Small Modular Reactor Materials R&D Program Materials Coordination Webinar William Corwin Office of Advanced Reactor Technologies U.S. Department of Energy August 2012
SMRs Are Strong Contenders to Augment Current Nuclear Power Capacity At 300 MWe, SMRs can provide flexible energy sources for multiple purposes at reduced financial risk Multiple modules can be licensed as a single plant Allow for sequenced construction and more acceptable cash flow Have strong potential for enhanced safety Can be sited in non-conventional locations and meet non-conventional missions Provide classic nuclear power benefits of low-fuel costs without significant carbon emissions
Different Types of SMRs Offer a Range of Advantages Integrated light water reactors (LWRs) may be deployed first and address nuclear industry inertia and licensing issues NuScale 45 MWe NuScalePower mpower 125 MWe Babcock & Wilcox IRIS 300 MWe Westinghouse
Advanced SMR Designs Are Being Developed for Additional Benefits New innovative technologies Liquid metal, gas and molten saltcooled designs Transformative reactor concepts Goals: Safer, simpler, extended fuel life, proliferation resistant reactors Broader applications Power production Process heat applications Oil refining and extraction Desalination Chemical and hydrogen production Biofuel & Synfuel production Support fuel-cycle options PRISM ANTARES Toshiba 4S
Different Types of SMRs Offer a Range of Advantages Sodium cooled fast reactors (SFRs) offer higher output temperatures and fast neutron spectra 4S 10 MWe Toshiba PRISM 311 MWe GE Hitachi
Different Types of SMRs Offer a Range of Advantages (cont) High temperature gas cooled reactors (HTGRs and VHTRs) offer still higher output temperatures NGNP 150-200 MWe ANTERES AREVA
Different Types of SMRs Offer a Range of Advantages (cont) Advanced (transformational) reactor concepts may offer longer lives, higher output temperatures, greater safety, better economics, etc. Gen4 Energy Lead bismuth cooled 500 C ROT 25-30 MWe AHTR Liquid salt cooled 700 C ROT 100-400 MWe
Structural Materials Are Critical for All Advanced Nuclear Reactor Technologies Development and qualification of advanced structural materials are critical to the design and deployment of the advanced nuclear reactor systems that DOE is developing High Temperature Gas Cooled Reactors (HTGRs) Sodium Cooled Fast Reactors (SFRs) Fluoride Salt Cooled High Temperature Reactors (FHRs) Lead and Lead-Bismuth Cooled Fast Reactors (LFRs) Structural materials must perform over design lifetimes for pressure boundaries, reactor internals, heat transfer components, etc. Performance of structural materials for high temperatures and radiation exposures associated with advanced reactor system development is supported by three separate reactor programs 8
NGNP, ARC, FCRD and SMR Programs Include Structural Materials R&D Activities Next Generation Nuclear Plant (NGNP) performs fuels & materials qualification and reactor systems R&D for high-temperature gas cooled reactors Development and qualification of graphite, high temperature metals, and composite materials for pressure boundaries, internals, and heat transfer equipment Advanced Reactor Concepts (ARC) and Fuel Cycle R&D Programs perform R&D on fundamental nuclear technologies that enable new uses of nuclear energy Improvement of ferritic-martensitic and austenic alloys and materials compatibility for SFR, FHR, and LFR systems Small Modular Reactor (SMR) supports smaller, advanced designs with longer-term licensing horizons through focused R&D High temp design methodology and ceramic composite codes & standards development and materials compatibility for LFR systems 9
Resolution of Materials Issues for HTGRs and VHTRs Is in Progress Qualification of currently available nuclear-grade graphites for core structures Strength, toughness & degradation of metals at high temperatures for IHXs (Intermediate Heat Exchangers), SGs (Steam Generators) & internals, as well as RPVs (Reactor Pressure Vessels), piping, and IHX & SG shells Fabrication, standard testing & NDE methods, irradiationeffects, and design codes for ceramics and composites for control rods, reactor internals, duct liners, etc. High Temperature Metals and Composites R&D, Plus Codes and Standards Development, Relevant to Multiple SMR Systems Now Being Addressed within SMR Program
High-Temperature Metals Usage in HTGR and VHTR Service Must Be Addressed High-temperature mechanical properties & environmental degradation processes in air & impure helium environments High-temperature metallurgical stability (thermal aging effects) Extension of ASME and similar design Code approval for metallic materials for higher HTGR & VHTR operating temperatures, (>800 C) longer service times, and complex loading conditions Validated methodologies for inelastic design analysis
Structural Composites Are Being Developed & Qualified for SMR Components SiC-SiC and C-C are candidates for non-metallic control rods C-C composites also evaluated for structural reactor internals applications at lower doses, as well as thermal duct and insulation covers Composite Boron Carbide Compact Control Rod Core Tie Rod Lateral Core Restraint Rings
Economic & Regulatory Concerns Dominate SFR Materials Needs Development and qualification of high-strength, advanced material alloys for structures and cladding to enable more affordable systems and longer life operation Ensure adequate environmental test facilities are available Address unresolved issues identified in existing reviews of high-temperature materials and design methodology needs performed over the past 25 years by NRC/ACRS/DOE for SFRs and HTGRs These issues are applicable to other candidate hightemperature SMRs
New Developments in High-Temperature Reactors Have Revitalized ASME Code Activities for Inelastic Materials & Designs Roadmaps for sodium- and gas-cooled reactor materials and design code needs have been developed and ISI needs for advanced reactors are being actively evaluated DOE s programs have funded numerous tasks to address identified Code deficiencies Entirely new Division 5 of Section III has been issued to incorporate Sec III NH & Code cases, plus rules for graphite and ceramics, to address all high-temperature reactor needs All high-temperature SMRs have materials, design, and ISI issues that need the updated ASME Code rules
Continued ASME Code Improvements for Advanced High-Temperature Reactors Are Essential for Design and Licensing Sec III Div 5 Graphite and composite issues Creep-fatigue, negligible creep, and weldment rules Incorporate recommendations from ASME-LLC tasks Simplified and all-temperature rules Environmental rule considerations (corrosion, irradiation, graphite, etc.) Extension of allowables for current and new materials Alloy 800H, 617, improved SS (316LN/FR, HT-UPS), modified grade 92, Hastelloy N, etc. Graphites and Composites Sec XI More reliance on advanced techniques (UT for volumetric exam, AE for crack or leak monitoring, phased arrays for micro-cracking, etc.) Creep-crack growth evaluation procedures Rules for compact heat exchangers (joint with Sec III)
SMRs Will Require a Re-examination and Likely Augmentation of Additional Existing ASME Nuclear Codes Construction Modularization Factory fabrication vs on-site assembly New materials and processes Modified stress allowables (time, temp, loading histories, environments, etc.) Pre-service and In-service inspection technologies Geometries and accessabilty New components (IHX, CHE, etc.) New coolants and heat transfer media (helium, sodium, liquid salt, etc.) Operating temperature and outage schedules Potential effects of multi-modules, staged deployments, remote siting, and other issues TBD
Examples of Active Code Issues: Negligible Creep Limits for RPV Materials Are Not Fully Defined Negligible creep is a condition where creep does not impact cyclic performance In LWR RPV design, no timedependent deformation is considered, hence no creepfatigue interactions are required For advanced reactors, RPVs will operate for longer times (60-year design) and possibly both higher operating and excursion temperatures, hence negligible creep is a potential concern HTGR RPV 17
Examples of Active Code Issues: More Accurate Predictions of Creep Fatigue Interaction Are Needed Understanding creep-fatigue interactions are critical for design and safety analysis of components in high temperature reactors operating under inelastic conditions of time, temperature, and loading Cumulative damage from creep and fatigue must be understood and predicted Many service applications generate more creep damage than fatigue damage, but most experimental CF data includes more fatigue damage Relative amounts of damage from either mechanism can affect the metallurgical failure processes as well as the numerical analysis correlations of the results AFR-100 Upper Internals 18
Even Limited Materials Issues for Small LWRs (ipwrs) Are Being Identified Evaluations of changes in materials needs from large LWRs as a function of specific designs Can use ipwrs to evaluate: Effects of geometry and outage intervals on inspection accessibility for PSI & ISI Potential for IGSCC in thin-walled tubes under compressive stress mpower for SGs Introduction and qualification of improved materials for longer life, better economics, etc., will be a continuing need
SMR Technologies Provide a Wide Range of Potential Benefits Safety Benefits Passive decay heat removal in most designs Smaller source term inventory - postulated accidents result in less public dose Simplified design eliminates/mitigates several postulated accidents - Loss of Coolant and Core Damage Accidents Below grade reactor siting Possible reduction in Emergency Planning Zone co-location with industrial applications may be viable Economic Benefits Reduced financial risk - smaller upfront investment Flexibility - modularity allows additional units to be added to meet increased demand Performing asset - operating units provide financing for future additional units Domestic forgings and manufacturing - revitalize nuclear industry in U.S. Job creation - domestic manufacturing, construction, and operation
Work Packages Addressing Five Major Materials Areas Will Be Active in the SMR Program During FY13 Composites Technology High dose SiC-SiC irradiation, SMR environmental effects, and Codes & Standards development High Temperature Design Methodology Develop creep-fatigue and weldment design methodologies, and methods to extrapolate design data to 60-year design life into ASME code Generate supplementary data development and prepare draft Code Case for Alloy 617 for nuclear application Design and Codification Basis for SMR Materials Develop and maintain web-based database on Gen IV reactor materials Materials Issues for SMR Operational Environments Assess specialized materlals issues for SMR-specific geometries and operating conditions, e.g., IGSCC under compressive loading in internal steam generators Materials and Component Studies for LBE SMR Concepts Examine corrosion and erosion of structural materials under flowing LBE
SMRs (<300MWe) Have a Wide Range of Advantages, but Numerous Materials Issues that Must Be Addressed Integrated LWRs may be deployed first (NuScale, mpower, Westinghouse: 45-200MWe) SFRs offer higher output temperatures and fast neutron spectra (4S, PRISM: 10-311MWe) HTGRs & VHTRs offer still higher output temperatures (NGNP: 150-200 MWe) Advanced (transformational) reactor concepts may offer longer lives, higher output temperatures, greater safety, better economics, etc. (Gen 4 Energy/leadbismuth: 25-30 MWe, AHTR/liquid salt: 100-400 MWe, etc.) Resolving Materials Issues Specific to SMRs Is Challenge