Fluoride Salt Cooled High Temperature...

Fluoride Salt Cooled High Temperature Reactors

Workshop on Advanced Reactors PHYSOR 2012

Knoxville, TN April 15, 2012

David Holcomb

[email protected]

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


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 –  …


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)


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


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


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


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


��

��

��

��

��"��

�� !�$��

��

� �� %��!��!" ��#

��

��#��!�

� �� %��!�"!��#

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


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


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


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


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)


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


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


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


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


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


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


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


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


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)


■  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


Fuel Handling Building

Heavy Lift Crane

Reactor Building

Workshop

Railroad Spur

AHTR Plant Layout Includes Construction Scheduling


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


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


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


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)


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


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


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

Date post:	06-Apr-2018
Category:	Documents
Upload:	dotram
View:	218 times
Download:	3 times