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TITAN Task 1-2_ Tritium Behavior in Blanket Systems

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TITAN Task 1-2_ Tritium Behavior in Blanket Systems Powered By Docstoc
					                Tritium transport properties in
                lead-lithium eutectic

                 Pattrick Calderoni
www.inl.gov




              Fusion Nuclear Science and Technology Annual Meeting
              August 2-4, 2010
              UCLA
Fusion Safety Program



R&D program objective


 Experimental determination
    of hydrogen isotopes             Tritium transport modeling in
  solubility in lead lithium         liquid metal blanket systems
        eutectic (LLE)

                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts

    Critical evaluation of
  completed and operating              Pre-conceptual design of
 experiments with hydrogen             forced convection liquid
  isotopes and lead lithium                   metal loop
            alloys
                                                                     2
Fusion Safety Program



Collaborative task


                                            US / Japan TITAN
                                              collaboration




                          Gen IV /
                                                                          IEA
                            VHTR
                                                                     implementing
                        activities on
                                                                     agreement on
                         sodium and
                                                                        fusion
                         molten salt
                                                                      technology
                          coolants




                                        ITER / TBM safety analysis




                                                                                    3
Fusion Safety Program



Near-term activities focus re-alignment

                           Experiments           Analysis
                        reduced effort and    database re-
                         focus on tritium    assessment and
                          solubility test      loop design




                                                              4
Fusion Safety Program



R&D program objective


 Experimental determination
    of hydrogen isotopes             Tritium transport modeling in
  solubility in lead lithium         liquid metal blanket systems
        eutectic (LLE)

                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts

    Critical evaluation of
  completed and operating              Pre-conceptual design of
 experiments with hydrogen             forced convection liquid
  isotopes and lead lithium                   metal loop
            alloys
                                                                     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
                                    Katsuta 85
                                                                         investigation (including
                                                                         title)
                                                 Aiello 06
                                         Chan 84                       • Based on permeation
              Fukada 09
                                                       Schumacher 90     through sealed iron
                        Fauvet 88                                        capsules
                                          Reiter 91
                                                                       • Most representative for
                                                                         T / LLE / Fe alloy
                                                                         systems

                                                                       • Reiter results mostly at
                                                                         400C and with 90%
                                                                         background retention in 9
Fusion Safety Program



R&D program objective


 Experimental determination
    of hydrogen isotopes             Tritium transport modeling in
  solubility in lead lithium         liquid metal blanket systems
        eutectic (LLE)

                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts

    Critical evaluation of
  completed and operating              Pre-conceptual design of
 experiments with hydrogen             forced convection liquid
  isotopes and lead lithium                   metal loop
            alloys
                                                                     10
  Fusion Safety Program



  TITAN experiments at INL – FY08
                                                                                                  Test tube 1
Alumina crucible and vacuum boundary            Tube   Mass
                                                           g
                                                               Tube ID Liquid v
                                                                     cm          cc
                                                                                      Liquid h
                                                                                             cm   25 g LLE from batch
No metal in heated zone
                                  1
                                  2
                                                       26.26
                                                       40.55
                                                               1.4
                                                               1.4
                                                                          2.77
                                                                          4.27
                                                                                      1.80
                                                                                      2.78
                                                                                                  1
                                  5                    24.56   2.4        2.59        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
                                         Tube   Mass    Tube ID Liquid v       Liquid h
                                                    g         cm          cc          cm
                                         1      26.26   1.4        2.77        1.80
                                         2      40.55   1.4        4.27        2.78
From EU report ‘High Temperature         5      24.56   2.4        2.59        0.57
                                                                                           From B. Pint (ORNL)
Corrosion of Technical Ceramics’, by               Test tube 2                             presentation at ICFRM14,
Coen (JRC Ispra):                                  40 g LLE from batch 1                   Sept 7-11 2009

 ‘Al2O3 reacts intensively with the
formation of both LiAlO2 and LiAl5O8’,
at 800C for 1500h




                                                                                                                      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             Tritium transport modeling in
  solubility in lead lithium         liquid metal blanket systems
        eutectic (LLE)

                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts

    Critical evaluation of
  completed and operating              Pre-conceptual design of
 experiments with hydrogen             forced convection liquid
  isotopes and lead lithium                   metal loop
            alloys
                                                                     14
                         Fusion Safety Program



               Tritium transport modeling

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




                                          Time [s]   15
        Fusion Safety Program



  Tritium transport modeling
                                                                                   Permeator T2 transport model
Schematic of TMAP DCLL test blanket system                                               Pb-17Li
                                                                                                              Membrane diffusion

  model                                                                                mass transport


(B. Merrill)                                                                       T  K m CT,Bulk  CT,S1  CT,S2
                                                                                                                                   Molecular
                                                                   He/H2O HXs                                                      recombination
                                                                                                CT,Bulk

                                                                                                QPb-17Li                           T2  rC2T,S3
              DCLL TBM
                                                                                                               CT,S1
                   PbLi core        Permeator

                                                PbLi/He HX
                                                                                                   CT,S2 KS, Nb        K m D tube
First                                                                                                                               0.0096 Re 0.913Sc 0.346
wall                                                                                               CT,S1 KS,Pb-17Li    D T, Pb17Li




                      Rib He                                     Tritium cleanup   Uncertainties to be resolved by
  Rib walls
                               Concentric
                                  pipe
                                                                      system
                                                                                   experiments:
                  Back plate                                 He pipes              • 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      Considers non-condensible
                                         losses, and choking                           gas effects


                                                                                                     Air
                                                                                                     atmosphere                      Leak
                                                                           Fog/vapor
                   Heat transfer to structures                                                                                       Filtered
                   from both liquid and vapor                                                                                        Dryed
                   phases accounting for
                   single phase convection,
                   pool boiling, and vapor                                                                                          Considers
                   condensation                                                                                                     Leakage
                                                                                                                                    from
                                                                                                                                    Volumes

                   Conservation of mass
                   and energy of liquid and
                   vapor phases inside volumes                                         Aerosol models
                   including inter-phases heat                                         consider agglomeration,
                                                                                       steam condensation,
                   and mass transfer, and                                              pool scrubbing, gravity
                   hydrogen combustion                                                 settling and other
                                                                                       deposition mechanisms
                                                                      Liquid Pool


                                                                                                                   Models exist for
                                                                                                                   suppression pools,
                                                                                                                   heat exchangers,
                                                                                                                   valves, pumps, etc.

                                                                                                                                                17
Fusion Safety Program



R&D program objective


 Experimental determination
    of hydrogen isotopes             Tritium transport modeling in
  solubility in lead lithium         liquid metal blanket systems
        eutectic (LLE)

                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts

    Critical evaluation of
  completed and operating              Pre-conceptual design of
 experiments with hydrogen             forced convection liquid
  isotopes and lead lithium                   metal loop
            alloys
                                                                     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.
                                                                              20
  Left :LLE flow rate is 300cc/min / Right :LLE flow rate is 1000cc/min
Fusion Safety Program



Outlook of near term activities




                  Design and experimental validation of
                   tritium extraction systems for LLE
                            blanket concepts
                                                          21

				
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