Transcript ITER Test Blanket Module (TBM) and ITER Nuclear Science and Engineering Mohamed Abdou VLT/USIPO Meeting, October 20-21, 2004

ITER Test Blanket Module (TBM)
and
ITER Nuclear Science and Engineering
Mohamed Abdou
VLT/USIPO Meeting, October 20-21, 2004
1
Outline
 Introduction
 Recent Progress in ITER TBM Program
 Issues and Recommended Approach
 FY05 and FY06 Plans
 Recommended Approach for Increased Effectiveness
2
Introduction
3
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, 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
International Team and representatives of the Parties.
4
Blanket Testing in ITER is one of ITER’s Key Objectives
Strong international collaboration among the ITER Parties is underway to provide the
science basis and engineering capabilities for ITER TBMs
Bio-Shield Plug
TBM Frame &
Shield Plug
Cryostat Plug
Breeder
Concentric
Pipe
Transporter
EU HCLL Test Module
FW
Cryostat
Extension
US Solid breeder submodule
Drain Pipe
5
Conceptual Liquid Breeder Port Layout and Ancillary equipment
ITER’s Principal Objectives Have Always
Included Testing Tritium Breeding Blankets
 “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 18–19
October 1987
 “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 (14–15 December
1994), and stated again in forming the Test Blanket Working Group
(TBWG)
6
ITER Blanket Testing is Essential to:
 Achieve a key element of the “ITER Mission”
 Establish the conditions governing the scientific feasibility of the
D-T cycle, i.e., determine the “phase-space” window of plasma,
nuclear, material, and technological conditions in which tritium
self-sufficiency can be attained
- The D-T cycle is the basis of the current world plasma physics and
technology program. There is only a “window” of physics and
technology parameters in which the D-T cycle is feasible. (If the D-T
cycle is not feasible the plasma physics and technology research would
be very different.)
- Examples of questions to be answered:
– Will tritium self-sufficiency allow low plasma-edge recycling?
– Are advanced physics modes acceptable?
– Is the “temperature window” for tritium release from solid breeders sufficient
for adequate TBR?
– Can circulated liquid metal breeders and coolants be used?
– Is there a blanket/material system that can exist in this phase-space?
– Are nuclear predictive capabilities (codes, data) adequate to predict accurate
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TBR in realistic system
ITER Blanket Testing is Essential to (cont’d):
 Achieve the most critical milestone in blanket and material
research: testing in the integrated fusion environment
(ITER construction and operation is for the next 30 years.
Without such fusion testing, material and blanket research
loses “focus”, relevance: Why are we doing any research in
these areas then?)
 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
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Nuclear Science and Engineering
is Critical to ITER and to the US Fusion Program
ITER will produce neutrons at a rate higher than any nuclear facility
ever constructed (including operating fission reactors). It will have
energetic neutrons, gamma rays, activated components, tritium,
etc.
 Challenges in predicting the radiation field:
• Complex geometry, large system
• “Deep radiation penetration” problem in predicting dose to external
components and radiation exposure to personnel
 ITER base design changes, TBMs, port-based diagnostic packages,
licensing, etc. will need more accurate nuclear simulation to
ensure robust design, safe operation, and correct interpretation of
experimental results.
 US has been a leader in Nuclear Science and Engineering for
fusion. Work on ITER base device and TBMs helps build US
knowledge, experience and competence needed to develop
practical and safe D-T fusion devices. (Building competence takes
decades.)
9
Recent Progress in ITER TBM Program
10
US Plasma Chamber Systems/Blanket Effort
has been redirected to support ITER
 With the US rejoining ITER, the Blanket/Chamber community concluded
that it is very important for the US to participate in the ITER Test Blanket
Module (TBM) Program (March 2003).
 Extensive deliberations have occurred in the US since March 2003 among
the community, DOE and VLT.
 Reached consensus on a general framework for the direction of activities
in the US Chamber/Blanket Program:
 Provide fusion nuclear technology (FNT) support for the basic ITER
device as needed
 Participate in ITER TBM program and redirect good part of resources
toward R&D for TBM
 Encourage partners in international collaborations, such as IEA and
JUPITER-II, to focus more on ITER TBM
 Important work has been carried out to implement the strategy.
• A study of ITER TBM issues and US options was initiated
• Some R&D was initiated
• Rejoined TBWG, strong participation
• The US interacted with all the other 5 parties to identify areas of
collaboration
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What should the US Blanket Options
be for ITER TBM?
 This has been a central question for the US community during FY 2004.
A study was initiated to select the two blanket options for the US ITER
TBM in light of new R&D results from the US and world programs over
the past decade.
 Key conclusions reached early in the study:
 Selection between solid and liquid breeders can not be made prior to fusion
testing in ITER.
 All Liquid Breeder Options have serious feasibility (“Go/No-Go”) issues, but
potential for higher performance. More assessment needed.
 Solid breeders are accepted by all parties.
 For the past year, the study has focused mostly on assessment of the
critical feasibility issues for liquid breeder concepts. Examples of
issues are MHD insulators, MHD effects on heat transfer, tritium
permeation, corrosion, SiC insert viability, and compatibility.
 The study has been led by the Plasma Chamber community with strong
participation of the Materials, Safety and PFC Programs. Many
international “Experts” in key areas participated in several meetings
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and provided important input.
US Selected Options for ITER TBM
 Select a helium-cooled solid breeder concept with ferritic steel structure
and neutron multiplier, but without an independent TBM.
 Support EU and Japan using their designs and their TBM structure and ancillary
equipment.
 Contribute only unit cell and sub-module test articles that focus on particular
technical issues of unique US expertise and of interest to all parties. (All ITER
Parties have this concept as one of their options.)
 Focus on testing Dual-Coolant liquid breeder blanket concepts with
potential for self-cooling. Develop and design TBM with flexibility to test
one or both of these two options:
 a helium-cooled ferritic structure with self-cooled LiPb breeder zone that uses
SiC insert as MHD and thermal insulator (insulator requirements in dual-coolant
concepts are less demanding than those for self-cooled concepts);
 a helium-cooled ferritic structure with low melting-point molten salt. Because of
the low electrical and thermal conductivity of molten salts, no insulators are
needed. (The key issues for MS are being addressed under JUPITER-II and no
additional work is planned under ITER-TBM.)
13
Helium-Cooled Pebble Breeder Concept for EU
Helium-cooled stiffening grid
Breeder unit
FW channel
- The US can provide small breeder units “inside” the EU SB structure.
- US Issues: Tritium Release and Thermomechanical Interactions
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ITER Test Blanket Module Activities
 Active participation in ITER test blanket working group
(TBWG) for test and infrastructure planning
 WSG participation
 Ancillary equipment and machine interface definition
 Evaluate blanket options for DEMO and evaluate R&D
results for key issues to select primary US blanket concepts
 Perform concurrently R&D on the most critical issues
required
 MHD flow with insulators and inserts
 Tritium recovery and control
 SiC inserts compatibility and failure modes
 Solid breeder / multiplier / structure / coolant interactions
 Develop engineering scaling and design, in collaboration
with ITER partners, for TBMs.
15
Test Blanket Working Group (TBWG)
 TBWG was proposed by SWG-1 and established by the ITER Council (IC-7
RECORDS, 14-15 December 1994)
 A formal and detailed charter of TBWG was developed
 During EDA, the US made major contributions to testing strategy,
engineering scaling, test port, frame design, and machine interface
 TBWG Main Objectives: Develop a Coordinated Blanket Test Program and
address the Interface between machine and the blanket modules
 The TBWG was re-established in new conditions (6 partners) in late 2003
and has had 3 meetings in FY04
TBWG Scope of Activities:
A. Provide the design documentation for the assessment of TBMs prepared by the
Parties, integration of TBMs into the ITER machine
B. Promote cooperation among the ITER Parties’ TBMs
C. Verify integration of the TBMs in the safety and environmental evaluations of
each ITER candidate site
D. Further develop the coordinated blanket test program
16
Two types of Solid Breeder TBMs have
been studied in FY04
1. Unit cells (3)
192.5 mm x 211 mm x 650 mm
2. Quarter-port Submodules
730 mm x 910 mm x 600 mm
• The proposal calls to share the port space to
test contemporaneously independent unit
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cells/submodules
Various analyses have been performed for TBM
designs to preserve key prototype parameters
Unit cell temperature
profile (edge-on design)
Be
Stress analysis shows smax =
268.4 MPa located at the
corner of the front inner wall
SB
He In
He out
The displacement profile
shows a non-uniform
distribution with a maximum
displacement of 3.51 mm
18
He in
He out
Sophisticated 2-D neutronics analysis shows
testing objective can be achieved for a proposed
NT TBM
5 10
-5
Right Configuration
Left Configuration
4 10
Layer#
-5
Layer#
1
3 10
1
2
3
-5
2
3
2 10
-5
5
6
3
4
7
4
1 10
-5
8
5
6
0
0
10
12
Left TBM Wall
Be Layer-Left Config.
Left VCP-Left Config.
Br1
Right TBM Wall
Be Layer- Middle
Be-Rt. Submdule
Be Layer-Rt. Config.
Rt. VCP- Left Config.
Left VCP-Rt. Config.
Rt. VCP-Rt. Config.
10
JA TBM
Finding:
Flat nuclear heating
and tritium production
profiles allow two
designs to be evaluated
in a ¼ port submodule
8
Breeder (Lft. Config.)
6
4
Be (Rt. Config.)
2
0
10
20
30
40
30
40
50
60
70
Distance from Frame, cm
Depth = 42 mm behind FW
Proposed NT TBM
20
9
50
60
70
Toroidal Distance from Frame, cm
80
Tritium production
profiles are nearly flat
over a reasonable
distance in the
toroidal direction
allowing accurate
measurements be
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performed
FS/He/PbLi Dual Coolant Blanket Concept:
Higher Performance Potential
with current generation ferritic structure
 The reason fusion pursued high temperature
structural materials is for high coolant
temperature.
 MHD effects in high-velocity channel flows
leads to very high primary stresses.
 IDEA of the Dual Coolant Concept*
– Cool structure with He so that FS can be
used. “Decouple” surface from bulk heating.
Self-Cooled
– Flow PbLi for self-cooling at low velocity
– Use a SiC insert to electrically and thermally
insulate the LM from the wall, so LM bulk
temperature can be higher than the wall
temperature (use the poor thermal and
electrical conductivity of SiC as an
advantage).
– Result: potential for high bulk temperature
with lower MHD pressure drop using
Ferritic Steel.
*Dual Coolant concepts proposed by ARIES and EU
Breeder Zone
ARIES-ST DCLL
Breeder Unit
20
Structure, Insert, and Breeder Temperatures
FS grid
Temperature
drop across the
FCI is 175 C
21
Complex geometry MHD codes already being
applied to DCLL blanket with SiC
Flow Channel Inserts
 2D and 3D codes
(developed for Liquid
walls) have been modified
for DCLL
 Initial results show strong
sidelayer jets at sSiC = 500
S/m with current DCLL
design
 2D and 3D codes give
conflicting results
concerning flow in the
“stagnant” gap region.
 Code improvements and
debugging, and continued
simulations planned for
FY05.
Strong negative flow jet near pressure
equalization slot not seen in 3D simulation
Velocity profile from
2D Simulation
Slice from 3D Simulation
Gap corner
jets not seen in 2D simulation
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Study of DCLL system and SiC FCI
feasibility has shown:
 SiC flow channel inserts in DCLL systems have an interesting
combination of effects that reduce pressure drop as compared to selfcooled LM systems
 Sensitivity to flaws and flow in complex geometry flow elements will dominate
pressure drop and must be carefully examined
 Flow balancing technique must be analyzed and tested in more detail.
 SiC FCI properties and fabrication must be established more definitively
 Compatibility of FCI in a PbLi/RAFM steel system is a critical
development issue
 Compatibility and corrosion are linked issues with DCLL design and MHD
behavior
 More definitive first and second level compatibility experiments and modeling
are necessary – definition of needed experiments could be an IEA Liquid
Breeder Topic
 Tritium control is a key issue which will also be affected by the baseline
MHD and subsequent FCI compatibility and corrosion
 Effective DCLL testing in ITER is possible and desirable
 MHD testing building towards integrated FCI performance
 Synergism with EU PbLi and RAFM program,
 Scaling and test program planning in the US underway
23
Technical Details from the US ITER-TBM
efforts this year were presented at ANS
TOFE Meeting in Madison
e.g.
 Abdou, "US Plans and Strategy for ITER Blanket Testing"
 Morley, “Thermofluid Magnetohydrodynamic Issues for Liquid
Breeders”
 Ying, “Engineering Scaling Requirements for Solid Breeder Blanket
Testing”
 Smolentsev, “MHD Effects on Heat Transfer in a Molten Salt Blanket
 Youssef, “Activation Analysis for Two Molten Salt Dual-Coolant
Blanket Concepts for the US Demo Reactor”
 Wong, “Assessment of Liquid Breeder First Wall and Blanket Options
for the Demo Design”
 Sawan, “Neutronics Assessment of Molten Salt Breeding Blanket
Design Options”
 Youssef, "On the Strategy and Requirements for Neutronics Testing
in ITER“
 …
24
J2 thermofluid experiments establish
reference database for molten salt
turbulence and heat transfer
Umean(DNS)
Umean(PIV)
Umean(DNS)
Umean(PIV)
20
20
Ave Velocity
u+
u+
Ave Velocity
10
10
0 0
10
Re=5300
2
1
10
y+
10
1
10
2
y+
1
Re=5300
uv(DNS)
uv(PIV)
0.8
uv(DNS)
uv(PIV)
0.8
0.6
0.6
uv

Re=11300
0 0
10
10
1
u v /u 
 Careful turbulence
measurements using PIV
technique show excellent
agreement with Direct
Numerical Simulation in
near wall and in channel
core (see figs right )
 Heat transfer reference
cases also showing good
agreement with standard
correlations.
 Special 2T gap magnet
with unobstructed viewing
access for laser
diagnostics designed and
nearing completion. To be
installed at UCLA in Dec.
2004.
0.4
0.4
0.2
0
0
0.2
Re stress
50
100
y+
Re = 5286
150
0
0
Re stress
Re=11300
100
200
300
y+
Re = 11,300
25
Two types of Solid Breeder TBMs have
been studied in FY04
1. Unit cells (3)
192.5 mm x 211 mm x 650 mm
2. Quarter-port Submodules
730 mm x 910 mm x 600 mm
• The proposal calls to share the port space to
test contemporaneously independent unit
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cells/submodules
Simulation and experimental
measurement of
CERAMIC BREEDER PEBBLE
BEDs
plate
Capacitance
displacement
sensor
Li4SiO4/Li2O pebble beds
0 .3
o
E xp e rim e nta l D a ta (7 0 0 C )
o
E xp e rim e nta l D a ta (7 5 0 C )
C re e p C o m p a ctio n stra in (% )
0 .2 5
o
E xp e rim e nta l D a ta (8 0 0 C )
o
F E M (7 0 0 C )
o
F E M (7 5 0 C )
0 .2
o
F E M (8 0 0 C )
0 .1 5
0 .1
0 .0 5
0
2
4
T im e (hour)
6
8
 Due to the high stress at the local contact
area, especially at the high temperature,
creep deformation occurs.
 Discrete Element Method (DEM) is able
to capture thermal deformation
characteristics and can provide the detail
stress distribution inside the pebble beds
at both steady and transient states.
Force distribution in the pebble bed simulated by DEM – line width
gives a measure of contact force
27
Lithium Ceramic (Solid Breeders) Pebble Bed
thermal conductivity data collected from all
world fusion program experiments
 effective conductivity of
Li-ceramic pebble bed
for temp. (400 to 850C)
 interface conductance
for temp. up to 550C.
2 0 .4 7 "
20
V acuum
V acuum
21
H e liu m
Ø 6 .0 0 "
6
2
1 .2 6 "
9 .3 8 "
3
4
1
1 .9 7 "
16
Thermal conductivity (W/m.K)
 Published studies show a wide spread in data, little error analysis, and
conflicting temperature dependence
 Significant work this
1.9
1.8
year at UCLA to
Li4SiO4, PF=62.5% [1]
1.7
design an
Li4SiO4, 0.5mm [13]
1.6
experimental apparatus
Li2TiO3, 60% [26]
1.5
Li2TiO3, 80% [26]
to measure
1.4
1.3
1.2
1.1
1.0
0.9
0.8
0.7
0.6
0.5
Li2TiO3, 59% [1]
Li2ZrO3, 53.4% [1]
Li2ZrO3, 63% [28]
Li2ZrO3, 60% [30]
Li2ZrO3, 63-65% [3]
Li2O, 62.1% [1]
Li2O, 48% [5]
Li4SiO4, PF=65% [27]
Li4SiO4, PF=64.4% [29]
17
18"
0 .9 7 "
0 .9 7 "
0 .5 "
5
1"
7
1"
8
1"
9
0 .8 8 "
18
19
0 .7 5 "
200
400
600
800
1000
1200
Temperature (C)
11
12
0 .7 5 "
0
10
13
14
0 .5 "
15
2"
28
Ø 9 .9 7 "
4 .0 0
1
P e b b le B e d
6
In s u la tio n (P a rtic le s B e d )
1 1 V a c u u m C F -F la n g e
1 6 V a c u u m E n v e lo p e
2
H e a te r
7
H e a t F lu x M e te r (S S 3 1 6 )
1 2 V H T G la s s -M ic a D is c
1 7 F le x ib le C o u p lin g
3
T h e rm o c o u p le
8
M a c o r D is c
1 3 V H T G la s s -M ic a D is c
1 8 T C F e e d th ro u g h
4
M a c o r R in g
9
M a c o r D is c
1 4 H T G la s s -M ic a D is c
1 9 V a c u u m P o rt
5
H e a t F lu x S e n s o rs D is c
1 0 V a c u u m C F -F la n g e
1 5 W a te r-C o o le d H e a t S in k
20 V acuum C ham ber
2 1 H e liu m C h a m b e r
Issues and
Recommended Approaches
29
Tritium Self-Sufficiency
Tritium self-sufficiency condition: Λa > Λr
Λr = Required tritium breeding ratio
Λr is 1 + G, where G is the margin required to account for tritium losses, radioactive
decay, tritium inventory in plant components, and supplying inventory for start-up of
other plants.
Λr is dependent on many system physics and technology parameters:



plasma edge recycling, tritium fractional burn-up in the plasma
tritium inventories (release/retention) in components
efficiency/capacity/reliability of the tritium processing system, etc.
Λa = Achievable tritium breeding ratio
Λa is a function of technology, material and physics.

FW thickness, amount of structure in the blanket, blanket concept (ITER detailed engineering is
showing FW may have to be much thicker than we want for T self sufficiency)




Presence of stabilizing/conducting shell materials/coils for plasma control and attaining
advanced plasma physics modes
Plasma heating/fueling/exhaust, PFC coating/materials/geometry
Plasma configuration (tokamak, stellerator, etc.)
Uncertainties in nuclear data required for accurate determination of TBR
For a few million dollars’ expenditure on test blanket modules, we will
acquire vital data and develop critical technologies
– an additional excellent return on the billions of dollars
invested in ITER.
30
Tritium Consumption and Production
Tritium Consumption in Fusion is HUGE!
55.8 kg per 1000MW fusion power per year
Production & Cost
• CANDU Reactors: 27 kg from over 40 years, $30M/kg (current)
• Fission reactors: 2-3 kg per year. It takes tens of fission reactors to supply
one fusion reactor.
$84M-$130M per kg, per DOE Inspector General*
Conclusions
• The cost of blanket development and ITER TBM is a fraction of the cost to
“purchase” tritium for a burning plasma facility such as ITER.
• “Availability” of external tritium supply for continued fusion development is
an issue.
• Large power DT facilities must breed their own tritium. (This is why ITER’s
extended phase was planned to install a tritium breeding blanket.)
*DOE Inspector General’s Audit Report, “Modernization of Tritium Requirements Systems”, Report DOE/IG-0632, December 2003,
available at www.ig.doe.gov/pdf/ig-0632.pdf
31
World Tritium Supply Would be Exhausted by 2025
if ITER Were to Run at 1000MW at 10% Availability
(OR at 500 MW at 20% availability)
Projected Ontario (OPG) Tritium
Inventory (kg)
30
25
CANDU Supply
20
w/o Fusion
15
1000 MW Fusion,
10% Avail, TBR 0.0
10
ITER-FEAT
(2004 start)
5
0
1995
2000
2005
2010
2015
2020
2025
2030
2035
2040
2045
Year
32
ITER Provides the First Integrated Experimental
Conditions for Fusion Technology Testing
• Simulation of all Environmental Conditions
Neutrons
Plasma Particles
Electromagnetics
Tritium
Vacuum
Synergistic Effects
• Correct Neutron Spectrum (heating profile)
• Large Volume of Test Vehicle
• Large Total Volume, Surface Area of Test
Matrix
33
Highlights of US Strategy for ITER TBM
(Evolved over the past year by the community, DOE and VLT)
 The US will seek to maximize international collaboration. There
is a need for all parties to collaborate, and to possibly consider
a more integrated plan among the ITER parties for carrying out
the R&D and construction of the test modules.
 ITER TBM should be viewed as a collaborative activity among
the VLT program elements. While the Blanket/Chamber
Program provides the lead role for ITER TBM, major
contributions from other programs (e.g., Materials, Safety,
PFC) are essential.
 The US community has now reached consensus on preferred
options for ITER TBM (see separate slide), following
assessment of new technical results obtained over the past
few years
34
Strategy for Testing Solid Breeder
(He/SB/Be/FS) Concepts

The US plans on unit cells and submodule tests to address
specific technical issues such as temperature window and
temperature control (not a fully independent TBM, uses other
Parties ancillary equipment)

The world program (particularly EU and Japan) in this area is
strong

There are very good reasons why the US program should
contribute to this area
1. the only universally accepted concept,
2. exciting issues with a lot of science,
3. Nearest term breeding capability for extended
phase in ITER
35
Present U.S. Solid Breeder R&D Effort
Carried out mostly in collaboration with other countries (IEA, JUPITER-II)
Solid Breeder Blanket Specifics:
 Focus on niche areas of solid breeder blanket material system
thermomechanics interactions (Primary organizations: UCLA,
Support: ORNL, PNL)
• design database on effective thermo-physical and mechanical
properties for breeder and beryllium pebble beds
• experiments and modeling development on evaluation of
thermomechanical states of blanket element pebble beds under
different loading conditions
Material/Blanket Experiments Interface:
 Development of Web based INTEGRATED FUSION MATERIALS
DATABASE (UCLA, UCBS, ANL, ORNL)
 Construction of VISTA (VIRTUAL INTERNATIONAL STRUCTURAL
TESTING ASSEMBLY) modelling tool, to evaluate a range of
potential interactions and failure paths (perform “Virtual
Experiments”). (UCLA, UCBS, FZK, ANL, ORNL)
36
DCLL Test Phases in ITER
H and D
Specific electromagnetic structure and MHD TBM:
 Structural TBMs reaction to field environment and various
transient plasma events – for instance “water hammer”
effect during rapid plasma current quench
 Scaled pressure drop tests, flow balance test and critical
velocity profiles affecting heat transfer in increasing field
strength
 SiC FCI integrity under MHD loading
Low Duty
DT
Thermofluid TBMs:
High Duty
DT
Partially-integrated Thermafluid TBMs:
 Scaled stress response of thermally loaded SiC inserts and
the effects of failures on pressure drop and thermal field
 Tritium production and permeation
 Corrosion and compatibility in nuclear field
 Radiation damage ~1 dpa in the inserts
 Tritium control
37
DCLL Concepts Issues:
SiC FCI Properties
A) Electrical and thermal conductivity of the SiC/SiC perpendicular to the
wall should be as low as possible to avoid velocity profiles with sidelayer jets and excess heat transfer to the He-cooled structure.
B) The inserts have to be compatible with Pb-17Li at temperatures up to
700-800°C.
C) Liquid metal must not “soak” into pores of the composite in order to
avoid increased electrical conductivity and high tritium retention. In
general “sealing layers” are required on all surfaces of the inserts.
- Even if the change in conductivity results in modest increase in
pressure drop, it could seriously affect flow balance.
D) There are minimum primary stresses in the inserts. However,
secondary stresses caused by temperature gradients must not
endanger the integrity under high neutron fluence.
E) The insert shapes must be fabricable and affordable.
38
DCLL Concept Issues:
MHD effects in self-cooled liquid
metal breeding regions
 Primary issue for blanket application and ITER testing is the
MHD pressure and flow distribution for complex geometry flow
elements:





SiC FCI overlap regions (stovepiping)
Defects in FCIs
Flow balancing sections
Turns, Field entrance/exit regions, Manifolds, Expansions/Contractions
Coaxial Pb-Li supply/return lines
 MHD velocities profiles can exhibit strong jets next to regions of
stagnation and even reversed flow
 Large temperature gradients can drive natural convection flows that
MHD effects do not damp – can swamp forced flow velocity in slow
moving breeder zone regions
 All of these MHD issues, through effects on temperature, velocity
shear and convection, and determination of LM-facing material, will
have effects on the corrosion and tritium permeation of the concept 39
Tritium Permeation
• Tritium Permeation is emerging as a
persistently serious issue for most (some
believe for ALL) concepts
• Developing acceptable solutions requires
integrated efforts among many programs:
Materials, Tritium, PFC, Safety, Plasma
Chamber
40
FY05 and FY06 plans
41
ITER TBWG Participation, Planning and
Documentation Requirements
 ITER TBWG ~quarterly meetings and contribute to the
final TBWG report on test strategy
 For both Solid Breeder and Dual Coolant Pb-17Li
Concepts, prepare and deliver to TBWG, a complete
preliminary Design Description Document containing





TBMs design
Fabrication
Qualification program plans
Optimization of TBM system
Provide initial TBM delivery date
42
Solid Breeder and Multiplier tasks
 Participate in the collaborative tasks that have been defined in the
IEA solid breeder (and liquid breeder subgroups), in which R&D
results improve/impact TBM designs
• design database on effective thermo-physical and mechanical
properties for breeder and beryllium pebble beds
• experiments and modeling development on evaluation of
thermomechanical states of blanket element pebble beds under
different loading conditions
43
IEA collaboration on solid breeder pebble bed time
dependent thermomechanics interactions/deformation
Primary Variables
• Materials
• Packing
• Loadings
• Modes of operation
Partially integrated
out-of-pile and
fission reactor tests
(NRG,ENEA)
Single/multiple effect experiments
(NRG, UCLA)
Finite Element Code
(ABQUS, MARC)
(NRG, FZK, UCLA)
Design Guideline and
Evaluation (out-of-pile &
in-pile tests, ITER TBMs)
Primary & Secondary
Reactants:
• Temperature magnitude/
gradient
• Differential thermal
stress/contact pressure
• Plastic/creep deformation
• Particle breakage
• gap formation
Discrete Element
Model (UCLA)
Thermo-physical and
Mechanical Properties
Consecutive equations
Database Experimental Program
(FZK, JAERI, CEA,UCLA)
Goal:
Performance/Integrity
prediction & evaluation
Irradiation Effect
(NRG)
44
Pb-17Li dual coolant
blanket concept TBM
Main R&D Tasks, Multi-year:
 Strong effort on development of simulation
capability (see slice of 3D simulation, right) and
experimental test plan, facility construction for LMMHD effects in closed channels with complex
geometry and heterogenous wall conductivity
Example simulation: MHD boundary
layer jets formed in a channel with
SiC flow channel insert
 Assessment, compatibility and property
experiments on the feasibility of the SiC insert
 DCLL Reference and scaled TBM design for
dual-coolant PbLi concepts in collaboration among
US Plasma chamber community
45
What experiments needed to
determine Pb-Li SiC compatibility?
 FY05 - More definitive capsule experiments required to show first
level of compatibility




More prototypic SiC composite should be used
Long exposure time to overcome any incubation period for wetting
Pressurization, impurities typical of RAFM steel
Careful post-examination of both SiC crucible for evidence of attack and
PbLi melt for accurate concentration of dissolved SiC
 As an intermediate step suitable models should be developed and the
basic material properties (most important diffusivity and saturation
concentration as functions of temperature) should be obtained from
the literature when available, and from dedicated experiments as
necessary.
 FY06 and beyond – Quantification of the allowable interface
temperature in a corrosion flow loop with relevant temperature
gradients, typical materials, and realistic flow conditions (Provided
SiC/PbLi passes first level of assessment )
46
JUPITER-II collaboration for
Plasma Chamber Systems
All experiments are directed to solve key feasibility issues for the
molten salt, Li/V and solid breeder ITER test modules.
1. Flibe REDOX control





Completed flibe purification process.
Completed flibe mobilization experiment.
Started hydrogen isotopes (D) permeation, diffusion and solubility measurements.
Preparation for the REDOX experiments.
Presented two papers at ICFRM and Be workshop.
2. Flibe heat transfer and flow mechanics




Measured straight pipe velocity profile and turbulent statistics.
Constructed 304SS heat transfer test section, with some initial data available.
New acrylic PIV attachment section constructed and tested.
Prepared for the MHD experiment with a US supplied magnet.
3. MHD coating development
 Coatings of AlN, Y2O3 and Er2O3 have been tested in Li up to 800C.
 Vacuum distillation system developed and tested to remove residue lithium from test coupons.
 Resistivity experiments conducted for Er2O3, Y2O3 and (Y,Sc)O3 to confirm sufficient resistivity for
MHD coating.
4. SiC/pebble bed thermomechanics experiments
 Two configurations were developed.
47
ITER Nuclear Analysis Needed in Three Areas
 Nuclear analysis for ITER TBM
• This effort is ongoing and funded as part of the US ITER TBM activities
 Nuclear support for basic ITER Device
• Nuclear support is needed particularly for US preferred procurement packages (diagnostics,
plasma heating and CD, module 18 (baffle), CS coils)
• This support will ensure that the US delivers components with great confidence in its successful
performance in ITER nuclear environment
• ~$200K (~1 FTE) a year needed. Small burden to the US ITER construction budget to support
the other programs
• The US neutronics community (primarily Universities) has strong interest in providing this
support. This allows training students and young scientists
 Development of CAD/Neutronics coupling
• Needed to facilitate the 3-D neutronics analysis of ITER components effectively
• Allows using geometric models directly from the actual design system, resulting in
more accurate results and reduce time between CAD-based design changes and
analysis
• ITER International Team requested that support
• Resources needed ~$100K (~0.5 FTE) per year for two years
48
Nuclear Analysis for TBM
 TBM designs will be developed and
modeled for 3-D neutronics
calculations with all design details
 The TBM 3-D model will be
integrated in the complete basic
ITER machine 3-D model
 Perform 3-D neutronics calculations
using the integrated model
 Neutronics calculations will provide
important nuclear environment
parameters (e.g., radiation damage,
tritium production, transmutations,
radioactivity, decay heat, and
nuclear heating profiles in the TBM)
that help in analyzing TBM testing
results
49
Nuclear Analysis for ITER Basic Machine
As ITER proceeds to construction, individual component designs become more
clearly defined and more accurate nuclear analyses are needed. Accurate
evaluation of the nuclear radiation level at these components is essential to
assess their survivability and performance capability
Special attention should be devoted to
components that are sensitive to radiation
and have penetrations with radiation
streaming concern
 Detailed 3-D nuclear
analyses needed in
support of US
procurement packages
 Radiation leakage
through holes and other
penetrations must be fully
assessed to establish
activation levels for
personnel access
Module 18
50
CAD-Based MCNP Development
 Use Sandia’s CGM interface to evaluate CAD directly from MCNP
»
CGM provides common interface to multiple
CAD engines, including voxel-based models
MCNP
MCNP
Native
Geometry
CGM
Pro/E
Voxels
 Benefits:
» Dramatically reduce turnaround time from
Clothespin, w/helical
ARIES-CS Plasma
CAD-based design changes
spring surface
– Identified as key element of ITER
Neutronics analysis strategy
» No translation to MCNP geometry commands
– Removes limitation on surface types
– Robustness improved by using same engine for CAD and MCNP
» Can handle 3D models not supported in MCNP
 Status: prototype using direct CAD query from MCNP
 Issues/plans:



ACIS
Benchmarking the current prototype version of MCNP/CGM for ITER analyses. Explore
various approaches to transferring geometry from the ITER CAD design system (possibly
CATIA) to the ACIS-based MCNP/CGM
The most immediate limitation is that it is slower than MCNP alone. The focus will be to
speed up the ray-tracing portion of the code (lots of acceleration techniques possible)
51
Goal: speed comparable to MCNP, but using direct CAD evaluation
Recommended approaches to
increased effectiveness
52
Programmatic Issues Requiring DOE, VLT, and
US ITER Management Attention
 Include ITER-TBM and Nuclear Science and Engineering in US
ITER Organization
 a “box/area” for TBM in the Organizational Chart
– this is essential to keep ITER link for TBMs that will
influence basic machine design and building space
– budget from DOE can remain separate from ITER
construction funds
 a “box/area” for Nuclear Science and Engineering
– Small budget should come from ITER funds to support
US procurement packages in diagnostics, module 18, etc.
 Recognize ITER-TBM as a US Program and organize it as a
community effort
 Led by Plasma Chamber but with major contributions from
Materials Program and significant contributions from PFC,
Safety, and Tritium Programs
 Form Steering Committee with members from TBWG ,
53
Plasma Chamber, Materials, PFC, Safety, etc.
Programmatic Issues Requiring DOE, VLT, and
US ITER Management Attention (Cont.)
 Resources Needed
 Technical expertise is available
 Required financial resources will be estimated over the next
few months, but preliminary estimates show they are modest,
not much beyond present budgets (mostly “refocus” and
“rebalance” with modest increases)
 Resources consist of:
a) annual R&D: starting now; already started under Plasma
Chamber with some support from Materials, Safety , and
PFC. Need to do more under other programs.
• Need a lot more from the Material Program in
particular (discussions to refocus parts of Materials
to serve ITER-TBM needs is underway)
b) cost of constructing Test Articles (to be inserted in
ITER): needed 7 years from now . It is modest and can
be adjusted based on the role US wants to play.
54