The effects of microstructures and radiation damage on the deformation behavior of a HT-9 alloy using microtensile testing

Vo, H. T.; Anderoglu, O.; Hosemann, P.; Maloy, S. A.

doi:10.1016/j.matchar.2021.110972

The effects of microstructures and radiation damage on the deformation behavior of a HT-9 alloy using microtensile testing

Journal Article · Sat Feb 20 23:00:00 EST 2021 · Materials Characterization

DOI:https://doi.org/10.1016/j.matchar.2021.110972· OSTI ID:1770079

Vo, H. T. ^[1]; Anderoglu, O. ^[2]; Hosemann, P. ^[1]; ^[3]

Univ. of California, Berkeley, CA (United States)
Univ. of New Mexico, Albuquerque, NM (United States)
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

Understanding the influence of radiation damage on the mechanical properties of HT-9 and other tempered martensitic alloys is part of the mission in developing radiation-tolerant materials for the next generation of nuclear reactors. Although there has been extensive data on the macroscopic mechanical properties of the irradiated HT-9 alloys and microstructural changes, it is not well-understood how radiation damage and the resulting microstructural changes influence the local mechanical properties. In this study, we utilized in situ SEM microtensile testing to investigate the deformation behavior of specific martensitic boundaries in 1 dpa proton-irradiated HT-9 and answered the question of what is the weakest link. Additionally, we provided the direct observation of the failure modes of microtensiles containing high angle and low angle martensitic boundaries. In the unirradiated condition, the deformation is ductile and no martensitic boundary failure is observed. In the irradiated condition, the high angle martensitic boundaries are more susceptible to radiation-induced boundary failure as compared to the low angle martensitic boundaries.

View Accepted Manuscript (DOE)

Research Organization:: Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE). Fuel Cycle Technologies; USDOE National Nuclear Security Administration (NNSA)

Grant/Contract Number:: 89233218CNA000001; NA0003180

OSTI ID:: 1770079

Alternate ID(s):: OSTI ID: 1778607

Report Number(s):: LA-UR--21-22121

Journal Information:: Materials Characterization, Journal Name: Materials Characterization Vol. 174; ISSN 1044-5803

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (38)

Review on slip transmission criteria in experiments and crystal plasticity models Bayerschen, E.; McBride, A. T.; Reddy, B. D. Journal of Materials Science, Vol. 51, Issue 5 https://doi.org/10.1007/s10853-015-9553-4	journal	November 2015
Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo Metallurgical and Materials Transactions A, Vol. 44, Issue S1 https://doi.org/10.1007/s11661-012-1565-y	journal	December 2012
The structure of high-angle grain boundaries Brandon, D. G. Acta Metallurgica, Vol. 14, Issue 11 https://doi.org/10.1016/0001-6160(66)90168-4	journal	November 1966
An overview of neutron irradiation effects in LMFBR materials Straalsund, J. L.; Powell, R. W.; Chin, B. A. Journal of Nuclear Materials, Vol. 108-109 https://doi.org/10.1016/0022-3115(82)90499-8	journal	July 1982
Irradiation creep and swelling of the US fusion heats of HT9 and 9Cr-1Mo to 208 dpa at ~ 400°C Toloczko, M. B.; Garner, F. A.; Eiholzer, C. R. Journal of Nuclear Materials, Vol. 212-215 https://doi.org/10.1016/0022-3115(94)90131-7	journal	September 1994
Comparison of swelling and irradiation creep behavior of fcc-austenitic and bcc-ferritic/martensitic alloys at high neutron exposure Garner, F. A.; Toloczko, M. B.; Sencer, B. H. Journal of Nuclear Materials, Vol. 276, Issue 1-3 https://doi.org/10.1016/S0022-3115(99)00225-1	journal	January 2000
Structural materials for fission & fusion energy Zinkle, Steven J.; Busby, Jeremy T. Materials Today, Vol. 12, Issue 11 https://doi.org/10.1016/S1369-7021(09)70294-9	journal	November 2009
Materials challenges in nuclear energy Zinkle, S. J.; Was, G. S. Acta Materialia, Vol. 61, Issue 3 https://doi.org/10.1016/j.actamat.2012.11.004	journal	February 2013
In situ micro tensile testing of He+2 ion irradiated and implanted single crystal nickel film Reichardt, Ashley; Ionescu, Mihail; Davis, Joel Acta Materialia, Vol. 100 https://doi.org/10.1016/j.actamat.2015.08.028	journal	November 2015
On the relationship between grain boundary connectivity, coincident site lattice boundaries, and intergranular stress corrosion cracking Lehockey, E. M.; Brennenstuhl, A. M.; Thompson, I. Corrosion Science, Vol. 46, Issue 10 https://doi.org/10.1016/j.corsci.2004.01.019	journal	October 2004
Microstructural evolution of proton irradiated T91 Gupta, G.; Jiao, Z.; Ham, A. N. Journal of Nuclear Materials, Vol. 351, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2006.02.028	journal	June 2006
On the effects of irradiation and helium on the yield stress changes and hardening and non-hardening embrittlement of ∼8Cr tempered martensitic steels: Compilation and analysis of existing data Yamamoto, Takuya; Odette, G. Robert; Kishimoto, Hirotatsu Journal of Nuclear Materials, Vol. 356, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2006.05.041	journal	September 2006
Role of grain boundary engineering in the SCC behavior of ferritic–martensitic alloy HT-9 Gupta, G.; Ampornrat, P.; Ren, X. Journal of Nuclear Materials, Vol. 361, Issue 2-3 https://doi.org/10.1016/j.jnucmat.2006.12.006	journal	April 2007
Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155dpa at 443°C Sencer, B. H.; Kennedy, J. R.; Cole, J. I. Journal of Nuclear Materials, Vol. 393, Issue 2 https://doi.org/10.1016/j.jnucmat.2009.06.010	journal	September 2009
Radiation-induced segregation and phase stability in ferritic–martensitic alloy T 91 Wharry, Janelle P.; Jiao, Zhijie; Shankar, Vani Journal of Nuclear Materials, Vol. 417, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2010.12.052	journal	October 2011
Tensile and impact properties of 9Cr tempered martensitic steels and ODS-FeCr alloys irradiated in a fast reactor at 325°C up to 78dpa Henry, J.; Averty, X.; Alamo, A. Journal of Nuclear Materials, Vol. 417, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2010.12.203	journal	October 2011
Helium transport, fate and management in nanostructured ferritic alloys: In situ helium implanter studies Odette, G. R.; Miao, P.; Edwards, D. J. Journal of Nuclear Materials, Vol. 417, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2011.01.064	journal	October 2011
Core materials development for the fuel cycle R&D program Maloy, S. A.; Toloczko, M.; Cole, J. Journal of Nuclear Materials, Vol. 415, Issue 3 https://doi.org/10.1016/j.jnucmat.2011.04.027	journal	August 2011
Dependence on grain boundary structure of radiation induced segregation in a 9wt.% Cr model ferritic/martensitic steel Field, Kevin G.; Barnard, Leland M.; Parish, Chad M. Journal of Nuclear Materials, Vol. 435, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2012.12.026	journal	April 2013
A dual ion irradiation study of helium–dpa interactions on cavity evolution in tempered martensitic steels and nanostructured ferritic alloys Yamamoto, Takuya; Wu, Yuan; Robert Odette, G. Journal of Nuclear Materials, Vol. 449, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2014.01.040	journal	June 2014
Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa Maloy, S. A.; Saleh, T. A.; Anderoglu, O. Journal of Nuclear Materials, Vol. 468 https://doi.org/10.1016/j.jnucmat.2015.07.039	journal	January 2016
In situ micro-tensile testing on proton beam-irradiated stainless steel Vo, H. T.; Reichardt, A.; Frazer, D. Journal of Nuclear Materials, Vol. 493 https://doi.org/10.1016/j.jnucmat.2017.06.026	journal	September 2017
TEM characterization of irradiated microstructure of Fe-9%Cr ODS and ferritic-martensitic alloys Swenson, M. J.; Wharry, J. P. Journal of Nuclear Materials, Vol. 502 https://doi.org/10.1016/j.jnucmat.2018.01.062	journal	April 2018
Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: effect of dose Zheng, Ce; Reese, Elaina R.; Field, Kevin G. Journal of Nuclear Materials, Vol. 523 https://doi.org/10.1016/j.jnucmat.2019.06.019	journal	September 2019
Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature Zheng, Ce; Reese, Elaina R.; Field, Kevin G. Journal of Nuclear Materials, Vol. 528 https://doi.org/10.1016/j.jnucmat.2019.151845	journal	January 2020
Applying a new push-to-pull micro-tensile testing technique to evaluate the mechanical properties of high dose Inconel X-750 Howard, C.; Judge, C. D.; Hosemann, P. Materials Science and Engineering: A, Vol. 748 https://doi.org/10.1016/j.msea.2019.01.113	journal	March 2019
Experimental methodology and theoretical framework in describing constrained plastic flow of FCC microscale tensile specimens Vo, H. T.; Still, E. K.; Lam, K. Materials Science and Engineering: A, Vol. 799 https://doi.org/10.1016/j.msea.2020.140124	journal	January 2021
Srim-2003 Ziegler, James F. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, Vol. 219-220 https://doi.org/10.1016/j.nimb.2004.01.208	journal	June 2004
Plasticity in small-sized metallic systems: Intrinsic versus extrinsic size effect Greer, Julia R.; De Hosson, Jeff Th. M. Progress in Materials Science, Vol. 56, Issue 6 https://doi.org/10.1016/j.pmatsci.2011.01.005	journal	August 2011
Contribution to size effect of yield strength from the stochastics of dislocation source lengths in finite samples Parthasarathy, Triplicane A.; Rao, Satish I.; Dimiduk, Dennis M. Scripta Materialia, Vol. 56, Issue 4 https://doi.org/10.1016/j.scriptamat.2006.09.016	journal	February 2007
Small-scale mechanical testing on nuclear materials: bridging the experimental length-scale gap Hosemann, Peter Scripta Materialia, Vol. 143 https://doi.org/10.1016/j.scriptamat.2017.04.026	journal	January 2018
Micropillar compression response of femtosecond laser-cut single crystal Cu and proton irradiated Cu Gigax, Jonathan G.; Vo, Hi; McCulloch, Quinn Scripta Materialia, Vol. 170 https://doi.org/10.1016/j.scriptamat.2019.05.004	journal	September 2019
Multiscale modelling of plastic flow localization in irradiated materials Diaz de la Rubia, Tomas; Zbib, Hussein M.; Khraishi, Tariq A. Nature, Vol. 406, Issue 6798 https://doi.org/10.1038/35022544	journal	August 2000
In situ nanocompression testing of irradiated copper Kiener, D.; Hosemann, P.; Maloy, S. A. Nature Materials, Vol. 10, Issue 8 https://doi.org/10.1038/nmat3055	journal	June 2011
Dislocation Models of Crystal Grain Boundaries Read, W. T.; Shockley, W. Physical Review, Vol. 78, Issue 3 https://doi.org/10.1103/PhysRev.78.275	journal	May 1950
Recent Developments in Irradiation-Resistant Steels Odette, G. R.; Alinger, M. J.; Wirth, B. D. Annual Review of Materials Research, Vol. 38, Issue 1, p. 471-503 https://doi.org/10.1146/annurev.matsci.38.060407.130315	journal	August 2008
Microstructure and Microdeformation Effects on IGSCC of Alloy 600 Steam Generator Tubing Bruemmer, S. M.; Charlot, L. A.; Henager, C. H. CORROSION, Vol. 44, Issue 11 https://doi.org/10.5006/1.3584948	journal	November 1988
Irradiation Effects of ht-9 Martensitic Steel Chen, Yiren Nuclear Engineering and Technology, Vol. 45, Issue 3 https://doi.org/10.5516/NET.07.2013.706	journal	June 2013

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37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CHEMISTRY
Microtensile testing
interface failure
radiation damage
tempered-martensitic

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