ITER Test Blanket Module (TBM)
Current U.S. Designs, Plans, Issues, and
Material R&D Needs
Prepared by:
M. Abdou, A. Ying, N. Morley, C. Wong
D. Sze, S. Malang, S. Smolentsev
M. Sawan, M. Dagher, P. Calderoni, B. Merrill
Presented at the MASCO meeting
Washington DC, August 23-24, 2005
1
Outline
I.
ITER Plans
•
ITER basic device components for TBM and support systems and
ITER/TBM interface.
•
What the International Partners (TBWG), including US, agreed to.
II. U.S. Current Plans (as they have evolved during the past year)
•
TBM options and strategy.
•
Description of designs of TBM test articles.
•
External loops and ancillary equipment (piping, heat exchangers,
tritium extraction, etc.).
III. TBM key requirements related to materials
IV. What is needed from the Materials Program
•
R&D for TBM.
•
R&D for higher performance TBM and Power Plants.
2
What is the ITER TBM Program?
Integrated testing of breeding blanket and first wall components and
materials in a Fusion Environment
• Breeding Blankets/FWs will be tested in ITER, starting on Day
One, by inserting Test Blanket Modules (TBMs) in specially
designed ports.
• Each TBM will have its own dedicated systems for tritium
recovery and processing, heat extraction, etc. Each TBM will also
need new diagnostics for the nuclear-electromagnetic environment.
• Each ITER Party is allocated limited space for testing two TBMs.
(Number of Ports reduced to 3. Number of Parties increased to 6).
• ITER’s construction plan includes specifications for TBMs
because of impacts on space (port, port area, hot cell, TCWS),
shielding, vacuum vessel, remote maintenance, ancillary
equipment, safety, availability, etc.
• The ITER Test Program is managed by the ITER Test Blanket
Working Group (TBWG) with participants from the ITER
3
International Team and representatives of the Parties. (However, this
entity may change under the new international agreement being negotiated.)
Blanket Testing in ITER is essential
•
Achieve a key element of the “ITER Mission”
“demonstrate the scientific and technological feasibility of fusion power for
peaceful purposes”
“test tritium breeding module concepts that would lead in a future reactor to
tritium self-sufficiency, the extraction of high grade heat, and electricity
production”
•
Achieve the most critical milestone in fusion nuclear technology
research: testing in the integrated fusion environment
– This has been the focus of the US Technology Program for a long
time. It is the first real opportunity to apply the results of R&D from
the past 30 years on blankets, materials, PFC, etc.
•
The ITER TBM project provides a driving force to bring Fusion Nuclear
Technology R&D the first step toward reality
•
Develop the technology necessary to install breeding capabilities to
supply ITER with tritium for its extended phase of operation
•
Resolve the critical “tritium supply” issue for fusion development
- and at a fraction of the cost to buy tritium for large D-T burning
plasma
4
TBM Mission
“Test tritium breeding module concepts that would lead in a
future reactor to tritium self-sufficiency, the extraction of high
grade heat, and electricity production.”
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 thermal predictions for strongly
heterogeneous breeding blanket concepts with
volumetric heat sources
5. demonstration and understanding of the integral
performance of the blanket components and material
systems
5
ITER Operation
• ITER operation starts in 2016. It has HH operation (~3yr), DD phase (~1yr), low duty
cycle DT (~3yr), high duty DT (~3yr)
• The ITER schedule shows Test Blanket Module operating in the device from Day 1
• The ITER International Partners agreed on a general strategy for each blanket concept:
4 sequential test articles corresponding to the 4 modes of ITER operation
6
• Average fluence: 0.09 MW•y/m2 after 10 years; 0.3 MW•y/m2 after 20 years
TBWG agreed to allocate the 3 test ports by
blanket concept, not by party
Port # 16:
Helium-cooled Ceramic Breeder
TBMs (all Parties)
Port # 18:
Helium-cooled Lithium Lead /Dual
Function Lithium Lead TBMs
(EU/China)
Water-Cooled Ceramic Breeder
TBM (Japan)
Port # 2:
Dual-Cooled Lithium Lead /Dual
Function Lithium Lead TBMs
(US/China)
Li-Breeder TBMs (RF, KO)
Helium-Cooled Ceramic Breeder
TBM (only if a liquid breeder
option does not make it)
Note: The interface with ITER device & facilities has been fixed (7/2005)
7
ITER test port configuration has been fixed (by ITER & TBWG)
(U.S. DCLL TBM is shown for illustration of how a TBM fits into this configuration)
VV – Cryostat Duct
Cryostat
VV Port Extension
Bio-Shield
TBM Frame Assy
Bio-Shield Port
Opening
Plasma
Transporter
Pb-Li Primary Coolant
Loop Ancillary system
TBM
VV Closure Plate
Pb-Li Concentric Pipe
Port Cell Area
8
Configuration (and size) of TBM within ITER Test Port
Port plug = Frame + TBMs
Frame = FW structure + Box structure + Backside shields
Vertical Port Frame
Opening space for TBM
20
Test Blanket
Module
Backside Shield
Cut-view
Vertical cross-section of Frame
(at the center of flexible supports)
20 mm gap all around
inside frame openings
20
1660
TBM Frame
Assembly
• Port can be divided vertically or
horizontally
• The maximum size of a test module
(half vertical port) is 1.66m x 0.484m
(TBM maximum first wall area is 0.8m2)
484
Unit: mm
9
II. Current U.S. TBM Plans
US TBM plans evolved during the past year through
technical studies, interactions with the community, VLT and
DOE, as well as interactions with the international ITER
partners and our work within TBWG.
• TBM selected concepts and strategy.
• Description of designs of TBM test articles.
• External loops and ancillary equipment
(piping, heat exchangers, tritium extraction, etc.)
10
US TBM Selected Concepts
1. The Dual-Coolant Pb-17Li Liquid Breeder Blanket concept with selfcooled 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 independent 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
11
Dual Coolant Lead-Lithium (DCLL)
FW/Blanket Concept
Idea of “Dual Coolant” concept –
Push towards higher
performance with present
generation materials (FS)
DCLL Typical Unit Cell
 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
 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
12
US DCLL TBM module
All structures are He-cooled @ 8MPa
self-cooled PbLi flows in poloidal direction
SS frame
Front
FCI is the Thermal and MHD
Insulator lining all PbLi channels
FS structure
PbLi in
Back
He in
He out
PbLi in
Be front face
PbLi out
PbLi out
FW He counter flow
13
Pb-Li Outlet Pipe
FW He Coolant
Manifolds
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
14
DCLL TBM Calculated Temperatures
• Nominal (reference) operation
Radial T distribution at the exit
high performance operation
nominal operation
(350C<He<450C;370C<PbLi<470C)
700
FS/PbLi=462C
SiC/PbLi=470C
Calculations performed at
k=15 W/m-K (not so important), =20
S/m
600
T, C
• High performance (example)
650
SiC/PbLi
PbLi/SiC
550
FS/PbLi
(350C<He<450C;450C<PbLi<620C)
FS/PbLi=482C
SiC/PbLi=607C
Calculations performed at
k=3 W/m-K (reduced to minimize heat
losses into He, =20 S/m
500
He/FS
450
400
-0.08
-0.04
0
y, m
0.04
0.08
15
DCLL Design Temperatures
For both TBM and power reactor design, the following temperature limits were
designed to and can be achieved:
TBM Reference operation
FS Tmax
≤ 550° C
FS/PbLi
< 500° C
SiC/PbLi < 500° C
SiC Tmax
< 500° C
350° C < He < 450° C
370° C < PbLi ≤ 470° C
Higher performance operation
≤ 550° C
< 500° C
< 700° C
< 700° C
350° C < He < 450° C
450° C < PbLi ≤ 650° C
For the DCLL TBM higher PbLi exit temperature ~650° C can be achieved via the
bypass loop without requiring high-temperature materials for external
piping/HX/TX.
High-temperature materials (compatible with Pb-17Li up to 700° C) will be needed
for the external lead lithium loop (HX/TX tubes) ONLY for the high performance
power plant (NOT for TBM, and NOT for moderate performance power plants)
16
DCLL TBM Bypass Loop Schematic
Pump
400 C
30.3 kg/s
Valve
off
DCLL
TBM
Concentric
pipe with
FCI
Tritium extraction tank
bypass
line
0 kg/s
470 C
PbLi mixing
tank
PbLi
loop
470 C
400 C
PbLi/He
Heat
Exchanger
180 C
8 MPa
Helium
loop
30.3 kg/s
0.4 MW
300 C
Higher PbLi exit temperature can be achieved without requiring hightemperature materials for external piping/HX/TX. This can be achieved by
turning the bypass valve “on” to allow mixing a lower temperature stream with
17
the high-temperature stream in the PbLi mixing tank
18
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.
Side Wall
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 19
for
tritium production and temperature control
Ceramic Breeder TBM
Inserting “US” unit cells into the EU HCPB structural box
Electromagnetics/Neutronics
unit cell design
Unit (mm)
Typical Operating Temperature
Helium Coolant
Ferritic
Ceramic breeder
(for tritium release)
Beryllium
In/Out
Max
Min
Max
Max
300/500 C
550 C
350-400 C
900-1000 C
600 C
20
Example: DCLL TBM Testing Schedule in ITER
(4 sequential test articles)
ITER Year
ITER
Operation Phase
Progressive
ITER
Testing
Conditions
Electromagnetic/
Structural
(EM/S) TBM
Nuclear Field/
Tritium Prod.
(NF/TP) TBM
Thermofluid/
MHD (T/M) TBM
Integrated (I)
TBM
-1
1
2
3
4
5
6
7
8
9
10
Magnet
testing &
vacuum
HHFirst
Plasma
HH
HH
DD
Low
Duty
DT
Low
Duty
DT
Low
Duty
DT
High Duty
DT
High
Duty
DT
High
Duty
DT
Toroidal
B field
Vacuum
• Install
• RH
• System
checkout
Heat
flux
NWL
Small
neutron
flux
B Field
Disruptions
Fluence
Accumula
-tion
Full
disruption
energy
• Transient EM Loading on
structure and FCIs
• FW heat flux loading
• ITER field perturbation
• LM-MHD tests
•
•
•
•

Finalize
Design
Nuclear field
Tritium production
Nuclear heating
Structure and FW
heating
• Thermal and
electrical
insulation
• Tritium
permeation
• Velocity profiles

Finalize
Design

Finalize
Design
• High temperature effects
• Tritium permeation/recovery
• Integrated function, reliability
21
DCLL TBM Pre-ITER Schedule to deliver test module
one year before ITER Day One of operation
ITER HCLL TBM Schedule
ITER Director appointed
First plasma
2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015
TBWG activities and ITER parties interface
R&D
Basic thermofluid MHD
SiC/SiC FCI Fab and Compatibility
EM/S TBM Specific
Module design specific R&D
Subsequent TBM Specific
Design and analysis
Conceptual design
Preliminary design
Final design
design review
final design
Development of TBM TSD (Technical specification data)
Mock-ups and Qualification tests
Facility definition and preparation
Sub-components verification test
1/4 to 1/2 scale mock-ups
TBM design and fabrication
Call for tender / Contract award
Manufacturing design (tooling and processing)
Material procurement
Fabrication and procurement
Delivery to ITER site installation and tests
Update: 08/19/05
22
HCCB TBM Pre-ITER Schedule to deliver test
submodule one year before ITER Day One of operation
ITER HCCB TBM Schedule
ITER Director appointed
First plasma
2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015
TBWG activities and ITER parties interface
R&D
Solid breeder thermonechanics and T recovery
T control and predictive capability
Virtual TBM (integrated modeling)
In-pile tritium release test
Diagnostics and instrumentation
Design and analysis
Conceptual design
Preliminary design
Final design
design review
final design
Development of TBM TSD (Technical specification data)
Mock-ups and Qualification tests
Facility definition and preparation
Sub-components verification test
1/4 to 1/2 scale mock-ups
TBM design and fabrication
Call for tender / Contract award
Manufacturing design (tooling and processing)
Material procurement
Fabrication and procurement
Delivery to ITER site installation and tests
Update: 08/17/05
23
III. TBM Key Requirements
Related to Materials
24
TBM Materials and Key Parameters
• Materials:
-- DCLL: Ferritic Steel, PbLi, He, FCI (SiC or Sandwich type)
-- HCCB: Ferritic Steel, Be, Ceramic Breeder (Li2TiO3 or Li4SiO4), He coolant,
He purge
• Maximum fluence < 0.09 MW•y/m2 (< 1 dpa)
• Neutron wall load: ~ 0.8 MW/m2
• Surface heat flux: maximum 0.5 MW/m2 (local)
design average 0.3 MW/m2
• maximum number of pulses: < 30,000 cycles
• Ferritic Steel: maximum temperature < 550°C
• FS/PbLi interface < 500°C
• SiC/PbLi < 600°C (~500°C also o.k.)
25
Relevant MHD issues for US DCLL that impact
material requirements and performance
 MHD Pressure drop is a serious concern for inboard LM
blankets in high field, high power density reactors. Even
moderate, but non-uniform, MHD pressure drops (arising from
flaws for example) can seriously affect flow balance between
parallel channels leading to hot channels
 MHD velocities profiles can exhibit strong jets next to regions
of stagnation and even reversed flow
 Non-uniform volumetric heating can cause natural convection
flows that MHD effects do not damp – can swamp forced flow
velocity in slow moving breeder zone regions
 Turbulence/stability modification and suppression by MHD
forces and joule dissipation will likely affect performance
Ha and Re numbers ~1-3 x 104
All of these MHD issues strongly influence heat transfer, corrosion, tritium
permeation, material requirements, and ultimate design.
26
Optimization of FCI Properties
 Properties of the FCI need to be optimized in order to address
competing requirements:





reduce MHD pressure drop to acceptable levels
minimize heat losses from PbLi to helium coolant
minimize temperature drop across the FCI (thermal stress)
limit RAFS/PbLi interface temperature
minimize corrosion rate ▪ minimize tritium diffusion
 Optimal mix of thermophysical & mechanical properties depends
strongly on the thermofluid MHD and will be determined by R&D and
design tradeoffs
 Properties given on another slide give some approximate goals and
requirements for SiC development for the FCI
=500 1/Ohm-m
100
20
5
Effect of on MHD
jet formation
Velocity normalized
to vave = 0.1 m/s
B=4T
27
Effectiveness of SiCf/SiC FCI as Electric/Thermal Insulator
depends on the Thermophysical properties
And Design of FCI (e.g PES VS. PEH and location relative to
magnetic field)
X (poloidal)
FCI
Pb-17Li bulk flow
He flows
Fe wall
Pressure equalization opening makes the
pressure on both sides of the FCI equal,
resulting in almost no primary stresses. The
opening is either a slot (PES) or a row of
holes (PEH).
500
Pb-17Li gap flow
PES
Analysis shows significant MHD pressure drop
reduction with FCI as compared to the case
without insulation:
a factor of 10 at  = 500 S/m (acceptable)
a factor of 200-400 at  = 5 S/m (desired)
(dP/dx)0 / (dP/dx)
B-field
PEH
PES
400
300
200
100
0
1
10
100
1000
Electrical conductivity, S/m
28
Desired Flow Channel Insert Properties
 Low transverse electrical and thermal conductivity of the SiC/SiC
 Goal ~1-10 S/m transverse electrical conductivity, Acceptable <500 S/m for TBM
operation
 Goal ~2-5 W/mK transverse thermal conductivity, Acceptable <15 W/mK for TBM
operation
 The inserts need to be compatible with flowing Pb-17Li
 Goal ~800°C, Acceptable 500°C for initial TBM operation
 Liquid metal must not “soak” into pores of the composite (or foam) in
order to avoid increased electrical conductivity. In general, closed porosity
and/or dense SiC layers are required on all surfaces of the inserts.
 Secondary stresses and deformation caused by temperature gradients
must not endanger the integrity of FCIs.
 Goal 200°C temperature difference, Acceptable 100°C temperature difference for
initial TBM operation
 The insert shapes needed for various flow elements must be fabricable
 Basic box-channel element
 Cross sections up to 150 x 150 mm for TBM
 Lengths 500-1000 mm for TBM
29
IV. What is needed from the
Materials Program
DCLL
A) R&D for TBM
B) R&D for Power Plants
B1) Moderate Performance
B2) Higher Performance
Solid Breeders
-- R&D for TBM
30
Material Application for FCI (and associated external
piping, HX, and TX) for TBM and Power Plants
TBM
Power Plants
Reference
Higher
Performance*
Moderate
Performance
Higher
Performance
FCI Sandwich
Yes
N/A
Yes
N/A
FCI Mod. Temp SiC (500°C) (X)
Yes
N/A
Yes
N/A
FCI High Temp SiC
No
Yes
No
Yes
Advanced Materials for HX and
TX tubes
No
No
(use bypass /
mixing tank)
No
Yes
(X) Requirements on σ and к for the moderate temp SiC FCI are less demanding
than those for high-temperature, high performance SiC FCI
* If development of high temperature SiC for power plants is available around 201531
2020, a higher performance TBM can be considered for later stages of ITER testing.
A. Needs for the DCLL TBM
a)
Development of flow channel inserts for thermal and
electrical insulation
•
•
•
b)
inserts made of SiC-composite suitable for moderate temperature
(< 500ºC)
inserts made of sandwich-structure (i.e. steel-alumina-steel,
welded at all edges), back-up solution in case the SiC inserts can
not be developed in time, suitable for lead lithium exit
temperatures up to 500ºC (defer this until after the first MHD tests
around 2008)
for higher performance TBM: SiC FCI for high temperature (up to
650°C) lead lithium at the exit
Fabrication of the TBM structure from RAFS (F82H or
EUROFER) including forming of the complicated structure
(i.e. FW panel), welding (diffusion, TiG, Laser, electron
beam), PWHT.
•
Most of these steps are under development in the EU and Japan,
and the alternative is therefore to either procure from their
industry, or to develop the required technology in the US
(Iteration with the detailed TBM design will be needed)
32
A. Needs for the DCLL TBM (cont’d)
c)
Code-qualification of the RAFS and the welds for ITER
applications
d)
Development of suitable FE-codes for the analysis of the
steel/Pb-17 Li compatibility tests and the extrapolation of
these results to the conditions in a blanket ( flow conditions,
impact of magnetic field, suitable criteria for determining
temperature limits)
e)
Data base for the compatibility of Pb-17 Li with RAFS and
SiC-composites (e.g. forced convection loop with
temperature gradients in flow direction, measurement of
depositions, impact of irradiation)
f)
Recommendations regarding suitable materials for the
external lead-lithium loop including limits on the allowable
interface temperatures (maximum lead lithium temperature
470ºC).
33
B. DCLL Material R&D for Power Plants
Two kinds of operation of the DCLL blankets in power plants can
be considered:
B1) Moderate Performance with lead lithium exit temperatures up to
500 C and a RANKINE cycle (steam turbine) power conversion
system
• B1 can be realized with sandwich-type flow channel inserts and a
primary loop build of well known ferritic steels
B2) Higher Performance: high temperature operation with exit
temperatures up to 700°C, allowing the use of a BRAYTON cycle
(He turbines) power conversion system.
• For B2 SiC-composite FCI are required, and the materials used in the
primary loop (tritium extraction system, intermediate HX) must be
compatible with Pb-17 Li up to 700°C.
• Certainly, the high power operation is more attractive, but a
DCLL blanket with lead-lithium exit temperatures of ~ 500°C has
already a number of advantages compared to a HCLL blanket
(simpler structure, no need for large internal heat transfer
surfaces, less void space in tokamak inboard, smaller pumping
power, tritium control easier).
34
B. DCLL Material R&D for Power Plants (cont’d)
Common issues for both B1 and B2 cases
For both kinds of power plants, the following issues have to
be addressed by the material program:
a) Qualification of the RAFS and its welds for high
fluence irradiation,
b) Development of suitable methods for the purification
of Pb-17 Li (corrosion products, on-line Bi removal,
Po control),
c) Tritium extraction and control (e.g. tritium
permeation barriers)
35
B. DCLL Material R&D for Power Plants (cont’d)
Issues specific to the B1 and B2 cases
B1) Issues to be addressed for moderate performance power plants
d)
e)
Long time behavior of sandwich-type FCI ,
Suitable tritium extraction system (main candidate: vacuum
permeator with ferritic steel tubes),
Tritium control, especially the required reduction of permeation losses
in the Pb-17Li/steam HX
f)
B2) Issues to be addressed for high performance power plants
g)
Qualification of the SiC flow channel inserts for high temperature
applications (Interface temperature SiC/Pb-17 Li up to 800 C)
Development of suitable tritium extraction methods from Pb-17 Li,
leading to low tritium partial pressure (< 100 mPa), (candidate
method: vacuum permeator with Nb or Ta tubes)
h)
•
•
i)
operational limits on the impurities in Pb-17 Li, maintain extremely low
impurity level in the vacuum chamber,
impact of different materials in the primary Pb-17 Li loop : RAFS, SiCcomposite, Nb-or Ta permeator tubes, HX tube material.
Development of HX between primary (Pb-17Li) and secondary (He)
loops, candidate material for the HX tubes include Nb or Ta, SiCcomposites, Ni-base alloy with high Al content for forming protective
coatings.
36
Materials Needs for Ceramic Breeder TBM
• Ferritic steel structure: needs are satisfied by
those for the DCLL
• Pebble ceramics and beryllium: evaluate US
strategy for procurement (Do we want to
collaborate with EU, Japan and other parties on
development or just purchase from
EU/Japanese industry?)
• There is an active R&D program on ceramic
breeder blankets worldwide. The US materials
program may wish to explore making
contributions to this area.
37
Additional Material Issues Common to All
TBM Concepts and Parties
• Development of Be/FS joining capability
• Transition elements between FS in the TBM and the
coolant access pipes (preferably austenitic steel for
remote welding)
38
There are Immediate Needs for Human Resources
from the Material Program to participate in the TBM
activities and to work as part of the ITER TBM TEAM
Those are in addition to performing the R&D described earlier
For Example:
• to help the team perform design trade offs, more detailed
design, and analysis for the TBM test articles and
associated systems
• To provide advice on all aspects of materials limits and
considerations
• To help the team decide on key areas of and strategy for
international collaboration
• To assist in calculating costs for the TBM Program
(including R&D, mockups, test articles, ancillary
equipment, etc) (URGENT)
Approximate for FY06: 1.5 – 2.5 FTE
39