Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.     

Influence of Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb binary alloys

The influence of the Nb concentration in the α-matrix on the corrosion behavior of Zr-xNb (x=0-0.6 wt%) binary alloys was evaluated using a static autoclave in the temperature range from 300 to 500 °C. Corrosion tests and precipitate analysis of Zr-xNb binary alloys showed that corrosion resistance increased with the increase of the Nb concentration in the α-matrix, and the best corrosion resistance was obtained when the Nb concentration was nearly at its equilibrium solubility limit at all test temperatures. The alloys containing a higher Nb concentration than their equilibrium solubility also showed good corrosion resistance, which could be attributed mainly to the formation of Nb-precipitates, resulting in an equilibrium Nb concentration in the α-matrix. These results imply that the corrosion resistance of Nb-containing Zr-alloys can be controlled by the Nb concentration in the α-matrix rather than the Nb-precipitates.     

Effect of microstructure on the cleavage fracture strength of low carbon Mn-Ni-Mo bainitic steels

The effects of the microstructure on the cleavage fracture strength of low carbon Mn-Ni-Mo bainitic steels were examined. A four-point bend test and double-notched bend specimens were used to measure the cleavage fracture strength of the alloys and identify the cleavage initiating micro-cracks, respectively. The cleavage fracture strength and DBTT of Mn-Ni-Mo bainitic steels were strongly affected by the alloy carbon content. The decrease in the alloy carbon content resulted in a decrease in the inter-lath cementite-crowded layers and higher cleavage fracture strength. Micro-cracks that formed across the inter-lath cementite-crowded layers were observed to initiate cleavage fracture. The width of these inter-lath cementite-crowded layers was accepted as a cleavage initiating micro-crack size in the micro-mechanical modeling of the cleavage fracture, and the measured cleavage strength values of the bainitic Mn-Ni-Mo steels were well represented by the modified Griffith relationship.     

Gas-driven hydrogen permeation in the surface-limited regime

Gas-driven permeation of hydrogen through metal membranes in the surface-limited regime (SLR) is analyzed. An analytical solution for the concentration and permeation flux as a function of time is given for permeation through the asymmetric membrane having different conditions on the inlet and outlet sides. The features of the steady state and transient permeation are discussed. Comparison of calculations with an experiment on deuterium permeation through vanadium, which is available from literature, is performed. It has been demonstrated that the parameters, extracted from the measurements of the permeation rate in SLR are very uncertain. Even the calculations with opposite asymmetries can be equally well adjusted to the experiment. Non-zero initial conditions in SLR experiments are additional source of uncertainty. Measurements of the accumulation in a closed volume instead of the permeation rate can seriously mislead in the interpretation of the experiment. Complementary experiments on permeation in two opposite directions and measurements of permeation decay could increase the reliability of the data obtained.     

The effect of grain boundary misorientation on the intergranular M23C6 carbide precipitation in thermally treated Alloy 690

The precipitation characteristics of chromium carbides on various types of grain boundaries in Alloy 690 thermally treated at 720 °C for 10 h were studied through transmission electron microscopy. Precipitation of the intergranular chromium carbides, identified as Cr-rich M₂₃C₆, was retarded on the low angle grain boundaries, compared to that on the random high angle grain boundaries on which coarse and discrete ones were found. They were rarely found on the coherent twin boundaries, however, needle-like ones were evolved on the incoherent twin and twin related Σ9 boundaries. Precipitation of the chromium carbides was also suppressed on the nearly exact coincidence site lattice boundaries such as Σ11 and Σ15, for which the Brandon criterion was fulfilled. The results of the intergranular M₂₃C₆ carbide precipitation were explained in terms of the influence of the grain boundary energy.     

Characteristics of microscopic strain localization in irradiated 316 stainless steels and pure vanadium

Characteristics of localized dislocation glide were investigated for 316 and 316LN stainless steels and pure vanadium after ion or neutron irradiation near room temperature and deformation by a uniaxial tensile load or by a multiaxial bending load. In the irradiated 316 stainless steels, both the uniaxial tensile loading and the multiaxial bend loading produced straight localized bands in the form of channels and twins. In vanadium specimens, on the other hand, curved channels were observed after tensile deformation, and these became a common feature after multiaxial bend deformation. No twin was observed in vanadium. A river pattern of channels was observed in the bent samples after irradiation to a high dose of 0.69 dpa. A highly curved channel can be formed by successive cross slip of screw dislocations. Also, the channel width was not constant along the channels; channel widening occurred when weak defect clusters were removed by the gliding screw dislocations changing their paths by cross slip. It is believed that the dissociation of dislocations into partials and high angles between easy glide planes suppresses the formation of curved channels, while a multiaxial stress state, or a higher stress constraint, increases the tendency for channel bending and widening.     

Influence of the Pb-Bi hydrodynamics on the corrosion of T91 martensitic steel and pure iron

The Pb-Bi eutectic liquid alloy is considered as spallation target material in hybrid systems due to its suitable nuclear and physical properties. One of the parameters which may have a significant influence on the corrosion of steels in contact with molten lead alloys is the hydrodynamic regime. Corrosion tests have been performed in the CICLAD device at 400 and 470 °C at low oxygen concentrations and for various cylinder rotating speeds with T91 martensitic steel. The results obtained show that increasing the rotating speed leads to an increase of the corrosion rate. Moreover, the need for controlling finely the Pb-Bi physico-chemistry as well as the surface state of the samples is also shown by these tests. Finally, a comparison is made between the experimental corrosion rates and calculated values obtained by using equations expressing the mass transfer coefficient.     

Kinetic Monte Carlo simulations of radiation induced segregation and precipitation

Kinetics of radiation induced segregation and precipitation in binary alloys are studied by Monte Carlo simulations. The simulations are based on a simple atomic model of diffusion under electron irradiation, which takes into account the creation of point defects, the recombination of close vacancy-interstitial pairs and the point defect annihilation at sinks. They can reproduce the coupling between point defect fluxes towards sinks and atomic fluxes, which controls the segregation tendency. In pure metals and ideal solid solutions, the Monte Carlo results are found to be in very good agreement with classical models based on rate equations. In alloys with an unmixing tendency, we show how the interaction between the point defect distribution, the solute segregation and the precipitation driving force can generate complex microstructural evolutions, which depend on the very details of atomic-scale diffusion properties.     

Ab initio calculations of self-interstitial interaction and migration with solute atoms in bcc Fe

The embrittlement of pressure vessel steels under radiation has been long ago correlated with the presence of solute Cu. Indeed the atom probe and the small angle neutron scattering, principally, have revealed the formation of Cu clusters under neutron flux in reactor pressure vessel (RPV) steels and dilute FeCu alloys. Other solutes such as Ni, Mn and Si which are also found within the clusters, are now suspected to contribute to the embrittlement. The interactions of these solutes with radiation induced point defects need thus to be characterized properly in order to understand the elementary mechanisms behind the formation of these clusters. We have investigated by ab initio calculations based on the density functional theory the interactions of self-interstitials with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si). Different possible configurations of solute-dumbbell complexes have been studied. Their binding energies are discussed, as well as their relative stability. The migration of dumbbells with a solute atom in their vicinity was also investigated. All these results are compared to some experimental ones obtained on dilute FeX model alloys. Our results indicate that for Mn solute atoms, diffusion via an interstitial mechanism is very likely.     

Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water

A formulation for the quantitative calculation of the stress corrosion cracking (SCC) growth rate was proposed based on a fundamental-based crack tip strain rate (CTSR) equation that was derived from the time-based mathematical derivation of a continuum mechanics equation. The CTSR equation includes an uncertain parameter r₀, the characteristic distance away from a growing crack tip, at which a representative strain rate should be defined. In this research, slow strain rate tensile tests on sensitized 304L stainless steel in oxygenated high temperature water were performed. By curve fitting the experimental results to the numerically calculated crack growth rate, the parameter r₀ was determined. Then, the theoretical formulation was used to predict the SCC growth rates. The results indicate that r₀ is on the order of several micrometers, and that the application of the theoretical equation in predicting the crack growth rate provides satisfactory agreement with the available data.     

The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

The impact of the composition of structural steels on their corrosion stability in liquid Pb-Bi at 500 and 400 °C with different oxygen concentrations

In recent years, heavy liquid metals have found exercise as possible coolants and targets in the conversion of radioactive elements in accelerator driven systems (ADS). Liquid lead-bismuth eutectic alloy is one of candidates for this using tanks to its suitable nuclear and physical properties. Performed examination was aimed at research of compatibility choice materials for parts of ADS with liquid Pb-Bi eutectic alloy, influence of composition choice materials on their corrosion resistance, influence of temperature and oxygen content. We performed corrosion tests of 1000 h each on approximately 20 types of structural steels (austenitic, ferritic and martensitic) in convection loops with flowing Pb-Bi at 500 and 400 °C and using different oxygen concentrations. The impact of Fe, Cr, Ni, Mn, Si, Al and Mo content on the corrosion stability of these steels was measured without and after preliminary passivation through creating thin spinel or oxide layers on their surface.     

Influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in PWR primary water

The influence of ageing heat treatment on alloy A-286 microstructure and stress corrosion cracking behaviour in simulated Pressurized Water Reactor (PWR) primary water has been investigated. A-286 microstructure was characterized by transmission electron microscopy for ageing heat treatments at 670 °C and 720 °C for durations ranging from 5 h to 100 h. Spherical γ′ phase with mean diameters ranging from 4.6 to 9.6 nm and densities ranging from 8.5 × 10²² m⁻³ to 2 × 10²³ m⁻³ were measured. Results suggest that both the γ′ phase mean diameter and density quickly saturate with time for ageing heat treatment at 720 °C while the γ′ mean diameter increases significantly up to 100 h for ageing heat treatment at 670 °C. Grain boundary η phase precipitates were systematically observed for ageing heat treatment at 720 °C even for short ageing periods. In contrast, no grain boundary η phase precipitates were observed for ageing heat treatments at 670 °C except after 100 h. Hardening by γ′ precipitation was well described by the dispersed barrier hardening model with a γ′ barrier strength of 0.23. Stress corrosion cracking behaviour of A-286 was investigated by means of constant elongation rate tensile tests at 1.5 × 10⁻⁷ s⁻¹ in simulated PWR primary water at 320 °C and 360 °C. In all cases, initiation was transgranular while propagation was intergranular. Grain boundary η phase precipitates were found to have no significant effect on stress corrosion cracking. In contrast, yield strength and to a lesser extent temperature were found to have significant influences on A-286 susceptibility to stress corrosion cracking.     

Pitting corrosion of the laser surface melted Alloy 600

The effect of laser surface melting (LSM) on the resistance to pitting corrosion of Ni-base Alloy 600 was investigated by a potentiodynamic polarization test in 1 M NaCl solutions at pH values of 4 and 10 at temperatures of 30, 60 and 90 °C. The pitting potentials of Alloy 600 were markedly increased by the LSM process, when compared with those of the non-laser treated Alloy 600. From the microscopic examination after the corrosion test, it was found that the pitting was initiated at the junction between a TiN inclusion and the matrix, possibly at the site of a sulfide physically associated with the TiN inclusion. The homogeneous micro-structure associated with the reduction of the inclusion size during the LSM process could be attributed to the improvement of the pitting corrosion properties in the LSM Alloy 600.     

Intergranular stress corrosion cracking initiation model of austenite stainless steels used in BWRs for optimized water chemistry control

A calculation model on intergranular stress corrosion cracking (IGSCC) initiation time of materials used in boiling water reactors (BWRs) has been developed to evaluate effectiveness of water chemistry control for mitigation of the IGSCC. The model was composed of four terms which determine passive film break time: (1) a chemical term based on electrochemical corrosion potential (ECP) and impurity concentration; (2) a mechanical term based on strain rate; (3) a material term based on sensitization; and (4) an irradiation term based on acceleration of corrosion by γ-rays and neutron irradiation. The contribution of the chemical term in the passive film break was calculated based on a deterministic local corrosion model. Then, the local corrosion model was modified by adding mechanical acceleration of the film rupture to treat the IGSCC phenomenon. The model could reproduce the behavioral tendency seen in the slow strain rate tensile test on high carbon contents with sensitization heat treatment (for example, 620°C × 24 h). Under BWR operating conditions, IGSCC initiation time could be extended by a factor of 5 by lowering the electric conductivity from 1.0 to 0.06 μS/cm. If the ECP was reduced below the critical potential by a mitigation method, the IGSCC initiation time was predicted to become sufficiently long for pipings and components.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.