Okla Inc at NRC December 14, 2016

Outline of Meeting

• Okla introduction ("'1Smin)

• Brief introduction to EBR-11 and IFR ("'1Smin)

• ANL presentation on fast reactor metallic fuel experience set and database ("'30min)

• Time for questions - can also be interspersed ("'15min)

2

Okla Intro

• Founded in 2013

•Awards

• Funded in 2015

• Growing team

• Press - overview/intentionally limited

• Designing compact fast reactor utilizing metallic fuel

3

Intro to EBR-11

• EBR-11 was a 62.SMWth, 19 MWe sodium-cooled fast reactor with metallic fuel

• EBR-11: • operated for 30 years

• sold power to the grid

• had higher capacity factor than fleet at the time

Intro to EBR-11

• Years of quality assured testing done with the EBR~ll reactor

• Wealth of data on U-Zr fuel, various claddings

• Oklo currently working with ANL to compile relevant data

EBR-11 Shutdown Heat Removal Tests (SHRT)

• Performed on the same day (April 3rd, 1986)

• Two types of unprotected loss-of-cooling accidents • Loss of Flow Without Scram

• Loss of Heat Sink Without Scram

• Performed on the actual, operating reactor at full power!

•Started back up after both tests without damage

6

EBR-11 Loss of Flow Without Scram

•Loss of Flow Without Scram (LOFWS): Primary coolant 800 -u pumps turned off while L 750 PREDICTION

w 0 MEASURED DA TA

operating at full power ~ 700 t-

~ 650 •Reactor shut down due to fuel w

0.. ~ 600

thermal expansion feedbacks w t-t- 550 w

•No damage to fuel or ..J t- 500 :::> 0 0

otherwise 450 0

-100 0 100 200 300 400 500 TIME INTO TRANSIENT (s)

.:.· {~

... -. 7

EBR-11 LOFWS Plots 800

Clod -'° \400

/\

OAO 50 700 I 00

IO 40 ~ &)() \100 I

1100

20 : soo ! . ... ....... .

I\' ~00 I\ .....

..................................

- 200 O 200 .00 IOO 100 tOOO rime into i'ansi«lt. 200 0 tOOO

8

EBR-11 Loss of Heat Sink Without Scram • Loss of Heat Sink Without Scram

(LOHSWS): Intermediate coolant pumps turned off while operating at full power

• Again, reactor shuts down without scram due to thermal expansion feedbacks

• No damage to fuel or otherwise

500

-t 450 w 0:

~ 400 <( a:. w ~ 350 w t-

300

-- PREDICTIONS

MEASURED DAT A

REACTOR OUTLET

REACTOR INLET o---1.Y

-500 0 500 1000 1500 2000 2500 TIME INTO TRANSIENT (s)

9

EBR~ll Safety Test Takeaways

"These are sensational results. Two of the most severe accidents that can threaten nuclear power systems have been shown to be of no consequence to safety or even operation of EBR-11. The reactor was inherently protected without requiring emergency power, safety systems, or operator intervention."

-J.I. Sackett "OPERATING AND TEST EXPERIENCE WITH EBR-11, THE IFR PROTOTYPE", Progress in Nuclear Energy 31, 1-2, pp. 111-129, 1997. . .

() 10

Metallic Fuel Experience in Fast Reactors

Oklo Inc. Public Meeting with the NRC

Presented by: Abdellatif M. Yacout

Argonne National Laboratory

December 14, 2016

Presentation Overview

• Presentation Purpose

• Metallic Fuel Experience

• Steady State Performance

• Transient Behavior

• GAIN Program Activities with Oklo Inc

• Summary

~'. :~ 12 "":- :::"

Presentation Purpose

• Familiarize the NRC with the performance of metallic fuel

• Present current GAIN program related activity to support licensing of Okla design

Metallic Fuel Experience

14

Metallic Fuel History • Over ?O years of irradiation

experience

• EBR-1, Fermi-1, EBR-11, FFTF

• U-Fs* U-Mo U-Pu-Fs* I I I

U-Zr, U-Pu-Zr, others

• EBR-11 • > 40,000 U-Fs* pins, > 16,000 U-Zrpins &

> 600 U-Pu-Zr pins irradiated, clad in 316 stain less steel, 09 & HT9

• FFTF • > 1000 U-Zr pins, mostly in HT9

• Vast experience with HT9 cladding

* Fs - Simulated Fission Products

EBR-11

FFTF

15 ,

- -- · -- --- -------- - ------- -----------

Sources of Metallic Fuel Data

TREAT

Metal Fuel Experiments

FFTF

IFR Experiments

Metallic Fuel Data

EBR-11

IFR Experiments

1mmlll llllllm! ~~

t' 16 ,

Metallic Fuel Experimental Database (Steady State)

• EBR-11 experiments to look at parameters and phenomena of interest to fuel performance

• Prototype fuel behavior • RBCB* and failure mode • Fuel swelling and restructuring • Lead IFR** fuel test • Fabrication • Design parameters • High clad temperature • Large fuel diameter • Blanket safety • Fuel qualification • Fuel impurities

*RBCB - Run Beyond Cladding Breach ** !FR- Integral Fast Reactor

• FFTF experiments to look at

• Fuel column length effects

• Lead metal fuel tests

• Metal fuel prototype

• Metal fuel qualification

, 17

Metallic Fuel Experimental Database (Transient)

• In-Pile • Run Beyond Cladding

Breach (RBCB) experiments: 6 RBCB tests U-Fs & U-Pu-Zr/U-Zr

• 6 TREAT tests: U-Fs in 316SS& U-Zr/U-Pu-Zr in D9/HT9

• Out-Pile • Whole Pin Furnace Tests

(WPF)

• Fuel Behavior Test Apparatus (FBTA)

• Diffusion compatibility tests

18 ,

Typical Metallic Fuel Design

i--------74.9Jcin Clll(RAl.L-------1

0 .4125~--J1 2'cnl SOOll.IM u::"'3. 0 Qk"' 0V..

ROOM TDIP(Rll: run SL UC

0.101.,.,, °"" 1$.2 .. m PITCH

Typical EBR-11 Metallic Fuel Pin

(Pahl, et al., 1990)

Gu Fill --.....

Gap

-

"

ii J

• i.

~

Cl~ .. ThicknHS

~

...IQ ~c:> ... w

l 19 ,

Design Parameters (nominal) of EBR-11 Fuel - =

llcm Mark-IA Mark-II Mark-UC Mark-llCS Ma1k·lll Mark-DIA Mark-IV· - = -Fuel :tlloy, wl % u-srs U-Slls U-lOZr tl-IOZr U-10'/.r U-!OZr U-IOZr

Enrichment wcighl, % mu 52 67 78 78 66.9 66.9 69.6 -Fuel-slug mass, g 64 S2 47 47 83 83 78

Fuel smeared density, % SS 75 15 15 ·1s 15 15

Cladding·wall lbiclm~. cm 0.023 0.030 0.030 0.030 0.03S 0.038 0.046

Cladding-wall OD, cm 0.442 0.442 0.442 0.442 0.584 O.S84 O.S84 -Length, cm 46.0 61.2 63.0 S'.\.6 74.9 74.9 74.9

Cladding malcrial 304L 316 316 316 CWH9 CW316 HT9 -Spacer-wire diamclcr, cm 0.124 0.124 0.124 0.124 0.10'/ 0.107 0.107

.,.,.,.-=.--~~"-~

C. E. Lahm, et al., "Experience with Advanced Driver Fuels in EBR-11," 1993.

20

Historical Fuel Design Parameters

Key Parameter EBR-11/FFTF

Peak Burnup, 104MWd/t 5.0-20

Max. linear power, kW/m 33-50

Cladding hotspottemp., °C 650

Peak centerline temp., °C <700

Peak radial fuel temp. difference, °C 100-250

Cladding fast fluence, n/cm2 up to 4 x 1023

Cladding outer diameter, mm 4.4- 6.9

Cladding thickness, mm 0.38-0.56

Fuel slug diameter, mm 3.33- 4.98

Fuel length, m 0.3 (0.9 in FFTF)

Plenum/fuel volume ratio 0.84to 1.45

Fuel residence time, years 1-3

Smeared density, % 75

,...----------------------------- -- ---- -

Fuel Material of Interest

• Fuel Alloy: U-lOZr • High thermal conductivity

• Compatible with sodium coolant

• Sufficiently high melting temperature for safety considerations

• Large experimental database

• Used as driver fuel in EBR-11 where over 16,000 pins were irradiated (Mark-Ill & Mark-II IA)

• Qualified as driver fuel in both EBR-11 and FFTF

• Previous comments from NRC review of PRISM (NUREG-1368) fuel qualification plan were related to U-Pu-Zr fuel and U-Pu-MA-Zr fuel.

Steady State Metallic Fuel Performance

23

Steady State Performance Topics • Fission Gas Release (FGR)

• Fuel Swelling

• Constituent Redistribution and Zone Formation

• Fuel-Cladding Chemical Interaction (FCCI} & Rare Earth Migration

• Fuel-Cladding Mechanical Interaction (FCMI}

• Cladding Material Performance

24

Fission Gas Release (FGR)

• Insoluble fission gases, Xe and Kr, accumulate in fuel until inter-linkage of porosity at sufficient burnup leads to release of large fraction of gas.

• The fission gases accumulate in plenum region and constitute t he primary clad loading mechanism.

OIMOb 0 IHIP1t-10b A V•'9Ptt-IOb

2 • 6 e 10 12 1" 16 1e 20 Ptok &I'• .. 1'

FGR vs. Burn up (Hofman & Walter, 1994): U·Sfs slightly lower because of beneficial effect of Si inclusion

IO

20

10

0

VII-Ft o~. 0 IM'lt-Zr

0 o 20 40 eo ao 100 120 140 r..iv-11cr-1o111

FGR vs. Fuel Swelling (Hofman & Walter, 1994): Independent of metal-fuel type

ii Liquid Na~.J=.-:-:-:":"1~-:-:-:-~

25

Fuel Swelling

• Driven by nucleation and growth of immobile fission-gas bubbles

• Low fuel smeared density ("'75%) combined with high swelling rate allow rapid swelling to "'33 vol% at "'2 at.% burnup where inter-linkage of porosity results in large gas release fraction which decreases the driving force for continued swelling

10

~ 8 .5

! 6 e ii ;;; <( 4

2 0 U-10Zr Cl U~PU-Zr A U-19PU-Zr

2 4 6 8 10 12 14 16 18 20 Peak Bumup in %

EBR-11 fuel length increase in various metallic fuels as a function of burnup where closed symbols correspond to FFTF data (ANL-AFCl-211)

26

Constituent Redistribution & Zone Formation

• Fuel melting temp. decrease in Zr-depleted region (th is zone happens off the fuel center) .

• Local fission rate change.

• Changes in swelling characteristics .

• Reliable predictive model has been developed.

900

v 7'°

! 700

i 650

eoo

$50

0 s 10

0 2 s

20 30

-10 IS 20 25 JO

ZlltC()NUj, w /o

eo

.,.

• O

Metallographic cross section with superimposed radia l microprobe scans at top of U-lOZr pin DP-81, experimentX447 (Hofman, et al ., 1995)

70

U-Zr Phase Diagram

JI# - 27 ,

Fuel Cladding Chemical Interaction (FCCI) & Fiss ion Product Migrat ion

• At steady state FCCI is characterized by solid state interdiffusion

• lnterdiffusion forms U/Fe alloys with lower eutectic temperature

• Decarburized zone at fuel-clad interface is expected in HT-9 cladding

• RE fission products (La, Ce, Pr, Nd) form a cladding brittle layer

• Penetration depth data are available from in and out-of-pile measurements

Fuel Clod<l<ng

FCCI of U-lOZr/HT-9 due to inter­diffusion of fuel/cladding constituents after 6 at% burnup at 620°C (Hofman & Walter 1994)

28 ,

Fuel Cladding Mechanical Interaction (FCMI}

• Due to accumulation of solid fission products at high burn ups

• Low fuel smeared density ("'75%) allows for a large amount of fuel swelling ("'30%) and gas release which reduces FCMI

• Porous fuel at low burnups (<5 at.%) does not exert significant force on cladding

~ 29 ,

Design Parameters (nominal) of EBR-11 Fuel

= Item Mark-IA Mark-n Mark-HC M.trk-IIC£ Mark-Ill Mark-fl IA -

Fuel. lloy, wt % U-Sfos U-SPs U-IOZr ll · IOZr U-10'/.c U-IOZr

Enrichment wcighl, ~ " 1U 52 67 78 78 66.9 (i6.9 --Fuel-slug ma.u, g 64 52 47 47 83 83

Fuel smeared dcnsit y, ~ 85 75 75 75 '15 75

Cladding·wall tbkkncu, cm 0.023 O.o30 0.030 0.030 0.0311 O.o38

Cladding-wall OD, cm 0.442 0.442 0.442 0.442 0.584 0.584

Lcn&th, cm 46.0 61.2 63.0 S:l.6 7'1 .9 74.9

Cladiling n~lcrial 304L 316 316 316 cw 1>9 CW316 - -Sp: er wire di:1111e1cr, c111 0.124 0.124 0.124 0.124 0.107 0.107 -C. E. Lahm, et al., "Experience with Advanced Driver Fuels in EBR-11," 1993:

Mark-IV

U-IOZr

69.6

78

75

0.046

0.584

74.9

HT9

0.107

30 ,

Cladding Material Performance • Experience with different types of steel

cladding: 304, 316, CW-316, 09, & HT9. r:f 1

• CW-316 steel (Mark-lllA} and 09 (Mark-Ill} with UlOZr were qualified in EBR-11 (over 16000) to 10at% BU.

• EBR-11 core was converted to Mark-Ill and Mark-lllA fuel during IFR program.

• Limited failure of Mark-lll&lllA fuel during operations (5 pins, where most failures attributed to fabrication defects that were corrected later)

• HT9 ferritic-Martensitic alloy with adequate swelling resistance, toughness, strength, and ductility.

• lOOs of HT9 clad pins were irradiated in EBR-11 & 1000s were irradiated in FFTF

Sl 6 ~

0 2 4 6 8 10 12 14 16 18 20 Peok Burn up in at.%

Void swelling behavior of HT9 cladding compared to other steels (Pahl, et al., 1992).

Run Beyond Cladding Breach (RBCB)

Thinned cladding

• EBR-11 RBCB experiments

• An area of cladding was machined down to 25-50 µm (<10% of cladding thickness is left).

• After a short period of irradiation, cladding failure occurred at the machined spot.

• Test Results

• No chemical interactions between fuel and coolant. U-19Pu-10Zr HT-9

• No fuel loss into coolant (under normal conditions) . Metallographic cross-section through cladding breach of a

• No significant liquid or solid FP escape. metallic fuel element after

• Only release of fission gas and Cs retained in the bond sodium was detected.

RBCB operation (Batte' and Hofman, 1990}

32 ,,

Transient Behavior

33

Metallic Fuel Characteristics

• Excellent transient capabilities • Does not impose restrictions on transient operations capabilities

• Sample history of a typical driver fuel irradiated during the EBR-11 inherent passive safety tests conducted in 1986;

• 40

• 5

• 3

• 45

start-ups and shutdowns

15% overpower transients

60% overpower transients

Joss-of-flow {LOF} and Joss-of-heat-sink tests including a LOF test from 100% without scram

No fuel failures 800

---~~....._,.~.,................,.........~

- 200 0 200 400 600 900 1000

Time nto Transient, s

Unprotected loss-of-flow test in EBR-11 demonstrated the benign behavior predicted (Mohr, et al., 1987)

34 ,

Transient Tests • In -pile TREAT (Transient Reactor Test

Facility) tests evaluated transient overpower margin to failure, pre­failure axial fuel expansion, and post­failure fuel and coolant behavior

• Hot cell furnace testing of pin segments (Fuel Pin Test Apparatus), and full length pins (Whole Pin Furnace) showed significant safety margin for particulartransient conditions.

- Penetration depth data were measured and provided the basis for penetration depth correlations

i j 1.e -01

i I r J 1 •.e<l2

i

..

. ! ., s.: • . .. . .

• • • • •• • . . •

• • .

tO M tOO 1H H O

lftWnie Tfflpef9'turt. 1tr(K}J;11•5

Effective cladding penetration rates from FBTA tests for speciments tested for 1.0 hour (Tsai, et al., 2007)

35

•

•

•

•

Eutectic Formation Temperat ure between Fuel and Clad

Critical parameter for metal fue l design

Onset of eutectic formation occurs between 650 - 725 °C

Rapid eutectic penetration at a much higher temperatures

Places lim its on the coolant outlet temperature to provide adequate margin to onset of eutectic fo rmation

-lll ·- • -... ~ r· i' ... _

-I-' --.. ... .. --Fe

L

" m

llO '° "° .. .. The Iron-Uranium Phase Diagram (Okamoto, 1990)

u

- ---- -------1

.!' 36 ti'

i

GAIN Program Collaboration

37

....------------------------------------------------------------~

GAIN (Gateway for Accelerated Innovation in Nuclear) Voucher Program

• Project: Legacy Metal Fuel Data Exploration for Commercial Scale-Up

• Objective: Support Oklo Inc. activities related to design and licensing of small portable fast reactor system for remote location deployment.

• ANL Scope: • Assess Oklo current fuel design and data needs, and identify subset of metallic fuel data

relevant to design & licensing activities

• Establish a platform for Okla access to the information

• Identify key design and licensing related information needed by a commercial entity to approach regulatory agencies.

• Oklo Fuel Design: • U-lOZrfuel alloy with steel cladding, low burnup, and fuel/cladding temperature within

envelope of existing database

Summary

39

Summary

•Over 30 years of in-reactor experience with metallic fuel irradiation.

• Extended database of metallic fuel behavior is available, with both excellent steady state and transient fuel behavior.

• Qu~lified U-lOZr fuel with steel cladding in both EBR-11 and FFTF

. •GAIN program initiated collaboration between national labs and Okla Inc to support licensing of their design

• Okla's operational fuel design parameters lie within the existing envelope of operation of the qualified Mark-Ill and Mark-lllA fuel (U-lOZr with 20% CW-316 steel and SS cladding).

·" f 40 •;;,: ~-

----------------~------------- -------- . .. -- - I

Conclusion of Meeting And Time For Questions

"' '~ 41