Introduction to Generation IV Nuclear Energy Systems

Transcription

1 Introduction to Generation IV Nuclear Energy Systems Dr. Ralph Bennett, Technical Director, Generation IV International Forum, and Director, International and Regional Partnerships, Idaho National Laboratory 16 Mar 2009

2 The Problem of Climate Change Global greenhouse gas (GHG) emissions have grown since pre-industrial times, increasing 70% between 1970 and 2004 With current climate change mitigation policies and practices, global GHG emissions will continue to grow The Earth is about to undergo long lasting changes in its climate, seas and land cover, including Temperature Precipitation Sea level Ocean circulation Ice/snow cover Storm frequency Storm intensity Desertification Global Warming (deg C) by 2100 (IPCC prediction) 2

3 The Challenge for Nuclear Energy Nuclear is a major contributor in the WEO Policy Scenario about 250 GWe more generation by 2030 (an 80% increase from today) Nuclear energy systems must continue their advances in order to unlock a potential on this scale 3

4 Generations of Nuclear Energy Generation IV Generation III+ Revolutionary Designs Generation I Early Prototypes Generation II Commercial Power Generation III Advanced LWRs Evolutionary Designs - Shippingport - Dresden - Magnox - PWRs - BWRs - CANDU - CANDU 6 - System AP600 - ABWR - ACR AP APWR - EPR - ESBWR - Safe - Sustainable - Economical - Proliferation Resistant and Physically Secure Gen I Gen II Gen III Gen III+ Gen IV 4

5 Creation of the International Forum Started in Jan 2000 by nine countries and established Jul 2001 Agreed that nuclear energy is needed to meet future needs Defined four goal areas to advance nuclear energy into its next, fourth generation: Sustainability Safety & reliability Economics Proliferation resistance and physical protection Will collaborate to make Generation IV systems deployable in large numbers by 2030, or earlier 5

6 Today s Membership 6

7 Overview of the Generation IV Systems System Neutron Spectrum Fuel Cycle Size (MWe) Missions R&D Needed Sodium Cooled Fast Reactor (SFR) Fast Closed Electricity, Actinide Management Advanced recycle options, Fuels Very-High- Temperature Reactor (VHTR) Gas-Cooled Fast Reactor (GFR) Thermal Open 250 Electricity, Hydrogen, Process Heat Fast Closed 1200 Electricity, Hydrogen, Actinide Management Fuels, Materials, H 2 production Fuels, Materials, Thermal-hydraulics Supercritical-Water Reactor (SCWR) Thermal, Fast Open, Closed 1500 Electricity Materials, Thermalhydraulics Lead-Cooled Fast Reactor (LFR) Fast Closed Electricity, Hydrogen Production Fuels, Materials Molten Salt Reactor (MSR) Epithermal or Fast Closed 1000 Electricity, Hydrogen Production, Actinide Management Fuel treatment, Materials, Reliability 7

8 Sodium-Cooled Fast Reactor (SFR) Characteristics Sodium coolant, pool or loop type 550C outlet temperature MWe large size, or MWe intermediate size 50 MWe small module option Metal fuel with pyroprocessing or MOX fuel with advanced aqueous separation Benefits High thermal efficiency Consumption of LWR actinides Efficient fissile material generation 8

9 EXHAUST TO VENT STACK 7m (23') IHX X-SECTION (FLATTENED FOR CLARITY) CONTROL RODS (7) PLAN VIEW OF IHX AND PUMPS IHX (2) 2 1.7m EACH PUMPS (2) ON Ø 142.5" B.C. DRACS (2) 2 0.4m EACH SECONDARY CONTROL RODS Na-CO2 HEAT EXCHANGER SODIUM DUMP TANK Ø 2.5 m x 3.8 m LONG (Ø 7.5' x 12.6' LONG) PRIMARY CONTROL RODS CORE BARREL Ø 266 / 268 cm (104.7" / 105.5") PLAN VIEW OF THE CORE METERS 10 TURBINE/GENERATOR BUILDING 3.25m (10'-8") 7m [23FT] 0.75m (29.5") THERMAL SHIELD IHX 1m TRAVEL DISTANCE OF THE CONTROL RODS 4.57m Primary Vessel I.D. [15FT] 5.08m Guard Vessel I.D. [16.7FT] m 3,186 gal. 1.89m [6.2FT] Ø 7.7m (Ø 25.5') SECTION A - A Na-Air HEAT EXCHANGER (2) ELEVATOR 3.5m 1m (11'-8") (39.4") GUARD VESSEL (1" THICK) PRIMARY VESSEL (2" THICK) CONTROL BUILDING 14.76m [48.4FT] 12.72m [41.7FT] 1.93m [6.3FT].61m [2FT] Hot Pool Normal sodium level Cold Pool Normal sodium level 2.29m [7.5FT] Pump off Sodium Level Sodium faulted level SFR Reactor Options Large-scale Loop AHX Chimney Intermediate-scale Pool Secondary Pump PDRC piping IHTS piping Steam Generator SG Primary Pump/IHX Small-scale Modular IHX DHX PHTS pump Reactor core IHTS pump In-vessel core catcher Reactor Vessel 9

10 SFR Technology Interests Minor actinide bearing fuel technology (fabrication, irradiation) Metal and oxide fuel performance Carbide fuel performance Nitride/Carbide fuel performance Inspection & repair technologies Ultrasonic and alternative techniques Replace/repair experience High temperature leak-before-break assessment technologies Creep-fatigue crack initiation and growth test results Advanced energy conversion concepts Basic design concept of supercritical CO2 Brayton cycle system Compact supercritical CO2-to-CO2 heat exchangers 10

11 Very-High-Temperature Reactor (VHTR) Characteristics He coolant >900C outlet temperature 250 MWe Coated particle fuel in either pebble bed or prismatic fuel Benefits Hydrogen production Process heat applications High degree of passive safety High thermal efficiency option 11

12 VHTR Reactor Options Pebble bed core Prismatic-fuel core 12

13 VHTR Hydrogen Options Sulfur-iodine cycle High temperature electrolysis 90 v /o HO v /o H 2 10 v /o H 2 O + 90 v /o H 2 4 e - H 2 O Porous Cathode, Nickel -Zirconia cermet 2 H e - 2 H O = H 2 2 O = Gastight Electrolyte, Yttria-Stabilized Zirconia 2 O = O e - O 2 Porous Anode, Strontium -doped Lanthanum Manganite H 2 O Interconnection H 2 O + H 2 Next Nickel-Zirconia Cermet Cathode H 2 13

14 VHTR Technology Interests Fuel and fuel cycle Particle fuel irradiations and fission product monitoring Materials Codes and standards extension Materials database extension Graphite dust behavior Hydrogen production Sulfur-iodine cycle High temperature electrolysis Coupling of H2 production process and reactor heat transport system Tritium transport Computational Methods Components and helium turbine Intermediate heat exchanger 14

15 Lead-Cooled Fast Reactor (LFR) Characteristics Pb or Pb/Bi coolant 550C to 800C outlet temperature Small transportable system MWe, and Larger station MWe year core life option Benefits Distributed electricity generation Hydrogen and potable water Replaceable core for regional fuel processing High degree of passive safety Proliferation resistance through long-life core 15

16 LFR Reactor Options Small, transportable module Large, stationary plant CLOSURE HEAD CO2 OUTLET NOZZLE (1 OF 8) CO 2 INLET NOZZLE (1 OF 4) Pb-TO-CO 2 HEAT EXCHANGER (1 OF 4) CONTROL ROD DRIVES CONTROL ROD GUIDE TUBES AND DRIVELINES THERMAL BAFFLE FLOW SHROUD RADIAL REFLECTOR ACTIVE CORE AND FISSION GAS PLENUM FLOW DISTRIBUTOR HEAD GUARD VESSEL REACTOR VESSEL Pb coolant (both) No intermediate loops 16

17 LFR Technology Interests Collaborations based on ELSY and SSTAR No formal agreement yet Conceptual design and safety Innovative components and design Compact, in-vessel steam generators Decay heat removal by air and water Refueling out-of-pb coolant Innovative structural design Buoyant fuel element support Seismic isolation of reactor building Fuel and core materials Many options ELSY: European Lead-cooled System; SSTAR: Small Secure Transportable Autonomous Reactor 17

18 Supercritical-Water-Cooled Reactor (SCWR) Characteristics Water coolant above supercritical conditions (374C, 22.1 MPa) C outlet temperature 1500 MWe Pressure tube or pressure vessel options Simplified balance of plant Benefits Efficiency near 45% with excellent economics Leverages the current experience in operating fossilfueled supercritical steam plants Configurable as a fast- or thermal-spectrum core 18

19 Gas-Cooled Fast Reactor (GFR) Characteristics He coolant 850C outlet temperature Direct gas-turbine cycle or supercritical CO2 cycle with optional combined cycles 2400 MWth / 1100 MWe Several fuel options Carbide in plates or pins Nitride Oxide Benefits High efficiency Waste minimization and efficient use of uranium resources 19

20 Molten Salt Reactor (MSR) Characteristics Fuel is liquid fluorides of U or Th with Li, Be, Na and other fluorides C outlet temperature 1000 MWe Low pressure (<0.5 MPa) Benefits Waste minimization Avoids fuel development Proliferation resistance through low fissile material inventory 20

21 Organization Policy Group Chair (France) Senior Industry Advisory Panel Experts Group Policy Secretariat Chair System Steering Committees Policy Director Technical Director Methodology Working Groups Proliferation Resistance and Physical Protection, Risk & Safety, Economics Co-Chairs Project Management Boards (multiple R&D projects) Technical Secretariat NEA, Paris 21

22 System Partners Mar 2009 VHTR GFR SFR SCWR LFR MSR Partners: NRCan JRC CEA JAEA, MEST, PSI DOE CAEA, DME ANRE KOSEF MOST ANRE CAEA CEA DME DOE JAEA JRC KOSEF MEST MOST NRCan PSI Agency for Natural Resources and Energy (JP) China Atomic Energy Authority (CN) Commissariat à l Énergie Atomique (FR) Department of Minerals and Energy (ZA) Department of Energy (US) Japan Atomic Energy Agency (JP) Joint Research Centre (EU) Korean Science and Engineering Foundation (KR) Ministry of Education, Science and Technology (KR) Ministry of Science and Technology (CN) Natural Resources Canada (CA) Paul Scherrer Institute (CH) VHTR GFR SFR SCWR LFR MSR Very-High-Temperature Reactor Gas-Cooled Fast Reactor Sodium-Cooled Fast Reactor Supercritical Water-Cooled Reactor Lead-Cooled Fast Reactor Molten Salt Reactor 22

23 Generation IV Annual Report Captures key information and accomplishments from System Steering Committee annual reports into one widely distributed report Captures brief summaries of working groups accomplishments, and background on the Forum Audience includes: World-wide Research and Development Community Governments sponsoring Generation IV R&D GIF committees, boards and working groups The 2008 Report has just issued 23

24 Working Toward the Future The GIF joined together to help assure a sustainable energy future Underscored by the advance of global climate change Based on advanced nuclear energy systems that are sustainable, safe, economical, proliferation resistant and physically secure Accelerated by the collaboration of the GIF members, industry, academia and non-member nations and institutions 24

25 Bibliography The web links provided on most slides lead to source documents, background materials or updates The full Generation IV Roadmap and all supporting documents are available at: Some technical papers are listed on the OECD NEA website (GIF website) at within each system Recent outlook articles on nuclear deployment: IEA (subscription) NEA (subscription) IAEA WNA EPRI (US R&D strategy and deployment outlook, respectively) My contact information: 25