Outlook on Gen IV Nuclear Systems and related Materials R&D Challenges

1. Outlook on Gen IV Nuclear Systems and related Materials R&D Challenges • Goals for innovative reactor systems- Requirements for structural materials: generic and specific- Synergies, crosscutting R&D areas and modelling- Significance of international collaboration • Frank Carré and Pascal Yvon • CEA – Nuclear Energy Divisionfranck.carre@cea.fr

2. Sustainable Nuclear Energy Technology Platform (SNE-TP) SNE-TP Objectives & Organization Kick-off meeting : September 21, 2007

3. Generation IVNuclear Systems Argentina Systems marketable from 2040onwards USA Brazil United Kingdom A closed fuel cycle Canada True potential for new Applications:Hydrogen,Syn-fuel, Desalinated water, Process heat Russia EU China Switzerland France Japan South Africa Internationally shared R&D New goals for sustainable nuclear energy • Break-throughs: • Natural resources conservation • Waste minimisation • Proliferation resistance • Continuous progress: • Economically competitive • Safe and reliable Membersof the Generation IVInternationalForum South Korea

4. Gas Fast Reactor Molten Salt Reactor Supercritical Water-cooled Reactor Sodium Fast Reactor Very High Temperature Reactor Generation IV Forum: selection of six nuclear systems Lead Fast Reactor (12-20y) R&D (~1 B€) before a 1st prototype or techno demo

5. Fast Reactors & recycling for Sustainable Nuclear Power Strategies for a flexible management of actinides in Gen IV fast neutron systems. • Natural resources conservation • Waste minimisation • Proliferation resistance (intl standards) • Type of nuclear materials • Detection, technical difficulty, cost, time… Udep Udep Udep R T FP R T FP R T FP MA MA U Pu U Pu U Pu MA U & Pu Recycling Homogeneous Recycling Heterogeneous Recycling  Implementation depending on international standards and national optimization criteria (economics & waste).

6. Requirements for materials in future nuclear systems (1/2) Technical challenges & Leading physical phenomena • 60-year lifetime • Fast neutron damage (fuel and core materials)  Effect of irradiation on microstructure, phase instability, precipitation Swelling growth, hardening, embrittlement  Effect on tensile properties (yield strength, UTS, elongation…)  Irradiation creep and creep rupture properties  Hydrogen and helium embrittlement • High temperature resistance(SFR > 550°C, V/HTR > 850-950°C)  Effect on tensile properties (yield strength, UTS, elongation…) High temperature embrittlement  Effect on creep rupture properties Creep fatigue interaction  Fracture toughness • Corrosion resistance(primary coolant, power conversion, H2 production) Corrosion and stress-corrosion cracking (IGSCC, IASCC, hydrogen cracking & chemical compatibility…)

7. Requirements for materials in future nuclear systems (2/2) Additional requirements • Material availability and cost • Fabricability, joining technology • In service inspection  Non destructive examination techniques • Safety approach and licensing Codes and design methods R&D effort needed to establish or complement mechanical design rules and standards • Decommissioning and waste management

8. Structural materials for Innovative Reactor Systems

9. U.S.A. 2007 + Russia China France South Korea Japan Sodium Fast Reactor(SFR) • A new generation of sodium cooledFast Reactors • Reduced investment cost Simplified design, system innovations(Pool/Loop design, ISIR – SC CO2 PCS) • Towards more passive safety features+ Better managt of severe accidents • Integral recycling of actinidesRemote fabrication of TRU fuel SFR Steering Committee  2009: Feasibility – 2015: Performance  2020+ : Demo SFR (FR, US, JP…) Euratom countries

10. New materials for sodium fast reactors (1/2) SFR Primary system • New 9-12%Cr F/M steel vs Advanced Austenitic • Good physical and thermal properties, dilatation, low cost • Better creep resistance(T91, T92 (Fe-9Cr-xW-V-N…)) • Compactness, mass reduction of components DBTT but Improved toughness  Weldability (%Cr dependent) • Good compatibility with sodium impurities (C, O, N) (Demonstrated in Phénix 2ry system & Steam generator + 150 000 h Irradiation experiments of T91 & ODS (SuperNova)) • Compact component and system designs (piping, IHX…) • Potential margin for temperature increase (< 600°C)(especially if using a gas turbine power conversion system) • Allowable departure from the negligible creep regime?

11. Sodium Fast Reactor structural materials: F/M Steels Great stability of fracture properties 9% Cr Martensitic steels J.L. Séran, A. Alamo, A. Maillard, H. Touron, J.C. Brachet, P. Dubuisson, O. Rabouille J. Nucl. Mater. 212-215 (1994) 588-593.

12. New materials for sodium fast reactors (2/2) Advanced fuel cladding • 316 Ti  15-15 Ti  F/M ODS • Reduced swelling with neutron fluence • EM10 & 15-15 Ti  100 dpa @ 400-700°C • T92/HC & ODS  200 dpa @ 480 – 750/800°C • Weldability & joining techniques • Good compatibility with sodium impurities (C, O, N) • Increased fuel burnup  200 GWd/t & 200 dpa • Increased safety with low sodium content in the core & low sodium void effect  Better prevention of severe accidents • Research in progress on hardening of F/M steels with micro/nano structures (dispersion / precipitates) • 2nd generation ODS • F/M steels with carbo/nitride precipitates

13. Swelling of advanced austenitic steels and ferrito-martensitic steels used as fuel cladding in Phenix Sodium Fast Reactor cladding material

14. Safety enhancement of Fast Reactor core • Low reactivity sodium void effect high BU core • Large diameter fuel pin with thin spacer wire • ODS cladding for low swelling (Experiments in Phenix (Supernova, Matrix1&2) + in Joyo) COEX COCA • MOX fuel fabrica-tion from co-precipita-ted UPu solution from the COEX process To be first tested in Phenix (Copix expt) • Various recycling modes of minor actinides in Fast Reactors: Homogeneous (~2% MA): GACID  Heterogeneous in blanket (10-20% MA): • Curios, Amboine2-Joyo expts

15. U.S.A. Switzerland France Japan Generation IVGas Fast Reactor (GFR) • A novel type of Gas-cooled Fast Reactor:  an alternative to the Sodium Fast Reactor, and a sustainable version of the VHTR • Robust heat resisting fuel (<1600°C) • 1200 MWe – THe ~ 850 °C - Cogeneration of electricity, H2, synfuel, process heat • Safe management of cooling accidents • Potential for integral recycling of Actinides GFR Steering Committee  2012 : Feasibility  ~2020 : ETDR (EU ?) 2020: Performance  2030+: GFR Prototype GCFR 5-6 EU PCRD Euratom countries • System Arrangement GFR signed Nov. 30 Nov.,2006 • Project Arrangements “Fuel “ & “Design-Safety-Integration” in 2007

16. Gas Fast Reactor fuel designs High density compartmented platelet Advanced particles Cladded pellets HTRs 25 50 0 75 100 %vol. of actinides compound in the volume dedicated to fuel

17. Fibre strenthened • Multi-layer materials TiC (HIP) trans-granulaire inter-granulaire 10 µm 10 µm Candidate ceramics materials for the GFR fuel Ceramics for Gas Fast Reactor Composite CERMET Matrix & concept in Phénix • Mixed CER & MET matrices Interfaces  Investigation and modelling of phenomena • Manufacturing and testing monolithic and composite ceramics(C/C, SiC/SiC) Usual low toughnessof ceramics • Characterization and optimization Objectives: Increase ceramics ductility and toughness

18. Goal: 3m (length) x 10mm (inner diameter) x 1mm (wall thickness) 2D SiC/SiC by NITE Process for GFR Fuel Pin or Plate Nite Process Kyoto University Fuel Pin 5.0mm Fuel Plate 43.0mm Wall thickness: 1.0mm

19. A few specific R&D areas on ceramics for GFR fuels Possible applicationsas matrix or interphase in SiC/SiCf composite… MAX Phases Nano-laminate structureTi3SiC2 Ti, C, Si • Special properties • Damage tolerant • Low density, machinable • High thermal and • electrical conductivities • Methods used to obtain large-scale bulk Ti3SiC2 • CVD, Arc melting, HIP, HP, SHS • High energy ball milling & reactive sintering to obtain bulk Ti3SiC2 with very fine grain • Synthesis of TiC from nano-powder • HIP without grain growth • Stable under irradiation (electrons & ions)

20. U.S.A. China Switzerland France 2007 + South Korea Japan South Africa Generation IVVery High Temperature Reactor (V/HTR) VHTR Steering Committee • A nuclear system dedicated to the production of high temperature process heat for the industry and hydrogen • 600 MWth -THe >1000 °CThermal neutronsBlock or pebble core concept • Passive safety features • I-S Cycle or HT Electrolysis for H2  2009: Feasibility – 2015: Performance ~ 2020: PBMR, NGNP & Other Near Term Projects Euratom

21. VHTR Vessel PWR Vessel 9Cr1Mo alloy for pressure vessel of gas cooled reactors VHTR vs PWR pressure vessel manufacturing techniques • Normal/off-normal service temperatures and vessel size dominate materials requirements Up to <450/550°C at 5-9 MPa Up to 1 x 1019 n/cm2 fluence • Very large vessel sizes require scale-up of ring forging & on-site joining technologies • Irradiation resistance to be demonstrated for licensing

22. 500 mm 60 mm 100 mm VHTR core material: Graphite & Composites • Graphite (PCEA (UCAR), NBG 17 (SGL)…) • Characterization:chemical, structural, thermal & mechanical properties (20-1000°C), corrosion tests (air; water, O2, CO2…) • Irradiation tests(T ~1050°C, 1-6 dpa G) • Optimization for waste minimization(14C) • Technical file for codification of design standards • C/C & SiCf/SiC composites  Manufacturing:2D & 3D woven fibres (C, Hi-Nicalon S), interphases, CVI or pitch densification, anti-oxidation coating (Si, B)…)  Characterization:chemical, structural, thermal & mechanical properties (20-1000°C), corrosion (air,water, O2, CO2), irradiation tests  Technical file for codification of design standards

23. VHTR Intermediate Heat Exchanger • Three IHX technologies identified: • Plate-machined Heat Exchanger (Fig. 1) • Plate-Fin Heat Exchanger (Fig. 2) • Tubular concept • Key issues to be addressed:  Materials development - Haynes 230 - Inconel 617 - Ni-ODS  Intermediate Heat Exchanger design - Compactness - High thermomechanical resistance - High thermal efficiency (95%) - Low pressure drop, no leakage  Properties required at 850°C - 950°C - Tensile, long term creep (Fig. 3), fatigue,creep-fatigue - Corrosion resistance - Fabrication and joining techniques FIG. 1 FIG. 2 FIG. 3 Creep strength: Haynes 230 and Inconel 617

24. Generation IV Systems related R&D Needs Structural materials & Components

25. Fusion Power Reactor Dual-Coolant T-Blanket He, 80 bars Pb-17Li, ~ bar 0 0 300, 480 C 480-700 C Dual-Coolant T-Blanket 0 Martensitic Steels (550 C) 0 ODS Ferritic steels (700 C) SiCf-SiC th. & elect. insulator 0 F W: T max= 625 C 0 Channel : Tmax= 500 C 0 Insert : Tmax~1000 C Typical Tokamak Configuration T-Breeding Blanket:Dual Coolant Lithium Lead Design and technology of T- breeding blanket

26. Innovative Reactor Systems & Requirements for Structural Materials Summary (1/2) • Materials science and new materials are key for optimizing 2nd & 3rd generation LWRs as well as to meet 4th nuclear systems’ objectives : • > 2040: Fast reactorswith a closed fuel cycle (SFR, GFR, LFR) • ~2025-30: High temperature reactors(V/HTR)for process heat (H2…) • More prospective nuclear systems (SCWR, MSR) • Incremental progress and breakthroughs are sought on a wide span of structural materials for fuel claddings, core structures, reactor cooling systems & components (RPV, IHX, SG…), power conversion systems (electricity, H2…): • Metals: Austenitic steels, 9-12Cr F/M steels, ODS, Ni-alloys… • Ceramics & composites: Graphite, C/C, SiCf/SiC & (TiC, ZrC, Ti3SiC2…) • Fabrication, characterization, manufacturing, ND examination • Mechanical design codification: ASME, RCCM-R + extensions / harmonization needed for fast neutrons, high temperature, lifetime… • Synergies between materials for 4th generation nuclear systems as well as with materials for Fusion (1st wall & blanket) • Ni-alloys, 9-12%Cr F/M steels, ODS, Ceramics (SiC…)

27. Innovative Reactor Systems & Requirements for Structural Materials Summary (2/2) • Increased role of Materials science (analytical research and modelling) for a more predictive R&D towards aimed materials properties Metals  Ceramics  Fuels • International cooperation to increase and share R&D work and achieve breakthroughs for 21st century nuclear power systems • Federate national programs into a consistent international roadmap • Enhancing R&D and technology demonstrations (Gen IV, EU FP7…) • Databases of materials properties • Multi-scale modelling of materials & fuels • Synergies between Fission and Fusion materials • Progressing towards harmonized international standards • Mechanical design rules and standards, Codification • Safety, non-proliferation, physical protection…