US ITER DCLL TBM Design US ITER TBM
US ITER DCLL TBM Design US ITER TBM Presented by Clement Wong DCLL TBM ASSESSMENT AND DESIGN TEAM U. of California, Los Angeles M. Abdou, M. Dagher, S. Smolentsev S. Sharafat, N. Morley, M. Youssef, A. Ying General Atomics C. P. C. Wong, D. Carosella, C. Baxi Consultant S. Malang, A. Rowcliffe, D. Sonn, S. Tourville U. of Wisconsin, Madison M. Sawan, G. Sviatoslavsky Oak Ridge National Laboratory P. Fogarty, Y. Katoh, B. Pint, T. Mann, S. J. Zinkle Idaho National Laboratory B. Merrill Pacific Northwest Laboratory R. J. Kurtz University of California, San Diego D. K. Sze Lawrence Livermore National Laboratory S. Reyes Los Alamos National Laboratory R. S. Willms Sandia National Laboratory R. Nygren, T. Tanaka, D. Youchison, M. Ulrickson 2006 US-Japan Workshop on Fusion High Power Density Components and System Inn on the Alameda, Santa Fe, New Mexico, USA November 15 -17, 2006 DCLL-TBM Pb. Li-loop on the transporter He-loops at the 077 -05/rs TCWS
Outline US ITER TBM – Key features of DCLL blanket concepts – Selection of DCLL TBM – R&D items for DCLL DEMO and HH-TBM – Utility of DCLL HH-TBM – DCLL HH-TBM and Ancillary Loops Design – Development Schedule 077 -05/rs
Key Features of Dual Coolant Lead-Lithium (DCLL) Concepts US ITER TBM • Helium is used to cool first wall and reduced activation ferritic steel (RAFS) e. g. F 82 H structure. Helium is also used for first wall/blanket preheat and tritium control. • Breeder is self-cooled Pb. Li moving at a slow velocity < 10 cm/s. – allowing high Tout (700°C) leading to ηth~ 40% (CCGT) EU design, FED, 61 -62, 2002 • Use Si. C flow channel inserts (FCI) to: – provide electrical and thermal insulation to reduce MHD pressure drop and to decouple high temperature Pb. Li bulk flow from cooler RAFS structure, also provide higher corrosion resistance temperature with only near stagnant Pb. Li in contact with the RAFS in the gap US ARIES ST A design in progress: • • Community recommendation to focus on RAFS mid-2002 Selected Pb. Li DCLL blanket as our reference option 12/04 Smaller TBM module recommended by TBWG 5/05 Improved Helium flow from parallel to series between FW and structure 5/06 • Adjustment on back plenum design DCLL 077 -05/rs TBM
DCLL Blanket Has Minimum Critical Issues and Can Become A High Performance DEMO Blanket US ITER TBM Neutron wall loading: 3 MW/m 2, FW Surface Loading: 0. 55 MW/m 2 Advantages of DCLL Concept • No need for separate neutron multiplier, like Be or Pb • No damage to breeder material by thermal effects and/or irradiation • Lower chemically reactive than Li • Self-cooled Pb. Li with velocity~0. 1 m/s to enhance Tout, minimize MHD effect • Flow channel insert (FCI) for MHD and thermal insulation • With Pb. Li Tout @ 700 C, projected CCGT thermal efficiency ~40% for DEMO • A high performance design with minimum critical issues • With ODFS FW layer, blanket performance can be enhanced With RAFS DCLL blanket can satisfy all DEMO design requirements: Nuclear performance, FW helium cooling, Waste disposal, Structural Design Requirements, Safety impacts including LOCA, Power Conversion with CCGT Clear R&D path to DEMO identified 077 -05/rs
DCLL Fusion Reactor (DEMO) Blanket Can Achieve High ηth US ITER TBM Max. neutron wall loading: 3 MW/m 2, max. surface heat flux: 0. 55 MW/m 2 Structural material: RAFS, Tmax@ 550 C, all structure cooled by 8 MPa Helium Breeder material: Pb. Li, MHD and thermal insulator: Si. C flow channel insert (FCI) Power Conversion, “a simple approach” Primary Helium 1. Secondary 2. Helium He Tin/Tout =350/450° C Pb. Li Tin/Tout =550/700° C CCGT 1 turbine 2 compressors @ ηth=26. 5% HX He Tin/Tout =500/650°C CCGT: 3 turbines 6 Compressors @ ηth=48. 7% 40% power from 1, 60% power from 2 combined gross thermal efficiency, ηth ~ 40% 077 -05/rs
DEMO DCLL Blanket R&D Items are Performance Related DCLL TBM HH-Module has Few Critical Issues R&D Items DEMO HH-TBM US ITER TBM Critical: RAFS fabrication development and materials properties Thermal fluid MHD Si. C FCI development Tritium extraction and control Including high fluence effects √…also safety √* √ √ NA DEMO DCLL Blanket Performance: RAFS/Pb. Li compatibility & chemistry √ √ He system subcomponents analyses and tests √ √ Piping connections at the back of the module √ NA Advanced CCGT development √ NA High temperature piping and ancillary equipment materials and design √ NA TBD 2 mm Be TBM diagnostics development √ √ Integrated mockup tests √ √ Development of engineering and materials (RAFS and Si. C) design codes √ NA Plasma facing first wall material Necessary for development: Safety: Pb. Li/H 2 O hydrogen production √ √ *√ means applicable Ferromagnetic effects of TBM to ITER have been analyzed by ITER: Effects from error fields can be corrected, effects on field line perturbation is small, further assessment is needed. ITER DCLL TBM 077 -05/rs
Four Phases of DCLL TBM for the 1 st 10 years in ITER US ITER TBM -1 1 2 3 4 5 6 7 8 9 10 ITER Operation Phase Magnet testing & vacuum HHFirst Plasma HH HH DD Low Duty DT High Duty DT Progressive ITER Testing Conditions Toroidal B field Vacuum Heat flux B Field Disruptions Small DD neutron flux NWL Full disruption energy ITER Year Fluence Accumula -tion • Transient EM Loading on structure Electromagnetic/ Structural (EM/S) TBM • Install • RH • System checkout • • and FCI FW heat flux loading ITERB-field perturbation LM-MHD tests FCI thermal and electrical performance • Nuclear field measurement Nuclear Field/ Tritium Prod. (N/T) TBM • Tritium production Finalize Design and processing • Nuclear heating • Structure and FW heating • FCI thermal and Thermofluid/ MHD (T/M) TBM electrical performance • Tritium transport • Pb. Li MHD flow • Activation product transport and chemistry control Finalize Design • Integrated synergistic effects of Integrated (I) TBM Finalize Design all systems at high temperature • On line tritium recovery and control and coolant purification • Long term operation effects including radiation 077 -05/rs
DCLL TBM HH-module can Simulate Selected Mechanical, MHD and First Wall Thermal Effects US ITER TBM Modules with similar geometry ITER-HH Heat flux=0. 11 MW/m 2 ITER-HH Heat flux=0. 3 MW/m 2 ITER-DT High duty factor Fusion Reactor (DEMO) Dimensions: W x H x depth, m x m 0. 48 x 1. 66 x 0. 41 1. 0 x 2. 0 x 0. 75 Neutron wall loading, MW/m 2 0 0 0. 56 (ave) 0. 78 -1. 09 (peak) 2. 2 (ave) 3. 08 (peak) First wall surface heat loading, MW/m 2 0. 11 (peak) for 600 cycles/year 0. 3 localized MARFE or other 0. 27 -0. 38 for 3000 cycles/y 0. 5 localized for 100 cycles/y 0. 396 (ave) 0. 55 (max) Thermal power of the module, MW 0. 09 0. 24 0. 86 7. 92 Expected burn time, s 100 -200, pulse ~400 Steady state Max. Disruption heat load during current quench 0. 72 MJ/m 2, 40 ms, 180 cycles/y 0. 72 MJ/m 2, 40 ms, 300 cycles/y TBD Heat load from ELM and blob Negligible (TBD) ? 8 MPa, He primary, Tin/Tout C 350/450 Adjustable 350/430 Adjustable 350/420 Adjustable 350/445 Not optimized He heat transfer coefficient, MW/m 2. k 1474 3866 6764 10966 First wall Tmax, C 538 544 555 Outboard mid-plane B, T 4 4 4 3. 5 <2 MPa, Pb. Li Tin/Tout C, T-melt @ 235° C 470/~450 360/470 550/700 Pb. Li channel velocity, m/s Adjustable ~0. 1 077 -05/rs
For DEMO Pb. Li Temperature is Strongly Dependent on σ of Si. C For DT-module Pb. Li/RAFS Interface Temperature <470° C (Integrated Testing in ITER to Establish Predictive Capability) US ITER TBM FCI thickness=0. 005 m, gap width=0. 002 m DEMO: B=4 T, NWL=3. 75 MW/m 2, FW heat flux=0. 5 MW/m 2 Pb. Li: Tin/Tout=460/700°C, V=0. 06 m/s h=4000 W/m. K σ=100 S/m is needed T Distribution in Pb. Li channel at exit for DEMO DT-module: B=4 T, NWL=0. 78 MW/m 2, FW heat flux=0. 5 MW/m 2 Pb. Li: Tin/Tout=360/470°C, V=0. 07 m/s h=4000 W/m. K T Distribution in Pb. Li channel at exit for DT DCLL TBM 077 -05/rs
HH-TBM Design and Ancillary Loops Design US ITER TBM • HH-TBM design • Ancillary Loops design • HH-TBM and Ancillary Loops • Schedule and Key Dates • Summary and Status 077 -05/rs
DCLL TBM Assembly US ITER TBM Top Plate Assembly FW Assembly TBM Frame Electric Strap Back Plate Assembly Flexible Support Assembly Helium Inlet/Outlet Pipes Pb. Li Inlet/Outlet Manifold Assembly DCLL TBM section 077 -05/rs view
DCLL TBM Assembly Exploded View, Major Components US ITER TBM Top Plate Cover Top Plate Assembly DCLL TBM section view Si. C FCI FW Assembly Pb. Li Outlet Manifold Grid Plate Assembly Back Plate Cover Assembly He Inlet Manifold Inner Back Plate Channel Assembly Back Plate & FW distribution Assembly Pb. Li Inlet Manifold Bottom Plate 077 -05/rs
Assembly and Si. C FCI US ITER TBM DCLL TBM section view U-shape FW Grid and diverter plates Si. C components Grid and divertor plates parts Bottom cross-section 077 -05/rs
DCLL TBM Design, Pb. Li Flow Scheme US ITER TBM Pb. Li Concentric Pipe 077 -05/rs
DCLL TBM Design, He Flow Scheme, Circuits/Passes US ITER TBM He Flow Distribution Manifold Pb. Li Flow Channels FW He Channels He inlet Manifold 077 -05/rs
DCLL TBM Design, He Flow Scheme, Inlet Manifold US ITER TBM He flow scheme: Series flow: FW & Bottom Plate to Top Plate to Grid & Divider Plates to Back Plate. First Wall He Flow: Two circuits cross flow, seven passes per circuit, variable number of channels per circuit. FW He Channels DCLL TBM section view He Distribution Manifold He Inlet Manifold He Outlet He Inlet FW counter flow Top plate He distributor to The grid and divider plates He Inlet Manifold 077 -05/rs
Tritium Production US ITER TBM Local TBR in the DCLL TBM is only 0. 741 because of the small thickness (42. 3 cm) Tritium generation rate in the TBM is 2. 1 x 1017 atom/s (1. 02 x 10 -6 g/s) during a DT pulse with 500 MW fusion power For a pulse with 400 s flat top preceded by 100 s linear ramp up to full power and followed by 100 s linear ramp down total tritium generation is 5. 1 x 10 -4 g/pulse For the planned 3000 pulses per year the annual tritium production in the TBM is 1. 54 g/year Tritium production in the Be PFC is 1. 4 x 109 g/s 7 x 10 -7 g/pulse 2. 1 x 10 -3 g/year Peak tritium production rate in Li. Pb is 2. 94 x 10 -8 kg/m 3 s during the D-T pulse 077 -05/rs
DCLL TBM Ancillary Equipment Loops US ITER TBM Test Port Primary He Coolant Loop TCWS Secondary He Coolant Loop TCWS Pb-Li Primary Coolant Loop Transporter, Port Cell Area DCLL TBM coolant circuits, Red-dot circuit shows the primary He loop cooling the first wall and all RAFS structures, Blue-dash circuit shows the Pb-Li loop and the Green-dash circuit shows the 077 -05/rs secondary helium loop.
Ancillary Loops Parameters US ITER TBM Design parameters: First wall surface heat flux 0. 5 MW/m 2 and neutron wall loading @ 0. 78 MW/m 2, module width x height = 0. 645 m x 1. 94 m Primary Helium loop Pb. Li loop Secondary Helium loop Total heat to be removed, MW 0. 57 1. 36 He temperature at module in/out, °C 380/440 340/440 180/300 Pressure, MPa 8 2 8 Mass flow rate, kg/s 1. 82 72 2. 2 Secondary coolant water Helium water Temperature at heat exchanger in/out, °C 35/43 180/300 35/70 17 (water) 2. 2 (helium) 13 (water) Mass flow rate, kg/s These loops are over-designed in power handling, due to earlier larger TBM dimensions and the allowance for testing flexibility He loops Pb. Li loop 077 -05/rs
Pb. Li Loop Schematic US ITER TBM Sump Pump Tank Expansion Tank Cold Trap Unit Tritium Extraction Pressure Control Unit Tritium Line Pb. Li He heat exchanger Dump Tank Pb. Li loop located on the transporter 077 -05/rs
US TBM PRIMARY AND SECONDARY Helium LOOPS TCWS Vault (He loops located in the TCWS Vault) Main design points: • He cooled with 35°C water • He/H 2 O HX with Al tubes to prevent T release • Water cooling to keep tube T<160° C • “U” tubes heat exchanger to reduce thermal stress US ITER TBM Sub-systems: • Helium pipe network • Pressure control unit • 170 k. W electric heater for system and FW heat-up • Circulator • Tritium extraction • Helium purification • Thermal insulation 077 -05/rs
DCLL TBM WBS Definition…costing was done beginning at the lowest WBS level US ITER TBM WBS# Title Responsible Person 1. 8 Test Blanket 1. 8. 1 DCLL TBM System and Testing Goals Wong 1. 8. 1. 1 Test Module 1. 8. 1. 1. 1 WBS# Title Responsible Person 1. 8. 1. 1. 3. 1 Preliminary Design and Analysis, Title I Wong 1. 8. 1. 1. 3. 1. 1 Mechanical Design Dagher Administration Wong 1. 8. 1. 1. 3. 1. 2 FM Steel Engineering and Fabrication Rowcliffe/Kurtz 1. 8. 1. 1. 2 R&D Morley 1. 8. 1. 1. 3 Nuclear Analsyis Sawan/Youssef 1. 8. 1. 1. 3 Engineering Wong 1. 8. 1. 1. 3. 1. 4 Thermofluid MHD Smolentsev 1. 8. 1. 1. 3. 1 Preliminary Design and Analysis, Title I Wong 1. 8. 1. 1. 3. 1. 4. 1 Smolentsev 1. 8. 1. 1. 3. 2 Detailed Design, Title II Wong Preliminary Assessment and Design of Si. C/Si. C FCI 1. 8. 1. 1. 3. 1. 4. 2 Title III Wong Preliminary Assessment and Design of Alternate FCI Smolentsev 1. 8. 1. 1. 3. 3 1. 8. 1. 1. 4 Prototype TBM Design and Fabrication Wong 1. 8. 1. 1. 3. 1. 4. 3 Preliminary Analysis and Design of Pb. Li Manifold Smolentsev 1. 8. 1. 1. 5 Prototype Assembly, Testing & Installation Wong 1. 8. 1. 1. 3. 1. 5 Thermofluid He Sviatoslavsky 1. 8. 1. 1. 6 TBM Fabrication Wong 1. 8. 1. 1. 3. 1. 5. 1 First Wall Thermofluid Analysis Sviatoslavsky 1. 8. 1. 1. 7 Acceptance Tests & Packaging Wong 1. 8. 1. 1. 3. 1. 5. 2 Grid/Top/Bottom/Back Plate Thermofluid Analysis Sviatoslavsky 1. 8. 1. 2 Helium Flow Loops Wong 1. 8. 1. 1. 3. 1. 5. 3 Fluid Distribution Analysis Sviatoslavsky 1. 8. 1. 2. 1 Primary Helium Loop Wong 1. 8. 1. 1. 3. 1. 6 Structural Analysis Sharafat 1. 8. 1. 2. 2 Intermediate helium loop Wong 1. 8. 1. 1. 3. 1. 6. 1 Normal Operation Sharafat 1. 8. 1. 3 Pb. Li Flow Loop Pint 1. 8. 1. 1. 3. 1. 6. 2 Transient Events Sharafat 1. 8. 1. 5 DCLL TBM/ITER System Integration Dagher 1. 8. 1. 1. 3. 1. 6. 3 Disruption Events Sharafat 1. 8. 3 Predictive Capabilities Abdou 1. 8. 1. 1. 3. 1. 7 Diagnostic/Instrumental/control Morley 1. 8. 4. 3 Safety and Regulatory Support Merrill 1. 8. 1. 1. 3. 1. 8 TBM Interface Dagher 1. 8. 4. 4 Quality Assurance Officer Abdou Definition This WBS includes experimental and modeling studies and their associated administration aiming at the design, testing and optimization of the inlet Pb. Li manifold. The goal of the studies is to give recommendations for designing a manifold that provides uniform flow distribution through all its legs, without significant increase in the MHD pressure drop. The reduced scale manifolds will be tested in a moderate (1 -2 T) magnetic field. 077 -05/rs
Primary Helium Loop Details US ITER TBM • Helium loop design were performed by a GA helium system engineers • Including all He-loop equipment, instrumentation, piping and safety systems… etc. • Assuming that the procurement packages had been supplied to qualified vendors in a competitive world market bidding process. • Some are first of a kind equipment. • Manufacturing inspection and qualification standards used were based on safety grade level Costing details include: • Circulator machine assembly • Bypass flow system • Flow control valves • He circulator instrumentation • Temperature instrumentation • Pressure instrumentation • Flow instrumentation • Safety valves • Safety piping • Safety system instrumentation • Helium BOP piping • Inlet and exhaust Piping to TCWS • Helium purge line piping • Helium piping thermal insulation…etc. 077 -05/rs
DCLL TBM Milestones and Key Dates US ITER TBM • • • • Preliminary Design Initiated 31 May 2006 Provide Latest Design Information for RPr. S Analysis 30 Sept 2006 Preliminary Design Midpoint Review 30 June 2007 Preliminary Design Review 30 June 2008 Select Fabrication Route 31 Dec 2008 Fabrication Bid Package Initiated 31 Aug 2009 Start Safety Classification Procedure for Systems Components 31 Oct 2009 Detailed Design Final Design Review 01 Sept 2010 Provide Final Design Specifications to Complete RFS 28 Feb 2011 Title III Design Review - Prototype Fabrication Initiated 30 June 2011 Complete Prototype Fabrication 30 Apr 2012 Final TBM Design Changes 31 Dec 2012 TBM Fabrication Initiated 28 June 2013 Complete TBM Fabrication 30 Apr 2014 Complete TBM Acceptance Tests 30 Jan 2015 TBM ready for shipping 31 Mar 2015 st (18 months before ITER 1 plasma) Blue ones are necessary inputs to safety analysis • RPr. S: Report Preliminary on Safety • RFS: Report Final on Safety 077 -05/rs
DCLL TBM Design Summary and Status US ITER TBM • • • DCLL blanket concept has been proposed and well-studied by US and EU blanket and reactor design experts US team selected the DCLL TBM after extensive reviews and assessments With RAFS, DCLL blanket concept has the potential to be a high performance blanket (gross ηth~40%) for DEMO with minimum critical issues Clear R&D paths have been identified for DEMO and DCLL TBM, many items are DCLL TBM common to the EU HCLL TBM Design section view DCLL TBM has strong common interest from China and Japan US has unique capabilities to address identified R&D items - RAFS fabrication is under development with inputs from US vendors - Si. C composite FCI is the key for high thermal and blanket performance Requirements on Si. C composite for FCI are much lower than for Si. C first wall and the development can piggyback on decades of study on Si. C as structural material and with input from US vendors - MHD analysis and experiments have made significant progress We are completing the second iteration of the DCLL TBM pre-conceptual design Primary and secondary helium loops and Pb. Li loop and corresponding components have been identified with well defined WBS structure DCLL TBM is ready to proceed to the preliminary design phase 077 -05/rs
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