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1 Fusion Nuclear Technology: Principles and Challenges Mohamed Abdou (web: http://www.fusion.ucla.edu/abdou/)http://www.fusion.ucla.edu/abdou/ Distinguished Professor of Engineering and Applied Science Director, Center for Energy Science and Technology (CESTAR) (http://www.cestar.seas.ucla.edu/)http://www.cestar.seas.ucla.edu/ Director, Fusion Science and Technology Center (http://www.fusion.ucla.edu/)http://www.fusion.ucla.edu/) University of California, Los Angeles (UCLA) Seminar at Grenoble, France March 23, 2007 Slide 2 2 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 Slide 3 3 ITER The World is about to construct the next step in fusion development, a device called ITER ITER will demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes ITER will produce 500 MW of fusion power Cost, including R&D, is 15 billion dollars Slide 4 4 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 Slide 5 5 ITER is a collaborative effort among Europe, Japan, US, Russia, China and South Korea Slide 6 6 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 Slide 7 7 The Deuterium-Tritium (D-T) Cycle World Program is focused on the D-T cycle (easiest to ignite): D + T n + + 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 Slide 8 8 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 Slide 9 9 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 Slide 10 10 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 Slide 11 11 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) Slide 12 12 phase I ARIES-CS HCCB Breeder Unit to be inserted into the space between the grid plates EU HCPB DEMO Module First Wall (RAFS, F82H) Tritium Breeder Li 2 TiO 3, Li 2 O (200 MPa will result just to remove nuclear heat Higher stress values will result when one considers the real effects of: 3D features like flow distribution and collection manifolds First wall cooling likely requiring V ~ 1 m/s Allowable Marginal Unacceptable ARIES-RS Slide 24 24 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 walls Slide 25 25 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 wallsInsulating 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 550C and do not support the BCSS reported corrosion limit of 5 m/year at 650C Tritium recovery and control Vanadium alloy development is very costly and requires a very long time to complete Slide 26 26 Proposed US DCLL TBM Cutaway PbLi Flow Channels He-cooled First Wall PbLi He SiC FCI 2 mm gap US DCLL TBM Cutaway Views 484 mm Slide 27 27 Flow Channel Inserts are a critical element of the high outlet temperature DCLL FCIs are roughly box channel shapes made from some material with low electrical and thermal conductivity SiC/SiC composites and SiC foams are primary candidate materials They will slip inside the He Cooled RAFS structure, but not be rigidly attached They will slip fit over each other, but not be rigidly attached or sealed FCIs may have a thin slot or holes in one wall to allow better pressure equalization between the PbLi in the main flow and in the gap region FCIs in front channels, back channels, and access pipes will be subjected to different thermal and pressure conditions; and will likely have different designs and thermal and electrical property optimization Slide 28 28 Example: Interaction between MHD flow and FCI behavior are highly coupled and require fusion environment PbLi flow is strongly influenced by MHD interaction with plasma confinement field and buoyancy-driven convection driven by spatially non-uniform volumetric nuclear heating Temperature and thermal stress of SiC FCI are determined by this MHD flow and convective heat transport processes Deformation and cracking of the FCI depend on FCI temperature and thermal stress coupled with early- life radiation damage effects in ceramics Cracking and movement of the FCIs will strongly influence MHD flow behavior by opening up new conduction paths that change electric current profiles Simulation of 2D MHD turbulence in PbLi flow FCI temperature, stress and deformation Similarly, coupled phenomena in tritium permeation, corrosion, ceramic breeder thermomechanics, and many other blanket and material behaviors Slide 29 Abdou Lecture 4 Key DCLL parameters in three blanket scenarios ParameterDEMOITER H-HITER D-T Surface heat flux, Mw/m 2 0.550.3 Neutron wall load, Mw/m 2 3.08-0.78 PbLi In/Out T, C 500/700470/~450360/470 2a x 2b x L, m 0.22x0.22x20.066x0.12x1.6 PbLi velocity, m/s 0.060.1 Magnetic field, T 444 Re 61,00030,500 Ha 11,6406350 Gr 3.52 10 12 7.22 10 9 *Velocity, dimensions, and dimensionless parameters are for the poloidal flow. DEMO parameters are for the outboard blanket. Slide 30 30 MHD / Heat Transfer analysis (DEMO) Parametric analysis was performed for poloidal flows to access FCI effectiveness as electric/thermal insulator SiC =1-500 S/m, k SiC =2-20 W/m-K Strong effect of SiC on the temperature field exists via changes in the velocity profile FCI properties were preliminary identified: SiC ~100 S/m, k SiC ~2 W/m-K FS GAP FCI Pb-Li 100 S/m 20 S/m FCI = 5 S/m Temperature Profile for Model DEMO Case k FCI = 2 W/m-K Slide 31 31 Molten Salt Blanket Concepts Lithium-containing molten salts are used as the coolant for the Molten Salt Reactor Experiment (MSRE) Examples of molten salt are: Flibe: (LiF) n (BeF 2 ) Flinabe: (LiF-BeF 2 -NaF) The melting point for flibe is high (460C for n = 2, 380C for n = 1) Flinabe has a lower melting point (recent measurement at SNL gives about 300C) Flibe has low electrical conductivity, low thermal conductivity Slide 32 32 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) Slide 33 Abdou Lecture 4 Summary of MHD issues of liquid blankets MHD issueSelf-cooled LM blanket Dual-coolant LM blanket He-cooled LM blanket MS blanket MHD pressure drop and flow distribution Very importantImportantInlet manifold ?Not important Electrical insulation Critical issueRequired for IB blanket Not needed ?Not needed MHD turbulence Cannot be ignored Important. 2-D turbulence Not applicableCritical issue (self-cooled) Buoyancy effects Cannot be ignored ImportantMay effect tritium transport Have not been addressed MHD effects on heat transfer Cannot be ignored ImportantNot importantVery important (self-cooled) Slide 34 34 Issues and R&D on Liquid Metal Breeder Blankets Fabrication techniques for SiC Inserts MHD and thermalhydraulic experiments on SiC flow channel inserts with Pb-Li alloy Pb-Li and Helium loop technology and out-of- pile test facilities MHD-Computational Fluid Dynamics simulation Tritium permeation barriers Corrosion experiments Test modules design, fabrication with RAFS, preliminary testing Instrumentation for nuclear environment Slide 35 35 Fusion environment is unique and complex: multi-component fields with gradients Neutrons (fluence, spectrum, temporal and spatial gradients) Radiation Effects (at relevant temperatures, stresses, loading conditions) Bulk Heating Tritium Production Activation Heat Sources (magnitude, gradient) Bulk (from neutrons and gammas) Surface Synergistic Effects Combined environmental loading conditions Interactions among physical elements of components A true fusion environment is ESSENTIAL to Activate mechanisms that cause prototypical coupled phenomena and integrated behavior Particle Flux (energy and density, gradients) Magnetic Field (3-component with gradients) Steady Field Time-Varying Field Mechanical Forces Normal/Off-Normal Thermal/Chemical/Mechanical/ Electrical/Magnetic Interactions Multi-function blanket in multi-component field environment leads to: -Multi-Physics, Multi-Scale Phenomena Rich Science to Study - Synergistic effects that cannot be anticipated from simulations & separate effects tests. Even some key separate effects in the blanket can not be produced in non-fusion facilities (e.g. volumetric heating with gradients) Slide 36 36 ITER Provides Substantial Hardware Capabilities for Testing of Blanket System Vacuum Vessel Bio-shield A PbLi loop Transporter located in the Port Cell Area He pipes to TCWS 2.2 m TBM System (TBM + T-Extrac, Heat Transport/Exchange) Equatorial Port Plug Assy. TBM Assy Port Frame ITER has allocated 3 ITER equatorial ports (1.75 x 2.2 m 2 ) for TBM testing Each port can accommodate only 2 Modules (i.e. 6 TBMs max) But, 12 modules are proposed by the parties Aggressive competition for Space Slide 37 37 Fusion Nuclear Technology Issues 1.Tritium Supply & Tritium Self-Sufficiency 2.High Power Density 3.High Temperature 4.MHD for Liquid Breeders / Coolants 5.Tritium Control (Permeation) 6.Reliability / Maintainability / Availability 7.Testing in Fusion Facilities Challenging Slide 38 38 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: a > r r = Required tritium breeding ratio 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. r is dependent on many system parameters and features such as plasma edge recycling, tritium fractional burn up in the plasma, tritium inventories, doubling time, efficiency/capacity/reliability of the tritium processing system, etc. a = Achievable breeding ratio 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. Slide 39 39 t d = doubling time Current physics and technology concepts lead to a narrow window for attaining Tritium self-sufficiency t d =10 yr t d =5 yr t d =1 yr Window for Tritium self sufficiency Max achievable TBR 1.15 Required TBR Fractional burn-up [%] Fusion power 1.5GW Reserve time2 days Waste removal efficiency0.9 (See paper for details) Slide 40 40 The D-T fuel cycle includes many components whose operation parameters and their uncertainties impact the required TBR Examples of key parameters: : 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 Only for solid breeder or liquid breeder design using separate coolant 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 liquid breeder as coolant design Slide 41 41 Slide 42 42 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) Slide 43 43 Slide 44 44 Slide 45 45 Fission (PWR) Fusion structure Coal Tritium in fusion Slide 46 46 Lessons learned: The most challenging problems in FNT are at the INTERFACES Examples: MHD insulators Thermal insulators Corrosion (liquid/structure interface temperature limit) Tritium permeation Research on these interfaces must be done jointly by blanket and materials researchers Slide 47 47 Need for High Power Density Capability A.To improve potential attractiveness of fusion power compared to other energy sources (e.g., fission) B.Some plasma confinement schemes have the potential to produce burning plasmas with high power density PWRBWRLMFBRITER-Type Average core power density (MW/m 3) 96562400.4 FW/Blanket/Divertor concepts developed in the 1970s and 80s have limitations on power density capability (wall load and surface heat flux) Substantial PROGRESS has been made in this area over the past several years (e.g. in the APEX Study in the US ) Slide 48 48 Reliability/Maintainability/Availability is one of the remaining Grand Challenges to Fusion Energy Development. Chamber Technology R&D is necessary to meet this Grand Challenge. timereplacementrate failure/1 )rate failure/1( Need Low Failure Rate: - Innovative Chamber Technology Need Short Maintenance Time: - Simple Configuration Confinement - Easier to Maintain Chamber Technology Need Low Failure Rate Energy Multiplication Need High Temp. Energy Extraction Need High Power Density/Physics-Technology Partnership -High-Performance Plasma -Chamber Technology Capabilities thfusion MP MOiC COE Availability & replacement cost Need High Availability / Simpler Technological and Material Constraints Slide 49 49 Availability = Current plasma confinement schemes and configurations have: Relatively long MTTR (weeks to months) Required MTBF must be high Large first wall area Unit failure rate must be very low MTBF ~ 1/(area unit failure rate) MTBF MTBF + MTTR Reliability requirements are more demanding than for other non-fusion technologies Slide 50 50 The reliability requirements on the Blanket/FW (in current confinement concepts that have long MTTR > 1 week) are most challenging and pose critical concerns. These must be seriously addressed as an integral part of the R&D pathway to DEMO. Impact on ITER is predicted to be serious. It is a DRIVER for CTF. Slide 51 51 Upper statistical confidence level as a function of test time in multiples of MTBF for time terminated reliability tests (Poisson distribution). Results are given for different numbers of failures. TYPICAL TEST SCENARIO Reliability Growth Example, To get 80% confidence in achieving a particular value for MTBF, the total test time needed is about 3 MTBF (for case with only one failure occurring during the test). Reference: M. Abdou et. al., "FINESSE A Study of the Issues, Experiments and Facilities for Fusion Nuclear Technology Research & Development, Chapter 15 (Figure 15.2-2.) Reliability Development Testing Impact on Fusion Reactor Availability", Interim Report, Vol. IV, PPG-821, UCLA,1984. It originated from A. Coppola, "Bayesian Reliability Tests are Practical", RADC-TR-81-106, July 1981. Slide 52 52 Summary of Critical R&D Issues for Fusion Nuclear Technology 1.D-T fuel cycle tritium self-sufficiency in a practical system depends on many physics and engineering parameters / details: e.g. fractional burn-up in plasma, tritium inventories, FW thickness, penetrations, passive coils, etc. 2. Tritium extraction and inventory in the solid/liquid breeders under actual operating conditions 3.Thermomechanical loadings and response of blanket and PFC components under normal and off-normal operation 4.Materials interactions and compatibility 5.Identification and characterization of failure modes, effects, and rates in blankets and PFCs 6.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 components 9. Remote maintenance with acceptable machine shutdown time Slide 53 53 Initial exploration of coupled phenomena in a fusion environment Uncover unexpected synergistic effects, Calibrate non-fusion tests Impact of rapid property changes in early life Integrated environmental data for model improvement and simulation benchmarking Develop experimental techniques and test instrumentation Screen and narrow the many material combinations, design choices, and blanket design concepts Uncover unexpected synergistic effects coupled to radiation interactions in materials, interfaces, and configurations Verify performance beyond beginning of life and until changes in properties become small (changes are substantial up to ~ 1-2 MW y/m 2 ) Initial data on failure modes & effects Establish engineering feasibility of blankets (satisfy basic functions & performance, up to 10 to 20 % of lifetime) Select 2 or 3 concepts for further development Identify lifetime limiting failure modes and effects based on full environment coupled interactions Failure rate data: Develop a data base sufficient to predict mean-time- between-failure with confidence Iterative design / test / fail / analyze / improve programs aimed at reliability growth and safety Obtain data to predict mean-time-to- replace (MTTR) for both planned outage and random failure Develop a data base to predict overall availability of FNT components in DEMO ITER TBM is the Necessary First Step to enable future Engineering Development Sub-Modules/ Modules Size Tests ~ 0.3 MW-y/m 2 Stage I Fusion Break-in & Scientific Exploration Stage II Stage III Engineering Feasibility & Performance Verification Component Engineering Development & Reliability Growth Module Size Tests Module / Sectors Size Tests DEMODEMO Role of ITER TBM 1 - 3 MW-y/m 2 > 2 - 4 MW-y/m 2 Slide 54 54 Tritium breeding capabilities are needed for the continued operation of successful ITER and fusion development A Successful ITER will exhaust most of the world supply of tritium ITER extended performance phase and any future long pulse burning plasma will need tritium breeding technology TBMs are critical to establishing the knowledge base needed to develop even this first generation of breeding capability 0 5 10 15 20 25 30 19952000 2005 20102015202020252030203520402045 Year Projected Ontario (OPG) Tritium Inventory (kg) CANDU Supply w/o Fusion 1000 MW Fusion 10% Avail, TBR 0.0 ITER-FEAT (2004 start) TBM helps solve the Tritium Supply Issue for fusion development (at a fraction of the cost of purchasing tritium from fission reactor sources!) Tritium Consumption in Fusion is HUGE! Unprecedented! 55.8 kg per 1000 MW fusion power per year Production & Cost: CANDU Reactors: 27 kg from over 40 years, $30M/kg (current) Fission reactors: 23 kg/year/reactor, $84M-$130M/kg (per DOE Inspector General*) *www.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdfwww.ig.energy.gov/documents/CalendarYear2003/ig-0632.pdf Slide 55 55 BACKUP SLIDES Slide 56 56 Will the development of high-temperature structural material lead to more attractive blankets? Not necessarily (unless we can solve the interface problems) 1.Vanadium alloys (high temperature capability) V is compatible only with liquid lithium Flowing liquid Li requires MHD insulators Tolerable crack fraction is estimated to be very low (< 10 -7 )much lower than can be achieved with real coatings Self-healing coating R&D results are negative for non-isothermic systems Laminated layer insulators (alternating layers of insulator and metallic protection layer) were proposed, but V layer needs to be too thin 2.High-temperature advanced ferritic steels (ODS, NFS) May potentially operate up to ~700C (compared to 550C for EUROFER and F82H) At present, we cannot utilize such advanced high-temperature ferritics Li and LiPb interface temperature (T int ) is limited by corrosion to ~500C Unless corrosion temperature limit is improved, EUROFER and F82H are satisfactory Solid breeder/structure interface cannot be increased much above 400 500C (to have adequate temperature window for T-release and TBR) Slide 57 57 Tritium Control and Management Tritium control and management will be one of the most difficult issues for fusion energy development, both from the technical challenge and from the public acceptance points of view. Experts believe the T-control problem is underestimated (maybe even for ITER!) The T-control problem in perspective: The scale-up from present CANDU experience to ITER and DEMO is striking: The quantity of tritium to be managed in the ITER fuel cycle is much larger than the quantities typically managed in CANDU (which represents the present-day state of practical knowledge). The scale-up from ITER to DEMO is orders of magnitude: The amount of tritium to be managed in a DEMO blanket (production rate ~400 g/day) is several orders of magnitude larger than that expected in ITER, while the allowable T-releases could be comparable. For more details, see: W. Farabolini et al, Tritium Control Modelling in an He-cooled PbLi Blanket paper in ISFNT-7 (this conference) Papers and IEA Reports by Sze, Giancarli, Tanaka, Konys, etc. Slide 58 58 Why is Tritium Permeation a Problem? Most fusion blankets have high tritium partial pressure: LiPb = 0.014 Pa Flibe = 380 Pa He purge gas in solid breeders = 0.6 Pa The temperature of the blanket is high (500700C) Surface area of heat exchanger is high, with thin walls Tritium is in elementary form These are perfect conditions for tritium permeation. The allowable tritium loss rate is very low (~10 Ci/day), requiring a partial pressure of ~10 -9 Pa. Challenging! Even a tritium permeation barrier with a permeation reduction factor (PRF) of 100 may be still too far from solving this problem! Slide 59 59 Pillars of a Fusion Energy System 1.Confined and Controlled Burning Plasma (feasibility) 2.Tritium Fuel Self-Sufficiency (feasibility) 3.Efficient Heat Extraction and Conversion (attractiveness) 4.Safe and Environmentally Advantageous (feasibility/attractiveness) 5.Reliable System Operation (attractiveness) Yet, No fusion blanket has ever been built or tested! The Blanket is THE KEY component Slide 60 60 Structure of TBM collaboration* Each 1/2 of a port is dedicated to testing of one TB concept design (one module or several connected sub-modules). Each Concept design is tested by a partnership of a TB concept leader + supporting partners One of the two TB leaders occupying the same port must play a role of the Port master responsible for integration of the given port Port Master has main responsibility of integration of 2 concepts in the same port frame and preparation of the integrated testing program (replacement strategy) for the port ITER has only 3 ports Only 6 TBMs can be tested 750 1248 Horizontal 524 1700 Vertical Each Port can hold two vertically or horizontally oriented half-port size integrated TBMs List of TBM Design Proposals for day-one (DDDs completed by Parties) Helium-cooled Lithium-Lead TBM (2 designs) Dual-coolant (He+Lithium-Lead) TBM (2 designs) He-cooled Ceramic Breeder/Beryllium multiplier TBM (4 designs) Water-cooled Ceramic Breeder/Beryllium multiplier TBM (1 design) He-cooled Liquid Lithium TBM (1 design) Self-cooled Liquid Lithium TBM (1 design) * Initial result of TBWG DH meeting, 18-19 July 2006 Slide 61 61 US advanced Ideas to Solve the TBM testing-space problem in ITER, while improving effectiveness of the tests KO Submodule JA Submodule US Submodule US experiments can focus on niche areas of US interest and expertise, While benefiting from world resources and testing results Different sub-modules can be tailored to address the many different: design configurations material options operating conditions (such as flow rates, temperatures, stresses) diagnostics and experimental focus The back plate coolant supply and collection manifold assembly incorporates all connections to main support systems. A Lead Party takes responsibility for back plate and sub-module integration Slide 62 62 0 5 10 15 20 25 30 20002010202020302040 Year Projected Ontario (OPG) Tritium Inventory (kg) CANDU Supply w/o Fusion Projections for World Tritium Supply Available to Fusion Reveal Serious Problems World Max. tritium supply is 27 kg Tritium decays at a rate of 5.47% per year Tritium Consumption in Fusion is HUGE! Unprecedented! 55.8 kg per 1000 MW fusion power per year Production & Cost: CANDU Reactors: 27 kg from over 40 years, $30M/kg (current) Fission reactors: 23 kg per year. It takes tens of fission reactors to supply one fusion reactor. $84M-$130M per kg, per DOE Inspector General* *DOE Inspector Generals Audit Report, Modernization of Tritium Requirements Systems, Report DOE/IG-0632, December 2003, available at www.ig.doe.gov/pdf/ig-0632.pdf Slide 63 63 Tritium Consumption in ITER Here is from a summary of the final design report. Link is: http://fusion.gat.com/iter/iter-ga/images/pdfs/cost_estimates.pdf http://fusion.gat.com/iter/iter-ga/images/pdfs/cost_estimates.pdf 9.4.3 Fuel Costs The ITER plant must be operated, taking into account the available tritium externally supplied. The net tritium consumption is 0.4 g/plasma pulse at 500 MW burn with a flat top of 400 s The total tritium received on site during the first 10 years of operation, amounts to 6.7 kg. whereas the total consumption of tritium during the plant life time may be up to 16 kg to provide a fluence of 0.3 MWa/m2 in average on the first wall This corresponds, due to tritium decay, to a purchase of about 17.5 kg of tritium. This will be well within, for instance, the available Canadian reserves. Slide 64 64 It is important to precisely understand the state-of-the- art in blanket and material research, and the role of ITER Over the past 30 years the fusion nuclear technology and materials programs have spent much effort on developing theories and models of phenomena and behavior. But, these are for idealized conditions based on understanding of single effects. There has never been a single experiment in a fusion environment! Are they scientifically valid? ITER will provide the 1 st opportunity to test these theories and models in a real fusion environment. Only the Parties who will do successful, effective TBM experiments in ITER will have the experimentally-validated scientific basis to embark on the engineering development of tritium breeding blankets Slide 65 65 Another Look at the DCLL Unit Cell Slide 66 66 A Simplified DCLL DEMO System Blanket Module Tritium Extraction Pump Heat Exchanger Cold Trap, Chem. Control 450C He 350C He 450C 650C From/To Tritium Processing System From/To Helium Loops and Brayton Cycle Power Conversion System Coaxial Feed Pipes PbLi Hot leg flows in inner pipe (700C) PbLi Cold leg flows in outer annulus (450C) Cold leg cools Pipe walls and TX/HX shells Slide 67 67 Coolant Routing Through HX Coupling Blanket and Divertor to Brayton Cycle Fusion Thermal Power in Reactor Core2650 MW Fusion Thermal Power in Pb-17Li1420 MW Fusion Thermal Power in Blkt He1030 MW Friction Thermal Power in Blkt He119 MW Fusion Thermal Power in Div He201 MW Friction Thermal Power in Div He 29 MW Total Power 2790 MW Overall Brayton Cycle Efficiency 0.42 Power Parameters for DCLL in ARIES-CS Blkt He Typical Fluid Temperatures in HX for DCLL ARIES-CS Blkt LiPb Blkt LiPb (711C) + Div He (711C) Cycle He ~681C 580C 369C 441C 452C 349C T Z HX Pb-17Li from Blanket He from Divertor He from Blanket Brayton Cycle He T HX,out He T HX,in Blkt He T in Blkt He T out (P th,fus +P frict ) Blkt,He (P th,fus ) Blkt,LiPb LiPb T in LiPb T out Div He T in Div He T out (P th,fus +P frict ) Div,He Slide 68 68 Blanket systems are complex and have many integrated functions, materials, and interfaces Tritium Breeder Li 2 TiO 3 (