ML050410494
| ML050410494 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/10/2005 |
| From: | Billone M, Burtseva T, Chung H, Yan Y Argonne National Lab (ANL) |
| To: | Office of Nuclear Reactor Regulation |
| scott h h | |
| References | |
| Download: ML050410494 (90) | |
Text
H.B. Robinson High-Burnup PWR Oxidation and Post-Quench Ductility M. Billone, Y. Yan, T. Burtseva and H. Chung Energy Technology Division Argonne National Laboratory February 10, 2005 Argonne National Laboratory A U.S. Department of Energy Office of Science Laboratory Office of Science U.S. Department of Energy Operated by The University of Chicago
Summary of ANL Results Advanced Alloy Program (NSRC-2004, Billone)
- 1000 & 1100ºC oxidation, slow-cooled to 800ºC & quenched Good RT ductility to >17% Cathcart-Pawel (CP)-ECR (Zry-4, ZIRLO, M5)
- 1200ºC-oxidized, slow-cooled to 800ºC & quenched (Zry-4, ZIRLO, M5)
RT embrittlement at 9% ECR (Zry-4) & 12% ECR (ZIRLO & M5)
Significant ductility improvement at 135ºC (embrittlement ECR>17%)
Severe embrittlement (135ºC): pre-H Zry-4 (<10% ECR, >300 wppm H)
- Plan to test high-burnp ZIRLO and M5 LOCA Integral Test Results at 1204ºC (NSRC-2004, Yan)
- High-burnup BWR Zry-2 (significant embrittlement in balloon region)
Non-uniform wall-thinning, 2-sided oxidation, high secondary H pickup
- High-burnup PWR Zry-4 (test details to be determined from data below)
Baseline ductility data for as-fabricated and prehydrided Zry-4 rings Data for high-burnup Zry-4 rings oxidized at 1206ºC to 3-10% CP-ECR Pioneering Nuclear Science and Regulatory Technology Commission
ZIRLO vs. Zry-4 4 6 8 10 12 14 16 18 20 22 60 60 ZIRLO 1200°C 55 55 Zry-4 1200°C 50 50 45 45 Offset Strain , %
40 40 p / D0 , %
35 35 30 30 25 25 20 20 15 15 10 10 5 5 0 0 4 6 8 10 12 14 16 18 20 22 Measured ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
M5 vs. Zry-4 4 6 8 10 12 14 16 18 20 22 60 60 M5 1200°C 55 55 Zry-4 1200°C 50 50 45 45 Offset Strain , %
40 40 p / D0 , %
35 35 30 30 25 25 20 20 15 15 10 10 5 5 0 0 4 6 8 10 12 14 16 18 20 22 Measured ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
135ºC Offset Strain for 17x17 Zry-4 Oxidized at 1200ºC 4 6 8 10 12 14 16 18 20 22 30 30 RT PQD Data 25 135°C PQD Data 25 Offset Strain, %
20 20 p / D0 , %
15 15 10 10 5 5 0 0 4 6 8 10 12 14 16 18 20 22 Measured ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
135ºC Offset Strain for 17x17 ZIRLO Oxidized at 1200ºC 4 6 8 10 12 14 16 18 20 22 60 60 RT PQD Data 50 135°C PQD Data 50 Offset Strain, %
40 40 p / D0 , %
30 30 20 20 10 10 0 0 4 6 8 10 12 14 16 18 20 22 Measured ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
135ºC Offset Strain for 17x17 M5 Oxidized at 1200ºC 0 2 4 6 8 10 12 14 16 18 20 22 60 60 RT PQD Data 50 135°C PQD Data 50 Offset Strain, %
40 40 p / D0 , %
30 30 20 20 10 10 0 0 0 2 4 6 8 10 12 14 16 18 20 22 Measured ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
Oxidized at 1200ºC Pioneering Nuclear Science and Regulatory Technology Commission
LOCA Integral Test Sequence & Time Pressurization Permeability t BJ t CP 1200 * *
- 1200 Cladding Temperature (°C) 0 - 5 min.
Oxidation 3°C/s Pressure (psig) 900 Burst
- 5°C/s *
- 600 600 Quench 300 *
- Steam 15 30 Time (minutes)
Pioneering Nuclear Science and Regulatory Technology Commission
Balloon and Burst Regions for High-burnup Tests Burst length: 14 mm Burst length: 13 mm Max. burst width: 3.5 mm Max. burst width: 3.0 mm ICL#1: Ramp-to-Burst test conducted in ICL#2: LOCA sequence with 5-minute oxidation argon at 1204ºC and slow-furnace cooling Burst length: 11 mm Burst length: 15 mm Max. burst width: 4.6 mm Max. burst width: 5.1 mm ICL#3: 5-min. oxidation at 1204ºC fFollowed by ICL#4: Full LOCA sequence (5-minute oxidation quench at 800°C (quartz tube failed at 480°C) at 1204ºC) with quench at 800°C Pioneering Nuclear Science and Regulatory Technology Commission
Post-test Characterization for ICL#3 Specimen 1.0 5/8 1-1/4 SEM SEM Bottom A B C D Top Sample ICL#3 was broken at locations A, B and C during the sample handling before the sectioning was performed at location D.
Pioneering Nuclear Science and Regulatory Technology Commission
H and O Analyses of ICL#2 and ICL#3 Hydrogen Pickup, wppm ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
for Prehydrided 15x15 Zry-4 Oxidized at 1204+/-10ºC Pioneering Nuclear Science and Regulatory Technology Commission
135ºC vs. CP-ECR for High-Burnup Zry-4 Cladding 60 670 +/- 210 wppm H, Tmax = 1140°C 600 +/- 90 wppm H, Tmax = 1185°C 780 +/- 70 wppm H, Tmax = 1204°C 50 750 +/- 100 wppm H, Tmax = 1206°C 545 +/- 80 wppm H, Tmax = 1206°C 40 Offset Strain, %
30 20 10 0
0 2 4 6 8 10 12 Predicted (CP) ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
for Non-Deformed 15x15 Zry-4 Rings
- Baseline Data for Non-Irradiated 15x15 Zry-4 Cladding
- Weight gain (1204ºC); PQD vs. ECR, H-content & test temperature
- HBR Rod Selection (F07) and Characterization
- Gamma scanning, corrosion thickness, hydride morphology
- Hydrogen- and oxygen-content profiles (axial and circumferential)
- HBR F07 Oxidation Test Samples and Results mm-long samples cut and defueled from F07 midplane (-40 to 100 mm)
- Two-sided steam oxidation tests conducted at 1204ºC Tmax to 3, 5, 7, 8, 10% ECR without quench 8% ECR with quench (to be conducted)
- Post-Oxidation (POD) and Post-Quench Ductility (PQD)
- POD offset & permanent strains for ring-compression tests at 135ºC
- PQD offset & permanent strains for ring-compression tests at 100-135ºC Pioneering Nuclear Science and Regulatory Technology Commission
Baseline Data for Non-irradiated 15x15 Zry-4 Cladding
- Weight Gain due to Steam Oxidation at 1204ºC
- Thermal benchmark results with 3 TCs welded onto sample and holder Subsequent tests conducted with 3 TCs welded onto holder above sample
- Data are in excellent agreement with Cathcart-Pawel (CP) correlation Hydrogen content and quench have no significant effect on weight gain Note 2 data points at about 8.7 mg/cm2 : one with & one without quench
- Post-Quench Ductility of 1204ºC-Oxidized/Quenched Zry-4
- Nonirradiated 15x15 Zry-4 cladding Offset & permanent strains vs. ECR based on measured weight gain Ductile-to-brittle trans. ECR: 8% at RT, 11.5% at 100ºC, 14% at 135ºC Offset strain vs. ECR calculated with CP correlation: trend curves
- Pre-hydrided, non-irradiated 15x15 Zry-4 cladding; 135ºC tests 5% CP-ECR: ductile-to-brittle transition H-content 600 wppm 7.5% CP-ECR: ductile-to-brittle transition H-content 400 wppm Pioneering Nuclear Science and Regulatory Technology Commission
Cladding Sample TC Readings Benchmark Test HBRU#20 at 1204°C, 6/16/2004 1400 1200 1000 Temperature (°C) 800 Holder TC (120°)
600 Holder TC (0°)
Holder TC (240°)
400 Sample TC (0°)
Sample TC (120°)
200 0
300 400 500 600 700 800 900 1000 Time (s)
Pioneering Nuclear Science and Regulatory Technology Commission
Cladding Sample TC Readings (Contd)
Benchmark Test HBRU#29 at 1204°C, 6/29/04 1400 1200 1000 Temperature (°C) 800 600 Holder TC (120°)
Holder TC (0°)
400 Holder TC (240°)
Sample TC (0°)
200 Sample TC (240°)
0 300 400 500 600 700 800 900 1000 Time (s)
Pioneering Nuclear Science and Regulatory Technology Commission
of 15x15 Zry-4 Post-Quench-Ductility Samples 1400 7.5% ECR 13% ECR 5% ECR 1200 1.3% ECR 10% ECR 1000 Temperature (°C) 800 600 TC on specimen at 120º 400 TC on specimen at 0º TC on specimen at 240º 200 0
400 500 600 700 800 900 Time (s)
Pioneering Nuclear Science and Regulatory Technology Commission
Cladding following Steam oxidation at 1204+/-10°C 16 As-received: about 10 wppm Prehydrided: 300-800 wppm 12 Measured WG, mg/cm2 8
4 0
0 4 8 12 16 CP-Predicted Weight Gain (WG), mg/cm2 Pioneering Nuclear Science and Regulatory Technology Commission
for 15x15 Zry-4 Oxidized at 1204+/-10ºC Pioneering Nuclear Science and Regulatory Technology Commission
for 15x15 Zry-4 Oxidized at 1204+/-10ºC Pioneering Nuclear Science and Regulatory Technology Commission
As-Received 15x15 Zry-4 Oxidized at 1204+/-10ºC Pioneering Nuclear Science and Regulatory Technology Commission
for Prehydrided 15x15 Zry-4 Oxidized at 1204+/-10ºC Pioneering Nuclear Science and Regulatory Technology Commission
Discussion of Baseline Data for 15x15 Zry-4
- Weight Gain Data not Plotted in Figure
- Results from 0304 test train benchmarked to 1200+/-17ºC Good agreement with CP model for 6 data points at 5-17% ECR
- Results from 0604 test train also generated at 1174, 1184 & 1194ºC
- Results from new (1204) test train at 1204+/-10ºC: 5% & 10% ECR
- Post-Quench Ductility (PQD) Data: 0604 Test Train
- As-Received Zry-4 PQD (low) the same for 1184-1204ºC oxidation/quench at 13% ECR PQD enhancement for 1174ºC oxidation temperature:
Permanent strain increase: 1.53.4% at 135ºC, 0.61.0% at 100ºC
- Pre-hydrided Zry Low Transition ECR Value Depends on T-Ramp At end of 5% ECR oxidation, Avg. T = 1193ºC & Max. T = 1204ºC At end of 7.5% ECR oxidation, Avg. T = 1208ºC & Max. T = 1218ºC Pioneering Nuclear Science and Regulatory Technology Commission
HBR Rod Selection for Oxidation, PQD and LOCA Tests
- Edge-Near-Corner Rods: A02, B01, R01, S02
- Characterization of A02 and R01 Effective flow and moderation area not uniform around rods Fuel & cladding temperatures not symmetric with fuel centerline axis Significant hoop variation in cladding hydrogen content ( +/-150 wppm)
Significant hoop variation in hydride morphology
- Rods at Corners of Guide Tubes: B05, E02, R05, E14
- May have more symmetric cladding temperatures than edge rods
- Gamma scanning completed for Rod E02 - may be okay for testing
- Interior Rods not Next to Guide Tubes: F07, G10
- Should have axi-symmetric power profile & cladding temperatures
- Should have more uniform hydrogen content and hydride morphology
- Select Rod F07 for initial characterization and testing Pioneering Nuclear Science and Regulatory Technology Commission
Final Configuration of HBR Rods Rod B01 Rod E02 Rod R01 Rod A02 Rod S02 Rod B05 Rod R05 Rod F07 Rod G10 Rod E14 Pioneering Nuclear Science and Regulatory Technology Commission
Characterization of HBR Rod F07
- Gamma Scanning of Middle Segment (C)
- Middle of grid spacer #4 is 250 mm from bottom of Segment C
- Fuel midplane is 160 mm below center of grid spacer #4
- Characterization, Oxidation, LOCA Sample Locations
- Metallography at 50 mm, 300 mm & 660 mm above fuel midplane
- H & O analysis at 25 mm, 320 mm and 650 mm above fuel midplane
- Oxidation samples at -40 to 100 mm above fuel midplane
- Characterization Results
- Fuel morphology: heat generation and T are not axisymmetric!
- Metallography at +50 mm : oxide layer thickness = 71+/-5 µm Hydride morphology: relatively uniform in hoop () direction
- H-content: 760+/-60 wppm at +13 mm; 670+/-80 wppm at +32 mm Too high and too much axial variation: pellet-pellet interfaces??
- O-content: 2.08+/-0.19 wt.% at +35 mm above midplane Pioneering Nuclear Science and Regulatory Technology Commission
Note: Dips at Pellet-Pellet Interfaces Every 7 mm HBR Rod F07 Gamma Scan Results 10000 GROSS200 9000 8000 7000 Grid Span #4 Gamma counts 6000 GS#4 5000 4000 3000 Fuel 2000 Midplane 1000 0
0 5 10 15 20 25 30 35 Position, in Pioneering Nuclear Science and Regulatory Technology Commission
Sectioning Diagram for HBR Rod F07 Segment C A/G 607C: 33-long C7 C6 C5 C4 C2 C3 C1 12.0 15.0 0.5 1.0 1.0 0.5 3.0 GS #4 Fuel Midplane A/G 607C7: 15-long 7H 7G 7F 7E 7D 7C 7B 7A 2-1/8 1.0 1.0 1.0 0.5 1.0 1.0 7-3/8 Oxidation & Characterization Samples Pioneering Nuclear Science and Regulatory Technology Commission
50 mm above Fuel Midplane Higher T Lower T Pioneering Nuclear Science and Regulatory Technology Commission
50 mm above Fuel Midplane Pioneering Nuclear Science and Regulatory Technology Commission
50 mm above Fuel Midplane Fuel-Cladding Bond: 11+/-4 µm (Remains after Defueling in RT HNO3)
Cladding Fuel Rim Region Pioneering Nuclear Science and Regulatory Technology Commission
50 mm above Fuel Midplane Pioneering Nuclear Science and Regulatory Technology Commission
at 50 mm above Fuel Midplane (Low Mag.)
Pioneering Nuclear Science and Regulatory Technology Commission
at 50 mm above Fuel Midplane (High Mag.)
Pioneering Nuclear Science and Regulatory Technology Commission
at 50 mm above Fuel Midplane (High Mag.)
Pioneering Nuclear Science and Regulatory Technology Commission
Prehydrided HBR Baseline Cladding (15x15 Zry-4)
- RT Ring Compression: Non-Irradiated 15x15 Zry-4 mm-long sample displaced to 2 mm and unloading at RT Loading stiffness (1.34 kN/mm) consistent with elastic model Unloading stiffness < loading stiffness Offset displacement is 0.13 mm (1.2%) > permanent displacement mm-long sample displaced to 2 mm and unloading at RT at 135ºC Larger decrease in yield strength than in stiffness (Youngs modulus)
- RT Ring Compression: Prehydrided Zry-4 mm-long rings compressed to failure (full-length through-wall crack)
Circumferential hydrides (300 wppm): high ductility = 45%
- Radial-Hydride-Treated (RHT) samples cooled from 400ºC under RHT (400ºCRT at 150 MPa, 750 wppm): low ductility = 3.1%
RHT (400ºCRT at 190 MPa, 300 wppm): brittle = 0% ductility Pioneering Nuclear Science and Regulatory Technology Commission
Irradiated, HBR Baseline 15x15 Zry-4 1
0.9 0.8 0.7 0.6 0.5 0.4 Initial Loading Slope = 1.34 kN/mm 0.3 0.2 Offset Displacement = 1.32 mm 0.1 Permanent Displacement = 1.19 mm 0
0 0.5 1 1.5 2 2.5 D isplacement ( mm)
Pioneering Nuclear Science and Regulatory Technology Commission
Results for Non-Irradiated, HBR Baseline 15x15 Zry-4 1
Ring-Compression Test 15x15 F-ANP Zircaloy-4 0.9 RT 135°C 0.8 0.7 0.6 Load (kN) 0.5 0.4 0.3 0.2 12.3% Offset 13.6% Offset 0.1 0
0 0.5 1 1.5 2 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
HBR Baseline 15x15 Zry-4 1.2 Room-Temperature Ring-Compression Results for Prehydrided Zry-4 Samples 1
0.8 15x15 Zry-4 (no RHT, 285 wppm H) 15x15 Zry-4 (400°C/150 MPa RHT, 265 wppm H)
Load (kN) 15x15 Zry-4 (400°C/190 MPa RHT, 760 wppm H) 0.6 0.4 0.2 0
0 1 2 3 4 5 6 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
Cladding (15x15 Zry-4)
- RT Load-Displacement Test Results
- 9.4-mm-long ring with 70 µm corrosion layer and 700 wppm H Near midplane of Rod F07 (see specimen 7E in Slide 15)
- Loading stiffness (1.92 kN/mm) consistent with increased modulus (E)
- Evidence of small load drops due to oxide cracking from 0.8-3.7% offset
- 1st large load drop (0.2 kN) at 3.7% offset partial wall crack
- 2nd large load drop (0.2 kN) at 8.3% offset crack extension Test stopped at 8.6% offset and 6.3% permanent strains Sample crack at 90º from loading line, extending 75% from OD to ID
- Test results for cladding with corrosion layer removed Smooth load-displacement curve with load drop at 10.6% offset strain 2 partial-wall cracks at +90º and -90º extending over full length of ring
- 135ºC Load-Displacement Test Results (more ductile)
Pioneering Nuclear Science and Regulatory Technology Commission
HBR Cladding Near Fuel Midplane 1.2 As-Irradiated H.B. Robinson Cladding
( 700 wppm H, 70 µm oxide layer)
Room-Temperature Ring Compression 1
0.8 Load (kN) 0.6 0.4 0.2 0.8% Offset 3.7% Offset 8.6% Offset 0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 1.6 Displacem ent (m m )
Pioneering Nuclear Science and Regulatory Technology Commission
Low-Magnification Images of Crack in HBR Sample 1X View of Crack 3X View of Crack at 90º from along Side of Ring Loading Line Pioneering Nuclear Science and Regulatory Technology Commission
High-Burnup HBR Cladding Near Fuel Midplane 1
HBR Ring Compression
(~700 wppm H) 0.9 RT 135°C 0.8 0.7 Normalized Load (kN) 0.6 0.5 0.4 0.3 0.2 0.1 8.5% Offset 19.5% Offset 0
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 2.2 2.4 2.6 2.8 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
Low Mag. Images of Crack in HBR Ring Tested at 135ºC Pioneering Nuclear Science and Regulatory Technology Commission
HBR Cladding with Corrosion Layer Removed 1.2 HBR cladding sample with corrosion layer removed from grid span 5 of Rod A02 540+/-40 wppm hydrogen 1
0.8 Load (kN) 0.6 0.4 0.2 10.6% Offset 0
0 0.5 1 1.5 2 2.5 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
with the Corrosion Layer Removed Prior to RT Testing 1X Image at 90º 3X Image of Crack Growth from Loading Line through Side Wall Pioneering Nuclear Science and Regulatory Technology Commission
Discussion of Characterization Results for HBR Rod F07
- Oxide Layer & Oxygen Content near Fuel Midplane
- Oxide layer thickness: F07 (71+/-5 µm) consistent with A02 (70+/-5 µm)
- Oxygen content of as-irradiated cladding Predicted assuming PB ratio of 1.75: 2.09 wt.%
Measured at +35 mm: 2.08+/-0.19 wt.% (good agreement)
Measured at +10 mm: 1.99+/-0.40 wt.% (some oxide lost during cutting)
- Hydrogen Content near Fuel Midplane
- Expected = 550-580 wppm based on A02 and R01
- Measured: 760+/-60 wppm at +13 mm; 670+/-80 wppm at +32 mm
- Re-measured over 14 mm (7H) just below midplane: 550+/-90 wppm
- Hydride Morphology near Fuel Midplane
- Appears symmetric with respect to the fuel centerline axis
- Large H variations with respect to should be visible in metallography Pioneering Nuclear Science and Regulatory Technology Commission
Re-Measurement of Hydrogen Content White dot A/G 607C7H Top A/G 607C7H9 H8 H7 H6 H5 H4 H3 H2 H1 1.5mm 1.5mm 1.5mm1.5mm 1.5mm1.5mm 1.5mm1.5mm 33mm H1: 545+/-85 wppm H; H2: 530+/-80 wppm H H3: 530+/-90 wppm H; H4: 535+/-85 wppm H H5: 555+/-85 wppm H; H6: 575+/-80 wppm H H7: 535+/-80 wppm H; H8: 565+/-70 wppm H Pioneering Nuclear Science and Regulatory Technology Commission
Steam Oxidation Tests of F07 Cladding Samples 0604 Test Train Benchmarked In-Cell (Sample Weight Gain)
- Nonirradiated 15x15 Zry-4 at 1204ºC to 10% ECR Excellent agreement with CP-model and out-of-cell results
- Thermocouple wires damaged during handling before HBR cladding sample could be run (December 2004)
New Test Train (1204) Benchmarks with As-Fabricated Zry-4
- Out-of-cell thermal benchmark for sample T vs. control T on holder
- Out-of-cell weight-gain benchmark at 5% and 10% ECR
- In-cell weight-gain benchmark at 10% ECR
- Out-of-cell and in-cell results in excellent agreement with CP model In-cell Oxidation Tests Conducted with HBR F07 Cladding
- CP-ECR = 3%, 5%, 7%, and 8% and 10%
- Metallography to determine steam-oxide, alpha, and beta layers; weight gain; and effects of corrosion layer on weight-gain kinetics Pioneering Nuclear Science and Regulatory Technology Commission
(0604 and 1204 Test Trains) 16 As-received: ~ 10 wppm Prehydrided: 300-800 wppm As-received: in-cell tests Measured Weight Gain, mg/cm2 12 8
4 0
0 4 8 12 16 2
CP-Predicted Weight Gain, mg/cm Pioneering Nuclear Science and Regulatory Technology Commission
Nonirradiated 15x15 Zry-4 Thermal History of In-Cell Oxidation Tests with New Test Train (12/29/2004) 1400 7% CP-ECR 3% CP-ECR T = 1204ºC T = 1140ºC 1200 1000 5% CP-ECR T = 1185ºC 8% CP-ECR 10% CP-ECR Temperature, °C T =1206 °C T = 1206ºC 800 600 400 TC on specimen at 0º 200 TC5 on specimen at 120º 0
400 500 600 700 800 900 1000 Time, s Pioneering Nuclear Science and Regulatory Technology Commission
3% CP-ECR Sample
- 3% CP-ECR with T = 1140ºC Prior to Cooling
- Post-test hydrogen: 670+/-210 wppm (2-mm rings sides of sample)
- Metallography Outer-surface: most of corrosion layer is missing; no evidence of steam-oxide layer O-diffusion: corrosion layer O-stabilized alpha alpha layer CP model thickness Inner-surface: partial fuel-cladding bond layer non-uniform steam-oxide layer??
O-stabilized alpha layer < CP model Summary: effects of corrosion layer ECRt < CP-ECR
- Ring-compression ductility > 45% (test stopped before failure)
- Assessment: very high ductility; comparable to as-fabricated Zry-4
- No intrinsic hydrogen embrittlement observed Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 3% CP-ECR Sample - Outer Surface Steam-Oxidation Layer???
Corrosion Layer Alpha Layer Prior-Beta Layer Pioneering Nuclear Science and Regulatory Technology Commission
(After Grinding off 2mm, Polishing and Etching)
Oxygen-Stabilized Alpha Layer Corrosion Layer No Steam Oxide Layer Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 3% CP-ECR Sample - Inner Surface Steam Oxide Layer & Fuel- Alpha Cladding Bond Layer Prior-Beta Layer Pioneering Nuclear Science and Regulatory Technology Commission
(After Grinding off 2mm, Polishing and Etching)
Alpha Layer Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 670+/-210 wppm H (Post-Oxidation) 0.7 HBR Rod F07 Cladding Oxidized to 3% ECR at Tmax = 1140°C
~ 70-µm Corrosion Layer; 670+/-210 wppm H Ring Compression Test at 135°C 0.6 0.5 0.4 Load (kN) 0.3 0.2 0.1 45% Offset 0
0 1 2 3 4 5 6 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
5% CP-ECR Sample
- 5% CP-ECR with T = 1185ºC Prior to Cooling
- Post-test hydrogen: 600+/-90 wppm (2-mm rings sides of sample)
Metallography Outer-surface: partial corrosion layer, uneven steam-oxide layer growth of uniform O-stabilized alpha layer alpha layer < CP model, but rate may be similar Inner-surface: growth of steam-oxide layer; some fuel-clad. bond growth of alpha layer ( same as outer surface)
Summary: effects of corrosion layer ECRt < CP-ECR
- Ring #1: test stopped after 40% load drop at 12% offset strain no through-wall crack seen at 3X mag.; need metallography
- Ring #2: test run to max. displacement; no sharp load drops; through-wall crack at bottom; ductility is 40%
- Assessment: medium-to-high ductility Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 5% CP-ECR Sample Inner Surface Steam Oxide Prior-Beta Alpha Outer Prior-Beta Surface Corrosion Steam Oxide??
Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 5% CP-ECR Sample (Contd)
Corrosion Steam Oxide Alpha Steam Oxide Fuel-Cladding Bond Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 600+/-90 wppm H (Post-Oxidation) 0.6 HBR Rod F07 Cladding Oxidized to 5% ECR at Tmax = 1185°C
~ 70-µm Corrosion Layer; 600+/-90 wppm H Ring Compression Test #1 at 135°C 0.5 0.4 Load (kN) 0.3 0.2 0.1 12% Offset 0
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 600+/-90 wppm H (Post-Oxidation) 0.6 HBR Rod F07 Cladding Oxidized to 5% EC R at Tm ax = 1185°C
~ 70-µm Corrosion Layer; 600+/-90 wppm H Ring Compression Test #2 at 135°C 0.5 0.4 Load (kN) 0.3 0.2 0.1 18% Offset 37% Offset 52% Offset 0
0 1 2 3 4 5 6 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 600+/-90 wppm H (Post-Oxidation) 0.6 HBR Rod F07 Cladding Oxidized to 5% CP-ECR at Tmax = 1185°C
~ 70-µm Corrosion Layer; 600+/-90 wppm H Ring Compression Tests at 135°C 0.5 HBR_5%CP-ECR_Ring1 HBR_5%CP-ECR_Ring2 0.4 Normalized Load (kN) 0.3 0.2 0.1 0
0 1 2 3 4 5 6 Displacement (mm)
Pioneering Nuclear Science and Regulatory Technology Commission
7% CP-ECR Sample
- 7% CP-ECR with T = 1204ºC Prior to Cooling
- Note: sample was heated to 900ºC, control TC failed, cooled to 300ºC, reheated to 1204ºC based on working TC
- Post-test hydrogen: 780+/-70 wppm (2-mm rings sides of sample) too high; probably not all in metal cladding
- Metallography Outer-surface: partial corrosion layer, growth of steam-oxide layer growth of uniform O-stabilized alpha layer Inner-surface: growth of steam-oxide layer; some fuel-clad. bond growth of O-stabilized alpha layer Summary: effects of corrosion layer ECRt < CP-ECR
- Ring-compression ductility = 3.8% offset strain, 2.6% permanent strain
- Assessment: low ductility Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 7% CP-ECR Sample - Outer Surface Corrosion Layer Steam Oxide Layer Prior-Beta Layer Alpha Layer Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 7% CP-ECR Sample - Inner Surface Steam Oxide Layer Prior-Beta Layer Alpha Layer Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 780+/-70 wppm H (Post-Oxidation) 0.6 HBR Rod F07 Cladding Oxidized to 7% CP-ECR at Tmax = 1204°C 70-µm Corrosion Layer; 780+/-70 w ppm H Ring Compression Tests at 135°C 0.5 0.4 Load (kN) 0.3 0.2 0.1 3.6% Offset 0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Displacem ent (m m )
Pioneering Nuclear Science and Regulatory Technology Commission
10% CP-ECR Sample
- 10% CP-ECR with T = 1206ºC Prior to Cooling
- Post-test hydrogen: 970+/-150 wppm (2-mm rings sides of sample) 750+/-100 wppm based on 8-mm-long ring too high; may not all be in the metal
- Metallography Outer-surface: corrosion layer??, growth of steam-oxide layer steam-oxide < O-stabilized alpha layer thickness Inner-surface: growth of steam-oxide layer; fuel-clad. bond??
steam-oxide > O-stabilized alpha layer thickness Summary: alpha layer thickness < pre-H, non-irr. sample need better metallography measured ECR
- Ring-compression ductility = 0.5% offset strain, 0.6% permanent strain
- Assessment: very brittle Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 10% CP-ECR Sample - Outer Surface Steam Oxide Layer Corrosion Layer???
Alpha Prior-Beta Pioneering Nuclear Science and Regulatory Technology Commission
Metallography of 10% CP-ECR Sample - Inner Surface Prior-Beta Prior-Beta Alpha Steam Oxide Layer Fuel-Clad Bond???
Pioneering Nuclear Science and Regulatory Technology Commission
Outer Surface Corrosion Steam Oxide Pioneering Nuclear Science and Regulatory Technology Commission
Inner Surface Fuel-Cladding Steam Oxide Bond??
Pioneering Nuclear Science and Regulatory Technology Commission
Hydrogen Content of 10% CP-ECR Sample Top of the ring 641 wppm C
B D 672 wppm 876 wppm A E 839 wppm Crack area 799 wppm Average Hydrogen Content - 751 wppm Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 970+/-150 wppm H (Post-Oxidation) 0.45 HBR Rod F07 Cladding Oxidized to 10% CP-ECR at Tmax = 1206°C 70-µm Corrosion Layer; 970+/-150 w ppm H Ring Compression Tests at 135°C 0.4 0.35 0.3 0.25 Load (kN) 0.2 0.15 0.1 0.05 0.5% Offset 0
0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 Displacem ent (m m )
Pioneering Nuclear Science and Regulatory Technology Commission
in 10% ECR Sample Pioneering Nuclear Science and Regulatory Technology Commission
8% CP-ECR Sample
- 8% CP-ECR with T = 1206ºC Prior to Cooling
- Note: sample was defueled very thoroughly prior to testing
- Post-test hydrogen: 545+/-80 wppm (2-mm rings sides of sample)
- consistent with pre-oxidation H content
- Metallography (in progress with vacuum-impregnation of epoxy)
Outer-surface:
Inner-surface:
Summary:
- Ring-compression ductility = 3.8% offset strain, 2.9% permanent strain
- Assessment: low ductility Pioneering Nuclear Science and Regulatory Technology Commission
Pre-Test Visual Appearance of Sample Sample Prior to ntested Testing ample Pioneering Nuclear Science and Regulatory Technology Commission
Post-Test vs. Pre-Test Visual Appearance of Sample Sample After ntested Oxidation ample Pioneering Nuclear Science and Regulatory Technology Commission
HBR Zry-4 Ring with 545+/-80 wppm H (Post-Oxidation) 0.6 HBR Rod F07 Cladding Oxidized to 8% CP-ECR at Tmax = 1206°C 70-µm Corrosion Layer; 545+/-80 w ppm H Ring Compression Tests at 135°C 0.5 0.4 Load (kN) 0.3 0.2 0.1 3.8% Offset Strain (2.9% Permanent Strain) 0 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Displacem ent (m m )
Pioneering Nuclear Science and Regulatory Technology Commission
Preliminary Results for (OD+ID)/2 Alpha Layer 50 Irradiated HBR Pre-H 15x15 Zry-4 Measured Alpha Layer Thickness, µm 40 30 20 10 0
0 10 20 30 40 50 CP-Predicted Alpha Layer Thickness, µm Pioneering Nuclear Science and Regulatory Technology Commission
Preliminary Results for Outer-Surface Alpha Layer 50 Irradiated HBR Measured Outer-Surface Alpha Layer, m Pre-H 15x15 Zry-4 40 30 20 10 0
0 10 20 30 40 50 CP Predicted Outer-Surface Alpha Layer, m Pioneering Nuclear Science and Regulatory Technology Commission
135ºC vs. CP-ECR for High-Burnup Zry-4 Cladding 60 670 +/- 210 wppm H, Tmax = 1140°C 600 +/- 90 wppm H, Tmax = 1185°C 780 +/- 70 wppm H, Tmax = 1204°C 50 750 +/- 100 wppm H, Tmax = 1206°C 545 +/- 80 wppm H, Tmax = 1206°C 40 Offset Strain, %
30 20 10 0
0 2 4 6 8 10 12 Predicted (CP) ECR, %
Pioneering Nuclear Science and Regulatory Technology Commission
Discussion of Results for High-Burnup HBR Cladding
- Characterization of As-Irradiated Rod F07 Cladding
- Significant hydrogen variation for the 1st 5 defueled segments
- Expected H content & variation for the next 2 defueled segments
- Reasonable to assume ductility data correspond to 550+/-100 wppm H
- Effects of Corrosion Layer on Steam Oxidation Kinetics
- Corrosion layer is a source of oxygen for embrittlement and a partial barrier to steam oxidation, especially for short times at temperature
- Metallography of 3-10% CP-ECR samples suggests that corrosion layer acts as partial barrier in slowing down oxidation
- Better mounting of 8% CP-ECR sample is in progress
- Other data on effects of corrosion layer French data (1990s): PWR Zry-4 corrosion layer is non-protective ANL data: BWR Zry-2 (10 µm, 300 s at 1204ºC): non-protective JAERI data: PWR Zry-4 (20-25 µm): partially protective at 1000ºC Pioneering Nuclear Science and Regulatory Technology Commission
(Contd)
- Effects of Quench
- One sample will be oxidized to 8% CP-ECR & quenched (in progress)
- Effects of Temperature Ramp Rate on PQD
- Significant effect expected for low-ECR 2-sided oxidation tests
- Range of ramp rates reported in the literature 20ºC/s from 1000 to 1200ºC 1% ECR for 2-sided oxidation 0.5ºC/s from 1000 to 1200ºC 12% ECR for 2-sided oxidation
- ANL test times at T = 1180-1206ºC (severe embrittlement expected) 3% CP-ECR, t = 0 s; 5% CP-ECR, t = 5 s; 7% CP-ECR, t = 44 s 8% CP-ECR, t = 77 s; 10% CP-ECR, t = 118 s (-layer saturation)
- Effects of Corrosion Layer on Cladding Temperature
- Low heat of oxidation may result in lower ramp temperatures
- Adsorption/emissivity coefficients: corrosion < steam oxide layer
- Neither of these would effect the 1204ºC hold temperature Pioneering Nuclear Science and Regulatory Technology Commission
Oxidized at 1200ºC to 6%, 8% and 14% ECR 500 Zry-4 at 8% ECR Zry-4 at 14% ECR Zry-4 at 6% ECR 400 Hardness, DPH 300 200 layer at 6% ECR 100 layer at 8% ECR layer at 14% ECR 0
0 100 200 300 400 500 600 700 800 Radial Position (Outer-to-Inner), µm Pioneering Nuclear Science and Regulatory Technology Commission
Unirradiated Zry-2 Sample 1400 1200 Temperature (°C) 1000 800 600 Free Standing ID TC OD TC 0° 400 OD TC 120° 200 0 100 200 300 400 500 600 700 800 900 1000 Time (s)
Pioneering Nuclear Science and Regulatory Technology Commission
Future Plans
- Post-Quench Ductility of High-Burnup Zry-4 sided oxidation: effects of quench on 8% CP-ECR ductility (2/05) sided oxidation: Th = 1204°C (3/05), 1000°C & 1100°C (5/05)
Fast ramp to Th - 50°C, slower ramp to Th CP-ECR = 3, 5, 7.5, 10% at 1204°C; 5, 10, 13% for 1000 & 1100°C Run tests without quench at 800°C; repeat one test with quench Ring-compression tests at 135°C & 80-100°C
- LOCA integral tests (8/05) 1 test to burst with slow cooling balloon-burst characterization 3 tests at 3, 5, 7.5% CP-ECR in non-ballooned region (TBD) 4-point-bend test (RT), ring-compression tests at 135°C & 80-100°C Pioneering Nuclear Science and Regulatory Technology Commission
Future Plans (Contd)
- Post-Quench Ductility of High-Burnup ZIRLO
- Perform Zry-4 oxidation/quench/compression test matrix ZIRLO from North Anna Studsvik ANL 1204°C tests (8/05); 1000°C & 1100°C tests after 9/05 Baseline data on nonirradiated, prehydrided ZIRLO?????
- Perform LOCA integral test matrix (>9/05)
Fuel at 48 GWd/MTU (<62 GWd/MTU)
- Post-Quench Ductility of High-Burnup M5
- Perform Zry-4 oxidation/quench/compression test matrix European M5 Studsvik ANL 1204°C tests (8/05); 1000°C & 1100°C tests after 9/05 Baseline data on nonirradiated, prehydrided M5 desirable (135°C)
- Perform LOCA integral test matrix (>9/05)
M5 from North Anna (nee ANL-W) ANL-E Pioneering Nuclear Science and Regulatory Technology Commission