Tritium transport properties in lead-lithium eutectic

1. Tritium transport properties in lead-lithium eutectic Pattrick Calderoni www.inl.gov Fusion Nuclear Science and Technology Annual Meeting August 2-4, 2010 UCLA

2. R&D program objective

3. Collaborative task

4. Near-term activities focus re-alignment

5. R&D program objective

6. 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

7. 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

8. Database evaluation – H solubility in LLE As presented by ItaloRicapito (F4E, then ENEA) in 2007 • Measurement technique • Equilibration time • Process interfaces • Passive interfaces • Velocity distribution • Temperature distribution Aiello

9. Database evaluation – H solubility in LLE • Chan and Veleckis work at ANL includes the widest parametric investigation (including title) • Based on permeation through sealed iron capsules • Most representative for T / LLE / Fe alloy systems • Reiter results mostly at 400C and with 90% background retention in Fe crucible Katsuta 85 Aiello 06 Chan 84 Fukada 09 Schumacher 90 Fauvet 88 Reiter 91

10. R&D program objective

11. TITAN experiments at INL – FY08 Test tube 1 25 g LLE from batch 1 Alumina crucible and vacuum boundary No metal in heated zone 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

12. TITAN experiments at INL – FY09 Fusion Safety Program 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 From B. Pint (ORNL) presentation at ICFRM14, Sept 7-11 2009 Test tube 2 40 g LLE from batch 1

13. 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

14. R&D program objective

15. Tritium transport modeling TMAP as tool for data analysis and experiments design (B. Merill) H2 release rate [Pa cc / s] Time [s]

16. Tritium transport modeling Permeator T2 transport model Schematic of TMAP DCLL test blanket system model (B. Merrill) Membrane diffusion Pb-17Li mass transport CT,S2 Molecular recombination CT,Bulk He/H2O HXs QPb-17Li CT,S1 DCLL TBM PbLi core Permeator PbLi/He HX First wall • Uncertainties to be resolved by experiments: • Tritium solubility and the mass transport correlation in flowing PbLi • Tritium behavior at PbLi/FS interface Rib He Tritium cleanup system Concentric pipe Rib walls Back plate He pipes

17. Conservation of momentum for 2 flow between volumes including friction, form losses, and choking Considers non-condensible gas effects Air atmosphere Leak Fog/vapor Filtered Heat transfer to structures from both liquid and vapor phases accounting for single phase convection, pool boiling, and vapor condensation 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 Aerosol models consider agglomeration, steam condensation, pool scrubbing, gravity settling and other deposition mechanisms Liquid Pool Models exist for suppression pools, heat exchangers, valves, pumps, etc. 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

18. R&D program objective

19. 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

20. 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

21. Outlook of near term activities Design and experimental validation of tritium extraction systems for LLE blanket concepts