Fluoride Salt Cooled High Temperature Reactors Workshop on Advanced Reactors PHYSOR 2012 Knoxville, TN April 15, 2012 David Holcomb [email protected]
Fluoride Salt Cooled High Temperature Reactors
Workshop on Advanced Reactors PHYSOR 2012
Knoxville, TN April 15, 2012
David Holcomb
2 Managed by UT-Battelle for the U.S. Department of Energy
FHRs Combine Desirable Attributes From Other Reactor Classes
Fluoride Salt Cooled Reactors• High temperature• Low pressure• Passive safety
Advanced Coal Plants• Supercritical water
power cycle• Structural alloys
Gas Cooled Reactors• TRISO fuel• Structural ceramics• High temperature power
conversion
Molten Salt Reactors• Fluoride salt coolant• Structural alloy• Hydraulic components
Light Water Reactors• High heat capacity
coolant• Transparent coolant
Liquid Metal Reactors• Passive decay heat
removal• Low pressure design• Hot refueling
3 Managed by UT-Battelle for the U.S. Department of Energy
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 – …
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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)
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Potential Benefits And Challenges Of FHRs Stem From Fundamental Coolant Properties
Coolant Tmelt (ºC)
Tboil (ºC)
Density (kg/m3)
Specific Heat
(kJ/kg K)
Volumetric Heat
Capacity (kJ/m3 K)
Thermal Conductivity
(W/m K)
Kinematic Viscosity
(m2/s) * 106
Li2BeF4 (Flibe) 459 1430 1940 2.42 4670 1.0 2.9
59.5NaF-40.5ZrF4 500 1290 3140 1.17 3670 0.49 2.6
26LiF-37NaF-37ZrF4 436 2790 1.25 3500 0.53
31LiF-31NaF-38BeF2 315 1400 2000 2.04 4080 1.0 2.5
8NaF-92NaBF4 385 700 1750 1.51 2640 0.5 0.5
Sodium 97.8 883 820 1.27 1040 62 0.12
22Na-44K -11 784 7420 0.87 6455 26.8 0.24
56Na-44K 19 826 7590 1.04 7894 28.4 0.25
Lead 328 1750 10540 0.16 1700 16 0.13
44.5Pb-55.5Bi 98 881 10020 0.15 1503 13.9 0.12
Helium, 7.5 MPa 3.8 5.2 20 0.29 11
Water, 7.5 MPa 0 290 732 5.5 4040 0.56 0.13
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Potential FHR Operating Temperatures Match Many Important Process Heat Applications
0 100 200 300 400 500 600 700 800 900 1000 110 1200 1300 1400 1500 1600 1700
Cogeneration of Electricity and Steam
Steam Reforming of Nat. Gas & Biomass Gasification
H2 Production & Coal Gasification
NaF-BeF2 (57-43)
LiF-NaF-KF (46.5-11.5-42)
LiF-BeF2 (67-33)
Temperature (ºC)
Melts Boils
Oil Shale/Sand Processing
Fluoride Salt Liquid Temperature Range
Petro Refining
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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 Dissolution to FormUranyl Tricarbonate
Anion Resin Exchange toSeparate Sodium Carbonate fromAmmonium Uranyl Tricarbonate
Thermal Decomposition (<100ºC)of Ammonium Carbonate
and Recycle
Resin Elution to RecoverTricarbonate and Regenerate
Resin in Carbonate Form
Thermal Decomposition of Tricarbonate toForm Oxygen and Regenerate U3O8
RemoveOxygen
Thermal Reduction of H2Oand Oxidation of Uranium
RoomTemperature Remove
Hydrogen
Heat650ºC
Water
Heat~400ºC
2U3O8 + 2H2O + 3Na2CO3 → 3Na2U2O7 + 2H2(g) + 3CO2(g)
12(NH4)2CO3 + 6CO2(g)
24NH4(+) + 6UO2(CO3)3
(-4) + 3Na2CO3
24NH4(+) + 6UO2(CO3)3
(-4)2U3O8 + 24NH3(g) + 18CO2(g) + 12H2O(g) + O2(g)
Dissolution to FormUranyl Tricarbonate
Anion Resin Exchange toSeparate Sodium Carbonate fromAmmonium Uranyl Tricarbonate
Thermal Decomposition (<100ºC)of Ammonium Carbonate
and Recycle
Resin Elution to RecoverTricarbonate and Regenerate
Resin in Carbonate Form
Thermal Decomposition of Tricarbonate toForm Oxygen and Regenerate U3O8
RemoveOxygen
Thermal Reduction of H2Oand Oxidation of Uranium
RoomTemperature Remove
Hydrogen
Heat650ºC
Water
Heat~400ºC
2U3O8 + 2H2O + 3Na2CO3 → 3Na2U2O7 + 2H2(g) + 3CO2(g)
12(NH4)2CO3 + 6CO2(g)
24NH4(+) + 6UO2(CO3)3
(-4) + 3Na2CO3
24NH4(+) + 6UO2(CO3)3
(-4)2U3O8 + 24NH3(g) + 18CO2(g) + 12H2O(g) + O2(g)
Solid
Wastes
Disposal
Methanolto
Gasoline
High Temperature
ReactorHeat
Grid Hydr
ogen
GasolineCnH2n+2
Oxygen
Hydrogen WaterWater
CO2CH
3 OH
Uranium Carbonate Hydrogen Production
650 °C
Methanol Production
250 °C - 5 MPa
Coal Fired Power Plant
Coal
Carbonate thermochemical hydrogen cycle efficiently couples to FHR heat
High temperature reactors can produce gasoline while removing carbon dioxide from the atmosphere
US Pat. 7,666,387 B2 Feb. 2010
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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
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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-ZrF4, and air)
■ Each DRACS is sized to reject 0.25% (8.5 MW) full power at operating temperature under fully developed flow
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Strong FHR Safety Case Makes Licensing Through Evolutionary Change to NRC Process Possible
■ 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
NRC Has Very High Inertia and Revolutionary Change is Unlikely to Succeed
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TRISO is the Only Near-Term High Temperature Fuel Technology
Fuel Particle
AGR testing spans power density anticipated for FHRs
TRISO = tri structural isotropic FIMA = fissions per initial metal atom
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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
■ Coolant viscosity decrease with temperature increases flow to hot spots
■ Forced flow is co-directional with natural circulation through core avoiding flow reversal requirement
Source IAEA-TECDOC-978
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FHR Concepts Are Being Developed For Diverse Applications
AHTR = Advanced High Temperature Reactor PB-AHTR = Pebble Bed Advanced High Temperature Reactor SmAHTR = Small Modular Advanced High Temperature Reactor
PB-FHR (410 MWe)
SmAHTR (125 MWt)
AHTR (1500 MWe)
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AHTR Sectional View
■ Advanced High Temperature Reactor (AHTR) is ORNL’s design concept for a central station type (1500 MWe) FHR
■ Objective is to demonstrate the technical feasibility of FHRs as low-cost, 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 Plant Layout
AHTR Concept is Primary DOE-NE Program Focus
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AHTR is Progressing Towards a Preconceptual Design Level of Maturity
Both reactor and power plant systems are included in the modeling
AHTR Properties
Thermal Power 3400 MW
Electrical Power 1500 MW
Top Plenum Temperature
700 °C
Coolant Return Temperature
650 °C
Number of loops 3
Primary Coolant 27LiF-BeF2
Fuel UCO TRISO
Uranium Enrichment 9%
Fuel Form Plate Assemblies
Refueling 2 batch 6 month
Vessel
Core
Pump
Prim
ary
to
Inte
rmed
iate
H
eat E
xcha
nger
CoolingTower
Inte
rmed
iate
to
Pow
er C
ycle
H
eat E
xcha
nger
Gen
erat
orTu
rbin
e
Decay Heat Cooling Tower
Natural DraftHeat Exchanger
Direct ReactorAuxilary Cooling
System HeatExchanger
Con
dens
er
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SmAHTR is A Cartridge Core, Integral-Primary-System FHR
Parameter Value
Power (MWt) 125
Primary Coolant 27LiF-BeF2
Primary Pressure (atm) ~1
Core Inlet Temperature (ºC) 650
Core Outlet Temperature (ºC) 700
Core coolant flow rate (kg/s) 1020
Operational Heat Removal 3 – 50% loops
Passive Decay Heat Removal 3 – 0.25% loops
Reactor Vessel Penetrations None
Overall System Parameters
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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 – 125 MWth – Plate assembly fuel – Cartridge core – Integral primary heat
exchangers
■ Technology development requirements for small and large FHRs is virtually identical
3.6 m
9 m
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Optimal Power Conversion Cycle is Not Obvious
■ Supercritical CO2 – 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
Reheated Supercritical Water Selected as Baseline AHTR Power Conversion Cycle
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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-of-
thumb ratios of time as 1:3:9 for factory to workshop to in-situ fabrication
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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 46.75 cm ■ Equivalent fueled core diameter
7.81 m ■ Volumetric power density 12.9 MW/
m3
■ Core makes extensive use of carbon fiber and silicon carbide fiber composites – Molybdenum alloy control blade is only
metallic material in core
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AHTR Design Calls for Coated Particle Plate Fuel Assemblies ■ Coated particle fuel is a uranium oxy-
carbide 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
Fuel Plate Cross Section
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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
Lower core support plate (50 cm)
Upper core support plate (30 cm)
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■ 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
Mechanically Integrated Design of the AHTR Reactor Building Systems and Structures is Underway
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Fuel Handling Building
Heavy Lift Crane
Reactor Building
Workshop
Railroad Spur
AHTR Plant Layout Includes Construction Scheduling
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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
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Gen IV – Economic Model Indicates that AHTRs Have the Potential For Lower LUEC than PWRs
Very significant uncertainty remains in the cost estimates
Capital cost
recovery, 22.77 O&M cost,
9.31
Fuel cycle costs, 10.74
D&D fund, 0.23
Levelized unit cost output from G4-ECONS (mills/kWh)
Reactor system PWR 12
better experience
AHTR
Capital cost recovery 29.66 22.77 Operation and maintenance
12.60 9.31
Fuel cycle costs 5.60 10.75 Decommissioning fund 0.32 0.23 Levelized unit cost of electricity
48.18
43.05
Total capital investment cost, $/kW(e)
4012
3149
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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
28 Managed by UT-Battelle for the U.S. Department of Energy
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-11-23076) – Available at www.inl.gov/technicalpublications/Documents/5144351.pdf
■ 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)
29 Managed by UT-Battelle for the U.S. Department of Energy
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
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Versatile Liquid Salt Loop Has Been Constructed To Support FHR Component Development
■ 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
Surge Tank Heat Exchanger
Pump Sump Tank
Storage Tank
■ 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
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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