Design of a Compact Fusion Reactor

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Design of a Compact Fusion Reactor

• MANE 4390
• Susan Kane
• Bill Schlichting
• Ben Schultz

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What is fusion?

• 2H 3H ? 4He n Q 17.6 MeV
• Atoms heated to high temperatures, a plasma is
  formed
• Three requirements
• high plasma density (n)
• high temperature
• long confinement time (t)
• Characterized by Lawson criterion
• nt 1020 sm-3

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Plasma Confinement

• Electrical repulsion tend to force ions away from
  each other
• Two approaches have been used
• Inertial confinement
• National Ignition Facility
• U of R Omega Facility
• Magnetic confinement
• ITER
• Princeton NSTX

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Conventional Tokamak Fusion Device (ITER)
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Project Description

• Fusion reactors that gear toward reduced size and
  high power density are known as compact designs
• One such design is the spherical torus
• A spherical torus is achieved by keeping only the
  indispensable components on the inboard side of
  a tokamak plasma
• Project Develop a rudimentary design for a
  compact Tokamak reactor based on the spherical
  torus concept and perform cost analysis

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Compact Fusion Device
ARIES Spherical Torus Design
NSTX Design
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Project Goals

• Create Working Design Code
• Design code incorporates physics and engineering
  constraints
• Develop Solidworks Model
• Model Key Reactor Components
• Perform Cost Analysis on Potential Designs and
  Identify Most Promising Configuration

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Our Design Code Has Been Able to Replicate
Previous Results to Within a Few Percent
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Solidworks Model
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Modeling Key Components
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Modeling of TF Coils

• Determine resistive losses in the copper center
  post and the copper return legs
• Develop cooling scheme in center post and return
  legs based on water coolant and resistive losses

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Calculation of Resistive Losses

• Center Post
• It Icp
• Pcp It2Rcp
• Rcp rL/prc2 rc Center Post Radius
• Return Legs
• It Icp IRLn
• n number of return legs
• PRL It2RRL/n
• Ptotal Pcp PRL

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Modeling of Center Post Cooling with Equivalent
Channel Approach
Req
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Heat Transfer Equations for Center Post
The temperature gradient is evaluated using
numerical analysis to determine the temperature
of the conductor at any point r.
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Equivalent Channel Results With Constant
Resistivity
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Equivalent Channel Results With Temperature
Dependent Resistivity
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Cooling Channel Optimization
?T22 K Tmax457 K
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Equivalent Channel Results
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Cooling Channel Optimization Contd

• Center post height increased to accommodate all
  components
• The new Maximum Temperatures are

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Torodial Field Coil Cost

• For pure copper, the fabricated cost is assumed
  to be 100 per kilogram.
• The final design uses dispersion strengthened
  copper with the same fabrication price.
• For a center post with 325 cooling channels and 6
  return legs, the cost is 32.7 million.

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Blanket and Shield Requirements

• Blanket Recover energy from plasma at a
  thermodynamic temperature and produce tritium to
  be used in the plasma
• Shield Limit nuclear heating and radiation
  damage to the magnetic field coils

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Blanket Characteristics

• Tritium breeding material For Example Lithium
  Oxide (Li2O) or Lithium-Lead (Li-Pb)
• Considerations
• High Atomic Density
• High Thermal Conductance
• Broad Operating Temperature Range
• Compatible with Center Post Coolant (H2O)
• 7Li 1n ? 3T 1n 4He
• 6Li 1n ? 3T 4He

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Shielding Characteristics

• Shielding materials Iron and Borated Water
• Both high-Z and low-Z materials absorb gamma rays
  and slow down neutrons that dont interact with
  the blanket

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Blanket and Shield

• The design of the blanket and shield would
  require much more time than just a semester
• Therefore, our blanket and shield design will be
  taken from previous research and experimentation

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Outboard Radial View
Plasma Blanket Shield
Coils
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Blanket Breeding

• In order to ensure that the needed amount of
  tritium is available, in design studies we aim at
  a calculated TBR of about 1.1
• Based on this requirement, the needed thickness
  of the Li-Pb blanket is about 50 cm

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Blanket Breeding Graph
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Shield Thickness

• In order to ensure that too much heating and
  radiation damage do not get to the coils, the
  shield must have enough material to protect the
  coils
• Based on nuclear heating, the shield must be
  about 22 cm thick
• Based on radiation (neutron fluence), shield
  thickness must be at least 26 cm

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Shield Thickness Graph 1
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Shield Thickness Graph 2
Overall thickness will be 28 cm to allow for
safety factor
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Water Percentage in Shield

• The shield will contain iron and borated water.
• Based on research for the ARIES-AT, the amount of
  water that should be included to minimize peak
  nuclear heating and radiation damage is about 55

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Graph of Water in Shield
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The Poloidal Field Coils Will be Superconducting
Magnets

• No electrical resistance
• Must maintain very low temperatures (4-10 K)
• About 500 Watts required to remove 1 Watt of heat
  deposited in coil

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Poloidal Field Coil Functions

• Vertical Field Coils
• Plasma Equilibrium

• Shaping Field Coils
• Plasma Elongation

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Coil Currents Relative to Plasma Current (ORNL)

• Low aspect ratios require lower current ratios
• Spherical Torus A1.3 2

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PF Coil Costing Algorithm

• Obtain cost of PF coils as a function of other
  parameters
• Aspect Ratio
• Plasma Current
• PF Coil Dimensions
• Superconducting Material Properties
• Current Density
• Material Density
• Unit Price
• Preliminary Estimate
• SF Coils 3 Million
• VF Coils 30 Million
• Total 33 Million

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Economic Analysis

• Analysis based on cost model by K. Evans, Jr.
• Model is relative to STARFIRE design
• Unit costs were doubled to account for inflation.

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Design Optimization
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Optimal Design

• Aspect Ratio 1.66
• Major Radius 1.99 m
• Blanket and Shield Thickness 0.93 m
• On-axis Torodial Field 4.78 T
• On-axis Polodial Field 4.61 T
• Fusion Power 3406 MW
• Max Temp in Center Post 475 K
• Net Electric Power 1020 MW
• Cost of Electricity 5.8 /kWh

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Questions?