full-scale bundle thermal-hydraulics tests in support of

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

Outline

• Introduction

• General ACR fuel specifications

• Full-scale bundle tests– Horizontal water flow– Freon flow

• Summary

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

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

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

Radial Heat-Flux Profiles in Various Bundles of a Fuel Channel

0

0.2

0.4

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

0.2

0.4

0.6

0.8

1.0

1.2

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

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

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)

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

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

Water Test Facility

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

Sliding Thermocouple Drive Unit

Sliding Thermocouple Assembly

Axial Power Profile in Water Tests

Axial Distance (m)0 1 2 3 4 5 6

0

0.5

1

1.5

2Coolant Flow

1-1 2 3 4 5 6 7

109

Axial Distance (m)0

108

107

106

105

104

103

110

Axial Flow Tube Diameter VariationFlow Direction

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

Experimental Data

• Onset of Nucleate Boiling

• Onset of Significant Void

• Single and Two-Phase Pressure Drops

• CHF

• Post-CHF Clad Temperature

Onset of Nucleate Boiling

1.4 mm BP, 5.1% Crept Channel270

280

290

300

310

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

Onset of Significant Void


0

10

20

30

40

50

60

70

80

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

Pressure Gradient Along Flow Channel


350

300

250

200

150

100

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%

Bundle I

Clad Temperature Map at Dryout

Bundle J

Bundle K

Critical Power in Uncrept Channel1.7 mm BP, Uncrept Channel, 11 MPa

5000

6000

7000

8000

9000

10000

11000

12000

13000

8 10 12 14 16 18 20 22 24 26

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

Critical Power in Crept Channel1.4 mm BP, 5.1% Crept Channel, 11 MPa

4000

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


Post-CHF Clad Temperature


300

320

340

360

380

400

420

440

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

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

Effect of Channel Creep1.7 mm Bearing Pads, Outlet Pressure: 11 MPa, Mass-Flow Rate: 21 kg.s-1

6000

7000

8000

9000

10000

11000

12000

13000

0 1 2 3 4 5 6

Channel Maximum Creep (%)

Dry

out P

ower

(kW

)

243268280290

Inlet-Fluid Temp. (oC)


Effect of Bearing-Pad Height on Bundle CHF


• 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

Effect of Bearing-Pad HeightUncrept Channel, Outlet Pressure: 11 MPa, Mass-Flow Rate: 17 kg.s-1

8000

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)


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

Freon Test Facility

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

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

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

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

0

0.2

0.4

0.6

0.8

1

1.2

1.4

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)

0

0.2

0.4

0.6

0.8

1

1.2

1.4

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)

Freon CHF Data for the 1.6% SEU and NU Radial Power Profiles

0

500

1000

1500

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

CHF Ratios for Various Radial Heat-Flux Distributions

0

0.2

0.4

0.6

0.8

1

1.2

1 1.05 1.1 1.15 1.2 1.25 1.3Bundle Imbalance Factor

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)

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

full-scale bundle thermal-hydraulics tests in support of

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