Fusion Nuclear Technology Research and Opportunities for ITER Utilization

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Fusion Nuclear Technology Research and
Opportunities for ITER Utilization

• Neil B. MORLEY and Mohamed ABDOU
• University of California, Los Angeles
• Fusion Power Associates
• Annual Meeting and Symposium
• Washington D.C.
• October 11 and 12, 2005

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Fusion Nuclear Technology (FNT)
Fusion Power Fuel Cycle Technology
FNT Components and Materials from the edge of the
Plasma to TF Coils (Reactor Core)
1. Blanket Components (FW)
2. Plasma Interactive and High Heat Flux
Components
a. divertor, limiter
b. rf antennas, launchers, wave guides, etc.
3. Vacuum Vessel Shield Components
Other Components affected by the Nuclear
Environment
4. Tritium Processing Systems
5. Instrumentation and Control Systems
6. Remote Maintenance Components
7. Heat Transport and Power Conversion
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Fusion Nuclear Technology Critical Issues for
Fusion Energy

1. Tritium Supply Tritium Self-Sufficiency
2. High Power Density
3. High Temperature
4. MHD for Liquid Breeders / Coolants
5. Tritium Control (Permeation)
6. Reliability / Availability / Maintainability
7. Testing in Fusion Facilities

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DT fusion is usually depicted to laymen by the
reaction in the plasma
Inexhaustible Limitless
Physics
Confinement Current Drive Heating
n
Tritium Consumption in Fusion is HUGE!
Unprecedented! 55.8 kg per 1000 MW fusion power
per year
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Tritium supply for the development of fusion
where does it come from?

• Production Cost
• CANDU Reactors 27 kg from over 40 years, 30M/kg
  (current)
• Fission reactors 23 kg / year. at a cost of
  84M-130M per kg, per DOE Inspector General

• Conclusions
• Availability of tritium supply for fusion
  development beyond ITER first phase is an issue
• Large power D-T facilities must breed their own
  tritium (this is why ITERs extended phase was
  planned to include the installation of a tritium
  breeding blanket)
• FW/Blanket are necessary in the near term to
  allow continued development of D-T fusion

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The DT FUSION ENERGY picture requires a closed
fuel cycle and nuclear technology

• Blanket / Shield Components and Materials
• Absorption
• Activation
• Multiplication
• Energy Extraction
• Shielding
• R/A/M

• Tritium Fuel Cycle
• Processing
• Decay
• Permeation
• Inventory

Physics
n
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Tritium Self-Sufficiency ?a gt ?r

• ?r Required tritium breeding ratio
• ?r is 1 G, where G is the margin required to
  account for
• tritium losses, radioactive decay
• inventory in plant components
• inventory in tritium processing system
• inventory stockpile for outages and for start-up
  of other plants
• ?r is dependent on many system physics and
  technology parameters.
• ?a Achievable tritium breeding ratio
• ?a is a function of technology, material and
  physics requirements, e.g.
• Efficient energy extraction
• FW armor and thickness
• Conducting shells, embedded coils, heating ports,
  etc.
• Reliability/maintainability concerns

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Current physics and technology concepts lead to a
narrow window for attaining tritium
self-sufficiency for DT fusion energy

• Tritium inventory in processing systems and
  reserves are closely tied to fueling rate and
  fractional burn-up in plasma strong influence
  on required TBR ?r
• 3D Analysis of current worldwide FW/Blanket
  concepts accounting for plasma support systems
  estimates an achievable TBR ?a 1.15
• Integral neutronics experiments in Japan and the
  EU showed that calculations consistently
  OVERESTIMATE experiments by an average factor of
  1.14

td doubling time
Required TBR
td1 yr
td5 yr
td10 yr
Fractional burn-up
Window for Tritium self sufficiency
Fusion power - 1.5GW Reserve time - 2 days Waste
removal efficiency - 0.9 (Sawan and Abdou,
ISNFT-7)
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Physics and Technology RD partnership needed to
determine the potential for achieving Tritium
Self-Sufficiency

• How do we Establish the conditions governing the
  scientific feasibility of the D-T cycle, i.e.,
  determine the phase-space of plasma, nuclear,
  material, and technological conditions in which
  tritium self-sufficiency can be attained
• RD on FW/Blanket/PFC and Tritium Processing
  Systems that emphasize
• Understanding and predicting behavior of
  components and materials in the integrated fusion
  environment under relevant conditions
• Minimizing Tritium inventory in components
• Faster tritium processing system, particularly
  processing of the plasma exhaust
• Improve reliability of tritium-producing
  (blanket) and tritium processing systems
• RD on physics concepts and operating modes that
• Maximize tritium fractional burn-up
• Reduce the requirements on space needed in the
  breeding region for heating, stabilization coils
  and conductors, etc.
• Ease peak requirements on surface heat loads and
  disruptions loads, etc.

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A technology/physics partnership is clearly
already a part of ITER
Many FNT components capabilities needed for
ITER basic machine
1. Blanket Components 2. Plasma Interactive and
High Heat Flux Components a. divertor,
limiter b. rf antennas, launchers, wave guides,
etc. 3. Vacuum Vessel Shield 4. Tritium
Processing Systems 5. Instrumentation and
Control 6. Remote Maintenance 7. Heat Transport
and Power Conversion
But FEW technology solutions for ITER are
compatible with TRITIUM SELF-SUFFICIENCY and
ENERGY NEEDS
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The ITER Test Blanket Module (TBM) Program is a
vehicle for utilizing ITER to advance the
scientific principals of tritium self-sufficiency!
The ITER should serve as a test facility for
neutronics, blanket modules, tritium production
and advanced plasma technologies. The important
objectives will be the extraction of high-grade
heat from reactor relevant blanket modules
appropriate for generation of electricity. The
ITER Quadripartite Initiative Committee (QIC),
IEA Vienna 1819 October 1987
Studying burning plasma physics
ITER
Studying breeding energy relevant technologies
n

• ITER should test design concepts of tritium
  breeding blankets relevant to a reactor. The
  tests foreseen in modules include the
  demonstration of a breeding capability that would
  lead to tritium self sufficiency in a reactor,
  the extraction of high-grade heat and electricity
  generation.
• SWG1, reaffirmed by ITER Council, IC-7 Records
  (1415 December 1994), and stated again in
  forming the Test Blanket Working Group (TBWG)

ITERs Principal Objectives Have Always Included
studying ENERGY relevant technologies and
materials
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The TBM in ITER is essential to
TBM Mission
Perform first wall and tritium breeding module
experiments to advance the understanding of the
competing requirements of tritium
self-sufficiency, extraction of high grade heat,
and controlled, ignited plasma operation.

• Achieve a key element of the ITER Mission
  demonstrate the scientific and technological
  feasibility of fusion power for peaceful
  purposes
• Achieve the most critical milestone in fusion
  nuclear technology research testing in the
  integrated fusion environment.
• Resolve the critical tritium supply issue for
  ignited plasma experiments and fusion development
  beyond ITER - and at a fraction of the cost to
  buy tritium for a large D-T burning plasma
• Access RD information from much larger (10-20M
  per year) blanket/PFC programs (EU and Japan) and
  other international partners
• Maximinize the return on the gt1B of US
  investment and capitalize on the gt10B of
  investment by international partners in ITER

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TBM Preparation and RD is proceeding
aggressively in the International Community
View of a typical TBM test port cell arrangement
TBM location in a ITER test port

• Several TBM proposals have
• been made by ITER Parties
• Helium-cooled Li-based Ceramic/Beryllium TBM (4
  variations)
• Helium-cooled liquid Lithium-Lead TBM (3
  variations)
• Water-cooled Li-based Ceramic/Beryllium TBM (1
  variation)
• Liquid natural Lithium TBM (2 variations)

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ITER plan includes the TBM Activities are
coordinated by the Test Blanket Working Group

• TBMs are to be installed from the first day
  ofH-H operation to check interfaces main
  operations, compatibility with ITER operations
  and to support to safety dossier

• 3 Midplane ports are reserved for TBM use, as
  well as space at the port cell, TCWS building,
  tritium building, and hot cell for necessary
  ancillary systems such as coolant loops, tritium
  processing, etc.

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ITER Environment for TBM Experiments

• large geometry of the test ports. (maximum height
  of TBM 2m, similar to the size of typical
  blanket modules in a power plant)
• plasma exposure with typical particle loads and
  off normal plasma events
• strong magnetic field ( 4 T), same order of
  magnitude as in power plants
• similar neutron energy spectrum as in power
  plants, however lower neutron flux (25- 30 of
  neutron wall loading in DEMO plant) and much
  lower fluence
• generation and confinement of radioactivity

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US TBM Selected Concepts
1. The Dual-Coolant Pb-17Li Liquid Breeder
Blanket concept with self-cooled Pb-Li breeding
zone and flow channel inserts (FCIs) as MHD and
thermal insulator
-- Innovative concept that provides pathway to
higher outlet temperature/higher thermal
efficiency while using ferritic steel.
-- US lead role in collaboration with other
parties (most parties are interested in Pb-Li as
a liquid breeder, especially EU and China).
-- Plan an TBM that will occupy half an ITER test
port with corresponding ancillary equipment.
2. The Helium-Cooled Solid Breeder Blanket
concept with ferritic steel structure and
beryllium neutron multiplier, but without an
independent TBM
-- Support EU and Japan efforts using their TBM
structure ancillary equipment
-- Contribute only unit cell /submodule test
articles that focus on particular technical issues
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Dual Coolant Lead-Lithium (DCLL) FW/Blanket
Concept

• Idea of Dual Coolant concept Push towards
  higher performance with present generation
  materials (FS)
• Ferritic steel first wall and structure cooled
  with helium
• Breeding zone is self-cooled Pb-17Li
• Structure and Breeding zone separated by SiCf/SiC
  composite flow channel inserts (FCIs) that

DCLL Typical Unit Cell
Self-cooled Pb-17Li Breeding Zone
SiC FCI
He-cooled steelstructure

• Provide thermal insulation to decouple Pb-17Li
  bulk flow temperature from ferritic steel wall
• Provide electrical insulation to reduce MHD
  pressure drop in the flowing liquid metal
• Pb-17Li exit temperature can be significantly
  higher than the operating temperature of the
  steel structure ? High Efficiency

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FW He Coolant Manifolds
Pb-Li Outlet Pipe
Pb-Li Inlet Pipe
Pb-Li Flow Separation Plate with He coolant
Channels
Pb-Li Inlet Manifold
Pb-Li Return Flow Channel
FCI
Plasma Facing First Wall
Pb-Li Inlet Flow Channel
FW He Coolant Channels
Bottom Plate He Coolant Channels
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Helium-Cooled Ceramic Breeder (HCCB)
Blanket/First Wall Concept for TBM

• Idea of Ceramic Breeder concepts Tritium
  produced in immobile lithium ceramic and removed
  by diffusion into purge gas flow
• First wall / structure / multiplier /breeder all
  cooled with helium
• Beryllium multiplier and lithium ceramic breeder
  in separate particle beds separated by cooling
  plates
• Temperature window of the ceramic breeder and
  beryllium for the release of tritium is a key
  issue for solid breeder blanket.

Schematic view of an example ITER HCCB test
blanket submodule showing typical configuration
layout of ceramic breeder, beryllium multiplier
and cooling structures and manifolds

• Thermomechanical behavior of breeder and
  beryllium particle beds under temperature and
  stress (and irradiation) loading affects the
  thermal contact with cooled structure and impacts
  blanket performance
• Nuclear performance and geometry is highly
  coupled and must be balanced for tritium
  production and temperature control

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Ceramic Breeder TBM Inserting US unit cells
into the EU HCPB structural box
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Specific TBM Test Objectives in ITER

1. validation of TBM structural integrity under
   combined and relevant thermal, mechanical and
   electromagnetic loads
2. validation of Tritium breeding predictions
3. validation of Tritium recovery process
   efficiency, tritium control and inventories
4. validation of thermofluid predictions for
   strongly heterogeneous breeding blanket concepts
   with volumetric heat sources and strong MHD
   interactions
5. demonstration and understanding of the integral
   performance of the blanket components and
   material systems

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Summary Remarks

• There are many remaining challenging FNT issues
  that need to be resolved for successful fusion
  development
• The D-T cycle is the basis of the current world
  plasma physics and technology program. If the D-T
  cycle is not feasible the plasma physics and
  technology research would be very different.
• Tritium self-sufficiency is a complex issue
  that depends on many system physics and
  technology parameters / conditions.
• Availability of external tritium supply for
  continued fusion development beyond ITERs first
  phase is an issue
• There is only a window of physics and
  technology parameters in which the D-T cycle is
  feasible. We need to determine this window.
• Conducting an effective Test Blanket Module (TBM)
  program is one of the main objectives of ITER and
  necessary to advance the understanding tritium
  breeding and tritium self sufficiency in fusion
  systems
• ITER will be the first real opportunity to apply
  the results of RD from the past 30 years on many
  aspects of blankets, materials, PFC, etc.