role of materials for lifetime extension · 2020. 4. 17. · atoms for future - october 24th 2012,...

« ROLE OF MATERIALS FOR LIFETIME EXTENSION»

Pascal YVON, Bernard MARINI and

Benoit TANGUY

Department of Materials for Nuclear

Applications, CEA SACLAY

| PAGE

1

Atoms for future - October 24th 2012, Paris

OUTLINE

Context

Effect of neutrons on materials

Role of R&D for lifetime extension through two

exemples

• Pressure vessel

• Internal structures

Conclusions


SOME MATERIALS ISSUES FOR A PWR

•Primary circuit

• Defects under liner of ferritc steel (including vessel)

• Stress Corrosion Cracking of alloy 600 steam

•Thermal Fatigue of thermal barriers of primary pumps

•Thermal ageing embrittlement of some casted components(coudes, corps

de pompes…)

• Thermal ageing embrittlement of some HAZ of ferritic steels with high P

content

•Auxiliary Circuits

• Corrosion in dead zones

• Thermal and vibrational Fatigue

• Internals

•Irradiation damage of bolts (IA-SCC)

• Dimensional changes of internals structures under swelling

• Containement barriers

• rapid degradation of some concretes

…etc


IRRADIATED COMPONENTS

~ 300 C 0.1 dpa

40 60 years

300 – 400 C 10/15 dpa 5 – 6 years

300 – 380 C 30 - 120 dpa

40 60 years

neutrons temperature mechanical stresses environment time

Core Internals Austenitic steels

Fuel Assemblies Zr alloys

Vessel Bainitic steel

16MND5 A508 Cl 3

Core Internals Nickel alloys

Control rods Austenitic steels

~ 320 C ~ 10 dpa few years

~ 320 C few 0.1 dpa

40 60 years

155 bars 293 C Water

H2, LiOH, B

155 bars 328 C


ROLE OF R&D AND METHODOLOGY

R&D outputs of two kinds

Existing materials : evaluation of ageing in order to perform

preventive maintenance on replaceable components

(internals) and justify the life extension of irreplaceable

components (RPV)

Future materials : understanding of ageing mechanisms to

propose optimized materials or improve the design

To perform R&D, we rely on

• Characterization (microstructural, mechanical,..)

• Simulation

• Modelling


EFFECT OF NEUTRONS

Depending on their energies, neutron can have

Nuclear effects (inelastic): - thermal neutrons

Fission

Capture (and subsequent nuclear reactions)

Ballistic effects (energy conservation) – fast neutrons

dpa – point defects

2 to 3 neutrons

2 atoms : Short life

radioactive fission products

+ énergy (~ 200 Mev)

neutron + heavy atom

Fission

Capture

2 to 3 neutrons

2 atoms : Short life

radioactive fission products

+ énergy (~ 200 Mev)

neutron + heavy atom

Fission

Capture


PKA : Primary

Knock on Atom

- For a transferred energy Et < Ed (threshold energy – typically 20-40 eV) -

> vibration of the crystal lattice -> heating

- For a transferred energy Et > Ed, the atom can be ejected from its atomic

site and move through the crystal to other atomic sites (mean free path ~

several atomic sites)

- This creates a vacancy + a self-interstitial atom. This is a Frenkel pair.

EFFECTS OF ELASTIC COLLISIONS INSIDE A CRISTAL


Lattice

distorsion

Interstitial

Vacancy

EFFECTS OF ELASTIC COLLISIONS INSIDE A CRISTAL

- For a transferred energy Et < Ed (threshold energy – typically 20-40 eV)

-> vibration of the crystal lattice -> heating

- For a transferred energy Et > Ed, the atom can be ejected from its atomic

site and move through the crystal to other atomic sites (mean free path ~

several atomic sites)

- This creates a vacancy + a self-interstitial atom. This is a Frenkel pair.


DISPLACEMENT CASCADE (ET >> ED)

For a transferred energy large compared to Ed, the ejected atom transfers part of its energy to other atoms of the crystal lattice...

... these other atoms can then displace other atoms.

The primary knock on atom induces a displacement cascade

Vacancy : yellow Interstitials : red Displaced atoms : blue


MACROSCOPIC EFFECTS OF IRRADIATION ON MATERIALS

After irradiation evolution of mechanical properties can be observed

For instance the tensile testing properties of steel, but also

embrittlement, dimensional changes, enhanced corrosion, precipitation,

segregation, amorphisation….

304-SA irradiated and tested at 325°C

Unirradiated

0,8dpa

1 dpa2 dpa

3,5 dpa

5,5 dpa

9 dpa

0

200

400

600

800

1000

0% 10% 20% 30% 40% 50%

Engineering strain (%)

En

gin

ee

rin

g s

tre

ss

(M

Pa

)


Arrêté de 1974 - Article 9: «… le constructeur montrera en particulier que

l'appareil ne présente aucun risque de rupture brutale en

exploitation. »

PWR VESSEL : SECOND SAFETY BARRIER

Operating life limited by the vessel embrittlement


Temperatures : 296 – 320 C

Coolant pressure: 155 bar

= 4400 mm

e = 220 mm

Gross weight 450 t

Steel A 508 Cl. 3 = 16 MND 5

(bainitic steel)

Internal cladding: 304 L = Z 2 CN 18 10

(austenitic stainless steel)

PWR PRESSURE VESSEL


Lee, 2000

= 4.5 1019 n/cm², Tirr = 288 C

Neutron irradiation embrittles the vessel material

Ductile to brittle transition depends on material

= 4.5 1019 n/cm², Tirr = 288 C

IRRADIATION EFFECTS


ASSESSMENT OF VESSEL INTEGRITY

Atoms for future - October 24th 2012, Paris T

( C)

0

50

100

150

200

-50 100 150 200

K (

MP

a.√

m)

KIc(0 y.)

INTEGRITY ASSESSMENT

FM (0 y.)

F M⩽C S

CS is depending on:

the transient category

the initiation mode (fragile / ductile)

FM = KIc / K > Cs


( C)

0

50

100

150

200

-50 100 150 200

TT

K (

MP

a.√

m)

KIc(0 y.) K

Ic(x years)


FM (0 y.) F

M (x y.)

F M⩽C S

CS is depending on:



FM = KIc / K > Cs


( C)

0

50

100

150

200

-50 100 150 200

TT

K (

MP

a.√

m)

KIc(0 y.) K

Ic(x years)


FM (0 y.) F

M (x y.)

F M⩽C S

CS is depending on:



FM = KIc / K > Cs

Effect of irradiation on DBDT depends

on material

This dependance has to be known in

order to know the life expectancy of

the vessel


The operator must be able to predict irradiation induced embrittlement in order to guarantee

the absence of risk of sudden break

Extrapolation is difficult given the empirical nature of embritittlement models based on PSI

and MTR irradiations

Short and mid termR&D:

- Understanding of embrittlement phenomena by experiments and numerical

simulations

- Qualitative evaluation of critical parameters (chemical composition, neutron flux…)

- Optimization of the empirical models

Long term R&D: base embrittlement models on multiphysical and multi scale models.

Enrichment of data base can lead to

significant changes



Ab initio

Molecular

Dynamics

Dislocation

Dynamics

Crystal

plasticity

Reference

values

Mechanisms:

Dislocation Mobility

Defect Strengths

Local Rules

Single crystal behaviour &

crystal constitutive law

Microstructure

Modeling

Microstrutural

representative

mesh

Macroscopic

behaviour

0

200

400

600

800

1000

0% 2% 4% 6% 8% 10% 12% 14%

Tru

e s

tress (

MP

a)

True strain

Euro material A, irradiated

-90°C

-50°C

+25°C

1

2

MULTISCALE MODELLING


0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0192

T= -150 °C

T= -125 °C

T= -100 °C

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C

EXAMPLE OF PREDICTION OF PWR VESSEL STEEL

EMBRITTLEMENT UNDER NEUTRON IRRADIATION


0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0135

T= -150 °C

T= -125 °C

T= -100 °C

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C




0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0108

T= -125 °C

T= -100 °C

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C

T= 25 °C




0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0099

T= -125 °C

T= -100 °C

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C

T= 25 °C




0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0099

T= -100 °C

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C

T= 25 °C

T= 50 °C




0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0113

T= -75 °C

T= -50 °C

T= -25 °C

T= 0 °C

T= 25 °C

T= 50 °C

T= 75 °C




0

50

100

150

200

250

-200 -150 -100 -50 0 50 100 150

Température (°C)

K (

MP

am

)

99%

95%

Master Curve

5%

1%

Gamma = 0,0119

T= -50 °C

T= -25 °C

T= 0 °C

T= 25 °C

T= 50 °C

T= 75 °C

T= 100 °C




JANNUS PLATFORM

Triple beam

chamber


Fe - 1% Mn MODEL ALLOY

Number density: 2,0 0,7 . 1021 m-3

Mean size: 21 nm

224 loops analysed

z = -101, g = 020, bright field

200 nm

39 % of <001> and 61 % of <111>

• Fe5+(10 MeV) • T = 400 C • Flux: 3.1 1015 ions. m-2.s-1

• Fluence: 1,55 1019 ions.m-2

0.5 dpa

Volume:42x42x86,3 nm3

Number density: 4.8 1.4 1022 .m-3

Radius: R=1.2 0.3 nm

Composition: cMn = 31 9.9 % at

8 nm

E. Meslin et al.

Experimental evidence that radiation-induced segregation (under saturated alloy)

can lead to formation of nanometre-scale solute clusters in ferritic alloys


MODELLING OF CU PRECIPITATION BY KINETIC MONTE CARLO

On each crystal site

Fe or Cu

Thermodynamics

for interactions

between species

Kinetics, according

to probability of

occurrence

Simulated time:

One century


INTERNALS STRUCTURE OF PWR

| PAGE

30

RPV

Fuel Assemblies

Lower

Internals

Upper

Internals

PWR (155 bars)

The internals lifetime has an important impact on the nuclear power plant lifetime because the cost and difficulty of their replacement

Design role of the Lower Internals:

•Support the core weight

•Circulation of the primary

coolant

•Positioning of the core and fuel

assemblies

•Protection of the RPV against

irradiation embrittlement

Design role of the Upper Internals :

• Align the rod control cluster

assemblies with the fuel assemblies

• Immobilize the fuel assemblies


INTERNALS N4 (PWR 1350MW)


INTERNALS STRUCTURE OF PWR

| PAGE

32

900-1000 bolts / reactor vessel

Material Chemical Composition (wt%): stainless steel

C Si Mn P S Ni Cr Mo Fe

304 0.06 0.78 0.96 0.011 0.003 9.3 18.6 / Bal.

316 0.047 0.72 1.12 0.028 0.019 10.65 16.83 2.28 Bal.

4m

Former

Baffle plates

3m •Choice of 304 and CW316 Austenitic Stainless steels for Internals Structures •Bolts : mechanical ties between formers and baffle plates

Baffle plates SS 304

Former

Core Barrel

Bolts CW316 SS

Lower Internals


INTERNALS: DESIGN AND AGEING MECANISMS


IDENTIFIED DEGRADATION MECHANISMS


SAFETY ISSUES OF BAFFLE –FORMER BOLTS

CRACKING


IRRADIATION EFFECTS

Consequences on mechanical properties and sensitivity to IASCC

Dislocation loops

Segregations at grain boundaries :

Cavities, helium bubbles…:

INCREASED SENSITIVITY TO SCC

(IASCC)

POTENTIAL SWELLING

HARDENING

IRRADIATION CREEP

LOCALIZATION-CHANNELLING

TOUGHNESS DECREASE

?


METHODOLOGY

Internal structures of the PWR in austenitic

stainless steels

Laboratory material selection and

characterization

Stress @0-220 MPa

Irradiation @ 320-390°C Primary water

Better understanding at the micro/nano-scale

R&D studies based upon the simulation of certain changes at temperatures used in PWRs

Changes in microstructure, microchemistry and mechanical properties and material degradation -Swelling -Hardening -Irradiation creep -Loss of ductility -Susceptibility to IASCC, -Etc…


OVERVIEW OF THE STUDIES RELATED TO INTERNALS AT

CEA

Microstructure and

radiation hardening

Radiation Induced

segregation

Localization of the

deformation

Swelling

Irradiation Creep

SCC of irradiated

material

Experimental Modelling

Mechanical tests

TEM

TEM-EDX, TAP

Mechanical tests, TEM

In-reactor creep tests,

TEM

SCC tests on

recirculation water

loop

Irradiation at high doses,

Swelling measurement, TEM

Mutiscale

s m

odelli

ng (

rate

theory

,

mo

lecu

lar

dynam

ics,

dis

locatio

ns d

ynam

ics,

cry

sta

l pla

sticity,

me

so

sco

pic

mech

anic

al beha

vio

r)

Swelling mandrel In-reactor IASCC S

imula

tion tool : JA

NN

US

CE

A (

irra

dia

tion

with p

art

icle

s)

Neutron irradiated materials


TESTS ON IRRADIATED STAINLESS STEELS

Neutron Irradiations (volume)

PWR : 6.10-8-1.10-7 dpa/s

Link?


IRRADIATION CREEP


Cracking of PWR bolts

200 m

25 m

Dose : up to 80 dpa

Temp: up to 370 C

Correlation (temp, dose) /

fissuration

IASCC


IASCC - METHODOLOGY

SCC tests on irradiated materials

K1 hot cell, LECI

(Nishioka, JNST,45,2008)

Sensitivity studies: SSRT tests (dynamic)

Crack initiation studies: Constant load tests

(~static)

Determination of a curve below which

there is no crack initiation (depends on

grade, environnement, temperature, …)


IASCC - METHODOLOGY

in situ SCC tests in représentative conditions

Crack initiation studies: Constant load tests

(~static)

(JMTR, Japan)

Comparison of in pile and out of

pile tests

Final validation


MECHANICAL BEHAVIOR OF IRRADIATED SS

Microstructural evolution → modification of mechanical

behaviour • Irradiation defects

• Radiation-induced segregation, second phase precipitation, etc.

• Formation of “clear bands” : localization of plastic deformation

• Evolution of mechanical properties

[Edwards et al., 2003; Pokor et al., 2004a]

Frank loops Black dots Gas bubbles Nano-voids

304L SA 316 CW

[Pokor, 2003]

[Nogaret, 2007]



45

Cristal scale modelling



Monocristal

Agrégat polycristallin simplifié

Agrégat polycristallin avec

tétraèdres de Voronoï

Monocristal poreux

Test en grands

transformations avec

1000 grains cubiques

Test sur un maillage

d’agrégat de 50 grains

Test sur un monocristal

poreux

Tests des lois cristallines

implantées à l’état non-

irradié et irradié

Tests polycristallins et identification

et validation des paramètres du

modèle

Behavior law for dense

monocristal

→ Simulation of

mechanical

behavior of

unirradiated and

irradiated materials

Identifier la loi d’endommagement

du monocristal

Agrégat polycristallin irradié

homogénéisé Etudier la croissance et la coalescence

d’un monocristal poraux

Behavior law for

porous monocristal

→ Evaluation of

swelling effect on

mechanical

properties

Homogenisation


Microstructural informations (input of the modelling)

Tensile curves as a function of irradiation (output of the modelling)

EXAMPLE OF MODELLING


CONCLUSIONS

• The evolution of materials properties under irradiation control the

component life time(Component life time is determined by

engineering approaches of the reactor safety )

• Mechanisms behind these evolutions are numerous and complex

• Experiments, simulations and modelling approaches are

simultaneously needed to study these mechanisms at the

different scales.

• The main objective of R&D programs is to improve engineering

approaches on existing material through qualitative understanding

and macroscopic data production

• Long term R&D (basic research) is dedicated to mechanism

understanding and their quantitative simulation through multi-

scale approach.

• Material design for future reactors is based on mechanisms

understanding and modelling


TO LEARN MORE ABOUT MATERIALS

Direction DEN

Départment DMNI

Commissariat à l’énergie atomique et aux énergies alternatives

Centre de Saclay | 91191 Gif-sur-Yvette Cedex

T. +33 (0)1 69 08 17 25| F. +33 (0)1 69 08 93 24

Etablissement public à caractère industriel et commercial | RCS Paris B 775 685 019

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role of materials for lifetime extension · 2020. 4. 17. · atoms for future - october 24th 2012,...

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