Transcript Tritium & compatibility experiments with PbLi

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