Transcript Clearance Issues for Advanced Fusion Power Plants

ARIES-CS
Radial Build Definition and
Nuclear System Characteristics
L. El-Guebaly,
P. Wilson, D. Henderson, T. Tautges,
M. Wang, C. Martin, J. Blanchard
and the ARIES Team
Fusion Technology Institute
UW - Madison
US / Japan Workshop on Power Plant Studies
January 24 - 25, 2006
UCSD
Nuclear Areas of Research
Radial Build Definition:
– Dimension of all components
– Optimal composition
Neutron Wall Loading Profile:
– Toroidal & poloidal distribution
– Peak & average values
Blanket Parameters:
– Dimension
– TBR, enrichment, Mn
– Nuclear heat load
– Damage to FW
– Service lifetime
High-performance
shielding module at min
Radiation Protection:
– Shield dimension & optimal
composition
– Damage profile at shield,
manifolds, VV, and magnets
– Streaming issues
Activation Issues:
– Activity and decay heat
– Thermal response to
LOCA/LOFA events
– Radwaste management
2
Nuclear Task Involves Active
Interaction with many Disciplines
Prelim. Physics
(R, a, Pf, ∆min, plasma
contour, magnet CL)
NWL Profile
Design
Requirements
Blanket Concept
1-D Nuclear Analysis
( peak, average, ratio)
(∆min, TBR, Mn, damage, lifetime)
no ∆min match
Radial Build Definition
or insufficient breeding
Activation Assessment
(Activity, decay heat, LOCA/LOFA,
Radwaste classification)
@ ∆min and elsewhere
(Optimal dimension and composition,
blanket coverage, thermal loads )
Blanket Design
Safety Analysis
3
Systems Code
(R, a, Pf)
Init. Magnet
Parameters
Init. Divertor
Parameters
3-D Neutronics
(Overall TBR, Mn)
CAD Drawings
Stellarators Offer Unique Engineering
Features and Challenges
• Minimum radial standoff at min controls machine size and cost.
 Well optimized radial build particularly at min
• Sizable components with low shielding performance (such as blanket and He
manifolds) should be avoided at min.
• Could design tolerate shield-only module (no blanket) at min?
Impact on TBR, overall size, and economics?
• Compactness mandates all components should provide shielding function:
– Blanket protects shield
– Blanket and shield protect manifolds and VV
– Blanket, shield, and VV protect magnets
• Highly complex geometry mandates developing new approach to directly
couple CAD drawings with 3-D neutronics codes.
• Economics and safety constraints control design of all components from
beginning.
4
ARIES-CS Requirements Guide
In-vessel Components Design
Overall TBR
1.1
Damage to Structure
200
3%
dpa - advanced FS
burn up - SiC
1
appm
1019
2
6x10-3
> 1011
n/cm2
mW/cm3
dpa
rads
Machine Lifetime
40
FPY
Availability
85%
(for T self-sufficiency)
(for structural integrity)
Helium Production @ Manifolds and VV
(for reweldability of FS)
S/C Magnet (@ 4 K):
Fast n fluence to Nb3Sn (En > 0.1 MeV)
Nuclear heating
dpa to Cu stabilizer
Dose to electric insulator
5
Reference Dual-cooled LiPb/FS Blanket
Selected with Advanced LiPb/SiC as Backup
Breeder
Multiplier
Structure
FW/Blanket
Coolant
Shield
Coolant
VV
Coolant
Be
FS
Flibe
Flibe
H2O
LiPb (backup)
–
SiC
LiPb
LiPb
H2O
LiPb (reference)
–
FS
He/LiPb
He
H2O
Li4SiO4
Be
FS
He
He
H2O
LiPb
–
FS
He/LiPb
He or H2O
He
Li
–
FS
He/Li
He
He
Internal VV:
Flibe
External VV:
6
FW Shape Varies Toroidally and Poloidally a Challenging 3-D Modeling Problem
Beginning
of Field
Period
3 FP
R = 8.25 m
min
=0
 = 60
Middle
of Field
Period
7
UW Developed CAD/MCNP Coupling Approach to
Model ARIES-CS for Nuclear Assessment
Neutronics
input file
CAD geometry file
CAD based Monte Carlo Method
CAD geometry engine
Monte Carlo
method
Ray object
intersection
• Only viable approach for ARIES-CS 3-D neutronics modeling.
• Geometry and ray tracing in CAD; radiation transport physics in MCNPX.
• This unique, superior approach gained international support.
• Ongoing effort to benchmark it against other approaches developed abroad
(in Germany, China, and Japan).
• DOE funded UW to apply it to ITER - relatively simple problem.
8
3-D Neutron Wall Loading Profile
Using CAD/MCNP Coupling Approach
=0
 = 60
Peak 
– R = 8.25 m design.
– Neutrons tallied in discrete bins:
– Toroidal angle divided every 7.5o.
– Vertical height divided into 0.5 m segments.
– Peak ~3 MW/m2 at OB midplane of  = 0o
– Peak to average = 1.52
9
IB
OB
Novel Shielding Approach Helps
Achieve Compactness
Benefits:
– Compact radial build at min
– Small R and low Bmax
– Low COE.
WC Shield-only Zone
(6.5% of FW area)
Transition Zone
(13.5% of FW area)
Challenges:
– Integration of shield-only and transition
zones with surrounding blanket.
– Routing of coolants to shield-only zones.
– Higher WC decay heat compared to FS.
10
Toroidal / Radial Cross Section
(R = 8.25 m )
Vacuum Vessel
Magnet
Gap
35
LiPb & He Manifolds
28
FS-Shield
5
Back Wall
28
WC-Shield-II
38
54
Non-uniform
Blanket
Divertor System
4
38
Blanket
(10% of FW area)
FW
13
WC-Shield-I
8
17
5 cm
SOL
Plasma
Nominal Blanket/shield
Zone (~80%)
|
Transition
Region (~13.5%)
11
|
WC-Shield only
Zone (~6.5%)
min = 119 cm
28
Radial Build Satisfies Design Requirements
(3 MW/m2 peak )
28
35
>2
28
25 cm
Breeding
Zone-II
1.5 cm FS/He
0.5 cm
SiC Insert
|
10
 > 181 cm
38
FW
SOL
Plasma
Shield
Manifolds
2
28
1.5 cm FS/He
VV
|
min = 119 cm
12
Thickness
(cm)
Blanket/
Shield
Zone
Thickness
(cm)
2
Gap + Th. Insulator
5
|
WC Shield-II
(permanent)
17
5
Magnet
≥2 2.2 31
External Structure
25 cm
Breeding
Zone-I
Permanent Components
External Structure
63
SOL
Plasma
5|
|
|
Gap + Th. Insulator
Replaceable
Shield
Only
Zone
@ min
Blanket Design Meets Breeding
Requirement
1.5
3.8cm He-Coole d FW
He -Coole d FS-Shie ld
1.0
0.5
0.0
Full
Blanke t
0
20
40
60
80
Thickness of Breeding Zone (cm)
• Local TBR approaches 1.3
• 3-D analysis confirmed 1-D local TBR estimate for full blanket coverage.
• Uniform and non-uniform blankets sized to provide 1.1 overall TBR based
on 1-D results combined with blanket coverage. To be confirmed with
detailed 3-D model.
13
Preliminary 3-D Results Using
CAD/MCNP Coupling Approach
Model*
1-D
3-D
Local TBR
1.285
1.316
± 0.61%
Energy multiplication (Mn)
1.14
1.143
± 0.49%
Peak dpa rate (dpa/FPY)
40
39.4
± 4.58%
FW/B lifetime (FPY)
5
5.08
± 4.58%
156
1572
13
71
18
1830
145.03
1585.03
9.75
62.94
19.16
1821.9
±1.33%
±1.52%
± 6.45%
± 2.73%
± 5.49%
± 0.49%
Nuclear heating (MW):
FW
Blanket
Back wall
Shield
Manifolds
Total
Future 3-D analysis will include blanket variation, divertor system
and penetrations to confirm 1.1 overall TBR and Mn.
____________________________
* Homogenized components. Uniform blanket everywhere. No divertor. No penetrations. No gaps.
14
25 cm
Breeding
Zone-II
He Tube
1.5 cm FS/He
(32 cm ID)
0.5 cm
SiC Insert
|
?
28
10
 > 181 cm
38
FW
SOL
Plasma
Shield
Manifolds
28
1.5 cm FS/He
VV
|
min = 119 cm
15
Thickness
(cm)
Blanket/
Shield
Zone
Thickness
(cm)
2
Gap + Th. Insulator
5
|
WC Shield-II
(permanent)
17
5
Magnet
≥2 2.2 30
External Structure
35
External Structure
25 cm
Breeding
Zone-I
28
Local Shield
|
63
SOL
Plasma
5|
Gap + Th. Insulator
He Access Tubes Raise Streaming Concern
Shield
Only
Zone
@ min
He Access Tube
Hot spots at
VV and magnet
40
19
Peak Fluence at Magnet (10
n/cm2 @ 40 FPY)
FW/Blanket
Local Shield Helps Solve
Neutron Streaming Problem
30
20
10
Limit
0
I-D
2-D
No Tube
No Local Shield
Ongoing 3-D analysis will
optimize dimension of local shield
16
with Tube
2-D w ith 20 cm
Local Shield
Key Design Parameters for
Economic Analysis
LiPb/FS/He
LiPb/SiC
1.19
1.14
(reference)
min
(backup)
Overall TBR
1.1
1.1
Energy Multiplication (Mn)
1.14
1.1
40-45%*
55-63%*
FW Lifetime (FPY)
5
6
System Availability
~85%
~85%
Thermal Efficiency (th)
__________
* Depending on peak .
Integral system analysis will assess impact of
min, Mn, and th on COE
17
Activation Assessment
• Main concerns:
– Decay heat of FS and WC components.
– Thermal response of blanket/shield during LOCA/LOFA.
– Waste classification:
- Low or high level waste?
-Any cleared metal?
– Radwaste stream.
18
Decay Heat
10
6
FW
3
Decay Heat (W/m )
WC Shie ld-I
10
5
10
4
FS of Blank e t-I
LiPb
10
3
1d
1h
1w
2
10
0
10
10
1
10
2
10
3
10
4
10
5
10
6
10
7
Time After Shutdown (s)
WC decay heat dominates at 3 h after shutdown
19
Thermal Response during
LOCA/LOFA Event
Nominal Blanket/Shield
Tmax ~ 711 oC
Shield-only Zone
Tmax ~ 1060 oC
740 C temp limit
740 C temp limit
1s
1m
1h
1d
1 mo
1s
1m
1h
1d
1 mo
• 3-coolant system  rare LOCA/LOFA event (< 10-8/y)
• Severe accident scenario: He and LiPb LOCA in all blanket/shield modules &
water LOFA in VV.
• For blanket, FW temperature remains below 740 oC limit.
• WC shield modules may need to be replaced. More realistic accident scenario
will be assessed.
20
Waste Management Approach
• Options:
– Disposal in LLW or HLW repositories
– Recycling – reuse within nuclear facilities
– Clearing – release to commercial market, if CI < 1.
• Repository capacity is limited:
– Recycling should be top-level requirement for fusion power
plants
– Transmute long-lived radioisotopes in special module
– Clear majority of activated materials to minimize waste
volume.
21
ARIES-CS Generates Only
Low-Level Waste
WDR
Replaceable Components:
FW/Blanket-I
WC-Shield-I
0.4
0.9
Permanent Components:
WC-Shield-II
FS Shield
Vacuum Vessel
Magnet
Confinement building
0.3
0.7
0.05
<1
<< 0.1
LLW (WDR < 1) qualifies for near-surface disposal
or, preferably, recycling
22
Majority of Waste (74%) can be
Cleared from Regulatory Control
8000
In-vessel components
cannot be cleared
74%
IAEA Clearance Index
3
Volume (m )
6000
10
11
10
9
10
7
10
5
10
3
10
1
Blank e t
Shie ld
Vacuum Ve s s e l
2000
14%
M agne t
7%
5%
0
Building
Blanket
Shield/VV
Magnet
Building
1y
1h
Lim it
1d
100y
Dispose
or Recycle
-1
10
4000
0
10
2
10
10
4
10
6
10
8
10
10
Time After Shutdown (s)
23
Clear
Building Constituents can be Cleared
1-4 y after Plant Decommissioning
Building composition: 15% Mild Steel, 85% Concrete
4
Inne rm os t
Se gm e nt
M ild Ste e l
Clearance Index
IAEA
10
10
10
2
US
0
Lim it
-2
1h
10
Clearance Index
10
3.5 y
1d
100y
-4
10
0
10
2
10
4
10
6
10
8
10
10
10
4
10
3
10
2
10
1
10
0
10
-1
10
-2
10
-3
10
-4
10
Time After Shutdown (s)
Inne rm os t
Se gm e nt
Concre te
IAEA
US
Lim it
1h
0
10
2
10
1.3 y
1d
4
100y
10
6
10
8
Time After Shutdown (s)
24
10
10
Stellarators Generate Large Radwaste
Compared to Tokamaks
not compacted, no replacements
Stellarators
3.0
FW
FS
V
2.5
Bucking
Cylinder
FS
Blanket
Back
Wall
3
3
Volume (10 m )
2.0
Shield
SiC
SiC
Coil
Structure
1.5
FS
V
Manifolds
V
3
(D- He)
1.0
SiC
0.5
0.0
ARIES -
I
III
II
IV SPPS
RS
ST
AT
R = 8.25 m
Blanket/Shield/Vacuum Vessel/Magnet/Structure
3.5
VV
WP
CS
Means to reduce ARIES-CS radwaste are being pursued
(more compact machine with less coil support and bucking structures)
25
Well Optimized Radial Build Contributed
to Compactness of ARIES-CS
m
ARIES-ST
Spherical Torus
3.2 m
8
|
6
4
ARIES-AT
Tokamak
5.2 m
2
0
Stellarators
2005
2000
2000
1996
1987
1982
|
ARIES-CS
7m
5
8.25 m
HSR-G
18 m
SPPS
14 m
FFHR-J
10 m
10
15
ASRA-6C
20 m
20
UWTOR-M
24 m
25
Average Major Radius (m)
Over past 25 y, stellarator major radius more than halved by advanced physics and
technology, dropping from 24 m for UWTOR-M to 7-8 m for ARIES-CS,
approaching R of advanced tokamaks.
26
Concluding Remarks
• Novel shielding approach developed for ARIES-CS. Ongoing
study is assessing benefits and addressing challenges.
• CAD/MCNP coupling approach developed specifically for
ARIES-CS 3-D neutronics modeling.
• Combination of shield-only zones and non-uniform blanket
presents best option for ARIES-CS.
• Proposed radial build satisfies design requirements.
• No major activation problems identified for ARIES-CS.
• At present, ongoing ARIES-CS study is examining more compact
design (R < 8 m) that needs further assessment.
27
ARIES-CS Publications
L. El-Guebaly, R. Raffray, S. Malang, J. Lyon, and L.P. Ku, “Benefits of Radial Build Minimization and Requirements Imposed on
ARIES-CS Stellarator Design,” Fusion Science & Technology, 47, No. 3, 432 (2005).
L. El-Guebaly, P. Wilson, and D. Paige, “Initial Activation Assessment of ARIES Compact Stellarator Power Plant,” Fusion Science
& Technology, 47, No. 3, 440 (2005).
M. Wang, D. Henderson, T. Tautges, L. El-Guebaly, and X. Wang, “Three -Dimensional Modeling of Complex Fusion Devices Using
CAD-MCNP Interface,” Fusion Science & Technology, 47, No. 4, 1079 (2005).
L. El-Guebaly, P. Wilson, and D. Paige, “Status of US, EU, and IAEA Clearance Standards and Estimates of Fusion Radwaste
Classifications,” University of Wisconsin Fusion Technology Institute Report, UWFDM-1231 (December 2004).
L. El-Guebaly, P. Wilson, and D. Paige, “Evolution of Clearance Standards and Implications for Radwaste Management of Fusion
Power Plants,” Journal of Fusion Science & Technology, 49, 62-73 (2006).
M. Zucchetti, L. El-Guebaly, R. Forrest, T. Marshall, N. Taylor, K. Tobita, “The Feasibility of Recycling and Clearance of Active
Materials from Fusion Power Plants,” Submitted to ICFRM-12 conference at Santa Barbara (Dec 4-9, 2005).
L. El-Guebaly, R. Forrest, T. Marshall, N. Taylor, K. Tobita, M. Zucchetti, “Current Challenges Facing Recycling and Clearance of
Fusion Radioactive Materials,” University of Wisconsin Fusion Technology Institute Report, UWFDM-1285. Available at:
http://fti.neep.wisc.edu/pdf/fdm1285.pdf
L. El-Guebaly, “Managing Fusion High Level Waste – a Strategy for Burning the Long-Lived Products in Fusion Devices,” Fusion
Engineering and Design. Also, University of Wisconsin Fusion Technology Institute Report, UWFDM-1270. Available at:
http://fti.neep.wisc.edu/pdf/fdm1270.pdf
R. Raffray, L. El-Guebaly, S. Malang, and X. Wang, “Attractive Design Approaches for a Compact Stellarator Power Plant” Fusion
Science & Technology, 47, No. 3, 422 (2005).
J. Lyon, L.P. Ku, P. Garabedian, L. El-Guebaly, and L. Bromberg, “Optimization of Stellarator Reactor Parameters,” Fusion Science
& Technology, 47, No. 3, 4414 (2005).
R. Raffray, S. Malang, L. El-Guebaly, and X. Wang, “Ceramic Breeder Blanket for ARIES-CS,” Fusion Science & Technology, 48,
No. 3, 1068 (2005).
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