1. Material Development and Testing for Extreme Reactor Applications
S.A. Maloy1,
O. Anderoglu1, T. Saleh1, M. Caro1, K. Woloshun1, F. Rubio1, M. Toloczko2, D. Hoelzer3, T.S. Byun3, G.R. Odette4
1Los Alamos National Laboratory, Los Alamos, NM 87545, USA
2Pacific Northwest National Laboratory, Richland, WA 99352, USA
3Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA
4University of California Santa Barbara, Santa Barbara, CA 93106, USA
Funded by Department of Energy-Nuclear Energy

2. DOE-NE is Performing Research to close the Nuclear Fuel Cycle – Fuel Cycle R&D (FCRD) Program

3. Advanced Fuels Campaign Mission & Objectives in the Fuel Cycle Research and Development Program
Develop and demonstrate fabrication processes and in-pile (reactor) performance of advanced fuels/targets (including the cladding) to support the different fuel cycle options defined in the NE roadmap.
Objectives
Development of the fuels/targets that
Increases the efficiency of nuclear energy production
Maximize the utilization of natural resources (Uranium, Thorium)
Minimizes generation of high-level nuclear waste (spent fuel)
Minimize the risk of nuclear proliferation
Grand Challenges
Multi-fold increase in fuel burnup over the currently known technologies
Multi-fold decrease in fabrication losses with highly efficient predictable and repeatable processes
Continuous
Recycle
Modified
Open
Once-
Through
Advanced Fuels
High-burnup LWR fuels
Accident Tolerant Fuels for improved Safety
Deep-burn fuels or targets after limited used fuel treatment
High burnup fuels in new types of reactors
Fuels and targets for continuous recycling of TRU in reactors (possibly in fast reactors)

4. Scientific Approach to Enabling a Multi-fold Increase in Fuel Burnup over the Currently Known Technologies
Ultra-high
Burnup
Fuels
FCCI
Corrosion
Radiation
Temperature
Coating
Liners
Advanced Alloys
F/M Steels
Advanced Alloys
F/M Steels
HT-9
Advanced F/M Steels, e.g. NF616
Advanced Alloys
ODS Steels Cr
Enhancements with
Fabrication Complexity Si
200 dpa dpa dpa Al
Increasing content
Different Reactor options to change requirements
LFR, GFR
500 C C C
Reduced embrittlement, swelling, creep
Enhancements with
Fabrication Complexity
Enhancements with Fabrication Complexity
Develop an advanced materials immune to fuel, neutrons and coolant interactions
under specific reactor environments

5. Outline STIP – IV irradiated Materials
LANL CMR Hot Cell Tensile Testing
ACO-3 Duct Analysis
Tensile Testing
Charpy Testing
Microstructural Analysis
STIP – IV irradiated Materials
Accident Tolerant Cladding Materials
Advanced Material Development for High Dose Applications
ODS steel Processing
Ion Irradiation Testing
Neutron Irradiation Testing to high dose
Development of Lead Corrosion Resistant Materials
Lead Fast Reactors
DELTA loop
Long Term corrosion tests (>3,000 hours)
Summary and Future Work

6. Our Unique Facilities and Infrastructure Providing Science and Technology to NE
Ion Beam Materials Laboratory, Structural
Materials and Fuels R&D
Actinide R&D: Separations, Integrated Safeguards
Test Lab, Characterization of Irradiated Materials
Information Sciences
World’s first petaFLOPS supercomputer
Hot-Cells for
Processing of
Isotopes
Fuels Research Lab:
R&D on Ceramic Fuels
Actinide Cross-Section Measurements
Irradiation of Targets for Isotope
Production
DELTA loop for LBE Corrosion
Nuclear Fuels R&D: MOX
Fuel Fabrication and Testing

7. Mechanical Testing in CMR Wing 9 Hot Cells
Tensile Specimen- dimensions are 4 mm x 16 mm x 0.25 mm thick (gage dims. are 1.2 x 5 x 0.25 mm3)
Tested at initial strain rate of 5 x 10-4/s.
Tested at 25 to 700ºC in ultra high purity argon
Shear Punch, 3 pt. bend and compression testing capabilities.
16 mm

8. Testing was Performed on the ACO-3 Duct, one of the most highly irradiated components in a fast reactor 1 2 3 5 4
FFTF, Hanford site, WA

9. Analysis of Specimens from ACO-3 Duct
Total specimens= 144 Charpy, 57 compact tension, 126 tensile specimens, 500 TEM
Charpy, Compact Tension testing and thermal annealing completed at ORNL.
Completed tensile testing at LANL from 6 different locations along the duct at 25, 200 and the irradiation temperature.
Completed Rate Jump Testing at 25C
Completed Microstructural Analysis using TEM, SANS and Atom Probe Tomography

10. Stress/Strain Curves – HT-9 Irradiated, Room Temp Tests
Decreased elongation in irradiated materials.
Increased hardening in lower irradiation temperature materials.

11. Results from Previous PIE Studies Agree Well With These Measurements
AC0-1 Duct and Cladding (HT-9) (total dose = ~ 88 dpa)
Maximum dilatation (swelling + precipitation + creep) = 0.5 %
Yield stress increase ~300 MPa, Tirr =~360C, Ttest= 25C, dose = 36 dpa
Previous studies (stars in figure, tested at 25C) show similar dependence of yield stress on irradiation temperature
HT-9 Control
Irradiation Temperature (C)
0.2% Offset Yield Stress (MPa)
25 C
200 C
Tirr

12. Charpy Impact Testing for ACO-3 Duct
Upper shelf energy is a function of irradiation temperature, dose, and specimen orientation, while the effect of irradiation temperature is dominant in the transition temperatures.

13. TEM analysis of ACO-3 Duct Material (B.H. Sencer, INL, O. Anderoglu, J. Van den Bosch, LANL)
T=450C, 155 dpa
Precipitation observed
• Dislocations of both
a/2<111> and a<100>
• Loops of a<100>
Void swelling observed (~0.3 %)
T=505C, 4 dpa
No precipitation or void swelling observed.
T=384C, 28 dpa
G-phase precipitates and alpha prime observed
No void swelling observed.
Small Angle Neutron Scattering Measurements
Obtain accurate measurement of ’ vs. dose and irr. Temperature
Measurements completed on 5 specimens from ACO-3 duct

14. Summary T °C dpa Cr solub at % (G. Bonny) Cr solub (Phase diagram) α′
G-phase
Loops
voids L nm d
x1021
m-3
X1021
380 20 9
8.1
7.8 72
11.3
9.3 14
0.93 X
410
100
9.1
10.1
2.6
16.2
1.4 - 23
440
155
12.4
9.6
9.5
26.5
1.1 18
0.5 28
0.25
466 92
9.9
14.7
505 2
12.5
18.5
475
10,000
hrs
10.3
15.5

15. He/dpa ratios of 50-60 appm/dpa
STIP- (SINQ Target Irradiation Program) Irradiations Provides an Understanding of the Effects of Helium and Irradiation Dose
Materials for STIP IV irradiation include the following in tensile and TEM specimens:
Structural: HT-9, EP-823, Mod 9Cr-1Mo, 9Cr-2WVTa, T122, 5Cr-2WVTa, A21N, ODS strengthened F/M steels-12YWT and 14YWT (Fe-12Cr-3W-0.4Ti-0.25 Y203, Fe-14Cr-3W-0.4Ti-0.25 Y203), V-4Cr-4Ti, High purity Ta, single crystal Fe (for modelling studies)
Fuels Matrices: ZrN, NiAl, FeAl, RuAl, MgO, Cubic ZrO2, Fissium
~570 MeV protons
He/dpa ratios of appm/dpa

16. Summary of STIP IV Tensile Testing at Room Temperature
Tirr=247C, dose=15 dpa
Tirr=120C, dose=8 dpa
Tirr=394C, dose=22 dpa
Tirr=247C, dose=15 dpa
Tirr=380C, dose=22 dpa
12YWT-Control
HT-9-Control
Highest Hardening observed at lowest irradiation temperatures
12YWT tests reached limits of testing machine before yielding
Significant helium accompanies dpa (up to 1300 appm He)
Comparison with Phenix irradiated specimens will help quantify helium effect on mechanical properties

17. Requirements for Accident Tolerant Fuels (ATF)

18. Measurements on hydrogen evolution performed in steam
Zirc-4 in N2 containing ~25% water vapor to 1100°C
304 in N2 containing ~25% water vapor to 1100°C
Hydrogen Production begins in Zircaloy-4 at ~700C and in 304L at ~1000C
Similar testing is underway on all advanced alloys in FY13

19. Advanced Material Development Activities
Characterizing and Testing MA-957 Irradiated to High Dose (>100 dpa)
Obtaining Irradiation data on Advanced Alloys (international collaborations)
MATRIX irradiations- Samples to be shipped in early 2013?
STIP irradiations – Samples from STIP IV to be shipped in next few weeks
Investigating Possible Future irradiations
Domestic Facilities (MTS (18 dpa/yr)) – Collaborating in ATR irradiations
International collaborations
Collaborating with Terrapower for irradiations in BOR-60 in Russia and DOE-RIAR collaborations for additional irradiations in BOR-60.
Initial discussions under way for future irradiation in the CEFR in China.
Advanced Material Development
Friction stir ODS material processing
Mechanical alloying ODS material processing

20. Oxide Dispersion Strengthened Alloys
Strength & damage resistance derives from a high density Ti- Y-O nano-features (NFs)
NFs complex oxides (Ti2Y2O7, Y2TiO5) and/or their transition phase precursors with high M/O & Ti/Y ratios (APT)
MA dissolves Y and O which then precipitate along with Ti during hot consolidation (HIP or extrusion)
Oxide dispersion strengthened alloys also have fine grains and high dislocation densities
Y-YO-Ti-TiO-O

21. Typical Processing Route for ODS Alloys
Ball milling
Alloy powder
Y2O3
Canning
MA powder
As extruded bar
Working
Heat treating
Hot consolidation (extrusion)

22. Ion Beam Materials Laboratory (IBML)
Fundamental irradiation studies performed at the IBML
Irradiations performed on interfaces characterized to the atomic scale
Post irradiation analysis will investigate the role of interfaces on defect formation and accumulation
Aids in model development and provides initial alloy irradiation results.

23. ODS strengthened Materials Show Excellent Resistance to Hardening under Ion Irradiation
Room temperature irradiation (1.5dpa)
Berkovich indenter, 200nm deep indents, constant displacement
HT-9 Martensitic
Dose [dpa]
HT-9 Ferritic
Nanohardness [GPa]
Dose
ion beam
Depth [mm]
MA957
MA956
Nanohardness [MPa]
Dose [dpa]
4mm
10mm
4mm
1mm
Beam
Depth [mm]

24. Analysis of highly irradiated MA-957 Tubes Underway at PNNL
Tensile testing of MA Pressurized tubes
Irradiation conditions in FFTF
(385°C, dpa)
(412°C, 110 dpa)
( °C, dpa)
( °C, dpa)
(750°C, dpa)
Testing will be performed at PNNL
Status
Specimens for testing were machined from pressurized tubes at LANL
Tensile testing and TEM work is underway at PNNL
Analysis of in-reactor creep response is complete.
Preliminary analysis of creep data
MA-957 is comparable to HT-9 in creep resistance to up to 550°C. At 600°C, MA-957 creep resistance remains high while HT-9 creep resistance begins to rapidly decline. 24

25. New ODS 14YWT heat produced with low N and C powder
New consolidation condition explored for 14YWT
2 cans heat treated to nucleate nanoclusters (1 750ºC & 850ºC)
Cans extruded at 1150ºC
Fabrication
Extruded bar cut into 3 sections
One section rolled parallel to extrusion axis at 1000ºC
One section rolled normal to extrusion axis at 1000ºC
Total reduction in thickness was 55% (~5.5 mm final thickness of 14YWT)
No cracking was observed

26. Core Materials Research and Development – 5 Year Plan
Qualify HT-9 for high dose clad/duct applications (determine design limitations)
Data to dpa on F/M and on Inn. Material
FFTF (ACO-3 and MOTA) Specimen Analysis
Rev. 6 of AFCI (FCRD) Materials Handbook
Re-irradiation of FFTF specimens in BOR-60
Advanced Material Development (improved radiation resistance to >400 dpa)
STIP- IV (PSI) Specimen PIE
Materials Test Station Irradiations
MATRIX-SMI and 2 (Phenix) Specimen PIE
Data on Advanced Materials to dpa
Innovative Material Development
Baseline Mech. Prop. Of Inn. Clad Material
Innovative Clad Material Downselect
ODS Ferritic Steel Material Development
Develop ODS Tubing and Weld specifications for innovative Weld material
Produce ODS Tubing
Advanced Materials Irradiation in BOR-60 and CEFR
Accelerated Aging of Advanced Materials (High Dose Ion Irradiations)
Advanced Material Development (improved FCCI resistance to >40 % burnup)
Development of Coated and Lined Tubes
Lined Tube for ATR irradiation
PIE on Lined Irradiated Tube
Innovative Coating Material Development
Fab. Innovative Coated Tube for ATR irradiation
FY’11
FY’12
FY’13
FY’14
FY’15
FY’16
Provides data for NEAMS model development of Cladding

27. All results provided to MBM
Development of a Physics-based Model of Radiation Damage in Nuclear Fuels
Motivation/Approach – Develop mechanistic materials models with improved accuracy and predictive power using atomic level simulation techniques for application in meso-scale and/or continuum models (MBM).
Analysis of segregation in terms of local strain
Compositional variations
Bulk diffusion
Bulk thermal conductivity
All results provided to MBM
Material specificity
MARMOT model development; mesoscale fission gas diffusion and segregation, thermal conductivity (INL)
Role of idealized grain boundaries in mass and thermal transport.
7/20/09

28. LBE as a Nuclear Coolant and Spallation Source Target28
LBE as a Nuclear Coolant and Spallation Source Target
Opportunity
Excellent neutron yield
Low neutron reaction cross sections
Low melting point
Excellent thermal properties
Not susceptible to radiation damage
Challenge
Highly corrosive to steel
Liquid Metal Embrittlement
Liquid Metal Enhanced Creep
XT-ADS, EU
MYRRHA
SVBR-75, Russia
Hyperion, US

29. Conventional and Innovative Materials Selection
Ferritic/Martensitic steels and graded composites
12 Cr F/M HT-9 steel and Russian EP823 steel (LANL)
New HT composite tube integrated to Delta Loop; provided by R. Ballinger (MIT, USA)
Oxide dispersed strengthened (ODS) and model alloys
PM 2000 and MA956 ODS steels and FeSi and FeSi model alloys (LANL)
Commercially available Fe-Cr-Al alloys
Kanthal Series steels (UCB, Berkeley)
APM, APMT, Alkrothal 14, Alkrothal 720 for high temperature applications
Alkrothal 3 (UU, KTH, Sweden)
Other possible candidate core materials
Proposed for fuel cladding, heat exchangers in ADS/LFRs
SCK.CEN, Belgium Collaboration
T91 9% Cr ferritic steel
D9 (DIN ) austenitic SS for in-vessel components
AISI 316 austenitic SS for vessel and in-vessel components

30. LANL Delta Loop Design and Capabilities
Design/Performance :
* Up to 2 m/s in 2.54 cm diameter ~ 3 m long test section
* Up to 100°C DT between heater section and heat exchanger exit
* Up to 550°C operation in the test section
* Capable of free convection flow
* All 316L construction
* Robust/Safe to operate continuously
* Gas injection system for oxygen control
DELTA isometric view

31. DELTA Corrosion Test Plan
* Corrosion resistance:
Exposure up to 3000h in flowing LBE at 500oC and Oxygen concentration wt%.
* Flow-rate resistance:
LBE velocities up to 3.5 m/s
* Grand total of 144 specimens tested
Thin test coupons are placed in a cylindrical holder that is lowered into the test section.
Canister loaded with 48 specimens
Gen-4 Module US Technological Impact: Understanding flow velocity and high-temperature effects on LBE steel corrosion properties for exposure times >2000h is critical in the conceptual design of advanced system 31

32. Conclusions and Future Work
* Achievements:
First specimens retrieved (May 2013) after ~ 900 h LBE exposure
Ongoing loading of second canister (2000 h) and preliminary corrosion studies
* Future Work:
Retrieval of specimens after 2000h and 3000h exposure
* Collaborations
UCB Post Exposure Studies:
Cross-section micro Raman, SEM/EDS/FIB/EBSD, nano-indentation studies
* Possible Future International Collaborations:
MYRRHA Accelerator Driven System - SCK.CEN, Belgium
ELECTRA European LBE Fast Reactor KTH, Uppsala Univ., Sweden
Publications: Contribution submitted to JOM August and Materials Selection Milestone LANL Reports
LANL Science Highlights PADSTE AOT MST (Oct. 2012)
F. Rubio et al., Rio Grande Symp. on Advanced Materials RGSAM 2012 32