Fusion Science and Technology
Mohamed Abdou, Neil Morley, Alice Ying
Mechanical and Aerospace Engineering Dept.
CESTAR: Center for Energy Science and Technology Advanced Research
WEB SITE: http://www.fusion.ucla.edu/
Presentation at KAIST/UCLA Joint Workshop, January 13-14, 2005
Fusion Science and Technology at UCLA
• Fusion Research is exciting and active worldwide
• UCLA has strong research programs in plasma physics, fusion
science and technology
• The largest part of the Fusion Science and Technology
Research at UCLA is in the Mechanical and Aerospace
Engineering Department
• UCLA leads the US program in Fusion Nuclear Technology
• We already have strong international collaborative programs
with Europe, Japanese Universities, JAERI, Korea (KAIST and
KAERI), China, and Russia
• Our research involves many technical disciplines: fluid
mechanics, heat transfer, MHD, tritium transport, neutronics,
materials, structural mechanics
• We have constructed world-class experimental facilities. Many
students do their Ph.D. research in these facilities. The facilities
also attract important international collaborations
Introduction
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)
Fusion Power & Fuel Cycle Technology
FNT Components from the edge of the
Plasma to TF Coils (Reactor “Core”)
1. Blanket Components
2. Plasma Interactive and High Heat Flux
Components
a. divertor, limiter
b. rf antennas, launchers, wave guides, etc.
3. Vacuum Vessel & Shield Components
Other Components affected by the
Nuclear Environment
4. Tritium Processing Systems
5. Instrumentation and Control Systems
6. Remote Maintenance Components
7. Heat Transport and Power Conversion
Systems
Shield
Blanket
Vacuum vessel
Radiation
Plasma
Neutrons
First Wall
Tritium breeding zone
Coolant for energy
conversion
Magnets
Blanket Concepts
(many concepts proposed worldwide)
A.
B.
Solid Breeder Concepts
–
Always separately cooled
–
Solid Breeder: Lithium Ceramic (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3)
–
Coolant: Helium or Water
Liquid Breeder Concepts
Liquid 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)
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.
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:
j   (E  V  B )
– 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:
V
1
1
 (V  )V   p   2 V  g  j  B
t


– 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
Large MHD drag results in large
MHD pressure drop
Conducting walls
Insulated wall
Lines of current enter the low
resistance wall – leads to very
high induced current and high
pressure drop
1
0.8
0.6
0.4
1
0.8
0.6
0.4
0.2
0.2
0
0
-0.2
-0.2
All current must close in the
liquid near the wall – net drag
from jxB force is zero
-0.4
-0.6
-0.8
-1
•
•
-0.6
-0.8
-1
-1
-1
•
-0.4
-0.8
-0.6
-0.4
-0.2
0
0.2
0.4
0.6
0.8
-0.8
-0.6
-0.4
-0.2
0
0.2
0.4
0.6
0.8
1
1
Net JxB body force p = cVB2
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)
ITER
U.S. In-kind Contributions to ITER
4 of 7 Central
Solenoid modules
Steady-state
power supplies
15% of port-based
diagnostic packages
44% of ICRH Antenna,
plus all transmission lines,
RF-sources, power supplies
Start-up gyrotrons,
all transmission lines,
and power supplies
Test Blanket Module
Tokamak exhaust
processing system
Pellet Injector
Baffle
Cooling for Divertor
and Vacuum Vessel
Roughing pumps,
standard components
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
Blanket Concepts for ITER-TBM Selected by the
Various Parties
• Solid Breeders
– He/SB/Be/FS: All parties are strongly interested
– H2O/SB/Be/FS: Only Japan (some interest from China)
• Liquid Breeders
– He/LiPb/FS (Separately cooled): EU lead (one of two main
concepts for EU, interest from other parties)
– Dual Coolant (He/LiPb/FS with SiC): US lead, strong
interest from EU and other parties
– Li/V (Self-cooled): Russia is main advocate (but no
significant resources on R&D!)
– Molten Salts: US and Japanese Universities want the
option to decide later whether to test
– He/Li/FS: Korea’s proposal
Blanket Testing in ITER is one of ITER’s Key Objectives
Strong international collaboration among the ITER Parties is underway to provide the
science basis and engineering capabilities for ITER TBMs
Bio-Shield Plug
TBM Frame &
Shield Plug
Cryostat Plug
Breeder
Concentric
Pipe
Transporter
EU HCLL Test Module
FW
Cryostat
Extension
US Solid breeder submodule
Drain Pipe
Conceptual Liquid Breeder Port Layout and Ancillary equipment
UCLA Activities
UCLA Program in
Fusion Engineering Research
Current UCLA Research Activities
–
–
–
–
–
ITER Test Blanket Module R&D
Molten Salt Thermofluid MHD (Jupiter-II)
Solid Breeder / SiC Thermomechanics (Jupiter-II)
Solid Breeder / Steel Thermomechanics (IEA)
ITER Basic Machine and Procurement Package
Support
– Free Surface MHD Flows for Plasma Facing
Components
– IFE Chamber Clearing Study
Experiments, Microscopic and Macroscopic Modeling efforts
simultaneously underway to Understand and Predict Solid Breeder
Blanket Pebble Bed Thermomechanics Interactions
Force distribution inside
Average stress exerted on the particles at the particles with 1%
compressive strain
initial time and at time 2000 minutes
Stress exerted on the wall at different
bed temperatures
0.5
0.25
-0.25
Normal Stress (MPa)
DEM calculations
Temp = 450 oC
Temp = 650 oC
Temp = 822 oC
0
-0.5
-0.75
-1
-1.25
-1.5
-1.75
-2
MARC calculations
-2.25
-2.5
0
10
20
30
40
50
Radial distance (mm)
Stress relaxed as creep initiated
Stress magnitude profiles at different times
1
Time = 0 hr
Time = 2 hr
Time = 24 hr
Time = 48 hr
Normal Stress (MPa)
0.5
0
Solid breeder
pebbles after
the tests
-0.5
-1
-1.5
MARC calculations
-2
0
10
20
30
Radial distance (mm)
40
50
Test Article for
Deformation Study
IEA collaboration on solid breeder pebble bed time dependent
thermomechanics interactions/deformation research
Primary Variables
• Materials
• Packing
• Loadings
• Modes of operation
Partially integrated
out-of-pile and
fission reactor tests
(NRG,ENEA)
Single/multiple effect experiments
(NRG, UCLA)
Finite Element Code
(ABQUS, MARC)
(NRG, FZK, UCLA)
Design Guideline and
Evaluation (out-of-pile &
in-pile tests, ITER TBMs)
Primary & Secondary
Reactants:
• Temperature magnitude/
gradient
• Differential thermal
stress/contact pressure
• Plastic/creep deformation
• Particle breakage
• gap formation
Discrete Element
Model (UCLA)
Thermo-physical and
Mechanical Properties
Consecutive equations
Database Experimental Program
(FZK, JAERI, CEA,UCLA)
Goal:
Performance/Integrity
prediction & evaluation
Irradiation Effect
(NRG)
UCLA is collaborating on
HIMAG 3D - a complex
geometry simulation code
for free surface MHD
flows
Simulations are crucial to both
understanding phenomena and exploring
possible flow option for NSTX Li module
Problem is challenging from a number of
physics and computational aspects requiring
clever formulation and numerical
implementation
Complex geometry:
Free surface flow
around cylindrical
penetration
Unstable MHD velocity profiles in gradient magnetic fields
breakdown into instability
Complex geometry MHD codes already being
applied to DCLL blanket with SiC
Flow Channel Inserts
• 2D and 3D codes
(developed for Liquid
walls) have been modified
for DCLL
• Initial results show strong
sidelayer jets at SiC = 500
S/m with current DCLL
design
• 2D and 3D codes give
conflicting results
concerning flow in the
“stagnant” gap region.
• Code improvements and
debugging, and continued
simulations planned for
FY05.
Strong negative flow jet near pressure
equalization slot not seen in 3D simulation
Velocity profile from
2D Simulation
Slice from 3D Simulation
Gap corner
jets not seen in 2D simulation
UCLA MTOR can be for basic flow
physics, free surface and TBM
module simulation experiments
 Large magnetic
volume for
complex geometry
modules
 Higher field
smaller volume
regions for higher
MHD interaction
experiments
 30 liter gallium
alloy flowloop
FC#1
FC#2
MTOR LM-MHD Facility
Experiments on film flows show
formation of 2D turbulence
structures

B
 Turbulent fluctuations organize into 2D structures with
U
vorticity along the magnetic field
 Corner vortices and small surface disturbances suppressed
 Flow can Pinch-IN in field gradients and separate from the wall
 Drag can be severe, slowing film down by 2x or 3x
B
Sophisticated 2-D neutronics analysis shows
testing objective can be achieved for a proposed
NT TBM
5 10
-5
Right Configuration
Left Configuration
4 10
Layer#
-5
Layer#
1
3 10
1
2
3
-5
2
3
2 10
-5
5
6
3
4
7
4
1 10
-5
8
5
6
0
0
10
12
Left TBM Wall
Be Layer-Left Config.
Left VCP-Left Config.
Br1
Right TBM Wall
Be Layer- Middle
Be-Rt. Submdule
Be Layer-Rt. Config.
Rt. VCP- Left Config.
Left VCP-Rt. Config.
Rt. VCP-Rt. Config.
10
JA TBM
Finding:
Flat nuclear heating
and tritium production
profiles allow two
designs to be evaluated
in a ¼ port submodule
8
Breeder (Lft. Config.)
6
4
Be (Rt. Config.)
2
0
10
20
30
40
30
40
50
60
70
Distance from Frame, cm
Depth = 42 mm behind FW
Proposed NT TBM
20
9
50
60
70
Toroidal Distance from Frame, cm
80
Tritium production
profiles are nearly flat
over a reasonable
distance in the
toroidal direction
allowing accurate
measurements be
performed
Pulsed electro-thermal plasma gun facility provides extreme high
heat flux capability for IFE super-heated vapor condensation study
Goal:
Time 0
assessing chamber
clearing issues in
Inertial Fusion
Energy systems
820 ms
1640 ms
Frame sequences recorded with high
Electrical network system provides a pulsed
speed camera - 10,000 frames per second
energy source simulating the pellet explosion for
and shutter speed of 100 ms
rapid vapor generation
Vapor pressure decay
curve
Droplet size ~ 1 to 2 mm
Condensed steel droplets on
top of deposited film
Vapor density decays exponentially Condensation characterization from
super-heated vapor (for Z-pinch)
with a time constant of 6.58 ms in the
Expansion chamber and diagnostics for super17
-3
range between 5x10 cm and
heated vapor consideration studies
2x1015 cm-3
Possibilities for Collaboration
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
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-ofpile 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