RADIATION DAMAGES IN MATERIALS PART II Dr. Celine Cabet CEA, DEN, - - PowerPoint PPT Presentation

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RADIATION DAMAGES IN MATERIALS PART II Dr. Celine Cabet CEA, DEN, DMN Service de Recherches de Mtallurgie Physique JANNUS laboratory +33 1 69 08 16 15 celine.cabet@cea.fr Outline 1. Background on alloys and radiation effects 2.

slide-1
SLIDE 1

RADIATION DAMAGES IN MATERIALS – PART II

  • Dr. Celine Cabet

CEA, DEN, DMN Service de Recherches de Métallurgie Physique

JANNUS laboratory

+33 1 69 08 16 15 celine.cabet@cea.fr

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SLIDE 2

Outline

  • 1. Background on alloys and radiation effects
  • 2. Radiation hardening: example of PWR pressure vessel steel
  • 3. Radiation swelling: example of fast reactor cladding
  • 4. Creep irradiation: example of fast reactor cladding… cont’d
  • 5. Radiation growth: example of LWR Zr-alloy cladding
  • 6. Conclusions

| PAGE 2

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SLIDE 3

Austénite Austénite +  ferrite A+M Martensite

Nieq Creq

Austénite Austénite +  ferrite A+M Martensite

Nieq Creq

316 15/15 EM10 F17 Austenite + Ferrite Martensite Ferrite

  • Steels can have different crystalline structures

depending on

  • the type and quantity of alloying elements
  • temperature
  • fabrication route (thermo-mechanical

treatments)

  • 1. Basics of crystalline structure

Structure of metals

Crystalline structure = precise pattern of atoms following a unit cell that is periodically reproduced

| PAGE 3

  • C. Cabet | Radiation damages in materials

Steels

  • Steels = Iron + Carbon + alloying major and minor elements (Cr, Ni, etc.) that

participate to the global mechanical and chemical properties and to the radiation resistance

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SLIDE 4
  • Two important steel types:
  • Austenitic (gamma)
  • Ferritic and martensitic (alpha and alpha’)
  • 1. Basics of crystalline structure

| PAGE 4

  • C. Cabet | Radiation damages in materials

Steels

  • Hexagonal close pack

structure (hcp)

Zirconium Structure of metals

Crystalline structure = precise pattern of atoms following a unit cell that is periodically reproduced

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SLIDE 5
  • 1. Basics of crystalline structure

Structure of metals

Crystalline structure = precise pattern of atoms following a unit cell that is periodically reproduced This crystal is not perfect !

  • Extra atoms (interstitials)
  • r lack of atoms (vacancies)

= point defects

  • Staking fault = dislocations - grain boundaries, interfaces

| PAGE 5

  • C. Cabet | Radiation damages in materials
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SLIDE 6
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

  • Neutrons dissipate energy in the matter by colliding atoms

| PAGE 6

  • Primary damage: atoms are expelled from their

equilibrium site and collide other atoms

  • Atomic displacement cascade :

interstitials + vacancies (Frenkel pair)  reorganization / atomic diffusion (thermally activate): some atoms go back to their initial site  others remains in the crystalline network as interstitials and vacancies

  • C. Cabet | Radiation damages in materials
slide-7
SLIDE 7

0,02 ps

  • 1. Basics of crystalline structure under irradiation

Material damage is quantified in displacement per atom = dpa

Radiation effects in metals

  • Computer simulation of

displacement cascades (DM)

  • Fast recombination

 Few surviving defects in the crystalline structure:

  • Point defects: interstitials and

vacancies

  • I and V clusters formed directly

in the cascade

| PAGE 7

  • C. Cabet | Radiation damages in materials

0,1 ps 0,28 ps

15 keV PKA

0,4 ps 0,8 ps

slide-8
SLIDE 8
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

  • Defects rapidly evolve with time. Depending on the dose, temperature, material

characteristics… these defects

  • Recombine together
  • Annihilate along dislocations and grain boundaries that acts as sinks for defects

 Driving force for interstitial annihilation at dislocation is (slightly) higher than for vacancies  bias

  • Group to form clusters

| PAGE 8

Vacancy clustering -> cavities Intersticial clustering -> faulted loop (Frank)

  • C. Cabet | Radiation damages in materials
slide-9
SLIDE 9
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

  • Defects can migrate and interact with the microstructure and dislocation network
  • Vacancies can be attracted to cavities
  • Frank loops can unfault into other types of loops
  • r as a dislocation line
  • Defects can drag solutes (coupling). This can

accelerate or modify precipitation

  • Extended defects: dislocations, cavities, precipitates

 change in the microstructure

  • with a direct impact on the material properties

| PAGE 9

Frank loop (~60 nm) dislocation lines

extended defects

cavities Precipitation/dissolution Segregation

  • C. Cabet | Radiation damages in materials
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SLIDE 10
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

| PAGE 10

Vacancy clusters

extended defects

cavities Precipitation/ dissolution Segregation

  • C. Cabet | Radiation damages in materials
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SLIDE 11
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

| PAGE 11

Impact on the material properties ?

  • C. Cabet | Radiation damages in materials

Dislocation loop (interstitials) Dislocation loop (vacancies) Screw dislocation Interstitial Precipitate Alloying element (substitution site) Alloying element (interstitial site) Cavity Vacancy

slide-12
SLIDE 12
  • 1. Basics of crystalline structure under irradiation

Radiation effects in metals

| PAGE 12

Impact on the material properties ?

  • C. Cabet | Radiation damages in materials
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SLIDE 13

Radiation embrittlement LWR vessel steel

| PAGE 13 CEA | 7 juin 2012

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SLIDE 14

from G. Was

TEMPERATURE ENERGY DUCTILITY OR FRACTURE APPEARANCE BRITTLE DUCTILE Fully Ductile 30% Ductility 75% Ductility Brittle Fracture

ΔTDBTT ΔUSE

  • 2. Radiation embrittlement – LWR vessel steel
  • Measured with Charpy impact

test and fracture toughness

  • Loss of ductility can lead to

loss of toughness or even failure in some alloys

Ductile-Brittle Transition Temperature

bcc bainitic steel with Mn, Ni, Mo…

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SLIDE 15
  • 2. Radiation embrittlement – LWR vessel steel

| PAGE 15

Embrittlement due to hardening

  • DBTT of irradiated steel
  • Higher strength increases the

probability of failure by cleavage, leading to higher transition temperature

  • DBTT increases with fluence
  • At high dose: occurrence of brittle Mn,

S, Ni enriched phases (late blooming phases)

  • The trend is not linear and saturates (?)
  • C. Cabet | Radiation damages in materials

ΔTDBTT

3.58 1018 n/cm²

7.05 1018 n/cm²

2.22 1019 n/cm²

ΔTDBTT

Data from the French surveillance program DBTT of vessel steel before and after irradiation, 290°C

  • J. Rist, EDF
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SLIDE 16
  • 2. Radiation embrittlement – LWR vessel steel
  • C. Cabet | Radiation damages in materials

Effect of steel purity on hardening and embrittlement

  • Effect of chemical composition through a large body of analytical studies
  • P, S segregate at grain boundaries
  • Cu, Ni clusters inside the grains
  • Cu content was shown to have a strong impact

from G. Was

| PAGE 16

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SLIDE 17

Fe Cu Ni Mn Si P atoms

Combination of experimental, modeling, and microstructural studies leads to advances in predictive capability.

20 keV 40 keV 10 keV

Neutron radiation produces an extremely high number density

  • f nanoscale copper-,

manganese-, nickel-, silicon-, and phosphorus-enriched precipitates.

Microstructure origins of embrittlement

  • Formation of nanoscale precipitates rich in Cu, Ni, Si, P, Mn
  • Composition and size don’t seem to change with dose
  • Number increases with dose and Cu content
  • 2. Radiation embrittlement – LWR vessel steel
  • C. Cabet | Radiation damages in materials

| PAGE 17

from G. Was

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SLIDE 18

| PAGE 18

  • 2. Radiation embrittlement – LWR vessel steel
  • C. Cabet | Radiation damages in materials

Recommended values for DBTT shift calculation

  • Several empirical estimates have been developed to account for the shift in DBTT with

dose and chemical composition

  • NUREG (Nuclear Regulation Board, USA)
  • EDF Framatome CEA

∆𝑆𝑈𝑂𝐸𝑈 = 22 + 556 %𝐷𝑣 − 0.08 + 2778 (𝑄 − 0.008)

𝐺 1019

1 2

∆𝑆𝑈𝑂𝐸𝑈 = 8 + 24 + 238 %𝐷𝑣 − 0.08 + 1537 𝑄 − 0.008 + 192 %𝑂𝑗2%𝐷𝑣

𝐺 1019 0.35

maximum mean

F: fast fluence concentration in weigth %

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SLIDE 19

Swelling SFR cladding

| PAGE 19 CEA | 7 juin 2012

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SLIDE 20

15/15Ti before irradiation (Phénix) 15/15Ti after irradiation (Rapsodie)

  • 3. Swelling – SFR cladding tubes

Swelling is a critical consequence of irradiation for austenitic steels… … and leads to steel embrittlement

  • Diameter increase
  • Elongation
  • Embrittlement at DV/V>6%

« twist » along the spacer wire

[FISSOLO, ASTM-STP 1046, 1988]

| PAGE 20

  • C. Cabet | Radiation damages in materials

Swelling (%) Elongation (%)

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SLIDE 21

Swelling is principally due to a bias…

  • Irradiation  Frenkel pairs are created = 1 interstitial + 1 vacancy
  • Defects evolve
  • Formations of clusters, small loops
  • Recombination/annihilation of defects HOWEVER

Preferential absorption of interstitials at sinks (dislocations)  Vacancies are in supersaturation  Nucleation and growth of cavities This mechanism was observed in the early SFR reactors Main consequences of swelling :

  • Changes in dimensions (elongation, loop deformation, arching/bending)
  • Build-up of internal stresses due to inhomogeneous swelling (under dose gradient,

temperature gradient)  creep is favored

  • Increase in the fuel pellet/clad gap  local heating and promotion of the oxide/clad

interaction (internal corrosion)

  • Embrittlement due to porosity at high dose

 swelling must be controlled in SFR cladding tube

| PAGE 21

  • C. Cabet | Radiation damages in materials
  • 3. Swelling – SFR cladding tubes

100 nm

304 SS SFR 450°C

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SLIDE 22

Cladding tubes in SA 316 from Phénix and DFR swelling as a function of temperature

T< 0,3 T fusion Low atomic mobility, vacancy supersaturation but cavity growth rate remains low  Moderate swelling. T> 0,5 T fusion High atomic mobility, high cavity growth rate but no nucleation of cavities  Moderate swelling.

Swelling follows a bell curve S = vacancy density × vacancy volume

| PAGE 22

  • Microstructure evolution (point defects, diffusion, precipitation…) depend on irradiation

dose (dpa) and temperature

Swelling is not linear in a reactor…

  • 3. Swelling – SFR cladding tubes
  • C. Cabet | Radiation damages in materials

swelling (%)

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SLIDE 23

| PAGE 23

Deformation gradient through tube wall Deformation gradient along a clad tube

  • Microstructure evolution (point defects, diffusion, precipitation…) depends on irradiation

dose (dpa) and temperature

  • Gradients of temperature and flux in the core induce microstructure and property

gradients

[MAILLARD, ASTM-STP 1175, 1994]

Swelling is not linear in a reactor…

  • 3. Swelling – SFR cladding tubes
  • C. Cabet | Radiation damages in materials

ΔD/D %

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SLIDE 24

Swelling shows an incubation time

  • Steel swelling is a phenomenon with a threshold,

which offers an operating widow below a given irradiation dose

  • The operating window/maximum dose depend on the steel (structure, chemistry,

metallurgical state).

  • Swelling is a key issue for austenitic steels but also appears in ferritic steels at high doses

 R&D for the last 40 years has allowed to push the threshold further to higher dose (improving defect recombination, increase sink density)

1 2 3 4 5 10 20 30 40 50 60

DV/V %

Dose dpa

SA 316 600°C Phénix Incubation Steady state Transient

0.5 1 1.5 2 2.5 3 20 40 60 80 100 120 140 160 180 200 220 Swelling (%) dose (dpa) EM10 T91 HT

Tirr : 390-420°C [Dubuisson 1993] [Van den Bosh 1994] [Gelles 1994]

[Henry 2013]

| PAGE 24

Note: irradiation induces other changes in microstructure (precipitation) and behavior

  • 3. Swelling – SFR cladding tubes
  • C. Cabet | Radiation damages in materials
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SLIDE 25

How can we improve swelling resistance?

  • In helping recombination between defects and defects annihilation

Any « fault » in the microstructure is a potential sink for irradiation defects

  • Interface with precipitates
  • Grain boundaries
  • Dislocations (cold work)
  • …
  • In adding swelling inhibitors in solid solution : C, Si, Ti, P, N which can surround the

dislocations and thus increase sink efficiency for vacancies

| PAGE 25

  • 3. Swelling – advanced austenitic steels for SFR
  • C. Cabet | Radiation damages in materials
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SLIDE 26

| PAGE 26

40 years

  • f R&D

Austenitic steels – improving metallurgy and chemistry

Cold work 316Ti

17%Cr 14%Ni

  • cladding and hex tubes
  • PX 197686, SPX 198698
  • Increase of the incubation dose

< 90 dpa

  • Cladding tube and HT in Rapsodie et Phénix  1980
  • Beneficial effect of cold work (dislocation = trap)
  • Dislocation network instability at T >550°C  swelling

Cold work 316 Solution annealed 316

  • 1st reference for Rapsodie-1968 and Phénix-1973
  • Poor resistance to swelling (~50dpa)

< 50 dpa

SA 316 DV/V=10% 316Ti CW DV/V=0%

hyp hyp

  • 3. Swelling – advanced austenitic steels for SFR

fine TiC, (M,Ti)2P precipitates pin the dislocations ΔD/D %

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SLIDE 27

Best of

  • Minor alloying element optimisation (P, Si…)

 French specification AIM1

< 130 dpa

40 years

  • f R&D
  • 3rd reference for PX cladding tubes (1982)
  • Beneficial effect of Cr/Ni = 1

CW 15/15-Ti

15%Cr 15%Ni

< 115 dpa

CW 316Ti

17%Cr 14%Ni

  • 2nd reference for cladding and hex tubes
  • PX 197686, SPX 198698
  • Increase of the incubation dose by Ti addition

< 90 dpa

Very Best of ? • Further optimisation toward AIM2

<< 130 dpa

?

| PAGE 27

Austenitic steels – improving metallurgy and chemistry

  • 3. Swelling – advanced austenitic steels for SFR
  • C. Cabet | Radiation damages in materials
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SLIDE 28

Embrittlement limit

best of 316

[Séran et al.]

Expected for advanced austenitic steels.

| PAGE 28

  • 3. Swelling – advanced austenitic steels for SFR

Main features of swelling (austenitic steels):

  • Swelling is due to vacancy supersaturation
  • Cavities are formed that embrittle the steel
  • Swelling varies with dose, dose rate, temperature and steel
  • Swelling resistance can be greatly improved by tuning the chemical composition and

thermomechanical treatments

  • C. Cabet | Radiation damages in materials
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SLIDE 29

Irradiation creep SFR cladding

| PAGE 29 CEA | 7 juin 2012

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SLIDE 30

Tube deformation results from several contributions

  • Swelling (= cavities)
  • Thermal creep: plastic deformation under load from fission

product pressure or fuel contact

  • Irradiation creep: plastic deformation under load increased

by irradiation

  • 4. Irradiation creep – SFR cladding tubes

347 clad (EBR-II)

[Appleby 1972]

| PAGE 30

irradiation creep swelling

  • C. Cabet | Radiation damages in materials

ΔD/D %

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SLIDE 31

Creep in a nutshell

  • Time-dependent plastic deformation
  • f an alloy under constant load

and at high temperature

  • Creep curve (ideally) shows 3 stages:

transient / steady state / tertiary

  • Temperature provides the energy for deformation

through the creation and diffusion of defects

  • In the steady state stage, creep can occur through

different mechanisms depending on stress and temperature (Ashby maps) like diffusion (at boundaries

  • r in the grain) and dislocation glide or climb
  • 4. Irradiation creep – SFR cladding tubes

| PAGE 31

  • C. Cabet | Radiation damages in materials
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SLIDE 32

Irradiation creep is athermal

  • Occurs under load and irradiation (+ T)
  • Radiation produced point defects increase diffusion and allow creep at lower temperature

 radiation induced creep or at a higher rate  radiation enhanced creep

  • 4. Irradiation creep – SFR cladding tubes

| PAGE 32

  • C. Cabet | Radiation damages in materials

Argon

6.0 6.5

10 20 30 40 50

Diameter mm Length mm

Tube Length Unirradiated 28 dpa 41 dpa 70 dpa 52 dpa

DL/L = 0 no swelling Deformation only under irradiation

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SLIDE 33

Argon

1 2 3 4 5 6 7 20 40 60 80 100 120

dose dpa

SA 304

220 MPa 188 MPa 150 MPa 127 MPa

330°C

e = B0 (sn ft – B1)

1 2 3 4 5 6 7 5000 10000 15000 20000 25000

s . ft MPa dpa

B0

Deformation only under irradiation

Irradiation creep is athermal

  • Linear deformation versus dose (dpa) and versus load (MPa)
  • In steady state domain, irradiation creep deformation

with B0 athermal coefficient

  • 4. Irradiation creep – SFR cladding tubes

| PAGE 33

  • C. Cabet | Radiation damages in materials

ΔD/D % ΔD/D %

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SLIDE 34

Radiation growth LWR cladding tubes

| PAGE 34 CEA | 7 juin 2012

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SLIDE 35
  • 5. Irradiation growth – LWR cladding

| PAGE 35

Growth of Zr4-cladding under irradiation

  • Deformation free of stress
  • Elongation of the cladding can reach several cm
  • Growth depends on temperature and neutron flux
  • Similar rods have different behavior (see picture). During cycle 5, rod #8 contacted both

assembly end-plates. It was somewhat twisted which may affect the thermal hydraulic conditions of the rod.

  • Margin is taken for free elongation of the PWR cladding
  • Other components are concerned : grid in PWR and channel box in BWR whose growth

can cause difficulty in fuel handling

Cycle 4 Cycle 5

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SLIDE 36
  • 5. Irradiation growth – LWR cladding

| PAGE 36

Growth of zirconium under irradiation

  • Specific to textured materials with anisotropic behavior like Zr or U
  • Elongation in the <a> direction and shortening in the <c> directions at constant volume
  • Acceleration of the growth at high doses (related to <c> loop formation)

Carpenter et al., 1981 a c a c Irradiation Growth Strain at 553K in Annealed Iodide and Zone-refined Zirconium Single Crystals

  • C. Cabet | Radiation damages in materials | PAGE 36
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SLIDE 37
  • 5. Irradiation growth – LWR cladding

| PAGE 37

Mechanism for growth under irradiation

  • Cladding tubes have a strong texture with basal pole

<c> along the radial direction (R) and <a> direction along the z-axis  Growth in the <a> direction with tube elongation

  • Interstitial preferential migration parallel to basal

planes and condensation on prism planes

  • Vacancy preferential condensation on basal planes

Crystal structure of -Zr

a a c c

  • C. Cabet | Radiation damages in materials | PAGE 37
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SLIDE 38

Limitation of irradiation growth

  • Increasing isotropy helps in limiting irradiation growth of Zr-alloys
  • In the annealed state, unit cells of Zr-alloy are more randomly distributed. Quench

also improves the behavior by keeping the high-temperature bcc structure

  • Cold work which favors recrystallization with a high texture appears to be negative
  • 5. Irradiation growth – LWR cladding

Irradiation Growth Strain at 353 and 553 K in Annealed and 25% Cold-worked Zircaloy-2 Rogerson A., JNM (1988)

  • C. Cabet | Radiation damages in materials | PAGE 38
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SLIDE 39

Radiation damage in Materials conclusion

| PAGE 39 CEA | 7 juin 2012

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SLIDE 40

Conclusions

  • Irradiation produces point defects in steels. Some interstitials and

vacancies evolve in clusters and extended defects like Frank loops, cavities. These features interact with the steel microstructure and change its properties.

  • Different mechanisms are induced or enhanced under irradiation

depending on environment (temperature, dose/fluence, dose rate, spectrum, load, fluid chemistry…) and materials (composition, thermomechanical treatments, history…)

  • Chemical changes
  • Radiation enhanced diffusion
  • Segregation at surface, grain boundaries,

dislocations, interfaces…

  • Precipitation / dissolution
  • Interstitial clustering
  • Dislocation loops
  • Cavities, voids
  • Bubbles (pressurized with He, H2)
  • Change in dislocation network

| PAGE 40

20 nm Black dot

320°C

Dislocation loops

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SLIDE 41

Conclusions

  • Irradiation produces point defects in steels. Some interstitials and vacancies evolve in

clusters and extended defects like Frank loops, cavities. These features interact with the steel microstructure and change its properties.

  • Changes in microstructure impact steel properties

 Change of dimension

  • Swelling
  • Growth
  • Irradiation creep

 Change in mechanical properties

  • Hardening
  • Embrittlement due to hardening, induced by segregation, due to cavities…

 Specific interactions with environment

  • Irradiation assisted stress corrosion cracking

In fact, all these aging processes under irradiation may be correlated !

  • Modelling coupled to fine characterizations and well controlled experiments is a

powerful tool to go further into the understanding of this complex metallurgy

| PAGE 41

from G. Was

  • C. Cabet | Radiation damages in materials
slide-42
SLIDE 42

Céline CABET CEA Paris-Saclay, DEN/DANS/DMN/SRMP Laboratoire JANNUS 91191 Gif Sur Yvette celine.cabet@cea.fr Tel : +33 672458311

| PAGE 42 CEA | 7 juin 2012