Rebak GE EPRI AFC Florida 2015 02 17

Imagination at work

Ferritic Steels as Accident Tolerant Fuel Cladding for Commercial Light Water Reactors

FOURTH EPRI/INL/DOE JOINT WORKSHOP ON ACCIDENT TOLERANT FUEL

17-18 February 2015, St. Petersburg, Florida

Raul B. Rebak

GE Global Research, Schenectady, NY

GE team is/will be focused on cladding material, no fuel development

Rebak GE-DOE, EPRI AFC Workshop, 17-February-2015 2

Approach of GE Research

•Demonstrate that stainless iron based bulk alloys or Advanced Steels can be used as fuel cladding materials in commercial nuclear reactors

•The proposed material should be as good as Zr alloys (or better than Zr alloys) under normal operation conditions

1. Resistant to general corrosion and environmental cracking under normal operation conditions

2. Resistant to radiation damage

•The proposed Advanced Steels should be able to outperform the containment the fuel in the case of an accident scenario (e.g. LOCA) as compared to the current Zr alloys

1. Better mechanical strength at higher temperature

2. Enhanced retention of fission products (no cracking)

3. Improved reaction kinetics with steam

4. Lower generation of hydrogen gas when reacting with steam

Rebak GE-DOE, EPRI AFC Workshop, 17-February-2015

We are currently in Phase 1B


Technology and Manufacturing Readiness Levels for Advanced Steel Cladding

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TRL 5 – Component in a Representative Environment

TRL 6 - Prototype in a Representative Environment

TRL 7 - Prototype in an Operational Environment

TRL 8 – System Qualification

TRL 1 – Basic Principles Observed

TRL 2 - Concept Formulation

TRL 3 - Proof of Concept

TRL 4 – Component in the Laboratory

End of Phase 1A

2022

Lead Fuel Assembly


Approach to Study and Rate the Steels

Regulatory Analysis, Fuel Economy

Fabrication Capabilities

Environmental Degradation Under Accident Scenarios

Environmental Degradation Under Normal Operation Conditions

6 Rebak GE-DOE, EPRI AFC Workshop, 17-February-2015

Work to be Done in Phase 1B to Mature Feasibility Concepts

Follow Guidelines per INL/EXT-13-29957, FCRD-FUEL-2013-000264, LWR ATF Performance Metrics

1. Environmental Confirmatory Testing (e.g. Quench Test, Fretting, etc.)

2. Irradiation Testing at ATR

3. Evaluate Tube Fabrication and Manufacturability

4. Early Economical Feasibility for the new fuel concept

5. Preliminary Performance and Safety Assessment

• Evaluate Neutronics

• Thermal Hydraulics

• Scoping Analysis for Bounding Transients or Accidents


Materials Selected to Study

Alloy Nominal Composition

A Zirc-2 UNS R60802 Zr + 1.2-1.7 Sn + 0.07-0.2 Fe + 0.05-0.15 Cr + 0.03-0.08 Ni

B Ferritic steel T91 K90901

Fe + 9 Cr + 1 Mo + 0.2 V

C Ferritic steel HT9 S42100

Fe + 12 Cr + 1 Mo + 0.5 Ni + 0.5 W + 0.3 V

D Nano ferritic alloys -NFA

e.g. 14YWT; Fe + 14 Cr + 0.4 Ti + 3 W + 0.25 Y2O3

E MA956 or UNS S67956 Fe + 18.5-21.5 Cr + 3.75-5.75 Al + 0.2-0.6 Ti + 0.3-0.7 Y2O3

G APMT Fe + 22 Cr + 5 Al + 3 Mo

H Super ferritic, e.g. 4C54, Ebrite S44627

Fe + 25-27.5 Cr + 1 Mo + 0.17 (Ni + Cu)

J Alloy 33 – UNS R20033 33 Cr + 32 Fe + 31 Ni + 1.6 Mo + 0.6Cu + 0.4 N


Behavior of Candidate Steels in Simulated Normal Operation Conditions

Behavior of Candidate Steels in Simulated Normal Operation Conditions

1) Crack Propagation Studies

2) One Year Exposure to High

Temperature Water

3) Electrochemical Tests


Crack Propagation Studies

GE Rebak DOE Project, FCRD AFC Integration Meeting, 26-August-2014 11

Specimen ID

Material Condition Hours Tested

C642 Nano-ferritic As-rec’d 4,970 hours

C646 HT-9, irradiated 38.7 dpa 4,533 hours

C647 HT-9 21% cold forged 8,585 hours

C648 APMT 23% cold forged 10,500+ hours

C649 T91 23% cold forged 4,624 hours

C680 HT-9, irradiated 7 dpa 1,312 hours

C689 HT-9, irradiated 38.7 dpa 4,300+ hours

C690 9% Cr ferritic 30% cold forged 2,457 hours

C691 20% Cr ferritic 20% cold forged 69 hours

C692 Alloy 33 20% cold forged 3,300+ hours

C709 APMT As-received 700+ hours

C710 Nano-ferritic As-received 500+ hours

Compact Tension Specimen of HT9 neutron irradiated to 36.9 dpa and tested in 288°C Water + 2 ppm O2 up to a stress intensity of 38 ksi√in

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In chromium ferritic alloys a crack may propagate under cyclic loading (fatigue) but it stops growing as the load cycling frequency is decreased


Compact Tension Specimen of 23% cold worked APMT tested in 288°C Water + 2 ppm O2 up to a stress intensity of 40 ksi√in

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In chromium ferritic alloys a crack may propagate under cyclic loading (fatigue) but it stops growing as the load cycling frequency is decreased


1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

00.0020.0040.0060.0080.01

Cra

ck G

row

th R

ate

, mm

/s

Frequency, Hz

K=40 ksi/in, R=0.5

K=30 ksi/in, R=0.5

Crack Arrest as FrequencyIs Decreased in Ferritic Steels

c648 - 0.5CT of APMT, 23% CW

30 ksiin, 2 ppm O2, 30 ppb SO4

1-Year Exposure of Coupons to High Temperature Water

Simulated BWR, Normal Water Chemistry (2000 ppb O2), 288°C

1) Zirc-2

2) T91

3) HT9

4) 14YWT NFA

5) MA956

6) APMT

7) Ebrite

8) Alloy 33

GE Rebak DOE Project, FCRD AFC Integration Meeting, 26-August-2014 14

1-Year Exposure at 288°C Water + 2 ppm O2

15

• The rate of mass gain decreased at the exposure time increased

• The highest mass gain was for Zirc-2 coupons

• APMT offers some of the lowest mass gains as a function of time


366 days exposure to 288°C Water

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T91, RB2, X10,000 HT9, RC2, X10,000 NFA, RD2, X10,000

MA956, RE2, X10,000 APMT, RG2, X10,000 Alloy 33, RJ2, X10,000


1-Year Exposure at 288°C Water + 2 ppm O2

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• Stable corrosion potentials as a function of immersion time

• The lowest Ecorr was for Zirc-2

• All the iron containing alloys had a similar Ecorr, the lowest being for T91


Redox kinetics on APMT, NFA and Alloy 33 alloys behaves similarly to 304 SS or X-750 in high temperature water

Corrosion Potential of Zirc-2, 304SS, X-750, APMT, Alloy 33, NFA and Pt under different water chemistry conditions

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Tip: To navigate between the first level copy and the first level bullet, use the Indent and Outdent button on the Home Ribbon. Once you are on the second, and third level bullets, you can Tab and Shift-Tab to go back and forth between the bullet levels.


Under UV illumination the ECP shift of the advanced steels is in the opposite direction as the one on Zirc-2 suggesting that APMT will not suffer Shadow Corrosion

Corrosion Potential response of several materials under UV illumination in 288°C water containing 1 ppm O2

19

Tip: To navigate between the first level copy and the first level bullet, use the Indent and Outdent button on the Home Ribbon. Once you are on the second, and third level bullets, you can Tab and Shift-Tab to go back and forth between the bullet levels.


Reactions with Steam, Resistance under Accident Conditions

Reactions with Steam, tested at GE Research


Appearance after 24 h in 100% steam at 800°C

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X100, Zirc-2

X500, APMT

Negligible steam attack in the Fe-Cr-Al alloy Polishing marks still visible


Fabrication Evaluations

Current Sandvik Fabrication Status

Evaluate the cold forming properties of APMT tubing

by cold pilgering and cold drawing.

The target for the study is to

1. Understand the limitations when cold forming

APMT tubing

• Manufacture a APMT tube with as small dimension as possible, the aim is to reach 15 x 1 mm

• If the cold forming trials are successful a tube of 1 m will be delivered with the smallest dimensions achieved

Irradiation Tests at the INL-ATR

Fuel Properties Used in Neutronics Analysis

Development

Lead Fuel Type

Cladding

Type

Fuel OD

(in)

Total

Fuel

Height

(in)

Enriched

Fuel

Height (in)

Enrichment

(% U-235)

Density

(g/cc)

GE UO2 APMT /

Alloy 33 0.3235 4 3.2 4.90% 10.631

Images of Fuel Provided by INL

INL ATR Tests to demonstrate chemical compatibility between fuel and cladding

• 300-400°C Cladding temperature

• 300-400 W/cm LHGR • 20, 80 GWd/mT Burnup

• < 1700°C fuel centerline

temperature

Rodlets fabricated by INL

using material provided by GE Research

Fuel Fabricated by GE-GNF


Tests at ATR started on 11-02-2015

Storage of Spent (Used) Fuel

Prediction for 100 Years Storage in Dry Casks

Repository

5-10 years

40-100 years

10,000 years? 1,000,000 years?

Summary and Conclusions

Summary and Conclusions

1) The proposed advanced steels are better than zirconium alloys under normal operation conditions

• Resistant to general corrosion and environmental cracking

• Resistant to shadow corrosion

• Resistant to radiation damage

2) The advanced steels outperform the zirconium alloys under accident conditions

• Better mechanical strength at higher temperature

• Enhanced retention of fission products

• Improved reaction kinetics with steam

• Lower generation of hydrogen gas when reacting with steam

3) Other Areas • Irradiation studies at ATR (dry and water loop)

• Neutronics, thermal hydraulics

• Tritium Release

• Long term storage of spent fuel in dry casks

• Fabrication of tubing • Economics

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Acknowledgements This material is based upon work supported by the Dept. of Energy [National Nuclear Security Administration] under Award Number DE-NE0008221. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

GE Tasks and Deliverables for FY2015-2016

Task Who 1QFY15 2QFY15 3QFY15 4QFY15 1QFY16 2QFY16 3QFY16 4QFY16

Task 1.1: Envormental Degradation Under Normal Opeation Conditions

Resitance to Environmental Cracking GE

Resistance to Fretting GE

Task 1.2: Behavior of materials under accident conditions

Superheated Steam Tests up to 1000°C GE

Superheated Steam Tests > 1000°C ORNL

Quenching tests GE, ORNL

Task 1.3: Behavior of materials under proton irradiation conditions

Effect of irradiation on microstructure U.Mich

Effect of irradiation on environmental cracking susceptibility U.Mich

Task 1.4: ATR Test

PIE analysis after intermediate irradiation INL, LANL

Task 1 Milestone: are advanced steels environmentally suitable?

Task 2.1: Develop tube fabrication parameters GE, Sandvik

Task 2.2: Evaluate optimal cladding composition, burst tests ORNL

Task 2 Milestone: Can FeCrAl tubes be fabricated in small diameter with thin walls

Deliverable: A Fabrication Feasibility Report

Task 3.1: Themal Hydraulics and Neutronic Calculations BNL, ORNL

Task 3.2: Data preparation for LTR/LTA Plan for 2022 Insertion GE, BNL, ORNL

Deliverable: A Report on LTR/LTA Insertion Plan GE

Deliverable: A Report on ATF advanced steel cladding feasibility vs. against metrics GE

Deliverable: Technical Quarterly Reports GE

Rebak GE EPRI AFC Florida 2015 02 17

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