Aerospace Nuclear Science & Technology Division - Safety...
Transcript of Aerospace Nuclear Science & Technology Division - Safety...
ORNL is managed by UT-Battelle
for the US Department of Energy
C.J. Hurt
University of Tennessee Nuclear Engineering Department (This material is based upon work supported under a Department of Energy Nuclear Energy University Programs Graduate Fellowship.)
James D. Freels, Frederick P. Griffin, Randy W. Hobbs, David Chandler ORNL, Research Reactors Division
Robert M. Wham ORNL, Nuclear Science & Engineering
Safety Analysis
Models for 238
Pu
Production at HFIR
Presented at
Nuclear and Emerging
Technologies for Space 2015
February 23-26, 2015 Albuquerque, NM
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Presentation Outline
• Experiment Safety Overview
• Model Physics
– Contact/Gap Conductance
– Irradiation Behavior
• Model Inputs
• Steady-State Model
• Transient Model
• Model Outputs
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238Pu Supply Requires Integration Across
Existing DOE Facilities
ProcessingORNL
Powder
PlannedPlutonium Fuel Production
IrradiationATR/HFIR
TargetFabrication
LANL
StoredNeptunium
• 238Pu is the fuel source for RTGs that power NASA deep space missions
• The final supply chain requires integration from Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL)
• This presentation discusses the safety analyses required for irradiation of preliminary targets at the High Flux Isotope Reactor (HFIR) at ORNL
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Target Qualification
• Four phase test program, each phase informing the next phase design
• Post-irradiation examination (PIE) results from each phase serve as a hold point for the following irradiations
pellet dimensional changes pellet clad interaction
fission gas release % 236Pu production
heat generation rates product yields
Post-Irradiation Characteristics
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Experiment Safety Review
• Target qualification at HFIR requires a safety review that assures target cooling in off-normal and nominal reactor operating conditions.
• Safety analyses using software that meets DOE requirements for SQA: COMSOL Multiphysics, RELAP5, ANSYS, MCNP5, ORIGEN-S and VESTA.
• Steady-State Analysis in COMSOL
• 50% reduced flow
• 130% overpower Bounding safety condition
• Transient Analysis in RELAP5:
• Small break loss of coolant
• Loss of offsite power
Off-Normal Safety Review Transport and
Depletion Analysis
Steady-State Thermal-Structure Analysis
Transient Heat Transfer
Analysis
Experiment Safety Approval Documentation
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Contact/Gap Conductance Physics
• Peak pellet temperatures are driven by heat transfer in the radial gap between the pellets and cladding
• Gap heat transfer (h) is strongly dependent on the pellet dimensional changes and fission gas release
Temperature
increase due to gap
heat transfer
between the pellet
and Al cladding
ℎ = ℎ𝑠 + ℎ𝑔 + ℎ𝑟
hs = constriction
conductance or solid spot
conductance
hg = gas gap conductance
hr = radiative conductance
δ
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Contact and Gap Conductance
• Depends on:
– Mean (harmonic) thermal conductivity of mating surfaces, km , Temperature
– Contact pressure (P), surface hardness (H) and elastic modulus (E)
– Function of effective surface roughness σ =√(𝜎12 + 𝜎2
2) and slopes of asperities (m = tanθ)
– Total apparent area of contact, Mean junction temperature
• Predicted by assuming a mode of ‘microscopic deformation’
– Plastic Deformation:
– Elastic Deformation (conservatively assumed):
– Alternatively, a ‘plasticity index’ may be used
• Gas gap conductance
• Portion of thermal contact conductance due to “conduction only” through “real” contact spots
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Gas Gap Conductance
• Thermal jump distance
– α = accommodation coefficient, λ = mean free
path
• Large changes in gap heat transfer solution can be caused by small changes in parameters:
– Accuracy of the model equations, inputs, solver, etc.
– Up-to-date pellet material properties (thermal expansion, thermal conductivity, etc.)
– The initial radial fabrication gap between pellet/cladding
– The burnup-dependent pellet radial shrinkage/swelling
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-0.10
0.00
0.10
0.20
Vo
lum
e F
ract
ion
Accumulated Fission Density/Burnup
Void Volume
Fission Product Swelling
Total Swelling
Dimensional Irradiation Behavior
• Swelling (+ΔV) increases linearly with the accumulation of fission products.
• Densification (-ΔV) occurs early due to the irradiation-induced sintering of the initial pellet porosity (P0).
• Net behavior is then an initial shrinkage of the pellet gradually overcome by swelling as the irradiation period progresses
• Other parameters (temperature, pressure, pore sizes) can be considered negligible for a simple model dependent on fuel burn up (BU) and constant coefficients (α,β)
∆𝑉𝑠𝑤𝑒𝑙𝑙.= 𝛼 ∗ 𝐵𝑈
∆𝑉𝑑𝑒𝑛𝑠.= 𝑃0 𝑒−𝛽∗𝐵𝑈 − 1
∆𝑉𝑡𝑜𝑡𝑎𝑙= ∆𝑉𝑑𝑒𝑛𝑠. + ∆𝑉𝑠𝑤𝑒𝑙𝑙.
Region of pellet densification/swelling
observed in post irradiation examination results
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Model Inputs
• PIE results
• Shrinkage/Swelling vs. burn-up for radial and height dimensions
• Fission gas release fractions
• Temperature- and composition-dependent material property data for the NpO2/Al pellet (and other materials):
• Thermal expansion coefficient
• Thermal conductivity
• Stress/strain curves (i.e. elastic modulus, yield strength)
• Density, Poisson’s ratio, etc.
• Target Design Drawings and Information
• Input heat generation rates, burnup and fission gas production from neutronics calculations
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Example Input: Shrinkage/Swelling PIE Curve
PIE data has progressed from reduced length pellet (top left), partial-target (top right) and full-target (bottom right) results for more detailed input to the model
-5
-4
-3
-2
-1
0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5
Dia
mete
r C
han
ge A
vera
ged
by I
nit
al
Po
rosit
y
Ratio to First Cycle Median Burnup
Partially Loaded Target PIE Results
Fully Loaded Target PIE Results
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Example Input: Pellet Property Data
0.000015
0.000017
0.000019
0.000021
0.000023
0.000025
400 450 500 550 600
Perc
en
tt T
herm
al E
xp
an
sio
n (
%)
Temperature (degC)
800 C
900c
1200c
Thermal expansion data for pellets sintered at different temperatures
Stress/strain curves for pellets evaluated at different
temperatures
0
2
4
6
8
10
12
14
0 5 10 15 20
Str
ess (
ksi)
Strain (%)
RT 200 °C
300 °C 400 °C
500 °C
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Example Input: Neutronics HGR for VXF-15 Cycle 2
0
50
100
150
200
250
300
350
400
450
-25 -15 -5 5 15 25
Mass
-Sp
ecif
ic H
eat
Gen
. R
ate
(w
/g)
Reactor Axial Position (cm)
EOC-1 VXF-15
EOC-1 VXF-3
Day 5
Day 10
Day 15
Day 20
Day 26
EOC-1 VXF-15 Fit
EOC-1 VXF-3 Fit
Day 5 Fit
Day 10 Fit
Day 15 Fit
Day 20 Fit
Day 26 Fit
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Steady-State Analysis in COMSOL
• Steady state coupled thermal-structure analysis of an individual Pu-238 capsule or single hot pin
• Primary codes: COMSOL v4.2a, v4.3 and v5.0
• Independent Reviews: Independent Reviews: ANSYS and hand calculations using Excel spreadsheets
• Dimensions: 3-D, 2-D axisymmetric
• Calculations:
• C-HFIR-2012-017, Rev. 1, Bare-Pellet Test Capsules
• C-HFIR-2012-036, Rev. 0, Reduced-Length Pellet Capsules
• C-HFIR-2013-007, Rev. 0, Partially Loaded Targets
• C-HFIR-2013-029, Rev. 4, Fully Loaded Targets
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single bare pellet, 2nd irradiation cycle, COMSOL 4.2a, 3D, ¼ pie slice
reduced-length bare pellet, 2nd irradiation cycle, COMSOL 4.2a, 3D, ¼ pie slice
partially-loaded (8 pellets) prototype production target, 2 irradiation cycles, COMSOL 4.3
fully-loaded prototype production target (52 pellets), COMSOL 4.3, 2D axisymmetric
individual pellet at maximum
temperature in stack:
stress contour with 10000x
deformation
Overview of calculations developed in
COMSOL
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Transient Analysis in RELAP5
• Transient thermal hydraulics analysis of seven pin Pu-238 target holder: single hot tube plus six average tubes
• Primary code: RELAP5
• Dimensions: 1D, Time-Dependent
• Calculations:
• C-HFIR-2012-015, Rev. 0, Bare-Pellet Test Capsules
• C-HFIR-2013-006, Rev. 1, Partially Loaded Targets
• C-HFIR-2013-030, Rev. 1, Fully Loaded Targets
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Model Outputs
• The primary output for the model is the maximum target temperature, which is compared against an assumed melting temperature of 650 °C.
• Other outputs including the pellet and cladding temperature profiles, component temperatures, net thermal expansion.
• Transient analysis outputs also include target surface and bulk coolant temperatures
• Determination of initial fabrication gap and pellet shrinkage values corresponding to the melting temperature:
• For different cycles, irradiation positions, and material property data
• Evaluation at 100% (85 MW) and 130% (110.5 MW) steady state powers
• Evaluation at 100% (5 gpm) and 50% (2.5 gpm) coolant flow rates
• Evaluation at SBLOCA and LOOP events for transient analysis
• A “best-estimate” evaluation was made for the target temperature profiles at various reactor operating times (or burn-ups)
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LOOP EOC2 – Target surface temperatures never exceed
adjacent coolant saturation temperatures
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RELAP5 Transient Results
Parameter VXF-15, EOC1 VXF-15, EOC2
NpO2/Al pellet thermal conductivity
and thermal expansion
1200-C oxide,
31% non-conductive volume
1200-C oxide,
31% non-conductive volume
Cold fabrication gap (average) 0.5 mil radial 0.2 mil radial
Cold shrinkage gap 2.1 mil radial
(4.2 mil diametric)
1.7 mil radial
(3.4 mil diametric)
Fission gas release fraction 10% Xe/15% Kr 10% Xe/17% Kr
LOOP max. centerline temp. 642.7 C 643.5 C
SBLOCA max. centerline temp. 640.1 C 639.3 C
1
2
3
4
5
6
7
8
9
100 150 200 250 300 350 400
Allo
wab
le P
ellet
Dia
metr
ical
Sh
rin
kag
e (
Mils)
Maximum Pellet Heating (W/g) at 85 MW Reactor Power
RELAP Allowable Shrinkage
COMSOL Allowable Shrinkage
• Maximum pellet temperatures remain below pellet melting point of 650 C
• Boiling does not occur because target surface temperatures never exceed adjacent coolant saturation temperatures
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Fully-loaded target qualification for 2nd cycle irradiation at 130% overpower conditions
(estimated based on partial-target PIE results)
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Temperature profile during second irradiation cycle Days 0, 5, and 10
t = 10 d, Tmax= 404.1 °C t = 5 d, Tmax= 405.1 °C t = 0 d, Tmax= 219 °C
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Temperature profile during second irradiation cycle Days 15, 20, and 26
t = 26 d, Tmax= 413.3 °C t = 20 d, Tmax= 399.8 °C t = 15 d, Tmax= 400.6 °C
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Change in gap size from the start of the 2nd cycle
0
0.5
1
1.5
2
2.5
3
-10 -5 0 5 10
Rad
ial
Gap
Siz
e (
mils)
Axial Position w/ respect to reactor midplane (in)
Day 0 Day 5
Day 10 Day 15
Day 20 Day 26
0
0.2
0.4
0.6
0.8
1
1.2
0 5 10 15 20 25 30
De
cre
ase
in A
vera
ge G
ap S
ize
(m
ils)
Second Irradiation Cycle (Days)
Net Gap -Δ
Swelling
Thermal Exp.Contributions to the close in the gap (right) show that fission product swelling is the driving mechanism after the initial heat up and thermal expansion.
The change in the radial gap along the pellet stack is shown for different irradiation times (left) for an alternative irradiation position.
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Safety Factor = Ratio of Housing Failure to
Force Equilibrium
Fully-Loaded Target Final Qualification for Pellet and Housing Stresses and Failure
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Summary and Conclusions
• Safety analysis progress through phase 3
– Phase 1 (single pellet) and Phase 2 (partial targets) safety analyses are complete
– Phase 3 safety analysis is complete: Fully-Loaded Targets have been irradiated for 2 cycles and further irradiation is currently underway
– A best-estimate analysis for Phase 3 was made to allow irradiation of previous designs
– The current design is bounded by previous analyses
– Steady-state cases are more thermally limiting than transient cases
• 238Pu throughput optimization will require further analysis
– Using more extensive PIE data and pellet property measurements as inputs, new safety analyses may be developed to examine the potential for 5-20% neptunium loading increases
– Safety analyses considering the effects of filling the HFIR external reflector with targets will need to be made
• A continuous mode of optimization/production is the intended final phase of the project at HFIR
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Acknowledgements
• HFIR RRD Staff
• ORNL
• DOE NEUP
• NASA
Questions?
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Supplementary Graphics…
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Revisions 3 & 4: Pellet Centerline Temperatures in the Fully-Loaded Target 2nd cycle
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Revisions 3 & 4: Pellet Centerline Temperatures in the Fully-Loaded Target 2nd cycle & EOC-1
300
320
340
360
380
400
420
440
460
480
-10 -8 -6 -4 -2 0 2 4 6 8 10
Tem
pera
ture
(°
C)
Axial Position along the pellet stack (in)
Day 5 Day 10
Day 15 Day 20
Day 26 EOC-1 VXF-15
EOC-1 VXF-3