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Overview of the ARIES Fusion Power Plant Studies Farrokh Najmabadi IAEA Technical Committee Meeting on Fusion Power Plant Studies March 24-28, 1998 – PowerPoint PPT presentation

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Title: Overview%20of%20the%20ARIES%20Fusion%20Power%20Plant%20Studies

Overview of the ARIESFusion Power Plant Studies
  • Farrokh Najmabadi
  • IAEA Technical Committee Meeting
  • on Fusion Power Plant Studies
  • March 24-28, 1998
  • Culham, United Kingdom

The ARIES Team Has Examined Several Tokamak and
non-Tokamak Power Plants in the Past 10 Years
  • TITAN reversed-field pinch (1988)
  • ARIES-I first-stability tokamak (1990)
  • ARIES-III D-3He-fueled tokamak (1991)
  • ARIES-II and -IV second-stability tokamaks (1992)
  • Pulsar pulsed-plasma tokamak (1993)
  • SPPS stellarator (1994)
  • Starlite study (1995)
  • ARIES-RS reversed-shear tokamak (1996)
  • ARIES-ST spherical tokamak (in progress)

Top-Level Requirements for Commercial Fusion
Power Plants
  • No public evacuation plan is required total
    dose lt 1 rem at site boundary
  • Generated waste can be returned to environment or
    recycled in less than a few hundred years (not
    geological time-scale)
  • No disturbance of publics day-to-day activities
  • No exposure of workers to a higher risk than
    other power plants
  • Closed tritium fuel cycle on site
  • Ability to operate at partial load conditions
    (50 of full power)
  • Ability to maintain power core
  • Ability to operate reliably with less than 0.1
    major unscheduled shut-down per year.
  • Above requirements must be achieved consistent
    with a competitive life-cycle cost of electricity

Tokamak Research Has Been Influenced by the
Advanced Design Program
Current focus of tokamak research
Conventional high-b tokamaks (Pulsed operation)
2nd Stability high-b tokamaks (Too much
Advanced tokamak (Balanced bootstrap)
PU Pulsed Operation SS 2nd Stability FS 1st
Stability, steady-state RS Reversed-shear
Engineering Design Options were Assessed Based
on High-performance Structural Material
Design Class Features Distinct Issues Ferritic
Steel Limited efficiency Ceramic or LM
breeder Large data base Limited heat flux He,
LM, or H2O coolant Large following Ferromagnetism
Vanadium-alloy Database/industry Li or
ceramic breeder High performance High unit
cost He or Li coolant Low afterheat Compatibilit
y Waste disposal Coatings SiC
composites Material form Ceramic breeder Very
high performance properties He
coolant Excellent safety High
fabrication Waste cost
ARIES-RS Design Emphasized those Features that
Maximize Attractiveness
  • Quantitative goals defined in the form of
    top-level requirements
  • Evolved from frequent interaction with customer.
  • Reversed shear mode of plasma operation
  • High b, high bN, high IBS, transport suppression.
  • High-performance self-cooled lithium with
    vanadium in high-temperature zones
  • Based on ARIES-II with numerous cost-saving
  • Availability a major engineering trust
  • Design for full-sector maintenance with detailed

Major Parameters of ARIES-RS Design
  • Aspect ratio 4.0
  • Major toroidal radius (m) 5.5
  • Plasma minor radius (m) 1.4
  • Plasma elongation 1.7
  • Plasma triangularity 0.5
  • Toroidal b 5
  • Electron density (1020 m-3) 2.1
  • ITER-89P scaling multiplier 2.3
  • Plasma current 11

Major Parameters of ARIES-RS Design
  • Current-drive power to plasma (MW) 81
  • On-axis toroidal field (T) 8
  • Peak field at TF coil (T) 16
  • TF-coil ohmic losses (MW) --
  • Peak/Avg. neutron wall load (MW/m2) 5.4 / 4
  • Fusion power (MW) 2,170
  • Gross electric power (MW) 1,200
  • Recirculating power fraction 0.17
  • Cost of electricity (mill/kWeh 76

ARIES-RS is a conceptual 1000MWe power plant
based on a Reversed-Shear tokamak plasma
Key Performance Parameters of ARIES-RS
Critical Physics Issues for Advanced Tokamaks
  • Wall-stabilization of kink modes.
  • Current drive near plasma edge and especially at
  • Current drive power is sensitive to constraint
    imposed by the divertor
  • High separatrix density
  • Impurity injection to radiate the plasma energy
  • Achieving the necessary density and temperature
  • Start-up, access, and

The ARIES-ST Study - Goals Schedule
  • The ARIES-ST study is a two-year project to
    investigate the potential of spherical tokamaks
    as commercial power plants as well as vehicles
    for fusion development.
  • The ARIES-ST study began in Jan. 1997. The
    effort has been focused on ST power plants. We
    have emphasized understanding the trade-off and
    identifying issues that have to be resolved.
  • The ARIES-ST study will be completed in Fall
    1998. The research reported in this meeting
    represents a progress report as in many areas
    design work is not completed.

The ARIES-ST Study - Background
  • Theoretical and experimental studies indicate
    that the MHD performance of a tokamak plasma is
    substantially improved with decreasing aspect
  • Tokamak power plants with superconducting TF
    coils, however, tend to optimize at A 4 as the
    gain in b at lower A is offset by gains at higher
    A through
  • Higher on-axis field for a fixed maximum field at
    the coil
  • Lower current-drive power (because of lower
    plasma current)
  • Engineering advantages of additional available
  • Question What is the optimum regime of
    operation for tokamaks with resistive coils
  • for power plants (Joule losses in TF is
  • for fusion development or non-electric
    applications (Joule losses in TF may not be as

Key Physics Issues for Spherical Tokamaks
  • Because of low aspect ratio, the area in the
    inboard is limited. A resistive TF coil is
    probably the only option because of the lack of
    space for a shield for a cryogenic
    superconducting coil.
  • In order to minimize the Joule losses in the TF
    coils (mainly the center-post), MHD equilibria
    with very high b are required.
  • Because there is no room for a central solenoid,
    steady-state operation is mandatory. Because of
    large plasma current, only MHD equilibria with
    almost perfect bootstrap alignment would lead to
    a reasonable current-drive power.

Key Physics Issues for Spherical Tokamaks
  • Because of unique magnetic topology, on-axis
    current drive with RF techniques is difficult.
    Current drive for profile control as well as
    start-up are additional challenges.
  • The divertor problem is more difficult than
    conventional and advanced tokamaks (higher P/R).
  • Extrapolation of present confinement data base
    (scaling) to fusion regime is questionable.

Key Engineering Issues for Spherical Tokamaks
  • The small area available for the inboard legs of
    the TF coils (center-post) make the design of
    center-post challenging.
  • Potential advantages of spherical tokamaks
    (compact and high wall load) make the engineering
    of fusion core difficult
  • Because of large recirculating power, a highly
    efficient blanket design is essential
  • Water-cooled copper coils further narrow the
  • High heat flux on in-vessel components further
    narrows the options
  • Highly shaped components (tall and thin) make
    mechanical design difficult.
  • Maintenance of the power core should include
    provisions for rapid replacement of center-post.

Spherical Tokamaks Are Quite Sensitive to
Physics/Engineering Trade-off
  • The physics and engineering trade-off are most
    evident in determining the inboard radial built
  • Smaller radial built -gt improved plasma
  • Larger radial built -gt engineering
  • Every centimeter counts!
  • Challenge maximize physics performance while
    maintaining a credible design.

Major Parameters of ARIES-ST Strawman
  • Aspect ratio 1.6
  • Major toroidal radius (m) 3.3
  • Plasma minor radius (m) 2.1
  • Plasma elongation 3.2
  • Plasma triangularity 0.57
  • Toroidal b 35
  • Electron density (1020 m-3) 3.0
  • ITER-89P scaling multiplier 2.7
  • Plasma current 32

Major Parameters of ARIES-ST Strawman
  • Current-drive power to plasma (MW) 57
  • On-axis toroidal field (T) 2.8
  • Peak field at TF coil (T) 11.5
  • TF-coil ohmic losses (MW) 871
  • Peak/Avg. neutron wall load (MW/m2) 8.2 / 5.4
  • Fusion power (MW) 4245
  • Gross electric power (MW) 2204
  • Recirculating power fraction 0.55
  • Cost of electricity (mill/kWeh) 111

TF Coil System Is Designed for Vertical Assembly
  • Water-cooled center-post is made of DS GlidCop
  • Outboard TF coil form a shell to minimize
    mechanical forces.
  • Center-post is connected to the TF shell through
    a tapered joint on the top and sliding joints at
    the bottom.
  • Insulating joint is located at the outboard
    mid-plane where the forces are smallest.
  • Another TF joint is provided for vertical
    maintenance of the power core.

Vertical Maintenance from the Bottom Is Preferred
  • Reduced building height size.
  • Radioactive material are confined to the
    maintenance area.
  • More accurate positioning with lifts compared to

The Fusion Core Is Replaced as a Unit
MHD Equilibrium and Stability
  • The MHD stability of ST discharges for a wide
    range of aspect ratio, elongation, triangularity,
    and kink wall location was examined (with 99
    bootstrap fraction).
  • There is a high leverage to operate at high
    elongations (and high d) in order to achieve a
    high b. It appears that operation at k 3 is
    possible. Detailed work in quantifying the
    feed-back power necessary for vertical
    stabilization of high-elongation ST plasmas is
  • Low-A free-boundary equilibria is unique and
    difficult to calculate
  • Strong B variation
  • Strong plasma shaping (k 3 , d 0.5)
  • High bp (2) and low li (0.15).

Free-Boundary Equilibria with Different
Current Drive
  • High-frequency fast wave (HFFW) can drive the
    current in the mid-plasma efficiently.
  • It appears that LFFW is the only plausible RF
    technique that drives current near the axis on
    high-b ST plasmas
  • Because wpe/ wce gtgt1, EC and LH waves cannot
    access the plasma center.
  • HFFW does not penetrate to the center because of
    strong electron and/or ion damping
  • ICRF fast wave suffer strong electron and a/ion
  • LFFW requires a large antenna structure for a
    well-defined spectrum (l 14 m). It generally
    has a fairly low current-drive efficiency.
  • We are also exploring NBI as an alternative

  • The divertor problem is more difficult than
    conventional and advanced tokamaks (higher PTR./R
  • To reduce the heat flux to a manageable level, a
    large fraction of the plasma power has to be
  • Radiative mantle
  • Impurity radiation in the divertor channel.
  • Impact of finite edge density and impurities on
    the MHD/current drive is under investigation.
  • This approach mainly transfers the divertor
    problem to the first wall!

Center-post Design -- Inboard Shield
  • A 20- to 30-cm thick inboard shield is required
  • To allow center-post to meet low-level waste
    disposal requirement
  • To reduce nuclear damage to the conductor
  • To limit Joule losses due to neutron-induced
  • To reduce nuclear heating in the center-post
  • To improve power balance by recovering high grade
    heat from shield
  • To prolong the center-post life time to up to 3
    years (same as first wall) in order to minimize
    impact on availability and replacement cost.

Transmutation of Cu Changes the Center-post
  • Dominant Cu transmutation products are Ni, Zn,
    and Co
  • 64Ni and 62Ni dominate the change in resistivity

Resistivity changes with a 30-cm, 80 dense
Ferritic Steel/He shield
Electrical Design of the Center-post
  • Leading conductor material is Glidcop AL-15.
  • It has adequate strength, ductility, low
    swelling, and thermal and electrical
  • Under irradiation, it suffers from severe
    embrittlement (at room temperature
  • Hardening and embrittlement are alleviated by
    operating above 180C but then it suffers from
    severe loss of fracture toughness.
  • Single-turn TF coils are preferred in order to
    reduce Joule heating
  • Higher packing fraction
  • Reduced shielding requirement (no insulation)
  • Requires high-current low-voltage supplies with
    massive busbars.

Mechanical Design of the Center-post
  • Sliding electrical joints are employed between
    center-post and other TF legs and bus-bars and TF
  • They allow relative motion in radial and vertical
    directions (which minimizes axial loads on the
  • They enhance maintainability
  • Several design options have been developed and
    tested successfully.
  • Center-post is physically separate from other
    components in order to avoid a complex interface.
  • We are currently assessing the degree to which
    the center-post can be flared to reduce Joule

Wedged Center-post Option for ARIES-ST
ARIES-ST Center-post Uses Sliding Joints
Thermal-hydraulic Design of the Center-post
  • Cry-cooling does not offer major improvement over
    cooling options at room temperature and above.
  • Water cooling is the leading option.
  • Low-temperature operation (Tinlet 35C)
    minimizes Joule losses but results in sever
    embrittlement of conductor
  • High-temperature (Tinlet 150 to 180C) avoids
    embrittlement but lose of fracture toughness and
    increased Joule losses are key issue.).
  • Liquid lithium (both conductor and coolant) is
    probably the best option for high-temperature
    operation. However, in addition to many
    challenging engineering issues, recovery of
    center-post heating does not offset increased
    Joule losses.

First Wall and Blanket Options
  • Design which include solid breeders require a
    major improvement in the thermal conductivity of
    solid breeders to handle high wall loads..
  • Self-cooled Li/V option can handle the high wall
  • The reference blanket design uses ferritic steels
    as structural material with helium as coolant and
    LiPb as the liquid breeder. SiC composite
    fillers are used to achieve a high-coolant outlet
    temperature and a reasonable power-conversion

High-Performance Ferritic Steels Blanket
  • Typically, the coolant outlet temperature is
    limited to the max. operating temperature of
    structural material (550oC for ferritic steels)
  • By using a coolant/breeder (LiPb), cooling the
    structure by He gas, and SiC insulators, a
    coolant outlet temperature of 700oC is achieved
    for ARIES-ST increasing the thermal conversion
    efficiency substantially.

Summary -- ARIES-ST
  • Several key challenging issues confront spherical
    tokamaks as fusion power plant.
  • We have proposed some potential solutions.
  • Some of these constraints are less sever in a
    non-electricity producing device.
  • It appears that spherical tokamak power plants do
    not offer major improvements over advanced
    high-aspect ratio tokamaks.
  • In the remainder of this year, we will complete
    our reference ARIES-ST design and examine
    potential of spherical tokamaks as vehicles for
    fusion development.
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