Lifetime prediction of components in Alloy 600 is a major concern for nuclear power plants. Alloy 600 components have been shown to be susceptible to stress corrosion cracking (SCC). In the 1990’s, an engineering model was developed in order to predict the life time as a function of the main macroscopic parameters (stress, environment, material), based on laboratory results. This model has since been used to predict the ranking of various Alloy 600 components, using the knowledges of the manufacturing and service conditions for each component. It was applied successfully in the case of forged control rod drive mechanism (CRDM) nozzles. However, it was found necessary to improve this model to account for the strain history of the different components. Predictions using the model, investigated from an array of test results on Alloy 600 in laboratory primary water, have demonstrated that the time for initiation differed significantly according to the strain path applied to the specimen. The present work is dedicated to assess SCC results from samples with different strain paths and different level of cold work in order to better understand the manufacturing conditions on SCC. The samples are machined in three different directions and tested at different durations in order to model the time for transition (transition between slow and fast propagation) as a function of cold work, strain path and stress. Thermomechanical treatments are also applied on two different heats of Alloy 600: forged WF675 (very susceptible to SCC in as received conditions) and rolled 78456/337 (non susceptible to SCC in as-received conditions) in order to transform the forged microstructure into a microstructure close to the rolled microstructure and vice-versa. These microstructures are then tested in primary conditions and the results are compared to the results obtained on as-received material in order to get a better understanding of manufacturing process and microstructure parameters regarding SCC behaviour.

Shear spinning of Haynes 282, Alloy 718, Alloy 600 and AISI 316L was done using several tool feeds and mandrel clearances. Multi passing of the materials was limited due to strain hardening and circumferential cracking except for AISI 316L. The effect of the tool feed and the mandrel clearance on the successful forming of the materials was established. The successfully spun samples were solution heat treated at varying temperatures and holding times to establish a range of grain sizes and hardness levels. An aging heat treatment process was performed for Haynes 282 and Alloy 718 to achieve precipitation strengthening. The micro hardness measurements were conducted for the materials prior to spinning and after spinning. The same was also done after the various heat treatment processes. Grain size mapping was conducted by the use of lineal intercept methods. Comparison of the results in terms of grain sizes and hardness values was done. The temperature ranges suitable for full recrystallization of the materials after the shear spinning were identified and the effect of the holding times on the grain growth established. Comparison with unspun samples showed that the heat treatment times required to achieve comparative hardness and grain sizes were distinctively different.

The stress corrosion cracking (SCC) condition for Alloy 600 was determined for various stress modes: constant-strain standard C-ring, and indentation, used to localize cracks for interrogation with x-ray techniques such as micro Laue diffraction (MLD). The SCC cracks nucleated on both the indentation edge, where finite element analysis showed that the maximum residual tensile stresses lie, and the surface in tension (bulge) on 150-kgf conically indented mill-annealed specimens (0.02 wt% C) in de-aerated solution of 10% caustic at 150 mVRE (pseudo-reference: A600), 315 OC for 48 hr. On the C-rings, the cracks nucleated at the lateral outer surface of apex, where maximum tensile stresses lie, in less than 12 hours, and propagated into the cross section. Also, corrosion tests on as-received A600 30-min ZrO2 surface mechanical attrition treated (SMAT) specimens suggested an intergranular attack type of behavior in 50% caustic at 210 mVRE (pseudo-reference: A600), 280°C for 24 hr.

The stress corrosion cracking (SCC) condition for Alloy 600 was determined for various stress modes: constant-strain standard C-ring, and indentation, used to localize cracks for interrogation with x-ray techniques such as micro Laue diffraction (MLD). The SCC cracks nucleated on both the indentation edge, where finite element analysis showed that the maximum residual tensile stresses lie, and the surface in tension (bulge) on 150-kgf conically indented mill-annealed specimens (0.02 wt% C) in de-aerated solution of 10% caustic at 150 mVRE (pseudo-reference: A600), 315 OC for 48 hr. On the C-rings, the cracks nucleated at the lateral outer surface of apex, where maximum tensile stresses lie, in less than 12 hours, and propagated into the cross section. Also, corrosion tests on as-received A600 30-min ZrO2 surface mechanical attrition treated (SMAT) specimens suggested an intergranular attack type of behavior in 50% caustic at 210 mVRE (pseudo-reference: A600), 280°C for 24 hr.

Stress corrosion cracking (SCC) of alloy 600 is regarded as one of the most important challenges to nuclear power plant operation worldwide. This study investigates two heats of alloy 600 (forged control rod drive mechanismnozzle and rolled divider plate) in order to obtain a better understanding of the effects of the material parameter on the SCC phenomenon. The experimental approach was designed to determine the effect of the manufacturing process (forged vs. rolled), the cold-work (as-received vs. cold-worked) and the strain path (monotonic vs. complex) on SCC of alloy 600. Specimens with different strain paths have been produced from two materials representative of plant components and tested in high temperature (360°C) high pressure primary water environment. The manufacturing process has been proven to have a great effect on the stress corrosion cracking behaviour of alloy 600. Indeed, the SCC susceptibility assessment has demonstrated that the rolled materialis resistant to SCC even after cold work, whereas the forged material is susceptible in the as-received state. Microstructural characterisations have been undertaken to explain these differences in SCC behaviour. The carbide distribution is the main microstructural parameter influencing SCC but the misorientation, in synergy with the carbide distribution, has been proven to give a better representation of the materials SCC susceptibilities.

Plusieurs composants présents dans les réacteurs à eau sous pression (REP) ont été fabriqués en alliage 600, un alliage base nickel contenant environ 16% de chrome. Le retour d’expérience, comme les études de laboratoire, montrent une sensibilité à la corrosion sous contrainte (CSC) de cet alliage en milieu primaire.Des études antérieures ont permis de développer un modèle d’amorçage basé sur une approche macroscopique et dépendant de différents paramètres tels que : la température, la contrainte et la microstructure du matériau. Cependant, ce modèle manque de robustesse car l’effet de la teneur en hydrogène dissous et l’effet de l’histoire de chargement mécanique ne sont pas considérés et les effets microstructuraux ne le sont que partiellement.Ces travaux de thèse ont comme objectif principal de développer un modèle local prévoyant le temps d’amorçage des fissures de CSC en fonction de paramètres locaux liés à la microstructure du matériau (précipitation intergranulaire), à l'environnement (température, et teneur en hydrogène dissous) et aux contraintes locales aux joints des grains. Cette étude comprend donc la caractérisation des matériaux (analyse chimique, microstructure et comportement mécanique) et la réalisation des essais d’oxydation et de corrosion sous contrainte, ainsi que leur interprétation.Le modèle local développé est basé sur des grandeurs physiques et enchaine les différentes étapes de CSC à savoir l’incubation, l’amorçage et la propagation des fissures. Pour construire ce modèle, nous avons considéré la formation de pénétrations d’oxyde aux joints de grains comme une étape-clé dans l’amorçage des fissures de CSC. Pour cela, une cinétique d’oxydation intergranulaire pour l’alliage 600 a été identifiée. De plus, un critère d’amorçage des fissures de CSC a été déterminé en couplant contrainte locale et profondeur d’oxydation intergranulaire critique. Enfin, l’étape de propagation des fissures a été modélisée à partir d’une base de données rassemblant les profondeurs de fissure atteintes en fonction du temps d’essai pour différentes conditions expérimentales. / Several components present in the primary circuit of Pressurized Water Reactors (PWR) of nuclear power plants were manufactured with Alloy 600, a nickel base alloy containing 16 wt.% chromium. Operating experience of PWRs and laboratory tests showed that Alloy 600 is susceptible to stress corrosion cracking (SCC).Previous studies have allowed developing an initiation model based on a macroscopic approach and depending on several parameters such as temperature, applied stress and material microstructure. However, this model suffers from a lack of accuracy: dissolved hydrogen content and mechanical loading history effects are not considered and the microstructure effects (such as intergranular precipitation) are only partially taken into account.The aim of this study is to develop a ‘local’ model predicting stress corrosion cracking initiation time, based on physical mechanisms and local parameters related to the material microstructure (intergranular precipitation), the environment chemistry (temperature and dissolved hydrogen content) and stress concentration at grain boundaries. The local model relies on a cracking scenario with three main steps: incubation, initiation and crack extension.The formation of intergranular oxide penetration was assumed to be a key step in SCC initiation. For this purpose, oxidation tests were performed in simulated primary water. The intergranular oxidation kinetics of Alloy 600 was studied and the effects of intergranular carbide precipitation, dissolved hydrogen content and temperature were investigated. In addition, a cracking criterion coupling a critical local stress and a critical intergranular oxide depth was estimated. Finally, a sigmoid crack growth law was used to simulate both the slow and fast propagation steps. The local model was validated using a database built from the results of SCC tests performed on Alloy 600 and gathering the crack depths reached as a function of test duration for different experimental conditions (material microstructure, loading conditions).

Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 and similar Ni-Cr-Fe alloys is regarded as one of the most important challenges to nuclear power plant operation. During the past decades the majority of research has focused on PWSCC crack growth rate measurements in order to assess the lifetime of real components and to develop empirical models for crack propagation. However, the incubation and initiation stages of PWSCC have the same or even greater importance than the propagation stage, particularly because SCC can be undetected for more than 20 years before the occurrence of a rapid and catastrophic failure. There is, therefore, the scientific need to understand the mechanisms playing a fundamental role in the formation and development of intergranular cracks embryo, the so-called SCC initiation "precursor events", in order to be able to predict and mitigate the occurrence of PWSCC. Amongst all the models proposed for SCC initiation, the internal oxidation mechanism proposed by Scott and Le Calvar in 1992 appears to be the most comprehensive. Although the internal oxidation mechanism is widely accepted, it still requires further elucidation, especially in terms of enhanced grain boundary diffusivity and the role of intergranular carbides on the oxidation mechanism. The present work has focused on the initial stages of intergranular oxidation of solution-annealed (SA) and thermally-treated (TT) Alloy 600 with the aim of understanding the active mechanism responsible for the enhanced intergranular oxide penetration kinetics. The material was tested in simulated PWR primary water at 320°C, high-pressure hydrogenated-steam at 400°C and low-pressure H2-steam environment at 480°C at potential more reducing than the Ni/NiO equilibrium. The detailed microstructural characterization was conducted using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and analytical transmission electron microscopy (ATEM) and demonstrated that Alloy 600SA is susceptible to diffusion-induced grain boundary migration (DIGM), preferential intergranular oxidation (PIO) and localised Cr and Fe depletions at the grain boundaries. The similar analyses performed on Alloy 600TT demonstrated reduced susceptibility to PIO and grain boundary migration. Further, detailed analyses confirmed that intergranular carbides were readily oxidized/consumed in all 3 environments and acted as Cr reservoir/O trap. These results shed additional light on the "precursor events" for PWSCC of Alloy 600, especially on the mechanism responsible for the enhanced Cr and O diffusivity and on the mechanism responsible for the enhanced Alloy 600TT SCC initiation resistance. Moreover, the strong similarities in the Alloy 600 oxidation behaviour observed for the 3 different environments and at the 3 different temperatures suggested that the same PIO mechanism is active in both steam and water and at temperatures between 320°C and 480°C. These results strongly support the possibility of using the low-pressure H2-steam environment as a substitute environment to accelerate PWSCC initiation without changing the mechanism.

Joints between carbon steel and Alloy 600, containing Alloy 82 weld metal, were exposed to two steam-hydrogen environments considered to simulate exposure to primary water conditions in a Pressurized Water Reactor (PWR) or Canada Deuterium Uranium (CANDU) reactor. The welds were found to have elevated and variable iron contents due to dilution by carbon steel during welding. This gave the Alloy 82 weld, near the inner surface of the component, an iron content approaching that of Alloy 800. A potentially protective external iron oxide film formed on the inner surface of the weld. However, the chromium content throughout the weld is below that which would form an external chromium oxide. The results indicate that low chromium content causes internal oxidation throughout the weld and potentially below the external iron oxide which could lead to Primary Water Stress Corrosion Cracking (PWSCC).

Joints between carbon steel and Alloy 600, containing Alloy 82 weld metal, were exposed to two steam-hydrogen environments considered to simulate exposure to primary water conditions in a Pressurized Water Reactor (PWR) or Canada Deuterium Uranium (CANDU) reactor. The welds were found to have elevated and variable iron contents due to dilution by carbon steel during welding. This gave the Alloy 82 weld, near the inner surface of the component, an iron content approaching that of Alloy 800. A potentially protective external iron oxide film formed on the inner surface of the weld. However, the chromium content throughout the weld is below that which would form an external chromium oxide. The results indicate that low chromium content causes internal oxidation throughout the weld and potentially below the external iron oxide which could lead to Primary Water Stress Corrosion Cracking (PWSCC).

Um dos principais mecanismos de falha que causam riscos de fratura a reatores de água pressurizada é a corrosão sob tensão de ligas metálicas em água do circuito primário (CSTAP). É causada por uma combinação das tensões de tração, meio ambiente em temperatura e microestruturas metalúrgicas susceptíveis. Ela pode ocorrer, dentre outros locais, nos bocais do mecanismo de acionamento das barras de controle. Essa fratura pode causar acidentes que comprometem a segurança nuclear através do bloqueio das barras de controle e vazamentos de água do circuito primário reduzindo a confiabilidade e a vida útil do reator. O objetivo desta Tese de Doutorado é o estudo de modelos e uma proposta de modelagem para fraturas por corrosão sob tensão em liga 75Ni15Cr9Fe (liga 600), em água de circuito primário de reator de água pressurizada nesses bocais. São superpostos modelos eletroquímicos e de mecânica da fratura e validados com dados obtidos em experimentos e na literatura. Na parte experimental foram utilizados resultados obtidos pelo CDTN no equipamento recém-instalado de ensaio por taxa de deformação lenta. Na literatura está proposto um diagrama que exprime a condição termodinâmica de ocorrerem diversos modos de CSTAP na liga 600: partiu-se de diagramas de potencial x pH (diagramas de Pourbaix), para a liga 600 imersa em água primária à alta temperatura (3000C a 3500C). Sobre ele, determinaram-se os submodos de corrosão, a partir de dados experimentais. Em seguida acrescentou-se uma dimensão adicional ao diagrama, correlacionando uma variável a que se denominou fração de resistência à corrosão sob tensão. No entanto, é possível acrescentar-se outras variáveis que exprimem a cinética de iniciação e/ou crescimento de trinca, provenientes de outras modelagens de CSTAP. A contribuição original deste trabalho se insere nessa fase: partindo-se de uma condição de ensaio de potencial versus pH, foram iniciadas as modelagens de um modelo empírico-comparativo, um semi-empírico-probabilístico, um de tempo de iniciação e um de taxa de deformação, a partir dos ensaios experimentais e superpostas a essa condição. Esses exprimem respectivamente a susceptibilidade à CSTAP, o tempo de falha, e nos dois últimos o tempo de iniciação de falha por corrosão sob tensão. Os resultados foram comparados com os da literatura e se mostraram coerentes. Através desse trabalho, obteve-se uma metodologia de modelagem a partir de dados experimentais. / One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism 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 accidents that reduce nuclear safety by blocking the rods displacement and may cause leakage of primary water, reducing the reactors life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickelbased Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found 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 Alloy 600 in high temperature primary water (3000C till 3500C). Over it, were located the PWSCC submodes, using experimental data. It was added a third parameter called stress corrosion strength fraction. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potencial versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our experimental data. The results were compared with the literature and it showed to be coherent. From this work was obtained a modeling methodology from experimental data.