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Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
Liquid Fluoride Reactors: A New Beginning for an Old Idea
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Liquid Fluoride Reactors: A New Beginning for an Old Idea

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Google Tech Talk by Dr. David LeBlanc about Liquid Fluoride Reactors. These are a class of nuclear reactor invented in the 1950s, abandoned in the 1970s, and becoming more interesting today.

Google Tech Talk by Dr. David LeBlanc about Liquid Fluoride Reactors. These are a class of nuclear reactor invented in the 1950s, abandoned in the 1970s, and becoming more interesting today.

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  • 1. Liquid Fluoride Reactors: A New Beginning for an Old Idea Feb 19 th 2009 Google Tech Talk Dr. David LeBlanc Physics Dept, Carleton University, Ottawa & Ottawa Valley Research Associates Ltd. [email_address]
  • 2. Can Nuclear Help Solve our Energy and Climate Change Problems?
    • The nuclear industry would be more than happy with building 400 reactors over the next 30 years
      • But that merely keeps the status quo
    • Thousands of large reactors are needed to truly make a difference.
    • Is there any way that existing or “ next generation ” designs can realistically accomplish this?
  • 3. Seven Reactor Design Priorities
    • Overall Power Costs
      • Capital Costs
      • Fuel plus O&M
    • Safety
    • Resource Utilization
      • Quantity and Quality of Starting Load
      • Annual Requirements
    • Proliferation Resistance
      • Subversion to a State Program
      • Theft and/or Terrorist Take Over
    • Long Term Waste Radiotoxicity
    • R&D Requirements
    • Capability for Rapid Deployment
  • 4. PWRs, BWRs and CANDUs
    • Overall Costs
      • High capital but arguably competitive with coal
    • Safety
      • Good through costly “Defence in Depth”
    • Resource Utilization
      • Moderate for startup, Very poor annual
    • Proliferation Resistance
      • Good if international safeguards maintained
    • Long Term Wastes
      • Once Through cycle extremely poor, Pu recycle not much better
    • R&D - Very Little Required
    • Capability for Rapid Deployment
      • Labour requirements alone limit growth
  • 5. What About Gas Cooled Pebble Beds?
    • Costs savings possible but no great reduction in overall complexity
    • Massive pressure vessel and needs of decay heat removal limit core size and power
    • Rad Waste production very similar (with radioactive graphite as an added bonus)
    • Reprocessing of fuel extremely difficult, so an even worse outlook for uranium resources
    • From safety outlook, excellent for decay heat management but criticality events still a possible issue*
    *The reverse situation is true for Accelerator Driven Systems which guard against criticality accidents but do not address decay heat
  • 6. What about other Generation IV Reactors?
    • Supercritical Water (LWR or CANDU)
      • Merely modest improvements possible and extremely challenging corrosion issues to solve
    • Sodium Cooled Fast Breeders
      • More costly and higher safety concerns
      • ~12 tonne fissile requirement
      • Doubling time upwards of 40 years
    • Lead Cooled Fast Breeders
      • Possible lower capital costs
      • ~20 tonne fissile requirement makes rapid fleet expansion even less likely*
    • Gas Cooled Fast Breeders
      • Similar cost advantages and fissile needs of Lead Cooled Fast Breeder
    *Entire U.S. supply of Plutonium from spent fuel and weapons program could only start 30 to 40 such reactors
  • 7. What about Thorium? We keep hearing about that
    • Yes, fertile thorium producing fissile 233 U has numerous advantages
    • Works well in any spectrum, not just a fast one like U-Pu
    • However any use in solid fuels, involves very difficult fuel reprocessing
    • What about fluid fuels? What about the Gen IV reactor that was made for thorium?
    • The Liquid Fluoride Reactor * Over 50 years old and never looked better
    * Historically known as Molten Salt Reactors
  • 8. Liquid Fluoride / Molten Salt Overview
    • What did the original Molten Salt Reactor Program at Oak Ridge National Labs (1950s to 1970s) end up with?
    • What factors guide design, how did the design evolve and why was it cutoff?
    • What has happened since?
    • Do modified priorities and new ideas change the outlook?
  • 9. The Single Fluid, Graphite Moderated Molten Salt Breeder Reactor (MSBR)
  • 10. General Attributes of the “Traditional” 1000 MWe Single Fluid MSBR Design
    • 2 7 LiF-BeF 2 - 12% ThF 4 - 0.3% 233 UF 4
    • 700 o C Outlet Temperature
    • 44.4% on Steam Rankine cycle
    • Up to 48% on Gas Brayton cycle
    • Fission Products removed on 20 day cycle
    • 233 Pa* removed on 10 day cycle
    • Startup Fissile Inventory of 1500 kg
    • Breeding Ratio of 1.06**
    • 20 year doubling time
    * 233 Pa is the 27 day half-life intermediate between thorium and 233 U ** Meaning it produces 6% more fissile fuel than it consumes
  • 11. General Benefits of Any Molten Salt Design
    • Volatile fission products including Xenon simply bubble out*, are removed and stored elsewhere. A spill will simply solidify
    • Salts have high boiling point and operate at low pressure
    • A temp increase instantly lowers reactivity
    • Fuel salt at the lowest pressure of the circuit, the opposite of a LWR
    • Control rods or burnable poisons not required. Almost no excess reactivity
    • A simple freeze plug melts upon fuel overheating to drain the salt to critically safe, passively cooled dump tanks
    *Assisted by bubbling helium through the flowing salt
  • 12. General Benefits of Any Molten Salt Design
    • Inherently Passive Safety
    • Low Fissile Inventory
    • Very High Thermal Efficiency
    • Ideal for LWR TRU waste destruction
    • Ability to use closed thorium cycle
      • Only consume 800 kg thorium per GWe year
      • TRU (transuranic = Np, Pu, Am, Cm) waste production extremely low
      • Much lower long term radiotoxicity
  • 13. Radiotoxicity PWR vs FBR * vs LFR * * Assuming 0.1% Loss During Processing Data and graph from Sylvain David, Institut de Physique Nucléaire d'Orsay PWR Ore Levels Turns waste management into 500 year job, not million year FPs Fission Products
  • 14. Proliferation Resistance The Pure Thorium – 233 U Cycle
    • 232 U present in significant quantities
      • 69 year half-life with strong 2.6 MeV gamma ray from daughter product 208 Tl
      • Makes illicit use difficult and highly detectable
      • No national program ever based on 233 U
    • 233 U can be instantly denatured by dumping 238 UF 4 into the molten fuel salt
    • 233 Pa removal can lead to “clean” 233 U and thus should be avoided
    • Only small amounts Plutonium are present, it is of poor quality and very hard to extract
    *HEU=Highly Enriched Uranium (>12% 233 U), LEU=Low Enriched Uranium
  • 15. Proliferation Resistance Denatured Cycles
    • The pure Th- 233 U cycle does represent the use of Highly Enriched Uranium
    • Running denatured by including 238 U makes any uranium useless for weapons
    • It does mean more plutonium present but still of poor quality and hard to remove
    • In many designs the conversion ratio will suffer somewhat
  • 16. Quality of Produced Plutonium 7.4% 24.3% 5.2% fertile 242 Pu 4.8% 13.6% 5.6% Fissile and adds hard gamma rays 241 Pu 4.4% 18.1% 24.3% Spontaneous fissions high 240 Pu 9.5% 31.1% 60.3% Main fissile Component 239 Pu 73% 12.6% 1.3% Generates heat from alpha emission 238 Pu MSBR Pure Th – 233 U cycle DMSR 30 Year Once Through PWR Reactor Grade Proliferation properties Isotope
  • 17. Problems Specific to the Single Fluid, Graphite Moderated MSBR
    • Limited graphite lifetime (4 years)
    • Fuel processing hindered by chemical similarity of thorium and the rare earth fission products
    • Problem with temperature reactivity coefficient recently discovered
    • 233 Pa separation adds a unique proliferation concern
  • 18. Design Basics: 2 Fluid Versus Single Fluid
    • A 2 Fluid design has separate fluids
      • A Fuel Salt containing 233 UF 4
      • A Blanket Salt containing ThF 4
    • Advantages;
      • Absence of thorium in fuel salt makes fission product removal far easier
      • 233 Pa removal unnecessary*
      • Strongly negative temperature and void coefficients for the fuel salt
    *Pa is effectively diluted in the blanket salt so it experiences a lower neutron flux and has a much greater probability to decay to 233 U
  • 19. Converter or “Once Through” Designs
    • Use a mix of Low Enriched Uranium and thorium in a Single Fluid
    • No need for any processing of fuel salt for the life of the plant
    • Not able to “Break Even” but require only a small fraction the annual uranium compared to LWR, CANDU or PBMR
  • 20. Choice of Neutron Spectrum
    • Harder Spectrum
      • Losses to fission products drop so may leave in longer
      • Prompt neutron lifetime is shorter
      • Typically requires higher fissile load
    • Softer Spectrum
      • Fissile inventory can be very low
      • Losses to fission products highest
      • Any extra structural material must be very limited or breeding not possible
      • Prompt neutron lifetime is increased
  • 21. Fuel Processing Basics
    • Fluoride Volatility Process
      • Removes uranium from carrier salt
    • Fission Product Removal
      • Essential if breeding desired
      • Several year cycle possible in some designs
    • Protactinium Removal and Hold Up
      • Essential only for Single Fluid designs wishing to breed or break even
  • 22. Fluoride Volatility
    • Simple bubbling of fluorine gas through carrier salt
    • Converts UF 4 to gaseous UF 6
    • Conversion back to UF 4 done by adding H 2
    • Well established process, known since 1950s
  • 23. Fission Product Removal
    • 1950 to 1964
      • Several potential methods but none proven to be practical
    • 1964 Vacuum Distillation
      • Simple system to evaporate off carrier salt and leave behind fission products
      • Thorium would remain with fission products, so only for a 2 fluid design with no thorium in the fuel salt
  • 24. Fission Product Removal
    • 1968 Liquid Bismuth Reductive Extraction
      • Ability to process for fission products with thorium present
      • Thorium behaves much like rare earth fission products which makes the process very “delicate”
      • Complex and costly but works for Single Fluid designs
  • 25. Fuel Processing Summary
    • For 2 Fluid design, choice of:
      • Vacuum Distillation
      • Salt Replacement
      • Simplified Liquid Bismuth Extraction
      • Can skip Pa removal since it resides in the low neutron flux blanket
    • For Single Fluid Design
      • Multi-staged Liquid Bismuth Extraction
      • Pa removal very difficult to avoid
    • Skip altogether for converter design
  • 26. Design Evolution at Oak Ridge National Laboratories (ORNL)
    • Crucial to remember that early design priorities very different
    • Long Term Wastes and Proliferation Resistance low priority
    • Uranium resources thought scarce, thus must “breed” fuel to expand reactor fleet
    • Short Doubling Time* top priority in competition with Sodium Cooled Fast Breeder
    * Time to breed enough excess fuel to start another reactor
  • 27. A Strange Beginning An Aircraft Reactor?
  • 28. The Aircraft Reactor Experiment
    • Test reactor of early to mid 1950s
    • Very high temperature 860 o C
    • Canned BeO moderator
    • NaF-ZrF 4 carrier salt
    • Points the way to possible power reactors (even if the idea of an airborne reactor far fetched)
  • 29. The Rise and Fall of the 2 Fluid Reactor The Late 1950s
    • Main focus on homogeneous reactors (no graphite or other solid moderator)
    • Looked at both converter reactors and thorium- 233 U breeders
    • Carrier Salt itself provides significant neutron moderation so a variety of neutron spectrums possible
    • All studies were spherical geometry with 1/3 inch Hastelloy N core walls
  • 30. Homogenous Molten Salt Reactor Late 50s ORNL
  • 31. The Rise and Fall of the 2 Fluid Reactor Early 1960s
    • Graphite now proven to be compatible with fluoride salts
    • Lower possible fissile inventory leads ORNL to change to a graphite moderated 2 Fluid design
    • Simplist Sphere-Within-Sphere 2 Fluid design would only allow ~ 1 meter core*
      • Obviously too low a total power output
    • Complex fluid intermixing deemed necessary
    *Without having the 233 UF 4 fraction so low as to preclude breeding
  • 32. 1960`s, 2 Fluid, Graphite Moderated MSBR
    • Intermixing of Fuel and Blanket salt in core done by graphite “plumbing”
    • Allows large core diameter for adequate power production
    • Graphite first shrinks and then swells under neutron irradiation
    • Extremely difficult “Plumbing Problem” to solve
  • 33.  
  • 34. Meanwhile, also in the mid 60s… Molten Salt Reactor Experiment MSRE
    • MSRE 8 MW(th) Reactor
      • Chosen to be Single Fluid for simplicity
      • Graphite moderated, 650 o C operation
      • Designed from 1960 to 1964
      • Start up in 1965
      • Ran very successfully for 5 years
      • Operated separately on all 3 fissile fuels, 233 U , 235 U and Pu
      • Some issues with Hastelloy N found and mostly resolved in later years
  • 35. 1968: The Start of the Single Fluid Era
    • New Liquid Bismuth Reductive Extraction, while more difficult, can work with thorium in the fuel salt
    • 2 Fluid concept abandoned
    • “ Plumbing Problem” left unsolved
    • Major funding for MSBR cancelled in early 70s. R&D extremely limited until quite recently
    • 2 Fluid concept largely forgotten
  • 36. Why was the program cancelled? First, what did ORNL do right ?
    • Early decision to focus on Th-U cycle instead of U-Pu
      • No need for a fast spectrum, less radiotoxicity much better resource potential of thorium over uranium
    • Brought the top minds together in one place
      • Engineers, chemists, physicists constantly update and reinforce each others efforts
    • Reactor safety kept at the forefront
      • Inherently passive over engineered solutions
      • Early recognition that events of low probability must be planned for
  • 37. What did ORNL do wrong ?
    • Early decision to focus on Th-U cycle instead of U-Pu
      • In 50s to 70s, production of Pu still a top military, government priority and by extension the AEC
    • Brought the top minds together in one place
      • MSR almost unknown to experts outside Oak Ridge. Government relies on opinions of experts from across the country
    • Reactor safety kept at the forefront
      • Oak Ridge director and inventor of the PWR, Alvin Weinberg drew the ire of the AEC`s infamous Milton Shaw by raising safety concerns of LWRs
      • Weinberg fired as director and funding for Molten Salt program cut off in the early 1970s. Topic almost forbidden since then by AEC and later DOE
  • 38. Post Program Developments
    • Late 70s ORNL work focuses on Denatured designs termed DMSRs
    • Little funding but many vocal advocates worldwide until being chosen as one of six Gen IV reactors in 2002
    • Major re-examination by large group in France
      • Find a reactivity problem with 70s MSBR
      • Look to decrease processing needs by allowing breeding ratio to drop
      • Move towards a harder spectrum without graphite and add a radial blanket salt
    • U.S. based effort to examine the use of clean salts as coolant for TRISO type fuels. Many advantages if positive void can be avoided
    • Much worldwide work on transuranic waste destruction using Molten Salt designs
  • 39. The French TMSR Thorium Molten Salt Reactor Design has a thorium blanket but only radial, not axially (which would be very difficult)
  • 40. Russian MOlten Salt Actinide Recycler and Transmuter MOSART
  • 41. Updating Priorities
    • Fissile material for starting a large fleet of liquid fluoride reactors not a problem so no need to “breed”
    • Proliferation Resistance and Long Term Waste Reduction now high priorities
    • Should minimise the need for fuel processing and its complexity
    • No two experts or two nations will rank priorities the same so multiple options best
  • 42. Keeping in Mind…
    • KISS, Keep It Simple Stupid
      • Idea people often drawn to complex solutions
      • New improvements are extremely expensive to verify in the nuclear field
    • OTS Off The Shelf where possible
    • RBB Rice Bowl Breaking
      • Unavoidable with new concepts but be aware and minimise if possible!
  • 43. New Ideas for Your Consideration
    • Reviving the 2 Fluid Concept
    • A Barrier Free, Possibly Fuel Processing Free Design
    • Simplified Converters, expanding on ORNL`s “30 Year Once Through” design
  • 44. Restating the 2 fluid Plumbing Problem
    • Sphere-Within-Sphere is simplest approach (or short cylinders)
    • BUT, if inner fuel salt lacks thorium then the critical diameter must remain small (approx 1 m with or without graphite)
    • Too small for power plants
    • “ Standard” conclusion was fuel and blanket salts must be interlaced
  • 45. What is the solution? Here’s a hint… 0.5 Infinite Slab If H = 10 R 0.772 + Finite Cylinder 0.766 Infinite Cylinder 1 Sphere Ratio to B sphere Buckling 2 Geometry
  • 46. Modified Geometry 2 Fluid Reactor* “Tube-Within-Shell” *Patent Pending Expands power producing volume while maintaining the small inner core needed for a simple 2 Fluid design
  • 47. New Concepts Advantages
    • Can use simple 2 fluid fuel processing without the “plumbing problem”
    • Very strongly negative fuel salt coefficients
    • Blanket will also have negative temp/void coefficient as it acts as a weak reflector
    • Simple, transportable cores
    • Ease of modeling and prototyping
    • Fissile inventory of 400 kg per GWe or even lower is possible.
  • 48. Cross Section Graphite Free
  • 49. Example: Graphite Free, Carbon or SiC composite for barrier
    • Using ORNL modeling for a 122 cm “spherical” core, 0.16% 233 UF 4 should be able to reach Break Even Breeding
    • A 122 cm sphere equates to 94 cm diameter in elongated cylindrical geometry
    • Assuming;
      • Core power density of 200 kW/L
      • 2 m/s salt velocity in core
      • Standard 565 C/705 C for Inlet/Outlet Temp
    • Gives 404 MWe (911 MWth), 6.6 m core
  • 50. Other Variations
    • Modestly higher concentration of 233 UF 4 (0.2% to 1%) gives excess neutrons to allow:
      • Metal barriers such as Hastelloy N, Stainless Steels, Molybdenum
      • Alternate carrier salts to reduce costs and tritium production
      • Even greater simplification of fission product processing. 20 year or longer removal time for fission products
  • 51.  
  • 52. Critical Issue: Core-Blanket Barrier
    • Viability of barrier materials in high neutron flux
      • Much recent work in the fusion field using same 2 7 LiF-BeF 2 salt as coolant
      • Molybdenum, SiC/SiC or simple carbon composites leading candidates
      • Hastelloy N and Stainless Steels possible with a modest temp reduction
      • Ease of “retubing” means even a limited lifetime still may be attractive
  • 53. Fusion Structural Materials Studied “ Operating Temperature Windows for Fusion Reactor structural Materials” Zinkle and Ghoniem, 2000
  • 54. Graphite Cores
    • Using graphite in the core leads to a similar inner critical diameter
    • Many advantages but much lower power density than pure salt cores
    • The following embodiment does not require an encompassing barrier and only graphite experiences a significant neutron flux
  • 55.  
  • 56.  
  • 57. Barrier Free, Fuel Processing Free?
    • Can we run break even without a barrier AND without fuel processing?
    • A harder spectrum without graphite lowers losses to fission products but neutron leakage can be excessive without a blanket salt and barrier
    • As well, the outer vessel wall needs protection against a high neutron flux
    • A common solution is a graphite liner/reflector
      • This leads to an increase in power near reflector and limited graphite lifetime
      • Can surprisingly increase neutron leakage
    • Boron often added at outer edge of the graphite reflector to stop final neutrons
  • 58. Power Distribution With and Without Graphite Reflector Reproduced from “Transmutation Capability of Molten Salt Reactors Feed with TRUs from LWR” M. Frantoni and E. Greenspan, AWRIF 2005
  • 59.  
  • 60. Liner = Blanket Concept
    • Core salt is Single Fluid (potentially denatured with 238 U)
    • In early years of operation the liner is completely absorbing but there are excess neutrons since few fission products
    • As fissile 233 U (and/or Pu) build up, the liner begins to send back neutrons which helps compensate for losses to fission products
    • Ideally the liner is permanent until the end of plant life or perhaps a major mid-life refurbishment. Contained fissile in liner and salt can then be collected for the replacement plant
  • 61. Liner = Blanket Concept
    • Concept is at very early stage and requires modeling
    • Graphite lifetime and ability to forego fuel processing increase with core diameter since power density drops
    • Question is what diameter would give a full 30 years or more for the graphite and be able to run without fuel processing
    • Too large and/or too high a fissile concentration may mean unacceptably large starting fissile load
  • 62. DMSR Converter Reactors
    • Starting Premise is Oak Ridge`s 30 Year Once Through Design (1980)*
      • 1000 MWe output
      • Startup with LEU (20% 235 U) + Th
      • No salt processing, just add small amounts of LEU annually
      • Low power density core gives 30 year lifetime for graphite (8m x 8m)
      • Similar startup fissile load as LWR (3450 kg/GWe)
      • Averages a Conversion Ratio above 0.8
    *ORNL TM 7207 and dozens of other relevant documents available at http://thoriumenergy.blogspot.com/
  • 63. Denatured Molten Salt Reactors
    • 1810 tonne Lifetime U Ore Requirement
      • 6400 tonnes for LWR
      • 4900 tonnes for CANDU
    • At end of 30 years
      • Uranium easily removed and reused which would drop the lifetime ore to under 1000 t
      • Transuranics should also be recycled
      • ~1000 kg TRUs in salt at shutdown
      • Assuming typical 0.1% processing loss, just 1 kg in 30 years! Same low long term radiotoxicity as pure Th- 233 U cycle
  • 64. Suggested Improvements
    • Graphite Pebbles as moderator
      • Removes need for flux flattening
      • Can go to smaller, higher power core
      • Pyrolytic coatings for increased safety
    • Higher power density, smaller cores
      • Ex, double and replace every 15 years
      • Much lower starting inventory and better uranium utilization
    • Elongated Cylindrical Geometry
      • Eases fabrication and dealing with axial plenums. Tricks to limit neutron leakage
  • 65. Suggested Improvements
    • Graphite Free “Tank of Salt” Core
      • Can retain softer spectrum by having very low fuel concentration and let the carrier salt act as moderator (Be, Li, F)
    • Alternate carrier salts
      • NaF-BeF 2 low cost, low melting point
        • May allow stainless steel throughout loop
      • NaF-RbF low cost, no tritium production
        • Simplification of entire primary loop
      • NaF very low cost, higher melting point
        • These alternatives will only roughly double the neutrons absorbed in the salt from 1.5% to 3%
  • 66. Based on 0.2% tails, 75% capacity factor, 30 year lifetime LWR data from “A Guidebook to Nuclear Reactors” A. Nero 1979 3.9 million$ annual enrichment costs for DMSR at 110$/SWU, Total Fuel Costs <0.001$/kwh At $5000/kg, uranium from sea water potentially feasible and unlimited resource 200 2.0 47 2140 0.74 DMSR alternate salt, triple the neutron losses 150 1.5 35 1000 0.8 DMSR single U recycle 150 0.02$/kwh 1.5 35 1820 0.8 DMSR Converter 1 2400 If start up on 235 U 1.3 Sodium Fast Breeder 530 5.3 125 4080 0.5-0.6 LWR with U-Pu Recycle 850 million 8.5 million 200 6400 0.5-0.6 LWR Annual Ore Costs 5000$/kg Annual Ore Costs 50$/kg U Annual Uranium Ore (t) Lifetime Uranium Ore (t) Conversion Ratio Reactor
  • 67. Peak Uranium?
    • Attractiveness of converter designs hinges on sustainability of uranium resources
    • As shown however, even an enormous price increase has little effect on fuel costs for a DMSR
    • DMSR also can be viewed as a first step with fuel processing added later if uranium is scarce
  • 68. Conclusions
    • Molten Salt designs have inherent features that favour overall safety, waste reduction, low cost and rapid deployment
    • They also have great flexibility to match varying priorities
      • Can attain much higher levels of proliferation resistance compared to current offerings
      • Can run on minute amounts of thorium, or modest amounts of uranium for the utmost in simplicity
  • 69. What Way Forward?
    • Corporate interest will always be difficult to attract
      • No lucrative fuel fabrication contracts
      • Min 15 year return on investment a tough sell to shareholders (no matter how big the return may be)
      • Existing nuclear players have their choices in place
  • 70. What Way Forward?
    • Other Corporate Players?
      • Big Oil
        • For a small fraction of current profits, can retain their position in the energy market after “Peak Oil”
      • Chemical Giants
        • A majority of the needed R&D and engineering work would fit their skill set
    • Individuals with Deep Pockets?
      • What better way for those such as Gates, Branson, Allen, Buffet to invest in the future
  • 71. What Way Forward?
    • International Cooperation is key way to spread the costs and rewards
    • ITER model as rough guide but with greater corporate involvement
    • Likely no one design will be best for all nations or utilities so best to move forward on several versions
      • 95% of R&D needed would serve entire community
      • Nothing like competition to yield the best results
  • 72. What is Needed Short Term
    • Neutronic modeling
    • Fuel Salt chemistry and corrosion studies of various carrier salts and materials for heat exchangers or potential 2 Fluid barriers
    • Non-nuclear component testing of pumps, valves, heat exchangers etc.
    • Minor levels of funding to support these efforts (the hardest part of all!)
  • 73. EXTRA SLIDES…
  • 74. Two Region Homogeneous Reactor Projected breeding ratios assume thicker blanket and alternate barrier. From ORNL 2751, 1958 1.112 1.066 1.054 1.105 Projected B.R. (carbon wall) 1.091 1.004 0.977 1.055 Projected B.R. (thinner wall) 1.078 0.929 0.856 0.972 Breeding ratio (Clean Core) 2.20 2.175 2.185 2.193 Neutron Yield 0.009 0.031 0.031 0.048 Leakage 0.025 0.109 0.140 0.090 Core Vessel 0.087 0.106 0.129 0.087 Salt Losses 0.603% 0.233% 0.158% 0.592% 233 U in fuel salt mole % 7 0.25 0 0 ThF 4 in fuel salt mole % 8 feet 4 feet 4 feet 3 feet Core Diameter
  • 75. Graduate (or 4 th year project) topics all would generate “publishable” results
    • Duplicate/refine/check early ORNL documents/conclusions (e.g., ORNL 2751) with a modern neutronic code
    • Determine (calculate) neutron energy distribution/total flux at the barrier of a hypothetical 0.4 GW e , tube-in-shell, 2-salt reactor – results relevant to its BR, tube durability/lifetime, and proliferation resistance ( 232 U generation)
    • Determine (calculate) that reactor’s long-term BR and fissile requirements as a function of salt clean-up frequency ( i.e., FP/ 234 U build-up in fuel salt - do not assume a salt-cleanup technological “breakthrough”
    • Ditto assuming that Th is added to its fuel salt (rendering it a 1 ½ salt system) - vary the concentration of that Th over a wide range
    • Ditto assuming that the 2-salt system’s uranium (both salt streams) must be “denatured” ( 233 U/U total ≤ 12%) with 238 U
    • Calculate the consequences of denaturation (how it affects radwaste generation, the reactor’s BR, and fissile requirement)
    • Design a barrier material-relevant test consistent with ATR’s* capabilities/limitations
    • Design a two-salt breeder with a simple-to-replace reactor tube
    *ATR is the Advanced Test Reactor of Idaho National Laboratories. This slide is from Dr. Darryl Siemer (retired) who is attracting student interest at Idaho State University

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