Module 15 Advanced Reactors LWR 3+ Generation IV

Transcription

1 Prof.Dr. H. Böck Vienna University of Technology /Austria Atominstitute Stadionallee 2, 1020 Vienna, Austria Module 15 Advanced Reactors LWR 3+ Generation IV

2 Generation 3 + Advanced Nuclear Power Reactors PWR & BWR APWR by Mitsubishi APR 1400 by South Korea ATMEA by AREVA with Mitsubishi MIR-1200 by Atomenergoproject ABWR by Toshiaba & General Electric ESBWR GE Hitachi Nuclear Energy KERENA AREVA SWR 1000 VBER-300 Russia OKBM RMWR JAERI & JAPCO

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5 Status of Gen III+ Reactors

6 Core damage frequencies

7 APWR (Advanced PWR) It is a 4-loop design with 257 fuel assemblies and neutron reflector Simpler, combines active and passive cooling systems 1538 MWe gross Was developed in collaboration with four utilities (Westinghouse was earlier involved). Basis for the next generation of Japanese PWRs Planned APWR+ is 1750 MWe and has full-core MOX capability US-APWR will be 1620 MWe net Due to longer (4.3m) fuel assemblies, higher thermal efficiency (37%), 24 month refueling cycle

8 Advanced PWR 1400 MWe net Two loop design APR 1400 (South Korea) Two units under construction 18 month fuel cycle Plant life 60 years

9 ATMEA (AREVA & Mitsubishi) Mid-size reactor targeted for nuclear embarking countries Three loop PWR 1150 MWe net using same steam generators as EPR 37% thermal efficiency, 60-year life time 157 fuel assemblies, 4,2m long, Fuel cycle flexible 12 to 24 months with short refuelling

10 MIR 1200 (Modernized International Reactor Russia) Improved VVER 1000 with 1170 MWe Four loop PWR 50 years life time 36,5% efficiency Passive safety features Resistant to external events About 20 under planning or construction? COUNTRIES

11 ABWR (Advanced Boiling Water Reactor) Derived from a General Electric BWR design Three NPPs are in commercial operation in Japan (1315 MWe net) One under construction in Japan and two in Taiwan Four more are planned in Japan another two in the USA Basically a 1380 MWe unit in Toshiba version Improvements in I&C system, maintanance, availability and safety, better fuel utilization pdf hibashi.pdf

12 ESBWR (Economic & Simplified BWR) GE Hitachi Nuclear Energy's is a Generation III+ technology Utilizes passive safety features and natural circulation principles, 25% fewer pumps, valves and motors The reactor cools down after an accident without the need for human intervention. The ESBWR (4500 MWt) will produce approximately 1535 MWe net with a design life 60 years Parallel on the market to ABWR more expensive to build and operate, but proven ESBWR is more innovative, with lower building and operating costs sbwr_nuclear_reactor.jsp NUCNET 2014/364

13 KERENA Areva NP with former Siemens 1290 MWe BWR formerly SWR year design life Passive safety features, including corecatcher Refuelling intervals of up to 24 months 37% net efficiency ready for commercial deployment 420/kerena-a-midpower-boiling-waterreactor.html

14 VBER-300 (Russia) OKBM's VBER-300 PWR is a MWe unit (917 MWt) developed from naval power plants Originally envisaged in pairs as a floating nuclear power plant with two steam generators for each reactor module Designed for 60 year life and 90% capacity factor Fuel cycle up to 10 years It now planned to develop it as a land-based unit with Kazatomprom, with a view to exports First unit will be built in Kazakhstan 300

15 RMWR (Reduced-Moderation Water Reactor) It is a light water reactor with the fuel packed in the core more tightly to reduce the moderating effect of the water Reduced moderation means that more fissile plutonium is produced and the breeding ratio is around 1 instead of about 0.6 More of the U-238 is converted to Pu-239 and then burned than in a conventional reactor Main rationale for RMWRs is extending the world's uranium resource and providing a bridge to fast neutron reactors c=report&doc_id=80&type_of_output=html

16 Advanced Nuclear Power Reactors (D 2 O, HTR, LMFBR) CANDU-6 Canada ACR-700 Advanced CANDU AHWR Advanced Heavy Water Reactor India (Thorium) HTR-PM Chinese HTR Pebble Bed Modular Reactor PBMR South Africa Pebble Bed Modular Reactor GT-MHR Gas Turbine-Modular Helium Reactor (US,Russia) JSFR Japan Standard Fast Reactor BN-800 (Russia) BREST Lead Cooled Reactor (Russia) ELSY European Lead Cooled System PRISM GE-Hitachi Liquid Metal Fast Breeder KALIMER Korea Advanced Liquid Metal Reactor

17 Generation IV Participants

18 Evolution of Nuclear Power Plants

19 Gen IV: Criteria and Goals

20 Goals of Generation IV Sustainability: Generation IV nuclear energy-systems will provide sustainable energy generation that meets clean air objectives and promotes longterm availability of systems and effective fuel utilization for worldwide energy production. Generation IV nuclear energy-systems will minimize and manage their nuclear waste and notably reduce the long-term stewardship burden, thereby improving protection for the public health and the environment. Economics: Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources. Generation IV nuclear energy systems will have a level of financial risk comparable to other energy projects.

21 Goals of Generation IV Safety and Reliability: Generation IV nuclear energy systems operations will excel in safety and reliability, they will have a very low likelihood and degree of reactor core damage and Generation IV nuclear energy systems will eliminate the need for offsite emergency response. Proliferation Resistance and Physical Protection: Generation IV nuclear energy systems will increase the assurance that they are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.

22 Application of Gen IV Reactors

23 Spent Fuel and Uranium Resources

24 Generation IV General Design (ATW 12/06) Neutron spectrum Coolant Temp. ( 0 C) Fuel Fuel cycle Size(s) (MWe) Sodium-cooled Fast Reactors (SFR) Very High Temperature Gas Reactors (VHTR) fast sodium 550 U-238 & MOX thermal helium 1,000 UO 2 prism or pebbles closed 150 1,500 open 275 Gas-cooled Fast Reactors (GFR) fast helium 850 U-238 closed 275 Supercritical Watercooled Reactors (SCWR) Lead-cooled Fast Reactors (LFR) Molten Salt Reactors (MSR) thermal or fast fast epithermal water 500 UO 2 or MOX open (thermal) or closed (fast) Pb or Pb-Bi fluoride salts U-238 closed UF in salt closed 1,000

25 Sodium Cooled Fast Reactor

26 Sodium Cooled Fast Reactor Reactor Parameters Outlet Temperature Pressure Rating Fuel Cladding Average Burnup Conversion Ratio Average Power Density Reference Value C -1 Atmospheres MWth Oxide or metal alloy Ferritic or ODS ferritic GWD/MTHM MWth/m³

27 Sodium Cooled Fast Reactor System Fast-spectrum reactor and closed fuel recycle system. Power from a few hundred MWe to MWe. Sodium core outlet temperatures are typically ºC. Either pool layout or compact loop layout is possible. Primary system operates at essentially atmospheric pressure, pressurized only to the extent needed to move fluid. Sodium reacts chemically with air and with water, design must limit the potential for such reactions and their consequences Secondary sodium system acts as a buffer between the radioactive sodium in the primary system and the steam or water If a sodium-water reaction occurs, it does not involve a radioactive release.

28 Gas Cooled Fast Reactor

29 Gas Cooled Fast Reactor Reactor Parameters Reactor Power Net plant efficiency (direct cycle helium) Coolant inlet/ outlet temperature and pressure Average power density Reference fuel compound Volume fraction, Fuel/ Gas/ SiC Conversion ratio Reference Value 600 MWth 48% 490xC/850xC at 90 bar 100 MWth/m³ UPuC/SiC(70/30%)with about 20% Pu content 50/40/10% Self sufficient

30 Gas-Cooled Fast Reactor System The GFR system features a fast-spectrum helium-cooled reactor and closed fuel cycle. The high outlet temperature of the helium coolant makes it possible to deliver electricity, hydrogen, or process heat with high conversion efficiency. The GFR uses a direct-cycle helium turbine for electricity and can use process heat for thermo-chemical production of hydrogen. The GFR s fast spectrum also makes it possible to utilize available fissile and fertile materials (including depleted uranium from enrichment plants) two orders of magnitude more efficiently than thermal spectrum gas reactors with once-through fuel cycles. The GFR reference assumes an integrated, on-site spent fuel treatment and re-fabrication plant.

31 Lead-Cooled Fast Reactor System

32 Lead Cooled Fast Reactor Reference values Reactor Parameters Reference Value Pb-Bi Battery (nearer-term) Pb-Bi Module (nearer-term) Pb-Large (nearer-term) Pb-Battery (nearer-term) Coolant Outlet Temperature( C) Pressure(Atmospheres) Power approx (MWth) Fuel Cladding Average Burnup (GWD/MTHM) Conversion Ratio Lattice Primary Flow Pb-Bi Metal Alloy or Nitride Ferritic Open Natural Pb-Bi Metal Alloy Ferritic d>1.0 Open Forced Pb Nitride Ferritic Mixed Forced Pb Nitrid Ceramic coatings or refractory alloys Open Natural

33 Lead-Cooled Fast Reactor System LFR systems are Pb or Pb-Bi alloy-cooled reactors with a fast-neutron spectrum and closed fuel cycle. Options include a range of plant ratings, including a long refuelling interval battery ranging from MWe, a modular system from MWe, and a large monolithic plant at 1200 MWe. It had the highest evaluations to the Generation IV goals among the LFR options, but also the largest R&D needs and longest development time. The nearer-term options focus on electricity production and rely on more easily developed fuel, clad, and coolant combinations and their associated fuel recycle and re-fabrication technologies. The longe term option seeks to further exploit the inherently safe properties of Pb and raise the coolant outlet temperature sufficiently high to enter markets for hydrogen and process heat, possibly as merchant plants.

34 Molten Salt Reactor

35 Molten Salt Reactor Reactor Parameters Net power Power density Net thermal efficiency Fuel - salt - inlet temperature - outlet temperature - vapour pressure Moderator Power Cycle Neutron spectrum turner Reference Value 1000 Mwe 22 MWth/m³ 44 to 50% 565 C 700 C (850 C for H-Production) <0.1 psi Graphite Multi - reheat recuperative HE Thermal - actinide

36 Molten Salt Reactor System The MSR produces fission power in a circulating molten salt fuel mixture fuelled with uranium or plutonium fluorides dissolved in a mixture of molten fluorides, with Na and Zr fluorides as the primary option. MSRs have good neutron economy, opening alternatives for actinide burning and/or high conversion High-temperature operation holds the potential for thermochemical hydrogen production Molten fluoride salts have a very low vapour pressure, reducing stresses on the vessel and piping Inherent safety by fail-safe drainage, passive cooling low inventory of volatile fission products in the fuel Refuelling, processing, and fission product removal performed online, potentially yielding high availability MSRs allow the addition of actinide feeds to the homogenous salt solution without blending and fabrication needed by solid fuel reactors.

37 Supercritical Water Cooled Reactor

38 Supercritical Water Cooled Reactor Reactor Parameters Plant capital cost Unit power and neutron spectrum Net efficiency Coolant inlet and outlet temperature and pressure Average power density Reference fuel Fuel structural materials Cladding structural materials Burnup / Damage Safety approach Reference Value $ 900/KW 1700 MWe; thermal spectrum 44% 280xC/510xC/25Mpa 100MWth/m³ UO 2 with austenitic or ferritic martensitic stainless steel or Ni alloy cladding Advanced high strength Metal alloys are needed 45GWD/MTHM; Similar to ALWRs

39 Supercritical-Water-Cooled Reactor System SCWRs are high-temperature, high-pressure water- cooled reactors that operate above the thermodynamic critical point of water (374 C, 22.1 MPa) These systems may have a thermal or fast-neutron spectrum, depending on the core design. SCWRs offer increases in thermal efficiency relative to currentgeneration LWRs. The efficiency of a SCWR can approach 44%, compared to 33 35% for LWRs. A lower-coolant mass flow rate per unit core thermal power.this offers a reduction in the size of the reactor coolant pumps, piping, and associated equipment, and a reduction in the pumping power. Steam dryers, steam separators, recirculation pumps,

40 Very High Temperature Reactor

41 Very High Temperature Reactor Reactor Parameters Reactor power Coolant inlet/outlet temperature Core inlet /outlet pressure Helium mass flow rate Average power density Reference fuel compound Net plant efficiency Reference Value 600 MWth 640 / 1000 C Dependent on process 320 kg/s 6 10 MWth/m³ ZrC coated particles in blocks, pins or pebbles > 50 %

42 Very High Temperature Reactor Systems The VHTR next step in development of HTR Graphite-moderated, helium-cooled reactor with thermal neutron spectrum able to supply nuclear heat, core-outlet temperatures of 1000 C, either prismatic block core or a pebble-bed core. VHTR may produce hydrogen from only heat and water using thermochemical iodine-sulfur (I-S) process Possible daily production over 2 million normal cubic meters of hydrogen. It may also generate electricity with high efficiency, over 50% at 1000 C, compared with 47% at 850 C in the GTMHR or PBMR. High core outlet temperatures would enable nuclear heat application to such processes as steel, aluminium oxide, and aluminium production. Direct cycle electricity generation with the helium gas turbine is possible. For nuclear heat applications such as process heat for refineries, petro-chemistry, metallurgy, and hydrogen production, the heat application process is generally coupled with the reactor through an intermediate heat exchanger (IHX), which is called an indirect cycle.

43 Candidate Materials

44 Combined Thermal and Fast Fuel Cycle

45 References -> search for Generation IV ADVANCED REACTORS Miniature Nuclear Power Plants Atomwirtschaft 2011 p IAEA Nuclear Technology Review