OMAR FERNANDES ALY
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Artigo IPEN-doc 22989 Reviewed software methodology to stress corrosion prediction2016 - ALY, OMAR F.; MATTAR NETO, MIGUELStress Corrosion Cracking (SCC) is a sudden and difficult-to-predict severe degradation mode of failure of nuclear, petrochemical, and other industries. This is a review of a development proposal for methodological software for modeling SCC based on: the failure propensity plus a kinetic model link which better describes its evolution.Artigo IPEN-doc 22132 Compared modeling study of primary water stress corrosion cracking at dissimilar weld of alloy 182 of pressurized wter nuclear reactor according to hydrogen concentration2015 - ALY, OMAR F.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICA M.A.M.; LIMA, LUCIANA I.L.Capítulo IPEN-doc 21188 Stress corrosion cracking2014 - ALY, O.F.; MATTAR NETO, M.Capítulo IPEN-doc 11449 Modeling of primary water stress corrosion cracking (PWSCC) at control rod drive mechanism (CRDM) nozzles of pressurized water reactors (PWR)in roadside soils2005 - ALY, O.F.; PAES de ANDRADE, A.H.; MATTAR NETO, M.; SZAJNBOK, M.; TOTH, H.J.Capítulo IPEN-doc 14659 Modeling of primary water stress corrosion cracking (PWSCC) at control rod drive mechanism (CRDM) mozzles of pressurized water reactors (PWR)2007 - ALY, O.F.; ANDRADE, A.H.P.; MATTAR NETO, M.; SZAJNBOK, M.; TOTH, H.J.One of the main causes of failure in pressurized water reactors (PWR) is the stress corrosion cracking (SCC) at control rods drive mechanism (CRDM) nozzles, produced by tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. Such cracks can cause accidents that reduce nuclear safety by blocking the rods displacement at CRDM and/or leakage of primary water. This paper will present a preliminary development of a model to predict such damage, including initiation and propagation of primary water SCC (PWSCC). The model assumes the Pourbaix potential-pH diagram for Alloy 600 on the typical PWR environment, primary water at high temperature. Over this diagram, the region where the SCC submodes can occur is plotted. Submodes are determined by regions of potential where various modes of surface material-environment interactions can occur, such as stress corrosion, pitting, generalized corrosion or passivation. Over these regions an empiricalprobabilistic is linked to a strain rate damage model that can evaluate the time to failure and the damage parameter, as a function of total stress at the material surface, its temperature and other factors depending on environment-material combination and thermomechanical treatment of this alloy.Artigo IPEN-doc 20214 A methodology for modeling stress corrosion cracking with an example2014 - ALY, OMAR F.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICA M de A.M.Artigo IPEN-doc 20210 Preliminary stress corrosion cracking modeling study of a dissimilar material weld of alloy (Inconel) 182 with stainless steel 316 in pressurized water nuclear reactor2014 - ALY, OMAR F.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICA M de A.M.Artigo IPEN-doc 20209 Modeling of tests of primary water stress corrosion cracking of alloy 182 of pressurized water reactor according to EPRI and USNRC recommendations2014 - ALY, OMAR F.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICA M de A.M.; LIMA, LUCIANA I.L.Artigo IPEN-doc 12367 Results on modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors2007 - ALY, OMAR F.; PAES de ANDRADE, ARNALDO H.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICAOne of the main failure mechanisms that cause risks to pressurized water reactors (PWR) is the primary water stress corrosion cracking (PWSCC) occurring in alloys like the alloy 600 (75Ni-15Cr-9Fe). It can occur, besides another places, at the control rod drive mechanism (CRDM) nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause problems that reduce nuclear safety by blocking the displacement of the control rods and may cause leakage of primary water that requires repair or replacement of the reactor pressure vessel head. In this work it is performed a study of the existing models and proposed a new approach to assess the primary water stress corrosion cracking in nickel-based Alloy 600 CRDM nozzles . The proposed model is obtained from the superposition of electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained from CDTN-Brazilian Nuclear Technology Development Center, in a SSRT equipment, according with Schvartzman et al.(2005). Staehle (1992) has built a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC submodes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Nickel in high temperature primary water (3000 C till 3500 C). The PWSCC submodes were located over it, using experimental data. Also, a third parameter called ìstress corrosion strength fractionî was added. However, it is possible to superimpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. It is important to mention that the main contribution of this work is from a specific experimental condition of potencial versus pH, it was superposed, an empiric-comparative, according with Staehle (1992), a semi-empiricalprobabilistic according with Gorman et al. (1994), an initiation time according with Garud (1997), and a strain rate damage according with Boursier et al.(1995)-models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. The results were compared with the literature and it showed to be coherent. From this work was obtained a modeling methodology from experimental data. The SSRT tests had been realized at a condition of potential =ñ621 mVSHE and pH= 7.3. The PWSCC strength fraction evaluated was 0.95: this initiates an empirical-comparative model. The initiation time model obtained was according Eq. (1) with ti in days, T in K, and σ in MPa. The model was planned for constant load, but some assumptions were done to obtain (1) from slow strain rate tests.Artigo IPEN-doc 16590 Modelos de processos de fratura por corrosão sob tensão em componentes de reatores nucleares - alguns resultados e exemplos2011 - ALY, OMAR F.; MATTAR NETO, MIGUEL; SCHVARTZMAN, MONICA M.A.M.