Full-Scale Bundle Thermal-Hydraulics Tests in Support of ...
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Full-Scale Bundle Thermal-Hydraulics Tests in Support of ACR
Fuel Design By Laurence Leung, Thermalhydraulics Branch
Presented to US Nuclear Regulatory CommissionWashington DC
February 18-20, 2004
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Outline
• Introduction
• General ACR fuel specifications
• Full-scale bundle tests– Horizontal water flow– Freon flow
• Summary
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Introduction
• Experimental data in support of ACR fuel design– Critical heat flux (CHF)– Fuel-string pressure drop– Post-CHF heat transfer
• Purpose– Establish operating margin– Quantify impact of separate effects
• Full-scale bundle tests– High-pressure steam-water flow– Freon flow
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General Bundle Specification
• CANFLEX ACR design– 43 rods (1, 7-, 14-, and 21-rod rings)– Two sizes of rods– 0.5 m (19.69 inch) nominal bundle length– Bearing pads at outer-ring rods– Spacers at midplane– Two planes of buttons– Endplates
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General Fuel Specification• Rods in rings contain
slightly enriched uranium (SEU) fuel
• Centre rod contains a mixture of dysprosium (Dy) and natural uranium (NU)
• Radial power distribution– Depends on enrichment and
fuel burnup– Steeper profile for high
enrichment and fresh fuel– Profile flattens with
decreasing enrichment and increasing burnup
SEU fuel
Mixture of Dy and NU fuel
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Radial Heat-Flux Profiles in Various Bundles of a Fuel Channel
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0.2
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0.6
0.8
1
1.2
1.4
Outer Intermediate Inner Centre Inner Intermediate Outer
Ring
Loca
l-to-
Ave
rage
Hea
t-Flu
x R
atio
Bundle 1Bundle 7Bundle 9Bundle 12
Bundle 7 Bundle 9 Bundle 12Bundle 1
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0.2
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0.6
0.8
1.0
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1.4
1.8
0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.00.0
Axial Location (m)
CANDU 6 (8-Bundle Shift)ACR (2-Bundle Shift)
1.6
Comparison of Axial Power Profiles
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Preliminary CHF Power Analysis
• Initial dryout occurrences– Either the 10th or 11th bundle – Close to discharge burnup– Radial power profiles are similar to that of natural
uranium fuel• Drypatch spreading
– Gradually to upstream and downstream bundles with increasing power
– Minor CHF reduction due to variation of radial power profile in upstream bundles
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Critical Heat Flux Experiments• Full-scale bundle simulators with junction and appendages• Detailed surface scans to detect the first dryout occurrence• Horizontal high-pressure water tests
– Primary database for licensing calculations– Relevant radial and axial power profiles– New and aged channel profiles– Separate effects (e.g., flow and power transients)
• Freon tests (simulating high-pressure water conditions)– Supplemental database for design and licensing calculations– Various radial and axial power profiles– Various channel profiles (uniform and non-uniform crept
channels)– Separate effects (e.g., channel orientation, appendage
configurations)
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Horizontal High-Pressure Water Tests
• 6-m long bundle simulator
• Non-uniform axial and radial power profiles
• Uncrept and crept flow tubes
• Well-instrumented– Loop– Bundle
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High Pressure Water Test Station
P4
Tout
P2
P3P1
dP1
Tin
Coolant Flow
dP2 dP13 dP3 dP14 dP4 dP5 dP7 dP8 dP9 dP10 dP11
dP12
dP6
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Water Test Facility
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Bundle Simulators• 6-m (20 ft) long full-
scale bundle strings with junction and appendages
• Non-uniform axial and radial power distributions
• Sliding thermocouples inside rods at several downstream bundles in the string
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Sliding Thermocouple Drive Unit
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Sliding Thermocouple Assembly
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Axial Power Profile in Water Tests
Axial Distance (m)0 1 2 3 4 5 6
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0.5
1
1.5
2Coolant Flow
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1-1 2 3 4 5 6 7
109
Axial Distance (m)0
108
107
106
105
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103
110
Axial Flow Tube Diameter VariationFlow Direction
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Test Conditions
10 to 75 oC
(18 to 135 oF)
7 to 23 kg.s-1
(15 to 51 lb.s-1)
6 to 11 MPa
(870 to 1600 psi)
CHF database
49 oC
(88 oF)
Inlet subcooling
26 kg.s-1
(57 lb.s-1)
Mass flow rate
12.5 MPa
(1800 psi)
Channel outlet pressure
ACR conditions
10 to 75 oC
(18 to 135 oF)
7 to 23 kg.s-1
(15 to 51 lb.s-1)
6 to 11 MPa
(870 to 1600 psi)
CHF database
49 oC
(88 oF)
Inlet subcooling
26 kg.s-1
(57 lb.s-1)
Mass flow rate
12.5 MPa
(1800 psi)
Channel outlet pressure
ACR conditions
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Experimental Data
• Onset of Nucleate Boiling
• Onset of Significant Void
• Single and Two-Phase Pressure Drops
• CHF
• Post-CHF Clad Temperature
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Onset of Nucleate Boiling
1.4 mm BP, 5.1% Crept Channel270
280
290
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320
330
0 2000 4000 6000 8000
Power (kW)
Cla
d Te
mpe
ratu
re (º
C)
01K104K107K109K112K122K129K140K142K143K1
Dimmick et al., Proc. 6th Int. Conf. on CANDU Fuel, Niagara Falls, Canada, September 26-30, 1999
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Onset of Significant Void
Dimmick et al., Proc. 6th Int. Conf. on CANDU Fuel, Niagara Falls, Canada, September 26-30, 1999
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10
20
30
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50
60
70
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90
0 2000 4000 6000 8000
Power (kW)
Pres
sure
Dro
p (P
a)
1.4 mm BP, 5.1% Crept Channel
DP8
DP9
DP10
DP11
DP12
DP13
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Pressure Gradient Along Flow Channel
Dimmick et al., Proc. 6th Int. Conf. on CANDU Fuel, Niagara Falls, Canada, September 26-30, 1999
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300
250
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50
0
400
2 3 4 5 6 7 8 9 111 12 13 1410Pressure Tap
Test ConditionsMass Flow Rate: 19 kg/sOutlet Pressure: 10.5 MPaInlet Temperature: 285 C
5.1% Crept Channel1.4 mm Bearing Pads Two-phase
Power: 5700 kWOutlet Quality: 2%
Single-phasePower: 2700 kWOutlet Quality: -10%
Two-phasePower: 6500 kWOutlet Quality: 5%
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Bundle I
Clad Temperature Map at Dryout
Bundle J
Bundle K
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Critical Power in Uncrept Channel1.7 mm BP, Uncrept Channel, 11 MPa
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6000
7000
8000
9000
10000
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12000
13000
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Mass Flow Rate (kg/s)
Crit
ical
Pow
er (k
W)
243 C268 C280 C290 C
Leung et al., Proc. 7th Int. Conf. on CANDU Fuel, Kingston, Canada, September 23-27, 2001
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Critical Power in Crept Channel1.4 mm BP, 5.1% Crept Channel, 11 MPa
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5000
6000
7000
8000
9000
10000
8 10 12 14 16 18 20 22 24 26
Mass Flow Rate (kg/s)
Crit
ical
Pow
er (k
W)
243 C255 C268 C280 C290 C
Dimmick et al., Proc. 6th Int. Conf. on CANDU Fuel, Niagara Falls, Canada, September 26-30, 1999
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Post-CHF Clad Temperature
Dimmick et al., Proc. 6th Int. Conf. on CANDU Fuel, Niagara Falls, Canada, September 26-30, 1999
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6050 6100 6150 6200 6250 6300Bundle power kW
Cla
d Te
mpe
ratu
re C
TC 06J1
1.4 mm BP, 5.1% Crept Channel
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Effect of Channel Diametral Creep on Bundle CHF
• Pressure-tube diameter increases due to creep
– Increase in total flow area– Increase in subchannel flow
areas at the top portion of the bundle
• Impact on critical channel power
– CHF reduction due to flow bypass effect
– Mass flow rate increase due to pressure-drop reduction, compensating the CHF reduction
• Experimental data available to quantify the impact
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Effect of Channel Creep1.7 mm Bearing Pads, Outlet Pressure: 11 MPa, Mass-Flow Rate: 21 kg.s-1
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7000
8000
9000
10000
11000
12000
13000
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Channel Maximum Creep (%)
Dry
out P
ower
(kW
)
243268280290
Inlet-Fluid Temp. (oC)
Leung et al., Proc. 7th Int. Conf. on CANDU Fuel, Kingston, Canada, September 23-27, 2001
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Effect of Bearing-Pad Height on Bundle CHF
Leung et al., Proc. 7th Int. Conf. on CANDU Fuel, Kingston, Canada, September 23-27, 2001
• Reduce bundle eccentricity – Bundle sits closer to
center of the pressure tube
– Reduce flow bypass at top subchannels between outer rods and pressure tube
• Increase gap size between bottom rods and pressure tube– Increase local subchannel
flow and heat transfer• Improve CHF at outer rods
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Effect of Bearing-Pad HeightUncrept Channel, Outlet Pressure: 11 MPa, Mass-Flow Rate: 17 kg.s-1
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8500
9000
9500
10000
10500
11000
1.3 1.4 1.5 1.6 1.7 1.8 1.9
Bearing-Pad Height (mm)
Dry
out P
ower
(kW
)
243268280
Inlet-Fluid Temp. (oC)
Leung et al., Proc. 7th Int. Conf. on CANDU Fuel, Kingston, Canada, September 23-27, 2001
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Freon Tests• Full-scale bundle simulators
• Uniform and non-uniform axial power profiles
• Non-uniform radial power profiles (variable)
• Uncrept and crept flow tubes (axially uniform variation)
• Vertical and horizontal channels
• Well-instrumented – Loop– Bundle
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Freon Test Facility
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Vertical Test Station• Steel pressure
boundary
• Fibreglass liner
• Upward co-current flow
• 14 pressure taps
Copper Extension Rods
Copper Bus Connection
Shell
HeaterSection
String
Liner
All dimensions in mm.
Nylon Grid Spacer
Sliding ThermocoupleExtension Rods
381PDT
PT- 361
PRESSURE
384PDT
380PDT
379PDT
378PDT
377PDT
376PDT
375PDT
374PDT
373PDT
372PDT
371PDT
No 1
No 2
No 3
No 4
No 5
No 6
No 7
No 8
No 9
No 10
No 11
No 12
No 13
No 14
382PDT
5944
1097652
135
OUTLET
PT- 357
PRESSUREINLET
383PDT
514
380
514
494
514
494
494
494
494
494
494
475
342232
7908
7940
494
495.3
Test Station
Fiberglas
Heater
Sliding ThermocoupleExtension Rods
Copper Bus ConnectionNylon Grid Spacer
Copper Extension Rods
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Horizontal Test Station
381PDT
PT- 361
PRESSURE
Test Station Shell
Electrical BusHeaterSection
Heater String Fiberglas Liner
CopperExtension
RodsAll dimensions in mm.
Nylon
384PDT
380PDT
379PDT
378PDT
377PDT
376PDT
375PDT
374PDT
373PDT
372PDT
371PDT
No 1 No 2 No 3 No 4 No 5 No 6 No 7 No 8 No 9 No 10 No 11 No 12 No 13 No 14
382PDT
ConnectionGrid Spacer
5944
1097
652
136
Sliding
Extension RodsThermocouple
Electrical Bus
CopperExtension
Rods
NylonConnection Grid Spacer
Sliding
Extension RodsThermocouple
OUTLETPT- 357
PRESSUREINLET
383PDT
514
380
514 494514 494 494 494 494 494 494 475 342
232
79087940
494
495.3
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Adjustable Radial Power Profile
+-
43-ELEMENTBUNDLE
1.7 MWattPOWER SUPPLY
CENTRE ELEMENT
Ω0.59
Ω0.085
Ω0.054
Ω0.029
INNER RING
INTERMEDIATE RING
OUTER RING
Ω0.012
Ω0.0
Ω0.0002
Ω0.0005
RESISTORBANK
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Effect of Radial Power Profile on CHF
• Radial power profile varies with enrichment and burnup
• A variety of radial power profiles have been tested
• Axially uniform-heated, Freon-cooled, bundle string– Base radial power profile
corresponding to natural uranium fuel
– Radial power profile adjustment using external variable-resistor bank
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Outer Intermediate Inner Centre Inner Intermediate OuterRing
Loc
al-
to-A
vera
ge
Hea
t-F
lux
Ra
tio
NU1.2% SEU (Fresh)1.2% SEU (Discharge)1.6% SEU (Fresh)2.7%-LVRF (Fresh)2.7%-LVRF (Mid-burn-up)2.7%-LVRF (Discharge)
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Outer Intermediate Inner Centre Inner Intermediate OuterRing
Loc
al-
to-A
vera
ge
Hea
t-F
lux
Ra
tio
NU1.2% SEU (Fresh)1.2% SEU (Discharge)1.6% SEU (Fresh)2.7%-LVRF (Fresh)2.7%-LVRF (Mid-burn-up)2.7%-LVRF (Discharge)
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Freon CHF Data for the 1.6% SEU and NU Radial Power Profiles
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2000
2500
0 0.05 0.1 0.15 0.2 0.25 0.3 0.35
Critical Quality
Crit
ical
Hea
t Flu
x (k
W.m
-2)
NU, 6.3SEU, 6.3NU, 7.6
SEU, 7.6
Mg.m-2.s-1
Mg.m-2.s-1
Mg.m-2.s-1
Mg.m-2.s-1
Leung, Proc. 22nd CNS Nuclear Simulation Symposium, Ottawa, Ontario, November 3-5, 2002
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CHF Ratios for Various Radial Heat-Flux Distributions
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CH
F R
atio
w.r.
t. Fr
esh
NU
RFD
NU (Fresh)1.2% SEU (Fresh)1.2% SEU (Discharge)NU (Fresh)1.6% SEU (Fresh)2.7%-LVRF (Fresh)2.7%-LVRF (Mid-burn-up)2.7%-LVRF (Discharge)
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Summary
• Water and Freon experiments were performed using full-scale CANFLEX bundle simulators
• Experimental data on CHF, pressure drop and post-CHF clad temperature are applicable for the ACR fuel design
• Clad temperature is higher at bottom elements than top elements in the bundle
• Flow stratification has not been observed in all tests
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