Overview of Fusion Technology Mohamed Abdou Distinguished Professor, Mechanical and Aerospace Engineering Department Director, Center for Energy Science and Technology Advanced Research (CESTAR) Director, Fusion Science and Technology Center University of California Los Angeles (UCLA) Seminar at Xi'an Jiaotong University May 2006

Outline World energy needs Fusion: Energy source for the XXI century ITER ITER Test Blanket Principles of Fusion Nuclear Technology Blanket concepts He-Cooled Ceramic Breeder Blanket (HCBB) Dual Coolant Lead-Lithium Concept (DCLL) SiC/SiC Flow Channel Insert (FCI) Liquid Walls MHD Code Development Summary

What is Nuclear Fusion? Nuclear Fusion is the energy-producing process taking place in the core of the Sun and stars The core temperature of the Sun is about 15 million °C. At these temperatures hydrogen nuclei fuse to give Helium and Energy. The energy sustains life on Earth via sunlight

Fusion Reactions Deuterium – from water (0.02% of all hydrogen is heavy hydrogen or deuterium) Tritium – from lithium (a light metal common in the Earth’s crust) Deuterium + Lithium → Helium + Energy This fusion cycle (which has the fastest reaction rate) is of interest for Energy Production

Figure 1.1 Energy use (in gigajoules) vs. GDP (on a purchasing power parity basis) for selected countries on a per capita basis. Data from the International Energy Agency. Upper line indicates ratio for the US; lower line indicates ratio for Japan and several Western European countries.

Figure 1.2. Human development index vs. per capita electricity use for selected countries. Taken from S. Benka, Physics Today (April 2002), pg 39, and adapted from A. Pasternak, Lawrence Livermore National Laboratory rep. no. UCRL-ID

The World, particularly in developing countries, needs a New Energy Source Growth in world population and growth in energy demand from increased industrialisation/affluence will lead to an Energy Gap which will be increasingly difficult to fill with fossil fuels Without improvements in efficiency we will need 80% more energy by 2020 Even with efficiency improvements at the limit of technology we would still need 40% more energy

World Energy Scene* (I) 1) The world uses a lot of energy Average power consumption = 13.6 TWs, or 2.2 kWs per person World energy market ~ $3 trillion/yr (electricity ~$1 trillion/yr) - very unevenly (OECD 6.2 kW/person; Bangladesh 0.20 kW/person; China 1.3kW/person) 2) World energy use is expected to grow - growth necessary to lift billions of people out of poverty 3) 80% is generated by burning fossil fuels  climate change & debilitating pollution - which won’t last for ever *See Sir Chris Llewellyn-Smith, FPA, October 11, 2005 Need major new (clean) energy sources - requires new technology

Future Energy Use The International Energy Agency (IEA) expects the world’s energy use to increase 60% by 2030 (while population expected to grow from 6.2B to 8.1B) - driven largely by growth of energy use and population in India (current use = 0.7 kWs/person, vs. OECD average of 6.2 kWs/person) and China (current use = 1.3 kWs/person) Strong link between energy use and the Human Development Index (HDI ~ life expectancy at birth + adult literacy and school enrolment + gross national product per capita at purchasing power parity) – need increased energy use to lift billions out of poverty

Carbon dioxide levels over the last 60,000 years - we are provoking the atmosphere! Source University of Berne and National Oceanic, and Atmospheric Administration

Meeting the Energy Challenge Requires: Fiscal measures to change the behaviour of consumers, and provide incentives to expand use of low carbon technologies Actions to improve efficiency (domestic, transport, …) Use of renewables where appropriate (although locally useful, not hugely significant globally) BUT only four sources capable in principle of meeting a large fraction of the world’s energy needs: Burning fossil fuels (currently 80%) - develop & deploy CO2 capture and storage Solar - seek breakthroughs in production and storage Nuclear fission - hard to avoid if we are serious about reducing fossil fuel burning (at least until fusion available) Fusion - with so few options, we must develop fusion as fast as possible, even if success is not 100% certain

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

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

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

ITER Location Caradache (France)Rokkasho (Japan) Cadarache was selected as the ITER construction site. There will be some facilities in Rokassho under the “Broader Approach” agreement.

Fusion Power Station Schematic

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 (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 Materials 1.Tritium Breeding Material (Lithium in some form) Liquid: Li, LiPb ( 83 Pb 17 Li), lithium-containing molten salts Solid: Li 2 O, Li 4 SiO 4, Li 2 TiO 3, Li 2 ZrO 3 2.Neutron Multiplier (for most blanket concepts) Beryllium (Be, Be 12 Ti) 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

Heat and Radiation Loads on First Wall Neutron Wall Load ≡ P nw P nw = Fusion Neutron Power Incident on the First Wall per unit area = J w E o J w = fusion neutron (uncollided) current on the first wall E o = Energy per fusion neutron = MeV Typical Neutron Wall Load ≡ 1-5 MW/m 2 At 1 MW/m 2 : J w = 4.43 x n · m -2 · s -1 Note the neutron flux at the first wall (0-14 MeV) is about an order of magnitude higher than J w Surface heat flux at the first wall This is the plasma radiation load. It is a fraction of the α-power q w = 0.25 P nw · f α where f is the fraction of the α-power reaching the first wall (note that the balance, 1 – f, goes to the divertor)

Tritium Breeding Blankets: Complex component submitted to very severe working conditions, Needed in DEMO, not present in ITER ► ITER is a unique opportunity to test breeding blanket mock-ups: Test Blanket Modules (TBMs) ► It is an ITER mission : “ITER should test tritium breeding module concepts that would lead in a future reactor to tritium self-sufficiency, the extraction of high grade heat and electricity production.” (ITER SWG Report to the IC)  TBMs have to be representative of a DEMO breeding blanket, capable of ensuring tritium-breeding self-sufficiency using high-grade coolants for electricity production ► The ITER TBM Program is therefore a central element in the plans of all seven ITER Parties for the development of tritium breeding and power extraction technology.

What is the ITER TBM Program? Integrated testing of breeding blanket and first wall components and materials in a Fusion Environment Breeding Blankets/FWs will be tested in ITER, starting on Day One, by inserting Test Blanket Modules (TBMs) 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 TBMs. (Number of Ports reduced to 3. Number of Parties increased to 7). ITER’s construction plan includes specifications for TBMs because of impacts on space (port, port area, hot cell, TCWS), shielding, vacuum vessel, remote maintenance, ancillary equipment, safety, availability, etc. The ITER Test Program is managed by the ITER Test Blanket Working Group (TBWG) with participants from the ITER International Team and representatives of the Parties. (However, this entity may change under the new international agreement being negotiated.)

Available TBM development time is limited: ITER Schedule calls for TBM testing from Day 1 of H-H plasma Operation “Day 1” H-H Phase TBM Testing in ITER is required for: –optimization of ITER plasma control, which can only be done with the TBMs installed because the use of ferromagnetic structure and effects on plasma-wall interactions. –licensing for ITER D-T operation, which requires TBM integrity under the severe ITER abnormal loading conditions (e.g. disruption) to be demonstrated in the H-H phase. Schedule from ITER documents show TBM starting from day 1 It is not certain how open ITER will be to fielding unproven TBMs in the D-T phases

TBMs Arrangement in ITER and Interfaces ► 3 ITER equatorial ports (opening of 1.75 x 2.2 m 2 ) devoted to TBM testing ► TBMs installed within a water-cooled steel frame (thk. 20 cm), typically half-port size verticalhorizontal TBM PORTSTBM PORTS The TBMs first wall is recessed 50 mm and protected with a Be layer Shield plug Frame Sample TBM (RF) TBM TBMs tests need a whole TBM system

US ITER Test Blanket Module Concepts  The Dual-Coolant Lead-Lithium (DCLL) and the Helium Cooled Ceramic Breeder (HCCB) concepts have been selected for ITER testing by the US community.  The DCLL is chosen as an innovative concept that provides a “pathway” to higher outlet temperature and higher efficiency while using current generation low-activation ferritic steel (FS) as a structural material and SiC composite only as a non-structural insulator.  The HCCB is chosen as the most likely candidate for near term tritium breeding blankets, e.g. in an extended performance phase of ITER, while providing high grade heat for electricity production.  Both concepts use reduced activation ferritic steel (RAFS) as a structural material and high pressure helium as a coolant. RAFS maximum operating temperature (550C) dictates the maximum helium coolant outlet temperature. DCLL Perspective cutaway Self-cooled Pb-17Li Breeding Zone He-cooled steel structure SiC FCI DCLL Typical Unit Cell Assembly of 3 HCCB sub-modules in 1 half-port Beryllium Pebble Bed Solid Breeder Pebble Bed Coolant Channel

US TBM Program has estimated its costs based on the degree of international collaboration / cost sharing  The high cost range scenario is for an independent US DCLL TBM and an independent HCCB TBM (similar to EU, Japan, and other parties independently testing two full modules.  The baseline scenario consists of (1) an independent US DCLL TBM, and (2) a supporting partnership with another party (Japan or EU) on the HCCB TBM providing only a submodules (size is 1/3 of a module)  The low cost range scenario is defined as a leading international partnership (with one or more ITER Parties) on DCLL TBM and a supporting partnership on the HCCB. All costs over the next 10 years up to shipping of first TBM systems Low - Partnership on both concepts Baseline - Partnership on one concept High - Little Collaboration ITER-TBM Estimated Cost (2006 k$)$60,392$89,676$119,666 Est. Escalation and Contingency$16,927$23,271$30,947 Total Program Cost (escalated k$ including contingency)$77,319$112,947$150,613

Comparison of Total Program Cost Ranges for the US TBM Program (including escalation and contingency)

Tritium Breeding Natural lithium contains 7.42% 6 Li and 92.58% 7 Li. The 7 Li(n;n’  )t reaction is a threshold reaction and requires an incident neutron energy in excess of 2.8 MeV. 6 Li (n,  ) t 7 Li (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: Λ a > Λ r Λ r = Required tritium breeding ratio Λ r 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. Λ r 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. Λ a = Achievable breeding ratio Λ a 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.

Neutron Multipliers Almost all concepts need a neutron multiplier to achieve adequate tritium breeding. (Possible exceptions: concepts with Li and Li 2 O) 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) – Be 12 Ti 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. 9 Be (n,2n) Pb (n,2n) Examples of Neutron Multipliers Beryllium, Lead

Plasma Plasma Facing Component PFC Coolant 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 FuelingFuel management Isotope separation 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 T i : 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

Blanket Concepts (many concepts proposed worldwide) A.Solid Breeder Concepts –Always separately cooled –Solid Breeder: Lithium Ceramic (Li 2 O, Li 4 SiO 4, Li 2 TiO 3, Li 2 ZrO 3 ) –Coolant: Helium or Water B.Liquid Breeder Concepts Liquid breeder can be: a) Liquid metal (high conductivity, low Pr): Li, or 83 Pb 17 Li b) Molten salt (low conductivity, high Pr): Flibe (LiF) n · (BeF 2 ), Flinabe (LiF-BeF 2 -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 (Li 4 SiO 4, Li 2 TiO 3, Li 2 O, 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)

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, 4 He)T Purge gas composition: He + 0.1% H 2 Tritium release composition: T 2, HT, T 2 O, HTO Breeder pebble

“Temperature Window” for Solid Breeders The operating temperature of the solid breeder is limited to an acceptable “temperature window”: T min – T max –T min, 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 –T max, 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

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)

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, 83 Pb 17 Li High conductivity, low Pr number Dominant issues: MHD, chemical reactivity for Li, tritium permeation for LiPb b)Molten Salt: Flibe (LiF) n · (BeF 2 ), Flinabe (LiF-BeF 2 -NaF) Low conductivity, high Pr number Dominant Issues: Melting point, chemistry, tritium control

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: –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

Large MHD drag results in large MHD pressure drop Net JxB body force  p = c  VB 2 where c = (t w  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 wallsInsulated wall

Li/Vanadium Blanket Concept Breeding Zone (Li flow) Primary Shield Secondary Shield Reflector Li Secondary shield (B 4 C) Primary shield (Tenelon) Reflector Lithium 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 wallInsulating 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 5  m/year at 650ºC Tritium recovery and control Vanadium alloy development is very costly and requires a very long time to complete

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

WHAT IS FCI ? FCI (Flow Channel Insert) is the key element of the DCLL blanket concept Both ITER and DEMO Made of 5-10 mm SiC f /SiC composite Pressure equalization openings (slot or holes) to nearly eliminate primary stress. Secondary (thermal) stress still exists The main functions are: - to reduce the MHD pressure drop (electrical insulation); - to reduce heat leakage into He (thermal insulation); - to separate hot PbLi (650  C) from Fe No serious feasibility issues have been identified yet. However tailoring SiC properties and fabrication of complex shape FCIs is still an issue. Pb-17Li exit temperature can be significantly higher than the operating temperature of the steel structure => High Efficiency

SiC f /SiC FCI REQUIREMENTS   SiC =1-100 S/m: reduction of MHD pressure drop k  SiC =1-10 W/m-K: heat leakage is <10% of the total power (DEMO) The optimal (   SiC,k  SiC )* is strongly dependent on the thermofluid MHD and should be determined by design tradeoffs, taking into account: -  P (<1-2 MPa) - heat leakage (<10-15% of the total power) - temperature gradient (< K per 5 mm FCI) - PbLi-Fe interface temperature (<  C) Suggested (DEMO): k  SiC ~2 W/m-K;   SiC ~100 S/m (S.Smolentsev, N.Morley, M.Abdou, MHD and Thermal Issues of the SiCf/SiC FCI, FST, July 2006 ) * Only k  and   (across the FCI) are important

FCI RELATED R&D Material science Development of low-conductivity grade 2-D woven SiC f /SiC with a thin surface sealing layer to avoid soaking of PbLi into pores (e.g. using CVD) Improvement of crack resistance Reliable measurements of SiC f /SiC properties at 300 to 800  C, including effect of irradiation Fabrication of complex shape FCIs with pressure equalization openings and overlap sections Thermofluid MHD Effectiveness of FCI as electrical and thermal insulator Pressure equalization (slot or holes ?) Effect of FCI on flow balancing in normal and abnormal (cracked FCI) conditions Optimal location of the FCIs in the module

THERMOFLUID MHD ANALYSIS recommended 5 mm FCI (  SiC ~10) reduces the MHD pressure drop by factor of 100 Reduction of MHD pressure drop by FCI Heat transfer is strongly affected by  SiC via flow modification Radial temperature distribution in the poloidal duct (DEMO) k SiC =2 W/m-K

MHD PHENOMENA in DCLL 1 Effectiveness of FCI as electrical/thermal insulator MHD pressure drop and flow balancing Buoyancy effects and 2-D MHD turbulence, and their effect on thermal behavior of the module US DEMO DCLL blanket module Intensive studies, including modeling and experiment, are being conducted at UCLA

MHD PHENOMENA in DCLL 2 The flow in a poloidal duct is strongly affected by cross-sectional currents Interaction of the currents with a toroidal field results in a jet-type flow Velocity profile and cross-sectional currents High velocity jets Flow in the gap

MHD PHENOMENA in DCLL 3 Strong influence of buoyancy effect and 2- D MHD turbulence on heat transfer Reduction of high velocity jets due to turbulent diffusion About 10-time increase in effective thermal conductivity Circulation motion LaminarTurbulent Reduction of high velocity jets in a turbulent flow Circulation motion

Many liquid wall reactor concepts for high power density were conceived & analyzed in APEX Thin liquid wall concept (blanket region behind LW not shown)  Many candidate liquids were studied: Li, Sn-Li, Sn, Flibe and Flinabe  Several liquid wall flow schemes were conceived: –Thick liquid walls –Thin fast flowing protection layer (CLIFF) –Inertial or EM assisted wall adhesion –Integrated or stand-alone divertors  Concept performance was analyzed from many perspectives –Liquid wall flow MHD and heat transfer –Breeding, shielding and activation potential –Simplicity of system design, maintenance  Interactions of liquid walls with plasma operation were emphasized –Plasma edge effects, impurities & recycling –Liquid metal motion coupling to plasma modes Surface Renewal Divertor Cassette Fast Flow Cassette Outboard Fast Flow Inboard Fast Flow Bottom Drain Flow

HIMAG simulation of the above experiment Flow Velocity : 3 m/s Average surface normal field gradient: 0.6 T/m New simulation tools and experimental facilities used to address flowing liquid metals in NSTX divertor fields – now being applied to DCLL-TBM  New phenomenon observed in both experiments and numerical simulation for film flows in NSTX divertor: the liquid film tends to ‘pinch in’ away from the wall under a positive surface normal magnetic field gradient. ‘Pinching in’ Gallium flow experiment at UCLA M-TOR facility  Simulation with MHD research code (at UCLA) shows tendency for strong reversed flow jets near slot or crack in flow channel insert (MTOR experiments in development) PbLi FCI

HyPerComp Incompressible MHD solver for Arbitrary Geometry The primary objective of HIMAG is to model the flow of liquid metals in nuclear fusion reactor design. These flows are characterized by very high magnetic field strength, complex geometry and fluid-solid coupling via electric current, heat and mass transfer. There is no other code (commercial or otherwise,) which is capable of producing accurate and reliable solutions in such flows.

 HIMAG is a parallel, unstructured mesh based MHD solver.  High accuracy at high Hartmann numbers is maintained even on non- orthogonal meshes  HIMAG can model single-phase as well as two-phase (free surface) flows  Multiple conducting solid walls may be present in the computational domain  Graphical User Interfaces are provided for the full execution of HIMAG  Heat transfer, natural convection, temperature dependent properties can be modeled  Extensive validation and benchmarking has been performed for canonical problems. Cases involving Ha > 1000 have never been demonstrated on non-rectangular meshes prior to HIMAG HIMAG: Technical Summary

Very high Hartmann number (>10,000) computed and verified Natural convection with MHD streamlines (above), current lines (below), against temperature contours Cylinder flow with MHD, Ha = 1000 HIMAG: Single-phase flow studies

HIMAG: Two-phase flow studies Plasma-liquid interaction Flows with large deformation Droplets and other 3-D phenomena Flows past obstacles, and Complex geometries Liquid metal jet in a magnetic field

 HIMAG is being actively used to assist current research projects (ALPS, ITER-TBM)  HIMAG can readily be applied to metallurgical processes which use magnetic fields e.g., continuous casting of Al and Steel  HIMAG is being extended to model physical phenomena usable in micro and nano flows where surface tension and capillarity determine fluid motion and EM interactions are important  HIMAG is being extended to model phase-change in the form of boiling, and solidification. Various industrial processes require this capability  HyPerComp is developing a multi-physical simulation capability with HIMAG being a key ingredient, in the modeling of flows in nuclear engineering, and aerospace (A “Virtual Test Blanket Module” using HIMAG is in the planning stage) HIMAG: Survey of applications

Summary The D-T Fusion process offers the promise of: –Virtually unlimited energy source from cheap abundant fuels; –No atmospheric pollution of greenhouse and acid rain gases; –Low radioactive burden from waste for future generations. Tremendous Progress has been achieved over the past decades in plasma physics and fusion technology. Fusion R&D involves many challenging areas of physics and technologies and is carried out through extensive international collaboration EU, JA, USA, RF, PRC, Korea, and India are about to construct ITER to demonstrate the scientific and technological feasibility of fusion energy (ITER will produce 500MW of fusion power )