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077-05/rs
DCLL Summary of TBM Design and AnalysisSignificant understanding on DCLL has been achieved by the team
General Atomics C. Wong, D. Carosella, M. P. LabarU. of California, Los Angeles M. Abdou, S. Smolentsev,
M. Dagher, S. Sharafat, N. Morley, P. Calderoni, A. Ying, M. Youssef
Consultant, Germany S. MalangU. of Wisconsin, Madison M. Sawan, G. SviatoslavskyORNL P. Fogarty, S. ZinkleIdaho National Laboratory B. MerrillUCSD D. K. SzePNL R. KurtzLLNL S. Reyes
ASSESSMENT AND DESIGN TEAM:
US ITER-TBM Meeting August 10-12, 2005 Idaho Falls, Idaho
0.0 100
5.0 10-9
1.0 10-8
1.5 10-8
2.0 10-8
2.5 10-8
3.0 10-8
0 5 10 15 20 25 30
Tri
tium
Pro
duct
ion
Rat
e (k
g/m
3 .s)
Radial Distance from FW (cm)
Radial Distribution of Tritium Production in LiPb Breeder
Neutron Wall Loading 0.78 MW/m2
DCLL TBM LiPb/He/FS90% Li-6
Front Channel Back Channel
077-05/rs
Outline
• Key features of DCLL concept• Reference reactor blanket• Conceptual TBM module design and analysis completed
Engineering designNeutronicsPbLi MHDHe thermal hydraulicsStructural analysis including disruption analysisTritium extraction and controlSafetyBypass loop
• PIE and hot-cell requirements• Key tasks for preliminary design• Summary
077-05/rs
Key Features of the DCLL Concept
• Cool FW and ferritic steel structure with 8 MPa He (also used for FW/blanket preheat and possible tritium control)
• Breeder is self-cooled Pb-17Li moving at a slow velocity, < 10 cm/s – allowing high Tout (700°C) leading to ηth> 40% (CCGT)
• Use flow channel inserts (FCIs), wherever possible to:– Provide electrical insulation to reduce
MHD pressure drop– Provide thermal insulation to decouple Pb-17Li
bulk flow temperature from wall/structure temperature
– Provide additional corrosion resistance since only nearly stagnant Pb-17Li is in contact with the ferritic steel structural walls
EU design, FED, 61-62, 2002
US ARIES ST
077-05/rs
DCLL Reference Tokamak FW/Blanket Module
First Wall
He Out
Pb-17Li Out
Reference Tokamak
Outboard max. NWL 3 MW/m2Maximum surface heat flux 0.5 MW/m2
Typical 2x3 m Blanket Module
Structural material: RAFS F82HFW/structure coolant: 8 MPa HeSelf-cooled breeder: Pb-17LiGross ηth > 40% with CCGT
Concentric pipes
077-05/rs
DCLL Reference Reactor Design Blanket Module
Cross-section
Exposed view, He flow
Exploded view
Pb-17Li flow
077-05/rs
US ITER-TBM DCLL Pb-17Li Module Initial Design Parameters with 10 cm frame thickness
• ITER scenario: Fusion power–500 MW, burn time-400 s
• Design heat flux: Average at 0.3 MW/m2, max at 0.5 MW/m2
• Design neutron wall loading: 0.78 MW/m2 (Under high poweroperation, the outboard mid-plane could see 1.09 MW/m2)
• Disruption load: 0.55 MJ/m2
• Disruption energy dump during current quench: 0.72 MJ/m2
• Pulse length: 400 s/
077-05/rs
Conceptual Engineering Design
077-05/rs
DCLL ITER-TBM Half Test Port Design
DCLL TBM
1 of 3 ITER Test Ports
077-05/rs
DCLL TBM Mechanical Design (Initial Design)
Test module
Cross-sectionPb-17Li flow
Sections, He flow
Radial thickness of PbLi zone islimited by the volume of PbLi to be
077-05/rs
4041424344454647484950
1 2 3 4 5 6 7 8
OUTLET CHANNELS
VELO
CIT
Y (m
/s)
Header Pass 1 to Pass 2
Header Pass 2 to Pass 3
MINUMUM TARGET VELOCITY42.6 m/s
Header Pass 3 to Pass 4
Header Pass 4 to Pass 5
Helium Gas Flow Distribution Analysis”For the determination of plenum size”
Header Flow Model Flow Distribution Results
Plenum size was increased by ~ 5 cm each until relatively uniform velocities are reached from different outlet channels
old new
077-05/rs
Top Plate Assembly
Bottom plate assembly
Outer back Plate Assembly
He Inlet/Outlet Pipe Assembly
Center divider late Assembly
Pb-Li Inlet/Outlet Pipe Assembly
1940
645
413
DCLL Module W/O FCIRadial depth increased from 308 mm to 413 mm
First wallpanel
077-05/rs
FW & Top/Bottom Plate He Flow Schematic
He Coolant Flow Circuit 1 Thru FW & Back Plate
He Coolant Flow Circuit 2 Thru FW & Back Plate
Top Plate He Flow
077-05/rs
Install the Cross Assembly of the PbLi Dividing Plates and weld to First Wall Module,then install the “Tee” Assembly and finish welding.
( welds are shown in yellow )
“Cross” Assembly“Tee” Assembly
077-05/rs
SiC InsertsFront Upper
Assembly
SiC InsertsRear Assembly
SiC InsertsBottom Assembly
SiC InsertsTop Assembly
Divider / Grid PlateAssembly
Install SiC/SiC Flow Channel Inserts
SiC InsertsFront Lower
Assembly
SiC inserts shownassembled with the
first wall module removed
SiC inserts shownfully assembled infirst wall module
Front View Back View
077-05/rs
First Wall
Grid Plates
Divider Plates
Inner Back PlateStructure
Inner Back PlateCover
Install Inner Back Plate AssemblyMake Internal
Seal Welds(shown in yellow)
077-05/rs
Install Top and Bottom Plate AssembliesTop Plate Cover
Top Plate Structure
BottomPlate Cover
BottomPlate
Structure
Make internal seal weldsto top and bottom of
the grid plates
Make external seal welds to
top and bottom plate assemblies
077-05/rs
Neutronics
077-05/rs
Neutronics for DCLL Reference Reactor Blanket
Blanket thicknessOB 75 cmIB 52.5 cm
Local TBR is 1.328OB contribution 0.995IB contribution 0.333
If neutron coverage for double null divertor is 12% overall TBR will be ~1.17 excluding breeding in divertorregion. Based on 3D results of DC molten salt design we expect blanket to have potential for achieving tritium self-sufficiency
Nuclear energy multiplication 1.136Peak nuclear heating in OB blanket
o FS 36 W/cm3o LL 33 W/cm3o SiC 29 W/cm3
077-05/rs
Tritium Production
Local TBR in the DCLL TBM is only 0.741 because of the small thickness (41.3 cm)
Tritium generation rate in the TBM is 3.2x1017 atom/s (1.59x10-6 g/s) during a D-T pulse with 500 MW fusion power
For a pulse with 400 s flat top preceded by 100 s linear ramp up to full power and followed by 100 s linear ramp down total tritium generation is 7.97x10-4 g/pulse
For the planned 3000 pulses per year the annual tritium production in the TBM is 2.4 g/year
Tritium production in the Be PFC is 2.2x10-9 g/s ⇒ 1.1x10-6 g/pulse ⇒3.3x10-3 g/year
Peak tritium production rate in LiPb is 2.94x10-8 kg/m3s during the D-T pulse
0.0 100
5.0 10-9
1.0 10-8
1.5 10-8
2.0 10-8
2.5 10-8
3.0 10-8
0 5 10 15 20 25 30
Tri
tium
Pro
duct
ion
Rat
e (k
g/m
3 .s)
Radial Distance from FW (cm)
Radial Distribution of Tritium Production in LiPb Breeder
Neutron Wall Loading 0.78 MW/m2
DCLL TBM LiPb/He/FS90% Li-6
Front Channel Back Channel
077-05/rs
Nuclear HeatingPower density in Be PFC 8.6 W/cm3
Peak power density in FS structure 8.2 W/cm3
Peak power density in Pb-17Li 18 W/cm3
Peak power density in SiC FCI 5.9 W/cm3
0
5
10
15
20
0 5 10 15 20 25 30 35 40
FSLiPbSiC
Pow
er D
ensit
y (W
/cm
3 )
Radial Distance from FW (cm)
Radial Distribution of Power Density in DCLL TBM Components
Neutron Wall Loading 0.78 MW/m2
LL
SiC FS
0.033Inlet/Outlet Pipes
0.103Back Wall
0.982Total
0.228Back Pb-17Li Channel
0.395Front Pb-17Li Channel
0.019Flow Channel Divider
0.020Radial Ribs
0.014Top/Bottom Walls
0.042Side Walls
0.128First Wall
Nuclear Heating (MW)
Component
0.014 MWTop/Bottom
0.033 MWPiping
0.020 MWRibs
0.019 MWDivider
0.042 MWSide Walls
0.128 MWFW
0.228 MWBack LL Channel
0.395 MWFront LLChannel
0.103 MWBack
077-05/rs
Radiation Damage in Steel Structure
For 0.57 MW/m2 average NWL and total fluence 0.3 MWa/m2 total irradiated time is 0.526 FPYPeak cumulative end-of-life dpa in FW is 5.7 dpa and He production is 64 He appmPeak cumulative end-of-life He production in inlet/outlet pipes is 0.34 He appm⇒ Pipe connections at back of TBM are reweldable
Determined radial variation of dpa, He and hydrogen production rates in structure
10-1
100
101
102
103
0 5 10 15 20 25 30 35 40
dpa/FPYHe appm/FPYH appm/FPY
Dam
age
Rat
e in
Ste
el S
truc
ture
per
FPY
Depth in Blanket (cm)
Radial Distribution of Damage Rate in Steel Structure
Neutron Wall Loading 0.78 MW/m 2
DCLL TBMLiPb/He/FS
90% Li-6
10-5
10-4
10-3
10-2
10-1
100
101
102
103
0 50 100 150
dpa/FPYHe appm/FPYH appm/FPY
Dam
age
Rat
e in
Ste
el S
truc
ture
per
FPY
Radial Distance from FW (cm)
Radial Distribution of Damage Rate in Steel Structure
Neutron Wall Loading 0.78 MW/m2
DCLL TBMLiPb/He/FS90% Li-6
Pipes Region Shield Plug
DCLLTBM
Dam
age
rate
40 cm 150 cm
077-05/rs
Total Radioactivity and After Heat and Contribution from Each Material
10-5
10-4
10-3
10-2
10-1
100
101
102
103
104
105
106
107
10-1 100 101 102 103 104 105 106 107 108 109 1010 1011
F82HLiPbSiCTotal
Time after Shutdown, Sec
1 s 1 m 1 h 1 d 1 mo 1 y 10 y 100 y 1000 y
TBM dimension64.5 cm Toroidal194 cm poloidal200.2 cm radial (including 128 cm shield)
10-14
10-13
10-12
10-11
10-10
10-9
10-8
10-7
10-6
10-5
10-4
10-3
10-2
10-1
10-1 100 101 102 103 104 105 106 107 108 109 1010 1011
F82HLiPbSiCtotal
Time after Shutdown, Sec
1 s 1 m 1 h 1 d 1 mo 1 y 10 y 100 y 1000 y
TBM dimension64.5 cm Toroidal194 cm poloidal200.2 cm radial (including 128 cm shield)
Radioactivity After Heat
At shutdown, the total radioactivity inventory and after heat are as low as ~2 MCi and 0.02 MW, respectively. They are mainly due to the F82H structure present in the shield, back plate, the back breeder channel, and the FW, in that order
077-05/rs
Waste Disposal Rating
Definition: The radwaste classification of TBM components was evaluated according to the Nuclear Regulatory Commission (NRC) 10FR61 and Fetter waste disposal concentration limits. The limits given are based on the assumption that all solid components are crushed before being disposed (no voids). Components having WDR >1, according to Class C limits, do not qualify for shallow land burial.
Results: The WDR values for F82H structure, the Pb-17Li breeder, and the SiC insert are 1.3x10-2, 8.7x10-3, and 2.1x10-4, respectively, based on Fetter’s limits. Although the Fetter limits are generally more conservative, still the values are much lower than unity and therefore these materials are qualified for shallow land burial according to the Class C limits and are within ITER guidelines.
077-05/rs
PbLi flow including MHD effects
077-05/rs
Effect of Electrical Conductivity and Pressure Equalization hole Geometry of FCI on Velocity Profile and Pressure Drop
With pressure equalization holes
With pressure equalization slot inverse flow generated
ΔP/dx at different σ for PEH and PES
B-fielddirection
077-05/rs
Pb-17Li MHD/Heat Transfer Issues(Pb-17Li, Tin=400° C, Tout=650° C)
1. Effectiveness of SiCf/SiC FCI as electrical/thermal insulator
2. 3-D MHD pressure drop
3. Flow/heat transfer in the concentric pipe
• In the poloidal channel flows, MHD pressure drop is caused by cross-sectional currents, which can be reduced with FCI. At the same time FCI serves as thermal insulator.
• 3-D MHD pressure drop is related to axial currents, which can’t be reduced by insulation. This type of pressure drop is associated with the flows in manifolds and a fringing magnetic field.
• “Hot” Pb-17Li flowing in the internal pipe will cause wall temperature rise. The goal of the analysis is to assess this rise in temperature.
077-05/rs
1. Poloidal Channel Flow with FCI
• 5 mm FCI with σ=20 S/m reduces the MHD pressure drop by a factor of 100
• Near-wall velocity jets exist, indicating that insulation is not perfect
• Velocity in the gap and jet flows will affect heat transfer
Velocity profile in the front poloidalchannel with FCI. Velocity is scaled with the mean velocity; z and y are scaled with b (half of the channel width).
B
077-05/rs
2. Pressure Drop in the Pb-17Li
• The MHD pressure drop is mostly contributed by 3-D flows in the manifolds, where axial currents appear due to changes in the flow direction
• High MHD pressure drop also appears in the flows in the concentric pipe, when the pipe goes through the fringing magnetic field
1000.302-0.442Total
23.30.07-0.14Outlet manifold
23.20.07-0.14Inlet manifold
19.30.0585Concentric pipe (annulus, fringing B-field)
19.30.0585Concentric pipe (internal, fringing B-field)
9.50.0286Concentric pipe (annulus, uniform B-field)
5.115.4×10-3Concentric pipe (internal, uniform B-field)
0.160.485×10-3Return channels
0.13 0.384×10-3Front channels
ΔPi/ΔP(%)
ΔPi(MPa)
Flow
077-05/rs
3. Heat Transfer in the Concentric Pipe
• Velocity profile was calculated first using 2-D MHD code
• Heat transfer calculations were then performed using FLUENT
• Maximum wall temperature does not exceed 475°C
Cross-sectional temperature distribution (°C) in the concentric pipe
B
077-05/rs
First wall design
077-05/rs
Results From First Wall Fluent Code Modeling
• With counter flow temperature variation between adjacent tubes is minimum
• One-sided wall roughening maintains FWtemperature well below 550oC
• One-sided wall roughening increases heat transfer coefficient but increases also pressure drop
C377Pa3716598w/m2-k3455smooth concurrent flow
C373Pa3716598w/m2-k3455smooth concurrent flow
C376Pa5000512w/m2-k9150Ks=0.000395 (one-side)
C376Pa3669589w/m2-k3404Smooth counter flow
He Outlet TP dropmax TFW(C)h
0.16 MW/m2 from PbLiSide
C373Pa5000514w/m2-k9150Ks=0.000395 (one-side)
C373Pa3669589w/m2-k3404Smooth counter flow
He Outlet TP dropmax TFW(C)hAdibataic PbLi Sice
077-05/rs
Counter Flow FW Thermal Performance
Plasma sideHeat Flux =0.3 MW/m2
Pb-17Li sideAdiabatic
First Wall Thermal Model
525 C
430 C First Wall Thermal Results
He Inlet Temp: 360° C
He Outlet Temp: 432° C
Max FW Temp: 523° C
077-05/rs
Power removed by Helium cooling:
0.375 MW
FW surface heat flux contributionAverage heat flux = 0.3 MW / m2FW area = 1.94 m X 0.64 m = 1.25 m2
+ 0.359 MW
Nuclear heating contributionTotal nuclear heat in TBM = 0.982 MWNuclear heat in Pb-17Li = 0.623 MWAssuming no leaks from Pb-17Li because of FCI
= 0.734 MW
Resulting total mass flow rate = 2.353 kg /sInlet T = 360 C / Outlet T = 420 CCp = 5200 J / kg K
(54% of total power)
10.1%0.811 MPaTotal
4.4%0.351 MPaFirst wall
1.7%0.14 MPaFlow distribution
4%0.32 MPaInlet / outlet pipes
Fraction of inlet pressurePressure drop Circuit component
Estimated Total Pressure Drop
DCLL He cooling Energy Balance and Pressure Drops
077-05/rs
Structural Analysis
077-05/rs
Temperature Distribution (Temperature can be controlled via mass flow rate)
(Kelvin)
FW-Helium out at 420oC
FW-Helium inat 360oC
Ribs and Back Plate Helium Tin = 438oC; Tout = 440oC
077-05/rs
FW Thermal Analysis Boundary Conditions
FW Helium makes 5 Passes: Pass 1: In 360oC Out 372oCPass 2: In 372oC Out 384oCPass 3: In 384oC Out 396oCPass 4: In 396oC Out 408oCPass 5: In 408oC Out 420oC
RIB Helium flow:Single PassTin = 420oCTout = 440oC
Heat Transfer Coefficient:FW (plasma side only)hcoef = 6979 W/m2-K*(1-side rib-roughened)All other surfaces
assume smooth wallshcoef = 3586 W/m2-K
5 Pass -Helium flowDirections
Tin = 360oC
Tout = 420oC Heat Loads:Plasma on FWq’’ = 0.5 MW/m2Volumetric Heatingq’’’=8.3 exp (-10 y) MW/m3Leakage from Pb-17Li into walls0.16 MW/m2, total ~ 40% of the heat
Simplified Geometry
077-05/rs
FW – One Sided Roughening Temperature Distribution
{
{Beryllium:Tmax = 580oCF82H:Tmax=559oC(top 1.2mm) h=6978 W/m2K
Higher h or mass flow rate can reduce the Tmax
077-05/rs
Primary + Secondary Stresses
Von Mises(MPa)
077-05/rs
DCLL 5-Channel Section Satisfies ISDC Design Rules
High Temperature Primary Membraneand Membrane Plus Bending Stress Limits are Satisfied
&
m
m
(T ) (T ,t )
mm
t l
SP
S⎧
≤ ⎨⎩
L b mP P K S+ ≤ bL tt
PP SK
+ ≤
Where:K is the bending shape factorand Kt = (K + 1)/2, K=1.15 [1].
1. Tavassoli et al Fusion Engineering and Design 61-62 (2002) 617-628.
077-05/rs
DCLL 5-Channel Section Satisfies ISDC Design Rules
High Temperature Ratcheting Primary Plus Secondary Stress Limits
1X Y+ ≤
Conservative Rule:Cyclic Ratcheting is Prevented
077-05/rs
LOCA Structural Analysis Results
Displacements
Maximum Displacements occur at the tips and is equal to
8.435 mm
Back View
IsoFront View
077-05/rs
Structural Analysis Results
Stress Distribution:
σ = 530 MPa
Critical section 1:(Rib-first wall)
View A
View A
077-05/rs
Reinforced Rib-FW Structure
Adding material (4x4 mm) Reduced Maximum Stresses
from 530 MPa to 450 MPa
σ = 450 MPa
First Wall
Added material
077-05/rs
Reinforced Top Plate Side Wall Contact
Adding material (4x15 mm) Reduced Maximum
Stresses from 616 MPa to 415 MPa
σmax = 415 MPa
077-05/rs
Disruption Eddy Current Analysis
ITER specified cases:18 ms exponential plasma current decay40 ms linear decay… max. JxB force generated
Method of analysis:OPERA E&M analysis code by VectorFieldsA 20° segment of ITERPoloidal field coils, vacuum vessel, TBM are includedPlasma represented by time varying filaments
Simulationvolume
Vacuumvessel
TBM
077-05/rs
• There is a considerable torque about the center of the TBM• There is a large twisting force trying to make a propeller
blade• The grids and the back plate have the greatest forces• Static finite element analysis to determine the stresses is
on-going, and is being iterated with the structural analysis group.
Disruption Analysis Summary“40 ms linear case”
N/mm3-Toroidal Direction (backplate)N/mm3-Radial Direction (backplate)
077-05/rs
Tritium Extraction and High Temperature Loop
077-05/rs
Primary Side Pb-17 Li Vacuum Permeator Scheme
He inletHe outlet
Vacuum pumpVacuum permeator
Blanket Concentric pipes
HeatExchanger
T2 outlet
Pressure boundary (90 °C)
Closed Brayton CyclePb-17Li (460 °C)
Pb-17Li (700 °C)
Pb-17Li pump
Inter-cooler Pre-cooler Recuperator
Turbo-compressor
Power turbine
For Power Reactor
077-05/rs
Single Tube Tritium Permeator Assessment
Mathematical model• A component balance describes the tritium
mass fraction along the membrane length
• Tritium transport to the membrane surface is described by a mass transfer coefficient
• The effective tritium partial pressure at the membrane surface is given by the solubility
• Permeation depends on the permeability
FGMWD
dzdx
ρπ−=
( )0xxMWk
N im
i −=ρ
2
0⎟⎟⎠
⎞⎜⎜⎝
⎛=
sr k
xp
( )prp pplk
AGG −==
_
Vacuum
Pb-17Li in Pb-17Li out
077-05/rs
Tritium Pressure for Various ParametersReasonable Outlet Partial Pressure can be Achieved
Mass transport coefficient 5.6 m/sDifferent tritium breeding ratio (TBR)
Mass transport coefficient 5.6 m/sNiobium permeation reduction factor (PRF)
.025
0 1 2 3 40.0
0.1
0.2
0.3
0.4
0.5
Tube length (m)
Pre
ssur
e (P
a)
.02
.06.025.02
0 1 2 3 4Tube length (m)
0.0
0.1
0.2
0.3
0.4
0.5
Pre
ssur
e (P
a)PRF = 10
PRF = 1
PRF = 50
TBR = 1
TBR = 1.17
Equilibrium results determined by the TMAP code
077-05/rs
While this initial analysis indicates that tritium permeatormay be feasible, many issues remain
• Measured mass transfer coefficients for the Pb-17Li T system• Compatibility of Pb-17Li with Nb at 700°C for power reactor• Additional resistances to tritium permeation such as surface
resistance?• At the Pb-17Li membrane interface, is the effective partial pressure
exerted by tritium indeed given by the solubility equation? (this may be a very different mechanism with a very different rate)
• What pressure can be practically maintained on the permeate side of the membranes?
• Will Nb tubes degrade due to reactions such as oxidation? • Will a surface treatment be needed?• For the DCLL TBM and for demonstration we will replace Nb with
a FS-permeator, and operate external Pb-17Li at T
077-05/rs
TBM Safety
077-05/rs
Reference Accidents Analyzed
1. Ex-vessel LOCA analysis to determine:– Pressurization of TBM vault– Behavior of TBM without active plasma shutdown
2. Coolant leak into TBM breeder zone analysis to assess:– Module and ancillary system pressurization– Chemical reactions and hydrogen formation– Subsequent in-vessel leakage
3. In-vessel TBM coolant leak analysis to demonstrate:– Minimal Pressurization of ITER first confinement barrier (i.e., VV)– Passive removal of TBM decay heat– Limited chemical reactions and hydrogen formation
077-05/rs
MELCOR Developed for TBM Accident Analyses
First wall
Concentric pipe
Permeator
PbLi/He HX
Back plate
He pipes
He/H2O HXs
Vacuum vessel
Be/FS/HE/FS/SiC
Drain tank
Port cell
TWCS vault
• 30 control volumes• 37 flow paths• 72 heat structures
(psuedo 3D TBM conduction)
• 6 valves• 1 rupture disk• 1 pump and 2
circulators
077-05/rs
Ex-vessel LOCA Pressure Results
• LOCA assumed to start at the end of a reactor pulse flat top (300 s)• Pressure port cell relief valve (set to open at 0.2 atm pressure differential with
TWCS vault and to re-seat at 0.01 atm pressure differential) limits test cell pressure to 1.5 atm, not exceeding confinement barrier design limits of 2 atm
0.0 0.5 1.0 1.5 2.0Time (hr)
0.0
0.5
1.0
1.5
2.0
Pre
ssur
e (a
tm)
Test cellTWCS vaultVacuum vessel
2600 2800 3000 3200Time (s)
0.8
1.0
1.2
1.4
1.6
Pre
ssur
e (a
tm)
2.0Test cellTWCS vaultVacuum vessel
Expanded view
077-05/rs
Ex-vessel LOCA Temperature and Oxidation Results
• FW beryllium evaporates and disrupts plasma ~ 90 seconds after LOCA starts• Beryllium on “hot strip” (region with a surface heating of 0.5 MW/m2) reacts with
steam from ITER FW cooling system producing 9 g of H2• In ~100 s, 0.32 m3, of Pb-17Li leaks into the VV; the quantity of hydrogen generated
from PbLi-water reactions based on data* is ~ 2.5x10-4 mol-H2/g LM × 2.92x106 g LM × 2 g-H2/ mol-H2 = 1460 g.
0
250
500
750
1000
1250
1500
Tem
pera
ture
(C)
0.0 0.5 1.0 1.5 2.0Time (hr)
FW
SW 2800 3000 3200Time (s)
0
500
1000
1500
FW
SW
0
1
2
3
FW b
eryl
lium
thic
knes
s (m
m)
FW
.00
.05
.10
.15
FW H
ydro
gen
Prod
uctio
n (k
g)
0.0 0.5 1.0 1.5 2.0Time (hr)
FW hot strip
*D. W. Jeppson, Nuclear Technology/Fusion, 4 (1983), p. 277-287
077-05/rs
TBM Tritium Permeation Analysis
• ITER allowed in-building permeation limit should be ~100 mg-T/a, based on an allowed TBM release to the environment of 1 mg-T as HTO/a and an assumed building detritiation system efficiency of 99%
• A TMAP model was developed to examine permeation from the TBM and ancillary system into the ITER test cell and vault, this model includes:
– All TBM structures (500°C), and the TBM Pb-17Li (temperature change of 400°C to 650°C)
– All piping (400°C or 450°C) and heat exchangers– T2 pulsed production with a flat top value of 1.59e-6 g/s during a 600 s pulse,
which translates into 2.33 g-T/a for 3000 pulses/a– A vacuum permeator, composed of 20, 5 m long, 0.01 m diameter tubes
077-05/rs
TBM Tritium Permeation Results
• TBM tritium concentration reaches an oscillatory equilibrium after ~50 pulses, with Pb-17Li tritium pressure decreasing from 1.5 Pa to 1.1 Pa through permeator by the end of a pulse
• Annual release based on 3000 consecutive pulses will be above ITER in-building annual limit of 100 mg-T/a, so alumina coatings (50 μm) will be applied to all piping to reduce permeation by a factor of 10 to 10,000
0 10 20 30 40 500.0
0.5
1.0
1.5
2.0
Triti
um p
ress
ure
abov
e Pb
-17L
i (Pa
)
Number of pulses
Expanded view
0 1000 2000 30000.0
0.5
1.0
1.5
2.0
Number of pulses
0 1000 2000 3000Number of pulses
0
100
200
300
400
Triti
um re
leas
e (m
g-T/
a) Helium piping
Pb-17Li piping
ITER limit
077-05/rs
TBM Meets ITER General Safety Requirements
• TBM pressurization of the VV, vaults and test cell is within ITER acceptance criterion
• TBM FW beryllium oxidation does not result in hydrogen generation that exceeds ITER limit of 2.5 kg
• TBM system Pb-17Li inventory is over 0.28 m3, but based on Pb-17Li/water reaction data the quantity of hydrogen generated will be less than the ITER limit of 2.5 kg
• Helium inventory is ~17 kg for the FW helium cooling loop, which is less than ITER limit of 40 kg
• TBM decay heat removal demonstrated
• Tritium permeation may require a permeation barrier for piping to meet ITER limits
077-05/rs
LM Pump
HX
TBM
TBM BypassValve
Cold PbLi Supply Pipe
Hot PbLi Return Pipe
Storage Tank
Tritium Extraction
Bypass control can be used to shunt coldPbLi away from the TBM operation, while themajority of the loop is operated at lowtemperature
Hot He to TCWS Building
Cold He from TCWS Building
Hot He to TCWS Building
Cold He from TCWS Building
Primary He flowrate andtemperature is alsocontrollable so thattemperature differencescan be studied even in lowtemperature experiments
PbLi flowrate and coldleg temperature arecontrollable so thatlow temperatureoperation is achievedand a wide variety ofexperimentalconditions can beproduced
Pb-17Li by-pass Loop Schematic(To control Pb-17Li temperature
077-05/rs
We followed closely the U.S. Position on PIE Needs
• A dedicated TBM hot cell is needed, but only limited PIE should be performed at the ITER site since each party already has extensive facilities for conducting specimen preparation, examination and testing.
• To avoid the expense and difficulty of shipping a large radioactive component the main function of the TBM hot cell should be to perform optical and other nondestructive examination, sectioningand preparation of samples for shipment, diagnostics instrument replacement, replacement of specimen cassettes, limited repairs,and perhaps simple mechanical testing of pre-machined specimens.
077-05/rs
TBM Hot Cell Requirements
• Capability to handle large activated metal/ceramic components, gas, and liquids.
• Remote visual equipment and camera to inspect and record condition of the TBM and guide cutting operations.
• Equipment for measuring dimensional changes.• Nondestructive inspection equipment such as UT and ET for
examination of large components.• Equipment for removal and temporary storage of specimen cassettes.• Cutting tools for sectioning the TBM into smaller, more manageable
pieces.• Capability to prepare pieces (and specimens) removed from a TBM
for shipment to the country of origin.• Simple (tensile, Charpy impact) mechanical testing of pre-machined
specimens over a range of temperatures and possibly under vacuum/inert gas.
077-05/rs
Key Tasks to be Addressed During Preliminary Design
Mechanical: More details including flow distributions and fabrication procedureNeutronics: More details and 3-D analysis including TBM shield designPbLi MHD design: More details for the whole loop, including natural convection
effectsHe thermal hydraulics: More details on first wall one-side roughening,
performance and flow distributions for the helium loopsTritium: Design of the tritium extraction and impurity control systemsDisruption analysis: Complete conceptual assessment and work on detailsSafety: Review and work with ITER, continue to provide safety design guidanceBy-pass loop: More design detailsPIE and hot-cell: Module needs to be designed to be compatible with PIE and hot-
cell requirementsAncillary equipment: More design and system detailsDiagnostics: Module design to accommodate measurement needs
077-05/rs
SUMMARY
• Using RAFS as the structural material, He as the FW/structure coolant, Pb-17Li as the self-cooled tritium breeding material and SiCf/SiC as FCI conceptual DCLL-TBM assessment was completed.
• For the reference reactor design, ηth>40% can be projected.• For the ITER TBM design, the following areas were evaluated: mechanical design, neutronics and
activation, He cooled FW/structure thermal hydraulic, Pb-17Li circuit including MHD effects with FCI, structural analysis, LOCA, disruption impact, by-pass Pb-17Li loop, tritium extraction, safety, and ancillary equipment on the transporter and TCWS, including PIE and hot cell requirements.
• SiCf /SiC composite FCI is the key for high thermal performance.• The DCLL TBM design can satisfy all ITER TBWG requirements.• A test plan for DCLL has been drafted.• The DCLL DDD report is to be finalized in August.• The DCLL TBM concept is ready to proceed to the preliminary design phase.
Summary