Petti IAEA - Nuclear Sciences and Applications - IAEA
Recent Accomplishments
and Future Directions in the
US Fusion Safety &
Environmental Programs
David Petti
IAEA TCM on Fusion Safety
Vienna, Austria
July 12, 2006
Outline
• STAR
• Dust
• Tritium Material Interactions and Permeation
• Fusion Safety Codes
• Risk
• Waste Management
• MFE Safety Design
• IFE Safety Design
• Summary
Dust/Debris Characterization
Home of the STAR Facility at the
Idaho National Laboratory
TFTR DIII-D C-MOD
Tore Supra
NOVA
Tritium Plasma Experiment TPE Plasma
W brush
samples
Tritium Uptake in Materials
Fusion Safety Chemical
Reactivity Experiments
STAR
A National User Facility Providing a
Valuable Resource for Researchers Across
the Nation and Around the World.
Be sample
after exposure
to ion beam
Molten Salt Tritium/
Chemistry Experiments
Tritium Infrastructure
Systems
Fusion Safety Mobilization Testing
Mo alloy samples after
exposure to air
Glovebox TCS
MS tritium exp
MS corrosion exp
Tritium SAS
STAR Floorplan Layout
Stack monitor
TPE
D-ion implantation
2LiF-BeF 2 preparation,
purification and testing
15,000 Ci tritium limit
Segregation of operations
Gloveboxes and hoods
Tritium cleanup system
Once-through room ventilation
Chemical reactivity
Star Tritium Storage and Assay System
SAS glovebox Setup
Secondary inlet
Primary outlet
U-bed-1
U-bed-2
SAS manifold with U-beds
SAS manifold and
vacuum pumps
STAR Tritium Cleanup System (TCS)
Condenser
MS water
Outlet IC
Blower
Inlet
Catalyst beds
Inlet sample loop
Control
Heat exchanger
Mole sieve bed
Tritium is now on-site at STAR
• Useable tritium inventory now 1300 Ci
– 300 Ci in equimolar H2 :D2 :T2 calibration standard
– 1000 Ci T2 available for experiments
– shipments from SRS limited to 1000 Ci with
standard TYPE-A shipping container
Shipping Vessel (1 available)
May 1995 - First tritium plasma
experiments at TSTA
September 2000 - Final tritium
plasma experiments at TSTA
February 2001 - Begin D&D
efforts of TPE
December 2001 - Final pump
out of system; close all valves
January 2002 - Preparation
for shipment
Timeline of TPE at LANL and INL
March 2002 - Extraction,
loading, and transport
April 8, 2002 - Depart LANL
April 10, 2002 - Arrival at INL
STAR Facility
Summer/
Fall 2002 -
Uncrate and
decon ancilliary
components
January 2003 -
Modify plans
for location of
experiment,
decide on PermaCon structure, initiate facility
design changes. June 2003- PermaCon installed,
TPE glovebox uncrated. 2004- Reassembly and
system interface design activities
Spring &
Summer 2005
Electrical
Service
re-design
& construction,
experiment
& facility
interface
assemblies
completed
Fall 2005 - Integrated systems testing initiated
December
2005 - First
plasma testing
(non-tritium)
Planned Research Agenda for TPE
• Study uptake, retention and permeation in PFCs
– Measurement of bulk tritium transport properties
(diffusivity, solubility, dissociation/recombinatino rates)
– Monomaterials (Be, W, C)
– Mixed materials
– Bonded and/or duplex structures
– Effects of neutron dose and irradiation temperature on
tritium trapping
• Certification of these structures for ITER
• Use of tritium in the plasma will enable low level
measurements needed for such research
Flibe Tritium Experiment: Design and testing of
tritium handling systems and diagnostics
Flow meter
Ar D 2
T 2
Vacuum
pump
Flow meter
Flow meter
Pressure gauge
Cap
exhaust
HF trap
High temperature salt
Flibe
Conceptual layout proposed by Fukada et al.
Ni
Gas chromatograph
or QMS
or ionization chamber
dip Be if Redox control is successful
• Tritium provided in
pressurized vessel
containing D 2 /T 2 mixture
• Glovebox setup to contain
potential leaks
• Localized tritium cleanup will
be connected
• GC column for H isotope
separation has been tested
with tritium; works well but
needs calibration
• Develop DF/TF generator if
schedule permits
Permeation Coating Barrier Experments
Thermal Cycle Performance of He Pipe Permeation Barriers
• simulates thermal stress
degradation of permeation barrier
coatings for He pipes
• configuration matched to TBM
design for coated components
• utilize tritium for barrier
technology qualification
• external thermal cycles followed
by testing in permeation rig for
integrated effects
• in-situ thermal cycling in
permeation rig for barrier dynamic
response
Dust generation
ITER Key Issues: chemical reactivity, radioactivity
content, dust explosions.
Science: understand and model underlying formation
mechanisms to estimate inventories expected in fusion
ITER Dust Strategy from EDA
Demonstration
of the filtered
vacuum
collection
technique
Key scientific issues
needing resolution
TFTR
JET
Tore Supra
LHD
NOVA
Measurements of dust characteristics are an active
area of research
Particle Size Distribution, Specific Surface Area, Surface Mass Density,
Composition, Shape and Tritium Content
DIII-D
Comparison of Size Distributions
Machine
Alcator-Cmod
TEXTOR
ASDEX-Upgrade
Lower
Regions
0.66 + 2.82
0.88 + 2.63
1.58 + 2.80
27 + (-)
5-20 + (-)
2.68 + 2.89
2.21 + 2.93
8.59 + 2.67
1.12 + 1.90
CMD (μm) + GSD
Middle
Regions
0.60 + 2.35
1.60 + 2.33
1.53 + 2.80
-
-
2.98 + 2.94
3.69 + 2.81
6.31 + 2.39
0.76 + 2.03
Upper
Regions
0.89 + 2.92
1.22 + 2.03
-
3.32 + 2.94
3.59 + 3.08
8.73 + 2.09
0.90 + 1.93
R&D continues to resolve the issues
-
-
Specific Surface Area, m 2 /g
100
10
1
Specific Surface Area
0.1
fully dense C
fully dense Mo
TFTR
DIII-D
Alcator-Cmod
Tore Supra
0.01
1
ASDEX-Upgrade
NOVA
JET
ATJ graphite
10
Mo
C
100
Mean Volume-Surface Diameter, d (μm)
MVS
Median Particle Diameter (μm)
15
10
5
0
10 0
LHD
Dust Size versus Surface Mass Density
ASDEX-Upgrade
Tore Supra
CMOD
10 1
10 2
Surface Mass Density (mg/m 2 )
LHD
ASDEX-Upgrade
Tore Supra
CMOD_(98)
DIIID_(98)
10 3
DIII-D
10 4
Research Agenda for Dust
• Continued characterization in existing tokamaks
• Mobilization testing
• Chemical reactivity of dust in grooves
• Monitoring and removal evaluation
• Improved strategy for ITER
Major Accomplishments in
Risk Assessment for Fusion
• Work on component failure rate data to support quantitative safety
assessment continues to be very successful.
– Initially, component failure rates were collected from handbooks
and applied to fusion.
– Now, through IEA Task 5, we collect fusion facility operating
experience data from TLK, TPL, and the former TSTA; and tokamak
data from DIII-D and JET.
– Independent data sets validate the failure rate values.
• Occupational safety is a new area for risk assessment.
– ITER IT has plans to perform a room-by-room overall assessment
of the ITER facility to identify occupational hazards
– Occupational injury rates have been collected from several US
tokamaks and large particle accelerators
– WE-FMEA method was developed to address highly hazardous
equipment failures in a fusion facility, such as high energy pipe
breaks that have caused worker fatalities in the power industry
The Hazard Zone of the Worker Exposure -
Failure Modes and Effects (WE-FMEA)
INL FSP Support of the ITER Project
• The FSP is supporting the ITER Project through two Implementing Task Agreements
(ITA), established in 2004.
• U. S. ITER Fusion Safety Code Support ITA
– Provide International Team (IT) with the latest fusion versions of the MELCOR and
ATHENA codes, documentation, validation, and support and assistance at
operation of the codes.
– Delivered MELCOR 1.8.5, upgraded ice layer model for cryogenic surfaces, and
developed a beryllium dust layer oxidation model for MELCOR
– Assist the ITER IT in producing the QA documentation for MELCOR and safety
analyses for ITER’s Report on Preliminary Safety (RPrS)
• U.S. ITER Magnet Safety Task Agreement
– Update the MAGARC code to current ITER design for TF and PF coils, include
new R&D results on insulation failure behavior at elevated temperatures, apply the
MAGARC code to various ITER magnet safety studies
– Upgraded insulation and magnet parameters, implemented arc limit model,
applied MAGARC to ITER-FEAT TF and PF magnet unmitigated quench events
– Develop magnet Busbar arc capability for MAGARC
MELCOR Code Heat Structure Dust Layer
Oxidation Model
• A beryllium dust layer oxidation
model was developed for ITER
to simulate oxidation of a dust
layer of dust that has settled
onto a heated surface inside of
a fusion device
• This model is based on
measured oxidation reaction
rates for fully dense beryllium,
binary gaseous diffusion of
oxygen or steam into the dust
layer, and BET measured
specific surface area for
beryllium dust
• Application is for slow vacuum
vessel pressurization events
from in-vessel loss-of-cooling
accidents (LOCAs) and loss-ofvacuum
accidents (LOVAs)
Beryllium oxidation rate (kg/m 2 -s)
10
10
10
10
10
0
-3
-6
-9
-12
INL92
88% dense MELCOR dust layer model
( Dust =0.7g/cm 3 , d p = 20μm)
INL fully dense
INL98 88% dense
INL Dust
5 10 15 20
10,000/T (K)
MAGARC Poloidal Field Coil Development
• The MAGARC code was recently
modified to analyze unmitigated
quench events in ITER poloidal
field (PF) magnets
• This modification include
the electrical
characteristics of the twoin-hand
winding pair of the
ITER PF coils and limits
on the number of arcs that
can form in the magnet
during unmitigated quench
events based on an
energy minimization
principle.
Voltage (V)
0 500 1000 1500
0.00
Radial direction
0.25
Height (m)
Time 110s
Winding pair
0.50 0.75 1.00
Axial direction
Inline arcs
voltage drops
0.00 0.18 0.36 0.540.72
Width (m)
Radial direction
Axial direction
Number of arcs
MAGARC Poloidal Field Coil Application
• MAGARC code
application to
unmitigated
quench events in
ITER poloidal field
(PF) magnets
8
6
4
2
Inline
Radial
Axial
Lead voltage drop (V)
0
0 150 300
Time (s)
450 600
Quench fraction
3000
2000
1500
1000
500
0
0.8
0.6
0.4
0.2
0.0
0 150 300 450 600
Time (s)
Coil current (kA)
0 150 300 450 600
Time (s)
Melt volume (m 3 )
4.0
3.0
2.0
1.0
0.0
0.12
0.08
0.04
0 150 300 450 600
Time (s)
0.00
0 150 300
Time (s)
450 600
Future Safety Code Activities
• MAGARC capabilities will be expanded to treat arcs in magnet
busbars by including the magnetic effects of the arc that forms
between the leads of a busbar. As part of an international
collaboration, this new capability will be validated against data that
has recently been obtained from the MOVARC experiment at FzK in
Germany
• We will be working with the ITER IT to provide the necessary quality
assurance documentation required for ITER licensing for the
MELCOR code
• We will continue in support of the licensing process for the US DCLL
TBM to complete Dossier on Safety for this TBM concept
• We will complete the safety assessment of the ARIES Compact
Stellarator and continue in support of the design activity as ARIES
takes on a new design vision
Recent Trends in Radwaste
Management
• Options:
– Disposal in repositories: LLW (WDR < 1) or HLW (WDR > 1)
– Recycling – reuse within nuclear facilities (dose < 3000 Sv/h)
– Clearance – release slightly-radioactive materials to
commercial market if CL < 1.
• Tighter environmental controls and the political difficulty of
building new repositories worldwide may force fusion designers
to promote recycling and clearance, avoiding geological disposal
No radwaste burden on future generation.
• There’s growing international effort in support of this new trend.
• Recycling may not be economically feasible for all fusion
components.
• Recycling of liquids and solids may generate limited amount of
radioactive waste that needs special treatment.
Blanket/Shield/Vacuum Vessel/Magnet/Structure
Volume (10 3 m 3 )
3.5
3.0
2.5
2.0
1.5
1.0
0.5
SiC
0.0
ARIES – I
1990
ARIES Project Committed to
Waste Minimization
FS
(D- 3 He)
III
1991
Tokamaks
II
1992
SiC
V V
IV
1992
RS
1996
ST
1999
SiC
AT
2000
Tokamak waste volume
halved over 10 y study period
FS
Blanket/Shield/Vacuum Vessel/Magnet/Structure
Volume (10 3 m 3 )
8
7
6
5
4
3
2
1
0
FS
UWTOR-M
24 m
1982
V
SPPS
14 m
1994
Stellarators
FS
ARIES-CS
8.25 m
2006
FS
ARIES-CS
7.5 m
2006
Stellarator waste volume
more than halved over
25 y study period
U.S. Clearance Index
Volume (10 3 m 3 )
1.0
0.9
0.8
0.6
0.5
0.4
0.3
0.1
0.0
10 12
10 10
10 8
10 6
10 4
10 2
10 0
10 -2
10 0
FW/Blkt/
BW
FW
Vacuum Vessel
10 2
80% of ARIES-CS Active Materials can be
Cleared in < 100 y after Decommissioning
Inter-Coil Structure
Steel of Bldg
Concrete of Bldg
Limit
Shld/
Mnfld
10 4
1d 1y
10 6
Time (s)
Cryostat
Not compacted, no replacements
Fully compacted with replacements
VV
Magnets &
Structure Cryostat
10 8
100y
10 10
Blanket
Shield
Vacuum
Vessel
Cryostat
2 m Bioshield
IAEA Clearance Index
Magnet
Manifolds
10 12
10 10
10 8
10 6
10 4
10 2
10 0
10 -2
10 0
FW
Vacuum Vessel
Inter-Coil Structure
Cryostat
Steel of Bldg
Concrete of Bldg
Clear
Limit
10 2
Recycle or
Dispose of
10 4
1d
10 6
Time (s)
1y
10 8
100y
10 10
Recycle or
Dispose of
B/S/VV/M
(20%)
Clear
Magnet w/o Nb Sn,
3
Cryostat & Bioshield
(80%)
Blanket
Shield
Vacuum
Vessel
Cryostat
All ARIES-CS Components can be
Recycled in 1-2 yr Using Advanced and
Conventional Equipment
Magnet
Manifolds
Recycling Dose Rate (Sv/h)
10 6
10 4
10 2
10 0
10 -2
10 -4
10 -6
10 -8
10 0
Shield
Inter-Coil Structure
Conc. of Bldg-I
Steel of Bldg-I
Time (s)
FS-based components:
– 54Mn (from Fe) is main contributor to dose.
– Store components for few years before recycling.
– After several life-cycles, advanced RH equipments may be needed to handle shield,
manifolds, and VV.
SiC-based components:
– 58,60 Co, 54 Mn, and 65 Zn contributors originate from impurities.
– Strict impurity control may allow hands-on recycling.
VV
10 2
10 4
FW
SiC
1d
Cryostat
10 6
1y
10 8
10 10
Advanced
RH RH Limit Limit
Conservative
RH RH Limit Limit
Hands-on Hands-on
Limit Limit
MELCOR Code Applied to ARIES Compact
Stellarator Reactor Safety Assessment
• ARIES-CS has Dual Cooled Liquid
Lead Lithium (DCLL) Blanket
• This blanket concept employees
reduced activation ferritic steel
(RAFS) for the structure, cooled by
8 MPa helium, and a self-cooled
breeding zone cooled by flowing
PbLi.
• Helium pressurization accidents will
be a concern for this concept, with
the reactor cryostat serving as
secondary radioactivity and
pressure confinement
• Because stellarator plasmas do not
disrupt, the beyond design basis
bypass accident (BDBA) of concern
may be a heat exchanger tube
breach in the PbLi/Brayton cycle
system, leaking 10 MPa helium into
the blanket causing a blanket break
and pressurization of the vacuum
vessel
Pressure (MPa)
0.25
0.20
0.15
0.10
0.05
Vacuum Vessel
Cryostat
0.00
3600 3700 3800
Time (s)
3900 4000
Initial results for in-vessel FW helium LOCA
MELCOR Code Applied to Advance Power
Extraction (APEX) Reactor Safety Assessment
• APEX studied an advanced Ferritic
Steel FLiBe breeder Reactor
Concept
• Evaluated a confinement bypass
accident
– Total loss of site power that
induces a beyond design basis
plasma disruption the resulting
in the loss of confinement
through a heating or diagnostic
duct to an adjoining room
• Source terms included structure
activation products by oxidation,
tungsten dust from the divertor
erosion, FLiBe activation products by
evaporation, and structure tritium by
diffusion
• Conservative weather conditions and
stack release assumed
Loss-of-Vacuum Accident
VV duct isolation valves fail
and reentrant flow is
established in duct
T 2 released from
FW forms HTO
in air humidity
Aerosols form from FW oxidation
and molten salt LOCA
Reentrant flow established in
HVAC duct transports
aerosols to the environment
Aerosols transported
to and deposition in
non-nuclear room
Temperature (C)
1200
1000
800
600
MELCOR Code Applied to Advance Power
Extraction (APEX) Reactor Safety Assessment
Temperature Response
LOCA without VV cooling
LOCA with VV cooling
LOFA with VV cooling
LOFA without
VV cooling
400
0 5 10 15
Time (d)
Dose (mSv)
0.5
0.4
0.3
0.2
0.1
Site boundary dose
Flibe
0.0
0 1 2 3 4 5 6 7
Time (d)
Tritium
AFS
Major contributors AFS dose are Mn-54, Ca-45, and Ti-45,plant
isolation must occur within four weeks to stay below the 10 mSv limit.
MELCOR Code Applied to US Test Blanket
Module Safety Assessment
• Evaluate consequences to ITER
from accidents in the proposed
US DCLL Test Blanket module
(TBM)
• To date three accident
scenarios have been
investigated:
– In-vessel TBM coolant leaks
– In-TBM breeding zone
coolant leaks
– Ex-vessel TBM cooling
system leaks
• No significant impacts on ITER
safety have been identified, but
assessment is still ongoing
• All FS structures
are He-cooled by
8 MPa
• PbLi self-cooled
flows in poloidal
direction
Internal
PbLi flow
He out
He in
PbLi
concentric
inlet/outlet
pipe
S&E considerations are critical
for the success of IFE
• IFE has both radiological and toxicological hazards:
– Tritium fuel, activated structural material, activated dust, activated
coolants or coolant impurities, and activated gases
– Chemically toxic materials (i.e.: Hg, Pb)
• Energy sources that can mobilize these hazardous materials include:
– chemical energy, decay heat, pressure energy, electrical energy and
radiation
• In the US, current IFE S&E activities, are focused on a few programs:
– The High Average Laser Program (HAPL)
– The Z-IFE Program
– The National Ignition Facility (NIF): not really an IFE program but
closely linked to the future of IFE research
In the recent years there has
been great progress in IFE S&E
• In order to maximize the S&E advantages of IFE, accident
consequences must be addressed realistically
• In early studies, safety analysis tools were not very refined which
often resulted in overly conservative safety analyses and safetyimportant
design details were not available to incorporate into the
safety assessment
• We have adopted and adapted computer codes traditionally used
by MFE, and integrated them in a set of state-of-the-art
codes/libraries for IFE safety analyses
• These tools have provided the first self-consistent analysis to
understand the integrated behavior of an IFE chamber under
accident conditions
• This methodology was applied to various IFE designs and a target
fabrication facility, demonstrating that the implementation of the
Fusion Safety Standards in IFE power plant designs was achievable
HYLIFE-II
SOMBRERO
S&E activities in support of the
HAPL Program
• We have completed preliminary S&E assessment for the HAPL design
• Dominant issue in accident scenario with Li chemical reactions is mobilization of
tritium and activated structural materials
Schematic of HAPL chamber using
self-cooled liquid Li blanket
Temperature (K)
MELCOR predicted temperature
evolution during Li fire
1.2E+03
1.0E+03
8.0E+02
6.0E+02
4.0E+02
2.0E+02
0.0E+00
FW
2nd wall
Back wall
Shielding
Building
0.0E+00 2.0E+05 4.0E+05 6.0E+05 8.0E+05 1.0E+06
Time (s)
1
H
3
Li
11
Na
19
K
37
Rb
55
Cs
Z-IFE RTL waste disposal rating
study: Goal is WDR < 1
4
Be
12
Mg
20
Ca
38
Sr
56
Ba
21
Sc
39
Y
57
La
22
Ti
40
Zr
72
Hf
58
Ce
23
V
41
Nb
73
Ta
59
Pr
24
Cr
42
Mo
74
W
60
Nd
WDR > 10
1 < WDR < 10
0.1 < WDR < 1
0.01 < WDR < 0.1
WDR < 0.01
Not studied
25
Mn
43
Tc
75
Re
61
Pm
26
Fe
44
Ru
76
Os
62
Sm
27
Co
45
Rh
77
Ir
63
Eu
28
Ni
46
Pd
78
Pt
64
Gd
00
Ex
29
Cu
47
Ag
79
Au
65
Tb
30
Zn
48
Cd
80
Hg
66
Dy
5
B
13
Al
31
Ga
49
In
81
Tl
67
Ho
6
C
14
Si
32
Ge
50
Sn
82
Pb
68
Er
Weekly recycling rating
Daily recycling rating
7
N
15
P
33
As
51
Sb
83
Bi
69
Tm
8
O
16
S
34
Se
52
Te
84
Po
70
Yb
9
F
17
Cl
35
Br
53
I
71
Lu
2
He
10
Ne
18
Ar
36
Kr
54
Xe
(
e
t
a
r
e
s
o
d
t
c
a
t
n
o
C
10 7
10 6
10 5
10 4
10 3
10 2
10 1
10 0
10 -1
10 4
Example of impact of safety
analyses in NIF
• Material selection is an important part of controlling worker doses
• Al-5083 was preferable to stainless steels at the “decay times” of greatest
interest for worker doses and decomissioning
• Addition of boron (0.14% by weight) to gunite shielding reduced t=5 day dose
rate by 3
5 days
10 5
10 6
Time (s)
Al-5083
SS-316
10 7
10 8
3 years
(
e
t
a
r
e
s
o
d
t
c
a
t
n
o
C
10 4
10 3
10 2
10 1
10 0
10 -1
10 -2
10 4
5 days
10 5
Concrete with no Boron
Concrete with Boron
10 6
Time (s)
10 7
3 years
10 8
Summary
• US S&E research continues to help improve fusion facility design in
terms of accident safety, worker safety, and waste disposal.
• The R&D underway and currently planned in the areas of dust and
tritium source terms will answer important questions for ITER and
future machines.
• Regulatory approval of ITER and the associated verification and
validation activities for our fusion safety codes and risk and
reliability methods will provide greater confidence in application of
these tools to evaluate public and worker safety of future fusion
facility designs.
• The resurgence of nuclear fission reactor construction activities
worldwide will cause increased attention to waste management
issues associated with nuclear power which in turn should help
fusion as it develops a long term waste management strategy
consistent with on-going US regulation.
• Safe and environmentally sound operation of both ITER and NIF
will be important public demonstrations of the S&E potential of
fusion.
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