Fluoride Salt Cooled High Temperature Reactors

Transcription

1 Fluoride Salt Cooled High Temperature Reactors Workshop on Advanced Reactors PHYSOR 2012 Knoxville, TN April 15, 2012 David Holcomb

2 FHRs Combine Desirable Attributes From Other Reactor Classes Gas Cooled Reactors TRISO fuel Structural ceramics High temperature power conversion Molten Salt Reactors Fluoride salt coolant Structural alloy Hydraulic components Liquid Metal Reactors Passive decay heat removal Low pressure design Hot refueling Fluoride Salt Cooled Reactors High temperature Low pressure Passive safety Light Water Reactors High heat capacity coolant Transparent coolant Advanced Coal Plants Supercritical water power cycle Structural alloys 2 Managed by UT-Battelle

3 FHRs Are Important to the World as a Potential Future Primary Electricity and Gasoline Energy Source Large FHRs have transformational potential to provide lower cost, high efficiency, large scale electrical power May be cheaper than LWRs due to higher thermal efficiency, low-pressure, and passive safety Small, modular FHRs can be cost effective, local process heat sources High temperature, liquid cooling enables efficient hydrogen production Domestic oil shale based gasoline production requires large-scale, distributed process heat FHRs have a high degree of inherent passive safety No requirement for offsite power or cooling water Low-pressure primary and intermediate loops Plant concept and technologies must be matured significantly before the potential for FHRs can be realized Lithium enrichment must be reindustrialized Tritium extraction technology must be developed and demonstrated Structural ceramics must become safety grade engineering material Safety and licensing approach must be developed and demonstrated Structured coated particle fuel must be qualified 3 Managed by UT-Battelle

4 FHRs are Central to the DOE-NE Advanced Reactor Concepts Program Mission ARC s mission is to develop and refine future reactor concepts that could dramatically improve nuclear energy performance (e.g., sustainability, economics, safety, proliferation resistance) The strategic approach is to: Tackle key R&D needs for promising concepts Fast reactors for fuel cycle missions Fluoride salt cooled thermal reactor for high-temperature missions Program includes both concept and technology development FHR technology support is also embedded throughout DOE-NE program structure University research Advanced gas reactor Nuclear Energy Enabling Technologies (NEET) Small Modular Reactors (SMR) 4 Managed by UT-Battelle

5 Potential Benefits And Challenges Of FHRs Stem From Fundamental Coolant Properties Coolant Tmelt (ºC) Tboil (ºC) Density (kg/m 3 ) Specific Heat (kj/kg K) Volumetric Heat Capacity (kj/m 3 K) Thermal Conductivity (W/m K) Kinematic Viscosity (m 2 /s) * 10 6 Li 2 BeF 4 (Flibe) NaF-40.5ZrF LiF-37NaF-37ZrF LiF-31NaF-38BeF NaF-92NaBF Sodium Na-44K Na-44K Lead Pb-55.5Bi Helium, 7.5 MPa Water, 7.5 MPa Managed by UT-Battelle

6 Potential FHR Operating Temperatures Match Many Important Process Heat Applications LiF-BeF 2 (67-33) LiF-NaF-KF ( ) NaF-BeF 2 (57-43) Melts Fluoride Salt Liquid Temperature Range Boils H 2 Production & Coal Gasification Steam Reforming of Nat. Gas & Biomass Gasification Cogeneration of Electricity and Steam Oil Shale/Sand Processing Petro Refining Temperature (ºC) 6 Managed by UT-Battelle

7 FHRs Are High Temperature Reactors and Can Support Industrial Process Heat Production High temperature reactors can efficiently produce large quantities of hydrogen Hydrogen production is key to enabling nuclear reactors to participate in the hydrocarbon energy cycle Disposal Coal Grid Solid Wastes Coal Fired Power Plant CO 2 Oxygen Methanol Production 250 C - 5 MPa Water Hydrogen CH 3 OH Hydrogen Water Methanol to Gasoline Gasoline C n H 2n+2 Heat 650ºC Dissolution to Form Uranyl Tricarbonate 24NH (+) 4 + 6UO 2 (CO 3 ) (-4) 3 + 3Na 2 CO 3 Room Temperature Remove Hydrogen Anion Resin Exchange to Thermal Reduction of H 2 O and Oxidation of Uranium Separate Sodium Carbonate from Ammonium Uranyl Tricarbonate 2U Thermal Decomposition (<100ºC) 3 O 8 + 2H 2 O + 3Na 2 CO 3 3Na 2 U 2 O 7 + 2H 2 (g) + 3CO 2 (g) of Ammonium Carbonate and Recycle Water 12(NH 4 )2CO 3 + 6CO 2 (g) Resin Elution to Recover Tricarbonate and Regenerate Resin in Carbonate Form High Temperature Reactor Heat Uranium Carbonate Hydrogen Production 650 C Thermal Decomposition of Tricarbonate to Form Oxygen and Regenerate U 3 O 8 24NH (+) 4 + 6UO 2 (CO 3 ) (-4) 2U 3 O NH 3 (g) + 18CO 2 (g) + 12H 2 O(g) + O 2 (g) 3 Heat ~400ºC High temperature reactors can produce gasoline while removing carbon dioxide from the atmosphere Remove Oxygen US Pat. 7,666,387 B2 Feb Carbonate thermochemical hydrogen cycle efficiently couples to FHR heat 7 Managed by UT-Battelle

8 FHR Safety Derives from Inherent Material Properties and Sound Design Inherent Large temperature margin to fuel failure Good natural circulation cooling Large negative temperature reactivity feedback High radionuclide solubility in salt Low pressure Engineered High quality fuel fabrication Effective decay heat sinking to environment Passive, thermally driven negative reactivity insertion Multi-layer containment 8 Managed by UT-Battelle

9 DRACS Rejects Decay Heat to Ambient Air In the Event of a Loss of Forced Flow Accident Loss of forced flow decay heat removal is via three, independent natural convection driven Direct Reactor Auxiliary Cooling System (DRACS) loops DRACS primary heat exchangers are located in vessel downcomer Bypass flow through DRACS is minimized by fluidic diode below heat exchanger DRACS employs three coupled buoyancy driven loops (FLiBe, KF- ZrF 4, and air) Each DRACS is sized to reject 0.25% (8.5 MW) full power at operating temperature under fully developed flow Primary Coolant Outflow Primary Coolant Return Flow DRACS Heat Exchanger (1 of 3) FLiBe Core Air Out KF-ZrF 4 Air In Fluidic Diode DRACS Flow Path 9 Managed by UT-Battelle

10 Strong FHR Safety Case Makes Licensing Through Evolutionary Change to NRC Process Possible NRC Has Very High Inertia and Revolutionary Change is Unlikely to Succeed 10CFR Part 50 Appendix A provides set of general design criteria (GDC) Current Appendix A criteria are LWR focused FHRs will require a custom set of GDCs Gas cooled reactors and liquid metal cooled reactors are already developing modified GDCs in the form of ANS safety standards (ANS 53.1 & ANS 54.1) FHR safety standard will provide the FHR customized set of GDCs Organizing meeting set for immediately following president s reception at the summer 2012 ANS meeting Will require NRC endorsement Will provide guidance on experimental demonstrations required Design specific safety analysis report will show how GDC are fulfilled E.g. - Fusible links versus buoyancy drive in negative reactivity insertion system FHR specific GDCs combined with safety standard will guide development of FHR specific standard review plan NUREG 0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition 10 Managed by UT-Battelle

11 TRISO is the Only Near-Term High Temperature Fuel Technology AGR testing spans power density anticipated for FHRs Fuel Particle 11 Managed by UT-Battelle TRISO = tri structural isotropic FIMA = fissions per initial metal atom

12 Liquid Cooling and Passive Decay Heat Removal Keeps Fuel Well Below Failure Temperature Fuel failure requires uncovering fuel or blocking core flow Large coolant volumetric change with temperature provides strong natural circulation cooling to keep fuel Source IAEA-TECDOC Managed by UT-Battelle Coolant viscosity decrease with temperature increases flow to hot spots Forced flow is co-directional with natural circulation through core avoiding flow reversal requirement

13 FHR Concepts Are Being Developed For Diverse Applications AHTR (1500 MWe) PB-FHR (410 MWe) SmAHTR (125 MWt) AHTR = Advanced High Temperature Reactor PB-AHTR = Pebble Bed Advanced High Temperature Reactor SmAHTR = Small Modular Advanced High Temperature Reactor 13 Managed by UT-Battelle

14 AHTR Concept is Primary DOE-NE Program Focus Advanced High Temperature Reactor (AHTR) is ORNL s design concept for a central station type (1500 MW e ) FHR Objective is to demonstrate the technical feasibility of FHRs as lowcost, large-size power producers while maintaining full passive safety Significant developments remain in almost all aspects of the reactor Recent program technical reports are available for download from the DOE Office of Scientific and Technical Information (OSTI) Core and Refueling Design Studies for the Advanced High Temperature Reactor Advanced High Temperature Reactor Systems and Economic Analysis AHTR Sectional View 14 Managed by UT-Battelle AHTR Plant Layout

15 AHTR is Progressing Towards a Preconceptual Design Level of Maturity Both reactor and power plant systems are included in the modeling Direct Reactor Auxilary Cooling System Heat Exchanger Decay Heat Cooling Tower Natural Draft Heat Exchanger Core Vessel Pump Primary to Intermediate Heat Exchanger Intermediate to Power Cycle Heat Exchanger Generator Turbine Condenser Cooling Tower Thermal Power Electrical Power Top Plenum Temperature Coolant Return Temperature AHTR Properties Number of loops MW 1500 MW 700 C 650 C Primary Coolant 2 7 LiF-BeF 2 Fuel Uranium Enrichment 9% Fuel Form Refueling UCO TRISO Plate Assemblies 2 batch 6 month 15 Managed by UT-Battelle

16 SmAHTR is A Cartridge Core, Integral- Primary-System FHR Overall System Parameters Parameter Value Power (MWt) 125 Primary Coolant 2 7 LiF-BeF 2 Primary Pressure (atm) ~1 Core Inlet Temperature (ºC) 650 Core Outlet Temperature (ºC) 700 Core coolant flow rate (kg/s) 1020 Operational Heat Removal Passive Decay Heat Removal Reactor Vessel Penetrations 3 50% loops % loops None 16 Managed by UT-Battelle

17 SmAHTR Design Shows Promise for High- Temperature Heat Production Small, modular Advanced High Temperature reactor (SmAHTR) has been designed for modular, factory fabrication, and truck transport 9 m 125 MW th Plate assembly fuel Cartridge core Integral primary heat exchangers Technology development requirements for small and large FHRs is virtually identical 3.6 m 17 Managed by UT-Battelle

18 Optimal Power Conversion Cycle is Not Obvious Reheated Supercritical Water Selected as Baseline AHTR Power Conversion Cycle Supercritical CO 2 Not mature or scaled Corrosion concerns Helium or helium-nitrogen Large components Open air Lower efficiency at 700 C Subcritical steam Mature Good cost models Supercritical water Highest proven efficiency Highest pressure 18 Managed by UT-Battelle

19 Modular Assembly Featuring Steel Plate Concrete Forms Employed to Shorten Schedule Steel plate concrete forms prefabricated off-site will save cost and schedule Assembled together and welded on-site where concrete is poured No rebar fabrication on-site Much of the site worked performed in local workshop General Dynamics rule-ofthumb ratios of time as 1:3:9 for factory to workshop to insitu fabrication 19 Managed by UT-Battelle

20 AHTR Core Consists of 252 Identical Hexagonal Fuel Assemblies Core surrounded by replaceable and permanent graphite reflector columns Fueled core height 5.5 m Total core height 6.0 m Fuel assembly pitch cm Equivalent fueled core diameter 7.81 m Volumetric power density 12.9 MW/ m 3 Core makes extensive use of carbon fiber and silicon carbide fiber composites Molybdenum alloy control blade is only metallic material in core 20 Managed by UT-Battelle

21 AHTR Design Calls for Coated Particle Plate Fuel Assemblies Coated particle fuel is a uranium oxycarbide variant currently being qualified under DOE-NE Advanced Gas Reactor (AGR) program Fuel particles are configured into stripes just below the surface of the fuel plates Minimizes heat conduction distance to coolant Fuel plates have a 5.5 m fueled length Fuel assemblies are surrounded by a C- C composite shroud to channelize coolant flow 21 Managed by UT-Battelle Fuel Plate Cross Section

22 AHTR Employs Ceramic Composite Core Support Plates to Minimize Temperature Impact Lower core support plate anchored to the vessel Accommodates differential thermal expansion between the vessel and the plate Flow channels are provided through lower core support plate to direct flow into the core Upper core support plate connected to top flange Upper core support plate raised during refueling Both plates are made from SiC-SiC composite 22 Managed by UT-Battelle Lower core support plate (50 cm) Upper core support plate (30 cm)

23 Mechanically Integrated Design of the AHTR Reactor Building Systems and Structures is Underway Focus is on integrating necessary systems, structures, and components A design focus is on maximizing the system economic performance Employing modular, open-top construction to minimize cost Availability increased through automated refueling and ease of access for inspection and maintenance Maintaining full passive safety when subjected to severe environmental challenges 23 Managed by UT-Battelle

24 AHTR Plant Layout Includes Construction Scheduling Reactor Building Fuel Handling Building Workshop Heavy Lift Crane 24 Managed by UT-Battelle Railroad Spur

25 Low Pressure Containment is Located Within the Shield Building Under normal operations the containment building serves as the outer boundary for beryllium and tritium Under accident conditions containment building provides radioactive material barrier All areas within the containment building have argon atmosphere All areas within the shield building, not within containment, have dry air atmosphere (e.g. electrical switch panel and manipulator maintenance areas) Tops of the shield building and containment building will be open for construction and for major maintenance Reactor component cooling system and cavity cooling system will use a forced flow argon system for cooling during normal operations Heat sink is provided through a chilled nitrogen cooling loop, which passes through containment For loss of forced flow accidents cavity passive cooling will be through the reactor building steel walls to the air trench surrounding the shield building Cavity heat load is small without pumps operating due to well insulated primary system To avoid the potential to overpressure containment due to phase change, no large-volumes of water will be used within containment 25 Managed by UT-Battelle

26 Gen IV Economic Model Indicates that AHTRs Have the Potential For Lower LUEC than PWRs Very significant uncertainty remains in the cost estimates Levelized unit cost output from G4-ECONS (mills/kwh) Reactor system PWR 12 better AHTR experience Capital cost recovery Operation and maintenance Fuel cycle costs Decommissioning fund Levelized unit cost of electricity Fuel cycle costs, O&M cost, 9.31 D&D fund, 0.23 Capital cost recovery, Total capital investment cost, $/kw(e) Managed by UT-Battelle

27 DOE-NE Has Awarded an FHR Focused University Integrated Research Project Massachusetts Institute of Technology Pre-conceptual design and material testing at the MIT research reactor University of California at Berkeley Thermal hydraulics and neutronics Safety and licensing University of Wisconsin Materials and corrosion Additional individual university FHR technology development projects are also underway The Ohio State University Direct Reactor Auxiliary Cooling System design and testing University of California at Berkeley Pebble bed fuel motion modeling and demonstration using simulant materials 27 Managed by UT-Battelle

28 FHR Material and Component Design Studies are Also Continuing Intermediate loop to power cycle heat exchanger is key component for successful FHR deployment Feasibility Study of Secondary Heat Exchanger Concepts for the Advanced High Temperature Reactor recently published (INL/ EXT ) Available at Assessment of current status of Alloy N for salt reactor deployment recently published Considerations of Alloy N for Fluoride Salt- Cooled High Temperature Reactor Applications ASME 2011 Pressure Vessels & Piping Division Conference Presentation available at info.ornl.gov/sites/publications/files/pub31145.pdf Cladding qualified structural alloys is a near-term approach to enable higher temperatures NGNP program is qualifying higher temperature structural alloys (800H, 617, perhaps 230) Cladding Alloys for Fluoride Salt Compatibility recently published (ORNL/ TM-2011/95 available on-line from OSTI) 28 Managed by UT-Battelle

29 U.S. is Providing Isotopically Selected Fluoride Salt to Czech Republic in Return for Criticality Measurements Criticality measurements provide confidence in neutronics predictions Fuel cycle length Reactivity feedbacks 29 Managed by UT-Battelle

30 Versatile Liquid Salt Loop Has Been Constructed To Support FHR Component Development Heat Exchanger Surge Tank Storage Tank Pump Sump Tank Experimental facility will include a salt purification system designed to remove moisture and oxides in the salt to minimize corrosion A fluidic diode (a leaky check valve with no moving parts) will be tested early in the experimental program Key decay heat removal component 30 Managed by UT-Battelle Follow on testing will focus on scaled AHTR components such as: Fuel heat transfer testing Improved pump designs Salt-to-salt or salt-to-gas heat exchanger Instrumentation Refueling components

31 FHR Reactor Class Shows Much Promise Still Requires Significant Research, Development, and Demonstration More complete reactor conceptual design required Needs to include all of the specialized systems and components Refueling mechanisms remain to be designed Replacement industrial scale lithium enrichment Salt chemistry control system requires basic design Structural ceramics must become safety grade nuclear engineering materials Process instrumentation requires further development Safety and licensing approach must be developed and demonstrated Plate fuel must be qualified 31 Managed by UT-Battelle