Fusion Safety Program
TITAN Task 1-2 Tritium Behavior in Blanket Systems INL Fusion Safety Program activities
www.inl.gov
Pattrick Calderoni
TITAN Joint Workshop Nov 1-2, 2010 Toyama University, Japan
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
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 4
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 5
Fusion Safety Program
Database evaluation What is the lithium lead eutectic?
15.7 at %, 235 C mp
Title, homogeneity and impurity content affect 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 6
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 Italo Ricapito (F4E, then ENEA) in 2007 7
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 8
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
9
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 10
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
11
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
12
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
13
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]
14
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 - DT
T Km CT, Bulk CT,S1 CT,S2
He/H2O HXs
CT x Molecular recombination
CT,Bulk
T2 rC2T,S3
QPb-17Li DCLL TBM PbLi core
CT,S1 Permeator PbLi/He HX
CT,S2 KS,Nb CT,S1 KS,Pb-17Li
First wall
Rib He Concentric pipe
Rib walls Back plate
Tritium cleanup system He pipes
K m D tube 0.0096 Re0.913Sc0.346 D T,Pb 17Li
Uncertainties to be resolved by experiments: • Tritium solubility and the mass transport correlation in flowing PbLi • Tritium behavior at PbLi/FS interface
15
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.
16
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
17
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
18
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. 19
Left :LLE flow rate is 300cc/min / Right :LLE flow rate is 1000cc/min
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 20
www.inl.gov
Pattrick Calderoni
TITAN Joint Workshop Nov 1-2, 2010 Toyama University, Japan
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
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 4
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 5
Fusion Safety Program
Database evaluation What is the lithium lead eutectic?
15.7 at %, 235 C mp
Title, homogeneity and impurity content affect 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 6
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 Italo Ricapito (F4E, then ENEA) in 2007 7
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 8
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
9
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 10
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
11
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
12
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
13
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]
14
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 - DT
T Km CT, Bulk CT,S1 CT,S2
He/H2O HXs
CT x Molecular recombination
CT,Bulk
T2 rC2T,S3
QPb-17Li DCLL TBM PbLi core
CT,S1 Permeator PbLi/He HX
CT,S2 KS,Nb CT,S1 KS,Pb-17Li
First wall
Rib He Concentric pipe
Rib walls Back plate
Tritium cleanup system He pipes
K m D tube 0.0096 Re0.913Sc0.346 D T,Pb 17Li
Uncertainties to be resolved by experiments: • Tritium solubility and the mass transport correlation in flowing PbLi • Tritium behavior at PbLi/FS interface
15
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.
16
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
17
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
18
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. 19
Left :LLE flow rate is 300cc/min / Right :LLE flow rate is 1000cc/min
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 20
