilization of Fission Reactors for Assessing Blanket Feasibility Issues (examples)

Submodule

10.8 cm

50 cm Module 1

Li4SiO4 T= 650 oC

Module 2 Li4SiO4 T= 850 oC

Module 3 Li2TiO3 T= 650 oC

Module 4 Li2TiO3 T= 850 oC

Pebble bed assemblies for thermomechanical

experiments at HFR in Petten

65

260

60

Open condition Close condition

Window of Hf neutron absorberCore Core

6565

Hollow cylinder (Hf)

Window angle

20

Fixed neutron absorber (Hf)

65

A

B

C

A and C sections

65

B section A : Cross section A B : Cross section B C : Cross section C : Hot junction point of multi-paired thermocouple : Self powered neutron detector (SPND)

Aluminium (Al)

Stepping motorHollow cylinder for neutron absorber(Hf)

Feature: a stepping motor used in a fission capsule test to simulate ITER pulsed

operations

Li2TiO3

pebbles

Tritium release experiments in JMTR

Possibilities for US-Korea Collaboration 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

*Similar possibilities exist also for molten-salt blankets