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Fusion Science and Technology
Mohamed Abdou, Neil Morley, Alice Ying
Mechanical and Aerospace Engineering Dept.CESTAR: Center for Energy Science and Technology Advanced Research
Presentation at KAIST/UCLA Joint Workshop, January 13-14, 2005
WEB SITE: http://www.fusion.ucla.edu/
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)
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
Plasma
Radiation
Neutrons
Coolant for energy conversion
First Wall
Shield
Blanket Vacuum vessel
Magnets
Tritium breeding zone
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)
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.
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
-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
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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 wall
-1 -0.8 -0.6 -0.4 -0.2 0 0.2 0.4 0.6 0.8 1
-1
-0.8
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1
ITER
4 of 7 Central Solenoid modules
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
BaffleRoughing pumps,
standard components
Tokamak exhaustprocessing system
Cooling for Divertorand Vacuum VesselPellet Injector
Steady-statepower supplies
U.S. In-kind Contributions to ITER
Test Blanket Module
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
• 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 Concepts for ITER-TBM Selected by the Various Parties
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
Cryostat Plug
Transporter
Cryostat Extension
TBM Frame & Shield Plug
Breeder Concentric
Pipe
Drain Pipe
FW
US Solid breeder submodule
EU HCLL Test Module
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
Radial distance (mm)
No
rma
lStr
ess
(MP
a)
0 10 20 30 40 50-2.5
-2.25
-2
-1.75
-1.5
-1.25
-1
-0.75
-0.5
-0.25
0
0.25
0.5
Temp = 450oCTemp = 650oCTemp = 822oC
Radial distance (mm)
No
rma
lStr
ess
(MP
a)
0 10 20 30 40 50-2
-1.5
-1
-0.5
0
0.5
1
Time = 0 hrTime = 2 hrTime = 24 hrTime = 48 hr
Stress relaxed as creep initiatedStress magnitude profiles at different times
Stress exerted on the wall at different bed temperatures
MARC calculations
MARC calculations
Solid breeder pebbles after the tests
107
108
109
0 0.005 0.01 0.015 0.02 0.025
Time = 0 minutes Time = 2000 minutes
Container Radius (m)
Average stress exerted on the particles at initial time and at time 2000 minutes
Test Article for Deformation Study
DEM calculations
Force distribution inside the particles with 1% compressive strain
IEA collaboration on solid breeder pebble bed time dependent thermomechanics interactions/deformation
research Primary Variables• Materials• Packing• Loadings• Modes of operation
Irradiation Effect(NRG)
Primary & Secondary Reactants:• Temperature magnitude/ gradient• Differential thermal stress/contact pressure• Plastic/creep deformation• Particle breakage• gap formation
Goal:Performance/Integrity prediction & evaluation
Partially integrated out-of-pile and fission reactor tests (NRG,ENEA)
Finite Element Code (ABQUS, MARC)
(NRG, FZK, UCLA)
Discrete Element Model (UCLA)
Design Guideline and Evaluation (out-of-pile & in-pile tests, ITER TBMs)
Database Experimental Program
(FZK, JAERI, CEA,UCLA)
Thermo-physical and Mechanical PropertiesConsecutive equations
Single/multiple effect experiments(NRG, UCLA)
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
Unstable MHD velocity profiles in gradient magnetic fields breakdown into instability
Complex geometry: Free surface flow
around cylindrical penetration
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.
Slice from 3D Simulation
Velocity profile from2D Simulation
Strong negative flow jet near pressure equalization slot not seen in 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
Turbulent fluctuations organize into 2D structures with 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
Experiments on film flows show formation of 2D turbulence structures
B
U
Sophisticated 2-D neutronics analysis shows testing objective can be achieved for a proposed NT TBM
0
1 10-5
2 10-5
3 10-5
4 10-5
5 10-5
0 10 20 30 40 50 60 70
Distance from Frame, cm
Left Configuration Right Configuration
Layer#
Layer#
1
2
3
4
5
6
7
8
9
1
3
2
4
5
6
3
2
4
6
8
10
12
0 10 20 30 40 50 60 70 80
Left TBM WallBe Layer-Left Config.Left VCP-Left Config.Br1Right TBM WallBe Layer- MiddleBe-Rt. SubmduleBe Layer-Rt. Config.Rt. VCP- Left Config.Left VCP-Rt. Config.Rt. VCP-Rt. Config.
Toroidal Distance from Frame, cm
Depth = 42 mm behind FW
Breeder (Lft. Config.)
Be (Rt. Config.)
Proposed NT TBM
JA TBMTritium production profiles are nearly flat over a reasonable distance in the toroidal direction allowing accurate measurements be performed
Finding:Flat nuclear heating and tritium production profiles allow two designs to be evaluated in a ¼ port submodule
Pulsed electro-thermal plasma gun facility provides extreme high heat flux capability for IFE super-heated vapor condensation study
Condensed steel droplets on top of deposited film
Droplet size ~ 1 to 2 m
Condensation characterization from super-heated vapor (for Z-pinch)
Electrical network system provides a pulsed energy source simulating the pellet explosion for
rapid vapor generation
Expansion chamber and diagnostics for super-heated vapor consideration studies
Vapor density decays exponentially with a time constant of 6.58 ms in the range between 5x1017 cm-3 and 2x1015
cm-3
Time 0
820 s
1640 s
Frame sequences recorded with high speed camera - 10,000 frames per second
and shutter speed of 100 ms
Vapor pressure decay curve
Goal:
assessing chamber assessing chamber clearing issues in clearing issues in Inertial Fusion Inertial Fusion Energy systemsEnergy systems
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-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