Quantitative study of solute segregation phenomenon is crucial in material science. In the present study, the influence of thermal ageing and ion irradiation (both at 400°C) on intergranular P segregation behavior was studied by a correlative Electron Backscattering Diffraction (EBSD)/Transmission Kikuchi Diffraction (TKD)/Atom Probe Tomography (APT) methodology. The co-segregation of P, C, Si, Mn, Ni, Cr and Mo appears at Grain Boundaries (GBs). There is an increased P segregation for all the detected GB types in both thermally-aged and ion-irradiated welds. Thermodynamic and kinetic predictions confirm the Radiation Enhanced Segregation (RES) mechanism of P at GBs under ion irradiation. The ternary Fe-P-C model overestimates P segregation level in a multicomponent system.

During operating lifetime of a nuclear power plant, the pressure boundary components, usually in 16MND5 (SA508 Cl. 3) steel, are subjected to thermal aging that may induce phosphorus intergranular segregation. This is known to decrease brittle fracture stress and induce intergranular fracture, which causes reversible temper embrittlement. Large components such as the reactor pressure vessel presents microstructure heterogeneity due to forging and welding. Tempered martensite can locally be found in the nominally bainitic steel. The question of microstructure susceptibility to reversible temper embrittlement has been raised in literature suggesting that tempered martensite has higher sensibility to phosphorus intergranular segregation than tempered bainite. In this paper, a comparative study of intergranular segregation in tempered bainitic and martensitic 16MND5 steel after thermal aging is conducted using energy dispersive X-ray spectroscopy on transmission electron microscope. Focus ion beam is used to precisely prepare lamellae at prior austenite grain boundaries on step cooled bainitic and martensitic samples. Different types of grain boundaries are identified using precession assisted crystal orientation mapping and then the local chemical composition is analyzed. Phosphorus, nickel, manganese and molybdenum were observed at all types of grain boundaries in both microstructures. The segregation levels are higher in prior austenitic grain boundaries than in other types of boundaries for both microstructures. Taking into account the segregation amount in different types of boundaries, the phosphorus bulk depletion is shown negligible. The results show similar segregation levels in the two studied microstructures, indicating that the susceptibility to thermally induced grain boundary segregation between the tempered bainite and tempered martensite is the same.

This paper reviews the prediction methods used in France for the assessment of irradiation and thermal aging effects. The irradiation embrittlement trend correlation (ETC) was revised in 2007, and a further revision depends on the significant surveillance database expansion that will occur over the next few years. New data will be obtained at high fluences (∼8 × 1023 n/m², E > 1 MeV). The thermal aging ETC was established in the 1990s, and laboratory aging programs were led for its validation. It appears to be too conservative when applied to underclad heat-affected zones.

Phosphorus intergranular segregation is known to influence the fracture properties of steels by decreasing grain boundary cohesion and induce intergranular fracture. Different techniques such as Angle-Resolved X-ray Photoelectron Spectroscopy (AR-XPS), Wavelength Dispersive Spectroscopy (WDS), Energy Dispersive X-ray Spectroscopy coupled with Scanning Transmission Electron Microscope (STEM-EDX), and Atom Probe Tomography (APT) can be used to quantify intergranular segregation. Although many studies of this phenomenon were conducted over the last decades, there are rarely direct comparisons between different techniques and there is still a need of reliable and comparable quantification methods for grain boundary segregation. This study cross compares four available techniques (AR-XPS, WDS, STEM-EDX, and APT) to quantify phosphorus interfacial segregation within the same grain of a sample. This was done by fabricating a Fe-P-Fe sandwich specimen with phosphorus segregated at the interface. Attention was paid to the way of expressing the results of the different techniques so that they can be compared with one another. The quantification results from the different techniques show reasonable agreement.

This study describes a method to quantify phosphorus grain boundary segregation by Energy Dispersive X-ray Spectroscopy in Scanning Transmission Electron Microscope (STEM-EDX). A "box-type method" is employed, removing the long-discussed problems of interaction volume and the beam broadening effect. The proposed methodology also introduces a novel way of subtracting the spectrum background to remove the influence of coherent Bremsstrahlung and spurious peaks. A Fe-P model alloy was used to compare the box method to the quantification results previously obtained by atom probe tomography on two high angle grain boundaries. The results are specifically reported in surface concentration (atom/nm2) to avoid additional hypotheses and allow the results between the two techniques to be directly compared. The measurements show that the box-type method can accurately measure phosphorus intergranular segregation in iron.

The sensitivity to ductility dip cracking was measured in two model high chromium nickel alloys, with the same composition, apart from different sulphur contents. The newly developed Refusion Cracking Test was used, that consists in repetitive refusion lines conducted at the specimen surface. Grain boundary cracks develop in the heat affected zone near the refusion line. Cracking is much more pronounced in the sulphur-enriched alloy, which demonstrates a strong deleterious effect of sulphur. However postmortem analyses using WDS and STEM-EDS revealed no segregation of sulphur at grain boundaries. In contrast, grain boundary fracture surfaces are covered with sulphur. This suggests a dynamic type of grain boundary embrittlement where sulphur acts as a surfactant, facilitating crack opening. Sulphur is efficiently provided to the crack tip as it propagates, due to accelerated diffusion by plastic deformation. This allows crack growth rates higher than 10 mu m/s. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

The macro-segregation of carbon and alloying elements in heavy forged nuclear components can lead to significant variations in mechanical properties, and especially in fracture toughness. In the current work, the prediction of fracture toughness was addressed for three model alloys chemically representative of the compositions encountered in macro-segregated zones. Characterisations of the microstructure and of the fracture toughness properties, measured by compact tension (CT) specimens, were performed. Tensile properties were measured around the reference temperature for each material for the calibration of elasto-plastic parameters. Carbide size distributions were determined thanks to scanning electron microscopy (SEM) image analysis. The microstructure-informed brittle fracture (MIBF) local approach model was applied here to predict the scatter of brittle fracture toughness. This model involves two sources of scatter: the stress distribution inside a representative volume of the bainitic microstructure and the size distribution of carbides, which are assumed to be the brittle facture initiators. The simulation results demonstrated the capabilities of the MIBF model for predicting the fracture toughness scatter and the shift towards higher temperature of the brittle-to-ductile transition with the increase of carbon and alloying elements. The only parameter of the MIBF model to be calibrated is the effective surface energy γf which was found very close to estimated values for ferrite. The observed variation of γf in these three model alloys also suggests a possible effect of microstructural evolution on fracture toughness, in addition to the constitutive law and carbide distributions.

This paper deals with Ductility Dip Cracking (DDC) during multipass welding of 690 filler metals. In addition to industrial alloys (152 and 52 M), model alloys of controlled purity were also used to unambiguously demonstrate the effect of elemental sulphur and carbon. The sensitivity of each alloy to DDC was measured with the recently developed Refusion Cracking Test (RCT). It is confirmed that the cracks observed in this study result from DDC, not from liquation nor solidification cracking. DDC cracks are in most cases observed after several welding passes. The detrimental effect of sulphur and beneficial effect of carbon on the resistance to DDC are clearly evidenced. The good resistance to DDC of the industrial alloys, compared to the model alloys, is related to their niobium content.

The solute segregation behavior at two types of carbide–ferrite interfaces was investigated by Atom Probe Tomography (APT) technique in a 16MND5 weld before and after ion irradiation. The two types of carbides are cementite and M2.0-3.2C carbides. Effects of these intergranular carbides on solute segregation behavior at the vicinal Grain Boundaries (GBs) were discussed. No influence was observed within the first tens of nm away from carbides. However, these carbides introduce effective sites for solute segregation (P, Mn, Si and Ni). The segregation of one or several chemical species was observed at interfaces. It is also revealed that the intersection region between carbide and the adjacent GB favors higher P segregation than any other region. After ion irradiation, there is a significant increase of P segregation level at both cementite-ferrite interfaces and M2.0-3.2C carbide–ferrite interfaces, which could be attributed to a Radiation Enhanced Segregation (RES) of P. The increased P segregation at interfaces could contribute to the radiation embrittlement. Under both conditions, there is a higher average P segregation level at cementite-ferrite interfaces than M2.0-3.2C carbide–ferrite interfaces.

Thermodynamic modeling of G-phase in iron-based multicomponent systems was revised and updated by means of a careful literature survey. The thermodynamic parameters were optimized based on DFT data. Additionally intermetallic phases with possible competition to G-phase in multicomponent systems were also optimized. The thermodynamic model parameters were added to multicomponent thermodynamic database and used for calculation of technological steels and possible competition between phases are tested and shown.

In the present study, a correlative Electron Backscattering Diffraction (EBSD)/Transmission Kikuchi Diffraction (TKD)/Atom Probe Tomography (APT) methodology was used to investigate the solute segregation behavior in a low-alloyed weld. Three dimensional chemical information was correlated to the crystallographic features of Grain Boundaries (GBs). The steel has a complex microstructure of acicular ferrite and intergranular carbides. About 78% High Angle Grain Boundaries (HAGBs, with about 12% Σ3 HAGBs) and about 22% Low Angle Grain Boundaries (LAGBs) were observed. APT analyses revealed the segregation of one element or several chemical species (C, P, Mn, Si, Ni, Cr and Mo) at GBs or carbide-ferrite interfaces. Taking into account the nature of segregants and the five-parameter GB crystallography, both interstitial and substitutional segregation behaviors were discussed. The results reveal a significant influence of misorientation angle (or deviation angle) on interstitial segregation for LAGBs (or Σ3 special HAGBs) and also a strong influence of the GB plane on substitutional segregation for Σ3 special HAGBs and general HAGBs.

Dissimilar welds close to the fusion boundary exhibit a variety of solidification microstructures that strongly impact their service behavior. Investigations were therefore undertaken to clarify the origins of the morphological and microstructural evolutions encountered in a 18MND5/309L dissimilar joint produced by submerged arc welding, using a combination of microstructural characterizations, thermodynamic computations, and solidification modelling. An unexpected evolution was observed in the solidification mode, from primary austenite towards primary ferrite with increasing growth rate. Solidification of austenite at the fusion boundary was assigned to its epitaxial growth on the metastable austenitic structure of the base metal resulting from an incipient melting mechanism. The evolution of the solidification mode toward primary ferrite was explained based on computations of the solute built up between austenite cells followed using the so-called “interface response function model”. Analyzing macro- and microstructural characteristic lengths with the published solidification model and data enabled evaluation of local values of the solidification rate, thermal gradient, and cooling rate close to the fusion boundary, thus providing useful data for numerical modelling of the submerged arc-welding process.

This paper describes the methodology of the French surveillance program as well as some aspects of the experience gained during its progress. Its main uses in the structural integrity assessment of the vessel consist of updating and validating the conservatism of the prediction of the lower-bound toughness curve used to assess the margins against brittle fracture. Its perspectives in the frame of the planned extended operation beyond 40 years are associated with the flux reductions that will be implemented in the 900-MWe fleet. Two surveillance capsules are dedicated to a specific monitoring in the 40 to 60-year range to complement the data set in the relevant fluence range.

Inconel® alloy 690 is nowadays commonly used instead of 600 for the manufacturing of certain components of the primary circuit of pressurized water reactor (PWR) nuclear power plants, due to its superior resistance to corrosion and stress corrosion cracking. However 690 alloy, and the corresponding welding filler metals (types 52 and 152), can be sensitive to a solid state hot cracking phenomenon during welding, called “ductility dip cracking” (DDC) associated to grain boundary cracking. This work is undertaken to determine more precisely the thermomechanical conditions of the occurrence of DDC in two types of materials: filler metals 52M and 152. To do this, we designed a simple hot crack susceptibility test. This test is based on multiple welding beads on a cuboidal mockup. This test clearly demonstrates the effect of multiple passes on the occurrence of DDC. In parallel, hot tensile tests following fast heating were performed to determine the DDC temperature range, to try and correlate DDC to the thermomechanical behavior.

The complex microstructures developed during post-welding heat-treatment in the vicinity of the fusion line between a ferritic and austenitic steel were examined in the case of submerged arc welded 18MND5/309L dissimilar joints. Quantitative measurements of the carbon distribution in the as-welded and post-weld heat-treated conditions were performed by both wavelength dispersive spectrometry and secondary ion mass spectrometry. The extent of carbon diffusion was confirmed by hardness profiles performed by nanoindentation. On the low-alloy ferritic side, decarburization resulted in cementite dissolution allowing the evolution of the bainitic structure toward a large-grained ferritic region. In the weld metal, the carbon content reached unusually high levels and an intense precipitation of chromium-rich carbides was observed in both the interfacial martensitic layer and the austenitic weld metal. The evolution of the precipitation as a function of the distance from the interface was analyzed in terms of crystallography, chemistry, volume fractions, and size distributions. Automated crystal orientation mapping in a transmission electron microscope allowed identification of the precipitates extracted on carbon replicas from both the martensitic and austenitic matrices. A 3D reconstruction of the carbides population in the martensitic layer was performed by serial cutting with a focused ion beam: M7C3 and M23C6 were found to coexist in the two carburized regions, but displayed different sizes, compositions, and morphologies, depending on their location with respect to the fusion line. This evolution in terms of precipitation was analyzed taking into account the local microstructure and composition.

In dissimilar welds between low-alloy steel and stainless steel, the post-weld heat-treatment results in a high variety of microstructures coexisting around the fusion line, due to carbon diffusion and carbides dissolution/precipitation. The local constitutive laws in the vicinity of the fusion zone were identified by micro tensile specimens for the sub-millimeter sized zones, equivalent bulk materials representing the decarburized layer using both wet H2 atmosphere and diffusion couple, and nano-indentation for the carburized regions (i.e. the martensitic band and the austenitic region). The decarburized zone presents only 50% of the yield strength of the low-alloy steel heat affected zone and a ductility doubled. The carburized zones have a yield strength 3–5 times higher than that of the low-alloy steel heat affected zone and have almost no strain hardening capacity. These properties result in heterogeneous plastic deformation happening over only millimeters when the weld is loaded perpendicularly to the weld line, affecting its overall behavior. The constitutive laws experimentally identified were introduced as inputs into a finite elements model of the transverse tensile test performed on the whole dissimilar weld. A good agreement between experiments and simulations was achieved on the global stress-strain curve. The model also well predicts the local strain field measured by microscale DIC. A large out-of-plane deformation due to the hard carburized regions has also been identified.

The EPR (initially European Pressurized water Reactor then Evolutionary Power Reactor and considered today as only a trademark) reactor pressure vessel (RPV) has been designed taking into account the requirement of an end-of-life (EOL) RTNDT lower than or equal to 30 degrees C after 60 years of operation. The maximum acceptable fluence at the RPV inner wall, calculated with RCC-M formula and emerging from that requirement, is 3.3 x 10(19) n/cm(2) (E > 1 MeV). To not exceed this fluence level, a heavy reflector has been introduced together with an increased water gap in the reactor downcomer. According to neutron calculations, the expected actual EOL fluence on EPR RPV in the presence of the heavy reflector is between 1 and 2.25 x 10(19) n/cm(2) (E > 1 MeV) depending on the fuel management route. The irradiation surveillance program consists of positioning representative specimens of core region materials in the capsules attached to the outer face of the RPV core barrel. In former reactors equipped with a conventional core baffling, the neutron energy spectrum affecting the surveillance capsules is quite similar to the one affecting the RPV inner wall. In EPR, the heavy reflector presence distorts the neutron spectrum at the capsules' location and, consequently, the EPR surveillance specimens used to monitor progress of the embrittlement in service would be subjected to a significantly different neutron energy spectrum from that on the RPV inner wall. The aim of this paper is to describe the approach followed by AREVA to limit this neutron spectrum effect and to design an appropriate capsule withdrawal schedule. This approach includes: (1) mitigation, with a new capsule basket design introducing an additional water gap between the reflector and the specimens, and (2) assessment of the embrittlement, not only with the conventional fast neutron fluence (E > 1 MeV), but also with the dose per atom as an additional dose damage parameter.

The goal of this study is to understand the complex microstructures which form at the interface between a low-alloy steel and a stainless steel during welding and subsequent heat-treatment at 610 ∘C. A dual approach with both experimental measurements and thermodynamic and kinetic modeling has been chosen for both solidification and solid-state phase transformations. A very good agreement has been obtained for the carbon profile across the interface and the nature of phases which form.

The Reactor Pressure Vessel (RPV) is with the concrete containment one of the two components of a NPP whose replacement is not considered as reasonably feasible. The RPV lifetime has thus an important impact on the lifetime of the whole NPP.One of the key issues concerning RPV lifetime is the radiation effect on the RPV steel in the core zones. The vessel steel becomes indeed more brittle in the RPV core region where radiation is high.Margins have been included at design and manufacturing stages taking into account the material's embrittlement. Moreover, operating measures have been taken to manage ageing of RPV in order to extend lifetime. The challenge is to preserve high margins and to provide the safety studies showing these margins.A large R&D program has been developed to support lifetime extension. The objective of the program is to develop tools and provide input data for the demonstration of the safe operation of the reactor pressure vessel significantly over a 40-year lifetime.The aim of the paper is to present an overview of the R&D program to support lifetime management on the fields of materials, mechanics and thermalhydraulics. Experiments are indeed performed on irradiated material in order to improve the knowledge on embrittlement for high fluences and to be able to determine embrittlement correlations for materials representative of French RPV. Actions are also planned to improve evaluation of the RPV mechanical behaviour and to describe physical phenomena such as crack arrest or warm pre-stressing effect. Last, studies are realized to improve the thermal loadings evaluations under hypothetical accidental scenarios. These studies are supported by thermalhydraulic numerical simulations whose validation is obtained by comparison to experimental results from experimental hydraulic loops representative of French RPV.