Relevant bibliographies by topics / Irradiated stainless steel

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Irradiated stainless steel'

Author: Grafiati
Published: 4 June 2021
Last updated: 1 February 2022

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Journal articles on the topic "Irradiated stainless steel"

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Kanne, W. R., G. T. Chandler, D. Z. Nelson, and E. A. Franco-Ferreira. "Welding irradiated stainless steel." Journal of Nuclear Materials 225 (August 1995): 69–75. http://dx.doi.org/10.1016/0022-3115(94)00439-0.

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Wang, C. A., M. L. Grossbeck, N. B. Potluri, and B. A. Chin. "Welding of irradiated stainless steel." Journal of Nuclear Materials 233-237 (October 1996): 213–17. http://dx.doi.org/10.1016/s0022-3115(96)00203-6.

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de Visser-Týnová, Eva, Stephen W. Swanton, Stephen J. Williams, Marcel P. Stijkel, Alison J. Walker, and Robert L. Otlet. "14C release from irradiated stainless steel." Radiocarbon 60, no. 6 (November 22, 2018): 1671–81. http://dx.doi.org/10.1017/rdc.2018.134.

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ABSTRACTRadiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.
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Mills, W. J. "Fracture Toughness of Irradiated Stainless Steel Alloys." Nuclear Technology 82, no. 3 (September 1988): 290–303. http://dx.doi.org/10.13182/nt88-a34130.

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Brimhall, J. L., J. I. Cole, and S. M. Bruemmer. "Deformation microstructures in ion-irradiated stainless steel." Scripta Metallurgica et Materialia 30, no. 11 (June 1994): 1473–78. http://dx.doi.org/10.1016/0956-716x(94)90248-8.

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Kenik, E. A., J. T. Busby, M. K. Miller, A. M. Thuvander, and G. Was. "Grain Boundary Segregation and Irradiation-Assisted Stress Corrosion Cracking of Stainless Steels." Microscopy and Microanalysis 5, S2 (August 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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Chen, Y., and E. Marquis. "Microstructural Characterization of an Irradiated 304 Stainless Steel." Microscopy and Microanalysis 19, S2 (August 2013): 1746–47. http://dx.doi.org/10.1017/s1431927613010726.

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Furutani, Gen, Nobuo Nakajima, Takao Konishi, and Mitsuhiro Kodama. "Stress corrosion cracking on irradiated 316 stainless steel." Journal of Nuclear Materials 288, no. 2-3 (February 2001): 179–86. http://dx.doi.org/10.1016/s0022-3115(00)00704-2.

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Jin, Hyung-Ha, Eunsol Ko, and Junhyun Kwon. "Microstructural Characterization of Hydrogen Irradiated Austenitic Stainless Steel." Microscopy and Microanalysis 21, S3 (August 2015): 1001–2. http://dx.doi.org/10.1017/s1431927615005802.

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Watanabe, K., S. Jitsukawa, S. Hamada, T. Kodaira, and A. Hishinuma. "Weldability of neutron-irradiated type 316 stainless steel." Fusion Engineering and Design 31, no. 1 (April 1996): 9–15. http://dx.doi.org/10.1016/0920-3796(95)00424-6.

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Dissertations / Theses on the topic "Irradiated stainless steel"

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Gupta, Jyoti. "Intergranular stress corrosion cracking of ion irradiated 304L stainless steel in PWR environment." Thesis, Toulouse, INPT, 2016. http://www.theses.fr/2016INPT0031/document.

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L’IASCC est un mécanisme de fissuration intergranulaire par corrosion sous contrainte (IGCSC) induite par l'irradiation. C’est un phénomène complexe qui peut avoir une influence significative sur le temps et le coût de maintenance des composants internes du coeur des réacteurs à eau pressurisée (REP) et est donc un sujet d'intérêt. Des études récentes ont proposé d'utiliser l'irradiation aux ions (protons) comme une alternative à l'irradiation neutronique afin d’améliorer la compréhension du mécanisme. L'objectif de cette thèse est d’étudier la sensibilité à la fissuration de l’acier austénitique SA 304L irradié aux ions ainsi que les facteurs contribuant à cette fissuration. Deux types d’irradiations aux ions ont été menées (fer et aux protons). Ces deux irradiations ont générées des défauts ponctuels dans la microstructure représentatifs de ceux crées par les neutrons provoquant ainsi le durcissement de l’acier austénitique 304L. Matériel (non irradié et le fer irradié) n'a montré aucune sensibilité à la fissuration intergranulaire sur la soumission à un essai de traction lente SSRT (Slow Strain Rate Test) commencer avec une vitesse de déformation de 5 × 10-8 s-1 jusqu'à 4% de déformation plastique dans un environnement inerte. Il est montré que les deux types d’irradiation aux ions (fer et protons) augmentent la sensibilité à la fissuration intergranulaire du matériau après un essai de SSRT dans un environnement simulé de REP à 340 ° C. La corrélation entre la sensibilité de fissuration et le degré de localisation de la déformation plastique a été étudiée. L’impact de l'irradiation aux ions fer sur l'oxydation du 304L a été aussi étudié grâce à des essais effectués pendant 360 h dans un milieu REP à 340 ° C. Les résultats de cette thèse indiquent que la fissuration intergranulaire de l'acier inoxydable 304L en milieu REP peut être étudiée en utilisant l'irradiation Fe malgré sa faible profondeur de pénétration dans le matériau. Par ailleurs, il est montré que le comportement vis-à-vis de la fissuration est similaire entre une irradiation aux protons et au fer, et ceux malgré une localisation de la déformation moins importante pour ces derniers. Par conséquent, l’irradiation au fer est utilisée pour étudier l'impact de la préparation de surface et des chemins de déformation sur la sensibilité de la fissuration intergranulaire de l’acier 304L
IASCC is irradiation – assisted enhancement of intergranular stress corrosion cracking susceptibility of austenitic stainless steel. It is a complex degrading phenomenon which can have a significant influence on maintenance time and cost of PWRs’ core internals and hence, is an issue of concern. Recent studies have proposed using ion irradiation (to be specific, proton irradiation) as an alternative of neutron irradiation to improve the current understanding of the mechanism. The objective of this study was to investigate the cracking susceptibility of irradiated SA 304L and factors contributing to cracking, using two different ion irradiations; iron and proton irradiations. Both resulted in generation of point defects in the microstructure and thereby causing hardening of the SA 304L. Material (unirradiated and iron irradiated) showed no susceptibility to intergranular cracking on subjection to SSRT with a strain rate of 5 × 10-8 s-1 up to 4 % plastic strain in inert environment. But, irradiation (iron and proton) was found to increase intergranular cracking severity of material on subjection to SSRT in simulated PWR primary water environment at 340 °C. Correlation between the cracking susceptibility and degree of localization was studied. Impact of iron irradiation on bulk oxidation of SA 304L was studied as well by conducting an oxidation test for 360 h in simulated PWR environment at 340 °C. The findings of this study indicate that the intergranular cracking of 304L stainless steel in PWR environment can be studied using Fe irradiation despite its small penetration depth in material. Furthermore, it has been shown that the cracking was similar in both iron and proton irradiated samples despite different degrees of localization. Lastly, on establishing iron irradiation as a successful tool, it was used to study the impact of surface finish and strain paths on intergranular cracking susceptibility of the material
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Patra, Anirban. "Modeling the mechanical behavior and deformed microstructure of irradiated BCC materials using continuum crystal plasticity." Diss., Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/50366.

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The mechanical behavior of structural materials used in nuclear applications is significantly degraded as a result of irradiation, typically characterized by an increase in yield stress, localization of inelastic deformation along narrow dislocation channels, and considerably reduced strains to failure. Further, creep rates are accelerated under irradiation. These changes in mechanical properties can be traced back to the irradiated microstructure which shows the formation of a large number of material defects, e.g., point defect clusters, dislocation loops, and complex dislocation networks. Interaction of dislocations with the irradiation-induced defects governs the mechanical behavior of irradiated metals. However, the mechanical properties are seldom systematically correlated to the underlying irradiated microstructure. Further, the current state of modeling of deformation behavior is mostly phenomenological and typically does not incorporate the effects of microstructure or defect densities. The present research develops a continuum constitutive crystal plasticity framework to model the mechanical behavior and deformed microstructure of bcc ferritic/martensitic steels exposed to irradiation. Physically-based constitutive models for various plasticity-induced dislocation migration processes such as climb and cross-slip are developed. We have also developed models for the interaction of dislocations with the irradiation-induced defects. A rate theory based approach is used to model the evolution of point defects generated due to irradiation, and coupled to the mechanical behavior. A void nucleation and growth based damage framework is also developed to model failure initiation in these irradiated materials. The framework is used to simulate the following major features of inelastic deformation in bcc ferritic/martensitic steels: irradiation hardening, flow localization due to dislocation channel formation, failure initiation at the interfaces of these dislocation channels and grain boundaries, irradiation creep deformation, and temperature-dependent non-Schmid yield behavior. Model results are compared to available experimental data. This framework represents the state-of-the-art in constitutive modeling of the deformation behavior of irradiated materials.
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Duff, Jonathon Andrew. "The influence of grain boundary structure in proton irradiated stainless steel on susceptibility to irradiation assisted stress corrosion cracking." Thesis, University of Manchester, 2008. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.496690.

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Gunter, Cameron Cornelius. "Feasibility of Friction Stir Processing (FSP) as a Method of Healing Cracks in Irradiated 304L Stainless Steels." BYU ScholarsArchive, 2016. https://scholarsarchive.byu.edu/etd/6111.

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The current US fleet of nuclear reactors has been in service for three decades. Over this period, existing welds in stainless steel (SS) shrouds have sustained stress corrosion cracking (SCC) and are in need of repair. Additionally, helium has formed interstitially as a byproduct of proton bombardment. Current repair technology, such as TIG welding, puts extreme amounts of heat into the material and allows for interstitial helium atoms to aggregate and form bubbles/voids at grain boundaries. This significantly weakens the material, proving to be a very counterproductive and ineffective repair technique. Much study has been done on friction stir processing (FSP), but none has explored it as an enabling repair technology for use in nuclear applications. Because of its relatively low energy input as a solid state joining technology, it is proposed that FSP could effectively heal SCCs in these existing welds without the negative side effect of helium bubble formation. A spread of speeds and feeds were initially tested using a PCBN-W-Re tool on 304L SS. Six of these parameter sets were selected as representations of high, medium, and low temperature-per-power outputs for this research: 2 IPM-80 RPM, 2 IPM-150 RPM, 4 IPM-150 RPM, 4 IPM-250 RPM, 6 IPM-125 RPM, and 6 IPM-175 RPM. These varied parameter sets were tested for their tensile, micro-hardness, and corrosion resistant properties. In general, the lower IPM and RPM values resulted in higher ultimate tensile strengths (UTS). Higher IPM and RPM values resulted in tunnel, pin hole, and surface void defects. These defects caused premature failure in tensile tests and could often be identified through microscopy. Micro-hardness testing demonstrated a strong correlation per the Hall-Petch relationship – finer grain sizes resulted in higher yield strength (hardness values) of the material. The tool temperature during FSP was a good indicator of the expected hardness – lower temperatures resulted in higher hardness values. Corrosion testing was performed with a 1000-hour alternate immersion test in a room temperature 3.5% NaCl solution. With these testing parameters, the results demonstrated that FSP had no effect on the corrosion resistance of 304L SS under these conditions.
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Fuller, Robert William. "Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels." Thesis, Mississippi State University, 2018. http://pqdtopen.proquest.com/#viewpdf?dispub=10747959.

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One of prominent issues related to failures in nuclear power components is attributed to material degradation due the aggressive environment conditions, and mechanical stresses. For instance, reactor core support components, such as fuel claddings, are under prolonged exposure to an intense neutron field from the fission of fuel and operate at elevated temperature under fatigue loadings caused by start up, shut down, and unscheduled emergency shut down. Additionally, exposure to high-fluence neutron radiation can lead to microscopic defects that result in material hardening and embrittlement, which significantly affects the physical and mechanical properties of the materials, resulting in further reduction in fatigue life of reactor structural components. The effects of fatigue damage on material deterioration can be further exacerbated by the presence of thermal loading, hold-time, and high-temperature water coolant environments.

In this study, uniaxial fatigue models were used to predict fatigue behavior based only on simple monotonic properties including ultimate tensile strength and Brinell hardness. Two existing models, the Bäumel Seeger uniform material law and the Roessle Fatemi hardness method, were employed and extended to include the effects of test temperature, neutron irradiation fluence, irradiation induced helium and irradiation induced swellings on fatigue life of austenitic stainless steels. Furthermore, a methodology to estimate fatigue crack length using a strip-yield based model is presented. This methodology is also extended to address the effect of creep deformation in a presence of hold- times, and expanded to include the effects of irradiation and water environment. Reasonable fatigue life predictions and crack growth estimations are obtained for irradiated austenitic stainless steels types 304, 304L, and 316, when compared to the experimental data available in the literature. Lastly, a failure analysis methodology of a mixer unit shaft made of AISI 304 stainless steel is also presented using a conventional 14-step failure analysis approach. The primary mode of failure is identified to be intergranular stress cracking at the heat affected zones. A means of circumventing this type of failure in the future is presented.

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Scherer, Jean-Michel. "Strain localization and ductile fracture in single crystals : application to irradiated austenitic stainless steels." Thesis, Université Paris sciences et lettres, 2020. http://www.theses.fr/2020UPSLM026.

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Pour leurs excellentes propriétés mécaniques et d’oxydation, les aciers austénitiques inoxydables sont largement utilisés dans l’industrie nucléaire, en particulier pour les structures internes de coeur des réacteurs. Toutefois, les niveaux d’irradiation neutronique importants auxquels ces matériaux sont exposés peuvent nuire à leurs propriétés mécaniques. Une forte baisse de la ténacité est en effet observée à mesure que la dose d’irradiation augmente. Selon les conditions d’irradiation (température, dose), on distingue principalement deux types de défauts induits par l’irradiation pouvant être responsables de ce comportement : des boucles de dislocations de Frank à basse température d’irradiation (∼300 ◦C) et des nano-cavités à haute température (∼600 ◦C). Comme ces défauts existent et agissent à des échelles inférieures à la taille de grain, leurs effets peuvent être étudiés à l’échelle du monocristal. Tout d’abord, ce travail vise à obtenir des données expérimentales sur le comportement mécanique des monocristaux d’acier inoxydable austénitique. Ensuite, la modélisation de la localisation de la déformation plastique induite par l’adoucissement survenant dans les aciers irradiés est étudiée. Les limites d’un modèle de plasticité cristalline à gradient sont exposées sur la base de prédiction analytiques de l’apparition de bandes de localisation. Une théorie étendue tenant compte de l’évolution de la longueur interne est proposée. Une attention particulière est alors accordée à l’efficacité numérique de la mise en oeuvre par éléments finis du modèle de plasticité à gradient susmentionné. Des formulations basées sur l’approche micromorphe ou sur une approche à multiplicateur de Lagrange sont décrites et comparées à l’aide de simulations par éléments finis. Enfin, un modèle de rupture ductile de monocristaux poreux est proposé – incluant à la fois la croissance et la coalescence des cavités – afin d’étudier l’impact des nano-cavités induites par irradiation sur le comportement mécanique des aciers austénitiques inoxydables. Le modèle est mis en place dans un formalisme à gradient afin de régulariser la rupture ductile
For their excellent mechanical and oxidation properties, austenitic stainless steels are widely used in the nuclear industry, in particular for structural applications inside the core of reactors. However the substantial neutron irradiation levels these materials can be exposed to can detrimentally affect their mechanical properties. A sharp drop of toughness is indeed observed as the irradiation dose increases. Depending on the irradiation conditions (temperature, dose), mainly two kinds of radiation-induced defects can be responsible for this behaviour: dislocation Frank loops at low irradiation temperature (∼300 ◦C) and nano-voids at higher temperature (∼600 ◦C). Since these defects exist and act at the subgrain level, it motivates to study their effects at the single crystal scale. First of all, this work aims at obtaining experimental data on the mechanical behaviour of austenitic stainless steel single crystals. Then, modeling of softening induced strain localization phenomena, as those taking place in irradiated materials, is investigated. The limitations of a reduced strain gradient crystal plasticity model regarding shear bands predictions are exposed on the grounds of analytical solutions and an enhanced theory accounting for internal length evolution is proposed. Thereupon attention is given to the numerical efficiency of the finite element implementation of the aforementionned strain gradient plasticity model. Micromorphic and Lagrange multiplier based formulations of the original theory are described and compared upon finite element simulations. Eventually, one of a kind ductile fracture model of porous single crystals is proposed – including both void growth and void coalescence – in order to investigate impact of radiation-induced nano-voids on the mechanical behavior of irradiated austenitic stainless steels. The model is set up in a strain gradient framework in order to regularize ductile fracture
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Seitzman, Larry Edward. "A study of the effects of oxygen on void stability in an ion-irradiated austenitic stainless steel." 1988. http://catalog.hathitrust.org/api/volumes/oclc/18854667.html.

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Books on the topic "Irradiated stainless steel"

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Chung, H. M. Irradiation-assisted stress corrosion cracking of model austenitic stainless steels irradiated in the Halden reactor. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.

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Chung, H. M. Irradiation-assisted stress corrosion cracking of model austenitic stainless steels irradiated in the Halden reactor. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.

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Chung, H. M. Irradiation-assisted stress corrosion cracking of model austenitic stainless steels irradiated in the Halden reactor. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1999.

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Fracture toughness and crack growth rates of irradiated austenitic stainless steels. Washington, D.C: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2003.

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Book chapters on the topic "Irradiated stainless steel"

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Gerke, R. David, and William A. Jesser. "Fracture in Helium-Irradiated Type 316 Stainless Steel Microtensile Specimens." In Effects of Radiation on Materials: 12th International Symposium Volume II, 605–18. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1985. http://dx.doi.org/10.1520/stp87019850005.

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Bruemmer, S. M., D. J. Edwards, B. W. Arey, and L. A. Chariot. "Microstructural, Microchemical and Hardening Evolution in LWR-Irradiated Austenitic Stainless Steel." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1078–87. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch113.

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Chen, Yimeng, Yan Dong, Emmanuelle Marquis, Zhijie Jiao, Justin Hesterberg, Gary Was, and Peter Chou. "Solute Clustering in As-irradiated and Post-irradiation-Annealed 304 Stainless Steel." In The Minerals, Metals & Materials Series, 973–91. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-68454-3_71.

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Chen, Yimeng, Yan Dong, Emmanuelle Marquis, Zhijie Jiao, Justin Hesterberg, Gary Was, and Peter Chou. "Solute Clustering in As-irradiated and Post-irradiation-Annealed 304 Stainless Steel." In The Minerals, Metals & Materials Series, 2189–207. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_147.

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Jiao, Zhijie, and Gary Was. "Oxidation of a Proton-Irradiated 316 Stainless Steel in Simulated BWR NWC Environment." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 1329–38. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_81.

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Jiao, Zhijie, and Gary Was. "Oxidation of a Proton-Irradiated 316 Stainless Steel in Simulated BWR NWC Environment." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1329–38. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch138.

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Freyer, Paula D., William T. Cleary, Elaine M. Ruminski, C. Joseph Long, and Peng Xu. "Hot Cell Tensile Testing of Neutron Irradiated Additively Manufactured Type 316L Stainless Steel." In The Minerals, Metals & Materials Series, 1021–38. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_64.

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Freyer, Paula D., William T. Cleary, Elaine M. Ruminski, C. Joseph Long, and Peng Xu. "Hot Cell Tensile Testing of Neutron Irradiated Additively Manufactured Type 316L Stainless Steel." In The Minerals, Metals & Materials Series, 1021–38. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-67244-1_64.

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Little, Edward A. "DynamicJ-lntegral Toughness and Fractographic Studies of Fast Reactor Irradiated Type 321 Stainless Steel." In Effects of Radiation on Materials: 12th International Symposium Volume II, 563–79. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1985. http://dx.doi.org/10.1520/stp87019850003.

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Boisson, M., L. Legras, F. Carrette, O. Wendling, T. Sauvage, A. Bellamy, P. Desgardin, L. Laffont, and E. Andrieu. "Comparative Study on Short Time Oxidation of Un-Irradiated and Protons Pre-Irradiated 316L Stainless Steel in Simulated PWR Water." In The Minerals, Metals & Materials Series, 899–918. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-68454-3_66.

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Conference papers on the topic "Irradiated stainless steel"

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Khalil, Sarah, and Tarek M. Hatem. "Hydrogen Embrittlement Characteristics in Irradiated Stainless Steel." In ASME 2020 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/imece2020-24081.

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Abstract Hydrogen is produced in nuclear reactors as a by-product of the corrosion reaction between the pressure vessel and the cooling water, where hydrogen produced may enter the metal in atomic form. During operation a reactor vessel is exposed to avalanche of neutron irradiation fluxes, in addition to corrosion from cooling water. A novel cluster dynamics model that accounts for off-stoichiometry of clusters and matrix was developed and applied to investigate the clustering behavior of Hydrogen-vacancy and Hydrogen-interstitial clusters in proton irradiated stainless steel has been developed. The differences in point defect migration energies and binding energy of H to lattice defects, makes it possible to have vacancy and interstitial clusters having compositions different from those of pure iron. The model predicts populations of Defect-Hydrogen complexes in iron. The model is applied to the early stage formation of voids and dislocation loops in stainless steel in the presence of atomic hydrogen. This study investigates the effect of irradiation dose and temperature on the concentration of vacancy-Hydrogen (VmHn) and Intersitial Fe-H (FemHn) complexes on bulk α-Iron. The re
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Tang, Wei, Maxim Gussev, Zhili Feng, Brian Gibson, Roger Miller, Jian Chen, Scarlett Clark, et al. "Friction Stir Welding and Preliminary Characterization of Irradiated 304 Stainless Steel." In ASME 2019 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/pvp2019-93899.

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Abstract The mitigation of helium induced cracking in the heat affected zone (HAZ), a transition metallurgical zone between the weld zone and base metal, during repair welding is a great challenge in nuclear industry. Successful traditional fusion welding repairs are limited to metals with a maximum of a couple of atomic parts per million (appm) helium, and structural materials helium levels in operating nuclear power plants are generally exceed a couple of appm after years of operations. Therefore, fusion welding is very limited in nuclear power plants structural materials repairing. Friction stir welding (FSW) is a solid-state joining technology that reduces the drivers (temperature and tensile residual stress) for helium-induced cracking. This paper will detail initial procedural development of FSW weld trials on irradiated 304L stainless steel (304L SS) coupons utilizing a unique welding facility located at one of Oak Ridge National Laboratory’s hot cell facilities. The successful early results of FSW of an irradiated 304L SS coupon containing high helium are discussed. Helium induced cracking was not observed by scanning electron microscopy in the friction stir weld zone and the metallurgical zones between the weld zone and base metal, i.e. thermal mechanical affected zone (TMAZ) and HAZ. Characterization of the weld, TMAZ and HAZ regions are detailed in this paper.
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Sato, Masatoshi, Masanori Kanno, Kiyotomo Nakata, Hidenori Takahashi, and Hiroshi Sakamoto. "The Study on the Applicability of Laser Surface Modification Technology to Irradiated Stainless Steel." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48312.

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Recently, occurrence of stress corrosion cracking has been reported at core shrouds in Boiling Water Reactor (BWR) nuclear power plants. Yttrium aluminum garnet (YAG) laser surface modification technologies (i.e. Laser Surface Melting Technology (LSM), Laser Cladding Technology (LC)) have been developed as promising preventive maintenance technologies to stress corrosion cracking (SCC) of austenitic stainless steel structures and components. On the other hand, it has been also well-known that the helium transmuted from nickel and boron is accumulated to neutron irradiated stainless steel, and that helium related cracks may occur at weld heat affected zone which were attributed to nucleation along grain boundaries, coalescence and growth of helium bubbles due to thermal cycle and thermal stress during welding. Then, the laser surface modification technologies to the irradiated stainless steels was developed and the applicability of these technologies was evaluated based on the results of various tests (e.g. dye-penetrant test, micro structure observation and bending test) to the laser surface modified Type 304 and Type 316L specimens containing up to about 10 appm helium. The laser surface modification applicability diagram was developed as a function of weld heat input and helium concentration, which was supported by numerical simulation on helium bubble formation and growth during welding for irradiated stainless steels.
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Chen, Y., B. Alexandreanu, W. J. Shack, K. Natesan, and A. S. Rao. "Cyclic Crack Growth Rate of Irradiated Austenitic Stainless Steel Welds in Simulated BWR Environment." In ASME 2011 Pressure Vessels and Piping Conference. ASMEDC, 2011. http://dx.doi.org/10.1115/pvp2011-57728.

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Reactor core internal components in light water reactors are subjected to neutron irradiation. It has been shown that the austenitic stainless steels used in reactor core internals are susceptible to stress corrosion cracking after extended neutron exposure. This form of material degradation is a complex phenomenon that involves concomitant conditions of irradiation, stress, and corrosion. Interacting with fatigue damage, irradiation-enhanced environmental effects could also contribute to cyclic crack growth. In this paper, the effects of neutron irradiation on cyclic cracking behavior were investigated for austenitic stainless steel welds. Post-irradiation cracking growth tests were performed on weld heat-affected zone specimens in a simulated boiling water reactor environment, and cyclic crack growth rates were obtained at two doses. Environmentally enhanced cracking was readily established in irradiated specimens. Crack growth rates of irradiated specimens were significantly higher than those of nonirradiated specimens. The impact of neutron irradiation on environmentally enhanced cyclic cracking behavior is discussed for different load ratios.
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Allen, Todd R., Hanchung Tsai, James I. Cole, Joji Ohta, Kenji Dohi, and Hideo Kusanagi. "Mechanical Properties of 20% Cold-Worked 316 Stainless Steel Irradiated at Low Dose Rate." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22189.

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To assess the effects of long-term, low-dose-rate neutron exposure on mechanical strength and ductility, tensile properties were measured on irradiated 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1–47 dpa at temperatures from 371–385°C and dose rates from 0.8–2.8 × 10−7 dpa/s. These dose rates are about one order of magnitude lower than those of typical EBR-II in-core experiments. Irradiation cuased hardening, with the yield strength (YS) following approximately the same trend as the ultimate tensile strength (UTS). At higher dose, the difference between the UTS and YS decreases, suggesting the work-hardening capability of the material is decreasing with increasing dose. Both the uniform elongation and total elongation decrease up to the largest dose. Unlike the strength data, the ductility reduction showed no signs of saturating at 20 dpa. While the material retained respectable ductility at 20 dpa, the uniform and total elongation decreased to <1 and <3%, respectively, at 47 dpa. Fracture in the 30 dpa specimen is mainly ductile but with local regions of mixed-mode failure consisting of dimples and microvoids. The fracture surface of the higher-exposure 47 dpa specimen displays significantly more brittle features. The fracture consists of maily small facets and slip bands that suggest channel fracture. The hardening in these low-dose-rate components differs from that measured in test samples irradiated in EBR-II at higher-dose-rate. The material irradiated at higher dose rate loses work hardening capactiy faster than the lower dose rate material, although this effect could be due to compositional differences.
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6

Sim, Jae Min, Yoon-Suk Chang, Byeong Seo Kong, and Changheui Jang. "Mechanical Properties of Ion-Irradiated Stainless Steel Determined by Nanoindentation Tests and Finite Element Analyses." In ASME 2020 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/pvp2020-21507.

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Abstract While austenitic stainless steels (ASSs) have been widely adopted for reactor vessel internals because of their excellent material properties, diverse ageing-related degradation may occur due to high temperature, corrosive and neutron radiation environments during operation. In particular, since the change of mechanical properties is a major concern in long-term operation but it is difficult to prepare and handle standard specimens influenced by neutrons, most of experimental researches for enhanced understanding of the radiation effects have been focused on high-energetic ion-irradiation and tests of small specimens. In this study, systematic finite element analyses were carried out to quantify changing mechanical properties based on both virgin and ion-irradiated nanoindentation test data of typical ASS material. First of all, numerical analysis was carried out to obtain unirradiated material constitutive parameters by using trial set along the miniature specimen and comparing test data, and then indentation stress-strain (ISS) curve was derived. Subsequently, ISS was converted into uniaxial stress-strain response taking into account simple correlation. Finally, with regard to the irradiated material, similar analytical procedures were established. 304 SS was irradiated with 2 MeV proton and radioactivity is being measured. Comparison between analysis result and experimental one will be carried out, of which details and key findings will be discussed.
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Kaji, Yoshiyuki, Hirokazu Ugachi, Takashi Tsukada, Yoshinori Matsui, Masao Ohmi, Nobuaki Nagata, Koji Dozaki, and Hideki Takiguchi. "In-Pile SCC Growth Behavior of Type 304 Stainless Steel in High Temperature Water at JMTR." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89338.

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Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs). It is, however, considered that the reproduced IASCC by PIEs must be carefully compared with the actual IASCC in nuclear power plants, because the actual IASCC occurs in the core under simultaneous effects of radiation, stress and high temperature water environment. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. A high temperature water loop facility was installed at the Japan Materials Testing Reactor (JMTR) to carry out the in-pile IASCC testing under a framework of cooperative research program between JAERI and the JAPC. In-pile IASCC growth tests have been successfully carried out using the compact tension (CT) type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1×1025n/m2 before the in-pile testing since 2004. The tests were carried out in pure water simulated boiling water reactor (BWR) coolant condition. In the paper, results of the in-pile SCC growth tests will be discussed comparing with the result obtained by PIEs from a viewpoint of the synergistic effects on IASCC.
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Andresen, Peter L. "Factors Influencing SCC and IASCC of Stainless Steels in High Temperature Water." In ASME/JSME 2004 Pressure Vessels and Piping Conference. ASMEDC, 2004. http://dx.doi.org/10.1115/pvp2004-2663.

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SCC growth studies were performed in high temperature, high purity water on various grades and various conditions of stainless steel. The synergistic effects of corrosion potential, sensitization, cold work (yield strength), temperature and irradiation were evaluated, and their implications to interpreting and modeling SCC in unirradiated and irradiated structures are discussed.
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Freyer, Paula D., Jonathan K. Tatman, Frank A. Garner, Greg J. Frederick, and Benjamin J. Sutton. "Hot Cell Pulsed Laser Welding of Neutron Irradiated Type 304 Stainless Steel With a Maximum Damage Dose of 28 DPA." In ASME 2019 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/pvp2019-93316.

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Abstract Radiation-induced degradation of reactor pressure vessels and internals is a concern to the aging nuclear fleet and welding solutions will be required if repair of these irradiated components is deemed necessary. However, the weldability of highly irradiated austenitic materials is significantly diminished due to the presence of irradiation induced helium in the material matrix. Helium-induced weld cracking is a complex phenomenon that is related to the concentration of helium, the heat input from the welding process, and stresses generated during cooling of the weld. During conventional high heat input welding methods such as gas tungsten arc welding, helium bubbles can coalesce and grow on base metal grain boundaries within the heat-affected zone which subsequently causes intergranular cracking. The objective of this work was to obtain weldability data by characterizing welds made on highly activated, neutron irradiated Type 304 stainless steel containing both radiation-induced helium and microstructural damage such as void swelling. All irradiated materials welding was performed inside a Westinghouse hot cell utilizing a pulsed Nd:YAG laser with welds made on three rectangular samples of highly activated Type 304 stainless steel. The rectangular samples were cut and milled in-cell from sections previously obtained from two neutron reflector hex blocks. The hex blocks are U.S. Department of Energy owned material and were irradiated for approximately 13 years in the EBR-II sodium cooled fast reactor from 1982 until 1995. The three samples selected for welding have nominal damage doses of approximately 0.4, 11, and 28 dpa with corresponding estimated helium contents of 0.2, 3 and 8 appm helium, respectively. A number of different weld parameter sets were utilized and included variations of travel speed, wire feed speed and lens-to-work distances. The parameter sets allowed for a range of effective weld heat input levels to be compared. Single pass and multiple pass as well as wire fed and autogenous welds were made. This paper presents the results from post-weld evaluations performed on the three welded irradiated samples, focusing on the reduced tendency for cracks to form adjacent to the weld as a function of weld parameters, lens-to-work distance and helium content.
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Tanguy, Benoit, Ce´dric Pokor, Anthony Stern, and Philippe Bossis. "Initiation Stress Threshold Irradiation Assisted Stress Corrosion Cracking Criterion Assessment for Core Internals in PWR Environment." In ASME 2011 Pressure Vessels and Piping Conference. ASMEDC, 2011. http://dx.doi.org/10.1115/pvp2011-58051.

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Irradiation assisted stress corrosion cracking (IASCC) is a problem of growing importance in pressurized water reactors (PWR). An understanding of the mechanism(s) of IASCC is required in order to provide guidance for the development of mitigation strategies. One of the principal reasons why the IASCC mechanism(s) has been so difficult to understand is the inseparability of the different IASCC potential contributors evolutions due to neutron irradiation. The potential contributors to IASCC in PWR primary water are: (i) radiation induced segregation (RIS) at grain boundaries, (ii) radiation induced microstructure (formation and growth of dislocations loops, voids, bubbles, phases), (iii) localized deformation under loading, (iv) irradiation creep and transmutations. While the development of some of the contributors (RIS, microstructure) with increasing doses are at least qualitatively well understood, the role of these changes on IASCC remains unclear. Parallel to fundamental understanding developments relative to IASCC, well controlled laboratory tests on neutron irradiated stainless steels are needed to assess the main mechanisms and also to establish an engineering criterion relative to the initiation of fracture due to IASCC. First part of this study describes the methodology carried out at CEA in order to provide more experimental data from constant load tests dedicated to the study of initiation of SCC on neutron irradiated stainless steel. A description of the autoclave recirculation loop dedicated to SCC tests on neutron irradiated materials is then given. This autoclave recirculation loop has been started on July 2010 with the first SCC test on an irradiated stainless steel (grade 316) performed at CEA. The main steps of the interrupted SCC tests are then described. Second part of this paper reports the partial results of the first test performed on a highly neutron irradiated material.
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Reports on the topic "Irradiated stainless steel"

1

Klueh, R. L. Tensile behavior of irradiated manganese-stabilized stainless steel. Office of Scientific and Technical Information (OSTI), October 1996. http://dx.doi.org/10.2172/414884.

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2

Hashimoto, N., J. P. Robertson, M. L. Grossbeck, A. F. Rowcliffe, and E. Wakai. Damage structure of austenitic stainless steel 316LN irradiated at low temperature in HFIR. Office of Scientific and Technical Information (OSTI), March 1998. http://dx.doi.org/10.2172/335399.

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3

Porollo, S. I., A. N. Vorobjev, Yu V. Konobeev, and F. A. Garner. Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C. Office of Scientific and Technical Information (OSTI), March 1998. http://dx.doi.org/10.2172/335397.

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4

Pawel, J. E., M. L. Grossbeck, and A. F. Rowcliffe. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment. Office of Scientific and Technical Information (OSTI), April 1995. http://dx.doi.org/10.2172/114931.

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5

Sindelar, R. L., and G. R. Jr Caskey. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components. Office of Scientific and Technical Information (OSTI), December 1991. http://dx.doi.org/10.2172/5084386.

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Chopra, O. K., B. Alexandreanu, E. E. Gruber, R. S. Daum, and W. J. Shack. Crack growth rates of irradiated austenitic stainless steel weld heat affected zone in BWR environments. Office of Scientific and Technical Information (OSTI), January 2006. http://dx.doi.org/10.2172/925223.

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7

Sindelar, R. L., and G. R. Jr Caskey. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components. Office of Scientific and Technical Information (OSTI), December 1991. http://dx.doi.org/10.2172/10164219.

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8

Baumann, E. W., and G. R. Jr Caskey. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens. Office of Scientific and Technical Information (OSTI), July 1993. http://dx.doi.org/10.2172/10185742.

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9

Gussev, Maxim, Gabriel de Bellefon, and T. M. Rosseel. Analysis of Localized Deformation Processes in Highly Irradiated Austenitic Stainless Steel through In Situ Techniques. Office of Scientific and Technical Information (OSTI), September 2019. http://dx.doi.org/10.2172/1661257.

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10

Porollo, S. I., A. N. Vorobjev, and Yu V. Konobeev. Extreme embrittlement of austenitic stainless steel irradiated to 75-81 dpa at 335-360{degrees}C. Office of Scientific and Technical Information (OSTI), April 1997. http://dx.doi.org/10.2172/543295.

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