Transcript Irradiation damage in Fe

MATERIAL ISSUES FOR ADS:
MYRRHA-PROJECT
A. Almazouzi
SCKCEN, Mol (Belgium)
On behalf of
MYRRHA-TEAM and MYRRHA-Support
MYRRHA – concept:
a multipurpose ADS
Proton
Accelerator
Subcritical
Neutron
Multiplier
Spallation
Source
•
•
•
•
Windowless design
Pb-Bi technology
Spallation products
MA transmutation
Proton
Source
• Material testing
• Radioisotope production
• Proton therapy
•
•
•
•
•
Neutron
Source
Material testing
Fuel irradiation
MA & LLFP transmutation
Radioisotope production
Neutron beams
2
Purpose of Myrrha
MYRRHA is intended to be:
 A full step ADS demo facility
 A P&T testing facility
 A flexible irradiation testing facility in
replacement of the SCKCEN MTR BR2
(100 MW)
 An attractive fast spectrum testing
facility in Europe
 An attractive tool for education and
training of young scientists and
engineers
 A medical radioisotope production
facility
3
Neutronic Design constraints
High energy flux
High dose accumulation
Fast neutron spectra
Important gaz production rate
a ppm o f g a s / d pa
1000
100
a ppm H / d pa
a ppm H e / d pa
10
1
0
3
6
9
Z - a x ia l po sitio n (cm )
12
15
4
Design constraints
Operating conditions
Properties needed
•Temperature (200 to 450°C)
•High thermal conductivity, heat
resistance, low thermal expansion
•High dose rate /high dose
(5.1015n/cm2.s and up to
40dpa/year)
•Low DBTT shift, sufficient strength,
limited loss of ductility and fracture
toughness, low swelling rate
•High He/dpa production rate
(3 to 8 appm)
•Adequate resistance to He and H
embrittlement
•Beam trips and
loading/reloading operations
•Resistance to fatigue in LBE
•High stress level (100 MPa),
operational trips
•High creep and fatigue resistance
•Aggressive conditions (LBE)
•Corrosion and liquid metal
embrittlement resistance
5
Materials for MYRRHA
The R&D program concerning the assessment
of the materials suitable to sustain the design
constraints should follow two routes:
•Experiments and tests to support the
engineering design: the material has to be
tested under condition relevant (as close as
possible) to the foreseen operational
conditions to ensure an economically viable
and safe operation of MYRRHA.
Standard materials
database
Design conception
and
engineering
•Research to build the ability to interpolate
and extrapolate the results from laboratory
tests to the real life.
Dedicated
engineering
database
Fundamental
understanding
6
MYRRHA materials
Bending magnet
Proton beam line
Secondary coolant
Secondary coolant
Spallation loop pump
Diaphragm
Main heat exchangers
Subcritical core – T91
Heat exchanger
Spallation target – T91
Main primary
circulation pump
Reactor vessel – SS316L
7
F/M Steels
FFTF-database
8
FP5-SPIRE Irradiated
material
• Materials: EM10, T91, HT9
E M 10
T 91
HT9
E M 10
T 91
HT9
C
0.099
0.099
0.204
Ti
0.01
< 0.005
Ni
0.07
0.24
0.66
N
0.014
0.03
Cr
8.97
8.8
11.68
P
0.013
0.02
0.020
EM10
- Normalised at 990°C/50’
- Tempered at 750°C/60’
Mo
1.06
0.96
1.06
Mn
0.49
0.43
0.63
Cu
0.05
0.05
O
0.001
Si
0.46
0.32
0.45
B
< 0.001
< 0.0005
S
< 0.003
0.004
< 0.003
W
< 0.002
< 0.01
0.47
Al
< 0.016
< 0.01
Sn
< 0.005
0.006
Nb
< 0.002
0.06
0.03
As
0.003
0.011
Co
0.03
0.03
Sb
0.01
0.012
V
0.013
0.24
0.29
Fe
bal.
bal.
bal.
T91
- Normalised at 1040°C/60’
- Tempered at 760°C/60’
HT9
- Normalised at 1050°C/30’
- Tempered at 700°C/120’
9
Irradiation effects on the yield
stress (hardening): high/low flux
BR2/HFR data
BOR60-data
600
T e s t te m p e ra tu re = Irra d ia tio n te m p e ra tu re
Y ie ld s tre n g th in c re a s e (M P a )
500
F ittin g c u rve E U R O F E R 9 7 d a ta
400
300
E U R O F E R 9 7 [B R 2 - T = 3 0 0 °C ]
200
E U R O F E R 9 7 [H F R - T = 3 0 0 °C ]
F 8 2 H [H F R - T = 3 0 0 °C ]
T 9 1 [B R 2 - T = 2 0 0 °C ]
100
E M 1 0 [B R 2 - T = 2 0 0 °C ]
0
0
2
4
6
8
10
12
D o s e (d p a )
Courtesy: SPIRE-program
Yamamoto et al. 2004
10
Impact test results
8
6
U n irra d ia te d (S C K )
7
S C K - 2 .4 7 d p a , 2 0 0 °C
5
6 .5 %
S C K - 3 .7 0 d p a , 2 0 0 °C
7 .3 %
A b s o rb e d e n e rg y K V (J )
A b s o rb e d e n e rg y K V (J )
6
5
4
3
4
41%
37%
3
2
1 6 9 °C
1
U n irra d ia te d T -L
2
1 3 1 °C
0
1 0 9 °C
1
1 6 3 °C
2 .4 3 d p a , 2 0 0 °C
3 .5 8 d p a , 2 0 0 °C
-1 5 0
-1 0 0
-5 0
0
50
100
150
200
250
300
T e m p e ra tu re (°C )
0
-2 0 0
-1 5 0
-1 0 0
-5 0
0
50
100
150
200
250
300
T e m p e ra tu re (°C )
250
250
M A N E T -II
M A N E T -I
200
S C K te sts T 9 1
(T irr = 2 0 0 °C )
D B T T s h ift (°C )
 D B T T (°C )
200
150
100
T 9 1 irra d ia te d a n d te ste d
b e tw e e n 5 0 a n d 3 0 0 °C
50
(A A A M a te ria ls H a n d b o o k
C h a p te r 1 9 )
S C K te sts E M 1 0
(T irr = 2 0 0 °C )
150
ORNL
F82H
O P T IF E R Ia
O P T IF E R IIa
M A N E T -I
M A N E T -II
EURO FER97
T 9 1 (T irr = 2 0 0 °C )
T91
O P T IF E R IIa
F82H
E97
O P T IF E R Ia
100
ORNL
50
Irra d ia tio n te m p e ra tu re : 3 0 0 °C
0
0
0
0
5
10
15
D o s e (d p a )
20
25
2
4
6
30
D o s e (d p a )
8
10
12
11
Fracture toughness
test results
T91
HT9
Embrittlement shows tendency to saturate FM steels
when irradiated in LWR type MTR
12
Irradiation under n&p:
He-effects
300
SP
250
8
6
4
CVN
Ti<380°C
F82H
T91
Optimax
Optifer-V
700
600
200
500
400
150
300
100
DBTTCVN (°C)
T91
He DBTT
(appm) (°C)
0.0 -54
265
54
450 165
DBTTSP (°C)
Absorbed Energy (J)
10
Dose
(dpa)
0.0
4.6
6.8
200
2
0
-100
50
100
0
-50
0
50
100
150
Temperature (°C)
200
250
0
2
4
6
8
10
12
14
16
18
0
20
Displacement (dpa)
13
Summary
• The existing database cannot be
rationalized to select unambiguously a
candidate material for ADS
• As long as there is no experimental facility
operating under representative conditions,
it is necessary to develop an in-depth
fundamental understanding of irradiation
damage and especially flux/spectra effects.
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Objectives
 To understand the basic mechanism of
radiation damage production and evolution in
Fe-Cr alloys
Standard materials
database
 To assess experimentally the effect of Cr –
Design conception
and
engineering
concentration on defect production and
accumulation in model alloys: how do they
compare with steels under irradiation?
 To investigate the mechanisms of irradiation
induced changes of the mechanical properties
of high Cr F/M steels
 Ultimate aim: Provide theoretical understanding
and reliable experimental database for model
validation
Dedicated
engineering
database
Fundamental
understanding
15
Multiscale computer simulation and
experimental validation of irradiation
damage in Fe-Cr based alloys
10-15 … 10-12 … 10-9
...
10-6 … 10-3
...
10-0 …
… 106 … 109
103 ...
Time scale (s):
SIMULATION
Kinetic Monte Carlo
Molecular Dynamics
Rate equations
Elasticity
Plasticity
Dislo/Defect Dynamics
Length scale (m)
10-9
Defect
production
Features of
intrinsic
irradiation
damage
10-8
10-7
Defect
evolution
Defect
accumulation
Positron
Annihilation
&
Electron
Microscopy
10-2
10-5
PhysicoChemical properties
Life time
assessment
Interactions
(Disl.-Def.)
/(Imp.-Def.)
/(Imp.-Disl.)
Mechanical
Internal
Friction
Tests
16
Approach
Experimental
 investigation of Fe, model alloys of different Cr content
and industrial steels after neutron irradiation (same
neutron flux, doses & temperature):
 I- Pure Fe and ultra-pure Fe-9Cr
 II- Industrial pure Fe and Fe-2,5,9,12 Cr
 III- Conventional and LA Ferritic martensitic steels
Theoretical
 Computer simulation of damage production in Fe-Cr binary system
 Investigation of defect mobility and interaction
 Simulation of defect accumulation kinetics using the state of the art
theoretical appraoch (LKMC, OKMC, RT,…)
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Main results
No significant influence of the
presence of Cr on:
-Collisional phase
-Number of Frenkel pairs
-Clustered fraction
0,5
0,4
0,3
0,2
1
Fe-Cr
Fe
0,4
0,3
0,2
0,1
0
10
Energy, keV
SIA clustered fraction
NRT efficiency
Fe-Cr
Fe
0,6
0,5
5
10 15 20 25 30 35 40 45 50 55
Energy, keV
0,8
0,7
V clustered fraction
Cascades in Fe and Fe10%Cr
0,7
0,6
0,5
0,4
Fe-Cr
Fe
0,3
0,2
0
5
10 15 20 25 30 35 40 45 50 55
Energy, keV
18
Main results
Principal effect of the
presence of Cr:
- High number of mixed
dumbbells
- Concentration of Cr higher in
SIA clusters than in matrix
100
Fe-Fe
Fe-Cr
Cr-Cr
80
60
40
20
0
10
20
30
40
50
Energy, keV
80
70
% Cr atoms
Number of dumbbells
Cascades in Fe and Fe10%Cr
Questions arising:
60
- How will SIA and SIA
cluster motion be affected?
50
Cr in SIA clusters
Cr in dumbbells
40
30
20
10
0
5
10
15
20
25
30
Energy, keV
35
40
45
- What the effect of this will
be on the long-term
evolution?
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Single SIA vs Cr
concentration
(preliminary)
 Low concentration: pure
trapping effect, at low T SIA
are trapped at Cr atoms and
diffusivity is reduced; effect
disappears at high T
1E-4
SIA
2
D , cm /sec
1E-3
Fe
0.2% Cr
7% Cr
12% Cr
1E-5
 High concentration: “jumping
from Cr to Cr” the SIA
reduces the binding energy to
Cr atoms to an effective
value, lower than for low
concentration: only slight
reduction of diffusivity
1E-6
5
10
15
20
25
30
35
40
Emig(Fe)/Emig(Fe-Cr)
1/kT, 1/eV
1,00
0,95
0,90
0,85
0,80
0
F. Garner et al. JNM 276 (2000) 123
2
4
6
% Cr
8
10
12
 Most effective diffusivity
reduction for 7% Cr (with this
potential …)
20
General conclusions
• Fast flux irradiation have shown that Fe-9%Cr based F/M
steels of the best candidates for future reactors
• Computer simulations demonstrate that Cr does not affect
the cascade efficiency but changes drastically the mobility
of defect clusters
• Low flux, low dose irradiation at BR2 do not show any
considerable effect of Cr on either defect density and
mechanical properties.
• Spectra / flux effects are still open issues on qualifying the
selected materials
21
Acknowledgements
L. Malerba
E. Lucon
M. Matjasevic
D. Terentyev
H. Ait Abderrahim
(MYRRHA-Team)
EU-FP5-SPIRE
EFDA-TTMS-007
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