Transcript INL experimental and analytical capabilities for DCLL

INL experimental and
analytical capabilities for the
DCLL concept
www.inl.gov
2nd EU-US DCLL Workshop
Brad J. Merrill, Fusion Safety Program
November 14-15th, 2014
University of California, Los Angeles
Presentation Outline
• Describe ongoing FSP experimental capabilities in tritium and
materials safety research (in particular PbLi) through international
collaborations
• Discuss computer codes developed by the FSP for accident
analyses
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Idaho National Laboratory…
• 890 square miles
• 111 miles of electrical transmission and
distribution lines
• 579 buildings
• 177 miles of paved roads
• 14 miles of railroad lines
INL-Site
Idaho Cleanup Project
Naval Reactors
Facility
…the National Nuclear Laboratory
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FSP’s STAR Lab located at the INL Site…
• The Safety and Tritium Applied Research
(STAR) facility is an Office of Science
facility commissioned in November 2001
• STAR Facility – DOE less-than Hazard
Category 3 Radiological Facility
– Allows handling of radioisotopes at
low radiation levels (e.g., W, Ni, Mo,
T2) typically specified as either:
 Contact dose (< 1.5 mSv/hr at 30
cm, normally < 100 µSv/hr),
 Annual worker dose (< 7 mSv/yr)
 Total inventory (1.5 g T2 or
5.6x105 GBq)
STAR
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Tritium Solubility testing in Lead-Lithium
Eutectic (LLE) under US-JA TITAN Collaboration
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Tritium Gas Absorption Permeation (TGAP)
Experiment
Unique capabilities
• Designed to measure transport properties (e.g. diffusivity, solubility, and permeability) of
tritium at realistic blanket conditions (e.g. low tritium partial pressure < 100 Pa)
• Using tritium should lead to accurate measurements since it is easily detected
• Uniform temperature (+/- 10 C) within the test section utilizing 12” tube furnace
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Tritium Concentration (Ci/m3)
Blank Tests Results - Qualify System w/o LLE
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TMAP model of TGAP
2.4 Pa
10-1
10-2
0.15 Pa
0.01 Pa
0.001 Pa
10-3
10-4
10-5
•
•
•
•
TMAP - Symbols
0
1000
2000
3000
4000
5000
Time (s)
Permeation results with empty test section for α-Fe membrane
can be matched well with TMAP’s surface kinetics model
(requires a multiplier of 0.1 on surface coefficients kd & kr)
Simple analytical expressions for permeation flux do not
accurately capture the response of the entire system
Low pressure (0.001 Pa) points are problematic – lower limits
of experimental resolution
The background concentration of hydrogen must be
quantifiable
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Role of H2 at Surface of Membrane in TLLE
Results for Blank Tests
Secondary H (100 Pa) changes T
2
permeation from surface to diffusion
limited permeation & reduces
membrane T concentration
2.4 Pa
H
Hydrogen
No hydrogen
H2
H2
HT
HT
0.15 Pa
T2
T
T2
• Primary H surface concentration >
40 x T, steepens T gradient
(speeds equilibrium in membrane)
• Secondary side H carries T off
surface back into secondary
(reduces T flux)
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Initial LLE Transport Properties Determined by
Analysis
Tritium permeation through
•
Due to the very low tritium pressures,
moderate sweep gas flow rates and
“comparatively” large TGAP test volumes,
data analysis must be accomplished though
computer code matching (Alice pointed out
yesterday that this experiment tests liquid
metal/solid metal interface assumption)
(1 mm) α-Fe + (6 mm) LLE
Modeling results (Preliminary analysis)
•
•
Tritium diffusivity in LLE:
• A factor of 2-3 higher value needed to fit
exp. data than found in literature
Tritium solubility in LLE:
• Similar to literature data (?)
TGAP is being reconfigured so that the
background hydrogen concentration will
be known during experiments
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TGAP Tritium Permeation Campaign for Tungsten
• Permeation test section developed under the Japan/US PHENIX collaboration by
Shizuoka University capable of testing 6 mm diameter tungsten disks up to
temperatures of 1000 C and low tritium partial pressures (< 100 Pa)
• With similar test sections under the NFRI-UCLA-INL collaboration, TGAP will be used to
study tritium permeation through RAFM at low pressures and release from functional
materials
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FSP Safety Codes Used in ITER and US DCLL
TBM Accident Analyses
• MELCOR for fusion - a fully integrated, engineering level thermalhydraulics computer code that models the progression of accidents in
fission and now fusion power plants, including a spectrum of accident
phenomena such as reactor cooling system and containment fluid flow,
heat transfer, and aerosol transport (various fluids, including PbLi, can
be modeled),
• ATHENA/RELAP – a multi-fluids thermal-hydraulics code developed
for design and accident analysis of cooling systems fusion reactor
systems,
• TMAP - a tritium migration code that treats multi-specie surface
absorption and diffusion in composite materials with dislocation traps,
plus the movement of these species from room to room by air flow
within a given facility.
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MELCOR:
Integrated Physics and Coupled
• During post ITER EDA , a multi-fluid version of MELCOR
1.8.5 was developed to assess the safety of Advanced
Phenomena
Fusion Reactor Concepts.
• This version of MELCOR uses the ATHENA/RELAP fluid
Fusion ITER Modifications
property subroutines which include 10 fluids
• EOS modifications
for water freezing & ice layer
– SPARC
FliBe, Hydrogen, Helium, Lithium, NaK, Nitrogen,
formation
Potassium,
BWRSAR PbLi, Sodium, Water
CORSOR
• C, Be, W oxidation (INL correlations)
– Flow choking changed to Fauske model
• Aerosol transport
module modifications for gas
MELCOR
–
Lithium
fire model
similar to LITFIRE
mixtures, turbulentpool
& inertial
deposition
CORCON
• Modifications
Multi-fluids
• Enclosure
thermal to
radiation
heat MELCOR
transport 1.8.5 for NGNP
MAEROS
made
• Flow application
Boiling heatwere
transfer
HECTR
VANESA
– Gaseous
binary diffusion was
to the
• HTO transport
model
• added
In 2008,
themass
NRC and
requested that all
energy
conservation
equations
Used in CONTAIN
ITER Safety Documents fusion modifications be included in
Graphite
andbeing
combustion
models 2.x F95 version
the MELCOR
TRENDS
• NSSR –
1&2,
GSSR, oxidation
RPrS (after
pedigreed
– Dust
re-suspension
version available. A
and placed
in INL
Software QA system• asPreliminary
a QL1
MELPROG
paperAPEX
at theEvolve
22nd International
• Used
in advanced reactor design studies:
safety
code)
VICTORIA
Conference
on Nuclear
Engineering
(Li/W), ARIES-AT & CS (Dual Coolant
Lead Lithium
–
IFCI
sodium
modeling by SNL-NM
DCLL), US ITER DCLL Test Blanket on
Module
(TBM)
DEBRIS
(2014)
Pool F.P. Scrubbing
BWR Plant Code
Fuel F.P. Release
Whole Plant Analysis
Molten Core/Concrete
Aerosol Mechnaics
Hydrogen Burn
MCCI Fission Products
Plant Containment Analysis
Iodine Chemistry
Detailed Plant Analysis
Fission Product Chemistry
Fuel/Coolant Interactions
Degraded Core
MELCOR Versions
1.7.1
1.8.1
1.8.3
1.8.5
02
20
20
00
1.8.6 (2005)
98
19
19
96
1.8.4
94
19
92
1.8.2
19
90
19
88
1.8.0
19
86
19
84
19
82
19
Ac TM
cid I-2
en
t
19
80
MACCS
2.x (2008)
MELCOR Code Applied to US Test Blanket
Module (TBM) Safety Assessment for ITER
• Evaluate consequences to ITER
from accidents in the proposed US
dual coolant lithium lead (DCLL)
TBM (recently for EU HCLL)
• To date several accident scenarios
have been investigated:
– In-vessel TBM coolant leaks
– In-TBM breeding zone coolant
leaks
– Ex-vessel TBM cooling system
LOCA, LOFA, LOHS
• No significant impacts on ITER
safety have been identified; a
preliminary safety report has been
published, INL/EXT-10-18169, and
transmitted to ITER IO for TBM
concept licensing
Internal PbLi flow
• All ferritic steel
structures are
He-cooled at 8
MPa, 350-410°C
• PbLi self-cooled
flows in poloidal
direction at 2
MPa, 360-470°C
He out
He in
PbLi concentric
inlet/outlet pipe
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Helium Leak into Inter-space
Pressure/Release Results
• Gallery overpressure reaches ~ 700 Pa
• Of ~20.6 g of dust, ~0.4 g of tritium as HTO, and ~0.03 g of ITER activated corrosion
products (ACP) transported into the gallery 11.7 g of dust, 0.03 g of tritium, as HTO,
and 0.01 g of ACP are predicted to be released to the environment.
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101.0
Cryostat room
101
Mass released (g)
Pressure (kPa)
Dust
100.5
Gallery
100.0
100
10-1
HTO
10-2
ACP
99.5
0
100
200
300
Time (s)
400
500
10-3
0
1000
2000
Time (s)
3000
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PbLi Leak into Inter-space
Pool Temperature/Mobilization Results
• Pool surface freezing in 23 hr, entire pool freezes by 130 hr, and by 250 hr the
temperature drops to 110 C.
• Pb-210 and Hg-203 are mobilized (< 3% 1.8 Ci of the Pb-210 and < 10% of the
36 Ci of Hg-203) by diffusion of these isotopes in the pool and release from the
surface by evaporation. Once the PbLi freezes, this diffusion and evaporation
process should drop dramatically.
0.10
Fraction mobilized
Temperature (C)
800
600
PbLi pool
400
200 Port plug
0
Inter-space wall
Hg- 203
0.08
0.06
Po- 210
0.04
0.02
Bioshield
0
100
200
Time (hr)
300
400
0.00
0
10
20
Time (hr)
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Future Directions in FSE Research
• Strategy based on FES guidance and 2013 FES Peer Review Comments
– Materials Research: Fusion materials, including tungsten irradiated, will be
studied at high temperature and heat flux to measure tritium retention and
permeation. Dust explosion measurements for fusion materials will
continue in support of licensing and computer code development activities.
New material diagnostics.
– Code Development: for the near term, a newer version of MELCOR for
ITER will to be completed that includes tritium transport and dust explosion
models. Long- term: Multi-dimensional safety code capabilities needs to be
developed that take advantage of parallel computing (example RELAP 7)
– Risk and Licensing: FSP’s evolving failure rate database will be expanded
to include maintenance data from existing tokamaks. Risk-informed safety
analysis methods (example RISMC Toolkit) will be studied for application to
an FNSF. Continue ASME codes and standards and licensing framework
development.
– Collaborations: Participation
existing Laboratory
collaborations to leverage other
The NationalinNuclear
institution's capabilities and reduce duplication of effort. STAR will move
towards being more effective FES User Facility.
Our website is at:
https://inlportal.inl.gov/portal/server.pt/community/fusion_safety
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