MATERIAL ISSUES FOR ADS:MYRRHA-PROJECT
A. Almazouzi
SCKCEN, Mol (Belgium)
On behalf of MYRRHA-TEAM and MYRRHA-Support
2
MYRRHA – concept: a multipurpose ADS
• Material testing• Fuel irradiation• MA & LLFP transmutation• Radioisotope production• Neutron beams
ProtonAccelerator
SubcriticalNeutron
Multiplier
SpallationSource
ProtonSource
NeutronSource
• Windowless design• Pb-Bi technology• Spallation products• MA transmutation
• Material testing• Radioisotope production• Proton therapy
Multipurpose hYbride Research Reactor for High-tech Applications
3
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 SCKCEN 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
4
Neutronic Design constraints
Fast neutron spectra
High energy flux High dose accumulation
1
10
100
1000
0 3 6 9 12 15
Z- axial position (cm)
appm
of
gas
/ dpa
appm H / dpa
appm He / dpa
Important gaz production rate
5
Design constraints
•High thermal conductivity, heat resistance, low thermal expansion
•Low DBTT shift, sufficient strength, limited loss of ductility and fracture toughness, low swelling rate
•Adequate resistance to He and H embrittlement
•Resistance to fatigue in LBE
•High creep and fatigue resistance
•Corrosion and liquid metal embrittlement resistance
•Temperature (200 to 450°C)
•High dose rate /high dose (5.1015n/cm2.s and up to 40dpa/year)
•High He/dpa production rate (3 to 8 appm)
•Beam trips and loading/reloading operations
•High stress level (100 MPa), operational trips
•Aggressive conditions (LBE)
Properties neededOperating conditions
6
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.
•Research to build the ability to interpolate and extrapolate the results from laboratory tests to the real life.
Standard materialsdatabase
Design conceptionand
engineering
Dedicated engineering
database
Fundamental understanding
7
MYRRHA materials
Proton beam lineBending magnet
Secondary coolant
Subcritical core – T91
Reactor vessel – SS316LSpallation target – T91
Main heat exchangers
Diaphragm
Heat exchangerMain primarycirculation pump
Spallation loop pump
Secondary coolant
9
• Materials: EM10, T91, HT9
FP5-SPIRE Irradiated material
C Ni Cr Mo Cu Si S Al Nb Co V EM10 0.099 0.07 8.97 1.06 0.05 0.46 <0.003 <0.016 <0.002 0.03 0.013 T91 0.099 0.24 8.8 0.96 0.05 0.32 0.004 <0.01 0.06 0.03 0.24 HT9 0.204 0.66 11.68 1.06 0.45 <0.003 0.03 0.29
Ti N P Mn O B W Sn As Sb Fe EM10 0.01 0.014 0.013 0.49 0.001 <0.001 <0.002 <0.005 0.003 0.01 bal. T91 <0.005 0.03 0.02 0.43 <0.0005 <0.01 0.006 0.011 0.012 bal. HT9 0.020 0.63 0.47 bal.
T91 - Normalised at 1040°C/60’- Tempered at 760°C/60’
EM10 - Normalised at 990°C/50’- Tempered at 750°C/60’
HT9- Normalised at 1050°C/30’- Tempered at 700°C/120’
10
Irradiation effects on the yieldstress (hardening): high/low flux
0
100
200
300
400
500
600
0 2 4 6 8 10 12
Dose (dpa)
Yie
ld s
tren
gth
incr
ease
(M
Pa)
EUROFER97 [BR2 - T = 300 °C]
EUROFER97 [HFR - T = 300 °C]
F82H [HFR - T = 300 °C]
T91 [BR2 - T = 200 °C]
EM10 [BR2 - T = 200 °C]
Fitting curve EUROFER97 data
Test temperature = Irradiation temperature
BR2/HFR data BOR60-data
Courtesy: SPIRE-program
Yamamoto et al. 2004
11
0
1
2
3
4
5
6
7
8
-200 -150 -100 -50 0 50 100 150 200 250 300
Temperature (°C)
Ab
sorb
ed e
ner
gy
KV
(J)
Unirradiated T-L
2.43 dpa, 200 °C
3.58 dpa, 200 °C109 °C
7.3%
131 °C
6.5%
Impact test results
0
1
2
3
4
5
6
-150 -100 -50 0 50 100 150 200 250 300
Temperature (°C)
Ab
sorb
ed e
ner
gy
KV
(J)
Unirradiated (SCK)
SCK - 2.47 dpa, 200 °C
SCK - 3.70 dpa, 200 °C
169 °C
41%
163 °C
37%
0
50
100
150
200
250
0 5 10 15 20 25 30
Dose (dpa)
D
BT
T (
°C)
T91 irradiated and testedbetween 50 and 300 °C
(AAA Materials HandbookChapter 19)
SCK tests T91(Tirr = 200 °C)
SCK tests EM10(Tirr = 200 °C)
0
50
100
150
200
250
0 2 4 6 8 10 12
Dose (dpa)
DB
TT
sh
ift
(°C
)
ORNL F82H
OPTIFER Ia OPTIFER IIa
MANET-I MANET-II
EUROFER97 T91 (Tirr = 200 °C)
Irradiation temperature: 300 °C
E97
F82HT91
MANET-II
MANET-I
OPTIFER Ia
OPTIFER IIa
ORNL
12
Fracture toughness test results
Embrittlement shows tendency to saturate FM steels when irradiated in LWR type MTR
T91 HT9
13
Irradiation under n&p: He-effects
-100 -50 0 50 100 150 200 2500
2
4
6
8
10T91
Dose He DBTT (dpa) (appm) (°C)
0.0 0.0 -54 4.6 265 54 6.8 450 165
Ab
sorb
ed
En
erg
y (J
)
Temperature (°C)
0 2 4 6 8 10 12 14 16 18 200
50
100
150
200
250
300
DB
TT
CV
N (
°C)
SP
DB
TT
SP (
°C)
Displacement (dpa)
0
100
200
300
400
500
600
700Ti<380°C
F82H T91 Optimax Optifer-V
CVN
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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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To understand the basic mechanism of radiation damage production and evolution in Fe-Cr alloys
To assess experimentally the effect of Cr –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
Objectives
Standard materialsdatabase
Design conceptionand
engineering
Dedicated engineering
database
Fundamental understanding
16
EXPER
IMENT
EXPER
IMENT
SIMULATIONSIMULATION
Multiscale computer simulation and experimental validation of irradiation
damage in Fe-Cr based alloys
Features of intrinsic
irradiation damage
Positron Annihilation
&Electron
Microscopy
... 10-6 … 10-3 ... 10-0 … 103 ... … 106 … 109 10-15 … 10-12 … 10-9
Time scale (s):
Molecular Dynamics Kinetic Monte Carlo
Dislo/Defect Dynamics
Rate equationsElasticityPlasticity
10-9 10-7 10-5 10-2
Length scale (m)
10-8
Defect production
Defect accumulation
Life time assessment
Interactions (Disl.-Def.) /(Imp.-Def.) /(Imp.-Disl.)
Defect evolution
Physico-Chemical properties
Internal Friction
Mechanical
Tests
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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,…)
Approach
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Cascades in Fe and Fe-10%Cr
0 5 10 15 20 25 30 35 40 45 50 55
0,2
0,3
0,4
0,5
0,6
0,7
SIA
clu
ster
ed f
ract
ion
Energy, keV
Fe-Cr Fe
0 5 10 15 20 25 30 35 40 45 50 55
0,1
0,2
0,3
0,4
0,5
V c
lust
ered
fra
ctio
n
Energy, keV
Fe-Cr Fe
1 10
0,2
0,3
0,4
0,5
0,6
0,7
0,8
NR
T e
ffic
ien
cy
Energy, keV
Fe-Cr Fe
Main resultsNo significant influence of the presence of Cr on:
-Collisional phase
-Number of Frenkel pairs
-Clustered fraction
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Cascades in Fe and Fe-10%Cr
10 20 30 40 500
20
40
60
80
100
Nu
mb
er o
f d
um
bb
ells
Energy, keV
Fe-Fe Fe-Cr Cr-Cr
Main resultsPrincipal effect of the presence of Cr:
- High number of mixed dumbbells
- Concentration of Cr higher in SIA clusters than in matrix
Questions arising:
- How will SIA and SIA cluster motion be affected?
- What the effect of this will be on the long-term evolution?
0 5 10 15 20 25 30 35 40 4510
20
30
40
50
60
70
80
% C
r at
om
s
Energy, keV
Cr in SIA clusters Cr in dumbbells
20
Single SIA vs Cr concentration
5 10 15 20 25 30 35 401E-6
1E-5
1E-4
1E-3
1/kT, 1/eV
DS
IA, c
m2 /s
ec
Fe 0.2% Cr 7% Cr 12% Cr
(preliminary) Low concentration: pure
trapping effect, at low T SIA are trapped at Cr atoms and diffusivity is reduced; effect disappears at high T
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
Most effective diffusivity reduction for 7% Cr (with this potential …)
F. Garner et al. JNM 276 (2000) 123
0 2 4 6 8 10 120,80
0,85
0,90
0,95
1,00
Em
ig(F
e)/E
mig(F
e-C
r)
% Cr
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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
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