U.S. Status of Fast Reactor

Research and Technology

Sal Golub

U.S. Department of Energy

Robert Hill

Argonne National Laboratory

IAEA Technical Working Group on Fast Reactors

May 16, 2017

The submitted manuscript has been created by UChicago Argonne, LLC as

Operator of Argonne National Laboratory ("Argonne") under Contract No. DE-

AC02-06CH11357 with the U.S. Department of Energy. The U.S. Government

retains for itself, and others acting on its behalf, a paid-up, nonexclusive,

irrevocable worldwide license in said article to reproduce, prepare derivative

works, distribute copies to the public, and perform publicly and display

publicly, by or on behalf of the Government.

GOVERNMENT LICENSE

Secretary Perry on Advanced Nuclear

3

“Fast reactors are one of the areas that I think that there’s some real

potential growth…The United States, I think, would be wise to use the

resources that we have here {at INL} to commit to the future of having our

ability to participate in that fast reactor technology and the potential that it

has for the future”

----Idaho National Laboratory, May 9, 2017

“…. We believe it is wise for countries to use and pursue

highly efficient energy resources.

They include… Advanced civil-nuclear technologies that

are proliferation resistant, produce little to no waste and

ensure safety.

Innovation is also a top priority for the Trump

Administration. We are committed to developing,

deploying and commercializing breakthrough

technologies…”

---G-7 Statement, April 10, 2017

4

Office of Nuclear Energy

Mission Areas

Sustaining the Current Fleet

of Light Water Reactors

Deploying Small Modular

Reactors

Demonstrating Advanced

Reactors

Nuclear Waste Management

Nuclear Science User

Facilities and Enabling

Capabilities

5

Advanced Reactor Vision and Strategy

GOAL

By the early 2030s, at least two non-light water advanced

reactor concepts have reached technical maturity,

demonstrated safety and economic benefits, and completed

licensing reviews by the U.S. Nuclear Regulatory Commission

(NRC) sufficient to allow construction to go forward.

VISION

By 2050, advanced reactors will provide a significant and

growing component of the nuclear energy mix both

domestically and globally, due to their advantages in terms of

improved safety, cost, performance, sustainability, and

reduced proliferation risks.

Growing National Interest in

Advanced Reactors

There is a growing interest in the development and deployment of

advanced (non-light water) nuclear reactor technologies

DOE-NE completed an Advanced Test/Demonstration Reactor Study as

directed by Congress in the FY2015 Omnibus Spending Bill

Secretary of Energy Advisory Board Task Force on the Future of Nuclear

Power recommended a major new deployment of nuclear power, to

include advanced reactors, in the 2030-2050 time period

Nuclear Energy Advisory Committee recommended prompt development

of a versatile fast spectrum test reactor

Held 3 NRC-DOE Joint Workshops on Advanced Non-Light Water

Reactors

Both Houses of the U.S. Congress have introduced and passed

legislation supporting innovative nuclear technologies

6

7

Engagement with Industry

Established the Gateway for Accelerating Innovation in Nuclear (GAIN)

initiative to provide developers with access to technical, regulatory and

financial support

Industry self established various advanced reactor working groups. The

Fast Reactor Technology Working Group consists of diverse concepts:

SFR – Oklo, General Electric, TerraPower, Advanced Reactor Concepts

LFR – Westinghouse, Columbia Basin Consulting Group

GFR – General Atomics

MCFR – Elysium, Southern/TerraPower

Group requests are generally broad in scope and identify capabilities useful

for multiple technology options 7

Priorities for Advanced Reactors

Work with NRC and industry to develop a framework for

advanced reactor licensing

Make DOE assets and expertise available to industry and

academia via the GAIN Initiative

Conduct cutting edge research to enable the commercial

deployment of advanced reactors by the 2030’s

Applying Modeling and Simulation tools suitable for analysis of

advanced reactor systems

8

9

Energy Production

Reactor

Recycle Reactor

Recycle Used

Uranium

Extend Uranium

Resources

Recycle Fuel

Fabrication

A wide variety of actinide management strategies possible – Waste management – Resource extension

Favorable features for small reactor applications – Compact (high power density) – Extended burnup and cycle length – Inherent safety

Favorable features for plutonium management – High loading and throughput possible

With key technology development, also intended for electricity, heat production, or other energy product missions

Actinide Management in Fast Reactors

10

Fast Reactor R&D Priorities

For the commercial deployment of fast reactor technology, two

recurring challenges are identified

– For advanced fuel cycles, capital investment in reactors is the dominant

cost (cost reduction is also vital for electricity production)

– A pathway must be established for non-LWR licensing

One research focus is capital cost reduction through

application of innovative technology solutions

– Improved design approach – components and maintenance

– Advanced structural materials to reduce commodities

– Advanced energy conversion to improve size/efficiency

– Advanced modeling and simulation to optimize performance

– Fuel development to improve fuel cycle costs

Another focus is the resolution of key licensing issues

– Safety R&D to validate tools and assure margins

– Qualification of fast reactor fuels

11

Outline

Technology Innovations for Cost Reduction

– Design, components, and maintenance

– Advanced materials

– Energy conversion options

(Advanced modeling and simulation)

(Recycle fuel performance)

Key Licensing Issues R&D

– Safety analysis methods and modeling

– Validation and knowledge preservation

– Fuel qualification

Mechanisms Engineering Test Loop

(METL) Facility

12

To test small or intermediate scale advanced liquid metal

components and instrumentation in sodium:

– Gear Test Assembly for Compact Refueling Machine

– Westinghouse Thermoacoustic Sensor Testing

To develop and provide performance data on systems and

components used in sodium and reduce risk of failures during

operations

Will provide much needed U.S. infrastructure (both personnel

and hardware) to test liquid metal systems and components

METL consists of:

– ~3,000 kg R-grade sodium – to be purified in cold trap

– Two 18 inch test vessels and two 28 inch test vessels

– Max system temperature = 1000°F (except for 28 inch test

vessels – 1200°F)

– Test vessels can be isolated from main loop

See: https://www.osti.gov/scitech/biblio/1334794-mechanisms-engineering-test-loop-phase-status-report

May 10, 2017

NE

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12

METL – Dump Tank and Main Loop

METL Status

We have the following systems and

components installed:

– Two large test vessels

– Two small test vessels

– Dump Tank

– Expansion Tank

– Large and small diameter valves

– Three pumps and flow meters

– Vessel Piping and Supports

– Plugging meter, cold trap, and vapor trap

– Control Cabinets

– Instrumentation and Heaters

– Valve Manifold

The following system and components are in

various stages of design and fabrication

– Control System is being developed and tested

– Insulation on piping and vessels – on going

– Checkout and commissioning will be starting soon

– Initial sodium fill – Summer 2017

– Test plan is being updated

METL – Test Vessels – piping insulated

Gear Test Assembly - Current Project Status

• Manufacturing of fabricated components

completed and received.

• Control system has been design and

constructed

• Electrical drawings for wiring diagrams and

system layout in production.

• System has been assembled

• Remachining was required of some parts

due to binding issues found after

assembly

• Performing some NDE of gears before

testing

• Developing plan to test in water before

sodium testing

• The gear test assembly will be the first

test assembly to be installed in METL

17

EM

Pump

EM Pump

Control

Unit

Gas/Vacuum

Control

Console

Translation

System

Test

Tank

Cold

Trap

Expansion

Reservoir

Sodium

Bypass

Cold Trap

Bypass

Flow

Meter

Dump

Tank

Under Sodium Viewing (USV)

System Temperature

Controller Unit

Develop an enabling ultrasonic technique for real-time operation monitoring and

nondestructive examination (NDE) under opaque liquid metals:

– Real-time operation and maintenance monitoring of SFRs at high temperature

and high radiation in-sodium

– In-service inspection and repair of components, structures, and systems

within reactor core or steam generators

– Improve reliability, ensure safety, and reduce operational costs

– Benefit inspection needs of various industries, particularly those requiring

inspection/monitoring in harsh environment

DAQ & Image

Processing

Sodium

Flow

Control

Ultrasonic

System

Gas Flow

Control

Bellows

Assembly

HT-TD

Target

Holder

Test

Tank

18

Submergible High Temperature Transducer

Under Sodium Viewing (USV) In-Sodium Test of UWT and Submergible HT Transducer

Simulation and Water Mockup of BUWT-PA

Ultrasonic Waveguide Transducer

Gold-plated

Focal Lens

Transducer Ultrasonic Waveguide

Achievements:

Temperature: 177C in sodium

Resolution: 0.5 mm in width and depth

Beam Size: < ~5 mm in sodium

Brush-type Ultrasonic Waveguide

Transducer Phased Array (BUWT-PA)

Achievements:

Temperature: 160-550C in sodium

Resolution: 0.5 mm in width and 1 mm depth

Piezo Element: PZT-5A or LiNbO3 @ 160°C Hot Oil

@ 177°C @344°C In-Sodium @350°C

Multiplexing

Composite

32-element in

Sodium (simulation) Multiplexing

Normalization

Water

Mockup

Code Qualification for Alloy 709

Alloy 709 (20Cr-25Ni) is an austenitic stainless steel with significant time-dependent

strength advantage over 316H stainless steel as a SFR construction material

Enhanced time-dependent strengths of Alloy 709 with respect to 316H can

o Reduce commodity requirements, and thereby decrease the capital cost of the reactor plant

o Permit structural components to withstand higher cyclic and sustained loading, leading to higher

safety margins, and the prospect of eliminating costly add-on hardware instituted in past designs,

and other design innovations and simplifications (Sham and Jetter, PVP2017-65102)

Developed an Alloy 709 code qualification plan to generate data to support

o Development of ASME Section III, Division 5 Code Case

o Licensing effort and long-term plant operations

Phase I Implementation

o ASME Section III, Division 5 Code Case for 650°C and 100,000 h

Engaged ASME Code committees on planned efforts

o Initiate very long-term thermal aging and creep tests to support future extensions of the code case to

longer design lifetimes (500,000 h)

Completed enhancement of creep testing capacities at the three national labs (ANL, INL and ORNL)

o Initiate very long-term sodium compatibility testing to generate data package to support licensing

effort and plant operations

Completed design effort for the construction of a sodium material test loop with large exposure vessels for

standard-sized test specimens 19

Alloy 709 Fabrication Scale-up

An Alloy 709 heat, ~55,000 lb, was melt by a vendor using commercial melt practice

Rectangular slabs and round ingots were produced

Slabs were hot-rolled to 1-inch plates using

vendor standard mill practice

Plates were solution annealed, and delivered

to the labs

Testing by the national labs and DOE-NEUP university

projects will commence after microstructure and mechanical

properties screening are complete

20

Extending Existing Materials:

Grade 91 Steel to 60-Year Design

Qualify Grade 91 for ASME Section III Division 5, Elastic, Perfectly Plastic (EPP)

Evaluation Methods

– Strain limits and creep-fatigue damage evaluation methods were developed as part of the ASME Code

Modernization effort to simply design analysis methods

– EPP methods do not require stress classification and take advantage of modern finite element technology

– Originally developed for stainless steels, additional qualification of Grade 91 for EPP is needed due to its

cyclic softening behavior that is different from stainless steels

– Testing of Grade 91 using special test articles that mockup thermal ratcheting (two-bar) and elastic follow-

up effects in actual structural component (SMT) to qualify Grade 91 for EPP has been initiated

Setup for two-bar testing SMT test setup and test article

21

Develop ASME Design Parameters for Grade

91 Steel to Support 60-Year Design Life

Thermal Aging Factors for Yield and Tensile Strengths in ASME Code

– A thermal aging model based on carbide coarsening mechanism has been developed to predict thermal

aging effects on yield and tensile strengths of Grade 91

– Efforts to verify and validate the model is ongoing

– Recovered 13 specimens from long-term creep tests of Grade 91 from ORNL (50,000 to 132,000 h)

– Also recovered an ex-service Grade 91 tube material from ORNL that had been exposed to steam

condition for 147,000 h at Kingston Fossil Plant, TN, USA

– Some specimens have been made

– Microstructure characterization and in-situ tensile tests will be performed to calibrate and validate the

thermal aging model

Section preserved for creep damage

(cavitation) characterization Section for specimens

preparation

22

Supercritical Transformational

Electrical Production (STEP)

STEP is a DOE crosscutting program supported by 3 DOE offices (EERE, NE, FE)

STEP Pilot Demonstration Facility Project awarded to Gas Technology Institute

– 10 MWe operation of RCBC at Southwest Research Institute, San Antonio, TX by 2020

ART Energy Conversion Team focuses on Sodium Fast Reactor Application

– Power Block R&D funded by STEP, Reactor to sCO2 Power Block funded by ART

– Technology Roadmap/Project Management Plan/System Engineering Model completed

• Systematic Risk Identification and Retirement from components to system configuration.

– Commercialization of the sCO2 system by 2030 through Collaborative Basic and Applied Research and

Development(CBARD)

– Operation Recompression Closed Brayton Cycle at a turbine inlet of 550C

• Brayton Development Platform working with industry to achieve high TRL components for system integration for

STEP facility.

– Development of Intermediate Sodium to CO2 Heat Exchanger (Primary Heat Exchanger)

• Sodium Drain, Fill, Plug in PCHE

• Sodium/CO2 Interaction

ART Energy Conversion Engages Industry to Advance TRL of Components to 550C

– FBO to CRADA Process yields Lab/Industry collaborations with national awards

– SNL procuring the worlds first 1 MW turbo compressor at 750C though Design/Build process

– Standing up 8 additional test configurations in addition to the RCBC Test Article to address:

• Heat Exchanger (SEARCH), Particle Imaging Velocimetry (PIV), Seals, Bearings, Turbcompressor (Core), Dry Heat

Rejection (Tall Loop), RCBC Parallel Compression

23

Supercritical CO2

Commercial Development Program

Summary to date: Advance bearings, seals and heat exchangers from TRL 3 to TRL 7 to be

scalable for a 10 MWe 700C s-CO2 RCBC demonstration.

1.5” prototype

Sample Flowserve compressor

seal geometry in CO2 service

Peregrine heat exchanger plates

96 Chewonki Neck Rd Wiscasset, ME 04578

24

sCO2 Recompression Closed

Brayton Cycle R&D Program

Demonstration of component and

cycle performance for

recompression closed Brayton

cycles (RCBCs)

Development and design of next

generation RCBC testing systems

- Motor driven compressors

- Leads into stable RCBC testing

- Measures thrusting forces

Parallel Compression Test Rig

Industry

collaboration

to test

commercial

scale

components

and systems

Development of RCBC

optimization and control programs

Peregrine Turbine1.0 MW Test Rig

Design

1.0 MW turbocompressor housing

1.0 MW

turbine

wheels

25

26

sCO2 Cycle Dynamic

Modeling and Simulation

ANL Plant Dynamics Code (PDC) is current

worldwide state-of-the art capability for system

level plant dynamic analysis of sCO2 Brayton

cycle power converters

– Coupled to SAS4A/SASSYS-1 LMR code

PDC is being used to investigate dry

air cooling for the sCO2 cycle

– An increase in plant $/kWe of only about 2% over

water cooling

– With commercially available technology

– Need to re-optimize cycle for dry cooling

PDC validation with IST and RCBC test data

– Power-neutral TAC shaft operation (no active

speed control) tests at IST

Sodium-CO2 Interaction Tests in

the Argonne SNAKE Facility

sCO2 – sodium interaction experiments studied

chemical reaction of sCO2 and sodium at prototypic

SFR conditions. sCO2 jet injected into open Na pool

and DBHX geometries.

In two experiments the solid reaction products

stopped the flow of sCO2 even at 200 atm (2900 psig)

pressure.

Lower temperature experiments resulted in

exothermic reaction producing agglomerated porous

solids (NaCO3, C, NaO2) and CO.

Acoustic leak detection feasible.

Test vessel post-experiment

Injection nozzle outlet pre-test & post-test

(solid reaction product plug)

r=3mm

DBHX Geometry

Open Na Pool

Geometry

28

Compact Heat Exchanger Development –

Key Phenomena

Three experiment facilities are being assembled to obtain data

essential for reliable design of intermediate heat exchangers

– Sodium Freezing and Remelting

– Sodium Draining and Refilling

– Sodium Plugging

Without the data and an understanding of the phenomena in the tests,

heat exchanger design constraints (e.g., a reliable safe value for the

minimum sodium channel size) will not be known

New Test Section

EM Pump

Reservoir

Direct Regulator Engagement

on Licensing Framework

U.S. NRC issued Draft Regulatory Guide DG-1330 (Proposed New Regulatory

Guide 1.232) as “Guidance for Developing Principal Design Criteria for Non-

LWRs” in February 2017

– Consistent with DOE proposal and structure of General Design Criteria for

LWRs (10-CFR-50 App. A)

– One set of technology-inclusive Advanced Reactor Design Criteria (ARDC)

– One set of SFR-specific Design Criteria (SFR-DC)

A new DOE initiative on revisions for the Standard Review Plan (NUREG-

0800) for SFRs (SFR-RP)

– Initial effort is limited to Chapter 4 (reactor core components)

– Most significant revisions to Section 4.2 on fuels: Proposed that the

applicant provides limits for cladding strain, fuel and cladding

temperatures, and time-at-temperature as fuel acceptance criteria for both

metallic or oxide fuel forms

– It is required that the "net reactivity feedback” (from all feedback

mechanisms) to be negative at temperatures elevated above the normal

operating range

29

Safety Analysis Code

Improvements

SAS4A/SASSYS-1 Modernization:

– Version 5.2 Completed (brand new source code and data management

structure compliant with Fortran 2003 standards)

– Development of SQA plan and its initial application for NQA-1

compliance

MELCOR-CONTAIN-LMR Integration

– Revive US sodium fire modeling capabilities by moving Contain-LMR

models into MELCOR

– Provide experiment validations including the use of past spray and pool

fire tests

– Validate the atmospheric chemistry model

– First expected supported release in October 2017 (code release and

updated manuals)

Reduced order thermal stratification modeling and testing for its validation via NE-

University Program

– University of Wisconsin/MIT and Kansas State University/UIUC

30

PRISM PRA Update

Multi-year project to update/modernize

the PRISM SFR probabilistic risk

assessment.

Goal: Address shortcomings of PRA

noted by the NRC during PSID submittal

in 1990s and meet requirements of new

ASME/ANS Non-LWR PRA standard.

Results: Project completion by end of

2016 (in interim, many conference papers

- ICONE/PSAM – on methodologies and

preliminary results).

31

GE Hitachi PRISM – SFR Design

Argonne: Examining passive safety

system reliability, the component

reliability database, and the mechanistic

source term assessment.

GE Hitachi: Developing PRA structure,

conducting system analysis and event

sequence identification/analysis, and

determining risk-insights.

SFR Mechanistic Source Term (MST)

Assessment

Assessment of methodologies for

mechanistically estimating the "source

term" from postulated accidents in the

residual risk domain for metal-fueled, pool-

type SFRs using best-estimate models

32

Advanced Modeling & Simulation

High-fidelity FEM neutron transport

module with matching cross-section

generation capabilities

Unstructured grid for complex geometries

• Deformed SFR cores with moving grid

On-line cross section generation

Excellent scalability for parallel computing

CFD with high-order spectral elements on an

unstructured (but conformal) hexahedral grid

Incompressible and weakly-compressible flow

DNS, LES, and RANS formulations for

turbulence

Excellent parallel scaling (1M+ ranks)

• Gordon Bell and R&D-100 awards

MC

2-3

/PR

OT

EU

S

Ne

k5

00

0

Advanced Modeling & Simulation

SFR challenge problems and V&V opportunities SHARP-zoom (analytic magnifier)

Coupled system-CFD analyses (to support bilateral

collaborations with Japan, France, China)

Wire-wrapped SFR pin bundle benchmarks (DOE-

funded, industry-led projects Areva/TerraPower, EU-

Sesame INERI, bilateral/trilateral collaborations with

Japan and France)

Reduced-order thermal stratification modeling

AFR hot-channel factor evaluation Numerous corrections to account for uncertainties In

modeling, experiments, instrumentation, manufacturing

tolerance, etc.

Reevaluations using SHARP:

– Intra-assembly flow maldistribution

– Cladding circumferential temperature

– Wire Wrap Orientation

– Subchannel Flow Area

– Film Heat Transfer Coefficient

– Cladding thickness and conductivity

– Coolant Properties

– Fissile Maldistribution

– Fuel Thermal Conductivity

– Inlet Flow Maldistribution

34

M&S for long-life cores: T/H simulations

and experiment for SFR fuel assemblies

Team

Previous work

• Previous efforts (recent and legacy

experiments) focused on nominal

geometries.

• Little data suited for CFD, none takes

into account deformations seen later

in fuel assembly life.

• Collaboration leverages extensive

simulation experience with wire-

wrapped rod bundles at Terrapower

and Argonne as well as experimental

excellence at Areva and TAMU

Sample 37 Pin Model (deformations)

Work in Progress

Expected Outcomes

Numerical Simulation Detailed Hydraulic Experiments Thermal Performance experiments

• High quality experimental data for:

a. Specified metallic pin deformation;

b. Specified duct deformation.

• Objective: generate CFD validation grade data for non-

deformed and deformed geometries

a. It is a benchmark of current approaches with

benchmark milestones, not a software development

effort.

• Impact: Validated CFD analysis will be used to assess

thermal performance of the fuel assemblies’

AREVA - Temperature measurements

of 61 pin fuel assemblies (19 distributed

heater rods + 42 unheated rods) with

straight duct and deformed duct

configurations

TerraPower, LLC –

Pre/post-test CFD predictions

Argonne National Laboratory -

Performing highly resolved

pre/post-test CFD predictions

Texas A&M University - Velocity

and pressure measurements of 61

pin assemblies (unheated) with

straight duct and deformed

pin/duct configurations

Fast Reactor

Knowledge Preservation

Identify areas where information is at risk of being lost or

destroyed

– Example – FFTF document preservation at PNNL

Collect and organize FR-related information

– EBR-II SHRT database

– FFTF Passive Safety Testing database

– TREAT Database

– EBR-II Fuels Irradiation Database

– Fast Reactor Reliability Database

Recover lost computer codes

– NUBOW-3D code recovery – supports core restraint design

– SWAAM (sodium-water interaction) code recovery

• Upgraded to include sodium-CO2 interactions

– SOFIRE – a sodium pool fire code

Make information accessible to U.S. and other Fast Reactor

technology development countries

– DOE’s Office of Scientific and Technical Information

• Over 8,600 documents have been scanned and made available

– IAEA CRP on EBR-II Passive Testing Benchmark

– IAEA CRP - NAPRO 36

37

FFTF – Information Recovery

Accomplishments

An FY16 FFTF Lessons Learned report was prepared discussing the

following technical areas:

1. Sodium Thermal Stratification,

2. Sodium Vapor Trap Design/Operation,

3. Thermal Transient Usage,

4. Sodium Pump Flooding, Shaft Bowing, and Seizure,

5. Sodium and NaK Fill Process,

6. Heat Exchanger Performance.

7. Deactivation of Primary System Isolation Valves

8. Sodium and NaK System Deactivation

These FFTF lessons learned reports are critical to passing on information

regarding the performance of U.S. fast reactor technology

This work is supported by personnel who were involved with the operations

and irradiation testing of the FFTF reactor located at Hanford, Washington

SFR Safety Testing Databases

EBR-II Safety Test Database: ~80 experiments from the comprehensive

shutdown heat removal, balance of plant, and inherent control testing program

conducted at EBR-II during 1984-97 period.

Including the landmark inherent safety demonstration test (unprotected

station blackout)

FFTF Passive Safety Testing Database: Natural circulation tests as a reliable

means of decay heat removal during unprotected loss-of-flow transient,

extending passive safety experience to a large-size SFR

Including impact of unique core restraint system design and GEM device

TREAT Test Database: Archive of documents, meta- and numerical data from

~800 one-of-a-kind tests as the basis of present knowledge of transient fuel

behavior on key phenomena related to transient fuel performance including fuel

failures.

SFR Component Reliability Database: Based on combination of original

CREDO data, as well as revisited EBR-II, FFTF, and FERMI run logs, to

support Fast Reactor PRA.

38

FFTF Passive Safety Benchmark

Benchmark specification for most severe Loss-

Of-Flow WithOut Scram (LOFWOS) test

– Core loading, assembly fuel compositions

– Major plant components and heat transport

system configuration

– Description of measurements – locations and

types of instruments

– Description of initial and boundary conditions

Demonstrated effectiveness of Gas Expansion

Modules (GEM) as passive reactivity control

devices

– Pump trip from 50% power without scram

– Measured coolant flow and temperatures to

natural convection using two fast-response

instrumented fuel assemblies

FFTF Reactor

400 MWth loop-type test reactor

Mixed UO2-PuO2 (MOX) fuel

Unique core restraint system 39

SFR Metal Fuel Qualification Databases

EBR-II Physics Analysis Database and IFR Materials Information System:

Pin-by-pin core loading data for each EBR-II cycle, fuels fabrication information,

and operating parameters for validation of depletion analysis capabilities.

EBR-II Metal Fuel Irradiation Database: Detailed pin-by-pin fuel irradiation

history in EBR-II (digitized micrographs for profilometry data, gamma scans,

and fuel cross section, and PIE data from Argonne Alpha-Gamma Hot Cell

(porosity, cladding strain and other microstructural characteristics)

FFTF Metal Fuel Irradiation Database: Archival of tests conducted at FFTF including fabrication data and QA documentation for IFR-1 and MFF series of metal fuel tests, reports and available operational data for irradiation cycles, results for impact of metal fuel tests on reactor operating parameters such as reactivity feedbacks, and direct measurement data (in-core assembly growth, assembly pull forces, IEM cell exams).

Out-of-pile Metal Fuel Test Database: Data from Whole Pin Furnace tests and Fuel Behavior Testing Furnace to characterize steady-state and transient margins to fuel failure for binary metallic alloy SFR fuel as part of IFR program.

40

Fast Reactor

Advanced Fuels Development

Scope: Advance the scientific understanding and engineering application of

fuels for use in future fast-spectrum reactors, including:

1) fuels for enhanced resource utilization (including actinide transmutation),

2) support for driver/startup fuel concepts.

Approach:

– DOE-funded development to focus on metallic fuels technology

– Industry interest manifested in metallic fuel-related CRADAs in critical

areas (e.g., new fabrication techniques)

– Continue international collaborations (e.g., CEA, JAEA) to track the

status of MOX fuels technology

– Support fuel technology needs of DOE-NE/ART and industry FOA

concepts relative to demonstration/test reactors

41

Develop/demonstrate advanced

fabrication methods capable of

fabricating fresh and recycled fuels

with <0.1% unrecoverable loss of

transuranic materials.

– Advanced casting (GACS)

– Continuous casting

– Robust crucibles and molds

– Extrusion

– Additive manufacturing

Fuel Fabrication

High Level Objectives

Demonstrate remote (i.e., hot cell)

fabrication of metallic fuels that

make use of recycled fuel

feedstock obtained from

separations processes.

– Design, fabrication, and installation of

advanced casting furnace in HFEF

– Demonstrate remote casting of

transmutation fuel alloys

– Remote casting of fuel for IRT

42

Fuel Performance

High Level Objectives

Demonstrate acceptable

performance of fast reactor fuels

which can accommodate Pu and

significant quantities of minor

actinides, with a view to provide a

fuel technology option for recycle

of used nuclear fuels and

management of actinide materials.

– Prepare actinide feedstock materials

– Test fuels with Pu, Am, Np contents

Develop/demonstrate fast reactor

fuels and associated fuel

technologies (i.e., cladding)

capable of reliable performance to

high/ultra-high burnup.

– Low smear density metallic fuels

– Annular, sodium-free metallic fuels

– Cladding coatings and liners

– Fuel alloy additions for lanthanide

control

– High-dose/strength cladding materials

43

Support development and

validation of an advanced fuel

performance code for use in

design/analysis of metallic and

MOX fuels in fast reactors.

– BISON is development platform

– Implement legacy metallic and MOX

fuel behavior models

– Develop new mechanistic models

– Systematize fuel characterization and

material property data (Handbook)

– Perform model validation

Fuel Modeling

High Level Objectives

BISON

BISON 1.3

44

Fuel Fabrication

Fuel Fabrication Activities

– Characterization Specimens (Handbook)

• Legacy and advanced metallic fuel alloys

1) Pu-10Zr

2) Pu-30Zr

3) U-20Pu-10Zr

4) U-20Pu-3Am-2Np-10Zr

5) U-20Pu-4Pd-10Zr

6) U-20Pu-4Pd-4RE-10Zr

7) U-19Pu-0.7Zr-4.3Ti-5Mo

– Fuel Fabrication for AFC-3F

• Fabrication Variables Experiment

U-20Pu-4Pd-4RE-10Zr

AFC-3F Fuel Slugs

AFC-3F

Nominal Design

Rodlet Composition Form SD

(%)

3F-1 U-10Zr (EBR-II) solid 75

3F-2 U-10Zr solid 75

3F-3 U-20Pu-10Zr solid 75

3F-4 U-10Zr solid 75

3F-5 U-20Pu-10Zr (EBR-II) solid 75

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Fuel Optimization Alloys-UPuZrPd

Palladium has been shown to be a promising additive to control

FCCI. Studies using Pd have been extended to include

transmutation fuel, i.e. U-20Pu-10Zr-3.86Pd

Microstructural characterization of the alloys is underway.

Diffusion couples against Fe have been started.

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U-20Pu-10Zr-3.86Pd U-20Pu-10Zr-3.86Pd-4.3Ln

Irradiation Status: AFC-3D

AFC-3D irradiation is complete, the experiment has been shipped to the Hot Fuels

Examination Facility (HFEF) for PIE.

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AFC-3D

Nominal Design As-Run Calculations (EOC 158B)

Rodlet Composition † Form SD

(%) EFPD

Average

LHGR

(W/cm)

Fission

Density

(1021

f/cm3)

Burnup (at.%)

HM 235U

3D-1 U-10Zr annular 55 195 302 1.61 4.6 6.8

3D-2 U-4Pd-13Zr solid 55 91 330 0.84 2.8 6.0

3D-3 U-10Mo solid 55 91 322 0.84 2.2 6.8

3D-4 U-10Mo annular 55 195 308 1.69 4.6 14.0

3D-5 U-4Pd-13Zr annular 55 91 349 0.90 2.9 4.8

† Fuel composition expressed in weight percent

Inserted cycle 154B, 23 Aug 2013

3D-2, 3, 5 discharged cycle 157A, 30 Aug 2014; 3D-1 and 3D-4 discharged cycle 158B, 1 Apr 2016

Industry interest in Advanced Reactor options has increased

– Wide variety of technology options being promoted

– Legislation for industry-led R&D and innovation R&D

To support near-term commercialization, R&D efforts are focused on

1) technology innovations for cost reduction

2) key licensing challenges

Recent progress and accomplishments technology R&D items focused

on innovations for performance improvement (cost reduction)

– METL sodium test loop and first test assembly in 2017

– Alloy 709 austenitic alloy ASME code qualification for SFR has started

– STEP Program to accelerate supercritical CO2 technology

Recent progress and accomplishments on licensing R&D items focused

on safety analysis tools/validation and advanced fuels

– PRA and mechanistic source term assessments completed

– Knowledge preservation in databases (e.g., FFTF benchmark proposal)

– Validation of fuel fabrication and irradiation performance

Summary

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