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

7

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.

9

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

41

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