Articles de revues : « Irradiation assisted stress corrosion cracking »

1

McNeil, M. B. « Irradiation assisted stress corrosion cracking ». Nuclear Engineering and Design 181, n^o 1-3 (mai 1998) : 55–60. http://dx.doi.org/10.1016/s0029-5493(97)00334-8.

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Kenik, E. A., R. H. Jones et G. E. C. Bell. « Irradiation-assisted stress corrosion cracking ». Journal of Nuclear Materials 212-215 (septembre 1994) : 52–59. http://dx.doi.org/10.1016/0022-3115(94)90033-7.

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Tsukada, Takashi. « Irradiation Assisted Stress Corrosion Cracking (IASCC) ». Zairyo-to-Kankyo 52, n^o 2 (2003) : 66–72. http://dx.doi.org/10.3323/jcorr1991.52.66.

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Kain, V., S. B. Chafle, D. Feron, B. Tanguy, C. Colin et C. Gonnier. « ICONE23-2044 IRRADIATION ASSISTED STRESS CORROSION CRACKING AND THE JULES HOROWITZ MATERIAL TEST REACTOR ». Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015) : _ICONE23–2—_ICONE23–2. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-2_18.

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Cui, Bai, Michael D. McMurtrey, Gary S. Was et Ian M. Robertson. « Micromechanistic origin of irradiation-assisted stress corrosion cracking ». Philosophical Magazine 94, n^o 36 (21 novembre 2014) : 4197–218. http://dx.doi.org/10.1080/14786435.2014.982744.

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Scott, P. « A review of irradiation assisted stress corrosion cracking ». Journal of Nuclear Materials 211, n^o 2 (août 1994) : 101–22. http://dx.doi.org/10.1016/0022-3115(94)90360-3.

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Was, Gary S., et Peter L. Andresen. « Irradiation-assisted stress-corrosion cracking in austenitic alloys ». JOM 44, n^o 4 (avril 1992) : 8–13. http://dx.doi.org/10.1007/bf03222812.

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Hojná, Anna. « Irradiation-Assisted Stress Corrosion Cracking and Impact on Life Extension ». CORROSION 69, n^o 10 (octobre 2013) : 964–74. http://dx.doi.org/10.5006/0803.

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Rossi, F., F. Fumagalli, A. Ruiz-Moreno, P. Moilanen et P. Hähner. « Membrane bulge test rig for irradiation-assisted stress-corrosion cracking ». Nuclear Instruments and Methods in Physics Research Section B : Beam Interactions with Materials and Atoms 479 (septembre 2020) : 80–92. http://dx.doi.org/10.1016/j.nimb.2020.06.012.

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Smith, Stuart A., Brock Gause, David Plumley et Masao J. Drexel. « Irradiation-Assisted Stress-Corrosion Cracking of Nitinol During eBeam Sterilization ». Journal of Materials Engineering and Performance 21, n^o 12 (17 octobre 2012) : 2638–42. http://dx.doi.org/10.1007/s11665-012-0396-8.

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Cookson, J. M., R. D. Carter, D. L. Damcott, M. Atzmon et G. S. Was. « Irradiation assisted stress corrosion cracking of controlled purity 304L stainless steels ». Journal of Nuclear Materials 202, n^o 1-2 (juin 1993) : 104–21. http://dx.doi.org/10.1016/0022-3115(93)90034-v.

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12

Zhou, Rongsheng, Elaine A. West, Zhijie Jiao et Gary S. Was. « Irradiation-assisted stress corrosion cracking of austenitic alloys in supercritical water ». Journal of Nuclear Materials 395, n^o 1-3 (décembre 2009) : 11–22. http://dx.doi.org/10.1016/j.jnucmat.2009.09.010.

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West, Elaine A., Michael D. McMurtrey, Zhijie Jiao et Gary S. Was. « Role of Localized Deformation in Irradiation-Assisted Stress Corrosion Cracking Initiation ». Metallurgical and Materials Transactions A 43, n^o 1 (20 juillet 2011) : 136–46. http://dx.doi.org/10.1007/s11661-011-0826-5.

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Kenik, E. A., J. T. Busby, M. K. Miller, A. M. Thuvander et G. Was. « Grain Boundary Segregation and Irradiation-Assisted Stress Corrosion Cracking of Stainless Steels ». Microscopy and Microanalysis 5, S2 (août 1999) : 760–61. http://dx.doi.org/10.1017/s1431927600017128.

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Irradiation-assisted stress corrosion cracking (IASCC) of irradiated austenitic stainless steels has been attributed to both microchemical (radiation-induced segregation (RIS)) and microstructural (radiation hardening) effects. The flux of radiation-induced point defects to grain boundaries results in the depletion of Cr and Mo and the enrichment of Ni, Si, and P at the boundaries. Similar to the association of stress corrosion cracking with the depletion of Cr and Mo in thermally sensitized stainless steels, IASCC is attributed in part to similar depletion by RIS. However, in specific heats of irradiated stainless steel, “W-shaped” Cr profiles have been observed with localized enrichment of Cr, Mo and P at grain boundaries. It has been show that such profiles arise from pre-existing segregation associated with intermediate rate cooling from elevated temperatures. However, the exact mechanism responsible for the pre-existing segregation has not been identified.Two commercial heats of stainless steel (304CP and 316CP) were forced air cooled from elevated temperatures (∽1100°C) to produce pre-existing segregation.

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Jacobs, A. J., G. P. Wozadlo et G. M. Gordon. « Low-Temperature Annealing : A Process to Mitigate Irradiation-Assisted Stress Corrosion Cracking ». CORROSION 51, n^o 10 (octobre 1995) : 731–37. http://dx.doi.org/10.5006/1.3293549.

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Was *, G. S., et J. T. Busby. « Role of irradiated microstructure and microchemistry in irradiation-assisted stress corrosion cracking ». Philosophical Magazine 85, n^o 4-7 (février 2005) : 443–65. http://dx.doi.org/10.1080/02678370412331320224.

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Konstantinović, M. J. « Probabilistic fracture mechanics of irradiation assisted stress corrosion cracking in stainless steels ». Procedia Structural Integrity 2 (2016) : 3792–98. http://dx.doi.org/10.1016/j.prostr.2016.06.472.

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Andresen, P. L., et F. P. Ford. « Use of Fundamental Modeling of Environmental Cracking for Improved Design and Lifetime Evaluation ». Journal of Pressure Vessel Technology 115, n^o 4 (1 novembre 1993) : 353–58. http://dx.doi.org/10.1115/1.2929541.

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This manuscript reviews an approach for improved design and lifetime evaluation for environmental cracking based on fundamental modeling of the underlying, operative processes in crack advance. In outlining this approach and its application in energy industries, the requirements for a life prediction methodology will be highlighted and the shortcomings of the existing design and lifetime evaluation codes will be discussed. Examples will be given of its development and application in a variety of cracking systems, such as environmental cracking of stainless steels and nickel alloys in hot water, and irradiation-assisted stress corrosion cracking.

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Tsukada, T., Y. Miwa, H. Tsuji et H. Nakajima. « Effect of irradiation temperature on irradiation assisted stress corrosion cracking of model austenitic stainless steels ». Journal of Nuclear Materials 258-263 (octobre 1998) : 1669–74. http://dx.doi.org/10.1016/s0022-3115(98)00317-1.

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Jenssen, A., L. G. Ljungberg, J. Walmsley et S. Fisher. « Importance of Molybdenum on Irradiation-Assisted Stress Corrosion Cracking in Austenitic Stainless Steels ». CORROSION 54, n^o 1 (janvier 1998) : 48–60. http://dx.doi.org/10.5006/1.3284828.

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FUKUYA, Koji, Morihito NAKANO, Katsuhiko FUJII et Tadahiko TORIMARU. « Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking ». Journal of Nuclear Science and Technology 41, n^o 5 (mai 2004) : 594–600. http://dx.doi.org/10.1080/18811248.2004.9715522.

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Jacobs, A. J., et G. P. Wozadlo. « Irradiation-assisted stress corrosion cracking as a factor in nuclear power plant aging ». Journal of Materials Engineering 9, n^o 4 (décembre 1988) : 345–51. http://dx.doi.org/10.1007/bf02834045.

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FUKUYA, Koji, Morihito NAKANO, Katsuhiko FUJII, Tadahiko TORIMARU et Yuji KITSUNAI. « Separation of Microstructural and Microchemical Effects in Irradiation Assisted Stress Corrosion Cracking using Post-irradiation Annealing ». Journal of Nuclear Science and Technology 41, n^o 12 (décembre 2004) : 1218–27. http://dx.doi.org/10.1080/18811248.2004.9726351.

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Wang, Rong-shan, Chao-liang Xu, Xiang-bing Liu, Ping Huang et Yu Chen. « The studies of irradiation assisted stress corrosion cracking on reactor internals stainless steel under Xe irradiation ». Journal of Nuclear Materials 457 (février 2015) : 130–34. http://dx.doi.org/10.1016/j.jnucmat.2014.11.019.

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Was, G. S., B. Alexandreanu et J. Busby. « Localized Deformation Induced IGSCC and IASCC of Austenitic Alloys in High Temperature Water ». Key Engineering Materials 261-263 (avril 2004) : 885–902. http://dx.doi.org/10.4028/www.scientific.net/kem.261-263.885.

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Grain boundary properties are known to affect the intergranular stress corrosion cracking (IGSCC) and irradiation assisted stress corrosion cracking behavior of austenitic alloys in high temperature water. However, it is only recently that sufficient evidence has accumulated to show that the disposition of deformation in and near the grain boundary plays a key role in intergranular cracking. Grain boundaries that can transmit strain to adjacent grains can relieve stresses without undergoing localized deformation. Grain boundaries that cannot transmit strain will either experience high stresses or high strains. High stresses can lead to wedge-type cracking and sliding can lead to rupture of the protective oxide film. These processes are also applicable to irradiated materials in which the deformation can become highly localized in the form of dislocation channels and deformation twins. These deformation bands conduct tremendous amounts of strain to the grain boundaries. The capability of a boundary to transmit strain to a neighboring grain will determine its propensity for cracking, analogous to that in unirradiated metals. Thus, IGSCC in unirradiated materials and IASCC in irradiated materials are governed by the same local processes of stress and strain accommodation at the boundary.

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Schmidt, Franziska, Peter Hosemann, Raluca O. Scarlat, Daniel K. Schreiber, John R. Scully et Blas P. Uberuaga. « Effects of Radiation-Induced Defects on Corrosion ». Annual Review of Materials Research 51, n^o 1 (26 juillet 2021) : 293–328. http://dx.doi.org/10.1146/annurev-matsci-080819-123403.

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The next generation of nuclear reactors will expose materials to conditions that, in some cases, are even more extreme than those in current fission reactors, inevitably leading to new materials science challenges. Radiation-induced damage and corrosion are two key phenomena that must be understood both independently and synergistically, but their interactions are often convoluted. In the light water reactor community, a tremendous amount of work has been done to illuminate irradiation-corrosion effects, and similar efforts are under way for heavy liquid metal and molten salt environments. While certain effects, such as radiolysis and irradiation-assisted stress corrosion cracking, are reasonably well established, the basic science of how irradiation-induced defects in the base material and the corrosion layer influence the corrosion process still presents many unanswered questions. In this review, we summarize the work that has been done to understand these coupled extremes, highlight the complex nature of this problem, and identify key knowledge gaps.

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Kaji, Yoshiyuki, Yukio Miwa, Keietsu Kondo et Nariaki Ohkubo. « 433 Simulation of irradiation assisted stress corrosion cracking initiation behavior by considering influence of residual stress ». Proceedings of The Computational Mechanics Conference 2008.21 (2008) : 742–43. http://dx.doi.org/10.1299/jsmecmd.2008.21.742.

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Kai, Ji-Jung, Chuen-Horng Tsai et Wen-Jen Liu. « Microstructural aspect of irradiation assisted stress corrosion cracking of LWR in-core structural materials ». Radiation Effects and Defects in Solids 144, n^o 1-4 (juin 1998) : 171–88. http://dx.doi.org/10.1080/10420159808229675.

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Chung, H. M., W. E. Ruther, J. E. Sanecki, A. Hins, N. J. Zaluzec et T. F. Kassner. « Irradiation-assisted stress corrosion cracking of austenitic stainless steels : recent progress and new approaches ». Journal of Nuclear Materials 239 (décembre 1996) : 61–79. http://dx.doi.org/10.1016/s0022-3115(96)00677-0.

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McMurtrey, M. D., G. S. Was, L. Patrick et D. Farkas. « Relationship between localized strain and irradiation assisted stress corrosion cracking in an austenitic alloy ». Materials Science and Engineering : A 528, n^o 10-11 (avril 2011) : 3730–40. http://dx.doi.org/10.1016/j.msea.2011.01.073.

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Song, Miao, Mi Wang, Xiaoyuan Lou, Raul B. Rebak et Gary S. Was. « Radiation damage and irradiation-assisted stress corrosion cracking of additively manufactured 316L stainless steels ». Journal of Nuclear Materials 513 (janvier 2019) : 33–44. http://dx.doi.org/10.1016/j.jnucmat.2018.10.044.

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Hojna, Anna, Jan Michalicka et Ondrej Srba. « Fracture of Irradiated Austenitic Stainless Steel with Special Reference to Irradiation Assisted Stress Corrosion Cracking ». Key Engineering Materials 592-593 (novembre 2013) : 569–72. http://dx.doi.org/10.4028/www.scientific.net/kem.592-593.569.

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This paper deals with fracture of neutron irradiated austenitic Ti-stabilized stainless steel 08Ch18N10T. The steel had been tested in air and in water environment (320°C) using several tests representing different stress strain conditions for crack initiation and growth; Slow Strain Rate and Crack Growth Rate tests were performed in the water. Without irradiation the steel did not suffer from stress corrosion cracking in the water, but on irradiated specimens appeared areas of intergranular fracture mixed with transgranular cleavage-like facets and secondary cracks typical for IASCC phenomenon. The differences between fracture of irradiated and non-irradiated specimens in air and in water are documented and discussed.

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Jacobs, A. J., G. P. Wozadlo, K. Nakata, T. Okada et S. Suzuki. « Grain Boundary Composition and Irradiation-Assisted Stress Corrosion Cracking Resistance in Type 348 Stainless Steel ». CORROSION 50, n^o 10 (octobre 1994) : 731–40. http://dx.doi.org/10.5006/1.3293462.

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Jacobs, A. J., G. P. Wozadlo et G. M. Gordon. « Use of a Constant Deflection Test to Evaluate Susceptibility to Irradiation-Assisted Stress Corrosion Cracking ». CORROSION 49, n^o 8 (août 1993) : 650–55. http://dx.doi.org/10.5006/1.3316096.

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Miwa, Y., T. Tsukada, S. Jitsukawa, S. Kita, S. Hamada, Y. Matsui et M. Shindo. « Effect of minor elements on irradiation assisted stress corrosion cracking of model austenitic stainless steels ». Journal of Nuclear Materials 233-237 (octobre 1996) : 1393–96. http://dx.doi.org/10.1016/s0022-3115(96)00255-3.

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Konstantinović, M. J. « Internal oxidation and probabilistic fracture model of irradiation assisted stress corrosion cracking in stainless steels ». Journal of Nuclear Materials 495 (novembre 2017) : 220–24. http://dx.doi.org/10.1016/j.jnucmat.2017.08.018.

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Burke, M. G., et R. Bajaj. « Irradiation-induced precipitation in direct-aged alloy 625 ». Proceedings, annual meeting, Electron Microscopy Society of America 54 (11 août 1996) : 994–95. http://dx.doi.org/10.1017/s0424820100167433.

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Direct-aged alloy 625 (DA-A625) is a precipitation-hardened nickel-base superalloy with good mechanical properties and excellent corrosion resistance. Its superior stress corrosion cracking (SCC) performance makes it a candidate material for applications in light water reactors where resistance to irradiation-assisted SCC (IASCC) is important. This alloy derives its strength from the intragranular precipitation of fine DO22 – ordered γ” precipitates. These precipitates have a disc-like morphology and are crystallo-grahically related to the γ matrix such that the [001] axis of the precipitates are oriented parallel to <100> directions in the matrix. This alloy is generally solution-annealed then aged within the temperature range ~600 to 750°C. In direct-aging, the alloy is immediately aged at ~ 660°C for 80 h following hot-working. In addition to the γ° precipitates, intergranular M23C6, M7C3, M6C, or MC carbides can form in this material.

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Kenik, E. A., R. D. Carter, D. L. Damcott, M. Atzmon et G. S. Was. « AEM and AES of radiation-induced segregation in proton-irradiated stainless steels ». Proceedings, annual meeting, Electron Microscopy Society of America 52 (1994) : 962–63. http://dx.doi.org/10.1017/s0424820100172541.

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Irradiation-assisted stress corrosion cracking (IASCC) of stainless steels has been attributed in part to radiation-induced segregation (RIS) of both major alloying and impurity elements at grain boundaries. There are phenomenological similarities observed between IASCC and intergranular stress corrosion cracking (IGSCC) of thermally-sensitized stainless steels. One concern for both IGSCC and IASCC is the localized loss of corrosion resistance associated with chromium depletion at grain boundaries. In order to avoid complications related to the long-term, induced radioactivity of neutron-irradiated specimens, four type 304L alloys were irradiated to 1 dpa (displacements per atom) with 3.4 MeV protons at 400°C. Both analytical electron microscopy (AEM) in a Philips EM400T/FEG and Auger electron spectrometry (AES) in a Perkin Elmer (PHI) 660 were employed to measure composition at or near grain boundaries in unirradiated and irradiated specimens of four controlled purity alloys [ultra-high purity (UHP), UHP+S (0.03 at.%), UHP+P (0.08 at.%), and UHP+Si (0.87 at.%)]. A sufficient number of boundaries were analyzed via AEM and AES to result in the standard deviation of the mean boundary composition of less than -0.5 at.%. Further experimental details are presented elsewhere.

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Kenik, E. A., et M. G. Burke. « Segregation in a neutron-irradiated Type 316 stainless steel ». Proceedings, annual meeting, Electron Microscopy Society of America 50, n^o 2 (août 1992) : 1218–19. http://dx.doi.org/10.1017/s0424820100130729.

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Radiation-induced segregation (RIS) and associated irradiation-assisted stress corrosion cracking (IASCC) of austenitic alloys may be a major factor in limiting component lifetimes in water-cooled nuclear reactors. There are some similarities between radiation-induced sensitization/IASCC and thermally-induced sensitization/intergranular stress corrosion cracking. Both processes are associated with chromium depletion at grain boundaries. Segregation to boundaries in a neutron irradiated type 316 stainless steel has been investigated with both energy-dispersive X-ray spectrometry (EDXS) and parallel detection electron energy loss spectrometry (PEELS).All specimens were from the same heat of cold-worked type 316 stainless steel. Both unirradiated control material and material irradiated at ∼300°C to a range of fluences 0.3 - 5 × 1026 neutrons/m2 (E>0.1 MeV) were available. The mass of irradiated material was minimized by mechanically polishing 3-mm-diam. disks to ∼75 μm thickness prior to electropolishing. However, the specific radioactivity of the specimens, which increased with neutron fluence, limited the application of EDXS to the unirradiated and the lowest fluence irradiated materials.

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Paccou, Elie, Benoît Tanguy et Marc Legros. « Irradiation-assisted stress corrosion cracking susceptibility and mechanical properties related to irradiation-induced microstructures of 304L austenitic stainless steel ». Journal of Nuclear Materials 528 (janvier 2020) : 151880. http://dx.doi.org/10.1016/j.jnucmat.2019.151880.

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McMurtrey, M., C. Sun, R. E. Rupp, C. H. Shiau, R. Hanbury, N. Jerred et R. O'Brien. « Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility ». Journal of Nuclear Materials 545 (mars 2021) : 152739. http://dx.doi.org/10.1016/j.jnucmat.2020.152739.

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Tanguy, Benoît, Maxime Sauzay, Christian Robertson et Stéphane Perrin. « The Irradiation-Assisted Stress Corrosion Cracking (IASCC) Issue : some Examples of Studies Carried out at CEA ». EPJ Web of Conferences 51 (2013) : 04002. http://dx.doi.org/10.1051/epjconf/20135104002.

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43

Wang, Mi, Miao Song, Calvin R. Lear et Gary S. Was. « Irradiation assisted stress corrosion cracking of commercial and advanced alloys for light water reactor core internals ». Journal of Nuclear Materials 515 (mars 2019) : 52–70. http://dx.doi.org/10.1016/j.jnucmat.2018.12.015.

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Tsukada, T., S. Jitsukawa, K. Shiba, Y. Sato, I. Shibahara et H. Nakajima. « Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR ». Journal of Nuclear Materials 207 (décembre 1993) : 159–68. http://dx.doi.org/10.1016/0022-3115(93)90258-z.

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Deng, Ping, Qunjia Peng, En-Hou Han, Wei Ke et Chen Sun. « Proton irradiation assisted localized corrosion and stress corrosion cracking in 304 nuclear grade stainless steel in simulated primary PWR water ». Journal of Materials Science & ; Technology 65 (février 2021) : 61–71. http://dx.doi.org/10.1016/j.jmst.2020.04.068.

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Margolin, B. Z., A. A. Sorokin, N. E. Pirogova, V. A. Potapova, Aki Toivonen, Faiza Sefta et Cédric Pokor. « Model of corrosion cracking of irradiated austenitic steels. Part 1. Analysis of damage mechanisms and formulation of the defining ». Voprosy Materialovedeniya, n^o 2(98) (11 août 2019) : 154–77. http://dx.doi.org/10.22349/1994-6716-2019-97-1-154-177.

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Mechanisms having a potential effect on irradiation assisted stress corrosion cracking (IASCC) of austenitic steels in the LWR environment have been analyzed. Based on the analysis and generalization of reference and original data on IASCC, an IASCC initiation criterion has been formulated. Conditions for grainboundary microcrack propagation by IASCC mechanism have been formulated. The nature of low-temperature creep of irradiated austenitic steels has been considered, constitutive equations have been derived. Relying on the formulated criterion of grain-boundary microcrack nucleation and the derived creep equations, an IASCC initiation model has been developed. The model allows one to predict the dependence of the threshold stress σIASCC th on neutron dose and also to calculate the IASCC initiation time with stresses exceeding σIASCCth .

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Busby, J. T., G. S. Was et E. A. Kenik. « Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels ». Journal of Nuclear Materials 302, n^o 1 (avril 2002) : 20–40. http://dx.doi.org/10.1016/s0022-3115(02)00719-5.

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KAJI, Yoshiyuki, Yukio MIWA, Takashi TSUKADA, Hirokazu TSUJI et Hajime NAKAJIMA. « Status of JAERI Material Performance Database (JMPD) and Analysis of Irradiation Assisted Stress Corrosion Cracking (IASCC) Data ». Journal of Nuclear Science and Technology 37, n^o 11 (novembre 2000) : 949–58. http://dx.doi.org/10.1080/18811248.2000.9714977.

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49

Busby, J. T., E. A. Kenik et G. S. Was. « The Measurement of Light Element Segregation Using EDS and EELS ». Microscopy and Microanalysis 4, S2 (juillet 1998) : 772–73. http://dx.doi.org/10.1017/s1431927600023989.

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Radiation-induced segregation (RIS) is the spatial redistribution of elements at defect sinks such as grain boundaries and free surfaces during irradiation. This phenomenon has been studied in a wide variety of alloys and has been linked to irradiation-assisted stress corrosion cracking (IASCC) of nuclear reactor core components. However, several recent studies have shown that Cr and Mo can be enriched to significant levels at grain boundaries prior to irradiation as a result of heat treatment. Segregation of this type may delay the onset of radiation-induced Cr depletion at the grain boundary, thus reducing IASCC susceptibility. Unfortunately, existing models of segregation phenomena do not correctly describe the physical processes and therefore are grossly inaccurate in predicting pre-existing segregation and subsequent redistribution during irradiation. Disagreement between existing models and measurement has been linked to potential interactions between the major alloying elements and lighter impurity elements such as S, P, and B.

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Busby, J. T., T. R. Allen, E. A. Kenik, N. J. Zaluzec et G. S. Was. « Beam-Broadening Effects in STEM/EDS Measurement of Radiation-Induced Segregation in High-Purity 304L Stainless Steel ». Microscopy and Microanalysis 3, S2 (août 1997) : 545–46. http://dx.doi.org/10.1017/s1431927600009612.

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Radiation-induced segregation (RIS) is the spatial redistribution of elements at defect sinks such as grain boundaries and free surfaces during irradiation. This phenomenon has been studied in a wide variety of alloys and has been linked to irradiation-assisted stress corrosion cracking (IASCC) of nuclear reactor core components. Therefore, accurate determination of the grain boundary composition is important in understanding its effects on environmental cracking. Radiation-induced segregation profiles are routinely measured by scanning-transmission electron microscopy using energy-dispersive X-ray spectroscopy (STEM-EDS) and Auger electron spectroscopy (AES). Because of the narrow width of the segregation profile (typically less than 10 nm full width at half-maximum), the accuracy of grain boundary concentration measurements using STEM/EDS depends on the characteristics of the analyzing instrument, specifically, the excited volume in which x-rays are generated. This excited volume is determined by both electron beam diameter and the primary electron beam energy. Increasing the primary beam energy in STEM/EDS produces greater measured grain boundary segregation, as the reduced electron beam broadening a smaller excited volume.

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