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
DOE-NE is Performing Research to close the Nuclear
Fuel Cycle – Fuel Cycle R&D (FCRD) Program
2
Advanced Fuels Campaign Mission & Objectives in the Fuel
Cycle Research and Development Program
•
•
Mission
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
OnceThrough
Modified
Open
Continuous
Recycle
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)
Scientific Approach to Enabling a Multi-fold Increase in Fuel
Burnup over the Currently Known Technologies
Ultra-high
Burnup
Fuels
Enhancements with
Fabrication Complexity
Different Reactor
options to change
requirements
LFR, GFR
Radiation
200 dpa
500 C
Corrosion
300 dpa
400 dpa
600 C
700 C
Advanced Alloys
F/M Steels
Advanced Alloys
ODS Steels
Temperature
Advanced F/M
Steels, e.g. NF616
F/M Steels
HT-9
Advanced Alloys
Liners
Coating
FCCI
Cr
Si
Al
Increasing content
Reduced embrittlement, swelling, creep
Enhancements with
Enhancements with Fabrication Complexity Fabrication Complexity
4
Develop an advanced materials immune to fuel, neutrons and coolant interactions
under specific reactor environments
Outline
•
•
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
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
Hot-Cells for
World’s first petaFLOPS supercomputer
Processing of
Fuels Research Lab:
Isotopes
R&D on Ceramic Fuels •Actinide Cross-Section Measurements
Nuclear Fuels R&D: MOX
Fuel Fabrication and Testing
• Irradiation of Targets for Isotope
Production
DELTA loop for LBE Corrosion
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.
ACO-3 HT-9 6E9 Tensile tests
Tirr=381°C, dose=22dpa
1100
1000
25C
900
200C
800
350C
Engineering Stress (MPa)
16 mm
700
600
500
400
300
200
100
0
0
5
10
15
Engineering Strain (%)
Testing was Performed on the ACO-3 Duct,
one of the most highly irradiated components in a fast reactor
5
FFTF, Hanford site, WA
4 3 2
1
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
9
Stress/Strain Curves – HT-9
Irradiated, Room Temp Tests
ACO3 HT-9 25C Tensile Tests
Tirr, Dose
381 C, 21 dpa
398C, 42 dpa
416C, 110 dpa
1000
Tirr, Dose
441C, 147 dpa
466C, 92 dpa
504C, 3 dpa
900
Control
800
700
Stress (MPa)
• Decreased
elongation in
irradiated
materials.
• Increased
hardening in
lower
irradiation
temperature
materials.
1100
600
500
400
300
200
100
0
0
5
10
15
Strain (%)
20
25
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
1200
0.2% Offset Yield Stress (MPa)
•
25 C
200 C
Tirr
1000
800
600
400
HT-9 Control
200
0
300
350
400
450
500
Irradiation Temperature (C)
550
11
Charpy Impact Testing for ACO-3 Duct
7
300
2E5-L(503°C,3.2dpa)
L-T Orientation
T-1J (L-T)
T-2J (L-T)
T-1J (T-L)
T-2J (T-L)
DBTT (L-T)
DBTT (T-L)
2E4-L(503°C,3.1dpa)
6C8-L(380°C,20.5dpa)
6
250
4C3-L(467°C,87.9dpa)
Absorbed energy, J
5C3-L(438°C,147.6dpa)
6C5-L(398°C,41.7dpa)
4
6C6-L(396°C,38.8dpa)
6C3-L(412°C,95.8dpa)
6C2-L(396°C,38.8dpa)
3
6C9-L(382°C,23.3dpa)
Non-irr. (Ht 84425)
2
Non-irr. (INL QA 145728)
1
o
4C2-L(465°C,95.7dpa)
Transition temperature, C
5C2-L(441°C,146.9dpa)
5
200
150
100
50
0
-50
0
-200
-100
0
100
200
Temperature, oC
300
400
-100
350
Non-irradiated
400
450
500
o
Irradiation temperature, C
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.
550
TEM analysis of ACO-3 Duct Material (B.H. Sencer,
INL, O. Anderoglu, J. Van den Bosch, LANL)
T=384C, 28 dpa
• G-phase
precipitates and
alpha prime
observed
•No void swelling
observed.
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.
Small Angle Neutron
Scattering Measurements
Obtain accurate measurement
of ’ vs. dose and irr.
Temperature
Measurements completed on 5
specimens from ACO-3 duct
Summary
Cr
solub
T
dpa
at %
°C
(G.
Bonny)
α′
G-phase
Loops
voids
Cr solub
at %
(Phase
diagram)
L
nm
d
x1021
m-3
L
nm
d
x1021
m-3
L
nm
d
X1021
m-3
L
nm
d
X1021
m-3
380
20
9
8.1
7.8
72
11.3
9.3
14
0.93
X
X
410
100
9.1
10.1
9
2.6
16.2
1.4
-
-
23
-
440
155
9.3
12.4
9.6
9.5
26.5
1.1
18
0.5
28
0.25
466
92
9.9
14.7
X
X
X
X
X
X
505
2
12.5
18.5
X
X
X
X
X
X
475
10,000
hrs
10.3
15.5
X
X
X
X
X
X
Slide 14
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 9Cr1Mo, 9Cr-2WVTa, T122, 5Cr-2WVTa,
A21N, ODS strengthened F/M steels12YWT and 14YWT (Fe-12Cr-3W0.4Ti-0.25 Y203, Fe-14Cr-3W-0.4Ti0.25 Y203), V-4Cr-4Ti, High purity Ta,
single crystal Fe (for modelling studies)
– Fuels Matrices: ZrN, NiAl, FeAl, RuAl,
MgO, Cubic ZrO2, Fissium
15
~570 MeV protons
He/dpa ratios of 50-60 appm/dpa
Summary of STIP IV Tensile Testing at
Room Temperature
STIP-IV TENSILE TESTS - HT-9
12YWT STIP-IV TENSILE TESTS
1800
1800
Tirr=247C, dose=15 dpa
1700
STIP-IV, 12YWT FA01, 247, 14.6, 16C
1700
STIP-IV, 12YWT control FA03,0, 0, 17C
STIP-IV, 12YWT FA04, 394, 21.8, 17C
1600
1600
1400
1400
1300
1300
1200
1200
1100
1100
Stress (MPa)
Tirr=394C, dose=22 dpa
1500
Stress (MPa)
1500
1000
Tirr=120C, dose=8 dpa
Tirr=247C, dose=15 dpa
Tirr=380C, dose=22 dpa
1000
900
12YWT-Control
800
900
800
700
700
600
600
500
500
400
400
300
300
200
200
100
100
0
HT-9-Control
0
0
5
10
Strain (%)
•
•
•
•
15
20
0
5
10
15
Strain (%)
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
20
Requirements for Accident Tolerant
Fuels (ATF)
Measurements on hydrogen evolution performed
in steam
Zirc-4 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
304 in N2
containing ~25%
water vapor to
1100°C
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
19
Oxide Dispersion Strengthened Alloys
•
•
•
•
Strength & damage resistance derives from a high density TiY-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
Typical Processing Route for ODS Alloys
Hot consolidation (extrusion)
Canning
Alloy powder
As extruded bar
MA powder
Y2O3
Ball milling
Heat treating
Working
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.
ODS Strengthened Materials Show Excellent Resistance to
Hardening under Ion Irradiation
Room temperature irradiation (1.5dpa)
HT-9 Martensitic
10.00
5
8.00
4
Dose [dpa]
Nanohardness [GPa]
6
6.00
3
HT-9 Ferritic
4.00
2
Dose
1
2.00
0
ion beam
0.00
0
1
2
HT-9 Ferritc RT
3
4
5
HT-9 Martensitic RT
6
7
8
Depth [mm]
dose
MA957
Dose [dpa]
Nanohardness [MPa]
Berkovich indenter, 200nm deep indents, constant displacement
MA956
4mm
4mm
10mm
Beam
1mm
Depth [mm]
Analysis of highly irradiated MA-957
Tubes Underway at PNNL
MA957 TX series - room temperature tests
tensile specimens cut from
unstressed creep tubes
385°C
1600
1400
 Tensile testing of MA-957
Pressurized tubes
–
Irradiation conditions in FFTF
– (385°C, 18-43 dpa)
– (412°C, 110 dpa)
– (500-550°C, 18-113 dpa)
– (600-670°C, 34-110 dpa)
– (750°C, 33-120 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.
1200
Stress (MPa)
–
550°C
495°C
603°C
670°C
unirradiated 750°C
1000
800
TX unirr
TX12 385°C, 43 dpa
TX15 495°C, 48 dpa
TX16 550°C, 71 dpa
TX18 603°C, 83 dpa
TX19 670°C, 103 dpa
TX20 750°C, 78 dpa
600
0.2% strain line
400
200
0
0
2
4
6
8
10
Strain (%)
12
14
16
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 h @ 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
Core Materials Research and Development –
5 Year Plan
Qualify HT-9 for high dose clad/duct applications (determine design limitations)
Rev. 6 of AFCI (FCRD) Materials Handbook
FFTF (ACO-3 and MOTA) Specimen Analysis
Re-irradiation of FFTF specimens in BOR-60
Data to 250300 dpa on
F/M and 100150 on Inn.
Material
Advanced Material Development (improved radiation resistance to >400 dpa)
Materials Test Station Irradiations
STIP- IV (PSI) Specimen PIE
Innovative Material Development
MATRIX-SMI and 2 (Phenix) Specimen PIE
Data on Advanced Materials to 80-100 dpa
Baseline Mech. Prop. Of Inn. Clad Material
Innovative Clad Material Downselect
Develop ODS Tubing and Weld specifications for innovative Weld material
ODS Ferritic Steel Material Development
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
Fab. Innovative Coated Tube for ATR irradiation
Innovative Coating Material Development
FY’11
FY’12
PIE on Lined Irradiated Tube
FY’13
Provides data for NEAMS model development of Cladding
FY’14
FY’15
FY’16
26
Development of a Physics-based Model
of Radiation Damage in Nuclear Fuels
– 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).
Motivation/Approach
Analysis of segregation in terms of local strain
Compositional
variations
All results
provided
to MBM
Bulk diffusion
Bulk thermal
conductivity
7/20/09
MARMOT model
development;
mesoscale fission
gas diffusion and
Role of idealized grain segregation,
boundaries in mass and thermal
conductivity (INL)
thermal transport.
28
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
XT-ADS, EU
MYRRHA
Challenge
Highly corrosive to steel
Liquid Metal Embrittlement
Liquid Metal Enhanced Creep
SVBR-75, Russia
Hyperion, US
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 1.4970) austenitic SS for in-vessel components
– AISI 316 austenitic SS for vessel and in-vessel components
29
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
30
DELTA Corrosion Test Plan
* Corrosion resistance:
Exposure up to 3000h in flowing LBE at 500oC
and Oxygen concentration 10-6-10-8
wt%.
* Flow-rate resistance:
LBE velocities up to 3.5 m/s
Thin test coupons are placed in a
cylindrical holder that is lowered into
the test section.
* Grand total of 144 specimens tested
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
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 2013 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