MATERIAL ISSUES FOR ADS: MYRRHA-PROJECT A. Almazouzi SCKCEN, 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 Proton Accelerator Subcritical Neutron Multiplier Spallation Source Proton Source Neutron Source 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 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

4 Neutronic Design constraints Fast neutron spectra High energy flux High dose accumulation 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 ( n/cm 2.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 materials database Design conception and engineering Dedicated engineering database Fundamental understanding

7 MYRRHA materials Proton beam line Bending magnet Secondary coolant Subcritical core – T91 Reactor vessel – SS316L Spallation target – T91 Main heat exchangers Diaphragm Heat exchanger Main primary circulation pump Spallation loop pump Secondary coolant

8 F/M Steels FFTF-database

9 Materials: EM10, T91, HT9 FP5-SPIRE Irradiated material 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 yield stress (hardening): high/low flux BR2/HFR dataBOR60-data Courtesy: SPIRE-program Yamamoto et al. 2004

11 Impact test results

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

14 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.

15  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 materials database Design conception and engineering Dedicated engineering database Fundamental understanding

16 EXPERIMENT SIMULATION 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 … … … Time scale (s): Molecular Dynamics Kinetic Monte Carlo Dislo/Defect Dynamics Rate equations Elasticity Plasticity Length scale (m) Defect production Defect accumulation Life time assessment Interactions (Disl.-Def.) /(Imp.-Def.) /(Imp.-Disl.) Defect evolution Physico- Chemical properties Internal Friction Mechanical Tests

17 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

18 Cascades in Fe and Fe- 10%Cr Main results No significant influence of the presence of Cr on: -Collisional phase -Number of Frenkel pairs -Clustered fraction

19 Cascades in Fe and Fe- 10%Cr Main results Principal 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?

20 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  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

21 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

22 Acknowledgements L. Malerba E. Lucon M. Matjasevic D. Terentyev H. Ait Abderrahim (MYRRHA-Team) EU-FP5-SPIRE EFDA-TTMS-007