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Fuel Cycle Research and Development: Advances in Nuclear Fuel and Materials
Stuart A. Maloy Transmutation Clad Materials Technical Lead for Fuels Ken McClellan Ceramic Fuels Technical Lead Los Alamos National Laboratory Steve Hayes Metallic Fuels Technical Lead Idaho National Laboratory
Advanced Fuels Campaign Mission & Objectives in the Fuel Cycle Research and Development Program
Mission Develop and demonstrate fabrication processes
and in-pile (reactor) performance of advanced fuels/targets (including the cladding) to support the different fuel cycle options defined in the NE roadmap.
Objectives Development of the fuels/targets that
– Increases the efficiency of nuclear energy production – Maximize the utilization of natural resources (Uranium, Thorium) – Minimizes generation of high-level nuclear waste (spent fuel) – Minimize the risk of nuclear proliferation
Grand Challenges – Multi-fold increase in fuel burnup over the currently known
technologies – Multi-fold decrease in fabrication losses with highly efficient
predictable and repeatable processes
Once- Through
Modified Open
Continuous Recycle
Advanced Fuels
High-burnup LWR fuels Accident Tolerant Fuels for improved Safety
- Deep-burn fuels or targets after limited used fuel treatment - High burnup fuels in new types of reactors
- Fuels and targets for continuous recycling of TRU in reactors (possibly in fast reactors)
Outline
Transmutation Fuel Development for Fast Reactors – Transmutation Cladding Development
• Qualify HT-9 to Radiation Doses >250 dpa • Develop Advanced Radiation Tolerant Cladding Materials • Develop Coatings and Liners to prevent FCCI
– Metallic Fuel Development • History of metallic fuel development • Improved Fabrication process • Performance of Metallic fuels with Minor Actinides • Future direction
Accident Tolerant Fuel Development for Light Water Reactors – Clad Development
• Steam testing • Tube development • Radiation testing • Future plans
– LWR Ceramic Fuel Development • Improved ATF fuels • Future testing
3
4
Approach to Enabling a Multi-fold Increase in Fuel Burnup over the Currently Known Technologies
Coa
ting
Line
rs
Adv
ance
d A
lloys
F/M
Ste
els
Adv
ance
d A
lloys
Cr Si Al
Increasing content
F/M
Ste
els
HT-
9
Adv
ance
d F/
M
Ste
els,
e.g
. NF6
16
OD
S S
teel
s
Adv
ance
d A
lloys
200 dpa 300 dpa 400 dpa
500 C 600 C 700 C
Reduced embrittlement, swelling, creep
Enhancements with Fabrication Complexity
Enhancements with Fabrication Complexity Enhancements with Fabrication Complexity
Different Reactor options to change
requirements LFR, GFR
FCCI
Radiation Temperature
Corrosion
Ultimate goal: Develop advanced materials immune to fuel, neutrons and coolant interactions under specific reactor environments
Ultra-high Burnup Fuels
Transmutation Cladding Development
Qualify HT-9 to >250 dpa
Significant data has been obtained on previously irradiated materials. How do we obtain data to dose levels out to 400 dpa?
300
350
400
450
500
550
600
650
700
750
800
0 50 100 150 200 250 300 350 400
Dose (dpa)
Irra
diat
ion
Tem
pera
ture
(C)
FFTF/MOTA Specimens
NSMH Data
ACO-3 Duct MATRIX
How do we get here?? International Collaborations New irradiations in CEFR FFTF/MOTA extended irradiations in other irradiation facilities (BOR60). Model Development High Dose Ion irradiations
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Analysis of Specimens from ACO-3 Duct
Total specimens= 144 Charpy, 57 compact tension, 126 tensile specimens, 500 TEM
Charpy and Compact Tension testing performed at ORNL. Then thermal annealing tests performed on Charpy halves.
Completed tensile testing at LANL from 6 different locations along the duct at 25, 200 and the irradiation temperature.
In addition, detailed microstructural analysis performed.
New specimens EDM machined for BOR-60 Irradiation
0
200
400
600
800
1000
1200
300 350 400 450 500 550
HT-9 Control
Irradiation Temperature (C)
0.2%
Offs
et Y
ield
Str
ess
(MPa
) 25 C 200 C Tirr
Tensile Specimens from ACO-3 Duct Under Irradiation in BOR-60
LANL recently shipped ACO3 and FFTF/MOTA samples to Russia for re-irradiation in BOR-60 through CRADA with Terrapower.
Re-irradiation of specimens began in January 2014
Diagram showing specimen cut plan from ACO-3 duct for re-irradiation in BOR-60
Terrapower, LANL, ORNL
Improved Radiation Response of New NQA1 Heat of HT-9
300 lb heat of HT-9 produced by Metalwerks following NQA-1 quality control Tensile specimens irradiated in ATR to 6 dpa at 290°C
– Hardening observed but excellent ductility retained after low temperature irradiation Ion irradiations performed to 600 dpa at 425°C
– Minimal swelling observed in tempered martensitic grains after ion irradiation to >500 dpa.
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INL heat ferrite (<5% volume fraction)
INL heat tempered martensite
33% CW heat – tempered martensite
0
200
400
600
800
1000
1200
0 5 10 15 20 25
Control vs. Irradiated, All Ferritic HT9 Control TB#1HT9 IRR TB01Control T91 TA#1Irr T91 TA04Control Eurofer ER07Eurofer Irr ER04NF616 Control NF08NF616 Irr NF04 F82H-IEA Irr HA05
Engi
neer
ing
Stre
ss M
Pa
Engineering Strain %
INL-HT-9 Heat, best ductility
INL (NSUF), LANL, UCSB
290°C 6 dpa
425-450°C
Transmutation Cladding Development
Develop Advanced Radiation Tolerant Materials
Nanostructured Ferritic Alloys
• Strength & damage resistance derives from a high density Ti-Y-O nano-features (NFs)
• NFs complex oxides (Ti2Y2O7, Y2TiO5) and/or their transition phase precursors with high M/O & Ti/Y ratios (APT)
• MA dissolves Y and O which then precipitate along with Ti during hot consolidation (HIP or extrusion)
• Oxide dispersion strengthened alloys also have fine grains and high dislocation densities
Y-YO-Ti-TiO-O
UCSB, LANL, ORNL
Excellent Retention of Ductility observed for ODS Steels after irradiation to 6 dpa at 293C
UCSB, LANL, INL (NSUF)
Ion Irradiation Induced Swelling Comparisons
All materials irradiated in the same facility. Tempered F-M steels reached a terminal 0.2%/dpa swelling rate that
has been observed ferritic alloys during neutron irradiations.
MA957 and 14YWT are exhibiting an extended nascient low swelling period.
Data suggests that 14YWT is exhibiting better resistance than MA957, but results are very similar.
Will 14YWT and MA957 abruptly transition to high swelling rate?
Transmutation Cladding Development
Develop Coatings and Liners to prevent FCCI
Directed laser light 45.0°
45.0°
Retreating tube withrotating target pellet holder
Roller to retreat tube
A customer-designed laser deposition system for inner wall coating of long tube is being used at Texas A&M University for the FY2012 coating work
Ce Fe Fe
Ceramic coating
Diffusion couple studies on the effect of ceramic coating on suppressing fuel-cladding-chemical-interaction (550 – 600 °C for 12-24 hours)
The effect of a 500 nm thin TiN coating on Fe-Ce interaction (550°C/48 hrs)
Coating and Diffusion Couple Study for FCCI Mitigation
High Dose Ion Irradiations
Core Materials Research and Development – 5 Year Plan
FY’16 FY’15 FY’14 FY’11 FY’13 FY’12
STIP- IV (PSI) Specimen PIE
MATRIX-SMI and 2 (Phenix) Specimen PIE
Re-irradiation of FFTF specimens in BOR-60
Use data for physics-based model development of cladding
FFTF (ACO-3 and MOTA) Specimen Analysis
Advanced Material Development (improved radiation resistance to >400 dpa)
Qualify HT-9 for high dose clad/duct applications (determine design limitations)
ODS Ferritic Steel Material Development
Produce ODS Tubing
Advanced Materials Irradiation in BOR-60 and CEFR
Advanced Material Development (improved FCCI resistance to >40 % burnup)
Development of Coated and Lined Tubes
Rev. 6 of AFCI (FCRD) Materials Handbook
PIE on Lined Irradiated Tube
Develop ODS Tubing and Weld specifications
Data to 250-300 dpa on F/M and 100-150 on Inn. Material
Data on Advanced Materials to 80-100 dpa
Metallic Fuel Development History & Benefits
History of Metallic Fuels in Fast Reactors
EBR-I (1951) • Unalloyed U • U-2Zr • Pu-1.25Al
UK Dounreay Fast Reactor (1963) • U-0.1Cr • U-7Mo • U-9Mo
Enrico Fermi FBR (1963) • U-10Mo
EBR-II (1964) • U-5Fs • U-10Zr • U-20Pu-10Zr
FFTF (1982) • Assembly testing of U-10Zr • Assembly testing of U-20Pu-10Zr
Key Features of Metallic Fuels
Metal fuel characteristics • U-Pu-Zr alloy base (good irradiation stability) • 75% smeared density (accommodate fuel swelling, mitigate FCMI) • Large fission gas plenum (accommodate high gas release) • Na bond in fuel-cladding gap (keep fuel temperatures low) • Low-swelling FMS cladding (minimize cladding/duct dimensional
changes)
Historic benefits • Outstanding fuel reliability to high burnup (~20 at.-%) • Compatibility with proliferation-resistant electrochemical recycle • Simple, compact (demonstrated remote) fabrication processes • Synergistic with passive approach to reactor safety
Metal Fuel Development Improve Fabrication Process
Charge Material Mass Wt%
Uranium 80.64 80.9%
Zirconium 3.77 10.0%
Plutonium 7.36 7.4%
Americium 1.66 1.7%
Zirconium 6.20
Total 99.63 100%
Advanced Fabrication Development
Minor actinide bearing fuel fabrication • Bottom pour crucible configuration • Casting under pressure
Casting Product
Mass Wt% (of total charge)
Cast Material 71.215 71.5%
Casting Heel 26.14 26.2%
Dross 2.231 2.2%
Total 99.589 99.96%
Greatly improved melt utilization, near-zero Am loss during fabrication.
Metal Fuel Development Performance of Metallic Fuels with Minor Actinides
AFC Test Series Irradiates Miniature Fast Reactor Fuel Rodlets
Capsule Assembly Completed Rodlets
AFC Series Irradiation Testing in ATR
East Flux Trap
4
2 5
1
3 6
7
AFC-1,2 Testing in EFT • Cd shrouds in 1, 2, 3, 4 • 6 rodlets per capsule • Up to 24 rodlets irradiated
simultaneously AFC-3,4 Testing in 4 OA’s
• Cd shrouds in up to 4 OA’s • 5 rodlets individually
encapsulated per OA • Up to 20 rodlets irradiated
simultaneously; cycle-by-cycle shuffling of rodlets
Capsule Limits • LHGR ≤ 500 W/cm • PICT ≤ 650°C • Capsule pressure ≤ 975 psi
Metallic Fuel Alloys Tested in ATR
A2A1 A2A2 A2A3 A2A5 A2A6 A2A4 A1D6
Summary of Irradiation Performance of Metallic Fuels with MAs
Metallic Fuel Performance
Anomaly: AFC-2A,2B
• Simulated Ln FP carryover from recycle • PICT 550°C; LHGR 355 W/cm; ~12% BU • Multiple failures during irradiation
Irradiation performance of AFC metallic fuels has been shown to be “typical” of historic understanding,
for wide variations of U, Pu, Zr, & MA contents.
Fuel Melting
Fission Product Transport and FCCI in Metallic Fuels
Fuel-cladding Chemical Interaction • Lanthanide fission products (La,
Ce, Pr, Nd) react with SS cladding • Interaction product brittle • Considered to be cladding wastage
• Low melting; limits transient overpower
Metal Fuel Development Future Direction
Immobilizing Lanthanide Fission Products
Lanthanide fission products migrate to fuel surface and interact with cladding, the primary source of FCCI
Use additives to chemically bind the lanthanides and immobilize in fuel matrix
Double Benefit:
• Homogenize/stabilize Ln’s carried over from recycle
• Mitigate FCCI during irradiation
R.D. Mariani, D.L. Porter, T.P. O'Holleran, S.L. Hayes, J.R. Kennedy, J. Nucl. Materials, 419 (2011) 263.
Back scattered electron image of
U-15Zr-3.86Pd-4.3Ln (Ln = 53Nd-25Ce-16Pr-6La, in wt%)
Palladium (Pd) combines with the lanthanides, as expected from analysis of thermodynamic and PIE data.
Accident Tolerant Fuel Development
Cladding Research
31
Grand Challenge for Core Materials for Next Generation LWR Fuels
Develop and test advanced alloys for Next Generation LWR Fuels with Enhanced Performance and Safety and Reduced Waste Generation – Low Thermal Neutron Crossection
• Element selection (e.g. Zr, Mg) • Reduce cladding wall thickness
– Irradiation tolerant to at least 15 dpa • Resists swelling and irradiation creep • Does not accumulate damage • Stable microstructure (resists RIS)
– Mechanically robust under loading and transportation conditions – Compatibility with Fuel and Coolant
• Resists stress corrosion cracking • Resists accident conditions (e.g. high temperature steam) • Resists abnormal coolant changes (e.g. salt water)
– Weldable and Processed into tube form • Maintain hermetic seal under normal/off-normal conditions
Prioritized Cladding Screening Tests
Metallic Cladding • Steam oxidation, post-steam ductility / strength tests • Environmental testing (corrosion, irradiation assisted SCC, SCC, erosion) • Basic chemical compatibility at normal operating conditions • Irradiation environmental testing (sample coupons; determine effects on mechanical properties) • Mechanical testing (assuming baseline unirradiated data is already available, additional mechanical
tests would not be required until after a material has passed the other screening tests identified)
Ceramic Cladding • Mechanical integrity tests (simple tests for hermeticity, break) • Steam oxidation, post-steam ductility / strength tests • Basic chemical compatibility at normal operating conditions • Environmental testing (LWR water exposure over a range of chemistry conditions) • Mechanical properties measurement • Irradiation environmental testing (sample coupons; determine effects on mechanical properties)
Hybrid Cladding / Coatings on Cladding Tube
• Assess coating adhesion – initial, thermal cycling, with flow (spall), wear (fretting) • Based on substrate and “coating” materials and configuration, follow metallic or ceramic cladding priorities as defined
above; additional details for coated cladding options include: - Environmental testing w/small exposed region (autoclave) - Chemical compatibility to include coating / substrate compatibility (establish temperature limit)
Objectives
Measure Kinetics of Oxidation in Steam • Steam oxidation testing up to 1300C (ORNL) • Fundamental oxidation studies in steam (LANL)
Develop Processing Techniques to produce thin-walled tubing • Producing tubing of MA-956 with 250 micron thick walls (LANL) • Measuring mechanical properties of thin walled tubes (ORNL) • Weld development on thin-walled tubing (INL and ORNL)
Measure Radiation Tolerance of ATF ferritic alloys • Ion irradiated materials (LANL) • ATR irradiated materials
- Tensile testing (LANL) - Fracture toughness testing (ORNL)
Developing Improved ATF Fe-Cr-Al alloy (ORNL) • Weld development for improved ATF alloy (INL) • Mechanical testing and ion irradiations (LANL)
Develop Advanced ATF (Molybdenum clad) (LANL) • Production of Mo tubing using FB-CVD processing
34
Properly alloyed metals as protective as Si-based ceramics at 1200°C
Example from FCRD experiments at 1200°C in steam at 3.4 bar (50 psia) for 8 h All low mass gain: 310SS (Cr2O3), FeCrAl, Kanthal APMT (Al2O3), CVD SiC (SiO2)
Mass change is net value: oxide growth, spallation (minor) and volatilization Commercial and model alloys included to fundamentally understand role of composition and minimum
amount of Cr (and Al) needed for protective behavior
Zirc
aloy
S
S-3
04L
35
Initial Neutron Irradiation Testing Results
Sample tested – MA-956;(Fe-20Cr-5Al-ODS alloy)
Irradiated to 6 dpa at 290 C in the ATR Reactor
Large reduction in ductility but still fails in a ductile manner.
36
Tube Produced from ORNL FeCrAl 1st Gen Advanced Alloys
Tube processing of 1st Gen ORNL FeCrAl alloys (T35 and T54) performed by Century tubing • Warmed initially to ~200 C to
make pointed end. • Annealed at 815C for 15
minutes in hydrogen • 16 cold reduction passes over a
mandrel through a carbide die of 14-24% per pass
• Intermediate anneals at 850C in hydrogen for 15 minutes
• Final dimensions for each alloy are 10 feet long, 0.374 in. OD, 0.014 in. wall thickness
Core Materials Research and Development ATF Clad Development - 5 Year Plan
FY’16 FY’15 FY’14 FY’13 FY’12
Develop Steel thin-walled tubing
Advanced Material Development (improved accident tolerance for LWR’s)
Oxidation testing in Steam on
Advanced Steel Alloys
Thin walled tube development for advanced steel alloys
ATR irradiation of ATF cladding with advanced fuels
Weld development for thin walled tubing
Oxidation testing in Steam on Advanced Steel Alloys
Irradiation testing and PIE on ATF clad materials
Corrosion testing on ATF clad materials
IASCC testing on ATF clad materials
Produce thin walled tubing for ATF irradiation
Oxidation testing in Steam on Advanced Steel Alloys
after irradiation
Qualified weld procedure for thin walled tubing
Report on initial results of LOCA testing on ATF clad materials
Final report of LOCA testing on ATF clad materials
Report on irradiation testing of clad materials to at least 15 dpa
Summary report on corrosion resistance of ATF clad materials
Coordinate and collaborate with ATF fuel development with industry through FOA and NEUP projects including the IRP on development of ATF
FY’17
LWR Ceramic Fuel Development
Approach and Requirements for Development of Ceramic Fuels With Enhanced Accident Tolerance
Approach: • Develop ceramic fuels with at least comparable performance under all conditions,
and improved performance in accident scenarios → fuel can enable cladding change (while maintaining/increasing linear power)
• Develop fabrication and evaluation techniques that can enable new fuels and can enable accelerated development
Preserve/Improve Characteristics for Normal Operation: • Burnup/Cycle length (criticality, fuel performance) • Operations (power distribution, peaking factors, margins, etc.) • Reactivity coefficients and control (shutdown margin, rod worths) • Handling, transportation, storage (isotopics, dose) • Compatibility with infrastructure (economics)
To warrant the label of “ATF” (enhanced) must Improve Response to Transients/Accidents: • Anticipated Operational Occurrences (AOOs) • Design Basis Events (LOCA, RIA) • Beyond Design Basis Events (Fukushima)
39
Advanced Ceramic Fuels Development Targets and Constraints
Target key performance aspects • Increase thermal conductivity • Increase toughness/engineer cracking • Decrease oxidation • Maintain current cycle length and performance while enabling
cladding changes Constraints:
• Compatibility with new ATF cladding • Must be economical
- Lower fuel cycle cost, “Low cost” manufacture, High power, Deeper burn
• Minimize deviation from existing infrastructure, experience and expertise
- Consider enrichment increases in stages: e.g. <5, 5-7, 8-10, <20% - Similar/same infrastructure: UF6 source, open powder processing or MOX-type line
40
Composite Ceramic Fuels
Engineering a ceramic fuel via a composite approach enables industry to utilize existing infrastructure (or at least mitigates deviation from their capabilities and experience base)
Design via high fissile density (U-based) constituents can enable low(er) enrichments relative to conventional composite (inert matrix) fuel concepts
Focus on maintaining/improving • Normal operations (steady state/transient) • AOO/DB accidents performance • BDB accident performance
Initial ATR irradiation mid/late 2014
Boride
Nitride Silicide
Oxide
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Full matrix of candidate systems
Initial screening by neutronics, thermodynamic calculations, and sintered composites
Systems proposed based upon fissile density, melting point, thermal conductivity, potential for compatibility with standard commercial fabrication infrastructure/expertise
10B enrichments identified as commercially available: 200000, 3000, 1000 & 100 ppm.
42
Uranium Borides
Compounds of Interest UB4
• Tm = 2495°C • Tetragonal structure
� ρU ~7.9 g/cm3 • Compatible with UO2 • B loss for melt synthesis • Some irradiation data
UB2 • Tm = 2385°C • Hexagonal structure
� ρU ~11.7 g/cm3 • B loss for melt synthesis • Risk of molten U above
668°C if U-rich
UB12 • Tm = 2145°C • Cubic structure
� ρU ~3.8 g/cm3 • B loss for melt synthesis • Risk of free B
H. Okamoto, Binary Alloy Phase Diagrams (1990) ρU ~9.7 g/cm3 for UO2
43
Uranium Silicides
Compounds of Interest U3Si2
• Tm = 1665°C • Tetragonal structure � ρU ~11.3 g/cm3 • Risk of molten U above
985°C if U-rich • Brittle, high hardness (742
Vickers)
U3Si5 • Tm = 1770°C • Hexagonal structure � ρU ~7.5 g/cm3 • Not compatible with UO2
USi • Peritectoid = 1580°C • Orthorhombic structure � ρU ~9.9 g/cm3
U3Si • Peritectoid = 930°C • Tetragonal structure � ρU ~15.0 g/cm3 • Risk of molten U above
668°C if U-rich • Tough/low hardness (265
H. Okamoto, Binary Alloy Phase Diagrams (1990)
U3S
i 5
U3S
i
ρU ~9.7 g/cm3 for UO2
44
ATF-1 Fuel Fabrication
Leading candidates for first irradiation under ATF-1 are UO2/UB2 and UN/U3Si5 • Initial scoping assembly-level calculations are promising • Initial scoping thermodynamic examination indicates viable stability
- Oxide/boride (see Voit/Besmann talk)
• Basic compatibility demonstrated via initial fabrication efforts
Promising for maintaining/improving performance while enabling cladding change
• Increased thermal conductivity, same/longer cycle length, same/higher linear power
Specifics of irradiation, e.g. composition, linear power, cladding, etc TBD
L2 for delivery of fuel pellets to INL 12/5/14 for insertion in ATR cycle 158B
45
Summary
Transmutation Fuel Development • Transmutation Cladding Development
- Developing cladding for high burnup fuel. - Research centered on HT-9 and Advanced ODS steels - Liners and coatings being developed to prevent FCCI
• Metallic Fuel Development - Metallic fuels being developed for transmutation of minor actinides in a fast spectrum. - Advances being made fabrication processes and fuel designs improve performance.
Accident Tolerant Fuel Development • Clad Development
- LWR claddings are being developed and tested for improved accident tolerance. - Development centered on Fe-Cr-Al alloys
• LWR Ceramic Fuel Development - Ceramic fuels are being developed for improved LWR performance including improved thermal
conductivity and toughness - A composite approach is being undertaken with combinations of borides and nitrides with uranium
dioxide.
46
Summary & Conclusions
Summary & Conclusions (Metallic Fuels with MA Additions)
Fabrication • Issue of Am volatility during casting has been resolved at bench-
scale using surrogate systems; validation with Am underway Irradiation Performance
• Wide spectrum of U-Pu-Am-Np-Zr fuel alloys have been irradiated in ATR
• Performance is good, “typical” of historic metallic fuel behavior • Comparison Report (FY15) will validate ATR Cd-shrouded test
results vs. data from EBR-II, FFTF, and Phénix Future Direction
• Development of the “Advanced Metallic Fuel Concept” - Additives for Ln FP stabilization and immobilization - Cladding coatings/liners - Fuel fabrication by extrusion (CRADA with TerraPower)
• Goal: Demonstrate reliable performance to ultra-high burnups
Screening: Initial UN/U3Si5 Processing
U3Si5 (15 vol%) in UN based on neutronics calculations
Method of fabrication • UN by CTR-N • U3Si5 by arc-melt • Crush grind and sieve
appropriate quantities of U3Si5 and UN powders to <325 mesh
• Blend 15 vol% U3Si5 in UN powder – spex mill in small nested containers
• Press pellets at 40 MPa Screening studies conducted to
evaluate compatibility and densification
• 1500ºC – 72.93% TD • 1800ºC – 91.4% TD
Dilatometer experiments to understand sintering kinetics
• 1650ºC – 1800ºC
49
H. Okamoto, Binary Alloy Phase Diagrams (1990)
U3S
i 5
Comparison of Spectra (ATR vs. SFR)
ATR Neutron Energy Spectrum – Highly thermal spectrum in EFT with no
neutron filter – Unaltered spectrum will result in
significant self-shielding in dense, highly-enriched fuels
Cd-shroud Integral with Experiment Basket
– Efficient removal of neutrons with energies below cadmium cut-off
Resulting Spectrum – Filtered spectrum in experiment does not
have prototypic fast neutron component – Epi-thermal component responsible for
most fissions; much more penetrating than thermal neutrons
– Test fuels are free of gross self-shielding
Radial Flux Depression and Temperature Profile in Test Fuels
How prototypic are AFC rodlets irradiated in the ATR?
• Assessed by analysis • Radial power profiles calculated
w/MCNP • Depletion in fuel and Cd shroud
calculated w/ORIGEN (MCWO) • 1-D thermal analysis using radial
powers
Resulting temperatures for AFC-2C,D oxide rodlets
• 3 cases: SFR, unshrouded ATR, ATR w/Cd shroud
• w/Cd shroud, peak-to-avg power at fuel periphery is 1.22; fuel central temperature 58°C less than SFR (~400°C less for unshrouded case)
Metallic Fuel Alloys Tested in ATR
PIE of AFC Test Series Rodlets
Baseline Non-destructive PIE • Visual examination • Metrology • Neutron radiography • Gamma ray spectroscopy
Baseline Destructive PIE • Fission gas collection and analysis • Optical metallography • Micro-hardness • Analytical chemistry (burnup analysis)
Additional PIE (i.e., not performed on all fuels)
• Property measurements (e.g., STDM) • Micro-XRD • SEM (w/EBSD, WDS) • FIB TEM
Advanced Metallic Fuel Concept (INL-ANL Collaboration)
Proposed innovations: 1. Decreased fuel smeared density 2. Coating or liner on cladding inner diameter 3. Vented fuel pin 4. U-Mo alloy base 5. Targeted fuel alloy additions 6. Advanced fabrication method (i.e., extrusion; CRADA
w/TerraPower) 7. Fuel recycle
Innovations proposed to deliver multifold advances in fuel reliability and performance, actinide utilization, and fabrication efficiency, while retaining all the historic benefits of the metallic fuel system.
Development of the “Advanced Metallic Fuel Concept” initiated as award under innovative fuel competition, but has become main line of metallic fuel development.
700°C Diffusion Couples
Nd Fe Nd/Pd
G.W. Egeland, R.D. Mariani, T. Hartmann, D.L. Porter, S.L. Hayes, J.R. Kennedy, Journal of Nuclear Materials, 432, (2013) 539.
Fe, , Pd
Nd5Fe17
Nd2Fe17
At 700°C, the Fe-Nd interface experienced liquefaction, while the Fe-(Nd/Pd) interface experienced no reaction.
Advanced Metallic Fuels Concept: Irradiation Testing Strategy
Fuel Fuel Alloy
Geometry & Bond
Smear Density
Alloy Additions
Burnup (HM) Cladding
U Zr Solid w/Na
55% none
8-10% HT-9 Pd
Mo In 65% U+Pu 15-20% liner Ga
Annular w/He Mo-Ti-Zr other
U+Pu+MA 75% 30-40% coating combos
• Short-term irradiations: early PIE to show feasibility • Long-term irradiations: demonstrate ultra-high burnup • Early materials testing of lined/coated cladding for incorporation in
subsequent short-term irradiations • Investigate alloy additions to immobilize lanthanide fission products
Start End EFPD Cycle A10-1 A10-2 A10-3 A10-4 A10-5 A11-1 A11-2 A11-3 A11-4 A11-5 A9-1 A9-2 A9-3 A9-4 A9-5 A12-1 A12-2 A12-3 A12-4 A12-5
10/15/2011 11/26/2011 42 150B
12/13/2011 2/11/2012 56 151A
2/28/2012 3/27/2012 10 151B
4/7/2012 5/5/2012 28 151B-2
11/27/2012 1/18/2013 52 152B MT-OA-1 MT-OA-2 Materials3A-3
MT-OA-4 MT-OA-5
3/23/2013 5/18/2013 56 154A
8/20/2013 10/12/2013 53 154B
11/5/2013 1/4/2014 60 155A
1/28/2014 3/22/2014 53 155B
5/31/2014 7/26/2014 56 156A
8/19/2014 10/18/2014 60 157A
11/11/2014 1/10/2015 60 157B
4/4/2015 5/27/2015 53 158A
6/20/2015 8/12/2015 53 158B
10/13/2015 11/28/2013 46 160A
12/22/2015 2/20/2016 60 160B
4/16/2016 6/11/2016 53 162A
7/5/2016 9/3/2016 60 162B
9/27/2016 11/22/2016 56 163A
6/2/2017 7/29/2017 57 164B
9/8/2001 1/16/2017 57 166A
11/21/2017 1/13/2018 53 167A
3/20/2018 5/12/2018 53 168A
FY14
FY15
FY16
FY17
FY18
U-10Zrannular65% SD
3E-1
U-10Mosolid
55% SD3E-2
U-10Moannular75% SD
3E-3
U-10Zrannular75% SD
3E-4
U-10Zrsolid
55% SD3E-5
FY12
Uannularwith Na60% SD
Zr/V coating10% BU
4B-3
U-Pu-MA-Moannular55% SD
4C-5
U-PuGa-1Pd-Zr
U-PuGa-2Pd-Zr
solid75% SD
4C-2
U-10Mosolid
75% SD3C-1
FY13
Uannular60% SD
Zr/V coating10% BU
4B-4
Uannular50% SD
ZrN/V liner5% BU4B-5
3T6
4D-1
3T7
4D-5
Uannular60% SD
Zr/Vliner
5% BU4B-1
Uannular50% SDZrN/V
coating10% BU
4B-2
U-10Mosolid
65% SD4C-3
U-2Pd-10MTZsolid
75% SD4A-4
U-10Moannular55% SD3 years
3D-4
U-4Pd-13Zrannular55% SD6 years
3D-5U-10MTZ
solid75% SD
4A-3
U-In-MoU-Pd-Mo
solid75% SD
4C-4
U-10Zrannular55% SD6 years
3D-1
U-4Pd-13Zrsolid
55% SD3 years
3D-2
U-10Mosolid
55% SD3 years
3D-3
U-10Moannular55% SD
3C-2
U-10Zrsolid
65% SD3C-3
U-10Zrannular55% SD
3C-4
U-1Pd-13ZrU-2Pd-13Zr
solid75% SD
3C-5
AFC-4AFC-3
U-10Mosolid
75% SD3A-1
U-10Moannular55% SD
3A-2
Outboard Position A10 (SE) Outboard Position A11 (SW)
U-4Pd-10Zrannular55% SD
3B-2
U-10Mosolid
55% SD3B-4
U-10Mosolid
55% SD3B-5
Outboard Position A12 (NW)
U-10Zrannular55% SD
3A-4
U-1Pd-10ZrU-2Pd-10Zr
solid75% SD
3A-5
Outboard Position A9 (NE)
U-4Pd-10Zrsolid
55% SD3B-1
ATR Cycle
U-10Moannular65% SD
4A-1
U-2.5W-10Mosolid
75% SD4A-2
U-10Zrsolid
75% SD4A-5
U-Pu-MA-Zrannular55% SD
4C-1
AFC-3/4 Irradiation Series: Test Matrix
Short-term, early PIE: demonstrate feasibility
Long-term: demonstrate “Grand Challenge”
TerraPower (CRADA)
Low Burnup, Feasibility, Additives
BaselineAlloysAdditivesGrand ChallengePu / MATerraPowerMaterials
• Any desired combination of powders: metals, alloys, and dispersoid, such as oxides, carbides, borides, etc.
Typical Processing Route for ODS Alloys
The conventional approach is to ball mill alloy and Y2O3 powders together
HIP near net shape final product
Y2O3
(Fe-14Cr-3W-0.4Ti)
Zoz
59
Densification and compatibility studies
Screening: Initial UO2/UB2 Processing
No visible interface cracking or reaction
10 vol% UB2 in UO2.16
Sintering Temperature (ºC)
%TD
1000 60.73
1350 75.77
1450 84.99
1600 86.73
1640 84.99
Sintered UB2-UO2.16 Pellet
“Best Practice” processing of NFA1:
APT (LEAP) Analysis (UCSB)
Y YO O Ti TiO Y-YO-O-TiO-Ti
Map
20 nm Y/Ti/O Y/Ti/Cr/O Number Density
(1023/m3) Diameter
(nm) Solute Fraction
(%)
13.7/41.8/44.5 10.5/32.0/23.6/34.0 6.86 2.02±0.78 0.74
High number density of nano-size O-enriched particles are consistent with previous 14YWT heats
Averages
• Unusual combination of high tensile strength and ductility • Very low brittle-ductile transition temperature (-150 to -175°C) –>
high isotropic strength and ductility in the presence of deep-sharp cracks (toughness)
Atypical isotropic toughness Averages
uniform
total
NFA-1 Strength, Ductility and Toughness
Deeply pre-cracked bend bars
UCSB
Fabrication of Cladding Tubes from ODS alloys
Working with PNNL (Curt Lavender) and CEA on Pilger processing of starting thick walled tubing and J. Lewandowski (CWRU) on hydrostatic extrusion 14YWT
CR6 14YWT CR6b
9YWTV PM2
3 cans were extruded with mandrel at 850ºC and decanned • 6-7 mm wall thickness; 31-32 mm diameter; 10.5-11.3 cm long
LANL, PNNL, CWRU
1. Hydrostatic EXTRUSION TEMP: 1500F (815C) 2. RAM SPEED: 0.5 in/min, however 1st 0.5” of extrusion,
speed was 0.7 in/min 3. SOAK TIME: 10 min 4. OVERALL EXTRUSION: 25 min 5. ER: 4:1, 45 DEG TAPER DIE (actual 0.495 diam) 6. CLAD/MANDREL DESIGN DIFF FROM PREVIOUS