Luca Capriotti, Jason M. Harp*

Idaho National Laboratory

(*now Oak Ridge National Laboratory)

Nufuel-MMSNF 2019 Workshop

PSI Auditorium, Switzerland

November 4-7, 2019

INL/CON-19-56301

Nuclear Technology Research and Development

Postirradiation Examination on Innovative

Advance Reactor Metallic Fuel Concepts

Contents

◼ Introduction

➢ Advance Fuel Campaign

➢ PIE strategy

➢ Metallic fuel background

◼ Advance fuel form / High burnup concepts test

➢ Recent engineering PIE results & comparison

➢ SEM preliminary results

◼ Summary & Prospective

2

NTRD Advanced Fuels Campaign

3

◼ Enhance the performance and safety of the USA current and future reactors

◼ Enhance proliferation resistance of nuclear fuel

◼ Effectively utilize nuclear energy resources

◼ Address the longer-term waste management challenges

History of the AFC Irradiations

2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 2016 2017 2018 2019

1B

1D

1F, 1G

1Æ, 1H

FUTURIX-FTA

FUTURIX-FTA

2A

2B

2C

2D

2E

3A, 3C

3B, 3D

3F

4A

4B/D

4C

4F

Future

AFC-1 36 rodlets

• non- or low-fertile compositions for

accelerators and fast reactors

• metallic and nitride fuels

• Am, Np for transmutation

• sister experiments for low and high

burnup comparison

• FUTURIX-FTA: Phénix fast reactor

experiment with select AFC-1

compositions

AFC-2 29 rodlets

• low-fertile compositions for fast

reactors

• metallic and oxide fuels

• Am, Np for transmutation

• Assume batch extraction of TRU

• RE additions to simulate recycling

carry-over

• sister experiments for low and high

burnup comparison

AFC-OA ~30 rodlets

• low-fertile compositions for fast

reactors

• metallic fuels

• innovative design features for ultra-

high burnup

• smear density, geometry, alloy,

additives, …

• short and long term tests

PIE Strategy –Moving towards a mechanistic understanding

of nuclear fuel performance

Scale of

Exam <1 µmMacro

TO

OL

SF

uel

Perf

orm

an

ce

Number of

Samples

Section

pinsA few

pinsSelect

samples to

EM

FIB Sample

Preparation

Visual Exams

Did it Fail

Catastrophically

How much did the

geometry change

Neutron Radiography

Dimensional Exams

Gamma Spectrometry

Fission Gas Release

Optical Microscopy

Chemistry Analysis

Fission Product Migration

Microstructure,

Restructuring, FCCI

Burnup, Actinide Balance

Electron Microscopy

SEM, TEM, EPMA

Focused Ion Beam

Atom Probe

Elemental distribution

3D microstructure

24 26 28 30 32

Rela

tive

Ac

tivit

y

Z Position in PGS

Co-60

Eu-154

Ce-144(Pr-144)Cs-137

BaselineBeyond

Baseline

Available PIE

◼ Baseline Nondestructive PIE

– visual

– neutron radiography

(thermal, epi-thermal)

– gamma scan (axial isotopic data)

– metrology (radial swelling)

◼ Baseline Destructive PIE

– fission gas analysis

– optical microscopy / metallography

– microhardness

– analytical chemistry / burnup analysis

– mechanical properties

◼ Advanced PIE

– Microstructural analysis (SEM)

– Microchemical analysis (EPMA, SEM)

– Phase analysis (pending deployment)

– Thermal properties (laser flash, TCM)

– XRD, Micro XT

– 3D Microstructure (FIB, PFIB)

– Lower scales (TEM, APT)

60.E+00

1.E+03

2.E+03

3.E+03

4.E+03

5.E+03

6.E+03

24 25 26 27 28 29 30 31 32

Rela

tive

Ac

tivit

y

Z Position in PGS

Co-60

Eu-154

Ce-144 (Pr-144)

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)

– Qualification of U-10Zr

– Assembly testing of U-20Pu-10Zr

Key Features & benefits of Metallic

Fuels

◼ Historic benefits

– Higher breeding ratio (fissile and fertile)

– Benign response to accident condition

– Hard neutronic spectrum

– 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 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) 39,000 EBR-II pins

fabricated remotely in

1960’s

Schematic of a metallic, Na bonded, fast reactor element

Carmack et al., J. Nucl. Mater. 392 (2009)

G.L Hofman and L. C. Walters, "Metallic

Fast Reactor Fuels," Materials Science

and Technology Vol. 10A, 1994.

Fast Reactor Fuel Performance Challenges

& New Concepts

◼ Historical Fuel Performance Issues

– Swelling - limited burnup to 3 at. %, Solved early in EBR-II testing

with lowering Smeared Density to 75% to allow for interconnected

porosity releasing fission gas, solid fission product build-up limits

fuel to 15-20 at.% burnup

– Alloying elements to raise the fuel melting temperature and tailor

the phase of U or U+Pu in the fuel (Zr, Fs, Mo, Ti)

– Fuel Cladding Chemical Interaction (FCCI)

• FCCI occurs at nominal operating conditions in U and U-Mo fuels and

limits burnup to 10at. % (U-Fe, U-Ni interaction typically)

• FCCI occurs at nominal operation conditions in U-Zr and U-Pu-Zr fuels

beyond 10at.% burnup (Lanthanide – Fe interaction typically)

– Fuel Constituent Redistribution – an effect of phase transitions

• U, U-5Fs, and U-10Mo do not redistribute

• U-10Zr does redistributes where Zr migrates to the center of the fuel

• U-Pu-10Zr redistributes with Zr migrating to the central region and the

periphery

9

Annular / low

smear density

New alloys

Additives

New alloys

New concepts

Fs – 49.8Mo-38Ru-6Rh-4Pd-2Zr-0.2Nb

Fast Reactor Testing in ATR –

Recent AFC tests

◼ Testing Fast Reactor Fuels in a

thermal reactor

– Rodlet – capsule – Cd basket system

– Comparison between true fast

spectrum vs ATR irradiations: proper

temperature profile is created in ATR

irradiations

– This allows for the study of fuel

performance phenomena that are

primarily dependent upon temp. /

temp. gradient

Typical AFC-3 Rodlet cross section

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

2.2

2.4

0 0.2 0.4 0.6 0.8 1

Po

we

r fa

cto

r

r/r0

Unshrouded

SFR

Cd-shrouded

AFC-3 A/B/C/D and AFC-4A series

◼ AFC-3A/B/C/D and AFC-4A is an alloy exploration test– Alternate alloys and forms to U-10Zr,

Sodium Bonded, 75% SD

– Pd additive to mitigate FCCI

– Annular Forms to eliminate Na treatment issues

◼ Meant to test early alloy performance against the historical experience

◼ Irradiation Issues– Capsule Fabrication 3A/B

– Reactor Uncertainty

11

Rodlet ID AlloyFuel

Form

Bond

Material

Nominal

Smear

Density

3C-R1 U-10Mo Solid Sodium 75%

3C-R2 U-10Mo Annular Helium 55%

3C-R3 U-10Zr Sodium Solid 65%

3C-R4 U-10Zr Annular Helium 55%

3C-R5A U-1Pd-13Zr Solid Sodium 75%

3C-R5B U-2Pd-13Zr Solid Sodium 75%

3D-R1 U-10Zr Annular Helium 55%

3D-R2 U-4Pd-13Zr Solid Sodium 55%

3D-R3 U-10Mo Solid Sodium 55%

3D-R4 U-10Mo Annular Helium 55%

3D-R5 U-4Pd-13Zr Annular Helium 55%

4A-R1 U-10Mo Annular Helium 65%

4A-R3 U-5Mo-4.3Ti-0.7Zr Solid Sodium 75%

4A-R4U-5Mo-4.3Ti-0.7Zr-

2PdSolid Sodium 75%

4A-R5 U-10Zr Solid Sodium 75%

Rodlet ID AlloyFuel

Form

Bond

Material

Nominal

Smear

Density

3A-R1 U-10Mo Solid Sodium 75%

3A-R2 U-10Mo Annular Helium 55%

3A-R4 U-10Zr Annular Helium 55%

3A-R5A U-1Pd-10Zr Solid Sodium 75%

3A-R5B U-2Pd-10Zr Solid Sodium 75%

3B-R1 U-4Pd-10Zr Solid Sodium 55%

3B-R2 U-4Pd-10Zr Annular Helium 55%

3B-R4 U-10Mo Solid Sodium 55%

3B-R5 U-10Mo Solid Sodium 55%

AFC-3C/3D –Irradiation History

◼ AFC-3C had irradiation temperature

– Peak Inner Cladding Temperature (PICT) exceeded 600°C for 3 or 4 rodlets

– PICT in R1 and R5 likely exceeded 650°C

– Experiment removed from ATR early after 3 cycles

◼ AFC-3D had reasonable PICT temperatures and appears to have better performance

540

560

580

600

620

640

660

680

700

0 20 40 60 80 100 120

PIC

T (°

C)

EFPD

3C-1

3C-2

3C-3

3C-4

3C-5A

3C-5B

500

520

540

560

580

600

620

640

0 50 100 150 200

PIC

T 9

°C)

EFPD

3D-1

3D-2

3D-3

3D-4

3D-5

Rodlet

Diameter

0.230 inch

Capsule Inner

Diameter .234±.001 in

Most Capsules are

near 0.235 inches

AFC-3C/3D – NDE highlights

5825

5845

5865

5885

5905

5925

5945

0 20 40 60 80 100 120 140 160

Dia

met

er (

µm

)

Distance from top of rodlet (mm)

Average AFC-3C R1

AFC-3C R2 Average

AFC-3C R3 Average

AFC-3C R4 Average

Average AFC-3C R5

Capsule Inner WallDiameterAFC-3D R1 Average

AFC-3D R2 Average

AFC-3D R3 Average

Rod Outer Diameter Range

AFC-3C R4

AFC-3C R5

AF

C-3

C R

4

AF

C-3

C R

5

AF

C-3

C R

3

AF

C-3

C R

2

AF

C-3

C R

1

AF

C-3

D R

4

AF

C-3

D R

5

AF

C-3

D R

3

AF

C-3

D R

2

AF

C-3

D R

1

◼ Thermal neutron radiography

– Annular fuel pins (e.g. 3C-R4)

maintained their annuli

– Visible change in grey scales shows

variation in density

◼ Profilometry

– AFC-3C presents some degree of

diametrical strain (max 0.7% strain)

– AFC-3D no measurable change in

diameter

AFC 3C/D - Fission Product

Distribution

14

◼ Axial Gamma spectrometry

– RhRu-106 flat distribution

– Annular fuel: Cs migrates

towards cooler axial ends

– Ce-144 some axial migration

0

50000

100000

150000

200000

250000

300000

0 5 10 15

Rel

ativ

e A

ctiv

ity

Distance from bottom of Rodlet (cm)

Cs-137

RuRh-106

CePr-144

Mn-54

Eu-154

0

20000

40000

60000

80000

100000

120000

140000

0 2 4 6 8 10 12 14

Rel

ativ

e A

ctiv

ity

Distance from bottom of Rodlet (cm)

Cs-137

RuRh-106

CePr-144

Mn-54

AFC 3C/D - Fission Product

Distribution

15

◼ Gamma Emission Tomography reveals the transverse distribution of fission products

◼ Signals exist for most major types of fission products

AF

C-3

C R

3A

FC

-D R

1

Cs-137 Ru-106 CePr-144U-10Zr 65% SD, solid

U-10Zr 55% SD, annular

Fission Gas Release

16

◼ The EBR-II data suggests that FGR is 70±10% after 1.5x1021 fiss/cc (~4 at.%)

◼ Other AFC tests and transmutation fuel (FUTURIX) tests also follow this general

trend

◼ The AFC-3 tests often also follow this trend

– There is some evidence that low smear density, annular fuel, and U-Mo fuel release fission

gas earlier

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

0 5E+20 1E+21 1.5E+21 2E+21 2.5E+21 3E+21 3.5E+21 4E+21 4.5E+21 5E+21 5.5E+21 6E+21 6.5E+21 7E+21

FGR

Fission density

Fission Gas release vs Fission density

EBR-II

AFC-1

FUTURIX

AFC-3

AFC 3C / 3D U-Mo Metallography

1717

AFC-3C R1 (U-10Mo,

75% SD, solid, Na, 2.3%FIMA, 650°C)

AFC-3C R2 (U-10Mo,

55% SD, annular, He, 3.3%FIMA, 625°C)

AFC-3D R3 (U-10Mo,

55% SD, solid, Na, 2.1%FIMA , 625°C)

AFC-3D R4 (U-10Mo, 55%

SD, annular, He, 4.5%FIMA, 600°C)

AFC 3C / 3D U-Zr solid

Metallography

AFC-3C R3 (U-10Zr,

65% SD, solid, Na, 2.9%FIMA, 630°C)

AFC-3D R2 (U-4Pd-

13Zr, 55% SD, solid, Na, 2.8%FIMA, 630°C)

AFC-3C R5A (U-1Pd-

13Zr, 75% SD, solid, Na, 2.6%FIMA, 680°C)

AFC-3C R5B (U-2Pd-13Zr, 75% SD, solid, Na, 2.6%FIMA, 660°C)

Fuel Mid-plane Fuel Top

AFC 3C / 3D U-Zr annular

Metallography

19AFC-3D R1 (U-10Zr, 55% SD, annular, He, 4.3%FIMA, 600°C)

AFC-3C R4 (U-10Zr,

55% SD, annular, He, 3.0%FIMA, 620°C)

AFC-3D R5 (U-4Pd-

13Zr, 55% SD, annular, He, 2.9%FIMA, 610°C)

AFC 3C / 3D U-Zr annular

Metallography

◼ Two of the Annular

U-Zr variants

performed

exceptionally well

◼ Two contributing

factors

– Appropriate PICT at

all times during

irradiation

– Well machined fuel

slugs

20

U-10Zr prelim. comparison between

AFC-series

21

U-10Zr,

annular, 55% sd

AFC-3DAFC-3CAFC-3A R4*

◼ U-10Zr annular: AFC-3D behaves better compare to -3A-3C

– Machining, gap ~50µm AFC-3A R4 vs 17µm AFC-3C R4 177ºC vs 60ºC

– Periphery temperature maintained below the critical temperature for Zr migration (Beta phase)

*AFC-3A: U-10Zr, bu 3.2 at.%, PICT 530°C / U-1/2Pd-10Zr, bu 2.5 at.%, PICT 585°C

AFC-3B R2*

U-4Pd-10/13Zr,

annular, 55% sd

SEM preliminary results

◼ AFC-3A/B/C/D electron microscopy examination– Challenges to prepare samples (high dose)

– A more quantitative understanding of the different phases and feature is required• New additives work?

• FCCI

• Temp / gap effect

22

Rodlet ID AlloyFuel

Form

Bond

Material

Nominal

Smear

Density

3C-R1 U-10Mo Solid Sodium 75%

3C-R2 U-10Mo Annular Helium 55%

3C-R3 U-10Zr Sodium Solid 65%

3C-R4 U-10Zr Annular Helium 55%

3C-R5A U-1Pd-13Zr Solid Sodium 75%

3C-R5B U-2Pd-13Zr Solid Sodium 75%

3D-R1 U-10Zr Annular Helium 55%

3D-R2 U-4Pd-13Zr Solid Sodium 55%

3D-R3 U-10Mo Solid Sodium 55%

3D-R4 U-10Mo Annular Helium 55%

3D-R5 U-4Pd-13Zr Annular Helium 55%

Rodlet ID AlloyFuel

Form

Bond

Material

Nominal

Smear

Density

3A-R1 U-10Mo Solid Sodium 75%

3A-R2 U-10Mo Annular Helium 55%

3A-R4 U-10Zr Annular Helium 55%

3A-R5A U-1Pd-10Zr Solid Sodium 75%

3A-R5B U-2Pd-10Zr Solid Sodium 75%

3B-R1 U-4Pd-10Zr Solid Sodium 55%

3B-R2 U-4Pd-10Zr Annular Helium 55%

3B-R4 U-10Mo Solid Sodium 55%

3B-R5 U-10Mo Solid Sodium 55%

AFC-3A R5 SEM (U-2Pd-10Zr) –

Alloying Element Behavior

◼The behavior of Zr and Pd are important

for this fuel to well perform

– Zr needs to stayed alloyed with U for fuel

performance

– Zr and Pd intermetallics need to be avoided

– Pd needs to form intermetallics with

Lanthanides (LnPd and Ln7Pd3)

– Alloy addition steps are important

U

Zr

Pd

Nd

AFC-3B R2 SEM (U-4Pd-10Zr) –

FCCI – Alloying behavior

◼ FCCI occurred in this sample similar to the

same levels observed in U-Mo

◼ Too much Pd can be detrimental to fuel

performance

– Pd and Zr will form high temperature intermetallics

leaving U unalloyed (PdZr2)

– Pd alloying addition order can significantly change

as-cast intermetallic behavior (understood after this

irradiation was fabricated)

◼ Cast annular fuel without a Na bond is

susceptible to local overheating Cu tape

Cladding

Pd, Zr, Fe

U only

Pd, Zr, Fe

U only

U, Fe mix (U6Fe)

U, Fe mix (UFe2)

Zr band

Original ID

AFC-3A R4 SEM (U-10Zr annular) –

FCCI - Helium gap effect

25

BSE Zr

Fe UAFC-3A R4 (U-10Zr, 55% SD, annular, He, 3.3%FIMA, 540-600+°C)

Summary & Prospective

26

◼ Advanced metallic fuel form

– Raise from the needs to elevate utilization and reach higher burnup

– Improve historical issue of metallic fuel

◼ AFC-3 and -4 experiments

– Alloys exploration test (annular, alternative alloys, additives)

– Some promising design and value data are examined for the first time

– U-Mo performance are quite below U-Zr

– Annual fuel (well machined) and Pd additives fuel form are somehow promosing

◼ Future and Prospective

– Extended examination are needed (also supported by modelling and simulation)

– Extended burnup / experiment on some promising fuel form are needed

– Safety / transient test on promising fuel from are the natural next step

Irradiated Material Characterization

Laboratory (2019-2020)

27Come work with us!

Contributors

◼ Doug Porter

◼ Steve Hayes

◼ Heather Chichester

◼ HFEF Staff

– Colt Killian

– Katelyn Wachs

– John Stanek

– Dave Sell

– Glen Papaioannou

– Francine Rice

– HFEF Operators

◼ AL Staff

– Jeff Giglio

– Dan Cummings

– Mike Rodriquez

– Brian Storms

– Larry Fulkrod

◼ IMCL/EML Staff

– Alex Winston

– Nick Bolender

– Brandon Miller

– Tammy Trowbridge

– Jim Madden

– Karen Wright

◼ And others