Overview of International Activities in

Accident Tolerant Fuel Development

for Light Water Reactors

Shannon Bragg-Sitton, Ph.D.

Idaho National Laboratory, United States

April 24, 2014

IAEA Technical Working Group on Fuel Performance and Technology

Presentation Overview

Definition of Accident Tolerant Fuel

Collaborative International activities

– OECD / NEA Expert Group on ATF

– Overview of countries participating in ATF research

Country snapshots:

– United States

– France

– Japan

– Republic of Korea

– People’s Republic of China

– Russian Federation

– Sweden

– Switzerland (analysis only)

– Belgium (proposal stage)

ATF R&D support activities

Upcoming international activities

2

LWR Enhanced Accident Tolerant Fuel:

Should Tolerate Loss of Active Cooling

for A Significant Period of Time

3

High

temperature

during loss of

active cooling

Improved Cladding Properties

- Clad fracture

- Geometric stability

- Thermal shock resistance

- Melting of the cladding

Improved Fuel Properties

- Lower operating temperatures

- Clad internal oxidation

- Fuel relocation / dispersion

- Fuel melting

Enhanced Retention of Fission Products

- Gaseous fission products

- Solid/liquid fission products

Improved Reaction Kinetics

with Steam and Slower H2 Generation

- Heat of oxidation

- Oxidation rate

- Hydrogen production

- Hydrogen embrittlement of the cladding

Fuels with enhanced accident tolerance are those that, in comparison with the standard

UO2 – Zr system, can tolerate loss of active cooling in the core for a considerably longer

time period (depending on the LWR system and accident scenario) while maintaining or

improving the fuel performance during normal operations.

New ATF Designs Must Meet the

LWR Operations, Safety and Fuel

Cycle Constraints

ATF will be evaluated over

all potential “performance

regimes”

Fabrication /

Manufacturability (to include

Licensibility)

Normal operations and

anticipated operational

occurrences (AOOs)

Postulated accidents

(Design Basis)

Severe accidents (Beyond

Design Basis)

Used fuel storage / transport

/ disposition

(to include potential for

future reprocessing)

See: U.S. DOE LWR ATF

Performance Metrics Report,

February 2014 4

Advanced Fuel Design, Operations and Safety Envelope

Backward Compatible

(qualified in existing reactor; operable with co-

resident fuel)

Operations

Safety (spectrum of DBAs and

possible BDBAs)

Fuel Cycle

Economics

High payoff technologies may require

additional development and time

5

High Fission

Product Retention

5 15 20

Time to Deployment

Perf

orm

an

ce

Mid-Term Technologies

Near-Term Technologies

10

Cladding

Coatings

Thin walled high

strength steel

alloy cladding

Ceramic Claddings

GEN 2 GEN 3 and 3+

High Density Fuels

(U2Si3, UN, etc.)

High Performance

UO2

Molybdenum

Claddings

Advanced steels (e.g. FeCrAl)

Refractory metals (e.g. Mo)

Ceramic cladding (SiC)

Innovative alloys with dopants

Zircaloy with coating or sleeve

• SiC CMC

• MAX-phase ceramics

• Other

Each concept has some pros and cons across the spectrum of operating and

transient conditions of interest. A systematic analytical and experimental

evaluation is being performed during the feasibility studies.

ATF cladding development efforts focus

on materials with more benign steam

reaction

6

Comparison of MELCOR predicted

cladding oxidation heating produced

during a TMI-2 accident sequence.

Higher density fuels

(metal, nitride, silicide)

• Higher thermal conductivity

• Higher fissile density to compensate for

neutronic inefficiency of some new

cladding concepts without increasing

enrichment limits

Oxide fuels with additives

• Higher thermal conductivity

• Fission product gettering

Microencapsulated fuels

• Particle fuel dispersed in a ceramic or

metallic matrix

Each concept has pros and cons across the spectrum of operating and transient

conditions of interest. A systematic analytical and experimental evaluation is being

performed during the feasibility studies.

Several advanced fuel concepts

are being considered

7

Uranium Silicide

Fabrication

Highest achieved

density:

94.6% theoretical

(11.5 g/cc)

INTERNATIONAL ATF R&D

ACTIVITIES

8

Coordination: OECD/NEA Expert Group on ATF

9

Expert Group on Increased Accident

Tolerance of Fuels for LWRs

10

Summary of ATF R&D Activities:

Cladding

Country Cladding

FeCrAl

Stainless Steels SiC* Advd.

Zr-alloys

Coated

Zr-alloys

U.S. x x x x x

France x x x

Japan x x

Rep. of Korea x x x x

P.R. of China x x x

Russia x

Sweden **see U.S. DOE-sponsored work led by Westinghouse

Switzerland analysis only

Belgium Proposal stage Proposal stage

11

*Non-fuel rod accident tolerant SiC/SiC activities:

BWR channel box – U.S., Japan

AT control rod (ATCR) – Japan

**Work in Sweden is via the Westinghouse research,

based out of the U.S. Research groups in the U.K. and

Australia also participate in U.S. University-led research.

***Bulgaria is actively conducting LOCA analysis on

VVER-1000 type fuel – no residual plastic deformation of

cladding and no fuel failure found.

****Brazil (INB) following all ATF development and will

adopt if/when licensed.

Summary of ATF R&D Activities:

Fuel

Country Fuel

Enhanced

UO2

Micro-

encapsulated

Uranium

Silicide

Uranium

Nitride

Ceramic

Composites

Metallic

U.S. x x x x x x

France

Japan x

Rep. of Korea x x

P.R. of China x x x

Russia

Sweden **see U.S. DOE-sponsored work led by Westinghouse

Switzerland

Belgium Proposal stage Proposal stage

12

**Work in Sweden is via the Westinghouse research,

based out of the U.S. Research groups in the U.K. and

Australia also participate in U.S. University-led research.

***Bulgaria is actively conducting LOCA analysis on

VVER-1000 type fuel – no residual plastic deformation of

cladding and no fuel failure found.

****Brazil (INB) following all ATF development and will

adopt if/when licensed.

U.S. DOE Advanced Fuels Campaign –

LWR Advanced Fuel Development

Activities beginning in 2010

O N D J F M A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S D J F M A M J J A S

FY2010 FY2011 FY2012 FY2013 FY2014

• MAX-Phase or other coatings on Zircaloy

• SiC cladding • FCM fuels for long-term • UN – focus on solving water

compatibility

Innovative fuels call for the national labs

FUKUSHIMA Events (Mar 11, 2011)

Working Group: Topic changed to enhanced accident tolerant fuels (Mar 16, 2011)

Senate language on accident tolerant fuels Actions: - Initiated ATF at DOE Labs - Initiated ATF NEUP projects - Issued FOA for Industry led

ATF projects - Issued call for University led

ATF IRPs

3 - Industry Accident Tolerant Fuel FOAs and 3 - University IRPs Initiated

Attributes/Metrics and R&D coordination and collaboration

meetings

ATF-1 Irradiation Initiated in INL Advanced Test Reactor (July 2014)

U.S. DOE Advanced Fuels Campaign –

LWR Advanced Fuel Development

Activities beginning in 2010

O N D J F M A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S O N D J F M A M J J A S D J F M A M J J A S

FY2010 FY2011 FY2012 FY2013 FY2014

• MAX-Phase or other coatings on Zircaloy

• SiC cladding • FCM fuels for long-term • UN – focus on solving water

compatibility

Innovative fuels call for the national labs

FUKUSHIMA Events (Mar 11, 2011)

Working Group: Topic changed to enhanced accident tolerant fuels (Mar 16, 2011)

Senate language on accident tolerant fuels Actions: - Initiated ATF at DOE Labs - Initiated ATF NEUP projects - Issued FOA for Industry led

ATF projects - Issued call for University led

ATF IRPs

3 - Industry Accident Tolerant Fuel FOAs and 3 - University IRPs Initiated

Attributes/Metrics and R&D coordination and collaboration

meetings

ATF-1 Irradiation Initiated in INL Advanced Test Reactor (July 2014)

2012

Feasibility studies on advanced fuel

and clad concepts

-- bench-scale fabrication

-- irradiation tests

-- steam reactions

-- mechanical properties

-- furnace tests

-- modeling

Workshops

2013 2014 2015 2016 2017 2018 2019 2020 2021

Assessment of new concepts

-- impact on economics

-- impact on fuel cycle

-- impact on operations

-- impact on safety envelope

-- environmental impact

Fuel Downselection

Steady State Tests

Transient Irradiation Tests

LOCA/Furnace Tests

Fuel Performance Code

Fuel Safety Basis

LTA/LTR Ready

Phase 1

Feasibility

Phase 2

Development/Qualification

Phase 3

Commercialization

2022

U.S. RD&D Strategy For Enhanced

Accident Tolerant Fuels – 10 Year Goal

2012 2013 2014 2015 2016

Roadmap

National workshop on attributes and metrics

International workshop on attributes and metrics

Selection of candidates for feasibility assessment

Sample Fabrication and Characterization

Oxidation Testing

High-temperature Furnace Testing

Irradiation Testing of Samples

Cladding Mechanical Properties Testing

Performance Code Upgrades Based on ATF Properties

Reactor Core, Safety, Fuel Cycle, and Economic Analyses

Technology down-

selection for development /

eventual qualification CD-1 for Transient Testing

Data Reduction Yearly reviews and reorientation

U.S. Near-term ATF Activities Aimed at

Completing the Feasibility Assessment

The U.S. Accident Tolerant Fuel

development is supported by a large

part of the U.S. nuclear complex

National Laboratories Universities Nuclear Industry

Summary of Major U.S. DOE

Funded ATF Projects

Lead Organization Category – Major Technology Area Additional Team Members

Oak Ridge National

Laboratory

Fuel: Fully Ceramic Microencapsulated

(FCM)-UO2, UN

FeCrAl cladding, SiC cladding

Los Alamos Nat. Lab.

EPRI + LANL

Fuel: Enhanced UO2, Composite Fuels

Cladding: Molybdenum

AREVA

(FOA, NEUP)

High conductivity fuel (UO2+Cr2O3, +SiC)

Cladding: Protective materials, MAX phase

U. Wisconsin, U. Florida, SRNL,TVA,

Duke

**Includes work with AREVA France

Westinghouse

(FOA, NEUP)

Fuel: U3Si2, and UN+U3Si2 fuel

Cladding: Coated Zr and SiC

General Atomics, MIT, EWI,

INL, LANL,TAMU, Southern Nuclear

Operating Company

**Includes work with WEC Sweden

GE Global Research

(FOA)

Advanced Steel (Ferritic / Martensitic)

Cladding

Global Nuclear Fuels, LANL,

U. Michigan

University of Illinois

(IRP) Modified Zr-based Cladding (coating or

modification of bulk cladding composition)

U. Michigan, U. Florida, INL,

U. Manchester, ATI Wah Chang

**UK contributions

University of Tennessee

(IRP) Ceramic Coatings for Cladding

(MAX phase and multilayer ceramic

coatings)

Penn State, U. Michigan, UC Boulder, LANL,

Westinghouse, Oxford, U. Manchester,

U. Sheffield, U. Huddersfield, ANSTO

**UK and Australia contributions

Georgia Institute

of Technology

(IRP, accident tolerant

reactor design)

Fuel: U3Si2,

Cladding: FeCrAl Details TBD for fuel component – likely to adopt

fuel developed under an above program

U. Michigan, Virginia Tech, U. Tennessee,

U. Idaho, Morehouse College, INL,

Westinghouse Electric, Southern Nuclear,

Polytechnic U. Milan, U. Cambridge

~M

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Initial irradiation of all

concepts at Idaho

National Laboratory

Summary of Major U.S. DOE

Funded ATF Projects

Lead Organization Category – Major Technology Area Additional Team Members

Oak Ridge National

Laboratory

Fuel: Fully Ceramic Microencapsulated

(FCM)-UO2, UN

FeCrAl cladding, SiC cladding

Los Alamos Nat. Lab.

EPRI + LANL

Fuel: Enhanced UO2, Composite Fuels

Cladding: Molybdenum

AREVA

(FOA, NEUP)

High conductivity fuel (UO2+Cr2O3, +SiC)

Cladding: Protective materials, MAX phase

U. Wisconsin, U. Florida, SRNL,TVA,

Duke

**Includes work with AREVA France

Westinghouse

(FOA, NEUP)

Fuel: U3Si2, and UN+U3Si2 fuel

Cladding: Coated Zr and SiC

General Atomics, MIT, EWI,

INL, LANL,TAMU, Southern Nuclear

Operating Company

**Includes work with WEC Sweden

GE Global Research

(FOA)

Advanced Steel (Ferritic / Martensitic)

Cladding

Global Nuclear Fuels, LANL,

U. Michigan

University of Illinois

(IRP) Modified Zr-based Cladding (coating or

modification of bulk cladding composition)

U. Michigan, U. Florida, INL,

U. Manchester, ATI Wah Chang

**UK contributions

University of Tennessee

(IRP) Ceramic Coatings for Cladding

(MAX phase and multilayer ceramic

coatings)

Penn State, U. Michigan, UC Boulder, LANL,

Westinghouse, Oxford, U. Manchester,

U. Sheffield, U. Huddersfield, ANSTO

**UK and Australia contributions

Georgia Institute

of Technology

(IRP, accident tolerant

reactor design)

Fuel: U3Si2,

Cladding: FeCrAl Details TBD for fuel component – likely to adopt

fuel developed under an above program

U. Michigan, Virginia Tech, U. Tennessee,

U. Idaho, Morehouse College, INL,

Westinghouse Electric, Southern Nuclear,

Polytechnic U. Milan, U. Cambridge

~M

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et

Initial irradiation of all

concepts at Idaho

National Laboratory

Summary of U.S. ATF

research provided in the

March 2014 edition of

Nuclear News, the monthly

publication from the

American Nuclear Society. file:///Users/bragsm/Downloads/nn_2014_3_2.pdf

Advanced Fuels – U.S. DOE-Supported

Industry Teams

AREVA

Develop coated Zr-alloy

cladding for improved

accident performance

Increased fuel pellet

conductivity: Fuel with

reduced stored energy that

must be accommodated

during DBE

Additives achieved:

– SiC powder or whiskers

– Diamond

– Chromia

dopant

GE

Develop advanced

ferritic/martensitic steel

alloys (e.g., Fe-Cr-Al) for

fuel cladding to improve

behavior under severe

accident scenarios

Objectives:

– Characterize candidate steels

– Study tube fabrication

methods, neutronics, fuel

economy, thermo-hydraulic

calculations, regulatory

approval path

– Initiate ATR testing with UO2

and two cladding materials.

20

Westinghouse

Develop and test cladding

concepts: SiC and SiC ceramic

matrix composites; coated Zr

alloys

High density/high thermal

conductivity fuel pellets (e.g.,

uranium nitride/silicides)

First batch of U3Si2 pellets

were sintered using finely

ground powder

Pellets were pressed using

pressures of 6,000-10,000 psi

and sintered at temperatures

of 1400°C

U.S. Program has Significant International

Collaboration in Advanced Fuels – Direct ATF

Collaborations Highlighted

21

France (March 11-13, Marcoule/Cadarache)

• FUTURIX-FTA shipment

• ATF

• Fuel Performance Code Comparison

• Joint Am irradiations in ATR

• Cladding materials

• Trilateral transient testing

Japan (CNWG Bilateral)

• Metallic fuel

• Oxidation kinetics

• ATF

South Korea (KAERI)

• Metal fuel fabrication technology and

irradiation performance

China (March 18 – 20, 2013)

• Materials and fuels irradiation in

CFTR

• Metal fuel fundamental properties and

fabrication

• ATF

Russian Federation

• Materials and fuels irradiation in BOR60

• Characterization and PIE methods

• In-pile instrumentation and testing

• Advanced LWR fuels and ATF

• Nitrides

Others

• Active participation in OECD/NEA,

IAEA, GENIV Advanced Fuel and

GACID working groups and projects

European Union (April 25, INL)

• Participation in European materials

program

• Characterization & PIE techniques

2013 Accident Tolerant LWR Fuel

Development: Major Accomplishments

Conducted Domestic and

International Metrics workshops

Industry teams are well underway

Oxidation and Steam testing

capabilities have been built and

deployed across the laboratories

FeCrAl, Mo, SiC cladding

development and testing

High density ceramic fuel

Planning and Design of ATF-1

irradiation experiment

22

FCM Fuel

U3Si2

ATF Irradiation Testing and Qualification

Test Series (Draft)

23

Test Series ATF-1 ATF-2 ATF-3 CM-ATF-x ATF-4

Test Reactor ATR ATR TREAT Commercial

Power Plant TREAT

Test Type Drop-in Loop Loop LTR/LTA Loop

Test Strategy

Scoping — Scoping — Focused

Compositions

Focused

Composition

Focused

Compositions Many Compositions Focused Compositions

Nominal

conditions

Nominal

conditions

Accident

conditions

Nominal

conditions

Accident

conditions

Fuel UO2, U3Si2,

UN

Down-selected

concepts

Fuel rodlets

from ATF-1

and test rods

from ATF-2

irradiations

Concepts

selected in

2016

Test rods from

LTR/LTA

irradiations Cladding

Zr w/coatings,

stainless steels,

advanced

alloys, SiC

Key Features Fuel-cladding

interactions

Fuel-cladding-

coolant

interactions

Integral testing Steady State

Irradiation Integral testing

Timeframe FY14 —

FY18+ FY16 — FY22 FY18 — FY25 FY2022 - ? FY26 — ?

ATF R&D in France – Focus on Cladding

Research on 2 attractive ATF cladding concepts:

– Coated Zr based Cladding In service, DBA and Post DBA benefits evaluation

– Sandwich SiC-SiC with liner (GFR) Evaluation of LWR applicability and benefits

Cladding characterizations and evaluation programs are now undertaken with the

support of the French Industrial partners AREVA and EDF

Targets : (Common to all ATF programs)

Decrease oxidation rate in service and in steam environment at high temperature

Reduce in service hydriding and Hydrogen release during accident sequence

Improve LOCA behavior (Peak clad T , post quench ductility )

Improve long term coolability - procure grace time in the early stage of the severe accident

Realistic approach based on three primary axes:

– Material development and innovation (CEA + collaborations, GFR)

• Long tubes, hermetic end joints

– Three Parties Institute (CEA-EDF-AREVA partnership)

• Characterization testing, experiments, numerical simulation; corrosion, erosion, LOCA

behavior, irradiation

– Involvement in international programs

• OECD/NEA; Halden Reactor Program (proposal for coated rodlet irradiation); proposal in

framework of Nugenia WP5 for Horizon 2020 call 24

ATF R&D in France – Coated Zr Cladding

COATED Zr Alloy

• Experiments made on the

most promising coatings

• 1100°C 850 s., 1200°C 300 s.

in steam water + quench

• Weight gain, microstructure

observations, Post-quench

flexural/tensile tests

25

Optimized Cr coated cladding appears as a candidate for safety improvement in current

reactors in a near future

Weight gain has been decreased

from ~10 mg/cm2 to ~0,5 mg/cm2

Zy4 First coatings Improved coatings Optimized coatings

[Idarraga, Lomello, Billard et al. 2012, unpublished results]

The optimized coating exhibits a very good resistance to both oxidation and hydrogen up-take:

(i) absence of significant zirconia formation nor oxygen diffusion into the metallic prior-βZr

substrate

(ii) no significant hydrogen pick-up (<80 wt.ppm)

(iii) Minimal weight gain (2.5 mg/cm2)

For the uncoated Zry-4, the weight gain (37-40 mg/cm2) and the associated hydrogen pick-up

(2000-3000 ppm) are fully consistent with the previous results obtained on low-tin Zry-4

cladding tubes

Results for “Slightly Beyond LOCA” Conditions (“Breakaway”, 4 hours at 1000oC)

Oxidation Testing (SiC/SiC samples, nominal conditions)

– First oxidation tests of SiC/SiC in LWR nominal conditions for

3500h are promising; oxidation mechanism must be clearly

understood

– Pyrocarbon interphase is not affected : no reduction of mechanical

strength

– SiC recession is estimated to be 0,5mm after 3500h of LWR

nominal oxidation conditions

Sandwich SiC/SiC cladding must be suitable for LWR

Short / Mid-term Prospects

– Numerical simulation of rod behavior in normal operation

(PCI, design)

– Fuel-SiC interface issues

– Corrosion /wear in nominal conditions (sandwich concept)

– High temperature oxidation (first results promising)

– Irradiation testing

International Collaborations (benchmarking and

evaluation)

– SiC-SiC sandwich with Ta liner is a candidate for experimental

comparison and numerical approach evaluations

26

ATF R&D in France – SiC/SiC Sandwich

Concept

Internal tube SiC/SiC: e~0.3mm

liner Ta : e<0.1mm

External tube SiC/SiC: e~0.6mm

CEA sandwich cladding

Patents

ATF Development in Japan

1.1 Toshiba, Kyoto Univ. and Hokkaido Univ. Program

1. Two projects for basic study and development of SiC composite

supported by MEXT* in 2012-2016

*Ministry of Education, Culture, Sports, Science and Technology

1.2 Muroran Institute of Technology (SCARLET Program)

- Fabrication of SiC samples by different processes

- Auto-clave corrosion test and high temperature steam oxidation test

- In-situ TEM observation of irradiation defects

- Development of NITE** process for SiC composite cladding

- Development of fuel rod fabrication technology with SiC composite cladding

- Irradiation test of SiC composite cladding samples in test reactors

- Test of SiC composite cladding samples under LOCA condition

**NITE (Nano-Infiltration Transient-Eutectic Phase) Process

2. Three projects for development of ATF supported by METI***

in 2012-2018

***Ministry of Economy, Trade and Industry

2.1 Toshiba, IBIDEN, NFI, Univ. of Tokyo and Tohoku Univ. Program

2.3 Kyoto Univ. as the representative of a team consisting of the

universities and industries program

- Development and testing of SiC composite channel box and cladding

fabricated by CVD/CVI process, in 2012-2014

2.2 Muroran Institute of Technology (INSPIRE Program) - Irradiation test for fuel rods with SiC composite cladding fabricated by NITE

process, in 2012-2016

- Irradiation test of SiC composite sample

- Development of stainless steel cladding

- Development of pebble bed fuel for LWR

- Test simulating severe accident condition

in 2014-2018

ATF Development in Japan

29

ATF Development in the Republic of Korea

New fuel cladding concepts

– Surface modification of Zr-alloy tubes

– Metal/ceramic hybrid cladding

– SiC triplex cladding

– FeCrAl/Zr duplex cladding

Advanced fuel for fission product retention

– Micro-cell UO2 pellet

Fully ceramic micro-encapsulated (FCM) fuel for LWRs

30

Coated Zr-alloy tubes

– Plasma spray of Cr on

Zr tubes

– LBS with Cr on Zr tubes

FeCrAl/Zr duplex cladding

– FeCrAl has excellent formability,

high strength at high T, high T

oxidation resistance

– BUT relatively low Tmelt, high n

capture cross section

31

ATF Development in the ROK – Metallic Cladding

Cr layer

Prior b-Zr

a-Zr(O)

ZrO2

Coated tube after HT high

temperature steam oxidation

at 1200oC for 2000 sec.

Hybrid cladding – SiC composite layer via polymer

impregnation and pyrolysis – samples fabricated

SiC triplex tubes – monolithic and composite layers

32

ATF Development in the ROK – Ceramic and Hybrid

Cladding

Hoop Strength of SiC Composite Tube

Micro-cell UO2

– Enhanced fission product retention

– Concept successfully implemented –

Cs preferentially reacts with grain boundary

– Pellets deform more easily than UO2 pellets

– Metal micro-cell

• Physical barrier – W: manufacturing feasibility and improved thermal diffusivity

demonstrated

• UO2 with CrO3 (3 wt%) – reduced thermal expansion andstored energy, low diffusivity with

physical barrier

FCM Fuel

– Coated TRISO particles

– Pellet: TRISO-SiC Matrix

– Cladding: SS310, SiC, SiC-Zr hybrid, FeCrAl

– AT benefits:

• Resistance to meltdown, fission product

release, H2 generation(low reaction rate

cladding), fuel thermo-mechanical degradation

• Proliferation resistant

• Flexible applications

– Must meet <20% enrichment screening

criteria

33

ATF Development in the ROK – Fuel

ATF R&D in China

Multiple organizations: CNNC, CGN, SNPTC, CAE

Advanced Cladding

– Advanced Zr-alloys

– MAX-phase coatings

• Sample tubes currently under irradiation test (10/2012-10/2014); 20 dpa at 200-400 C

– SiC cladding

Advanced Fuel

– UO2 w/BeO – preliminary R&D shows enhanced thermal conductivity

– TRISO fuel

– Uranium silicide

Theoretical studies and simulation of fuel performance

34

China General Nuclear – Fields of

Interest for Collaboration

Innovative Cladding

– SiCf, SiC-coated Zr-alloy

– Bench-scale fabrication, mechanical properties testing,

furnace tests, modeling, irradiation test and characteristics

Innovative Fuel

– U3Si2-Al dispersion fuel – fabrication and test facilities

could be used for new fuel development

– TRISO coated fuel particles (used in HTR-10 at Tsinghua

University) – fabrication and test capability

Metrics Identification

Fabrication and experiments

– Performance testing for fuel and cladding: High-T thermal analysis system; material test system;

universal microscope

– SiC cladding manufacturing process and properties: sintered powder molding; microstructure

analysis

Conceptual design and assessment

Reactor operation and fuel cycle constraints

Regulatory issues

35

ATF R&D in the Russian Federation

Initiating R&D on advanced cladding:

Research in SiC for cladding and structural materials

– M.V.Frolova, PN Alekseev, PS Teplov AV Chibinyaev. Impact of the use of composite

material based on SiC / SiC as a cladding on the neutronic characteristics of the reactor ASS-

600 / / VAST, Ser. Physics of nuclear reactors number 2, 2012, p. 64-74.

– PN Alekseev, PS Teplov AV Chibinyaev, VM Mahin, MV Frolova, "Improving the performance

and safety of fuel use in reactors ASS-600-670 and SKDI using structural materials based on

silicon carbide”, Abstracts of the 8 IRTC "Safety Assurance of NPP with WWER" IRTC 2013,

OKB "Hydraulic", Podolsk, 28 - 31 May 2013

– Bezumov V.N., Novikov V.V., Kabanov A.A., Zakharov R.G., Pimenov Y.B., Design Issues of

Fuel Rod Cladding Made of Composite Material Based on Carborund (SiC) for Concept of

Water Cooled Reactor Safety under Accident Conditions, VVER 2013, Experience and

Perspectives after Fukushima, Nov 2013.

36

ATF R&D in Switzerland

NPP Leibstadt

– Preliminary severe accident analysis using MELCOR to estimate the impact of

accident tolerant materials for BWR channels and fuel cladding

– Initial interest in FeCrAl materials

– Estimation of additional “grace time” during a severe accident – semi-quantitative

assessment of ATF benefits relative to current system

Paul Scherrer Institue

– Following international ATF research

– No current plans for experiments

37

Potential EU-Funded Projects: Horizon 2020

Currently in proposal stage

Several ATF proposals being prepared for EU-funded projects under

Horizon 2020 Research and Innovation Program

Proposals due September 2014

Possible submissions from Belgium (SCK.CEN), United Kingdom

(National Nuclear Laboratory), Sweden (KTH), France…

38

ATF R&D SUPPORT

ACTIVITIES

39

40

Halden Reactor Project (Norway)

Interactions/collaborations with Halden are ongoing

– Advanced Modeling and Simulation

– Potential ATF irradiation

2015-2017 program currently in discussion

Coating tests currently ongoing for grid-to-rod fretting (not ATF)

Karlsruhe Institute of Technology

(Germany)

Experimental infrastructure

Evaluation techniques:

– 2D x-ray inspection and 3D computed tomography

Separate effects testing:

– Separate-effects tests on high-temperature oxidation,

quenching and mechanical behavior of AT claddings

• Quench facility: Investigate core reflood; DBA LOCA

experiments (out-of-pile; T>2000oC)

• High T oxidation testing in steam w/quench:

SiC material testing (a-SiC, SiC Triplex)

– Bundle tests with prototypic LOCA and BDBA scenarios

• One bundle experiment with SiC-SiC claddings

(if available) planned within the next evaluation

period of KIT

– Steam furnace for thermogravimetric

analysis (TGA)

– BOX rig: investigation of materials

at high T in defined atmospheres

41

SiC/SiCf

cladding

TECHNICAL MEETING “ACCIDENT TOLERANT FUEL CONCEPTS FOR

LIGHT WATER REACTORS”

13-17 October, 2014 Oak Ridge National Laboratory, Oak Ridge, TN, USA

The TM will include a round-table discussion on outlining IAEA Coordinated Research Project “Analysis of options and experimental examination of accident tolerant fuels for water-cooled reactors”

(CRP ACTOF planned for 2015-2019)

THANK YOU

43

ATF-1 Irradiation Test Series

(begin mid-2014)

Objective:

Assess the performance of proposed Accident Tolerant Fuel concepts

under normal operating conditions common to existing LWRs.

– ATF-1 test series is appropriate for testing irradiation behavior of new fuels and

their interaction with cladding

– Test rodlets are isolated from ATR primary coolant, so ATF-1 test series is not

appropriate for testing interaction of cladding (or coatings) with coolant

– All post-irradiation examination work for

ATF-1 planned to be conducted at INL –

preliminary plan currently being

discussed

Preparing for Steady State Loop

Irradiation in ATR – (ATF-2)

Conceptual water loop for ATF Transient

Testing (ATF-4)

Reduced oxidation and hydrogen

generation is a key benefit of alternative

cladding and materials

0 5 10 15 20 25 30 35 40 45 50

0

5

10

15

20

25

30

CVD SiC

NITE SiC

FeCrAl

Th

ickn

ess C

on

su

me

d [

mm

]

Time [hrs]

1200C

0.34MPa Steam

Time (hrs)

Surf

ace R

ecessio

n (

mm

)

30

25

20

15

10

5

0 0 5 10 15 20 25 30 35 40 45 50 310SS FeCrAl NITE-SiC CVD SiC Zr

0.1

1

10

100

1000

Th

ickn

ess C

on

su

me

d [

mm

]

1200C

1300C

1350C

8 hour tests

Source: Snead, et al. 2013, Oak Ridge National Laboratory, Advanced Fuel Campaign