recent accomplishments and future directions in … · recent accomplishments and future directions...
TRANSCRIPT
Recent Accomplishmentsand Future Directions in the
US Fusion Safety &Environmental Programs
David PettiIAEA TCM on Fusion Safety
Vienna, AustriaJuly 12, 2006
Outline
• STAR
• Dust
• Tritium Material Interactions and Permeation
• Fusion Safety Codes
• Risk
• Waste Management
• MFE Safety Design
• IFE Safety Design
• Summary
Fusion Safety ChemicalReactivity Experiments
Molten Salt Tritium/Chemistry Experiments
Fusion Safety Mobilization Testing
Tritium Uptake in Materials
Home of the STAR Facility at theIdaho National Laboratory
Tritium Plasma Experiment TPE Plasma
STAR
Be sampleafter exposureto ion beam
Mo alloy samples afterexposure to air
Dust/Debris Characterization
TFTR DIII-D C-MOD
Tore SupraNOVA
A National User Facility Providing aValuable Resource for Researchers Across
the Nation and Around the World.
W brushsamples
Tritium InfrastructureSystems
STAR Floorplan Layout
15,000 Ci tritium limit
Segregation of operations
Gloveboxes and hoods
Tritium cleanup system
Once-through room ventilation
2LiF-BeF2 preparation,
purification and testing
D-ion implantation
TPE
Chemical reactivity
Glovebox TCS
Tritium SAS Stack monitor
MS tritium exp
MS corrosion exp
Star Tritium Storage and Assay System
U-bed-2
U-bed-1
Secondary inlet
Primary outlet
SAS manifold with U-bedsSAS manifold and
vacuum pumpsSAS glovebox Setup
STAR Tritium Cleanup System (TCS)
Inlet
Control
Outlet IC
Heat exchanger
Catalyst beds
MS water
Blower
Inlet sample loop
Condenser
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)
Timeline of TPE at LANL and INL
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
April 8, 2002 - Depart LANL
March 2002 - Extraction,loading, and transport
December
2005 - First
plasma testing
(non-tritium)
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
April 10, 2002 - Arrival at INL
STAR Facility
Spring &
Summer 2005
Electrical
Service
re-design
& construction,
experiment
& facility
interface
assemblies
completed
Fall 2005 - Integrated systems testing initiated
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 ontritium trapping
• Certification of these structures for ITER
• Use of tritium in the plasma will enable low levelmeasurements needed for such research
Flibe Tritium Experiment: Design and testing of
tritium handling systems and diagnostics
• Tritium provided in
pressurized vessel
containing D2/T2 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
Ar D2
Flow meter
Flow meter
Flow meter
Gas chromatograph
or QMS
or ionization chamber
HF trap
exhaust
High temperature salt
Flibe
Cap
Ni
T2
Vacuum
pump
Pressure gauge
dip Be if Redox control is successful
Conceptual layout proposed by Fukada et al.
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 generationITER Key Issues: chemical reactivity, radioactivity
content, dust explosions.Science: understand and model underlying formation
mechanisms to estimate inventories expected in fusion
Demonstration
of the filtered
vacuum
collection
technique
ITER Dust Strategy from EDA
Key scientific issuesneeding resolution
0.01
0.1
1
10
100
1 10 100
fully dense Cfully dense MoTFTRDIII-DAlcator-CmodTore SupraASDEX-UpgradeNOVAJETATJ graphite
Spe
cific
Sur
face
Are
a, m
2/g
Mean Volume-Surface Diameter, dMVS
(μm)
M o
C
Specific Surface Area
Dust Size versus Surface Mass Density
0
5
10
15
100 101 102 103 104
LHDASDEX-UpgradeTore SupraCMOD_(98)DIIID_(98)
Me
dia
n P
art
icle
Dia
me
ter
(μm
)
Surface Mass Density (mg/m2)
LHD
ASDEX-Upgrade
Tore Supra
CMODDIII-D
Comparison of Size Distributions
3.32 + 2.942.98 + 2.942.68 + 2.89Tore Supra
0.90 + 1.930.76 + 2.031.12 + 1.90NOVA
8.73 + 2.096.31 + 2.398.59 + 2.67LHD
3.59 + 3.083.69 + 2.812.21 + 2.93ASDEX-Upgrade
--5-20 + (-)TEXTOR
--27 + (-)JET
1.22 + 2.031.53 + 2.801.58 + 2.80Alcator-Cmod
-1.60 + 2.330.88 + 2.63TFTR
0.89 + 2.920.60 + 2.350.66 + 2.82DIII-D
UpperRegions
MiddleRegions
LowerRegionsMachine
CMD (μm) + GSD
Measurements of dust characteristics are an activearea of research
Particle Size Distribution, Specific Surface Area, Surface Mass Density,Composition, Shape and Tritium Content
R&D continues to resolve the issues
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 inRisk Assessment for Fusion
• Work on component failure rate data to support quantitative safetyassessment continues to be very successful.
– Initially, component failure rates were collected from handbooksand applied to fusion.
– Now, through IEA Task 5, we collect fusion facility operatingexperience data from TLK, TPL, and the former TSTA; and tokamakdata 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 assessmentof the ITER facility to identify occupational hazards
– Occupational injury rates have been collected from several UStokamaks and large particle accelerators
– WE-FMEA method was developed to address highly hazardousequipment failures in a fusion facility, such as high energy pipebreaks that have caused worker fatalities in the power industry
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 LayerOxidation 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-of-
vacuum accidents (LOVAs)
1210
10
10
10
10
-
-9
-6
-3
0
Be
rylli
um
oxid
atio
n r
ate
(kg
/m2-s
)
5 10 15 2010,000/T (K)
INL Dust
INL fully dense
INL98 88% dense
INL92
88% dense MELCOR dust layer model( Dust=0.7g/cm3, dp= 20μm)
MAGARC Poloidal Field Coil Development
• This modification include
the electrical
characteristics of the two-
in-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.
Ax
ial directio
n
Radial direction
• The MAGARC code was recently
modified to analyze unmitigated
quench events in ITER poloidal
field (PF) magnets
voltage drops
Time 110s
0.18
0.00
0.360.540.720.25
0.00
0.500.75
1.00
500
01000
1500
Voltage (
V)
Width (m)
Height (m)
Inline arcs
Axial directionRadial direction
Winding pair
MAGARC Poloidal Field Coil Application
• MAGARC code
application to
unmitigated
quench events in
ITER poloidal field
(PF) magnets
0.0
0.2
0.4
0.6
0.8
Que
nch
frac
tion
0 150 300 450 600Time (s)
0
500
1000
1500
2000
Lead
vol
tage
dro
p (V
) 3000
0 150 300 450 600Time (s)
0 150 300 450 600Time (s)
0.0
1.0
2.0
3.0
4.0
Coi
l cur
rent
(kA
)0 150 300 450 600
Time (s)
0
2
4
6
8
Num
ber
of a
rcs Inline
RadialAxial
0 150 300 450 600
Time (s)
0.00
0.04
0.08
0.12
Mel
t vol
ume
(m3 )
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 RadwasteManagement
• 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 tocommercial market if CL < 1.
• Tighter environmental controls and the political difficulty ofbuilding new repositories worldwide may force fusion designersto 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 fusioncomponents.
• Recycling of liquids and solids may generate limited amount ofradioactive waste that needs special treatment.
ARIES Project Committed toWaste Minimization
Tokamak waste volumehalved over 10 y study period Stellarator waste volume
more than halved over25 y study period
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
Bla
nket/
Sh
ield
/Vacu
um
Vessel/M
ag
net/
Str
uctu
re
Vo
lum
e (
10
3 m
3)
ARIES – I1990
III1991
II1992
IV1992
RS1996
ST1999
AT2000
SiCSiC
SiC
V V
FS
FS(D-
3He)
0
1
2
3
4
5
6
7
8
Bla
nket/
Sh
ield
/Vacu
um
Vessel/M
ag
net/
Str
uctu
re
Vo
lum
e (
10
3 m
3)
UWTOR-M24 m1982
SPPS14 m1994
ARIES-CS 8.25 m
2006
V FS
FS
ARIES-CS 7.5 m2006
FSTokamaks Stellarators
80% of ARIES-CS Active Materials can beCleared in < 100 y after Decommissioning
10-2
100
102
104
106
108
1010
1012
100
102
104
106
108
1010
U.S
. C
leara
nce I
nd
ex
Time (s)
1d 1y
Inter-Coil Structure
Limit 100y
FW
Vacuum Vessel
Cryostat
Steel of Bldg
Concrete of Bldg
10-2
100
102
104
106
108
1010
1012
100
102
104
106
108
1010
IA
EA
Cle
ara
nce I
nd
ex
Time (s)
1d 1y
Inter-Coil Structure
Limit
100y
FW
Vacuum Vessel
Cryostat
Steel of Bldg
Concrete of Bldg
0.0
0.1
0.3
0.4
0.5
0.6
0.8
0.9
1.0
Not compacted, no replacementsFully compacted with replacements
Vo
lum
e (
10
3 m
3)
FW/Blkt/BW
Shld/Mnfld
VV Magnets &Structure
Cryostat
Recycle orDispose ofB/S/VV/M
(20%)
Clear Magnet w/o Nb
3Sn,
Cryostat & Bioshield(80%)
2 m Bioshield
Cryostat
Blanket
Manifolds
Shield
VacuumVessel
Magnet
Recycle orDispose of
Clear
All ARIES-CS Components can beRecycled in 1-2 yr Using Advanced and
Conventional Equipment
10-8
10-6
10-4
10-2
100
102
104
106
100
102
104
106
108
1010
Recyclin
g D
ose R
ate
(S
v/h
)
Time (s)
Advanced RH Limit
Conservative RH Limit
Hands-onLimit1y1d
FW
SiCShield
Inter-Coil Structure
Steel of Bldg-IConc. of Bldg-I
VV
Cryostat
Advanced RH Limit
Conservative RH Limit
Hands-onLimit
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:
– 58,60Co, 54Mn, and 65Zn contributors originate from impurities.
– Strict impurity control may allow hands-on recycling.
Cryostat
Blanket
Manifolds
Shield
VacuumVessel
Magnet
MELCOR Code Applied to ARIES CompactStellarator Reactor Safety Assessment
• ARIES-CS has Dual Cooled LiquidLead Lithium (DCLL) Blanket
• This blanket concept employeesreduced activation ferritic steel(RAFS) for the structure, cooled by8 MPa helium, and a self-cooledbreeding zone cooled by flowingPbLi.
• Helium pressurization accidents willbe a concern for this concept, withthe reactor cryostat serving assecondary radioactivity andpressure confinement
• Because stellarator plasmas do notdisrupt, the beyond design basisbypass accident (BDBA) of concernmay be a heat exchanger tubebreach in the PbLi/Brayton cyclesystem, leaking 10 MPa helium intothe blanket causing a blanket breakand pressurization of the vacuumvessel
3600 3700 3800 3900 4000Time (s)
0.00
0.05
0.10
0.15
0.20
0.25
Pre
ssu
re (
MP
a)
Vacuum VesselCryostat
Initial results for in-vessel FW helium LOCA
MELCOR Code Applied to Advance PowerExtraction (APEX) Reactor Safety Assessment
• APEX studied an advanced FerriticSteel FLiBe breeder ReactorConcept
• Evaluated a confinement bypassaccident
– Total loss of site power thatinduces a beyond design basisplasma disruption the resultingin the loss of confinementthrough a heating or diagnosticduct to an adjoining room
• Source terms included structureactivation products by oxidation,tungsten dust from the divertorerosion, FLiBe activation products byevaporation, and structure tritium bydiffusion
• Conservative weather conditions andstack release assumed
Aerosols transported
to and deposition in
non-nuclear room
VV duct isolation valves fail
and reentrant flow is
established in duct
Reentrant flow established in
HVAC duct transports
aerosols to the environment
T2 released from
FW forms HTO
in air humidity
Aerosols form from FW oxidation
and molten salt LOCA
Loss-of-Vacuum Accident
MELCOR Code Applied to Advance PowerExtraction (APEX) Reactor Safety Assessment
Temperature Response Site boundary dose
0 5 10 15
Time (d)
400
600
800
1000
1200
Tem
pera
ture
(C
)
LOCA without VV cooling
LOCA with VV cooling
LOFA without VV cooling
LOFA with VV cooling
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.
0 1 2 3 4 5 6 7
Time (d)
Dos
e (m
Sv)
0.0
0.1
0.2
0.3
0.4
0.5
Flibe
AFS
Tritium
MELCOR Code Applied to US Test BlanketModule Safety Assessment
• All FS structures
are He-cooled by
8 MPa
• PbLi self-cooled
flows in poloidal
direction
He out
He in
Internal
PbLi flow
PbLi
concentric
inlet/outlet
pipe
• Evaluate consequences to ITERfrom accidents in the proposedUS DCLL Test Blanket module(TBM)
• To date three accidentscenarios have beeninvestigated:
– In-vessel TBM coolant leaks
– In-TBM breeding zonecoolant leaks
– Ex-vessel TBM coolingsystem leaks
• No significant impacts on ITERsafety have been identified, butassessment is still ongoing
S&E considerations are criticalfor the success of IFE
• IFE has both radiological and toxicological hazards:
– Tritium fuel, activated structural material, activated dust, activatedcoolants 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 andradiation
• 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 butclosely linked to the future of IFE research
In the recent years there hasbeen great progress in IFE S&E
• In order to maximize the S&E advantages of IFE, accidentconsequences must be addressed realistically
• In early studies, safety analysis tools were not very refined whichoften resulted in overly conservative safety analyses and safety-important design details were not available to incorporate into thesafety assessment
• We have adopted and adapted computer codes traditionally usedby MFE, and integrated them in a set of state-of-the-artcodes/libraries for IFE safety analyses
• These tools have provided the first self-consistent analysis tounderstand the integrated behavior of an IFE chamber underaccident conditions
• This methodology was applied to various IFE designs and a targetfabrication facility, demonstrating that the implementation of theFusion Safety Standards in IFE power plant designs was achievable
HYLIFE-II
SOMBRERO
S&E activities in support of theHAPL Program
• We have completed preliminary S&E assessment for the HAPL design
• Dominant issue in accident scenario with Li chemical reactions is mobilization oftritium and activated structural materials
Schematic of HAPL chamber using
self-cooled liquid Li blanket
0.0E+00
2.0E+02
4.0E+02
6.0E+02
8.0E+02
1.0E+03
1.2E+03
0.0E+00 2.0E+05 4.0E+05 6.0E+05 8.0E+05 1.0E+06
Time (s)
Te
mp
era
ture
(K
)
FW
2nd wall
Back wall
Shielding
Building
MELCOR predicted temperature
evolution during Li fire
Z-IFE RTL waste disposal ratingstudy: Goal is WDR < 1
1
H
2
He
3
Li
4
Be
5
B
7
N
6
C
8
O
9
F
10
Ne
11
Na
12
Mg
13
Al
14
Si
15
P
16
S
17
Cl
18
Ar
19
K
37
Rb
55
Cs
20
Ca
31
Ga
32
Ge
33
As
34
Se
35
Br
36
Kr
25
Mn
26
Fe
27
Co
28
Ni
29
Cu
30
Zn
21
Sc
22
Ti
23
V
24
Cr
38
Sr
49
In
50
Sn
51
Sb
52
Te
53
I
54
Xe
43
Tc
44
Ru
45
Rh
46
Pd
47
Ag
48
Cd
39
Y
40
Zr
41
Nb
42
Mo
56
Ba
81
Tl
82
Pb
83
Bi
84
Po
75
Re
76
Os
77
Ir
78
Pt
79
Au
80
Hg
57
La
72
Hf
73
Ta
74
W
67
Ho
68
Er
69
Tm
70
Yb
61
Pm
62
Sm
63
Eu
64
Gd
65
Tb
66
Dy
58
Ce
59
Pr
60
Nd
71
Lu
WDR > 10
1 < WDR < 10
0.1 < WDR < 1
0.01 < WDR < 0.1
WDR < 0.01
Not studied
00
Ex
Weekly recycling rating
Daily recycling rating
10-2
10-1
100
101
102
103
104
104
105
106
107
108
Concrete with no BoronConcrete with Boron
Contact dose rate (
Time (s)
10-1
100
101
102
103
104
105
106
107
104
105
106
107
108
Al-5083SS-316
Contact dose rate (
Time (s)
Example of impact of safetyanalyses in NIF
• Material selection is an important part of controlling worker doses
• Al-5083 was preferable to stainless steels at the “decay times” of greatestinterest for worker doses and decomissioning
• Addition of boron (0.14% by weight) to gunite shielding reduced t=5 day doserate by 3
5 days
3 years5 days
3 years
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 andtritium source terms will answer important questions for ITER andfuture machines.
• Regulatory approval of ITER and the associated verification andvalidation activities for our fusion safety codes and risk andreliability methods will provide greater confidence in application ofthese tools to evaluate public and worker safety of future fusionfacility designs.
• The resurgence of nuclear fission reactor construction activitiesworldwide will cause increased attention to waste managementissues associated with nuclear power which in turn should helpfusion as it develops a long term waste management strategyconsistent with on-going US regulation.
• Safe and environmentally sound operation of both ITER and NIFwill be important public demonstrations of the S&E potential offusion.