Challenges of structural materials for innovative nuclear systems in Europe

Transcription

1 Challenges of structural materials for innovative nuclear systems in Europe Marta Serrano, Dolores Gomez-Briceño Structural Material Division CIEMAT Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5 9 October 2009, Fusion for Energy, Barcelona, Spain Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 1

2 Scope Introduction Description of systems Common issues High Temperature Radiation damage Compatibility Open points Summary Recommendations Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 2

3 Introduction At the beginning of 2007 the European Commission include for the first time on its Energy Policy for Europe the benefits of nuclear energy: low carbon emissions, competitiveness and stable prices The Strategic Research Agenda of Sustainable Nuclear Energy Technology Platform (SNETP) addresses the key issues of fission technologies including waste management Fast reactors: Sodium Fast Reactor (SFR) the main option and the Leadcooled Fast Reactor (LFR) and the Gascooled Fast Reactor (GFR) as alternative Accelerator Driven Systems (ADS) are also included for transmutation Very high temperature reactor, VHTR is envisage for production of nuclear heat Supercritical water reactor SCWR is considered as one option for efficiency improvement of current operation LWR Supercritical Water Reactor SCWR Very High Temperature Reactor VHTR Fast Reactor (GFR, SFR, LFR) Transmutation ADS Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 3

4 Introduction FUTURE NUCLEAR REACTORS LWR (Plant life extension) EPR SFR (EU prototype 2020) Gen IV Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 4

5 Introduction FUTURE NUCLEAR REACTORS LWR (Plant life extension) EPR SFR (EU prototype 2020) Gen IV ITER DEMO Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 5

6 Introduction FUTURE NUCLEAR REACTORS LWR (Plant life extension) EPR SFR (EU prototype 2020) Gen IV ITER DEMO Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 6

7 Description of systems SFR GFR LFR/ADS VHTR SCWR Fusion Coolant Liquid Na He Pb alloys He Water He/Pb 17Li Temperature C Dose Cladding 200 dpa dpa 100 dpa 7 25 dpa 20 dpa 100 dpa 10 ppm He/dpa 45 ppm H/dpa Core structures Wrapper F/M steel Cladding AFMA F/M ODS Fuel and core structures SiCf SiC composite Core components ODS steels Target and cladding F/M steels ODS Core graphite Control rods C/C SiC/SiC Cladding and core structures Ni-based alloys and F/M steels ODS steels First wall and blanket: F/M steels ODS SiCf SiC F. Carre SMINS 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 7

8 Description of systems SFR GFR LFR/ADS VHTR SCWR Fusion Coolant Liquid Na He Pb alloys He Water He/Pb 17Li Temperature C Dose Cladding 200 dpa dpa 100 dpa 7 25 dpa 20 dpa 100 dpa 10 ppm He/dpa 45 ppm H/dpa Core structures Wrapper F/M steel Cladding AFMA F/M ODS Fuel and core structures SiCf SiC composite Core components ODS steels Target and cladding F/M steels ODS Core graphite Control rods C/C SiC/SiC Cladding and core structures Ni-based alloys and F/M steels ODS steels First wall and blanket: F/M steels ODS SiCf SiC F. Carre SMINS 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 8

9 Description of systems SFR GFR LFR/ADS VHTR SCWR Fusion Coolant Liquid Na He Pb alloys He Water He/Pb 17Li Temperature C Dose Cladding 200 dpa dpa 100 dpa 7 25 dpa 20 dpa 100 dpa 10 ppm He/dpa 45 ppm H/dpa Core structures Wrapper F/M steel Cladding AFMA F/M ODS Fuel and core structures SiCf SiC composite Core components ODS steels Target and cladding F/M steels ODS Core graphite Control rods C/C SiC/SiC Cladding and core structures Ni-based alloys and F/M steels ODS steels First wall and blanket: F/M steels ODS SiCf SiC F. Carre SMINS 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 9

10 Description of systems SFR GFR LFR/ADS VHTR SCWR Fusion Coolant Liquid Na He Pb alloys He Water He/Pb 17Li Temperature C Dose Cladding 200 dpa dpa 100 dpa 7 25 dpa 20 dpa 100 dpa 10 ppm He/dpa 45 ppm H/dpa Core structures Wrapper F/M steel Cladding AFMA F/M ODS Fuel and core structures SiCf SiC composite Core components ODS steels Target and cladding F/M steels ODS Core graphite Control rods C/C SiC/SiC Cladding and core structures Ni-based alloys and F/M steels ODS steels First wall and blanket: F/M steels ODS SiCf SiC F. Carre SMINS 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 10

11 Common issues High temperature operation High dpa Aggressive coolant: Gas, Na, HLM Focus on F/M (inc. RAFM) steel and their ODS variants dpa at 600 o C Swelling He (ADS) RIS Cladding, Internals Creep Irradiate d 316 Coolant Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 11

12 Elevated temperature behavior Increasing Temperature 565ºC EM12 HT9 593ºC HCM12 T91 620ºC NF616 E ºC NF12 SAVE12 ODS, NFAs, Thermomechanical treatments Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 12

13 Elevated temperature behavior Increasing Temperature 565ºC EM12 HT9 593ºC HCM12 T91 620ºC NF616 E ºC NF12 SAVE12 ODS, NFAs, Thermomechanical treatments Evolution to nanostructured ferritic alloys ODS (NFAs) Nano-size grain size and high number density of Nanocluster Y-Ti-O New generation ORNL 14YWT, 14%Cr and 0.3 wt.% Yttrium. Fabrication modification: Temperature of hot extrusion Reduction in thickness per pass during hot rolling Final heat treatment. Very low transition temperature Bright-field TEM images of (a) 12YWT and (b) 14YWT microstructures (McClintock 2009, Stoller 2009) Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 13

14 Elevated temperature behavior Thermo-mechanical treatments: Austenization ( ºC) Hot roll ( ºC) Annealing + Air Cooling Klueh 2007 Strength from distribution of nano-sized MX nitride and/or carbonitride precipitates Yield Stress (MPa) N-containing 9Cr TMT 12YWT Mod 9Cr-1Mo (N&T) N-containing 9Cr TMT Strain (%) Mod 9Cr-1Mo (N&T) 650ºC 138 MPa New 9Cr-1Mo TMT Temperature (ºC) Time (h) Strength and ductility comparable to high-strength experimental ODS steel Conventional and cheaper fabrication route Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 14

15 Radiation damage at low Tirrad Radiation damage mechanisms for F/M steels depend on temperature T irrad < ºC hardening and embrittlement Higher T irrad Phase instabilities, irradiation creep, volumetric swelling. Effect of Cr content on DBTT shift Some cautions!!! Main concerns Thermomechanical treatments not optimized (Zinkle 2007) Definition of DBTT for F/M steels (Chaouadi 2007) Lack of physical meaning (Malerba 2008) He effects underlined?? Possible reason: For high Cr content precipitation of Cr-rich brittle phase at temperatures around 475ºC. Some SANS results show that for irradiation at 325ºC Cr-rich phase appears in 8%Cr in matrix (Mathon 2003) Kohyama dpa, 365ºC 7 dpa, 365ºC 36 dpa, 410ºC High Cr ODS steels do not follow this trend Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 15

16 Radiation damage at high Tirrad Swelling F/M steels has better resistance than austenitic steels Recent studies pointed out that this better resistance is due to a longer cavity incubation time of ferritic steels, and that austentic and ferritic steel have similar swelling rates. The swelling incubation time could be affected by temperature, dpa rate and gas production. This late would make the predictions of low swelling of F/M steels non-conservatives for fusion environment with a high He and H produced by transmutation Carre 2007 Kimura 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 16

17 Compability Sodium: Mature technology Helium: GFR impurities are different than VHTR Heavy liquid metals (LBE, Pb, Pb-Li): Corrosion protection In situ growth and control of a self-healing protective oxide layer on the steel surface: Pb-Li is less inert than elements in the steels so lithium oxides form preferentially than chromium or iron oxides and this method of protective oxide layer formation is not possible in fusion devices Deposition of surface coatings Supercritical water Materials undergo a weight gain instead of a weight loos as observed in water corrosion processed Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 17

18 Compabitibility 1m/s 2 m/s FeCrAlY alloy coating (GESA) -T91 tubes exposed to LBE at 600 C and to different flow rates No liquid metal attack No influence of the flow velocity 3 m/s ODS F/M steels LBE: Effect of Cr and Al Kimura 2007 (a) Weisenburger 2008 (b) Smaller grains (14YWT) lead to a more homogeneous oxidation while larger grains (MA957) lead to grain boundary oxidation Sienicki 2006 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 18

19 Compabitibility SCWR corrosion SCW 510ºC Heikinheimo 2007 Kimura 2007 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 19

20 Compabitibility Pb-17Li In general, ferritic martensitic and RAFM steels present corrosion rates in Pb-Li higher than acceptable values (20 m/year) at operating conditions. Preliminary tests indicate that there is not a significant effect of the magnetic field on corrosion and deposition for austenitic steels but it seems to have an influence on the type and place of deposition for martensitic steels Konys 2004 Eurofer 97 Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 20

21 OPEN POINTS Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 21

22 Open points Cyclic softening It is not a new issue - Reported in the 80 Decrease in yield point and reduction in strain hardening during cyclic deformation are responsible for cyclic softening Cyclic softening is a design issue at elevated temperature, which reduces the design margin significantly. Transferability to plant?? Li 2007 JFL-1 600ºC Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 22

23 Open points RIS and RIP It is generally accepted that for irradiation temperatures above ºC no embrittlement occurs for F/M steels. In a recent study of F82H, a non hardening embrittlement was identified after irradiation temperature 500ºC at 5 and 20 dpa (Klueh 2009) These observations were attributed to Lavesphases precipitation and M23C6 coarsening that occurring during irradiation Radiation Induced Segregation (RIS) can lead to local concentrations that exceed solubility limits which leads to radiationinduced precipitation (RIP). Trends in RIS of F/M steels are unclear and more analysis and RIS data is needed to properly evaluate this phenomena Klueh 2009 DBTT shift 30ºC Non-hardening The use of computational thermodynamics calculations can predict the amounts of precipitates expected and can also be used to design optimal composition to minimize the effects of precipitations on mechanical properties Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 23

24 He effect Open points Transmutated He on structural material of fusion reactor leads to a loss of hightemperature creep strength, increased swelling and irradiation creep at intermediate temperatures and a potential loss of ductility and fracture toughness at low temperatures Helium effects at low temperature are still a matter of controversy. Recent results suggest a synergistic interaction between high levels of He (> appm), leading to a grain boundary weakening, combined with large hardening for irradiation temperatures bellow 400ºC. These combined mechanisms may results in the occurrence of intergranular fracture and a big DBTT shift DBTT/ Yamamoto 2006 Henry 2007 T91 specimens after implantation at 250 and testing at 25 C Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 24

25 Open points Compatibility Modelisation of oxidation/corrosion in HLM - How to transfer to Pb, higher temperatures and ODS Data on ODS + HLM lacking Cr content Al Compromise Corrosion Aging embrittlement SCWR water chemistry and radiolysis Confirmation of the effect of magnetic field on corrosion and deposition on Fusion devices Effect of HLM on mechanical properties Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 25

26 Summary and recomendations Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 26

27 Summary Classical F/M steels (9Cr and 12Cr) are well characterized in the fossil community, but some open issues exists: Cyclic softening Irradiation at high temperature RIS, RIP, Non-hardening embrittlement Helium effects Effect of HLM on mechanical properties ODS steels are promising in terms of high temperature strength, radiation resistance and compatibility, but there are some issues to be solved: Thermal stability of oxide particles Irradiation effects High variability between heats Compatibility with HLM Large-scale production of high quality ODS steel must be developed. No EU industrial steel makers Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 27

28 Recommendations Ferritic/Martensitic steels - Besides open issues, qualification is in a short term Industrial partners are needed to lead qualification ISI methods should be developed ODS Less developed. Qualification is a long way Lab-scale optimization of fabrication processes Characterization including irradiation Joining and welding An earlier connection with industry would accelerate the qualification process Simulation tools will help to Optimization of material design Understanding of degradation mechanisms for 60 years of operation Take advantage from non-nuclear communities High temperature materials Joint EC-IAEA Topical Meeting on Development of New Structural Materials for Advanced Fission and Fusion Reactor Systems 5-9 October 2009, Fusion for Energy, Barcelona, Spain 28