Characterization of Long-Term, In-Reactor Zircaloy-4 ......Characterization of Long-Term, In-Reactor Zircaloy-4 Corrosion Coupons and the Impact of Flux, Fluence, and Temperature on

Characterization of Long-Term, In-Reactor Zircaloy-4 Corrosion Coupons and the

Impact of Flux, Fluence, and Temperature on Oxide Growth, Stress Development,

Phase Formation, and Grain Size

Brendan Ensora, Gene Lucadamoa, John Seidenstickera, Ram Bajaja,

Zhonghou Caib, and Arthur Mottac

The Naval Nuclear Laboratory is operated for the U.S. Department of Energy by Fluor Marine Propulsion, LLC,

a wholly owned subsidiary of Fluor Corporation

aNaval Nuclear Laboratory, bArgonne National Laboratory, cThe Pennsylvania State University

19th International Symposium on Zirconium in the Nuclear Industry

Outline

2

• Introduction

• Effects of temperature, neutron flux, and neutron fluence on Zircaloy-4 corrosion, transition, and irradiated induced breakaway

• Results

• SEM, FIB-SEM serial sectioning, crack morphology, metal-oxide interface morphology

• Synchrotron µXRD experiments (grain size and phase fraction)

• Discussion

• Tetragonal phase stability

• Proposed theory on cause of irradiation induced breakaway/transition corrosion behavior

• Summary and Conclusions

Zirconium Alloy Corrosion• Zr alloys used as structural materials

• Good high temperature corrosion resistance and neutron transparent

• As alloys corrode:𝑍𝑟𝑚𝑒𝑡𝑎𝑙 + 2𝐻2𝑂𝑙𝑖𝑞𝑢𝑖𝑑 → 𝑍𝑟𝑂2 + 2𝐻2 𝑔𝑎𝑠

• Oxide layer insulates against heat transfer

• Hydrogen ingress leads to embrittlement

• Oxide forms long columnar monoclinic phase grains (some tetragonal phase)

• Transition: loss of protectiveness

• Cyclic process

3

0

20

40

60

80

100

120

140

0 100 200 300 400 500

We

igh

t G

ain

(m

g/d

m2)

Days @ 360°C in Autoclave

Transitions

Cyclic Process

Zircaloy-4, 360°C water

Oxide

Metal100032, Zircaloy-4

1804 days, 360°C, 54.9 µm

1B. Ensor, A. M. Lucente, M. J. Frederick, J. Sutliff and A. T. Motta, "The role of hydrogen in

zirconium alloy corrosion," Journal of Nuclear Materials, vol. 496, pp. 301-312, 2017.

Zirconium Alloy Corrosion - Irradiation

4

0

500

1000

1500

2000

2500

0 500 1000 1500 2000 2500 3000 3500 4000 4500W

eig

ht G

ain

(m

g/d

m2)

Exposure time (days)

Neutron Irradiated Zircaloy-4 Weight Gain (>10x1013 n/cm2/s)

AAHH1 C000PE176 H920J

310°C

275°C

353°C

1 B. F. Kammenzind, J. A. Gruber, R. Bajaj and J. D. Smee, "Neutron irradiation effects on the

corrosion of Zircaloy-4 in a PWR environment," 18th International Symposium on Zirconium in

the Nuclear Industry ASTM STP 1597, pp. 448-490, 2018

• Under irradiation, Zircaloy-4

undergoes an irradiation

induced corrosion

acceleration

• This varies as a function of:

• Neutron flux

• Temperature

• Generally cyclic corrosion

layers still observed

• A distinct rate change in life

appears to occur10 µm

0

500

1000

1500

2000

2500

0 500 1000 1500 2000 2500 3000 3500 4000 4500W

eig

ht G

ain

(m

g/d

m2)


Neutron Irradiated Zircaloy-4 Weight Gain (>10x1013 n/cm2/s)

AAHH1 H381J H920J

310°C

275°C353°C

Zirconium Alloy Corrosion - Irradiation

5




• Under irradiation, Zircaloy-4

undergoes an irradiation

induced corrosion

acceleration

• This varies as a function of:

• Neutron flux

• Temperature

• Generally cyclic corrosion

layers still observed

• A distinct rate change in life

appears to occur

Why?

10 µm

Samples

and

Electron Microscopy

6

Synchrotron µXRD Experiments – Irradiated Samples

7

• Eleven recrystallized alpha-annealed (RXA) Zircaloy-4

samples were corroded isothermally in the ATR over a

range of temperatures and fluxes then prepared in cross-

section and examined at the APS

Sample Alloy

Average

Corrosion

Temperature*

Avg. Flux

(1013n/cm2/s)

E > 1 MeV

Total Fluence

(1020n/cm2)

E > 1 MeV

Time

(days)

Oxide

Thickness&

H1169J RXA Zircaloy-4 272°C 1.94 24.8 1474 1.97 µm







H337J RXA Zircaloy-4 355°C 0 0 1139 25.6 µm




8


9

H920J, 353°C, 328 days, 11.48x1013 n/cm2/s, 21.8 µm H1240J, 337°C, 1189 days, 2.36x1013 n/cm2/s, 29.3 µm

H977J, 314°C, 1206 days, 4.23x1013 n/cm2/s, 25.1 µm

10 µm 10 µm

10 µm


H381J, 272°C, 2395 days, 10.49x1013 n/cm2/s, 15.8 µm

10 µm

All Zircaloy-4

10


11

H1490J, 274°C, 1138 days, 8.29x1013 n/cm2/s, 1.60 µm,12.5 dpa

• Needles of oxide observed at the metal-oxide interface• Seen elsewhere…(Zircaloy-2 BWR 9-cycle1, Baris et al, Journal of

Nuclear Materials, 2018)

• Perhaps a precursor to unstable oxide growth?• Similar morphology seen in Crystal Bar Zr in air (600°C) and autoclave

(360°C) prior to breakaway oxidation2

1A. Baris, R. Restani, R. Grabherr, Y.-L. Chiu, H. E. Evans, K. Ammon, M. Limback and S.

Abolhassani, "Chemical and Microstructural Characterization of a 9 cycle Zircaloy-2 cladding

using EPMA and FIB Tomography," Journal of Nuclear Materials, vol. 504, pp. 144-160, 2018. 2B. Ensor, "The Nature of Unstable Oxide Growth in Zirconium and Zirconium Alloys," PhD

Thesis in Nuclear Engineering: The Pennsylvania State University, 2016.

FIB/SEM Characterization – Irradiated Samples

12


13

Figure 1- Left, reconstructed FIB serial section acquired from H337J (no flux, 1139 days, 355°C, 25.6 µm) and Right, H421J

(1.37x1013 n/cm2/s, 537 days, 355°C, 25.7 µm). The capping layer is red, the cracks are dark blue, and the Zircaloy-4 is rendered

in green. The oxide is semi-transparent.

5 mm 5 mm

• Three thick oxide film Zircaloy-4 samples were examined using

FIB serial sectioning

All images on

same scale

H421J, 355°C, 573 days,

1.37x1013 n/cm2/s, 25.7 µm H337J, 355°C, 1139 days,

0x1013 n/cm2/s, 25.6 µm

Flux

H381J, 272°C, 2395 days,

10.49x1013 n/cm2/s, 15.8 µm


14

Figure 1- Left, reconstructed FIB serial section acquired from H337J (no flux, 1139 days, 355°C, 25.6 µm) and Right, H421J

(1.37x1013 n/cm2/s, 537 days, 355°C, 25.7 µm). The capping layer is red, the cracks are dark blue, and the Zircaloy-4 is rendered

in green. The oxide is semi-transparent.

5 mm 5 mm

• Three thick oxide film Zircaloy-4 samples were examined using

FIB serial sectioning

All images on

same scale

H421J, 355°C, 573 days,

1.37x1013 n/cm2/s, 25.7 µm H337J, 355°C, 1139 days,

0x1013 n/cm2/s, 25.6 µm

Flux

H381J, 272°C, 2395 days,

10.49x1013 n/cm2/s, 15.8 µm

No crack spacing difference,

yet 2x the corrosion rate in-

flux, suggests in-cycle effect

Initial layer spacing similar, but

different crack morphology

observed after


Technique

15

16

• Motivation: Answer questions about oxide growth to inform use of better alloys and models

• Goal: To better understand mechanistically the corrosion behavior of zirconium alloys using µXRD at the APS

• High brilliance enables small beam size

• Allows for location specific examination on sub-micron level

• Ability to vary and tune energy permits variety of experiments

Advanced Photon Source (APS)

2-ID-D Beamline

Microbeam (0.2 µm x 0.2 µm)

Simultaneous XRD and XRF, XANES

Variable energy (5-32 keV)

Synchrotron µXRD Experiments - Beamline Data

7

H381J

272°C, 2395 days, 10.5 x 1013 n/cm2/s

Metal

Planes // to

Metal-oxide

interface

Planes ┴ to

Metal-oxide

interface

(10-10)α-Zr

(10-11)α-Zr

(10-12)α-Zr

Chi (χ)

Two-theta (2θ)

Oxide layer

Artifact

H381J

272°C, 2395 days, 10.5 x 1013 n/cm2/s

Oxide

Metal

0

500

1000

1500

2000

2500

3000

-548 -546 -544 -542 -540 -538

Inte

nsi

ty

x (µm)

Zirconium X-Ray Fluorescence Data H1187R xscan3

Zirconium

Epoxy Oxide Metal

• Garvie-Nicholson Tetragonal Phase Fraction

𝑇𝑒𝑡 𝐹𝑟𝑎𝑐, 𝑓𝑇 =𝐼 101 𝑡

𝐼 101 𝑡 + 𝐼 ത111 𝑚 + 𝐼 111 𝑚

• Debye-Scherrer

𝑑 =0.9𝜆

𝐵𝛳𝑐𝑜𝑠𝛳where 𝐵𝛳 = (𝐵2 − 𝐵𝑖

2)1/2

Synchrotron µXRD Experiments – Analysis Methods

8

Instrumental BroadeningX-ray Wavelength

Phase Fraction

Grain Size

Diffraction Patterns

Corre

late

d to

Tem

pera

ture

, Flu

x,

Flu

ence, a

nd O

xid

e T

hic

kness

ResultsGrain Size

19

Results – Grain Size (Autoclave)

20

0

5

10

15

20

25

30

35

40

270 280 290 300 310 320 330 340 350 360 370 380 390 400 410 420 430

Gra

in S

ize

Dia

met

er (

nm

)

Corrosion Temperature (°C)

Grain Diameter as a Function of Temperature (Outer Oxide, Zircaloy-4)

Monoclinic Phase Oxide

Tetragonal Phase Oxide

- From previous work1

- Increasing temperature leads to increasing monoclinic grain size

- Temperature has no effect on tetragonal grain size- Indicative of a critical size, above which tetragonal transforms to monoclinic

1B. Ensor, D. J. Spengler, J. R. Seidensticker, R. Bajaj, Z. Cai and A. T. Motta, "Microbeam

synchrotron radiation diffraction and fluorescence of oxide layers formed on zirconium alloys

at different corrosion temperatures," Journal of Nuclear Materials, vol. TBD, p. TBD, 2019.


21

0

5

10

15

20

25

30

35

40

45

3.53.02.52.01.51.00.50.0

Gra

in S

ize

Dia

met

er (

nm

)

Distance from Metal-Oxide Interface (µm)

H1187R, H2287J, H1190R, and H1406J Grain Size Diameter

360°C

274°C

Monoclinic Grains

Tetragonal Grains

- From previous work1

- Smaller monoclinic grains at metal-oxide interface (MOI), larger away

- Larger tetragonal grains at MOI, smaller away- Indicative of stress stabilization of the tetragonal phase at the MOI

1B. Ensor, D. J. Spengler, J. R. Seidensticker, R. Bajaj, Z. Cai and A. T. Motta, "Microbeam

synchrotron radiation diffraction and fluorescence of oxide layers formed on zirconium alloys at

different corrosion temperatures," Journal of Nuclear Materials, vol. TBD, p. TBD, 2019.

Results – Grain Size (ATR)

22

271-274°C

352-355°C

352-355°C

271-274°C

- Still no change in

tetragonal phase grain size

- Hard to see temperature

effect

- Instead, much larger

neutron flux/fluence effect

observed

- Increasing the neutron flux

or fluence leads to larger

monoclinic grain size

0

5

10

15

20

25

30

35

40

45

50

H381J

(272°C)

H977J

(314°C)

H1240J

(337°C)

H944J

(339°C)

H337J

(355°C)

H421J

(355°C)

H657J

(352°C)

H920J

(353°C)

Gra

in s

ize

(nm

)

Average Tet d Average 111m d

Results – Grain Size (ATR)

23

- Monoclinic and

tetragonal grain size

evolution through oxide

consistent with

autoclave observations

- Same stable (~15 nm)

tetragonal grain size as

autoclave

- Noticeable effect of

neutron flux or fluence

on monoclinic grain

size

- No observed effect on

tetragonal grain size

ResultsPhase Composition

24

Results - Phase Composition

25

- Highest fT at MOI with fTdecreasing away from

MOI- As observed previously1

- No effect of temperature1

- Observable effect of

neutron flux and fluence- Higher fT at MOI

- Lower fT in bulk

- Decrease away from

MOI consistent with

grain growth

- fT at MOI thought to be

related to metal stress

accommodation

(Irradiation hardening?)

R² = 0.3951

R² = 0.51490

0.05

0.1

0.15

0.2

0.25

0.3

0 10 20 30 40 50 60 70 80 90

Tet

ragon

al F

ract

ion

, f T

Fluence (1020 n/cm2; >1 MeV)

Tetragonal Fraction (Bulk and Max) vs Fluence

Max

Bulk

1B. Ensor, D. J. Spengler, J. R. Seidensticker, R. Bajaj, Z. Cai and A. T. Motta, "Microbeam synchrotron radiation diffraction and fluorescence of

oxide layers formed on zirconium alloys at different corrosion temperatures," Journal of Nuclear Materials, vol. TBD, p. TBD, 2019.

Grain growth

Stress at MOI

Results – Phase Composition

26

1 B. Ensor, D. J. Spengler, J. R. Seidensticker, R. Bajaj, Z. Cai and A. T. Motta,

"Microbeam synchrotron radiation diffraction and fluorescence of oxide layers formed

on zirconium alloys at different corrosion temperatures," Journal of Nuclear Materials,

vol. TBD, p. TBD, 2019.

- Two key factors: grain

size and stress

- Previous results have

linked the tetragonal

phase fraction1 to stress

in the oxide layer2

- Macroscopic

compressive stress

influences the formation

of fT

- Resulting t → m and

cracks influence

observations

2 H. Swan, M. S. Blackmur, J. M. Hyde, A. Laferrere, S. R. Ortner, P. D. Styman, C. Staines, M. Gass, H. Hulme, A. Cole-Baker

and P. Frankel, "The measurement of stress and phase fraction distributions in pre and post-transition Zircaloy oxides using

nano-beam synchrotron X-ray diffraction," Journal of Nuclear Materials, vol. 479, pp. 559-575, 2016.

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14

0.16

0.18

0 1 2 3

Tet

ragon

al P

has

e F

ract

ion

, f

T

Distance from Oxide-Water Interface (µm)

Zircaloy-4 Average Tetragonal Phase Fraction

Average of all

Samples

Average of all

360°C Samples

Average of

H1406J (360°C)

Peak at metal-

oxide interface

DiscussionTheory

27

Stress and Thermodynamics

28

∆𝐺𝑡→𝑚 = ∆𝐺𝑐ℎ𝑒𝑚𝑖𝑐𝑎𝑙𝑡→𝑚 + ∆𝑊𝑚𝑎𝑡𝑟𝑖𝑥

𝑡→𝑚 + 𝑆𝑡𝑤𝑖𝑛𝑡→𝑚 + 𝜎𝑚𝑎𝑐𝑟𝑜𝑁𝐴𝜐

𝑡𝜅0 +𝑁𝐴𝜂(𝜐

𝑡)1/3[ 𝜐𝑚23𝛿𝑚 − 𝜐𝑡

23𝛿𝑡]

(𝑉𝑡)1/3

1 Qin, W., C. Nam, H.L. Li, and J.A. Szpunar. "Tetragonal phase stability in ZrO2 film formed on

zirconium alloys and its effects on corrosion resistance." Acta Materialia 55 (2007): 1695-1701.

Free energy to

change from the

tetragonal to the

monoclinic

phase

Chemical free

energy

driving force

Matrix

strain

Work against the

macroscopic

compressive

stress

Tetragonal phase

grain size term

Alloying elements Stress Grain Size

No effect seen

at MOI

• Why higher tetragonal fraction at MOI?

• Thermodynamics1 of t → m

Redistribution2 of

Fe…but so much

more Sn in matrix

Irradiation

hardening of

matrix

Twin

boundary

energy term

2 B. Ensor, "The Nature of Unstable Oxide Growth in Zirconium and Zirconium Alloys," PhD Thesis in

Nuclear Engineering: The Pennsylvania State University, 2016.


29

• Why higher tetragonal fraction at MOI?

• Irradiation hardening• Metal matrix less able to accommodate oxide growth stresses, leading to

higher stresses at MOI and higher fT• Levels off around 4-5 dpa

• What about irradiation induced creep?• Irradiation induced creep would relieve σ, thus leading to lower fT• Larger enough to offset hardening?

• All material here is recrystallized alpha-annealed Zircaloy-4

• Cold-worked/stress-relief annealed Zircaloy-4 creeps more in-flux

• So would expect less fT in irradiated CW/SRA Zircaloy-4

Theory

30

- A proposed model includes the basics of protective oxide layer growth and is

consistent with the findings from these studies

- The oxide growth can be split into four distinct phases:

(1) Oxide nucleation from the metal grains

(2) Growth of protective oxide layer (i.e., barrier layer)

(3) Breakdown of protective oxide during transition and formation of a new barrier

layer

(4) Irradiation induced transition (distinct from the cyclic transition process

described in steps 1-3) to post-transition oxide growth

0

50

100

150

200

250

300

350

400

450

0 500 1000 1500 2000 2500

Weig

ht G

ain

(m

g/d

m2)


Neutron Irradiated Zircaloy-4 Weight Gain

0

20

40

60

80

100

120

140

0 100 200 300 400 500

We

igh

t G

ain

(m

g/d

m2 )

Days @ 360°C in Autoclave

Transitions

Cyclic Process

Zircaloy-4, 360°C water

43

12

3

12

3

12

Discussion – Cyclic Corrosion Regimes

31

(1) Oxide nucleates as small, equiaxed grains, postulated to be driven partly by matching to the

metal lattice; higher corrosion rate• Equiaxed oxide grains, higher fT, thinner suboxide layer, and more uniform grain size (larger tetragonal

phase grains and smaller monoclinic phase grains) as compared to oxide grown in the next phase

(2) Favorably-oriented monoclinic grains grow into larger, columnar oxide grains, driven by the

need to reduce stress accumulation, slower corrosion rate• Lower fT (increases as the oxide approaches transition), a significant suboxide layer (the size increases

as transition is approached and oxide growth slows), and a divergence in the grain size of the

monoclinic and tetragonal phases (monoclinic grains become larger, while tetragonal grains transform to

monoclinic phase above a critical size)

(3) Oxide transition; the protective nature of the oxide layer is lost due to a breakdown in the oxide• Leads to easy access of the metal to the oxidizing species (made possible by a percolation of porosity)

• Occurrence of the transition is characteristic of an alloy and oxidation conditions

• Once it occurs, the process begins again at (1).

➢Phases (1) and (2) governed by the alloying elements and precipitates➢Efforts to model this behavior based on the potential across the oxide and space charge compensation1

➢Phase (3), oxide transition, is governed by stress accommodation and a critical stress of the

protective oxide layer

1A. Couet, A. Motta and A. Ambard, "The coupled charge compensation model for zirconium

alloy fuel cladding oxidation, I. Parabolic oxidation of zirconium alloys," Corrosion Science, vol.

100, pp. 73-84, 2015

Discussion – Irradiation Induced Transition

32

(4) An irradiation induced transition to enhanced corrosion rates

• Described well by Kammenzind et al. for Zircaloy-41

• Appears to be related to a synergistic effect between thick oxide films and irradiation effects

on SPPs/alloying element redistribution• Heterogeneous radiolysis could not cause a transition prior to the existence of the thick film →

magnitude of the acceleration once in the post-transition regime

• Led to 30-40x accelerated corrosion rates in post-transition at temperatures similar to those

examined above

• Exposure to various levels of neutron fluence did accelerate the time at which the material

transitioned from pre-transition (1)-(3) to post-transition (4)• SPPs/alloying element could also effect the onset to (4)

➢ The two corrosion regimes (pre-transition (1)-(3) and post-transition (4)) and the time at which

the corrosion transitions from one to the other are the key descriptors of in-reactor Zircaloy-4

corrosion

• Note distinction between ‘irradiation induced transition (4)’ and the ‘transition (3)’ ascribed to

cyclic behavior often observed out-of-reactor




Discussion – Irradiation Induced Transition Cause

33

➢ It is hypothesized that the oxide growth induced stress accommodation

of the Zircaloy-4 metal while irradiated could be the cause of the

transition from the pre-transition regime (1)-(3) to the post-transition

regime (4)

• Further testing is needed to understand and describe this mechanism

in detail and how other alloy systems (such as Zr-Nb) show

resistance to this enhanced corrosion regime

➢ Additional accelerating factors, such as acceleration due to hydrogen

concentration1, can also impact the corrosion rates observed in all

regimes (1)-(4)

1B. Ensor, A. M. Lucente, M. J. Frederick, J. Sutliff and A. T. Motta, "The role of hydrogen in

zirconium alloy corrosion," Journal of Nuclear Materials, vol. 496, pp. 301-312, 2017.

Ability of Metal to

accommodate oxide

growth stresses

Effects on

Precipitates

Creep

properties

Irradiation

Hardening

Irradiation

Induced

Transition

Else?

Summary of Observations

34

Characterization

- Irradiated Zircaloy-4 samples had similar oxide peaks, intensities, and variations as the samples

that were corroded in autoclave, suggesting similar corrosion mechanisms

- Two samples exposed at the same temperature to the same oxide thickness (355°C, ~26 µm),

but at different neutron flux levels (0 vs. 1.37x1013 n/cm2/s) had similar oxide layer spacing and

crack morphology despite the 2x corrosion rate of the in-flux sample; - Suggests that the in-cycle corrosion kinetics, rather than the transition oxide layer spacing, was most

influenced by the irradiation at this temperature and low total fluence (~1 dpa)

- One sample corroded at a low temperature (272°C) had accelerated corrosion after ~26 dpa,

which was correlated with a decrease in crack spacing, and change in crack morphology- Needles observed in sample close to entering similar regime (~12.5 dpa)

Grain Size

- Monoclinic grain size is larger for higher neutron fluxes and fluences, growing continually away

from the metal-oxide interface

- Average tetragonal grain size constant as a function of temperature and neutron flux/fluence;

above ~15 nm, the tetragonal phase is not stable and transforms to the monoclinic - Supported by a decrease in fT in the bulk of the oxide as a function of neutron fluence

Tetragonal Phase Fraction, fT- Maximum fT occurs near the MOI and increases as a function of neutron fluence

- Saturates around ~4-8 dpa, and likely occurs because of irradiation hardening limiting the

amount of stress the metal can accommodate, leading to higher stress accumulation at the MOI

Acknowledgements

35

- Ashley Lucente and Robert Etien III for their help coordinating the

sample preparation.

- Arash Parsi and Robert Rees for their help in sample preparation for

APS, acquiring the FIB serial section images, SEM imaging, and

conventional XRD measurements at Westinghouse, Churchill.

- This research has been authored by the Naval Nuclear Laboratory

under Contract No. DOE-89233018CNR000004 with the U.S.

Department of Energy.

- This research used resources of the Advanced Photon Source, a

U.S. Department of Energy (DOE) Office of Science User Facility

operated for the DOE Office of Science by Argonne National

Laboratory under Contract No. DE-AC02-06CH11357.

- This research was performed (B. Ensor) under appointment to

the Rickover Fellowship Program in Nuclear Engineering sponsored

by Naval Reactors Division of the U.S. Department of Energy.

Thank you for your attention

Questions?

36

37

EXTRA SLIDES

Synchrotron µXRD Experiments - Beamline Data

14

H657J

352°C, 230 days,

8.1 x 1013 n/cm2/s

9.50 µm

Metal-oxide

Interface

(-111)m

(111)m

(101)T

(002)T

(020)m

(002)m

(0002)α-Zr

{110}m

(002)m

(200)m

(020)m

Planes // to

Metal-oxide

interface

Planes ┴ to

Metal-oxide

interface

(002)t in

irrad too

{200}m chi

location

H977J

314°C, 1206 days,

4.23 x 1013 n/cm2/s

25.1 µm


39

• All material here is RXA Zircaloy-4

• Cold-worked/SRA Zircaloy-4 has stronger substrate out-of-flux…

• …creeps more in-flux

• So would expect less fT in CW/SRA Zircaloy-4

• …but less fT in autoclave tests too?

1B. Ensor, "The Nature of Unstable Oxide Growth in Zirconium and Zirconium Alloys," PhD Thesis in Nuclear

Engineering: The Pennsylvania State University, 2016.

Synchrotron µXRD Experiments – Autoclave Samples• Tested 11 Zircaloy-4 (some β-quenched) samples corroded at a variety of

temperatures, all prepared in cross-section

*H1190R, H1312J, and H1313J were corroded for 3 days at 360°C prior to listed exposure, <0.8 µm of the

oxide for all

**Oxide thickness is average thickness based on the weight gain of the sample

Exposure Condition* Time (days) Oxide thickness (μm)**

274°C water 3003 2.05, 2.35

316°C water 3113 13.1

360°C water 20 1.14

360°C water 120 2.09

360°C water 170 3.03, 4.02

400°C steam 41, 259 3.5, 9.1

413, 427°C steam 200, 203 7.8, 13.6

0

3

6

9

12

15

0

45

90

135

180

225

0 500 1000 1500 2000 2500 3000

Oxid

e T

hic

kn

ess

(µm

)

Wei

gh

t G

ain

(m

g/d

m2)

Time (days)

H2287J

H1187R

H1406J

H1312J

H1313J

H1405J

N2513

H1372J

H3017J

H3035J

H1190R

400-413°C

360°C

316°C

274°C

●: Zircaloy-4

427°C

15

B. Ensor, D. J. Spengler, J. R. Seidensticker, R. Bajaj, Z. Cai and A. T. Motta, "Microbeam



Results – Grain Size (ATR and Autoclave)

41

20

25

30

35

40

45

50

270 290 310 330 350 370 390 410 430

Gra

in S

ize

(nm

)

Corrosion Temperature (°C)

Grain Size Data

ATR data Autoclave data Autoclave Fit




42


0

5

10

15

20

25

30

35

40

0 2 4 6 8 10 12 14 16 18 20

Gra

in S

ize

Dia

met

er (

nm

)

Average Oxide Thickness (µm)

Monoclinic Grains

Tetragonal Grains

427ºC

413ºC

400ºC

360°C

316ºC

274°C




Results - Phase Composition (Autoclave)

43

0

0.05

0.1

0.15

0.2

0 0.5 1 1.5 2 2.5 3 3.5 4 4.5

Tet

ragon

al F

ract

ion

Distance from the Metal-Oxide Interface (µm)

Average Tetragonal Fraction for Different Corrosion Temperatures

427ºC

413ºC

400ºC

360°C

316ºC

274°C

All

0.04

0.06

0.08

0.1

0.12

0.14

0 0.5 1 1.5 2 2.5 3

Tet

ragon

al P

has

e F

ract

ion

Distance from the Oxide-Water Interface (µm)

Towards metal-oxide interface

(oxide growth direction)

Oxide first formed Approximate

transition location

- Highest fT at MOI- Around 15%

- Decreases away from MOI- Around 7.5%

- No effect of temperature- Creep could have made fT

less for higher temp

- Higher corrosion rate

previously linked to higher fT- Competing?

- Plot as function of O/W

interface to see change

within transition layer

- Noticeable periodicity

(“dinosaur bones” theory)




Results - Phase Composition (ATR)

44

0.0E+00

5.0E+05

1.0E+06

1.5E+06

2.0E+06

2.5E+06

26 27 28 29 30 31 32 33 34 35 36 37 38 39

Inte

nsi

ty (

cps)

Position (°2θ)

H381J 1° i.a.

H381J 2.5° i.a.

H920J 1° i.a.

H920J 2.5° i.a.

(-111)m

(101)T

(111)m

272°C

353°C

Results - Phase Map (Autoclave & ATR)

45

(a) (-111)m Diffraction Intensity

(d) Tetragonal Fraction, fT

(b) (111)m Diffraction Intensity

(c) (101)T Diffraction Intensity

Metal-oxide interface

Oxide-water interface

Oxide

Metal





Oxide

Metal

Metal-oxide

interface

Oxide-water

interface - Created µXRD/XRF

maps on two samples

(one autoclave, one

irradiated)

- Shows typical metal-

oxide interface

undulations

- Can create oxide peak

intensity maps

- And grain size,

tetragonal fraction, etc.

- Highlights variability,

reproducibility, and

consistency of data

locally




Results - Phase Map (Autoclave & ATR)

46

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0 1 2 3 4

Tet

ragon

al F

ract

ion

, f T

, at

Met

al-O

xid

e In

terf

ace

Oxide Layer Thickness (µm)

Irradiated

H1490J

8.29x1013 n/cm2/s, 1138 days

274°C, 1.60 µm0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0 1 2 3 4

Tet

ragon

al F

ract

ion

, f T

, at

Met

al-O

xid

e In

terf

ace

Oxide Layer Thickness (µm)

Non-Irradiated

H1187R

Autoclave, 170days

360°C, 3.03 µm

0

0.05

0.1

0.15

0.2

0.25

0 0.5 1 1.5 2 2.5 3 3.5

Tet

rago

nal

Ph

ase

Fra

ctio

n


H1187R Map Average Tetragonal Phase Fraction

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0 0.5 1 1.5 2 2.5

Tet

rago

nal

Fra

ctio

n, fT


H1490J Map Tetragonal Fraction• Irradiated vs. Autoclave comparison

1B. Ensor, "The Nature of Unstable Oxide Growth in Zirconium and Zirconium Alloys,"

PhD Thesis in Nuclear Engineering: The Pennsylvania State University, 2016.

Results – Previously published by Kammenzind et al

47




Precipitate Oxidation

(Presented at NuMat 2018, October 2018)

48

49

µXANES Results - Fe in Zircaloy-4 - Autoclave

0%

25%

50%

75%

100%

-2 -1 0 1 2 3 4 5

Per

cen

tage

Met

allic

, No

rmal

ized

Distance to the Metal-Oxide Interface (mils)

274°C316°C360°C

Metal Oxide

- Zr oxidizes preferentially

to Fe

- Overall similar behavior

of Fe metallic fraction

when corroded in water

- Difference between 274-

360°C water and 400-

425°C steam, with

plateau of unoxidized Fe

- Attributed to increased

corrosion kinetics in

steam at higher

temperatures0%

25%

50%

75%

100%

-2 -1 0 1 2 3 4 5 6 7 8

Per

cen

tage

Met

allic

, No

rmal

ized

Distance to Metal-Oxide Interface (µm)

Metal Oxide

316°C 400°C413°C 425°C

2B. Ensor, A. T. Motta, R. Bajaj, J. R. Seidensticker, Z. Cai, "XANES analysis of iron in Zircaloy-4 oxides formed at different temperatures

studied with microbeam synchrotron radiation," ANS LWR Fuel Performance Meeting, TopFuel 2015, A0151 (2015).



Synchrotron µXRD Experiments – Autoclave Samples

50

Sample AlloyExposure

Condition

Time

(days)

Oxide Thickness

(μm)

C1757P Zr-2.5Nb360°C water

680°F20 1.55

C1758P Zr-2.5Nb360°C water

680°F170 3.12

C5031P Zr-2.5Nb454°C steam

850°F200 39.9

• Three Zr-2.5Nb samples were corroded in autoclave with pure

water or steam for up to 200 days at 360°C and 454°C

B. Ensor, M. Moorehead, J. R. Seidensticker, R. Bajaj, A. Couet, and A. T. Motta, “XANES

Study of Fe and Nb Oxidation in Zr-2.5Nb Oxide Layers," Transactions of the American

Nuclear Society, vol. 117, p. 527-530, 2017.

51

µXANES Results – Fe in Zircaloy-4 and Zr-2.5Nb

0%

25%

50%

75%

100%

-1 0 1 2 3

Pe

rce

nt

of

Fe M

etal

lic, N

orm

aliz

ed

Distance to the Metal-Oxide Interface (µm)

Metal Oxide

Zr-2.5Nb

Zircaloy-4

360°C

• Fe oxidizes closer to the Metal-Oxide Interface in Zr-2.5Nb than Zircaloy-41

• Smaller precipitates (surface area) and more Fe in solid solution

1B. Ensor, A. T. Motta, R. Bajaj, J. R. Seidensticker, Z. Cai, "XANES analysis of iron in

Zircaloy-4 oxides formed at different temperatures studied with microbeam synchrotron

radiation," ANS LWR Fuel Performance Meeting, TopFuel 2015, A0151 (2015).

B. Ensor, M. Moorehead, J. R. Seidensticker, R. Bajaj, A. Couet, and A. T. Motta, “XANES Study of Fe and Nb

Oxidation in Zr-2.5Nb Oxide Layers," Transactions of the American Nuclear Society, vol. 117, p. 527-530, 2017.

52

µXANES Results – Fe & Nb in Zr-2.5Nb

0%

25%

50%

75%

100%

-1 0 1 2 3 4 5 6 7

Perc

ent

of

Fe o

r N

b M

etal

lic, N

orm

aliz

ed

Distance to Metal-Oxide Interface (µm)

360°C450°C

● NbΔ Fe

Metal Oxide

• Fe and Nb oxidize at similar distances to the Metal-Oxide Interface in Zr-2.5Nb

• Difference seen in water (360°C) vs. steam (454°C), more “delayed”

• Consistent with studies of ZIRLO®1

• Space charge (Fe3+ vs. Nb5+)

1A. Couet, A. T. Motta, B. de Gabory, Z. Cai, "Microbeam X-ray Absorption Near-

Edge Spectroscopy study of the oxidation of Fe and Nb in zirconium alloy oxide

layers," J. Nucl. Mater, 452, 614 (2014).

B. Ensor, M. Moorehead, J. R. Seidensticker, R. Bajaj, A. Couet, and A. T. Motta, “XANES Study of Fe and Nb

Oxidation in Zr-2.5Nb Oxide Layers," Transactions of the American Nuclear Society, vol. 117, p. 527-530, 2017.

53

µXANES Results – Effect of Irradiation

0

0.25

0.5

0.75

1

-2 -1 0 1 2 3 4 5 6 7 8 9 10

Fra

ctio

n o

f M

eta

llic F

e


Non-irradiated

Zircaloy-4

(H337J)

Irradiated

Zircaloy-4

0

0.25

0.5

0.75

1

-1 0 1 2 3 4 5

Fra

ctio

n o

f M

eta

llic F

e


352-355°C

Fe in non-irradiated sample is

oxidized further into the oxide

than Fe in the irradiated samples

0

0.25

0.5

0.75

1

-1 0 1 2 3 4 5

Fra

ctio

n o

f M

eta

llic F

e

Distance to the Metal-Oxide Interface (µm)

274°C

316°C

360°C

Autoclave

Iso

therm

al



54


0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0 20 40 60 80 100

Oxid

e P

ositio

n w

he

re F

e is m

ostly O

xid

ize

d (

µm

)

Neutron Fluence (1020n/cm2, E > 1 MeV)

Oxide Position where Fe is mostly Oxidized vs Neutron Fluence

Average of autoclave

samples corroded in water

One additional

sample at (217.1, 1)

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0 4 8 12

Oxid

e P

ositio

n w

he

re F

e is m

ostly O

xid

ize

d (

µm

)

Neutron Flux (1013n/cm2/s, E > 1 MeV)

Oxide Position where Fe is mostly Oxidized vs Neutron Flux

Average of autoclave

samples corroded in water

352-355°C

337-339°C

314°C

271-274°C

- There is a clear Neutron Flux and Neutron Fluence correlation to the

position in the oxide where Fe is mostly oxidized



352-355°C

337-339°C

314°C

271-274°C

55


Hypothesis: redistribution of Fe from precipitates

to matrix caused by irradiation leads to higher

surface to volume ratio and more Fe oxidation

closer to the MOI



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