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Influence of Thermo-mechanical Cycling on Pre-hydrogenated Zircaloy-4 … · 2019-06-26 ·...

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IRSN/FRM-296 ind 5 Influence of Thermo-mechanical Cycling on Pre-hydrogenated Zircaloy-4 Embrittlement by Radial hydrides Jean Desquines*, Manfred Puls**, Stéphane Charbaut* and Marc Philippe* *IRSN **MPP Consulting
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Page 1: Influence of Thermo-mechanical Cycling on Pre-hydrogenated Zircaloy-4 … · 2019-06-26 · Thermo-mechanical Cycling on Pre-hydrogenated Zircaloy-4 Embrittlement by Radial hydrides

IRS

N/F

RM

-29

6 in

d 5

Influence of

Thermo-mechanical Cycling

on

Pre-hydrogenated Zircaloy-4 Embrittlement

by

Radial hydrides

Jean Desquines*, Manfred Puls**, Stéphane Charbaut*

and Marc Philippe*

*IRSN

**MPP Consulting

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2 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Context & Objectives▌ Situations where thermo-mechanical cycling of fuel rods is expected:

- Dry cask storage,

- Spent fuel transportation from one spent fuel pool to another one

▌ Expected influence of cycling based on literature data

o No clear trend (Kearns, Mishima, Chu, Sakamoto, Billone,...)

o Possibly more radial hydrides and rather stronger embrittlement

▌ Main Goals of the study: clarify the expected influence of cycling on radial

hydride precipitation:

- Relying on a large data set on single cycle test results using « C »-shaped samples,

- Then cycle samples and evaluate the radial hydride precipitation and cladding

embrittlement.

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3 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

1- Experimental protocol

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4 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

“C”-Shaped Compression Tests in the Elastic Range

0

1

2

3

4

5

6

7

8

0 0.1 0.2 0.3 0.4 0.5 0.6

fqq= 0.0

X: location within the cladding thickness

(0: Outer Diameter, 1: Inner diameter)

fqq= 0.0

s:

cu

rvil

ine

ar

dis

tan

ce

to

A(m

m)

𝜎𝑡 𝑥, 𝑠 = 𝜎𝑡 𝐴 . 𝑓𝜃𝜃 𝑥, 𝑠

𝜎𝑡 𝐴 =𝐹 𝑁

0.013 𝐿 𝑚𝑚

Stress free area

Stress free area

A q

r

CCT sample

Load

Load

x

1s

0

A

F

𝐿

Equator

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5 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Radial Hydride Heat Treatments (RHT) considered

0

100

200

300

350

0 4 8 12 16 20

Com

pre

ssio

n load

Tem

pera

ture

(°C

)

Time (h)

5 N

315°C

FRHT

20°C/h

2h1h

Tem

pera

ture

(°C

)Time (h)

5 N

365°C

FRHT

20°C/h

0

100

200

300

400

0 4 8 12 16 20

2h1h

Com

pre

ssio

n load

350°C Max temperature RHT 400°C Max temperature RHT

• The load is applied as late as possible to limit the sample creep but

significantly before hydride precipitation

• The creep is expected to increase linearly with cycle number

F[H] contents – 50 to 236 wppm

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6 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Equator

Post-test metallography

Using CCT to study radial hydride precipitation

Hydrided sampleF

RHT, 1 to 30 cycles

(Constant Load)

F.d

at RT

Failure test: CCT

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7 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Experimental device

500 N load cellFan cooling

the load cell

FurnaceElectromechanical tension-

compression machine

L1

L2

F.d

2.Re

Furnace

Simple model of the test

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8 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

𝑊𝑒𝑙𝑙 𝑘𝑛𝑜𝑤𝑛𝑚𝑎𝑡𝑒𝑟𝑖𝑎𝑙 𝑝𝑟𝑜𝑝𝑒𝑟𝑡𝑦𝐹𝐸 𝑐𝑎𝑙𝑐𝑢𝑙𝑎𝑡𝑖𝑜𝑛

Interpretation of the RHT load displacement record

F.d

Furnace

𝛿 = 𝛿𝑜𝑓𝑓𝑠𝑒𝑡 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑒𝑙 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒

𝑡ℎ𝑒𝑟𝑚𝑎𝑙 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑐𝑟𝑒𝑒𝑝

+ 𝛿𝑟𝑜𝑑𝑠𝑡ℎ𝑒𝑟𝑚𝑎𝑙

Reasonably good agreement is obtained between calculations and

measurement assuming 0 creep

RE

OR

-103 –

firs

t cycle

𝑎𝑑𝑗𝑢𝑠𝑡𝑒𝑑

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9 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Interpretation of the RHT load displacement record

𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑐𝑟𝑒𝑒𝑝

= 𝛿 − 𝛿𝑚𝑜𝑑𝑒𝑙

Creep deformation remains below ~1% diameter change after 30 cycles

F.d

Furnace

time (h)-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 100 200 300 400 500

𝛿 − 𝛿𝑚𝑜𝑑𝑒𝑙(𝑚𝑚)

REOR-103 –30 RHT cycles

Creep displacement (temperature stabilized)

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10 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

RHT & CCT Test matrix

350 & 400°C Max temperature – 1 to 30 cycles – [H] content below 240 wppm

Maximum stress at equatorial location: between 190 and 220 MPa

RHT

parameters

REOR-XXX ([H](wppm)) – Test ID and associated hydrogen content

Italic: the hydrogen content of tested conditions in past programs.

Cycles Number 1 2 5 10 30

Max R

HT t

em

pera

ture

(°C

)

350

53 ; 63 ;

69 ; 74 ;

REOR-117 (109)

127 ; 141 ;

177 ; 217 ;

309 ; 322 ;

525 ; 540

REOR-114 (176)

REOR-118 (95)

REOR-115 (101)

REOR-110 (139)

REOR-106 (203)

REOR-119 (86)

REOR-116 (94)

REOR-111 (138)

REOR-112 (209)

REOR-113 (210)

400

REOR-131 (61)

REOR-133 (141)

REOR-134 (187)

217 ; 218 ;

398 ; 478

REOR-126 (62)

REOR-120 (78)

REOR-125 (147)

REOR-123 (236)

REOR-121 (75)

REOR-124 (148)

REOR-135 (177)

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11 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

2- Analysis of Metallographic data

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12 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

RHT at 350°C max temperature

• When cycling: higher density but comparable fraction of radial hydrides

• Saturated influence of cycling over 5 cycles

1 5 10 30

130

200

REOR-17 [11]

REOR-14 [11] REOR-110

REOR-106

REOR-111

REOR-112 REOR-113

Cycles

[H]

(wppm)

Meta

llogra

phs

at equ

ato

rial lo

cation

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13 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

RHT at 400°C max temperatureM

eta

llogra

phs

at equato

rial lo

cation

1 5 10

150

200

REOR-134

REOR-125

REOR-123

REOR-124

Cycles[H]

(wppm)

75

REOR-120

REOR-133

REOR-135

REOR-131 REOR-121

At H~200 wppm: higher density but comparable fraction of radial hydrides

Below H~150 wppm: no influence of cycling

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14 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Influence of cycling and stress on incipient radial

hydride precipitation

Slight influence of cycling at 350°C maximum temperature

No clear influence at 400°C

Ta

nge

ntialstr

ess (

MP

a)

[H] (wppm)

Mixed radial & circumferential hydrides

Circumferential hydrides

0

20

40

60

80

100

120

0 100 200 300 400 500 600

Temp.

12

35

0 5

10

30

RHT cycles

400°C RHT350°C RHT

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15 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Influence of cycling and stress on radial hydride

precipitation

Limited influence of cycling on radial hydride precipitation stress thresholds

Circumferential hydrides

Ta

ng

en

tia

lstr

ess (

MP

a)

[H] (wppm)

Mixed radial & circumferentialhydrides

Radial hydrides

(Eq.10)

0

20

40

60

80

100

120

140

160

180

200

0 100 200 300 400 500 600R

HT

max

imu

m t

emp

erat

ure

(°C

) 12

35

04

00

45

0

5

10

30

5

10

1

RHT cycles0%

10

0%

1

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16 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

3- Radial hydride

Embrittlement

F, d

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17 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Load displacement records – an illustration

Crack nucleation is associated with the first significant load-drop and

deviation to the elastic-plastic trend

0

2

4

6

8

10

12

14

16

18

20

0 0.5 1 1.5 2 2.5Displacement (mm)

CCT-112 (209wppm)Elasticity

Plasticity

CCT-115 (101wppm)

Cra

ck n

ucle

ation

Brittle

cra

ck p

ropagation

Cra

ck n

ucle

ation

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18 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Load displacement records

Surprisingly ductile nucleation was associated with brittle crack

propagation and brittle crack nucleation with ductile crack propagation

Rather high [H] contents

&low fracture toughness

Rather low [H] contents

&high fracture toughness

Elasticity

Plasticity

ducti

le

bri

tlle

bri

tle

ducti

le

Cra

ck

nucle

ati

on

Cra

ck

pro

pagati

on

F/L

(N/m

m)

Displacement (mm)

0

2

4

6

8

10

12

14

16

18

20

0 0.5 1 1.5 2

REO

R-1

10

REO

R-1

06

REO

R-1

11

REO

R-1

12

REO

R-1

13

REO

R-1

14

REO

R-1

15 REOR-116

REO

R-1

17

REO

R-1

18

REO

R-1

19

REOR-120

REOR-121 REO

R-1

23

REO

R-1

24

REO

R-1

25

RC

T-1

26

REOR-131

REO

R 1

33

REO

R-1

34

REO

R-1

35

2.5

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19 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Crack nucleation displacement – 1 cycle

Maximum sensitivity to crack nucleation at about [H]= 100 wppm,

No obvious influence of maximum RHT temperature

Cra

ck n

ucle

atio

nd

isp

lace

me

nt

(mm

)

[H] (wppm)

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

0 100 200 300 400 500 600

350°C

400°C

450°C

Maximum RHT temperature

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20 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Crack nucleation displacement – several cycles

[H] (wppm)

0.5

1.0

1.5

2.0

0 100 200 300 400 500 600

Protective influence of RHT cycling

Detrimental influence of RHT cycling

2

350

400

5

10

30

5

10

RHT cyclesTemp.

Cra

ck n

ucle

atio

nd

isp

lace

me

nt

(mm

)

350°C RHT cycling trend

• Regions associated with higher density of radial hydrides have a lower

sensitivity to crack nucleation

• Protective influence of RHT cycling on crack nucleation

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21 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Conclusion▌ The thermo-mechanical cycling was studied relying on CCT tests

▌ Analysis of metallographs

- At 350°C for H contents between 75 and 230 wppm, and at 400°C for H contents

close to 200 wppm: shorter and denser hydrides observed,

- A 400°C below 200 wppm, no influence of cycling is observed

- The denser and short hydrides were associated with slightly easier radial hydride

precipitation

▌ A protective influence of cycling was observed on radial hydride

embrittlement.

▌ Extrapolation of these results to irradiated claddings:

If there is any location containing 100 wppm, then no influence of cycling is

expected

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22 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

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23 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Stress at crack nucleation– several cycles

Brittle crack nucleation region remains at [H] contents of about 100 wppm

and no influence of cycling is observed in this region

Cra

ck n

ucle

atio

nta

ng

en

tia

lstr

ess

(MP

a)

[H] (wppm)

Plasticity

Elasticity

500

700

900

1 100

1 300

1 500

0 100 200 300 400 500 600

2

350

400

5

10

30

5

10

RHT cyclesTemp.

350°C RHT cycling trend

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24 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

-3.0

-2.0

-1.0

0.0

0.5

0 2 4 6 8 10 12 14 16 18

ModelRecord

d(mm)

50 N loading

Cooling @ 20°C/h

2h dwell @ 350°C

heating

Time (h)

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25 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑒𝑙 =

𝐹 1 − 𝜈2

2.6410−3𝐸. 𝐿𝑒

𝑅𝑒 − 𝑒

2.7644

Interpretation of the RHT load displacement record

L1

L2

F.d

2.Re

Furnace

𝛿 = 𝛿𝑜𝑓𝑓𝑠𝑒𝑡 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑒𝑙 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒

𝑡ℎ𝑒𝑟𝑚𝑎𝑙 + 𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑐𝑟𝑒𝑒𝑝

+ 𝛿𝑟𝑜𝑑𝑠𝑡ℎ𝑒𝑟𝑚𝑎𝑙

𝛿𝑟𝑜𝑑𝑠𝑡ℎ𝑒𝑟𝑚𝑎𝑙

= −15.510−6 𝑇 − 𝑇0 . 𝐿1 + 𝐿2

𝛿𝑠𝑎𝑚𝑝𝑙𝑒𝑡ℎ𝑒𝑟𝑚𝑎𝑙

= −5.610−6 𝑇 − 𝑇0 · 2𝑅𝑒

Reasonably good agreement is obtained between calculations and

measurement assuming 0 creep

RE

OR

-103 –

firs

t cycle

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26 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Load displacement records

Surprisingly ductile nucleation was associated with brittle crack

propagation and brittle crack nucleation with ductile crack propagation

0

2

4

6

8

10

12

14

16

18

20

0 0.5 1 1.5 2 2.5Displacement (mm)

CCT-112 (209wppm)

𝐹 𝐿𝑁/𝑚

𝑚

Elasticity

Plasticity

CCT-115 (101wppm)

Cra

ck n

ucle

ation

Brittle

cra

ck p

ropagation

Cra

ck n

ucle

ation

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27 19TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE NUCLEAR INDUSTRY

Elasticity

Plasticity

ducti

le

bri

ttle

bri

ttle

ducti

le

Cra

ck

nucle

ati

on

Cra

ck

pro

pagati

on

F/L

(N/m

m)

Displacement (mm)

0

2

4

6

8

10

12

14

16

18

20

0 0.5 1 1.5 2

REO

R-1

10

REO

R-1

06

REO

R-1

11

REO

R-1

12

REO

R-1

13

REO

R-1

14

REO

R-1

15 REOR-116

REO

R-1

17

REO

R-1

18

REO

R-1

19

REOR-120

REOR-121 REO

R-1

23

REO

R-1

24

REO

R-1

25

RC

T-1

26

REOR-131

REO

R 1

33

REO

R-1

34

REO

R-1

35

2.5


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