Tritium transport properties in lead-lithium eutectic www.inl.gov Pattrick Calderoni Fusion Nuclear Science and Technology Annual Meeting August 2-4, 2010 UCLA Fusion Safety Program R&D program objective Experimental determination of hydrogen isotopes solubility in lead lithium eutectic (LLE) Tritium transport modeling in liquid metal blanket systems Design and experimental validation of tritium extraction systems for LLE blanket concepts Critical evaluation of completed and operating experiments with hydrogen isotopes and lead lithium alloys Pre-conceptual design of forced convection liquid metal loop 2 Fusion Safety Program Collaborative task US / Japan TITAN collaboration Gen IV / VHTR activities on sodium and molten salt coolants IEA implementing agreement on fusion technology ITER / TBM safety analysis 3 Fusion Safety Program Near-term activities focus re-alignment Experiments reduced effort and focus on tritium solubility test Analysis database reassessment and loop design 4 Fusion Safety Program R&D program objective Experimental determination of hydrogen isotopes solubility in lead lithium eutectic (LLE) Tritium transport modeling in liquid metal blanket systems Design and experimental validation of tritium extraction systems for LLE blanket concepts Critical evaluation of completed and operating experiments with hydrogen isotopes and lead lithium alloys Pre-conceptual design of forced convection liquid metal loop 5 Fusion Safety Program Database evaluation • Reports on hydrogen solubility and transport properties prepared in 2000 by A. Pisarev (Moscow Technical Un.) on ENEA contract • Provided by F4E through IEA Implementing Agreement on Nuclear Technology for Fusion Reactors • Contain critical evaluation of experimental facilities, procedures and data analysis • Summarized by I. Ricapito at Int. Workshop on Liquid Metal Breeder Blankets at INL in 2007 • FZK TRITEX experiment report 6 Fusion Safety Program Database evaluation What is the lithium lead eutectic? 15.7 at %, 235 C mp Title, homogeneity and impurity contentaffect Li activity and therefore hydrogen isotopes solubility – up to 5 orders of magnitude difference between pure elements TRITEX op experience: PbLi at phase boundaries and 20-60 at% Li in condensate composition 7 Fusion Safety Program Database evaluation – H solubility in LLE • Measurement technique • Equilibration time Aiello • Process interfaces • Passive interfaces • Velocity distribution • Temperature distribution As presented by ItaloRicapito (F4E, then ENEA) in 2007 8 Fusion Safety Program Database evaluation – H solubility in LLE • Chan and Veleckis work at ANL includes the widest parametric investigation (including title) Katsuta 85 Aiello 06 Fukada 09 Chan 84 Schumacher 90 • Based on permeation through sealed iron capsules Fauvet 88 Reiter 91 • Most representative for T / LLE / Fe alloy systems • Reiter results mostly at 400C and with 90% background retention in Fe crucible 9 Fusion Safety Program R&D program objective Experimental determination of hydrogen isotopes solubility in lead lithium eutectic (LLE) Tritium transport modeling in liquid metal blanket systems Design and experimental validation of tritium extraction systems for LLE blanket concepts Critical evaluation of completed and operating experiments with hydrogen isotopes and lead lithium alloys Pre-conceptual design of forced convection liquid metal loop 10 Fusion Safety Program TITAN experiments at INL – FY08 Alumina crucible and vacuum boundary 1 No metal in heated zone 2 Tube Mass Tube ID 5 g 26.26 40.55 24.56 1.4 1.4 2.4 Liquid v cm Liquid h cc 2.77 4.27 2.59 cm Test tube 1 25 g LLE from batch 1 1.80 2.78 0.57 Desorption test rely on the assumption of complete equilibration during charge phase. Initial evaluation of procedure parameters was not validated by TMAP modeling results. PVT technique require assumptions for gas temperature continuous desorption measurement not feasible, rate-step introduces further parameters complicating analysis 11 Fusion Safety Program TITAN experiments at INL – FY09 From EU report ‘High Temperature Corrosion of Technical Ceramics’, by Coen (JRC Ispra): ‘Al2O3 reacts intensively with the formation of both LiAlO2 and LiAl5O8’, at 800C for 1500h Tube Mass 1 2 5 g 26.26 40.55 24.56 Tube ID Liquid v cm 1.4 1.4 2.4 Liquid h cc 2.77 4.27 2.59 cm 1.80 2.78 0.57 Test tube 2 40 g LLE from batch 1 From B. Pint (ORNL) presentation at ICFRM14, Sept 7-11 2009 12 Fusion Safety Program TITAN experiments at INL – ongoing LLE in quartz crucibles showed evidence of strong interaction both in resistive and induction heating tests Tritium test configuration: W crucibles (99.97%, smooth forged) induction heating Ameritherm Ekoheat 10kW 13 Fusion Safety Program R&D program objective Experimental determination of hydrogen isotopes solubility in lead lithium eutectic (LLE) Tritium transport modeling in liquid metal blanket systems Design and experimental validation of tritium extraction systems for LLE blanket concepts Critical evaluation of completed and operating experiments with hydrogen isotopes and lead lithium alloys Pre-conceptual design of forced convection liquid metal loop 14 Fusion Safety Program Tritium transport modeling H2 release rate [Pa cc / s] TMAP as tool for data analysis and experiments design (B. Merill) Time [s] 15 Fusion Safety Program Tritium transport modeling Permeator T2 transport model Schematic of TMAP DCLL test blanket system model (B. Merrill) Membrane diffusion Pb-17Li mass transport T K m C T, Bulk C T, S1 CT,S2 Molecular recombination He/H2O HXs CT,Bulk T2 r C T, S3 2 QPb-17Li DCLL TBM PbLi core CT,S1 Permeator PbLi/He HX C T, S2 First wall C T, S1 Rib He Concentric pipe Rib walls Back plate Tritium cleanup system He pipes K S, Nb K m D tube K S, Pb -17Li D T, Pb 17Li 0.0096 Re 0.913 Sc 0.346 Uncertainties to be resolved by experiments: • Tritium solubility and the mass transport correlation in flowing PbLi • Tritium behavior at PbLi/FS interface 16 Fusion Safety Program Tritium transport modeling • MELCOR can be used to give a more detailed engineering thermal-hydraulic experimental design analysis if needed • MELCOR is a engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants, including reactor cooling system and containment fluid flow, heat transfer, and aerosol transport. (Developed by Sandia National Laboratory) • Modification have been made to MELCOR at the INL for fusion applications, including the addition of PbLi as a working fluid Conservation of momentum for 2 flow between volumes including friction, form losses, and choking Fog/vapor Considers non-condensible gas effects Air atmosphere Heat transfer to structures from both liquid and vapor phases accounting for single phase convection, pool boiling, and vapor condensation Leak Filtered Dryed Considers Leakage from Volumes Conservation of mass and energy of liquid and vapor phases inside volumes including inter-phases heat and mass transfer, and hydrogen combustion Liquid Pool Aerosol models consider agglomeration, steam condensation, pool scrubbing, gravity settling and other deposition mechanisms Models exist for suppression pools, heat exchangers, valves, pumps, etc. 17 Fusion Safety Program R&D program objective Experimental determination of hydrogen isotopes solubility in lead lithium eutectic (LLE) Tritium transport modeling in liquid metal blanket systems Design and experimental validation of tritium extraction systems for LLE blanket concepts Critical evaluation of completed and operating experiments with hydrogen isotopes and lead lithium alloys Pre-conceptual design of forced convection liquid metal loop 18 Fusion Safety Program Forced convection liquid metal loop design Current effort is mainly at program level and leveraged with activities related to advanced power plant concepts within DoE NE • Conceptual design of an engineering scaled facility to investigate heat transfer properties of molten salt coolants • Conceptual design of a sodium components Test Complex • Planning for nuclear technology development facilities at INL, in particular related to the decommissioning of secondary loops of EBR-II 19 Fusion Safety Program Forced convection liquid metal loop design Preliminary parametric investigation of main loop parameters (K. Katayama) Hydrogen concentration in flowing LLE at the test section. Averaged leak rate from F82H main pipes to atmosphere. H2 partial pressure in outer gas phase of the test section. Left :LLE flow rate is 300cc/min / Right :LLE flow rate is 1000cc/min 20 Fusion Safety Program Outlook of near term activities Experiments Database evaluation Modeling Loop design T solubility Ongoing H solubility activities Database data analysis LLE loop parameters Single effect H transport properties H transport properties T extraction concepts Advanced coolants test facility Design and experimental validation of tritium extraction systems for LLE blanket concepts 21