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Tensile properties and microstructure of austenitic steels irradiated in different reactors

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Tensile properties and microstructure of austenitic steels irradiated in different reactors Ph. Dubuisson X. Averty V.K. Shamardin V.I. Prokhorov – PowerPoint PPT presentation

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Title: Tensile properties and microstructure of austenitic steels irradiated in different reactors


1
Tensile properties and microstructure of
austenitic steels irradiated in different reactors
Ph. Dubuisson X. Averty
V.K. Shamardin V.I. Prokhorov
J.P. Massoud C. Pokor
M. Žamboch
Y. Bréchet
Influence of Atomic Displacement Rate
on Radiation-induced Aging of Power
Reactor Ulianovsk, Russia - October 2 - 8,
2005
2
Irradiations in Experimental Reactors
Temperature
  • Reach "rapidly" the end-of-life doses ?
    FBR
  • Mechanical properties
  • SCC
  • Microstructure
  • Modelling

370C
Tensile specimens
360C
  • Materials
  • Representative of Core Internals of the
    PWRs
  • SA 304L Baffle plates, Former, Core barrel
  • CW 316 Baffle bolts

PWR 40 years
320C
300C
C Cr Ni Mn Mo Si Cu
SA 304L 0.022 18.6 9.9 1.8 0.06 0.36 0.25
CW 316 0.054 16.6 10.6 1.1 2.25 0.68 0.24
Dose
30 dpa
95 dpa
3
Irradiations in Experimental Reactors
Irradiations in FBR ? Spectrum effect ?
Flux, Gaz He, H
Temperature
370C
360C
PWR 40 years
  • BOR-60 / Osiris Tensile 10 dpa

320C
  • He effect
  • Mechanical properties
  • SCC tests

300C
SAMARA
Dose
  • Modelling

30 dpa
95 dpa
4
Tensile properties Fast Breeder Reactor BOR 60
320C
Te 330C 3 10-4 s-1
SA 304L
BOR 60
CW 316
Total Elongation 5 10
U.T.S. ? YS0.2
SA 304L gt 5 dpa CW 316 ? 10 dpa
Saturation
CW 316 gt SA 304L - 125 dpa
5
Tensile properties BOR 60 - Osiris
Te 330C 3 10-4 s-1
5 dpa
10 dpa
SA 304L
No difference between Osiris and BOR 60
No effect of neutron spectrum Saturation of
mechanical properties gt 5 dpa
6
Tensile properties BOR 60 - Osiris
320C
Te 330C 3 10-4 s-1
CW 316
SA 304L
BOR 60
SA 304L
Osiris
CW 316
No difference between Osiris and BOR 60
No effect of neutron
spectrum Saturation of mechanical properties
SA 304L 5 dpa CW 316 10 dpa
7
Tensile properties Helium effect SM 2
300C
Te 330C 3 10-4 s-1
Same Flux
6 dpa
17 dpa
SA 304L
600 appm He
10 appm He
14 appm He
300 appm He
SM 2
No obvious effect of Helium (H2)
content Saturation of mechanical properties lt 6
dpa
8
Tensile properties BOR 60 Osiris SM 2
Te 330C 3 10-4 s-1
CW 316
BOR 60
SA 304L
SA 304L
Osiris
SM 2
CW 316
  • No obvious effect of helium (H2) on tensile
    characteristics
  • Tensile characteristics similar to those
    measured after irradiation in Bor-60 (FBR) at
    320C both for CW 316 and SA 304L

No flux effect
9
Tensile properties
  • Saturation dose at 5 dpa for SA 304L 10 dpa
    for CW 316
  • No evolution between 10 and 125 dpa for both SA
    304L - CW 316
  • CW 316 gt SA 304L hardness residual
    ductility
  • No neutron spectrum effect on tensile
    characterictics Gaz content and flux

10
Hardening Model
Evolution of the Yield Stress after
irradiation Temperature, fluence, neutron spectrum
Model of the population of point defects
clusters (dislocation loops)
Microstructural data of neutron irradiated
materials TEM
Model of hardening by a cluster population Ds
proportional a, L
Yield Strength of neutron irradiated
materials Tensile tests
EBR II, Osiris, BOR 60
11
Microstructure
Frank Loops
Black dots
No precipitation No more dislocation lines
EBR II Osiris BOR 60
Main feature Frank loops formation
Saturation for dose about 5 - 10 dpa size SA
304L ? 316 density SA 304L gt CW 316
12
Microstructure Modelling Frank loop
Loops
Chemical kinetic Model "Cluster Dynamics"
Evolution of the concentration of point defects
External source of irradiation defects
Neutron spectrum Flux - EPKA
  • Production
  • - Recombination (v-i)
  • Loss of v and i at sinks
  • Agglomeration

Neutrons
Dislocation network evolution
Homogeneous medium
13
Microstructure Modelling Frank loop
Loops
Chemical kinetic Model "Cluster Dynamics"
External source of irradiation defects
Neutrons
Homogeneous medium
14
Microstructure Modelling Frank loops
Chemical kinetic Model "Cluster Dynamic"
Emv 1.35 eV Emi 0.45 eV Eb2i 0.6 eV r0 1010 cm-2
  1. Adjust material parameters of the model on low
    dose data EBR II - Osiris
  2. Predict the behavior at higher doses EBR
    II Osiris BOR 60
  3. Comparison with experimental data BOR 60
  4. Comparison with future results high doses
    BOR 60 (90 dpa) and Osiris (10 dpa)

15
Microstructure of expertised components
21


/

18,7

18 10
21

21

21

SA 304

310

35

10

15 10
1

1 10
10,2

80 10

In agreement with results from experimental
reactors
In terms of interstitial loops size and
density, the results of the model are in
relatively good agreement with the results
obtained from field experience in PWRs.
IV Back to field experience
Chap IV-1
16
Hardening - Orowan Model
Dislocations network evolution
Cluster Dynamics Model
Good agreement at low dose Data / Model
aloops ? 0,4
Same for CW 316
17
Modified Orowan Model
  • Hardening due to Frank Loops
  • Defaulting of the Frank loops
  • Transformation in perfect loop under a applied
    stress
  • Critical shear stress for defaulting a
    Frank loop of diameter f One relation between
    r and f
  • r and f increase with dose
  • Perfect loop glide and annihilate
  • Saturation at the critical stress
  • Critical dose for the mechanism of hardening

dc
Main parameters g Stacking Fault Energy, rd
One adjustable parameter Number of dislocations
in the pile-up
18
Modified "Orowan" model
Hardening model permitting defaulting of Frank
Loops ? Saturation of Hardening
Voids
r0 1014 cm-2 g 42 Jm-2
r0 1010 cm-2 g 26 Jm-2
330C Good description of experimental
data 375C Need data at high dose to verify
Voids ? data / Model in SA 304
? 2 steels r0 g
Well description of experimental data by the model
19
Slow Strain Rate Tests (SSRT) PWR environment
Te 320C 5 10-8 s-1
Flow rate 2 autoclave vol./h
SA 304L
s MPa
CW 316
s MPa
300C
Air
PWR
Air
5 dpa
PWR
  • T.E. of SSRT specimens strongly reduced
    (compared to tensile tests in air)
  • lower for the specimens with low helium
  • Hardening lower for SA 304 after tests in PWR
  • No significant difference in susceptibility
    between SA304L and CW 316

Slight effect of He content
SA 304L CW 316
20
Fracture surfaces
5 facets
3 facets
SA 304L
CW 316
He (appm)
294
9
297
15
brittle fracture
33,3
71,1
38,0
50,8
Low He
facet initiation
TG
IG
TGIG
IG
fracture in facets
TGIG
IGgtTG
TGIG
IGgtTG
high He
Transgranular
Intergranular
21
Conclusions - Perspectives
  • Tensile properties
  • Saturation dose at 5 dpa for SA 304L 10 dpa
    for CW 316
  • CW 316 gt SA 304L hardness residual
    ductility
  • No neutron spectrum effect on tensile
    characterictics Gaz content and flux
  • Microstructure
  • High density of small Frank loops
    Voids at high temperature in SA 304L
  • Disappearance of the initial dislocations network
  • No precipitation
  • Reproduce Microstructure observed on PWR
    components
  • Hardening Model
  • Cluster Dynamic Model
  • Good agreement with TEM quantification Frank
    loops No real saturation of loop number density
    and diameter
  • Hardening Model
  • Orowan Model No saturation of hardening
  • Modified Orowan Model permitting the defaulting
    of Frank loops Saturation of Hardening

No evolution between 10 and 125 dpa
22
Conclusions - Perspectives
  • In simulated PWR water
  • Total Elongation of the SSRT specimens strongly
    reduced
  • Fracture surface partly intergranular
  • T.E. lower - Fracture surfaces more
    intergranular low helium content
  • Further examinations and SCC tests will be
    performed on more highly irradiated materials
  • Mechanical properties saturate
  • He content increases
  • Intergranular fracture SM 2 gt BOR 60
  • Flux effect ? Medium ?

23
  • This work was performed through a collaboration
    between EDF, CEA and RIAR partly sponsored by
    EPRI
  • Authors are grateful to
  • HT Tang (EPRI),
  • V. Golovanov and G. Shimansky (RIAR),
  • P. Brabec and A. Brožova (NRI)
  • F. Rozenblum, J.C. Brachet and A. Barbu (CEA).
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