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Modeling and Measurement ofModeling and Measurement ofModeling and Measurement of Interfacial Area Concentration in
T h Fl
Modeling and Measurement of Interfacial Area Concentration in
T h FlTwo-phase FlowTwo-phase Flow
Mamoru Ishii and Takashi Hibiki
Thermal-Hydraulics and Reactor Safety LaboratorySchool of Nuclear Engineering
P d U i itPurdue UniversityWest Lafayette, IN 47907
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
IntroductionIntroduction
The modeling philosophy of thermal-hydraulic system analysis codes treatsThe modeling philosophy of thermal-hydraulic system analysis codes treats interface structure using flow regimes and transition criteria that cannot dynamically represent the changes in interfacial structure (no time or length scale is incorporated into the transition criteria).
This leads to instantaneous changes in flow regime, which can not only induce non physical oscillations in system behavior but can also hamperinduce non-physical oscillations in system behavior but can also hamper code accuracy and robustness.
To better represent the effects of interfacial structure and regime transition, the use of a first order equation to characterize interfacial area transporthas been recommended (Ishii, 1975).has been recommended (Ishii, 1975).
ai : interfacial area concentrationvi : interfacial velocityΦ
( )ii i
aa
tΦ
∂+ ∇ ⋅ =
∂v
Φ : sink and source termst∂
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Formulation of Interfacial Area Transport Equation (K t f ll i d I hii 1995)(Kocamustafaogullari and Ishii, 1995)
Boltzmann Transport Equation of Particles
( ) j ph
f dVf f S S
t V dt⎛ ⎞∂ ∂ ⎟⎜+∇ ⋅ + = +⎟⎜ ⎟⎜⎝ ⎠∂ ∂ ∑v
jt V dt⎝ ⎠∂ ∂
( ), , : distribution function, : volume, and : particle source and sink rates per unit
mixture volume due to particle interaction and phase changej phf V t V S Sx
mixture volume due to particle interaction and phase change
Interfacial Area Transport Equation
( ) ( ){ }23
i ii i g ph j ph
j
a aa
t tα
α η Φ Φα
∂ ∂⎛ ⎞⎟⎜+∇ ⋅ − +∇⋅ − = +⎟⎜ ⎟⎝ ⎠∂ ∂ ∑v vj
: rate of volume generated by nucleation source per unit mixture volume, and :
interfacial area source and sink rates per unit mixture volume due to particle interactionph j phη Φ Φ
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
p p
and phase change
Development of Two-group Interfacial AreaT t E ti (I hii d Hibiki 2005)Transport Equation (Ishii and Hibiki, 2005)
Fluid particle number density transport equation analogous to Interfacial Area Transport Eqtransport equation analogous to Boltzmann’s transport equation
Interfacial Area Transport Eq.
Two-group approachTwo group approach
Group 1Spherical/distorted bubble
Group 2Cap/slug/churn-turbulent
b bblgroup bubble group
4d,maxDgσΔρ
=Maximum Distorted Bubble Size Limit
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
, gΔρ
Two-group Interfacial Area Transport Equation(I hii d Ki 2004 I hii d Hibiki 2005)(Ishii and Kim, 2004; Ishii and Hibiki, 2005)
Two group Void Fraction Transport Equation
( ) ( ) ( ) 121gk g kgk g gk gk m
t
α ρα ρ Γ Δ
∂+∇⋅ = + −
∂v
Two-group Void Fraction Transport Equation
t∂
( ) ( )3 112 2 1 1 1 1 1
1
* *, , g crit
g j c g g ph cj Sm
Dm D D
t D
αΔ ρ η χ α η
⎡ ⎤⎧ ⎫∂⎪ ⎪⎪ ⎪⎢ ⎥= + +∇⋅ − ≡⎨ ⎬⎢ ⎥⎪ ⎪∂⎪ ⎪⎩ ⎭⎣ ⎦∑ vj⎣ ⎦
( ) ( ){ } ( )2 11 12 *a a α⎡ ⎤∂∂ ⎢ ⎥ ∑
Two-group Interfacial Area Transport Equation
( ) ( ){ } ( )2 11 11 1 1 1 1 1 1 1
1
23
*,
gi ii i c g g ph j ph
jg
a aa D
t t
αχ α η φ φ
α∂∂ ⎢ ⎥+∇ ⋅ = − +∇⋅ − + +⎢ ⎥∂ ∂⎣ ⎦
∑v v
( ) ( )22 22 gi ia aa
αα η
⎡ ⎤∂∂ ⎢ ⎥+ ∇ +∇v v( ) ( )
( ) ( )
2 22 2 2 2 2
2
2 111 1 1 1 2 2
3
*
gi ii i g g ph
g
gic g g ph j ph
at t
aD
t
α ηα
αχ α η φ φ
⎢ ⎥+ ∇ ⋅ = +∇⋅ −⎢ ⎥∂ ∂⎣ ⎦
⎡ ⎤∂⎢ ⎥+ +∇⋅ − + +⎢ ⎥∂ ∑
v v
v( ) ( )1 1 1 1 2 21
,c g g ph j phjg t
χ η φ φα ⎢ ⎥∂⎣ ⎦
∑
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Two-group Momentum Equation(S t l 2003)(Sun et al., 2003)
Two group Momentum Equation
( ) ( ) ( )1 11 1 1 1 1 1 1 1 1
g g g Tg g g g g g g g g gp
tμ
α ρα ρ α α α ρ
∂ ⎡ ⎤+∇⋅ = − ∇ +∇⋅ + +⎢ ⎥⎣ ⎦∂
vv v gT T
Two-group Momentum Equation
( ) ( )
( )1 12 1 1 1 1g gi gi ig
tmΓ α
⎣ ⎦∂+ −Δ −∇ ⋅ +v MT
( )( ) ( ) ( )
( )
2 22 2 2 2 2 2 2 2 2
2 12 2 2 2 2
g g g Tg g g g g g g g g
i i i
pt
m
μα ρ
α ρ α α α ρ
Γ α
∂ ⎡ ⎤+∇⋅ = − ∇ +∇⋅ + +⎢ ⎥⎣ ⎦∂+ +Δ −∇ ⋅ +
vv v g
v M
T T
T( )2 12 2 2 2 2g gi gi igmΓ α+ +Δ ∇ +v MT
( )( ) ( ) ( )( )
11 1 1
g f f Tg f f f g f g f fp μ
α ρα ρ α α
⎡ ⎤∂ −⎢ ⎥⎣ ⎦ ⎡ ⎤ ⎡ ⎤+ ∇ ⋅ − = − − ∇ +∇⋅ − +⎢ ⎥ ⎢ ⎥⎣ ⎦ ⎣ ⎦v
v v T T( ) ( ) ( )( )
( ) ( )
1 1 1
1 1
g f f f g f g f f
g f f fi if g fi
pt
α ρ α α
α ρ Γ α
+∇ ∇ +∇ +⎢ ⎥ ⎢ ⎥⎣ ⎦ ⎣ ⎦∂+ − + + −∇ − ⋅
v v
g v M
T T
T
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
One-dimensional One-group Interfacial AreaT t E tiTransport Equation
Bubbly Flow Regime2-G Interfacial Area Transport Eq. → 1-G Interfacial Area Transport Eq.
One-dimensional Interfacial Area Transport Equation
( )a∂ ∂ ( )ii iz HE WE BB VT BC CDa
aa v
t zΦ Φ Φ Φ Φ Φ
∂ ∂+ = + + + − −
∂ ∂
HE: bulk liquid boiling, WE: bubble nucleation from active cavities,C C
( )dnWEWE DfN ,,ΦΦ =
BB: bubble breakup, VT: void transport, BC: bubble coalescence, CD: condensation
N : active nucleation site density f : bubble generation frequencyNn: active nucleation site density, f : bubble generation frequency,Dd: bubble departure size
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Classification of Possible Interactions ofT B bbl (Hibiki d I hii 2000)Two-group Bubbles (Hibiki and Ishii, 2000)
Coalescence Breakup
Group 1(1+1 & 1)
Inter-group(1+2 & 2)
Inter-group(1+1 & 2)
Group 2(2+2 & 2)
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Bubble Coalescence & Breakup Mechanism(K t f ll i d I hii 1995)(Kocamustafaogullari and Ishii, 1995)
Coalescence Mechanisms
Random Collision Wake Entrainment
Breakup MechanismsBreakup Mechanisms
Turbulent impact Shearing-off Surface instability
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Modeling of Bubble Coalescence(Hibiki d I hii 2000)(Hibiki and Ishii, 2000)
Probable Coalescence Bubble random collision induced byProbable CoalescenceMechanism
Bubble random collision induced by turbulence in a liquid phase
Bubble Coalescence Rate
Bubble Collision Frequency
Coalescence Efficiency= x
Bubble Collision Frequency
(1) Turbulence is isotropic,(2) Bubble size lies in the inertial subrange.
Coalescence
Coalescence efficiency is an exponential function of time required for bubble
l i b li id fil thi iCoalescence Efficiency coalescence given by liquid-film-thinning
model and a contact time for two bubbles given by dimensional consideration.
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Sink Term of Interfacial Area Concentration Due to B bbl C l (Hibiki d I hii 2000)Bubble Coalescence (Hibiki and Ishii, 2000)
Bubble Collision Coalescence
1 3 1 2 5 6 1 3K D⎛ ⎞⎛ ⎞
Bubble Collision Frequency
Coalescence Efficiency
( )
1 3
2 3,max
CC
b C
fD
γ αεα α
=− 1 2exp exp C f bC
CC
K Dt ρ ελ
τ σ
⎛ ⎞⎛ ⎞ ⎟⎜⎟⎜ ⎟⎜⎟= − = −⎜ ⎟⎜⎟ ⎟⎜ ⎟⎜ ⎜ ⎟⎜⎝ ⎠ ⎟⎝ ⎠
2 2⎛ ⎞ ⎛ ⎞
Rate of IAC Change
1 2 5 6 1 3
1 13 3C C C b C
i i
f na aα α
Φ φ λψ ψ
⎛ ⎞ ⎛ ⎞⎟ ⎟⎜ ⎜⎟ ⎟= =⎜ ⎜⎟ ⎟⎜ ⎜⎟ ⎟⎜ ⎜⎝ ⎠ ⎝ ⎠⎛ ⎞K =1 29
( )
1 2 5 6 1 31 32
5 3 1 2,max
exp C f bC
b C
K D
D
ρ εΓ α εα α σ
⎛ ⎞⎟⎜ ⎟⎜= − ⎟⎜ ⎟⎜ ⎟⎜ ⎟− ⎝ ⎠
KC=1.29
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Modeling of Bubble Breakup(Hibiki d I hii 2000)(Hibiki and Ishii, 2000)
Probable Breakup Bubble eddy random collision inducedProbable BreakupMechanism
Bubble-eddy random collision induced by turbulence in a liquid phase
Bubble Breakup Rate
Bubble-Eddy Collision Frequency
Breakup Efficiency= x
Bubble-Eddy Collision Frequency
(1) Turbulence is isotropic,(2) Eddy size lies in the inertial subrange.(3) Eddy with size from cD to D can break
Breakup
(3) Eddy with size from cDb to Db can break up bubble with size of Db.
Breakup efficiency is an exponential function Breakup Efficiency
p y pof average energy of a single eddy and average energy required for bubble breakup
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Source Term of Interfacial Area Concentration Due to B bbl B k (Hibiki d I hii 2000)Bubble Breakup (Hibiki and Ishii, 2000)
Bubble Eddy BreakupBubble-Eddy Collision Frequency
Breakup Efficiency
1 3E K⎛ ⎞⎛ ⎞
( )
1 3
2 3,max
BB
b B
fD
γ αεα α
=− 5 3 2 3exp expB B
B
f b
E Ke D
σλ
ρ ε
⎛ ⎞⎛ ⎞ ⎟⎜ ⎟⎟⎜ ⎜= − = − ⎟⎟⎜ ⎜⎟⎜ ⎟⎝ ⎠ ⎜ ⎟⎜ ⎟⎝ ⎠
Rate of IAC Change
2 21 1⎛ ⎞ ⎛ ⎞
K =1 59 1 3
1 13 3B B B e B
i i
f na aα α
Φ φ λψ ψ
⎛ ⎞ ⎛ ⎞⎟ ⎟⎜ ⎜⎟ ⎟= =⎜ ⎜⎟ ⎟⎜ ⎜⎟ ⎟⎜ ⎜⎝ ⎠ ⎝ ⎠⎛ ⎞KB=1.59 ( )
( )
1 3
5 3 5 3 2 31
,max
expB B
b B f b
K
D D
Γ α α ε σα α ρ ε
⎛ ⎞− ⎟⎜ ⎟⎜= − ⎟⎜ ⎟⎜ ⎟⎜ ⎟− ⎝ ⎠
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Sink Term of Interfacial Area Concentration Due to B bbl C d ti (P k t l 2007)Bubble Condensation (Park et al., 2007)
Heat Transfer Control Region
Inertial-Control Region
( )2
4 1 BCD HC IC b c t Nuc Ja
c
Dn P N N ,
tΦ Φ Φ π α
⎧ ⎫⎪ ⎪⎪ ⎪= + = − +⎨ ⎬⎪ ⎪⎪ ⎪⎩ ⎭
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
c⎪ ⎪⎩ ⎭
Modeling of Wall Nucleation TermModeling of Wall Nucleation Term
W ll N l ti S T
2 : heated perimeter : cross sectional arean HN fD A
ξφ π ξ=
Wall Nucleation Source Term
, : heated perimeter, : cross-sectional areaWE d H cc
D AA
φ π ξ=
Key Models to Estimate Wall Nucleation Source Term
Active Nucleation Site Density, Nn
Key Models to Estimate Wall Nucleation Source Term
y, n
Bubble Departure Diameter, Dd
Bubble Departure Frequency, f
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Active Nucleation Site DensityActive Nucleation Site Density
rr*
r
r1
r
Active nucleation site density images by Basu et al. (2002).
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Active Nucleation Site Density (Hibiki and Ishii, 2003)Active Nucleation Site Density (Hibiki and Ishii, 2003)
Knowledge of Size and Cone Angle Distributions of Cavities Modelg g
( )2
21 exp exp 18n nN N f ,
Rθ λ
ρμ
+⎡ ⎤⎧ ⎫ ⎧ ⎫⎛ ⎞⎪ ⎪ ⎪ ⎪⎪ ⎪ ⎪ ⎪⎟⎜ ⎢ ⎥= − − ⎟ −⎨ ⎬ ⎨ ⎬⎜ ⎟ ⎢ ⎥⎜ ⎟⎜⎪ ⎪ ⎪ ⎪⎝ ⎠ ⎪ ⎪⎪ ⎪ ⎩ ⎭⎣ ⎦8 cRμ⎝ ⎠ ⎪ ⎪⎪ ⎪ ⎩ ⎭⎩ ⎭ ⎣ ⎦
( ) ( )2 30 01064 0 48246 0 22712 0 05468 log *f ρ ρ ρ ρ ρ ρ+ + + + += + + ≡
5 2 -6=4.72 10 sites/m , =0.722 radian, =2.50 10 m, : contact angle,nN μ λ θ× ×
( ) ( ) -0.01064 0.48246 0 22712 0 05468 logf . . , ,ρ ρ ρ ρ ρ ρ= + − + ≡
*gρ Δρ ρ≡
( ){ }( ) ( ){ }
2 1
exp 1
g f f
c
fg g sat g sat
pR ,
i T T RTT
σ ρ ρ+=
− −{ } : gas temperature, : saturation temperature, : latent heat,
: gas constant based on a molecular weight. For example, the valueg sat fgT T i
R
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
of for water vapor is 462 J/(kg K)(=8.31 J/(mol K)/(18.0 10-3 kg/mol))
Active Nucleation Site Density (Hibiki and Ishii, 2003)Active Nucleation Site Density (Hibiki and Ishii, 2003)
107Data: Basu, Warrier & Dhir (2002)
θ=30o, Water-Copper θ=57o, Water-Copper
o otes/m
2 ]
10111012
Basu et al., θ=30o
Basu et al., θ=57o
Basu et al θ=80otes/m
2 ]
106
θ=75o-85o, Water-Zr-4 θ=90o, Water-Copper Hibiki & Ishii Kocamustafaogullari & Ishii (1983) Basu, Warrier & Dhir (2002)
en.,
Nn
[sit
107108109
1010 Basu et al., θ=80 Basu et al., θ=90o
Borishanskii et al., P=0.101 MPa Borishanskii et al., P=0.450 MPa Borishanskii et al., P=0.980 MPa Borishanskii et al., P=1.48 MPa Borishanskii et al., P=1.94 MPa
n.,
Nn,
cal
c. [s
it
105
θ=90o
θ=90o
θ=75o-80o
θ=57oNuc
l. Si
te D
e
104105106107
+50 %
Borishanskii et al., P=2.95 MPa Borishanskii et al., P=4.90 MPa Borishanskii et al., P=7.30 MPa Borishanskii et al., P=9.80 MPaBorishanskii et al P=14 7 MPaN
ucl.
Site
Den
100 101 102104θ=30o
Act
ive
N
Wall Superheat ΔT [K]
10210310
102 103 104 105 106 107 108 109101010111012
-50 %Borishanskii et al., P=14.7 MPa Borishanskii et al., P=19.8 MPa Zeng & Klausner, θ=5o
Benjamin & Balakrishnan, θ=38o
2
Act
ive
N
Wall Superheat, ΔTw [K] Active Nucl. Site Den., Nn, meas.
[sites/m2]
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Bubble Departure DiameterBubble Departure Diameter
Time (ms): 0.0 0.2 0.4 0.6 1.0 1.2 Lift-offLift-off
Diameter (mm): 0.000 0.397 0.440 0.460 0.507 0.544 SlidingSliding
Time (ms): 1.4 1.6 1.8 2.0 2.2 2.4
CondensationVaporizationHeat Flux
CondensationVaporizationHeat Flux
i ( ) llDiameter (mm): 0.574 0.588 0.609 0.617 0.615 0.605 DepartureFlow DepartureFlow
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Bubble Departure Diameter (Situ et al., 2008)Bubble Departure Diameter (Situ et al., 2008)
Balance of Forces on Bubble at Nucleation Site Model
FpFqs : surface tension force at -directionsxF x
y
: unsteady drag force (growth force)
at -directionduxF
x
FduxFslθi
x F
: shear lift force
: surface tension force at -direction
: unsteady drag force at
sl
sy
F
F y
F directiony
Fduy Fg
Fsx
Fsy
: unsteady drag force atduyF -direction
: pressure force
: gravity forcep
y
F
F : gravity force
: quasi-steady force g
qs
F
F
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Bubble Departure Diameter (Situ et al., 2008)Bubble Departure Diameter (Situ et al., 2008)
∑ y sy duy p g qsF F F F F F .= + + + +∑
0F 0syF =Surface Tension Force
4 2444 fb α
U d D F 4443
fduy Jae i
bF N sin
αθ
π= −Unsteady Drag Force
4 ( ) 343p b f g bF F grπ ρ ρ+ = − −Pressure and Gravity Forces
1/12 12 0 796
6 3
/ nn
qs n
f f r b Reb
F.
v r Nπρ ν
−⎡ ⎤⎛ ⎞⎟⎜⎢ ⎥⎟= + +⎜ ⎟⎢ ⎥⎜ ⎟⎜⎝ ⎠⎢ ⎥⎣ ⎦Quasi-Steady Force
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Bubble Departure Frequency (Situ et al., 2008)Bubble Departure Frequency (Situ et al., 2008)
Non-Dimensional Analysis Correlationy
0 8034 06 .fd qNBN . N .=
2dfDN ≡
Non-Dimensional Bubble Departure Frequency
fdf
N ,α
≡
Non-Dimensional Heat Flux Representing Nucleate Boiling Heat Transfer
qNB dqNB
f g fg
q DN ,
iα ρ
′′≡
f g fg
: bubble departure diameter,
: nucleate boiling heat flux calculated by using Chen's correlation (1966)b
qNB
D
q ′′
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
g y g ( )qNBq
Bubble Departure Frequency (Situ et al., 2008)Bubble Departure Frequency (Situ et al., 2008)
106
Basu's data Situ et al.'s data Thorncroft et al.'s upflow dataThorncroft et al 's downflow datae f [-
]
106
Basu's data Situ et al.'s data Thorncroft et al.'s upflow data Thorncroft et al.'s downflow dataN =10 7N0.634
d=fdD
2 d /αf [-
]
105
Thorncroft et al. s downflow data
lds N
umbe
r, R
103
Nfd=10.7NqNB
Prediction error = ±113%
Freq
uenc
y N
fd
104
Liqu
id R
enol
100
-D. D
epar
ture
F
+100%
-100%
0 10 20 30 40 50 60103
Jacob Number, Jaw [-]10-3 100 103 10610-3N
on-
Non-D. Nucl. Boil. Heat Flux, NqNB=q"NBDd/(αfρgifg) [-]
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Local Interfacial Structure CharacterizationLocal Interfacial Structure Characterization
• Instrument: Multi-sensor Conductivity Probe• Measured Variables (Local)
1 1– Void fraction– Interfacial area concentration
1 1=
Δ ∑ti
j nij
aT v
– Interfacial velocity– Bubble number frequency and bubble chord length
T Di t ib ti f Th V i bl• Transverse Distribution of These Variables• Measurements for Two Bubble Groups Separately
– Group 1: Spherical and distorted small bubbles– Group 2: Taylor and churn-turbulent large bubbles
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Multi-Sensor Conductivity ProbeMulti Sensor Conductivity Probe
~250 mm
Epoxy and conductive ink Stainless-steel tubing,OD: ~3 mm
Thermocouplewires
~50-70 mm
Epoxy
OD: 3 mmStainless-steel tubing
wires
Coated sensor body
< 50-micron
01Sensor tip (uncoated)~2.5 mm
2 3
~ 0.5-0.8 mm
Not scaled
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Interfacial Area Measurement (Cont’d)Interfacial Area Measurement (Cont d)
impedanceΔ /Δt
t4
downstream sensor(sensor 2) sensor 1
vi=Δs/Δtt3
t2upstream sensor
(sensor 1)Δs
sensor 2
t2t1
timeΔt
Δs vb t1 t2 t3 t4
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Conductivity Probe PortConductivity Probe Port
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Benchmark with Image Analysis(50.8 mm ID Pipe Upward Flow: <jf > = 0.321 and <jg> = 0.179 m/s)
40 80
[%
]
imageexperiment
Frac
tion
AC 40
a i
[1/m]imageexperiment
α [
L sl = 10.5 cmL s = 49.8 cmj g =0.179m/sj =0 321m/s
Void
F IA
0
0.0 0.5 1.0r/R
j f =0.321m/s
Radial Position, r/R
0
0.0 0.5 1.0r/R
Radial Position, r/R
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Database to Evaluate Interfacial Area Transport E ti Adi b ti T h FlEquation -Adiabatic Two-phase Flow-
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Database to Evaluate Interfacial Area Transport E ti B ili T h FlEquation -Boiling Two-phase Flow-
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Summary of Interfacial Area DatabaseSummary of Interfacial Area Database
• Test Section Geometry: Round pipe, Confined channel, Annulus and Rod BundleT t S ti Si 1 t 102• Test Section Size: 1 mm to 102 mm
• Flow Regime: Bubbly, Cap-bubbly, Slug and Churn-turbulent Flowsturbulent Flows,
• Flow Condition: <jg> up to 10 m/s, <jf> from -3.1 m/s to 5.0 m/s
• Thermal Condition: Adiabatic and Diabatic Flows• Gravity Condition: Normal and Micro Gravity Conditions
P C diti At h i P• Pressure Condition: Atmospheric Pressure
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Database for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
dW
d
S
x
Subchannel CenterRod Gap CenterRod Gap Center
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Database for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Database for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
s]
101 Mishima and Ishii (1984) Bubbly to Cap Bubbly Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-TurbulentFlow Conditions for Local Data<j
f> [
m/s
B S
100
Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19loci
ty,
< B-S
S-A
10-1
14
20
17 19
318215871
Liqu
id V
e
1211
perf
icia
l L
S-C
C-A
10-2 10-1 100 101 10210-2
Sup
Superficial Gas Velocity, <jg> [m/s]
S C
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
g
Local Data for 8 X 8 Rod Bundle Geometry at z/D=200( j 0 2 / d j 0 02 / )(<jf > = 0.2 m/s and <jg> = 0.02 m/s)
0.20G 1 C
300Group 1 at Centerm
-1]101
Mishima and Ishii (1984) Bubbly to Cap Bubbly [m
/s]
0.10
0.15
Group 1 at Center Group 1 at Gap
on,
<α>
[-
]
150
200
250Group 1 at Center Group 1 at Gap
Con
c.,
<a i>
[m
100
y p y Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19
Vel
ocity
, <
j f>
B-S
S-A
0 00
0.05V
oid
Frac
tio
0
50
100
Inte
rfac
ial A
rea
C
10-2
10-120
1211
318215871
uper
ficia
l Liq
uid
S-C
C-A
0.0 0.2 0.4 0.6 0.8 1.00.00
Location from Center, x/W [-]0.0 0.2 0.4 0.6 0.8 1.0
0I
Location from Center, x/W [-]2.0
Group 1 at CenterG 1 t G/s
]
8
10Group 1 at CenterGroup 1 at Gapm
> [
mm
]
10-2 10-1 100 101 10210 2
Su
Superficial Gas Velocity, <jg> [m/s]
Wd
S
1.0
1.5 Group 1 at Gap
ocity
, <
v g> [
m/
4
6
8 Group 1 at Gap
Dia
met
er,
<D
Sm
x
S
0.0 0.2 0.4 0.6 0.8 1.00.0
0.5
Gas
Vel
o
0.0 0.2 0.4 0.6 0.8 1.00
2
Saut
er M
ean
D
Subchannel Center
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Location from Center, x/W [-] Location from Center, x/W [-]Subchannel CenterRod Gap Center
1-D Data for 8 X 8 Rod Bundle Geometry( j 0 2 / d j 0 02 / )(<jf > = 0.2 m/s and <jg> = 0.02 m/s)
0.20 Group 1Spacer Grid Location 250
300 Group 1Spacer Grid Location[m
-1]
0.10
0.15
Sp ce G d oc o
actio
n,
<α>
[-]
150
200
250 p
ea C
onc.
, <
a i>
101
m/s
]
0 50 100 150 2000.00
0.05V
oid
Fra
0 50 100 150 2000
50
100
Inte
rfac
ial A
re
100
10 Mishima and Ishii (1984) Bubbly to Cap Bubbly Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19loci
ty,
<j f>
[m
B-S
S-A
0 50 100 150 200Axial Position, z/DH [-]
0 50 100 150 200Axial Position, z/DH [-]
0 8
1.0 Group 1 Spacer Grid Location
m/s
]
4
5 Group 1 Spacer Grid Location
m>
[m
m]10-1
14 16
20
1211
17 19
318215871
rfic
ial L
iqui
d V
el
C-A
0.4
0.6
0.8
city
, <
<vg>>
[m
2
3
4
Dia
met
er,
<D
Sm
10-2 10-1 100 101 10210-2
Supe
r
Superficial Gas Velocity, <jg> [m/s]
S-C
0 50 100 150 2000.0
0.2
Gas
Vel
o
0 50 100 150 2000
1
Saut
er M
ean
D
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Axial Position, z/DH [-] Axial Position, z/DH [-]
Local Data for 8 X 8 Rod Bundle Geometry at z/D=200( j 1 0 / d j 0 02 / )(<jf > = 1.0 m/s and <jg> = 0.02 m/s)
0.20G 1 t C t
300Group 1 at Centerm
-1]
101 Mishima and Ishii (1984) Bubbly to Cap BubblyCap-Bubbly to Cap-Turbulent>
[m
/s]
0.10
0.15
Group 1 at Center Group 1 at Gap
on,
<α>
[-
]
150
200
250Group 1 at Center Group 1 at Gap
Con
c.,
<a i>
[m
100
Cap Bubbly to Cap Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19
d V
eloc
ity,
<j f>
B-S
S-A
0 00
0.05V
oid
Frac
tio
0
50
100
Inte
rfac
ial A
rea
C
2 1 0 1 210-2
10-120
1211
318215871
Supe
rfic
ial L
iqui
d
S-C
C-A
0.0 0.2 0.4 0.6 0.8 1.00.00
Location from Center, x/W [-]2.0
Group 1 at CenterGroup 1 at Gap/s
]
0.0 0.2 0.4 0.6 0.8 1.00I
Location from Center, x/W [-]
8
10Group 1 at CenterGroup 1 at Gap
m>
[m
m]
10-2 10-1 100 101 10210S
Superficial Gas Velocity, <jg> [m/s]
Wd
S
1.0
1.5p p
ocity
, <
v g> [
m/
4
6
8 p p
Dia
met
er,
<D
Sm
x
S
0.0 0.2 0.4 0.6 0.8 1.00.0
0.5
Gas
Vel
o
0.0 0.2 0.4 0.6 0.8 1.00
2
Saut
er M
ean
D
Subchannel Center
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Location from Center, x/W [-] Location from Center, x/W [-]Subchannel CenterRod Gap Center
1-D Data for 8 X 8 Rod Bundle Geometry( j 1 0 / d j 0 02 / )(<jf > = 1.0 m/s and <jg> = 0.02 m/s)
0.05Group 1
200Group 1m
-1]
0.03
0.04
Group 1 Spacer Grid Location
on,
<α>
[-
]
100
150
Group 1 Spacer Grid Location
Con
c.,
<a i>
[m
101
m/s
]
0 00
0.01
0.02
Voi
d Fr
actio
0
50
Inte
rfac
ial A
rea
C
100
10 Mishima and Ishii (1984) Bubbly to Cap Bubbly Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19loci
ty,
<j f>
[m
B-S
S-A
0 50 100 150 2000.00
Axial Position, z/DH [-]0 50 100 150 200
0I
Axial Position, z/DH [-]2.0
Group 1 Spacer Grid Location
m/s
]
4
5 Group 1 Spacer Grid Location
m>
[m
m]10-1
14 16
20
1211
17 19
318215871
rfic
ial L
iqui
d V
el
C-A
1.0
1.5
city
, <
<vg>>
[m
2
3
4
Dia
met
er,
<D
Sm
10-2 10-1 100 101 10210-2
Supe
r
Superficial Gas Velocity, <jg> [m/s]
S-C
0 50 100 150 2000.0
0.5
Gas
Vel
o
0 50 100 150 2000
1
Saut
er M
ean
D
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Axial Position, z/DH [-] Axial Position, z/DH [-]
Local Data for 8 X 8 Rod Bundle Geometry at z/D=200( j 0 2 / d j 0 5 / )(<jf > = 0.2 m/s and <jg> = 0.5 m/s)
1.0G 1 t C t
600G 1 t C tm
-1]101
Mishima and Ishii (1984) Bubbly to Cap Bubbly [m
/s]
0.6
0.8Group 1 at Center Group 1 at Gap Group 2 at Center Group2 at Gap
on,
<α>
[-
]
300
400
500Group 1 at Center Group 2 at Gap Group 2 at Center Group2 at Gap
Con
c.,
<a i>
[m
100
y p y Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19
Vel
ocity
, <
j f>
B-S
S-A
0 0
0.2
0.4
Voi
d Fr
actio
0
100
200
Inte
rfac
ial A
rea
C
10-2
10-120
1211
318215871
uper
ficia
l Liq
uid
S-C
C-A
0.0 0.2 0.4 0.6 0.8 1.00.0
Location from Center, x/W [-]0.0 0.2 0.4 0.6 0.8 1.0
0I
Location from Center, x/W [-]
2.5
3.0Group 1 at CenterGroup 1 at Gap
m/s
]
80
100 Group 1 at CenterGroup 1 at Gap
m>
[mm
]
10-2 10-1 100 101 10210 2
Su
Superficial Gas Velocity, <jg> [m/s]
Wd
S
1 0
1.5
2.0
p pGroup 2 at CenterGroup 2 at Gap
ocity
, <
v g> [
m
40
60
80 p p Group 2 at Center Group 2 at Gap
Dia
met
er,
<DSm
x
S
0.0 0.2 0.4 0.6 0.8 1.00.0
0.5
1.0
Gas
Vel
o
0.0 0.2 0.4 0.6 0.8 1.00
20Sa
uter
Mea
n
Subchannel Center
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Location from Center, x/W [-] Location from Center, x/W [-]Subchannel CenterRod Gap Center
Local Data for 8 X 8 Rod Bundle Geometry at z/D=200( j 0 2 / d j 0 5 / )(<jf > = 0.2 m/s and <jg> = 0.5 m/s)
1.0Group 1
600
m-1
]101 Mishima and Ishii (1984) Bubbly to Cap Bubbly [m
/s]
0.6
0.8
Group 1 Group 2
on,
<α>
[-
]
300
400
500 Group 1 Group 2
Con
c.,
<a i>
[m
100
y p y Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19
Vel
ocity
, <
j f>
B-S
S-A
0 0
0.2
0.4
Voi
d Fr
actio
0
100
200
Inte
rfac
ial A
rea
C
10-2
10-120
1211
318215871
uper
ficia
l Liq
uid
S-C
C-A
0.0 0.2 0.4 0.6 0.8 1.00.0
Location from Center, d/S [-]0.0 0.2 0.4 0.6 0.8 1.0
0I
Location from Center, d/S [-]
2.5
3.0 Group 1Group 2/s
]
80
100Group 1Group 2>
[m
m]
10-2 10-1 100 101 10210 2
Su
Superficial Gas Velocity, <jg> [m/s]
Wd
S
1 0
1.5
2.0
Group 2
ocity
, <
v g> [
m/
40
60
80 p
Dia
met
er,
<D
Sm
x
S
0.0 0.2 0.4 0.6 0.8 1.00.0
0.5
1.0
Gas
Vel
o
0.0 0.2 0.4 0.6 0.8 1.00
20Sa
uter
Mea
n D
Subchannel Center
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Location from Center, d/S [-] Location from Center, d/S [-]Subchannel CenterRod Gap Center
1-D Data for 8 X 8 Rod Bundle Geometry( j 0 2 / d j 0 5 / )(<jf > = 0.2 m/s and <jg> = 0.5 m/s)
0.6Group 1
500Group 1-1
]
101
m/s
]
0.3
0.4
0.5G oup Group 2 Spacer Grid Location
on,
<α>
[-]
300
400
Group 1 Group 2 Spacer Grid Location
Conc
., <
a i> [
m-
100
10 Mishima and Ishii (1984) Bubbly to Cap Bubbly Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19loci
ty,
<j f>
[m
B-S
S-A
0 0
0.1
0.2V
oid
Frac
tio
0
100
200
Inte
rfac
ial A
rea
C
10-1
14 16
20
1211
17 19
318215871
rfic
ial L
iqui
d V
el
C-A
0 50 100 150 2000.0
Axial Position, z/DH [-]
0 50 100 150 2000I
Axial Position, z/DH [-]
2 0
2.5 Group 1 Group 2Spacer Grid Locationm
/s]
40
50 Group 1 Group 2Spacer Grid Locationm
> [
mm
]
10-2 10-1 100 101 10210-2
Supe
r
Superficial Gas Velocity, <jg> [m/s]
S-C
1.0
1.5
2.0 p
city
, <
<vg>>
[m
20
30
40 Spacer Grid Location
Dia
met
er,
<D
Sm
0 50 100 150 2000.0
0.5
Gas
Vel
o
0 50 100 150 2000
10
Saut
er M
ean
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Axial Position, z/DH
[-] Axial Position, z/DH
[-]
1-D Data for 8 X 8 Rod Bundle Geometry1 D Data for 8 X 8 Rod Bundle Geometry
5G 1m
m]
101
m/s
] 3
4
Group 1 Spacer Grid Location
met
er,
<D
Sm>
[m
100
10 Mishima and Ishii (1984) Bubbly to Cap Bubbly Cap-Bubbly to Cap-Turbulent Cap-Turbulent to Churn-Turbulent Flow Conditions for Local Data
9 14 16
4 10 13 5 6
17 19loci
ty,
<j f>
[m
B-S
S-A1
2
aute
r Mea
n D
iam
10-1
14 16
20
1211
17 19
318215871
rfic
ial L
iqui
d V
el
C-A4
5 Group 1 Spacer Grid Location
m>
[m
m]
0 50 100 150 2000Sa
Axial Position, z/DH [-]
40
50 Group 1 Group 2Spacer Grid Locationm
> [
mm
]
10-2 10-1 100 101 10210-2
Supe
r
Superficial Gas Velocity, <jg> [m/s]
S-C
2
3
4
Dia
met
er,
<D
Sm
20
30
40 Spacer Grid Location
Dia
met
er,
<D
Sm
0 50 100 150 2000
1
Saut
er M
ean
D
0 50 100 150 2000
10
Saut
er M
ean
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Axial Position, z/DH [-] Axial Position, z/DH
[-]
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Benchmarking One-DimensionalI t f i l A T t E tiInterfacial Area Transport Equation
2 iaΦ Φ∂ ∂
+
One-Dimensional One-Group IATE under Steady Bubbly Flow Conditions
3i
i i TI RC gaa v v
z zΦ Φ α
α= − +
∂ ∂
Sink and Source Terms
1 2 5 6 1 32 1 3
exp RC f bRCK Dρ εΓ α ε
Φ⎛ ⎞⎟⎜ ⎟⎜= − ⎟⎜
Sink and Source Terms
( )5 3 1 2expRC
b RC ,maxDΦ
α α σ= ⎟⎜ ⎟⎜ ⎟⎜− ⎝ ⎠
( ) 1 31 KΓ α α ε ⎛ ⎞⎟⎜( )
( )1 3
5 3 5 3 2 3
1 expTI TITI
b TI ,max f b
K
D D
Γ α α ε σΦ
α α ρ ε
⎛ ⎞− ⎟⎜ ⎟⎜= − ⎟⎜ ⎟⎜ ⎟⎜− ⎝ ⎠
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Benchmarking One-Dimensional Interfacial Area T t E ti (Hibiki d I hii 2002)Transport Equation (Hibiki and Ishii, 2002)
30
40 Experimental Data (Hibiki & Ishii, 1999) Total Void TransportB bbl R d C lli i
[m-1] 600
800 Experimental Data (Hibiki & Ishii, 1999) Total Void Transport[m
-1]
10
20 Bubble Random Collision Turbulent Impact
> - <
a i> z/D
=12
200
400Bubble Random Collision Turbulent Impact
> - <
a i> z/D
=12
-10
0
<j >=0 872 m/s <j >=0 0813 m/sCha
nge,
<
a i>
400
-200
0
<j >=3 49 m/s <j >=0 702 m/sCha
nge,
<
a i>0 20 40 60 80 100 120 140
-20
<jf> 0.872 m/s, <jg,0> 0.0813 m/sD=25.4 mmIA
C C
Axial Position, z/D [-]0 20 40 60 80 100 120 140
-600
-400 <jf> 3.49 m/s, <jg,0> 0.702 m/sD=25.4 mmIA
C C
Axial Position, z/D [-]
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Benchmarking Interfacial Area Transport EquationS iti it A l i (Hibiki d I hii 2002)-Sensitivity Analysis- (Hibiki and Ishii, 2002)
1200
1400 Experimental Data
(Hibiki et al., 2001a)a i> [
m-1]
1200
1400 ΛD=0.50 ΛD=0.80Λ =1 00
<jf>=0.491 m/s<j >=0 190 m/s<a
i> [
m-1
]
800
1000
ΛD=0.50
entra
tion,
<
600
800
1000ΛD=1.00 ΛD=1.25 ΛD=1.50
<jg,0> 0.190 m/sD=50.8 mm
entra
tion,
<
200
400
600 ΛD=0.80
ΛD=1.00
ΛD=1.25<jf>=5.00 m/s<jg 0>=3.90 m/sA
rea
Con
ce
200
400
600
Are
a C
once
0 10 20 30 40 50 600
200 ΛD=1.50
g,0
D=50.8 mmIn
terf
acia
l
Axial Position, z/D [-]0 10 20 30 40 50 60
0
00 Experimental Data (Hibiki et al., 2001a)
Inte
rfac
ial
Axial Position, z/D [-]
0
0
b, ,setD
b,
D
DΛ ≡
0
0
: initial bubble diameter utilized for computation,
: initial bubble diameter observed in the experimentb, ,set
b,
D
D
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
0b, 0b,
Benchmarking One-DimensionalI t f i l A T t E tiInterfacial Area Transport Equation
One-Dimensional One-Group IATE under Steady Bubbly Flow Conditions
2 ii i TI RC CD
aa v vΦ Φ Φ α
∂ ∂= − − +
with Condensation
3i i TI RC CD gaa v v
z zΦ Φ Φ α
α+
∂ ∂
Sink and Source Terms
( )
1 2 5 6 1 32 1 3
5 3 1 2exp RC f bRCRC
b RC ,max
K D
D
ρ εΓ α εΦ
α α σ
⎛ ⎞⎟⎜ ⎟⎜= − ⎟⎜ ⎟⎜ ⎟⎜− ⎝ ⎠( )b RC ,max ⎝ ⎠
( )
( )1 3
5 3 5 3 2 3
1 expTI TITI
K
D D
Γ α α ε σΦ
α α ρ ε
⎛ ⎞− ⎟⎜ ⎟⎜= − ⎟⎜ ⎟⎜ ⎟⎜⎝ ⎠( )b TI ,max f bD Dα α ρ ε⎜ ⎟⎜− ⎝ ⎠
( )2
4 1 BCD HC IC b c t Nuc Ja
Dn P N NΦ Φ Φ π α
⎧ ⎫⎪ ⎪⎪ ⎪= + = − +⎨ ⎬⎪ ⎪
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
( )CD HC IC b c t Nuc Jact
+ +⎨ ⎬⎪ ⎪⎪ ⎪⎩ ⎭
Benchmarking One-DimensionalI t f i l A T t E tiInterfacial Area Transport Equation
0
50
[m
-1] p0=0.170 MPa, Tf0=102 oC, G=492 kg/m2s
0
50 p0=0.103 MPa, Tf0=98.0 oC, G=139 kg/m2s
[m
-1]
-100
-50
i> - <
a i> z/D
=1
150
-100
-50
i> - <
a i> z/D
=1
-200
-150 Experimental Data (Zeitoun, 1994) Total Heat Transfer Inertia Void TransportRandom CollisionC
hang
e,
<a i
-250
-200
-150 Experimental Data (Zeitoun, 1994) Predicted IAC Change Heat Transfer Inertia Void TransportRandom CollisionC
hang
e,
<a i
0 1 2 3 4 5 6-250
Random Collision Turbulent Impact
IAC
C
Axial Position, z/D [-]0 2 4 6 8
-300
Random Collision Turbulent Impact
IAC
CAxial Position, z/D [ - ]
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Future DirectionF l ti f I t f i l A T t E tiFormulation of Interfacial Area Transport Equation
Formulation of Interfacial Area Transport EquationFormulation of Interfacial Area Transport Equation
1975 Ishii Basic concept of necessity of interfacial area transport equation
1995 Kocamustafaogullari and Ishii
Foundation of interfacial area transport equation
2003c Hibiki et al. Formulation of one-dimensional interfacial area transport equation in subcooled boiling flow
2003a Sun et al. Formulation of modified two-fluid model for two-gas momentum equations
2004 Ishii and Kim Formulation of two-group interfacial area transport equation
Future work • Extension of interfacial area transport equation to churn-turbulent-to-annular fl t itiflow transition
• Extension of interfacial area transport equation to annular and annular-mist flow regimes
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Future DirectionD l t f M t T h iDevelopment of Measurement Techniques
Development of Measurement TechniquesDevelopment of Measurement Techniques1986 Kataoka et al. Mathematical foundation of interfacial area concentration to be measured by local
probe technique
1992 Revankar and Ishii Demonstration of double-sensor probe technique 1993 Revankar and Ishii Demonstration of multi-sensor probe technique1998 Hibiki et al. Development of improved double-sensor probe technique1998 Hibiki et al. Application of hot-film anemometry to liquid velocity measurement
1999 Wu and Ishii Monte Carlo simulation of double-sensor probe technique2000 Kim et al. Development of improved multi-sensor probe technique2004c Sun et al. Application of laser Doppler anemometer to liquid velocity measurement
Future work • Improvement of local probe technique to be applicable to highly three-dimensional flow
• Application of film thickness probe to measure annular flow characteristicsA li i f d l h i l i fl• Application of droplet measurement technique to measure annular-mist flow characteristics
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Future DirectionD t b C t tiDatabase Construction
Database Construction
1998 Hibiki et al. Upward bubbly flow in vertical pipe (gas and liquid phases)
1999 Hibiki and Ishii Upward bubbly flow in vertical pipe (gas and liquid phases)p y p p (g q p )
2001 Bartel et al. Upward boiling bubbly flow in vertical annulus
2001a Hibiki et al. Upward bubbly flow in vertical pipe (gas and liquid phases)
2002 Sun et al. Upward bubbly flow in vertical large diameter pipe
2003a Hibiki et al. Downward bubbly flow in vertical pipe
2003b Hibiki et al. Upward bubbly flow in vertical annulus
2003 Kim et al. Upward bubbly flow in confined channel
2003b Sun et al. Upward cap-turbulent and transition to slug flows in vertical large diameter pipe
2003 Takamasa et al Bubbly flow in pipe under microgravity conditions2003 Takamasa et al. Bubbly flow in pipe under microgravity conditions
2004 Situ et al. Upward boiling bubbly flow in vertical annulus
2004a Sun et al. Upward cap-turbulent and churn-turbulent flows in confined channel
2005 Situ et al. Bubble lift-off and departure diameters
2007 Hazuku et al. Upward annular flow in vertical pipe
2007 Hibiki et al. Upward bubbly flow in vertical mini-channel
2008 Jeong et al. Upward cap-turbulent and churn-turbulent flows in vertical annulus
2008 Situ et al Bubble departure frequency2008 Situ et al. Bubble departure frequency
Future work • Development of extensive slug, churn-turbulent and annular flow data• Development of extensive data at elevated pressure• Development of extensive data in various flow channels (geometry, orientation and size)• Development of extensive wall nucleation data (active nucleation site density, bubble departure size and
frequency)• Development of extensive condensation and boiling data
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
• Development of extensive condensation and boiling data
Future DirectionSi k d S T M d liSink and Source Term Modeling
Sink and source term modelingSink and source term modeling1983 Kocamustafaogullari and
IshiiActive nucleation site density
1989 Riznic and Ishii Flashing source term1998 Wu et al. One-group model in pipe2000a Hibiki and Ishii One-group model in pipe2000b Hibiki and Ishii Two-group model in pipe2001b Hibiki et al. One-group model in small-diameter pipeg p p p2003a Fu and Ishii Two-group model in pipe2003 Hibiki and Ishii Active nucleation site density2004b Sun et al. Two-group model in confined channel2005 Situ et al Bubble lift off diameter2005 Situ et al. Bubble lift-off diameter2007 Park et al. Condensation sink term2008 Situ et al. Bubble departure diameter2008 Situ et al. Bubble departure frequency
Future work • Improvement of two-group model• Improvement of bubble departure diameter model• Improvement of bubble departure frequency model• Development of bulk boiling source model
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Future DirectionI l t ti i t CFD C dImplementation into CFD Codes
Implementation into CFD codeImplementation into CFD codeFuture work • Implementation of interfacial area transport equation into CFD code
• Benchmarking CFD code against data showing fully 3-D behavior
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Future DirectionM d li f I t f i l F d T b l M d lModeling of Interfacial Forces and Turbulence Models
• Lift Force Model • Zero-Equation Model
Interfacial Force Models Turbulence Models
• Tomiyama et al. (2002)• Hibiki and Ishii (2007)
• Wall Lubrication Force Model
• Sato et al. (1981) • One-Equation Model
• Kataoka and Serizawa (1995) • Antal et al. (1991)• Tomiyama (1998)
• Turbulence Dispersion Force Model
( )• Two-Equation Model
• Lopez de Bertodano et al. (1994)Turbulence Dispersion Force Model
• Lahey et al. (1993)• Burns et al. (2004)
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
Presentation OutlinePresentation OutlineIntroductionFormulation of Two-Fluid Model with Interfacial Area Transport Eq.p q
Two-Group Interfacial Area Transport Eq.Two-Group Momentum Eq.
Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Modeling of Sink and Source Terms in Interfacial Area Transport Eq.Sink and Source Due to Bubble Breakup and CoalescenceSink and Source Terms Due to Phase ChangeSource Term Due to Wall NucleationSource Term Due to Wall Nucleation
Database to Evaluate Interfacial Area Transport Eq.Local Interfacial Area MeasurementDatabase for 8 X 8 Rod Bundle GeometryDatabase for 8 X 8 Rod Bundle Geometry
Benchmarking Interfacial Area Transport Eq.Benchmarking 1-D IATE in Adiabatic SystemsBenchmarking 1 D IATE in Condensation SystemsBenchmarking 1-D IATE in Condensation Systems
Future DirectionsConclusions
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University
ConclusionsConclusions
In relation to the modeling of the interfacial transfer terms in the two-fluid model, the concept of the interfacial area transport equation has been proposed to develop a constitutive relation for the interfacial area concentration. The changes in the two-phase flow structure can be g ppredicted mechanistically by introducing the interfacial area transport equation.(1) The basic concept of the interfacial area transport equation and its(1) The basic concept of the interfacial area transport equation and its
formulation have been briefly explained.(2) Available models of interfacial area sink and source terms and
existing databases have been reviewed.(3) Newly obtained data for 8 X 8 rod bundle geometry has been
presentedpresented.(4) The interfacial area transport equation has been benchmarked
using adiabatic bubbly flow and condensation bubbly flow data. (5) Future direction for this research has been also suggested.
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University