Advanced LWRs Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering
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Advanced LWRs Jacopo Buongiorno Associate Professor of Nuclear Science and Engineering 22.06: Engineering of Nuclear Systems Outline Performance goals for near-term advanced LWRs Technical features of advanced LWRs: LWRs: - US-EPR (Evolutionary Pressurized Reactor) - US US-APWR -APWR (Advanced Pressurized Water Reactor) Reactor) - AP1000 (Advanced Passive 1000) - ABWR (Advanced BWR)) - ESBWR (Economic Simplified BWR) Summaryy of common characteristics Conclusions 2 Nuclear Reactor Timeline 3 Mission/Goals for Advanced LWRs Baseload generation of electricity (hydrogen is not emphasized) Improved economics. Targets: - Increased plant design life (60 years) - Shorter construction schedule (36 months months*) ) - Low overnight capital cost ($1000/kWe** for NOAK plant) - Low O&M cost of electricity ( 1¢/kWh) * First concrete to fuel loading (does not include site excavation and pre-service testing) ** Unrealistic target set in early 2000s. Current contracts in Europe, China and US have overnight capital costs >$3000/kWe Improved safety and reliability reliability - Reduced need for operator action - Expected to beat NRC goal of CDF<10-4/yr - Red duced d l large rellease prob babilit bility - More redundancy or passive safety 4 U.S S. NRC Certifi tificatition off Ad Advanced d LWRs WR Design Applicant Type AP1000 WestinghouseToshiba Advanced Passive PWR 1100 MWe ABWR GE-Hitachi Advanced BWR 1350 MWe ESBWR GE-Hitachi US-EPR AREVA US APWR US-APWR Mit bi hi Mitsubishi Advanced Passive BWR 1550 MWe Advanced PWR 1600 MWe Ad Advanced d PWR 1700 MWe Status Certified, amendment under review Certified, Constructed in Japan/Taiwan Under review Under review U d review Under i U.S. Economic Pressurized Reactor (US (US-EPR) EPR) b A by Areva 6 US-EPR Overview 1600 MWe PWR Typical PWR operating conditions in primary system, pressure, temperatures linear power temperatures, power, etc. 4 loops Hi h pressure iin SG Higher SGs results in somewhat higher efficiency (35% net) Safety systems are active High redundancy 7 US EPR Parameters US-EPR Parameter Design life, life yrs Net electric output, MWe Reactor power, MWt Plant efficiency, y, % Cold/hot leg temperature, C Reactor pressure, MPa Total RCS volume, m3 Number of fuel assemblies Type of fuel assemblies Active length, m Li Linear power, kW/m kW/ Control rods Steam pressure, MPa Radial reflector SG secondary inventory, ton Current 4-loop PWR 40 1100 3411 32.2 292/325 15.5 350 193 17x17 3.66 18 3 18.3 53 6.7 No 46 EPR 60 1600 4500 35.6 296/329 15.5 460 241 17x17 4.20 16 4 16.4 89 7.7 Yes 83 US-EPR Safetyy Four identical diesel- driven trains, each 100%, provide redundancy for maintenance or single- failure criterion (N+2) Physical separation against internal hazards (e.g. fire) Shield building extends airplane crash and external explosion protection to two safeguard buildings and fuel building (blue walls) 9 US-EPR Safetyy (2) U.S. NRC Safety Goal 1 x 10-4 Current U.S. LWR Plants 5 x 10-5 EPRI Utility Requirement 1 x 10-5 4 x 10-7 Core Damage Frequency Per Year 10 US-EPR Containment Inner wall pre-stressed concrete with i h steell liner li Outer wall reinforced concrete Protection against airplane crash Protection against external explosions Annulus sub-atmospheric and filtered to reduce radioisotope release 11 US-EPR US EPR Severe Accidents Mitigation IRWST Corium Spreading Area Ex-vessel core catcher concept (passive) - Molten core is assumed to breach vessel - Molten core flows into spreading area and is cooled by IRWST water - Hydrogen recombiners ensure no detonation within container 12 EPR is being g built now Olkiluoto – September 2009 Taishan – September 2009 Olkiluoto 3 (Finland) - construction start 2004 Flamanville 3 (France) - construction start 2007 Taishan (China) – construction start 2008 Flamanville – October 2009 © source unknown. All rights reserved. This content is excluded from our Creative Commons license. For more information, see http://ocw.mit.edu/fairuse. U.S. Advanced PWR (US-APWR) by Mitsubishi 14 US-APWR Overview (fundamentally similar to US-EPR) 1700 MWe PWR Typical PWR operating conditions in primary system, pressure, temperatures etc temperatures, etc. Long (14 ft.) fuel assemblies with reduced power density for 24 months operation operation 4 loops High efficiency turbine (70" blades) results in >35% thermal efficiency of plant RPV with no bottom penetrations Safety systems are active with high high redundancy 15 US-APWR Safety y HP 16 US-APWR Safetyy (2) ( ) Accumulator design with flow damper eliminates need for active high-pressure injection system Severe accid identts miti itigatition based d on core-cattcher h conceptt similar to US-EPR 17 Advanced Passive 1000 (AP1000) (AP1000) b W by Westingh ti house-Toshiba T hib 18 AP1000 Overview 1100 MWe PWR Typ ypical PWR op perating g conditions, pressure, temperature, flow rates, linear power, etc. RPV with no bottom penetrations 2 loops, 2 SGs 4 recirculation pumps (canned motor pumps, no shaft seals) Large pressurizer 50% larger than operating plants All safety-grade systems are passive 19 Courtesy of Westinghouse. Used with permission. AP1000 Passive Core Cooling System PRHR HX Natural circ. heat removal Passive Safety Injection Core Makeup Tanks (CMT) Accumulators Kick in at intermediate pressure IRWST Injection Full press, natural circ. inject Replaces HPCI pumps Low press (replaces LPCI pumps) Automatic RCS Depressurization Courtesy of Westinghouse. Used with permission. 20 AP1000 Passive Containment Cooling gS yystem 21 Courtesy of Westinghouse. Used with permission. AP1000 Severe Accidents Mitigation Core-Concrete Interaction In-Vessel Retention (IVR) / ex-vessel cooling cooling Means of cooling damaged core REACTOR VESSEL Tests and analysis of IVR reviewed by U S NRC U.S. NRC High Pressure Core Melt REACTOR VESSEL SUPPORT STEEL STEAM VENTS TYPICAL 4 PLACES CORE Eliminated by redundant, diverse ADS SHIELD WALL Hydrogen Burn, Detonation Hydrogen vent paths from RCS kept away from containment shell shell Redundant, diverse igniters 4.8 m Vessel INSULATION Water Corium melt WATER INLET Steam Explosions Ex-vessel prevented by IVR 22 cm W Water Courtesy of Westinghouse. Used with permission. 22 AP1000 videos ECCS http://www.ap1000.westinghousenuclear.com/ap1000_psrs_pccs.html PCCS http://www.ap1000.westinghousenuclear.com/ap1000_psrs_pcs.html IVR http://www.ap1000.westinghousenuclear.com/ap1000_safety_ircd.html AP1000 Safety Margins and Risk Typical Plant AP1000 • Loss Flow Margin to DNBR Limit ~ 1 - 5% ~16% o • Feedline Break ( F) Subcooling Margin >0o F ~140o F Operator actions required in 10 min Operator actions NOT required • Small LOCA 3 LOCA 3" core uncovers PCT ~1500oF <8 8" LOCA NO core uncovery o • Large LOCA PCT ( F) with uncertainty 2000 - 2200o F <1600o F (1) • SG Tube Rupture Core Damage Frequency At-Power Shutdown Internal Events Internal Floods Internal Fires Sub-Totals Grand-Totals NRC Safety Goals Large Release Frequency At-Power Shutdown 2.41E-07 /yr 8.80E-10 /yr 5.61E-08 /yr 1.23E-07 /yr 3.22E-09 /yr 8.52E-08 /yr 1.95E-08 /yr 7.10E-11 /yr 4.54E-09 /yr 2.05E-08 /yr 5.40E-10 /yr 1.40E-08 /yr y 2.98E-07 /yr 2.11E-07 /yr y 2.41E-08 /yr y 3.50E-08 /yr y 5.09E-07 5.92E-08 1 E-4 1 E-6 24 Courtesy of Westinghouse. Used with permission. Use of passive safety systems simplifies the plant plan plant Safety Valves Pumps Safety Pipe Seismic Building Volume Cable Reduced Number of Components: 1000 MW Reference Safety Valves AP 1000 Reduction 2844 1400 51% Pumps 280 184 34% Safety Piping 11.0 x 104 feet 1.9 x 104 feet 83% Cable 9.1 mil. feet 1.2 mil. feet 87% Seismic Building Volume 12.7 mil. ft3 5.6 mil. ft3 56% Image by MIT OpenCourseWare. 25 …and dR Reduces d S Safety/Seismic f t /S i i B Building ildi V Volume l LEGEND: 1. CONTAINMENT/ SHIELD BUILDING 6 2. SAFEGUARD BUILDING 3. FUEL BUILDING 4. AUXILIARY BUILDING 5. DIESEL GENERATOR BUILDING 6. ESSENTIAL SERVICE WATER/ CIRCULATINGWATER PUMP HOUSE 7. LIQUID RADWASTE BUILDING 2 2 2 4 5 5 1 2 2 4 3 EPR Safet Saf et y / Seismic Sei smi c St Structures r ct res 0 20 40 60 80 100M 1 7 3 7 AP1000 Safety S f / Seismic Structures S Courtesy of Westinghouse. Used with permission. 26 AP1000 Construction Simplification of Systems Reduction in bulk materials and field labor Maximum Use of Modularization 300 rail-shippable equipment and piping modules 50 llarge structural t t l mod dulles (assembl bled don-s itite)) - Under construction at Taishen (China) since 2008 Courtesy of Westinghouse. Used with permission. - 4 P&E orders in US Advanced BWR (ABWR) and Economic Simplified BWR (ESBWR) b G by Generall El Electric-Hitachi t i Hit hi 28 ABWR Overview Steam dryer Steam nozzle Steam separator Feedwater nozzle Fuel assemblies Pressure vessel Vessel support skirt Control rod guide tubes Control rod drives Recirculation pump Courtesy of GE Hitachi Nuclear Systems. Used with permission. 1350 MWe BWR Typical BWR operating conditi ditions, pressure, temperature, linear power, etc. Internal recirculation pumps (no jet pumps) = no external loop Large vessel with large water inventory + no large piping = no core uncovery Redundant active safety systems Proven technology (built and operated for over ten years in Japan and Taiwan) 29 ESBWR Overview Closure Head Steam Dryer DPV/ IC Outlet Main Steam Line Flow Restrictor Steam Separators RPV Stabilizer RWCU/ SDC Outlet Feedwater Inlet IC Return Chimneyy GDCS Inlet Chimney Partitions Support (sliding block) Core Top Guide Fuel Assemblies GDCS Equalizing Li Inlet Line l Co C ore Plate Shroud Support Core Shroud CRD Housing Control Rod Guide Tubes Control Rod Drives (CRD) In-core Housing ESBWR Reactor Courtesy of GE Hitachi Nuclear Systems. Used with permission. 1550 MWe BWR T i l BWR operatiting Typical conditions, pressure, temperature, linear power, etc. etc Natural circulation reactor = No reactor pumps Large vessel with large water inventory Core at lower elevation within vessel All safety-grade systems are passive 30 BWR Primaryy System y Evolution Dresden 1 KRB Oyster Creek Dresden 2 ABWR SBWR ESBWR 31 Courtesy of GE Hitachi Nuclear Systems. Used with permission. ABWR & ESBWR Balance of Plant is Traditional Courtesy of GE Hitachi Nuclear Systems. Used with permission. 32 ABWR & ESBWR Parameters BWR/4-Mk I(Browns Ferry 3) BWR/6-Mk III (Grand Gulf) ABWR ESBWR 3293/1098 3900/1360 3926/1350 4500/1550 Vessel height/dia. (m) 21.9/6.4 21.8/6.4 21.1/7.1 27.7/7.1 Fuel Bundles (number) 764 800 872 1132 Active Fuel Height (m) 3.7 3.7 3.7 3.0 Power density (kW/L) 50 54 2 54.2 51 54 Recirculation pumps 2(large) 2(large) 10 zero Number of CRDs/type 185/LP 193/LP 205/FM 269/FM Safety system pumps 9 9 18 zero Safety diesel generator 2 3 3 zero Core damage freq freq./yr /yr 1E 5 1E-5 1E 6 1E-6 1E 7 1E-7 1E 7 1E-7 Safety Bldg Vol (m3/MWe) 115 150 160 <100 Parameter Power (MWt/MWe) 33 Courtesy of GE Hitachi Nuclear Systems. Used with permission. ABWR Safetyy Steam line CSS CSS Feedwater line Feedwater line HPCF HPCF LPFL LPFL LPFL SPC SPC Supression Pool Supression Pool Supression Pool LPFL LPFL Supression Pool LPFL Condensate Storage Pool RHR RCIC G D Division 1 Division 2 G D Division 2 RHR Division 1 RHR Division 3 G D Division 3 RHR ECCS Courtesy of GE Hitachi Nuclear Systems. Used with permission. 34 ESBWR Enhanced Natural Circulation 4.5 ABWR LUNGMEN ABWR 4.0 Average Powerr per Bundle (MWt) BWR/6 CLINTON 3.5 ESBWR ESBWR 1132 - a 3.0 2.5 2.0 1.5 1.0 0.5 0.0 0.0 1.0 2.0 N Power Flow - 1132-4500.XLS Chart1 (5) 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 Average Flow per Bundle (kg/s) • Higher driving head • Chimney/taller vessel • Reduced flow restrictions •Shorter core • Increase downcomer area Courtesy of GE Hitachi Nuclear Systems. Used with permission. 35 ESBWR Stability y ESBWR is designed to operate with significant margin to any instability regions Courtesy of GE Hitachi Nuclear Systems. Used with permission. 36 ESBWR Passive Safety y 37 Courtesy of GE Hitachi Nuclear Systems. Used with permission. ESBWR Passive Safety y Decay Heat HX’s Above Drywell All Pipes/Valves Inside Containment High Elevation Gravity Drain Pools Raised Suppression Pool 38 Courtesy of GE Hitachi Nuclear Systems. Used with permission. ESBWR Passive Systems Isolation Condensers System (ICS) High pressure residual heat removal Safety Relief Valves (SRV) Prevent reactor overpressurization discharging steam into suppression pool Suppression Pool Absorbs blowdown energy during LB-LOCA. Gravity Driven Cooling System (GDCS) Low pressure residual heat removal following LB-LOCA. Keep ps the core covered. Passive Containment Cooling System (PCCS) Long-term heat removal from containment No operator action needed for 72 hours 39 ESBWR Severe Accident Mitigation Containment filled with inert gas In-vessell rettenti tion is complilicatted d by b CRDM penetrations, so it is not done. Quench molten core byy delu ge from the GDCS tanks If molten material drips through vessel, there is a sacrificial concrete shield (core catcher) on the containment floor Easy to refill PCCS pool and continue to remove the heatt from th f the th vessell ind i definit fi itelly Fission Product Control Hold up and filtering filtering 40 Comparison p of Safety y Syystem - Passive vs. Active Emergency Bus Loading Program DG Room Ventilation System Diesel Generator Room 1 of 3 Crankcase Ventilation G Generator t Control and Protection Engine Governing Control DG Lubrication Oil System Plant Service Water Diesel Starting Air Generator Air Intake & Exhaust Breaker Closes < 10 s Plant Service Water Breaker Pump Motor DC Pwr DG Fuel Oil Storage and Transfer System Loads ADS Logic ADS Aux. Aux Water Source Water Source HVAC Reactor Component Cooling Water Breaker P Pump Motor M t RPV Q S/P Loads Core M RCCW HVAC DG Fuel Oil System DC Pwr Q HVAC Emergencyy Bus DG Cooling Water System Initiation Signal Emergency Core C li Systtem Cooling Breaker Pump Motor Typical of HPCS, LPCS, & RHR Plant Service Water A M ECCS Logic A Initiation Signal Passive Plant Conventional Active Plant 41 Courtesy of GE Hitachi Nuclear Systems. Used with permission. Reduction in Systems & Buildings with Passive Systems ABWR ESBWR (higher power, smaller building) 42 Courtesy of GE Hitachi Nuclear Systems. Used with permission. Summary Features of Advanced LWRs Reactor US-EPR US-APWR AP1000 ABWR ESBWR Neutron spectrum Thermal Thermal Thermal Thermal Thermal Coolant/moderator H2O/H2O H2O/H2O H2O/H2O H2O/H2O H2O/H2O Fuel LEU pins LEU pins LEU pins LEU pins LEU pins ++ ++ + ++ + Use of proven technology Plant simplification ++ ++ M d l construction Modular t ti + + Economy of scale ++ ++ High thermal efficiency + + Passive safety g of severe Mitigation accidents + + Core catcher Core catcher In-vessel retention ++ + - Core catcher Potential Issues for Dep ployment y of Advanced LWRs in the U.S. •No capabilities for manufacturing heavy components left. Need to buy from overseas. •Shortage of specialized workforce experienced in nuclear construction ((e.g., g , welders)). •Slow licensing process •Financial Fi i l riisk k iin deregulated l t d markets k t 44 MIT OpenCourseWare http://ocw.mit.edu 22.06 Engineering of Nuclear Systems Fall 2010 For information about citing these materials or our Terms of Use, visit: http://ocw.mit.edu/terms.
