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

8

F/M SteelsFFTF-database

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

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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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Acknowledgements

L. MalerbaE. LuconM. MatjasevicD. Terentyev

H. Ait Abderrahim

(MYRRHA-Team)

EU-FP5-SPIRE

EFDA-TTMS-007