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Development of Improved Radiation Tolerant Ferritic Steels for Fusion Applications

This research project focuses on the development of radiation-tolerant ferritic steels for use in fusion applications. The goal is to improve the mechanical properties and reduce radiation damage in these materials. The project involves various contributors from different research organizations.

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Development of Improved Radiation Tolerant Ferritic Steels for Fusion Applications

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  1. Development of Improved Radiation Tolerant Ferritic Steels for Fusion Applications Stuart A. Maloy Advanced Reactor Core Materials Technical Lead Nuclear Technology Research and Development Program LANL

  2. Contributors • ATI Inc.- M. Ferry, Jean Stewart • Kharkov Institute of Physics and Technology- V.V. Bryk • UCB: P. Hosemann • CWRU: J.J. Lewandowski • PSI: Y. Dai • Texas A&M: Lin Shao (NEUP) • UCSB: G.R. Odette, N. Cunningham, T. Yamamoto, M. E. Alam, S. Pal (NEUP) • LANL: Tarik Saleh, Toby Romero, Matthew Quintana, Bill Crooks, Ed Garcia, Osman Anderoglu (now at UNM), E. Aydogan, T. Lienert,B. Eftink • PNNL: Mychailo Toloczko, David Senor, Curt Lavender, T. S. Byun • INL: Jim Cole, Randy Fielding, Jian Gan, Mitch Meyer, Bulent H. Sencer, Emmanuel Perez, Michael Teague • ORNL: David Hoelzer • Radiation Effects Consulting: F. Garner

  3. Outline • Materials in Nuclear Reactors • Radiation Damage as a Function of Temperature • Improving ductility after low temperature irradiation • TMS • NFAs • Reducing void swelling and Helium Embrittlement • Update on present research efforts • Tube development studies • High dose irradiations underway

  4. Radiation Damage can have Significant Effects on Mechanical Properties Later, but still qualitative, version of the displacement spike (also called depleted zone). Original model of the displacement spike, J.A. Brinkman, Amer. J. Phys 24 (1956) 24. Displacement damage occurs when enough energy (approximately 25 eV) is transferred to an atom producing a or many Frenkel defects. Energy can be transferred from many different particles (neutrons, protons, fission products, etc). Particles with higher energies (>5 MeV) can cause spallation in materials.

  5. Materials in nuclear systems can fail 53 dpa, V/V=28% 52 dpa, V/V=30% 34 dpa, V/V=14% Fast Reactor Duct Failure Grid-to-Rod Fretting CRUD Davis-Besse Reactor Vessel Head Degradation

  6. Operating Temperature Windows for Structural Alloys 10 to 50 dpa Most Promising Alloys ° • Lower temperature limit of alloys based on radiation hardening/ fracture toughness embrittlement (K1C< ~30 MPa-m1/2) • Upper temperature limit based on 150 MPa creep strength (1% in 1000 h); chemical compatibility considerations may cause further decreases in the max operating temp. S.J. Zinkle and N.M. Ghoniem, Fus. Eng. Des. 51-52 (2000) 55; S.J. Zinkle et al. STAIF2002

  7. Fundamentals of Void Swelling • Void Swelling has an incubation period, transient period followed by a steady state regime. • During the incubation and transient periods void nucleation occurs. This requires: • Helium or other gas accumulation • Dislocation evolution • Precipitate-microchemical evolution • These periods can be affected by initial dislocation density, impurity levels and dpa rate. • During the steady state period, a parity exists between dislocation density and void density. Dislocations attract interstitials slightly more than vacancies (called the net bias) allowing the voids to grow in size. • The net bias determines the maximum swelling rate and depends on interstitial and vacancy relaxation volumes and tends to be larger for FCC metals (~1%/dpa) than for BCC metals (~0.2%/dpa) • In general void swelling occurs between 1/3 and 3/5 of the melting point of an alloy (stainless steel ~300-700C).

  8. Effects of Initial Cold Work on Swelling for 304 Austenitic Stainless Steel • Larger amount of cold work tends to increase incubation period, (Dupouy et al., 1978, Busboom et al., 1975).

  9. 3 2 5 4 1 The ACO-3 duct was analyzed after irradiation in the Fast Flux Test Facility FFTF, Hanford site, WA

  10. Mechanical Test Results on ACO-3 Duct Show strong Effects of Irradiation Temperature 25 C 200 C Tirr 0.2% Offset Yield Stress (MPa) HT-9 Control Irradiation Temperature (C)

  11. Improvements in Low Temperature Ductility

  12. Improved Radiation Response of New NQA1 Heat of HT-9 • 300 lb heat of HT-9 produced by Metalwerks following NQA-1 quality control • Tensile specimens irradiated in ATR to 6 dpa at 290°C • Hardening observed but excellent ductility retained after low temperature irradiation • Ion irradiations performed to 600 dpa at 425°C • Minimal swelling observed in tempered martensitic grains after ion irradiation to >500 dpa. INL-HT-9 Heat, best ductility 290°C 6 dpa INL (NSUF), LANL, UCSB Maloy, S.A., Saleh, T.A., Anderoglu, O., Romero, T.J., Odette, G.R., Yamamoto, T., Li, S., Cole, J.I., Fielding, R., JNM, 468 (2016) 232-239.

  13. Ductility Retention in MA957 after irradiation to 6 dpa at 290C LANL, UCSB, NSUF

  14. MA-957 tested at Room Temperature after STIP-V Irradiation to 14.6 dpa maximum dose. • Greatest hardening at lowest irradiation temperature (t=135C, dose=6.1 dpa) • Large ductility (UE > 5%) observed for all specimens. • High hardening and work hardening rate observed for sample irradiated to 14.6 dpa at 479 C may be caused by high helium content ( 875 appm He) Dose 5.8 dpa 6.1 dpa 10 dpa 14.6 dpa Tirr 153 C 135 C 305 C 479 C Control LANL, PSI

  15. Improvements in Void Swelling Resistance

  16. Improved Radiation Response of New NQA1 Heat of HT-9 • 300 lb heat of HT-9 produced by Metalwerks following NQA-1 quality control • Tensile specimens irradiated in ATR to 6 dpa at 290°C • Hardening observed but excellent ductility retained after low temperature irradiation • Ion irradiations performed to 600 dpa at 425°C • Minimal swelling observed in tempered martensitic grains after ion irradiation to >500 dpa. INL heat ferrite (<5% volume fraction) 33% CW heat – tempered martensite 425-450°C INL heat tempered martensite LANL, PNNL, KIPT

  17. Ion Irradiation Induced Swelling Comparisons • All materials irradiated in the same facility. • Tempered F-M steels reached a terminal 0.2%/dpa swelling rate that has been observed ferritic alloys during neutron irradiations. • MA957 and 14YWT are exhibiting an extended nascient low swelling period. • Data suggests that 14YWT is exhibiting better resistance than MA957, but results are very similar. • Will 14YWT and MA957 abruptly transition to high swelling rate? PNNL, KIPT, LANL, UCSB, ORNL

  18. Updates on Present Research

  19. Improved Radiation Response of New NQA1 Heat of HT-9 • 300 lb heat of HT-9 produced by Metalwerks following NQA-1 quality control • Tensile specimens irradiated in ATR to 6 dpa at 290°C • Hardening observed but excellent ductility retained after low temperature irradiation • Ion irradiations performed to 600 dpa at 425°C • Minimal swelling observed in tempered martensitic grains after ion irradiation to >500 dpa. • Four new heats of HT-9 were produced: Two by Metalwerks and two by Sophisticated Alloys with controlled interstitial content. INL heat ferrite (<5% volume fraction) INL-HT-9 Heat, best ductility 33% CW heat – tempered martensite 425-450°C 290°C 6 dpa INL heat tempered martensite INL (NSUF), LANL, UCSB Maloy, S.A., Saleh, T.A., Anderoglu, O., Romero, T.J., Odette, G.R., Yamamoto, T., Li, S., Cole, J.I., Fielding, R., JNM, 468 (2016) 232-239.

  20. Complex Component Produced from T91 using Additive Manufacturing LANL (LDRD)

  21. Fine Microstructure Observed for Normalized and Tempered AM T91 • Note the finer prior austenite grain size and martensite lath size of the N&T AM deposit. N&T AM Deposit N&T Wrought LANL (LDRD)

  22. Additively Manufactured Mod9Cr-1Mo has fine distribution of gas bubbles 0.14% porosity 19 ± 9 nm diameter average size 0.09% porosity 24 ± 6nm diameter average size Normalized and Tempered Tempered LANL (LDRD)

  23. Nanostructured Ferritic Alloys • Strength & damage resistance derives from a high density Ti-Y-O nano-features (NFs) (6 x 1023/m3) • 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) • Nanostructured ferritic alloys also have fine grains and high dislocation densities Y-YO-Ti-TiO-O UCSB, LANL, ORNL

  24. HIP near net shape final product Typical Processing Route for NFA Alloys • Any desired combination of powders: metals, alloys, and dispersoid, such as oxides, carbides, borides, etc. (Fe-14Cr-3W-0.4Ti) Y2O3 Zoz The conventional approach is to ball mill alloy and Y2O3 powders together UCSB, LANL, ORNL

  25. Production of plates for tube fabrication and joining studies • Bars cross-rolled at 1000ºC to 50% reduction in thickness • Procedures • Material: FCRD-NFA1 and 14YWT (SM13) heat • Cans filled with ball milled powder and extruded through rectangular die • Bars cut and then annealed at 1000ºC in vacuum 10 mm thick plate with can Extruded Bars Decanned plates ORNL, UCSB, LANL

  26. Fabrication of thin plates of 14YWT for FSW joining studies • Starting plate thickness of 10 mm (~50% cross-rolled) • Cross-rolled at 1000ºC to 5 mm thickness (50 % RIT) and cut in half • One half was parallel-rolled at 1000ºC to t = 2.2 mm thickness (56% RIT) and cut in half • One half was cross-rolled at 1000ºC to ~1 mm thickness (54.5% RIT) and cut in half • ~95% RIT total with no edge cracking t = 1.0 mm t = 5 mm t = 2.2 mm ORNL, UCSB, LANL

  27. Production of rods of FCRD-NFA1 for tube fabrication studies First two extruded rods • Three extruded rods were produced • Rods were extruded through 1.75 in. dia. die • Rods were cut to reveal ODS sections • Two longest rods had slight bows • Straightened by hot pressing at 1000ºC • Rods were decanned • Final diameter of rods were 1.0 - 1.1 in. • Final lengths were: 17.2 in. long x 1.0 in. diameter • 17.2 in. • 09.7 in. • 04.7 in. Sent to Japan • Three rods were recently shipped to LANL for machining core section to produce thick wall mother tubes ORNL, UCSB, LANL

  28. Hot Hydrostatic Extrusion at CWRU Hot Extrusion Parameters • Hydrostatic EXTRUSION TEMP: 1500F (815C) • RAM SPEED: 0.5 in/min, however 1st 0.5” of extrusion, speed was 0.7 in/min • SOAK TIME: 10 min • OVERALL EXTRUSION: 25 min • ER: 4:1, 45 DEG TAPER DIE (actual 0.495 diam) Ram CWRU

  29. Final NFA Extruded tubes after Etching to remove liner and mandrel 025 026 027 Start of extrusion CWRU, LANL

  30. Joining study: FSW bead-on-plate on 1 mm thick plate of 14YWT • Bead-on-plate stir zone (SZ) using modified pin tool design • Microstructure characterization (SEM, S/TEM) • Grains in SZ: similar in size but isotropic compared to grains in rolled plate • Ti-Y-O particles in SZ: slight increase in size Stir zone • Tensile properties (DIC) 100 nm Stir Zone Gauge: 3.55x0.8x0.4-0.6t(mm) ORNL

  31. Joining study: DIC analysis of stir zone in 1 mm thick 14YWT plate • Stir zone (SZ) is softer than plate in unaffected zone (UZ) • However, only small differences in the ductility Plate Stir Zone Stir zone • Location of 3 SS-Mini-2 specimens in stir zone Engineering stress – plastic strain curves Stir Zone • Specimen #2 failed in the elastic region during the tensile test ORNL

  32. ACO3 Duct MATRIX Significant data has been obtained on previously irradiated materials. How do we obtain data to dose levels out to 400 dpa? How do we get here?? International Collaborations New irradiations in CEFR FFTF/MOTA extended irradiations in other irradiation facilities (BOR60). Model Development High Dose Ion irradiations FFTF/MOTA Specimens NSMH Data Irradiation Temperature (C) Dose (dpa)

  33. BOR-60 Re-Irradiation of Samples through Collaboration with Terrapower • Goal: Obtain High dose irradiation data on ferritic/martensitic steel, HT-9 up to 450 dpa through re-irradiation of samples previously irradiated in FFTF • Irradiation Temperatures (370, 400, 450, 525 and 600°C) • Pre-irradiation doses (9-155 dpa) at irradiation temperatures (370-630°C). Additional irradiation will be performed at original irradiation temperature (within 10-20°C) • Specimens • Pressurized tubes (50) • TEM (420) • Tensile (148) • Charpy specimens (14) • Compact Tension (20) • 3 pt. Bend (60) • Target dose up to 450 dpa • Post Irradiation Examinations (to be performed at LANL, ORNL and PNNL) LANL, Terrapower

  34. Summary • Improved Radiation Tolerant Materials • Retention of low temperature ductility observed in new heats of HT-9 • Retention of low temperature ductility observed in MA-957 (ferritic ODS) after Low Temperature Reactor (ATR) and Spallation (SINQ) irradiations • Ion irradiations show low void swelling in HT-9 and 14YWT to doses up to 500 dpa with best void swelling resistance in 14YWT • Fabrication Studies • Thin walled tubing and sheet material being produced out of new heat of 14YWT through LANL, UCSB and ORNL collaboration. • Complex components being made by additive manufacturing out of T91 with fine microstructure in normalized and tempered material. • High Dose Irradiations • Very high dose irradiations underway in BOR-60 to obtain irradiation data out to 450 dpa in previously irradiated HT-9 • Facilities with prototypic spectra and neutron flux levels are needed to test these materials. UCSB, LANL, ORNL, PNNL, KIPT, Terrapower, PSI

  35. 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

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