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Kenik, E. A., R. H. Jones, and G. E. C. Bell. "Irradiation-assisted stress corrosion cracking." Journal of Nuclear Materials 212-215 (September 1994): 52–59. http://dx.doi.org/10.1016/0022-3115(94)90033-7.
Kain, V., S. B. Chafle, D. Feron, B. Tanguy, C. Colin, and 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.
Cui, Bai, Michael D. McMurtrey, Gary S. Was, and Ian M. Robertson. "Micromechanistic origin of irradiation-assisted stress corrosion cracking." Philosophical Magazine 94, no. 36 (November 21, 2014): 4197–218. http://dx.doi.org/10.1080/14786435.2014.982744.
Was, Gary S., and Peter L. Andresen. "Irradiation-assisted stress-corrosion cracking in austenitic alloys." JOM 44, no. 4 (April 1992): 8–13. http://dx.doi.org/10.1007/bf03222812.
Rossi, F., F. Fumagalli, A. Ruiz-Moreno, P. Moilanen, and 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 (September 2020): 80–92. http://dx.doi.org/10.1016/j.nimb.2020.06.012.
Smith, Stuart A., Brock Gause, David Plumley, and Masao J. Drexel. "Irradiation-Assisted Stress-Corrosion Cracking of Nitinol During eBeam Sterilization." Journal of Materials Engineering and Performance 21, no. 12 (October 17, 2012): 2638–42. http://dx.doi.org/10.1007/s11665-012-0396-8.
Дисертації з теми "Irradiation assisted stress corrosion cracking":
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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.
Hilton, Bruce A. "Irradiation assisted stress corrosion cracking susceptibility of low fluence stainless steels evaluated by in-flux slow strain rate tests." Thesis, Massachusetts Institute of Technology, 1996. http://hdl.handle.net/1721.1/42810.
Le, Millier Morgane. "Fragilisation des aciers inoxydables austénitiques sous irradiation : évolution de la microstructure et amorçage de la corrosion sous contrainte assistée par l'irradiation en milieu REP." Thesis, Paris, ENMP, 2014. http://www.theses.fr/2014ENMP0047/document.
Cette étude porte sur l'évolution de la microstructure des aciers inoxydables austénitiques sous irradiation et les conséquences de cette évolution sur leur comportement en milieu REP. Un acier 304L a été irradié aux protons à 360°C à 5 et 10 dpa. Suite à ces irradiations, la sensibilité du matériau à l'IASCC a été étudiée en milieu primaire simulé à 350°C, avec suivi par microextensométrie des champs locaux de déformation. Parallèlement à ce travail, des lames minces ont été irradiées in situ aux ions Ni++ à 500°C jusqu'à 2 dpa avec implantation simultanée d'hélium. Ces expérimentations nous ont permis (i) grâce au couplage microstructure /champs mécaniques /fissuration de mieux comprendre les paramètres responsables de l'amorçage de l'IASCC en milieu réducteur (ii) de définir le rôle joué par l'hélium sur l'évolution des défauts d'irradiation. Il s'avère que, dans les conditions d'étude, l'implantation d'hélium n'a qu'un effet limité sur les populations de boucles de dislocation et de cavités pour des rapports inférieurs à 800 appm He/dpa. Des cavités ont été observées avec et sans implantation d'hélium, y compris dans les joints de grains ce qui pourrait être un facteur de fragilisation. L'ensemble des essais de corrosion sous contrainte ont validé que la densité de fissures augmente avec l'augmentation du taux de déformation et qu'un chargement séquentiel conduit à une plus grande ouverture et propagation en surface des fissures. Ces fissures se propagent en profondeur dans la couche irradiée notamment du fait de la surcontrainte générée par le fort gradient de propriétés entre la zone irradiée et non irradiée du matériau. Les mécanismes de déformation activés sont complexes et du maclage a été observé après 2 et 10% de déformation macroscopique. La déformation après irradiation est fortement localisée sous forme de bandes intragranulaires et autour de certains joints de grains, mais la déformation de ces joints ne semble pas constituer un critère d'amorçage. L'absence de transmission de la déformation de part et d'autre des joints fissurés est par contre systématiquement observée et la connaissance de l'état de contrainte local s'avère indispensable pour décrire l'amorçage de l'IASCC en milieu réducteur. Une méthodologie basée sur l'exploitation des résultats expérimentaux (champs d'orientation cristallographique, champs cinématique) appliquée à une simulation aux éléments finis permet d'estimer l'état local de contrainte, seul à même de discriminer un critère d'ouverture de fissure This work deals with the microstructure evolution of austenitic stainless steels under irradiation and the consequences of this evolution on their behavior in PWR environment. 304L steel was proton-irradiated at 360°C to 5 and 10 dpa. Following these irradiations, IASCC was studied in a 350°C simulated primary water, with strain fields measurements using digital image correlation. In parallel, thin foils were irradiated in situ with Ni++ ions at 500°C up to 2 dpa with simultaneous helium implantation. These experiments allowed us (i) to have a better understanding of the key parameters responsible of the IASCC initiation in reducing environment thanks to the coupling between microstructure, mechanical fields and cracking (ii) to define the role of helium on the nucleation and evolution of radiation defects. It turns out that, in the studied conditions, the implantation of helium has only a limited effect on the dislocation loop and cavity populations for ratios lower than 800 appm He/dpa. Cavities were observed with and without helium, including in the grain boundaries which could be a factor of embrittlement. The stress corrosion cracking tests resulted in an increase of the crack density with the increase of the macroscopic deformation and in a bigger opening and on-surface propagation of cracks after a sequential loading. These cracks propagate deeply in the irradiated layer partly because of the overstress generated by the strong gradient of mechanical properties between the irradiated and non-irradiated zones of the material. The activated deformation mechanisms are complex and twinning was observed after 2 and 10% of macroscopic deformation. The deformation after irradiation is strongly localized in transgranular bands and around some grain boundaries, but it appears that the strong deformation around boundaries is not an initiation criterion. Deformation discontinuity on both sides of cracked boundaries is systematically observed and evaluation of the local stress state appears to be essential to describe IASCC initiation. This local stress state could be calculated by finite elements, taking into account the experimental results in terms of crystallographic orientation fields or Kinematics fields strong heterogeneity of local deformation quantified in this work
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Riad, Soukaina. "Vers une modélisation de la corrosion sous contrainte assistée par l'irradiation du superalliage 718." Electronic Thesis or Diss., Ecole centrale de Nantes, 2022. http://www.theses.fr/2022ECDN0039.
Le superalliage base nickel 718 est réputé pour présenter une excellente tenue à la corrosion, une très forte résistancemécanique et une bonne tenue sous irradiation. De ce fait, il s’agit d’un matériau de choix au sein d’un réacteur électronucléaire pour les pièces soumises à des sollicitations extrêmes (ressorts, systèmes de maintien. . . ).Pourtant des ruptures en service ont été observées de ce matériau sous le phénomène de corrosion sous contraintes assistée par l’irradiation. La présente thèse vise à apporter de nouveaux éléments de compréhension de ce phénomène complexe sous l’angle de la modélisation numérique. Le processus de fissuration par corrosion sous contrainte est modélisé par la méthode des champs de phase. Une implémentation unifiée, apte à traiter lesfissurations intra et intergranulaires, est proposée et permet de coupler efficacement différentes échelles de travail (du milieu continu au polycristal) et différents physiques (mécanique des milieux continus et généralisés et oxydation interne). Cette modélisation permet de proposer des simulations des étapes complexes de la corrosion sous contrainte, à savoirl’amorçage, la coalescence et la propagation Inconel 718 alloy is renowned for having excellent corrosion resistance, very high mechanical strength and good resistance to irradiation. Thus, it is a material of choice within a nuclear power reactor for parts subjected to extreme stresses (springs, retaining systems,...). However, failures in service have been observed in this material under irradiationassisted stress corrosion cracking phenomenon. This thesis aims to bring new elements of understanding of this complex phenomenon from the point of view of numerical modeling. The stress corrosion cracking process is modeled by the phase field fracture method. A unified implementation, able to deal with inter and intergranular fracture, is proposedand allows to couple efficiently different scales of work (from continuous medium to polycrystal) and different physics (mechanics of continuous and generalized media and internal oxidation). This modeling allows to propose simulations of the complex stages of stress corrosion cracking, namely initiation, coalescence and propagation
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Ghasemi, Rohollah. "Hydrogen-assisted stress corrosion cracking of high strength steel." Thesis, KTH, Skolan för kemivetenskap (CHE), 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-50416.
In this work, Slow Strain Rate Test (SSRT) testing, Light Optical Microscopy (LOM) and Scanning Electron Microscopy (SEM) were used to study the effect of micro-structure, corrosive environments and cathodic polarisation on stress corrosion cracking (SCC) of two grades of high strength steels, Type A and Type B. Type A is manufactured by quench and tempered (Q&T) method. Type B, a normalize steel was used as reference. This study also supports electrochemical polarisation resistance method as an effective testing technique for measuring the uniform corrosion rate. SSRT samples were chosen from base metal, weld metal and Heat Affected Zone (HAZ). SSRT tests were performed at room temperature under free corrosion potential and cathodic polarisation using 4 mA/cm2 in 1 wt% and 3.5 wt% NaCl solutions. From the obtained corrosion rate measurements performed in 1 wt% and 3.5 wt% NaCl solutions it was observed that increased chloride concentration and dissolved oxygen content enhanced the uniform corrosion for all tested materials. Moreover, the obtained results from SSRT tests demonstrate that both Q&T and normalized steels were not susceptible to SCC in certain strain rate(1×10-6s-1) in 1 wt% and 3.5 wt% NaCl solutions under free corrosion potential. It was con-firmed by a ductile fracture mode and high reduction in area. The weld metal of Type A with acicular ferrite (AF), pro-eutectoid (PF) and bainite microstructure showed higher susceptibility to hydrogen assisted stress corrosion cracking compared to base metal and HAZ. In addition, typical brittle intergranular cracking with small reduction in area was observed on the fracture surface of the Type A due to hydrogen charging.
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Fegan, J. J. H. "Environment assisted cracking of deaerator steels in high temperature water." Thesis, University of Newcastle Upon Tyne, 1995. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.260856.
Thesis (Ph.D.)--Mechanical Engineering, Georgia Institute of Technology, 2008. Committee Co-Chair: Preet M. Singh; Committee Co-Chair: Richard W. Neu; Committee Member: Hamid Garmestani; Committee Member: Timothy Patterson; Committee Member: W. Steven Johnson.
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Cano-Castillo, U. "Environment-assisted cracking of spray-formed Al-alloy and Al-alloy-based composite." Thesis, University of Oxford, 1995. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.260730.
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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Hejman, Ulf. "On initiation of chemically assisted crack growth and crack propagation paths of branching cracks in polycarbonate." Licentiate thesis, Malmö högskola, Teknik och samhälle, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:mau:diva-7790.
Stress corrosion, SC, in some cases gives rise to stress corrosion cracking, SCC, which differs from purely stress intensity driven cracks in many aspects. They initiate and grow under the influence of an aggressive environment in a stressed substrate. They grow at low load and may branch. The phenomenon of SCC is very complex, both the initiation phase and crack extension itself of SCC is seemingly associated with arbitrariness due to the many unknown factors controlling the process. Such factors could be concentration of species in the environment, stress, stress concentration, electrical conditions, mass transport, and so on.In the present thesis, chemically assisted crack initiation and growth is studied with special focus on the initiation and branching of cracks. Polycarbonate plates are used as substrates subjected to an acetone environment. Experimental procedures for examining initiation and branching in polycarbonate are presented. An optical microscope is employed to study the substrate.The attack at initiation is quantified from pits found on the surface, and pits that act as origin for cracks is identified and the distribution is analysed. A growth criterion for surface cracks is formulated from the observations, and it is used to numerically simulate crack growth. The cracks are seen to coalesce, and this phenomenon is studied in detail. Branching sites of cracks growing in the bulk of polycarbonate are inspected at the sample surface. It is found that the total width of the crack branches are approximately the same as the width of the original crack. Also, angles of the branches are studied. Further, for comparison the crack growth in the bulk is simulated using a moving boundary problem based algorithm and similar behaviour of crack branching is found.
Книги з теми "Irradiation assisted stress corrosion cracking":
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M, Chung H., Argonne National Laboratory, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., eds. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys. Washington, DC: U.S. Nuclear Regulatory Commission, 2000.
M, Chung H., Argonne National Laboratory, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., eds. Irradiation-assisted stress corrosion cracking of model austenitic stainless steel alloys. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2000.
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.
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.
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.
R, Rungta, American Society of Mechanical Engineers. Winter Meeting, American Society of Mechanical Engineers. Pressure Vessels and Piping Division. Materials and Fabrication Section., American Society of Mechanical Engineers. Pressure Vessels and Piping Division. Design and Analysis Committee., and Metal Properties Council, eds. Predictive capabilities in environmentally assisted cracking: Presented at the Winter Annual Meeting of the American Society of Mechanical Engineers, Miami Beach, Florida, November 17-22, 1985. New York, N.Y. (345 E. 47th St., New York 10017): ASME, 1985.
Kane, Russell D. Environmentally Assisted Cracking: Predictive Methods for Risk Assessment and Evaluation of Materials, Equipment, and Structures (ASTM Special Technical ... (Astm Special Technical Publication// Stp). Astm Intl, 2000.
Частини книг з теми "Irradiation assisted stress corrosion cracking":
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Chung, H. M., W. E. Ruther, R. V. Strain, W. J. Shack, and T. M. Karlsen. "Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steels." In Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 931–39. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2013. http://dx.doi.org/10.1002/9781118787618.ch98.
Wang, Mi, Miao Song, Gary S. Was, L. Nelson, and R. Pathania. "Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-Base Alloys in Light Water Reactors Environments Part II: Stress Corrosion Cracking." In The Minerals, Metals & Materials Series, 2177–88. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_146.
Wang, Mi, Miao Song, Gary S. Was, L. Nelson, and R. Pathania. "Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-Base Alloys in Light Water Reactors Environments Part II: Stress Corrosion Cracking." In The Minerals, Metals & Materials Series, 961–72. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-68454-3_70.
Hojná, Anna, Miroslava Ernestová, Ossi Hietanen, Ritva Korhonen, Ludmila Hulinová, and Ferenc Oszvald. "Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 1257–75. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_77.
Hojná, Anna, Miroslava Ernestová, Ossi Hietanen, Ritva Korhonen, Ludmila Hulinová, and Ferenc Oszvald. "Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1257–72. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch131.
Fyfitch, Steve, Sarah Davidsaver, and Kyle Amberge. "Irradiation-Assisted Stress Corrosion Cracking Initiation Screening Criteria for Stainless Steels in PWR Systems." In The Minerals, Metals & Materials Series, 2211–20. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_148.
Teysseyre, S., J. H. Jackson, P. L. Andresen, P. Chou, and B. Carter. "Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X-750 Exposed to BWR Environments." In The Minerals, Metals & Materials Series, 2243–53. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_151.
McMurtrey, M. D., and G. S. Was. "Role of Slip Behavior in the Irradiation Assisted Stress Corrosion Cracking in Austenitic Steels." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 1383–95. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_85.
Mcmurtrey, M. D., and G. S. Was. "Role of Slip Behavior in the Irradiation Assisted Stress Corrosion Cracking in Austenitic Steels." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 1383–94. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch144.
Fyfitch, Steve, Sarah Davidsaver, and Kyle Amberge. "Irradiation-Assisted Stress Corrosion Cracking Initiation Screening Criteria for Stainless Steels in PWR Systems." In The Minerals, Metals & Materials Series, 995–1004. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-68454-3_72.
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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Rebak, Raul B. "Resistance of Ferritic Steels to Stress Corrosion Cracking in High Temperature Water." In ASME 2013 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/pvp2013-97352.
Austenitic stainless steels such as type 304 and 316 are susceptible to stress corrosion cracking in high temperature water environments typical of boiling water reactors (BWR) and pressurized water reactors (PWR). The accumulation over time of irradiation dose on the austenitic materials increases further their susceptibility to environmental cracking. Ferritic steels containing chromium are less susceptible to irradiation damage such as void swelling. Ferritic steels also offer desirable higher thermal conductivity and lower thermal expansion coefficient. Little is known however about the stress corrosion cracking behavior of ferritic steels in high temperature water. Crack propagation rate studies were conducted using four types of wrought and welded ferritic steels (5 to 17% Cr) in high purity water at 288°C containing dissolved oxygen or dissolved hydrogen. Results show that the ferritic steels are notably more resistant to environmental assisted cracking than the austenitic materials.
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Eason, Ernest D., and Raj Pathania. "Disposition Curves for Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments." In ASME 2015 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/pvp2015-45323.
This paper presents irradiation-assisted stress corrosion cracking (IASCC) disposition curves developed in a multi-year international data collection, data review and modeling project. More than 800 IASCC crack growth rate (CGR) data points were collected from six laboratories worldwide, and an international panel of experts reviewed and ranked the data. The better-ranked data were used to calibrate empirical models for IASCC CGR in boiling water reactor (BWR) normal water chemistry (NWC) and hydrogen water chemistry (HWC) environments and in pressurized water reactor (PWR) primary water environments. The mean models were shifted upward to the 75th percentile of the calibration data for use as crack disposition curves. The disposition curves are presented in this paper and compared with data used for fitting and data not used for fitting, including field data from BWR core shrouds and additional laboratory data. The paper is intended as a basis document for possibly incorporating the new disposition curves in the ASME code.
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Matsubara, Toru, and Yuichi Mogami. "Stress Evaluation Method of Baffle Former Bolt and its Maintenance Program." In ASME 2016 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/pvp2016-63971.
Baffle Former Bolt (BFB) is a fastening part of Reactor Vessel Internals (RVI) of PWR. BFB is made of type 347 or 316CW (cold work) stainless steel and it is known to have the risk of cracking caused by Irradiation Assisted Stress Corrosion Cracking (IASCC) under high neutron flux and tensile stress. To evaluate the time to crack of BFB, BFB’s time-dependent stress change caused by irradiation creep (relaxation) and by the swelling deformation of a baffle structure should be obtained. The authors have developed the finite element (FE) analysis method to calculate time-dependent stress of BFB considering the irradiation effects. The method combines two kinds of models; “global model” to calculate the deformation of whole baffle structure and “local model” to calculate the peak stress at the stress concentrated area under the bolt head. Incorporating the above calculation method, a new BFB inspection and evaluation guideline has been established in Japan. The concept of the guideline is also outlined in the paper.
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Li, Yongkui, Yoshiyuki Kaji, and Takahiro Igarashi. "Study of Weld Residual Stress Field in the Girth Seam H6A of Core Shroud of Boiling Water Reactor." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29269.
Many accidents have occurred in nuclear power plants due to the intergranular stress corrosion cracking (IGSCC) in the heat affected zone (HAZ) of welded joint of the core shroud of boiling water reactors (BWRs) in past years. The IGSCC is considered to be caused by the synergistic roles of corrosion environment, neutron irradiation and the welding residual stress. After several decades, the degradation of Type 316L low carbon stainless steel used in the core shroud occurs due to the neutron irradiation and thermal cycles. The degradation can be referred to the irradiation hardening, segregation of the local chemical composition at grain boundaries and swelling. The synergistic effects of those eventually lead to the initiation and propagation of the irradiation-assisted stress corrosion cracking (IASCC) in core shroud for long operation. The HAZ of the girth seams H6a in the core shroud are sensitive to the stress corrosion cracking. We are focusing on the weld residual stress field around the girth seam H6a in the core shroud as weld. The analysis work adopted different approaches in ABAQUS to simulate the weld residual stress, and they are Static General Analysis (SGA) and Fully Coupled Temperature-Displacement Analysis (FCTDA) respectively. The former is much simple to finish the progress while cannot obtain much accurate results at the boundaries of beads due to the discontinuous temperature field in the model. The later analysis gave the much accurate results comparing with the experimental results. The axial stress field in the crossing section of the wall of the core shroud was also clarified.
6
Elliot, Barry J., Vikram N. Shah, and Yung Y. Liu. "Effective Approaches for Managing Aging Effects in BWR Reactor Coolant System Components for License Renewal." In ASME 2003 Pressure Vessels and Piping Conference. ASMEDC, 2003. http://dx.doi.org/10.1115/pvp2003-2164.
This paper discusses management of aging effects for reactor coolant pressure boundary components in boiling water reactors (BWRs): loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of vessel internals made of cast austenitic stainless steel; cracking of the top guide due to irradiation-assisted stress corrosion cracking; cracking of the core shroud and reactor coolant system piping due to intergranular stress corrosion cracking; cracking of the small bore piping due to high-cycle thermal fatigue; and loss of preload in the pressure boundary bolting. The applicants for license renewal of BWR plants have proposed different approaches for managing these aging effects such that the intended functions of the affected components will be maintained, consistent with the current licensing basis, for the period of extended operation. The NRC staff has performed safety evaluation of these approaches and found them acceptable for adequately managing the aging effects during the period of extended operation. The technical bases for the acceptance are presented in this paper.
7
Takakura, Kenichi, Kiyotomo Nakata, Noboru Kubo, Koji Fujimoto, and Kimihisa Sakima. "IASCC Evaluation Method of Irradiated Cold Worked 316SS Baffle Former Bolt in PWR Primary Water." In ASME 2009 Pressure Vessels and Piping Conference. ASMEDC, 2009. http://dx.doi.org/10.1115/pvp2009-77279.
Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as degradation of core internal components in light water nuclear reactor. To clarify the IASCC initiation conditions of baffle former bolt (BFB), constant load stress corrosion cracking (SCC) tests were carried out in simulated PWR primary water (290, 320, 340°C) using C-ring type specimens. Based on the SCC test results, IASCC initiation time becomes shorter with increasing fluence and increasing applied stress, IASCC initiation threshold stress becomes lower with increasing fluence. A test temperature effect was observed in SCC initiation time, but it was not clear the effect of test temperature for SCC initiation threshold stress. These results suggest that IASCC initiation threshold criteria can be described with stress in specimen and fluence. This paper describes the whole evaluation procedure to secure structural integrity of irradiated baffle structure in PWR primary environments, including the threshold stress diagram of IASCC initiation and the irradiation creep formula.
8
Kiss, E. "Component Reliability Considerations for New Designs and Extended Operation of Boiling Water Reactor (BWRs)." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48864.
To achieve high reliability for new designs and extended operation of Reactor Pressure Vessels and Internals it is mandatory to apply the technical knowledge gained during operation of the existing Plants to assure that sufficient “Margin” is built into the new design. This paper discusses the importance of four key structural degradation mechanisms that have been shown by operational experience to affect the reliability of the BWR. These are: 1) Stress Corrosion Cracking (IGSCC) of Stainless Steel and Nickel-based Alloys; 2) Irradiation Assisted SCC (IASCC) of Stainless Steel and Nickel-based Alloys; 3) Irradiation Embrittlement of RPV low alloy Steel; 4) Corrosion Assisted Fatigue of Carbon and Low Alloy Steel. While the focus of this paper is the BWR, the mechanisms discussed are equally applicable to the PWR, although the water chemistry effects and mitigations will be different.
9
Ge´rard, Robert, and Fre´de´ric Somville. "Situation of the Baffle-Former Bolts in Belgian Units." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75445.
The baffle to former bolts are used in Pressurized Water Reactors to attach the baffle plates to the former plates in the reactor vessel lower internals. The resulting structure forms a boundary for the flow of coolant and provides lateral support to the fuel assemblies. Some edge bolts are also present, assembling together the baffle plates. After an operating time of the order of 120 000 hours, some bolts exhibit cracking at the junction of the head and the shaft of the bolt. Examinations of failed bolts have made it possible to identify the cause of cracking as irradiation assisted stress corrosion cracking (IASCC). Up to now, baffle bolt cracking has been detected in units older than 15 years, where the baffle bolts are not cooled (no holes in the former to allow a water flow on the bolt shaft). In Belgium the concerned unit are Tihange 1 and Doel 1–2. The paper summarizes the experience with baffle bolts cracking in Belgian units and the strategy implemented to mitigate this problem, consisting of structural integrity analyses, baffle bolts inspections and replacement, and research programs in the field of IASCC, including examinations of highly irradiated replaced bolts.
10
Fedorova, Valentina, and Boris Margolin. "Method for Estimation of Pressure Vessel Internals Lifetime on IASCC Criterion." In ASME 2013 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/pvp2013-97949.
Austenitic stainless steels are used extensively as structural materials in the internal components of reactor pressure vessels. However, high neutron doses lead to a significant reduction in the fracture resistance of these steels in water environment. Irradiation assisted stress corrosion cracking (IASCC) of internals has been observed in pressurized water reactors (PWRs). In the present work the IASCC model of the irradiated austenitic steels in PWR water has been developed. On the basis of analysis of available experimental data IASCC mechanism is proposed. Based on this mechanism, the dependence of fracture stress under IASCC on neutron dose is derived. For its construction the following assumptions were made. 1. Creep rate due to grain boundary sliding does not depend on neutron dose. 2. Fracture strain due to grain boundary sliding decreases when neutron dose increases. 3. There is an apparent stress threshold below which IASCC initiation does not occur in PWR environment. Life prediction analysis for IASCC is performed on the basis of linear rule of damage accumulation.
Звіти організацій з теми "Irradiation assisted stress corrosion cracking":
1
Bell, G. (Irradiation assisted stress corrosion cracking). Office of Scientific and Technical Information (OSTI), April 1990. http://dx.doi.org/10.2172/7010172.
Teysseyre, S. Effect of Swelling on Irradiation-Assisted Stress Corrosion Cracking. Office of Scientific and Technical Information (OSTI), July 2017. http://dx.doi.org/10.2172/1483829.
Simonen, E. P., R. H. Jones, and S. M. Bruemmer. Irradiation-assisted stress corrosion cracking considerations at temperatures below 288{degree}C. Office of Scientific and Technical Information (OSTI), March 1995. http://dx.doi.org/10.2172/46591.
Chen, Y., O. K. Chopra, Eugene E. Gruber, and William J. Shack. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments. Office of Scientific and Technical Information (OSTI), June 2010. http://dx.doi.org/10.2172/1224951.
Teysseyre, Sebastien Paul. Study of the Effect of Swelling on Irradiation Assisted Stress Corrosion Cracking. Office of Scientific and Technical Information (OSTI), September 2016. http://dx.doi.org/10.2172/1364496.
Jackson, J. H., S. P. Teysseyre, and M. P. Heighes. Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions. Office of Scientific and Technical Information (OSTI), June 2017. http://dx.doi.org/10.2172/1408502.
Gary S. Was. Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking. Office of Scientific and Technical Information (OSTI), March 2009. http://dx.doi.org/10.2172/950834.
Teysseyre, Sebastien. Irradiation Programs and Test Plans to Assess High-Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility. Office of Scientific and Technical Information (OSTI), March 2015. http://dx.doi.org/10.2172/1177229.
Chen, Y., O. K. Chopra, W. K. Soppet, Nancy L. Dietz Rago, and W. J. Shack. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Halden Phase-II Irradiations. Office of Scientific and Technical Information (OSTI), September 2008. http://dx.doi.org/10.2172/1224948.
Chung, H. M., and W. J. Shack. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals. Office of Scientific and Technical Information (OSTI), January 2006. http://dx.doi.org/10.2172/915725.
