Ex-situ and in-situ studies of radiation damage mechanisms in Zr-Nb - - PowerPoint PPT Presentation

ASTM 19th International Symposium on Zirconium in the Nuclear Industry

Ex-situ and in-situ studies of radiation damage mechanisms in Zr-Nb alloys

Junliang Liu1, Guanze He1, Anne Callow1, Kexue Li1, Sergio Lozano- Perez1, Angus Wilkinson1, Michael Moody1, Chris Grovenor*1,Jing Hu2, Mark Kirk2, Meimei Li2, Anamul Haq Mir3, Jonathan Hinks3, Stephen Donnelly3, Jonna Partezana4 and Heidi Nordin5

1 Department of Materials, Oxford University, Oxford, UK. 2 Argonne National Laboratory, Argonne, IL, US 3 School of Computing and Engineering, University of Huddersfield, UK 4 Westinghouse Electric Company, Pittsburgh, PA, US 5 Canadian Nuclear Laboratories, Chalk River, ON, K0J 1J0 Canada

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What are we interested in? The use of advanced analytical techniques to study the response of Zr cladding materials to corrosion and radiation damage

• J. Hu et al. Acta Materialia May 2nd 2019 https://doi.org/10.1016/j.actamat.2019.04.055

Materials Autoclave corrosion: Zircaloy-4 and Zr-0.5Nb: 360°C, 18MPa, Pure water Zr-2.5Nb: 300°C, 10 Mpa, D2O, PH=10.5(LiOD)

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Materials

CNL Halden reactor samples

• In-flux
• Out-of-flux but in-reactor

water chemistry

• Static autoclave with D2O

(pH=10.5, LiOD), and at 300°C and 10 MPa.

• In flux 325 oC samples that

have been extensively analysed

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20 µm

2 µm

Zr-1Nb sheet SRA Zr-2.5Nb tube

AD RD* TD* RD TD ND

Materials

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Aim of study Material Conditions Sample source In-situ heavy ion irradiation

• 1. Irradiation effects in Zr
• xides

RX Zr-0.5Nb Autoclave corrosion Westinghouse and CNL

• 2. Irradiation induced

elemental redistribution RX Zr-1Nb RX and SR Zr-2.5Nb Metal Ex-situ characterisation of in-reactor corroded alloys

• 3. Microstructure of in-

reactor formed oxide SR Zr-2.5Nb In-reactor corrosion CNL

• 4. Irradiation-induced

elemental redistribution 3D mapping of deuterium distribution using NanoSIMS

• 5. The transportation of

hydrogen/deuterium through the oxide layer Ziraloy-4 SR Zr-2.5Nb Autoclave and In-reactor Corrosion Westinghouse and CNL

Experiments undertaken

RX: Recrystallised; SR: Stress Relieved

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Sample exchange Analysis chamber Duoplasmatro n source Cs+ source Multicollection chamber Magnetic sector Electron beam Electron beam Ion beam Ion beam

SEM/FIB/EDX 3D APT In-situ TEM STEM/EELS/EDX NanoSIMS 50

Experimental methods

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Experiment Temperature (K) Ions Flux (ions.cm-2.s-1) Damage rate (dpa/s) Facility In-situ irradiation in oxides 50 1 MeV Kr++ 8 x 1011 1.5x10-3 IVEM 293 700 keV Kr++ 1-5 x 1012 0.5-2.5x10-3 MIAMI2 In-situ irradiation in SPPs and metal matrix 50 1 MeV Kr++ 8×1011 1.5x10-3 IVEM 293 1 MeV Kr++ 8×1011 1.5x10-3 IVEM 623 1 MeV Kr++ 8×1011 1.5x10-3 IVEM 873 1 MeV Kr++ 8×1011 1.5x10-3 IVEM 623 350 keV Kr++ 6×1011 1-3x10-3 MIAMI2 In-Reactor 600, 520 Neutrons 4.3-4.7×1013 ~10-7 Halden Reactor

Irradiation Parameters

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In-situ Ion Irradiation of bulk monoclinic-ZrO2 on 0.5 %-Nb (210 days)

Evolution of oxide structure under in situ 700 keV Kr++ irradiation at room temperature

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0 dpa 3 dpa (5.6x1015 ions.cm-2) 10 dpa (1.9x1016 ions.cm-2)

700 keV Kr++ implantation at 293 K to different damage levels; pre-irradiation, 5.6x1015 ions.cm-2 (3 dpa) and (c) 1.9x1016 ions.cm-2 (10 dpa)

In-situ Ion Irradiation damage in monoclinic-ZrO2

Simulated patterns Rotationally averaged experimental patterns

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In-situ Ion Irradiation damage in monoclinic-ZrO2

Atomic-resolution HAADF STEM image from an oxide grain post-irradiation, 1.9x1016 ions.cm-2 (10 dpa), with corresponding FFT from the whole region, (b) direct measurement of lattice parameters based on the HAADF STEM images

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In-situ Ion Irradiation damage in monoclinic-ZrO2

TKD pattern quality and phase maps from typical regions of in-situ irradiated Zr oxide. (a and d) 0 dpa, (b and e) 4 dpa, 7.4x1015 ions.cm-2 (c and f) 10 dpa, 1.9x1016 ions.cm-2

0 dpa 4 dpa 10 dpa monoclinic

97.5% 10.5% 0.4%

tetragonal

2.5% 4.1% 4.4%

cubic

85.4% 95.2%

Horizontal grain size (nm)

63±6 66±11 97±24

Vertical grain size (nm)

142±28 76±13 162±49 12

In-situ Ion Irradiation damage in ZrO suboxide

In-situ TEM images of suboxide region and metal substrate in the Zr-0.5Nb alloy: (a) pre-irradiation (b) 1014 ions/cm2 Pre-irradiation HAADF STEM image of the region followed during in-situ irradiation and (d) O/Zr atomic ratio map from EELS analysis of the pre-irradiation sample.

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Preferential amorphisation of ZrO suboxide

HRTEM images and FFTs from the interface region (a) pre-irradiation (b) irradiated at 293 K to 1.9x1016 ions.cm-2 (c) irradiated at 50 K to 4 x 1015 ions.cm-2

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J Liu et al. Journal of Nuclear Materials 513, 226-231 2019

RT DF g, 3g 350℃ BF

In-situ Ion Irradiation damage in SPPs in metal matrix Following the same particles during in-situ irradiation

• Morphology of SPPs can easily be obscured by dislocation loops, surface oxide and

bend contours

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In-situ Ion Irradiation damage of β-Nb SPPs in metal matrix

0002 α ത 2110 α ത 2112 α B= 01ത 10 α−Zr 01ത 1 β 110 β 101 β B= ത 111 β−Nb

(a) pre-irradiation BF (b) post-irradiation BF (c) pre-irradiation SAD (d) post-irradiation SAD

EDX line-scan profiles of Nb Ka for SPPs irradiated at 293 K, 623 K, and 873 K with 1 MeV Kr++ to 6.4x1015 ions.cm-2 (15 dpa). Before and after irradiation by 1 MeV Kr++ to 6.4x1015 ions.cm-2 (15 dpa) at 293 K .

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Size changes in ion-irradiated β-Nb SPPs?

• 20%
• 10%

0% 10% 20% 30% 40% 50% 10 20 30 40 50 60 50 293 623 873

Relative size change Radius (nm) Irradiation temperatures (K) β-Nb SPP size after irradiation at different temperatures before after relative size change

20 40 60 80 100 120 140 160 180 200 50 100 150 200 250 300 350

Zr Kα1 Nb Kα1 Simulated pre Nb Zr Kα1 Distance (nm)

Post irradiation 17

In-situ Ion Irradiation damage in Lave phase SPPs in metal matrix

(c)

SPP Zr(Nb, Fe)2 α-Zr Matrix

(a) (b)

FFT from SPP FFT from matrix

HRTEM images and inset FFTs showing the amorphisation of a Laves phase SPP irradiated at 50 K to 6.8x1015 ions.cm-2 (16 dpa), with EDX line-scans over the same SPP before and after irradiation

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How can we study composition changes in in the matrix? APT tips irradiated in Huddersfield MIAMI2 with 650 keV Kr2+ ions

CNL B166 Zr2.5Nb

TEM Before TEM after 15 dpa

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Metal matrix average Nb content of 0.42 at% (cf 0.45 at% in un-irradiated material. No Clusters detected Fe Nb Zr ZrO2

Sample B166 Zr-2.5%Nb 650 keV Kr+ 5 dpa Zr Nb Fe Cr C O Al 98.5 0.42 0.03 0.01 0.12 0.89 0.01

Kr ion irradiation to 5 dpa

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In flux and out of flux CNL samples have been studied to:

• Compare pore distributions between in flux and out of flux

samples

• Analyse differences in oxide grain texture in autoclave and in-

reactor samples

• Study the growth of nano-scale b-Nb precipitates during n-

irradiation

• Analyse the rate at which Nb is oxidised in the oxide under different

conditions

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Does neutron-irradiation damage create extra porosity in the ZrO2?

Fresnel contrast (±500 nm) bright field TEM images from 2000-day autoclave-corroded samples and 2700-day in-reactor sample (1022 n.cm-2, ~8 dpa). Grain boundary nano-porosity is formed in both oxides (yellow arrows) and but voids in the grain interior only in the n-irradiated sample (blue circles).

Autoclave In reactor ~ 8 dpa

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Why are we interested in porosity? Because we can map directly deuterium distributions in oxides. (See Jones et al this afternoon)

Distribution of 2H- and 18O- in a 61-day Zircaloy-4 sample showing interconnected pathways for deuterium. Distributions of 2H- in a 700-day CNL Zr-2.5Nb sample

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• K. Li et al. Applied Surface Science

464, 311-320 2019

Comparing pore distribution at different stages of oxidation Autoclave 0.5%Nb samples after 75 and 165 days

See Poster: Couet et al

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Neutron-irradiation damage in ZrO2: grain size and shape

TKD analysis of oxides on CNL Zr-2.5Nb tubes (a) In-flux, 190 days, 7.6x1020 n.cm-2 (2 dpa), (b) Out-of-flux, 185 days, (c) autoclave, 150 days and (d) autoclave, 700 days.

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All images taken at with B close to [11-20], g = (0002), 4-5g

Neutron induced nano-Nb precipitates

B70 B74

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B56: 250 oC, in flux, low damage No detectible nano-Nb B70: 325 oC, in flux , low damage Small nano-Nb particles B74: 325 oC, in flux , high damage Larger nano-Nb particles and numerous hydrides

Morphology of Ir Irradiation in induced nano nano-precipitates

~ 4.5 nm ~ 2 nm ~ 1.5 nm B70, In flux 190 days, 325⁰C

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Material Number density (No./cm^3) Average long axis (nm) Average short axis (nm) Aspect ratio B70 1.9 dpa 1016 5 ± 1.3 2.4 ± 0.5 2.09 B74 25.2 dpa

• 5. 1015

8.3 ± 3.7 3 ± 0.9 2.71

Siz Size and sh shape of nano-precip ipit itates versus damage le level

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5 10 15 20 2 4 6 8 10 12 14 16 18 20 22 24

Count long axis (nm)

B74 B70

1 2 3 4 5 6 2 4 6 8 10 12 14 16 18

Count short axis (nm)

B74 B70

B62 B74 B56 B70

The same CNL Halden alloys have been analysed by APT

250 oC 190 days 250 oC 2400 days 325 oC 190 days 325 oC 2750 days

Nb

Residual b-Zr

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B70, , APT & TEM result lts Nb Fe

Grain boundary

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B74, APT & TEM results

Nb Fe

Line 1

EDX Nb line profile

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Precipitate number density

Sample Number Density (No. per cm^3) By APT By TEM B70 1.3 x 1017 1016 B74 4.9 x 1016 5.5 x 1015

The factor of ~ 10 between APT and TEM results suggest that the precipitates/clusters inspected by APT and TEM bright field image are different. Only those large, incoherent precipitates got picked up by the TEM bright field image, while all the large and small, coherent and incoherent clusters were counted by APT.

4 nm B74: Nb and Fe do not precipitate together

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Nb enri richment at t c lo loops aft fter 1.9 .9 dpa dpa

B70

EDX analysis of Nb segregation to c loops

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Matrix and cluster compositions from APT Nb in matrix at % Fe in matrix at % Nb in clusters at %

B166 Un-irradiated

0.45 0.07

• B56 250 oC 190 days

0.31 0.07 77

B62 250 oC 2400 days

0.31 0.06 66

B70 325 oC 190 days

0.27 0.06 76

B74 325 oC 2750 days

0.17 0.04 73

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2100 2200 2300 2400 2500 2600 2700 1000 2000 3000 4000 5000 6000

Nb L2 Nb L3 Zr L2 counts Energy (eV) Spectrum 1 Spectrum 2 Zr L3

Id Identify ifyin ing th the oxid xidatio ion rate of f Nb Nb in in se second phase partic icle les by EELS Edge shift of a specific edge indicates the change of oxidation state.

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Oxide Metal

Does the Nb oxidise at the same rate in different samples? In particular, does more n-irradiation result in faster oxidation?

50 100 150 200 1 2 3 4 5 high damage low damage no damage

Nb L3 Edge Shift (eV) Days in oxide for each individual Nb particles

Oxidation State of Nb in Beita Phase

• Oxidation rate of Nb in
• xide: Out of flux > in

flux low damage > in flux high damage

• Oxidation of Nb in β-Nb is

slower than in β-Zr, so the decomposition of β-Zr results in slower oxidation

• f Nb

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Conclusions

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• We have used in situ ion irradiation to study the stability of ZrO2 and metastable ZrO
• xide phases to high damage levels.

The susceptibility of both phases to phase changes has been shown.

• The extraordinary resistance of b-Nb SPPs to radiation damage has been confirmed
• Detailed analysis of the nano-structure of CNL 2.5% Nb samples after n-irradiation has

revealed the details of precipitation and changes in matrix chemistry during reactor exposure.

• The oxidation rate of Nb (and so doping of the oxide phase) is remarkably slowed

down by n-irradiation.

Acknowledgements

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• EPSRC grants (EP/K040375/1, EP/N010868/1 and EP/M018237
• access to the Culham Materials Research Facility.
• electron Physical Science Imaging Centre (ePSIC) on the Harwell campus for access

to the JEOL ARM300CF instrument.

• Access to the IVEM facilities at ANL was provided through the NSUF RTE

scheme

• Access to the MIAMI2 facilities was provided through the EPSRC UK National

Ion Beam Centre (http://www.uknibc.co.uk/).