Irradiated Pressure Vessel Research Articles

Ab-initio calculations of substitutional co-segregation interactions at coherent bcc Fe-Cu interfaces

Cu rich precipitates (CRPs) are commonly observed to form core-shell structures in neutron irradiated pressure vessel steels. The core region is typically composed of pure Cu whilst the shell is formed of other solute elements including Ni and Si. In this work we calculate the segregation energies for Ni and Si substitutional defects segregating to coherent Fe-Cu interfaces decorated with different types of point defect (vacancies and other solute substitutional defects). By comparing these values to those found for Ni and Si segregation to clean Fe-Cu interfaces we can establish how the presence of point defects on the interface influences solute segregation behaviours. Interestingly we find that the segregation of Ni to Si decorated interfaces and the segregation of Si to Ni decorated interfaces demonstrate a relatively strong co-segregation interaction. We additionally observe that both solute species experience more attractive segregation to vacancy decorated Fe-Cu interfaces. These findings suggest that mixed solute interactions and the presence of vacancies may both play an important role in assisting the formation of large solute shell regions in CRPs. We further observe that the presence of Ni and vacancies on coherent Fe-Cu interfaces both result in substantial reductions in interfacial energy density. These findings support experimental predictions and indicate both features may significantly contribute to CRP formation.

Assessment of the Generalization Ability of the ASTM E900-15 Embrittlement Trend Curve by Means of Monte Carlo Cross-Validation

The standard ASTM E900-15 provides an analytical expression to determine the transition temperature shift exhibited by Charpy V-notch data at 41-J for irradiated pressure vessel materials as a function of the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. The 26 free parameters included in this embrittlement correlation were fitted through maximum likelihood estimation using the PLOTTER—BASELINE database, which contains 1878 observations from commercial power reactors. The complexity of this model, derived from its high number of free parameters, invites a consideration of the possible existence of overfitting. The undeniable goal of a good predictive model is to generalize well from the training data that was used to fit its free parameters to new data from the problem domain. Overfitting takes place when a model, due to its high complexity, is able to learn not only the signal but also the noise in the training data to the extent that it negatively impacts the performance of the model on new data. This paper proposes the resampling method of Monte Carlo cross-validation to estimate the putative overfitting level of the ASTM E900-15 predictive model. This methodology is general and can be employed with any predictive model. After 5000 iterations of Monte Carlo cross-validation, large training and test datasets (7,035,000 and 2,355,000 instances, respectively) were obtained and compared to measure the amount of overfitting. A slightly lower prediction capacity was observed in the test set, both in terms of R2 (0.871 vs. 0.877 in the train set) and the RMSE (13.53 °C vs. 13.22 °C in the train set). Besides, strong statistically significant differences, which contrast with the subtle differences observed in R2 and RMSE, were obtained both between the means and the variances of the training and test sets. This result, which may seem paradoxical, can be properly interpreted from a correct understanding of the meaning of the p-value in practical terms. In conclusion, the ASTM E900-15 embrittlement trend curve possess good generalization ability and experiences a limited amount of overfitting.

A study of the pressure vessel steel of the WWER-440 unit 1 of the Kozloduy nuclear power plant

A comparison between highly neutron irradiated samples from the region of weld № 4 and low irradiated samples from weld № 1 taken from the pressure vessel of the WWER-440 Unit № 1 of the Kozloduy NPP has been performed. Measurements of the residual activity of samples from the outer surface of the reactor pressure vessel bottom corpus reveal very low activity of 60Co. Insofar as there the base and weld metal appear to be exposed to a very low neutron fluence, the samples from these locations can be considered as practically not affected and may serve as a reference basis for comparison with highly irradiated pressure vessel regions. The Mossbauer parameters isomer shift (IS) and quadrupole splitting (QS) were found to be absolutely irradiation insensitive. A stepwise reduction of the internal hyperfine magnetic field Bhf, each by about 2.6 T, was observed. This can be attributed to the replacement of one or two surrounding iron atoms as first nearest neighbors by non-iron alloying atoms. The Mossbauer experimental line widths for irradiated and non-irradiated samples are practically the same, which is a quite unexpected result. The area fraction ratio for the three main Zeeman sextet subspectra S1:S2:S3 shows very high irradiation sensitivity. For the bottom low irradiated region of the reactor vessel the values are S1:S2:S3 = 50.1:40.0:9.4. After seven years of operation between the pressure vessel annealing in 1989 and the autumn of 1996 when the samples from weld № 4 were taken the ratio changes strongly to S1:S2:S3 = 56.4:34.7:8.5. A possible explanation of this result is that neutron irradiation gives rise to a precipitation process involving predominantly alloying atoms as Ni, Mn, Cr, Mo and V which become mobile and precipitate in the form of carbides and/or P-rich phases and alloying atom aggregates. This “refinement” process lowers the partial area of subspectra S2 and S3 where alloying atoms are involved and leads to a higher area fraction of the pure iron component S1, which is the major experimental result. For a more complete Mossbauer investigation on the processes of generation of structure defects caused by the neutron fluence, a new series of measurements will be performed by using a set of so-called surveillance specimens with different irradiation histories which are available only for the WWER-1000 reactors of the Kozloduy NPP.

On flux effects in a low alloy steel from a Swedish reactor pressure vessel

This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

Irradiation hardening of reactor pressure vessel steels due to the dislocation loop evolution

The irradiation hardening of reactor pressure vessel steels due to the formation of dislocation loops is analyzed. The analysis is based on the original model for the nucleation and subsequent evolution of dislocation loops in irradiated materials. The loop formation in displacement cascades is taken into account, along with the homogeneous clustering of point defects. The loop evolution is shown to contribute mainly to the athermal component of the yield stress, which is determined by interaction of gliding dislocations with strong barriers. Irradiation-induced hardening is evaluated as a function of irradiation dose and temperature, dose rate, material parameters and initial microstructure. The model results are compared with experimental data for neutron irradiated pressure vessel steels of various grades and with empirical low power expressions of the yield stress increase with increasing irradiation dose.

Microstructural evolution under high flux irradiation of dilute Fe–CuNiMnSi alloys studied by an atomic kinetic Monte Carlo model accounting for both vacancies and self interstitials

Under neutron irradiation, a large amount of point defects (vacancies and interstitials) are created. In the irradiated pressure vessel steels, weakly alloyed, these point defects are responsible for the diffusion of the solute atoms, leading to the formation of solute rich precipitates within the matrix. Ab initio calculations based on the density functional theory have been performed to determine the interactions of point defects with solute atoms in dilute FeX alloys (X = Cu, Mn, Ni or Si). For Mn, the results of these calculations lead to think that solute transport in α-Fe can very likely take place through an interstitial mechanism as well as via vacancies while the other solutes (Cu, Ni and Si) which establish strong bonds with vacancies diffuse more likely via vacancies only. The database thus created has been used to parameterize an atomic kinetic Monte Carlo model taking into account both vacancies and interstitials. Some results of irradiation damage in dilute Fe–CuNiMnSi alloys obtained with this model will be presented.

Small-angle neutron scattering study of post-irradiation annealed neutron irradiated pressure vessel steels

Neutron irradiated reactor pressure vessel steels of two different Cu contents exposed to isochronal post-irradiation annealing treatments at stepwise increasing levels of temperature have been investigated. Results of small-angle neutron scattering experiments are reported and compared with the respective unirradiated and as-irradiated conditions. Volume distributions of scatterers and ratios of total-to-nuclear scattering intensities ( A-ratio) were calculated from the measured differential macroscopic scattering cross-sections. Finally, total volume fractions of scatterers, peak radii of the volume distribution and partial information on the composition of scatterers were extracted and discussed as a function of Cu level and annealing temperature. Our measurements indicate that irradiation-induced clusters dissolve completely for the lower Cu level of 0.15 wt% but dissolve only partly prior to the onset of coarsening for the higher Cu level of 0.29 wt%. An Arrhenius type behaviour and a remarkable reversibility of the irradiation-induced cluster distribution are observed in the complete dissolution regime.

The Technical Foundations of a Unified Adjusted Reference Temperature for RPV Fracture Toughness

Abstract The use of a reference temperature to define the transition from ductile-to-brittle behavior is a well-established practice in reactor pressure vessel integrity analysis. Current procedures for analysis of irradiated pressure vessel steels use an Adjusted Reference Temperature (ART or RTPTS) that shift the unirradiated RTDNT value based on Charpy data and apply additional margins. The recent adoption of an ASTM Standard Test Method for Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range (E 1921) has made direct determinations of fracture toughness reference temperature in irradiated materials using Master Curve feasible. ASME Code Case N629 recognizes the potential of this new technology as an alternative means of determining an ART value for use in reactor pressure vessel integrity analysis. This alternative approach has been used in recent plant specific submittals to the NRC. ASTM sub-committee E10.02 is currently considering adopting a standard to provide a unified definition of the Adjusted Reference Temperature. This unified definition would provide consistent procedures for using either RTNDT or Master Curve based definitions of ART. This unified definition would provide common margins for the two approaches and assure a consistent level of confidence in the analysis. The statistical basis of the Master Curve makes it possible to establish a more rational basis for the margins applied in both approaches.

Dependence of Barkhausen noise in the neutron irradiated reactor pressure vessel steel

Barkhausen noise (BN) and ferromagnetic resonance (FMR) experiments were conducted in the neutron irradiated commercial nuclear pressure vessel steel. The energy of BN decreased by neutron irradiation and the rates increased with increasing neutron dose. The resonance field and linewidth determined from FMR spectra increased with increasing neutron dose. The experimental results were explained as domain wall pinning of irradiation induced precipitates and change of anisotropy energy associated with neutron irradiation.

03/02626 Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel: Gómez, F. J. et al. Nuclear Engineering and Design, 2003, 219, (2), 111–125

03/02626 Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel: Gómez, F. J. et al. Nuclear Engineering and Design, 2003, 219, (2), 111–125

Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel

The cohesive fracture process zone model was used to account for the neutron irradiation embrittlement of a pressure vessel steel. The tensile testing and fracture of axisymmetrically notched round specimens were numerically modelled assuming a rectangular traction separation law and the irradiation effects were introduced by due modification of this law. The results corroborate those of the experiments performed in a previous work. The cohesive strength and the cohesive energy of the cohesive model were not considered as adjusting parameters, but they were determined from the data of conventional tensile tests and fracture toughness tests on the assumption that the failure of the specimens in these tests also follows the cohesive model.

Atomic Level Characterization of Neutron Irradiated Pressure Vessel Steels

ABSTRACTAtom probe tomography provides one of the most effective tools to characterize the solute distribution and precipitation that occurs in pressure vessel steels and associated model alloys during irradiation. The three-dimensional atom probe is able to experimentally determine the elemental identities of the atoms and their spatial coordinates with near atomic resolution so that their distribution within small volumes of the specimen can be reconstructed and analyzed. This technique together with conventional atom probe field ion microscopy has been applied to many different types of pressure vessel steels and model alloys and has revealed and characterized several different nanostructural transformations. These radiation induced or enhanced processes lead to the formation of copper-nickel-manganese-silicon-enriched precipitates, and solute segregation to dislocations, dislocation loops, nanovoids and boundaries.

A computational microscopy study of nanostructural evolution in irradiated pressure vessel steels

Nanostructural features that form in reactor pressure vessel steels under neutron irradiation at around 300°C lead to significant hardening and embrittlement. Continuum thermodynamic-kinetic based rate theories have been very successful in modeling the general characteristics of the copper and manganese nickel rich precipitate evolution, often the dominant source of embrittlement. However, a more detailed atomic scale understanding of these features is needed to interpret experimental measurements and better underpin predictive embrittlement models. Further, other embrittling features, believed to be subnanometer defect (vacancy)-solute complexes and small regions of modest enrichment of solutes are not well understood. A general approach to modeling embrittlement nanostructures, based on the concept of a computational microscope, is described. The objective of the computational microscope is to self-consistently integrate atomic scale simulations with other sources of information, including a wide range of experiments. In this work, lattice Monte Carlo (LMC) simulations are used to resolve the chemically and structurally complex nature of CuMnNiSi precipitates. The LMC simulations unify various nanoscale analytical characterization methods and basic thermodynamics. The LMC simulations also reveal that significant coupled vacancy and solute clustering takes place during cascade aging. The cascade clustering produces the metastable vacancy-cluster solute complexes that mediate flux effects. Cascade solute clustering may also play a role in the formation of dilute atmospheres of solute enrichment and enhance the nucleation of manganese-nickel rich precipitates at low Cu levels. Further, the simulations suggest that complex, highly correlated processes (e.g. cluster diffusion, formation of favored vacancy diffusion paths and solute scavenging vacancy cluster complexes) may lead to anomalous fast thermal aging kinetics at temperatures below about 450°C. The potential technical significance of these phenomena is described.

A lattice Monte Carlo simulation of nanophase compositions and structures in irradiated pressure vessel Fe-Cu-Ni-Mn-Si steels

A self-consistent analysis, combining thermodynamic calculations, recent small angle neutron scattering and atom probe field ion microscopy measurements and lattice Monte Carlo (LMC) simulations, is used to characterize the ultra fine coherent precipitates that form in irradiated reactor pressure vessel steels. These nanofeatures are typically rich in impurity copper precipitated from highly supersaturated solution at vessel operating temperatures around 300 °C at rates greatly accelerated by radiation enhanced diffusion. Although some copper appears to be needed to catalyze their formation, manganese nickel-rich precipitates may replace copper-rich precipitates at high concentrations of these alloying elements and/or lower temperatures. Thermodynamic and kinetic models are generally consistent with observations on the number densities, sizes, compositions and nucleation, growth and coarsening behavior of the nanofeatures. However, the LMC atomic scale simulations of complex nanofeature structures are needed to fully unify experiment and theory.

APFIM studies of the phase transformations in thermally aged ferritic FeCuNi alloys: comparison with aging under neutron irradiation

An atom probe field ion microscopy characterization has been performed on a ternary Fe1.28wt%Cu1.43wt%Ni model alloy. Two sets of heat treatments at 400 and 500°C were selected to follow the decomposition of the solid solution. The atom probe results suggest that the precipitation of copper from the body-centered cubic matrix involves classical nucleation, growth and coarsening processes. Nickel does not appear to be involved in these processes. The nickel and iron detected in the small particles in the early stage is rejected from the particles during subsequent growth and coarsening. The comparison of these data with those observed in neutron-irradiated pressure vessel steels and in Fe0.28wt%Cu0.7wt%Ni model steel indicates that thermal aging of ternary FeCuNi model alloys is not an appropriate process to elucidate the phase transformations in neutron-irradiated pressure vessel steels, and that the high nickel level detected in the clusters in the irradiated pressure vessel steels may be due to a “radiation-induced” rather than a “radiation-enhanced” precipitation process.

On the Composition and Structure of Nanoprecipitates in Irradiated Pressure Vessel Steels

AbstractNanoscale Cu rich precipitates (CRPs) are widely believed to be the dominant hardening feature resulting in severe embrittlement in irradiated reactor pressure vessel (RPV) steels. However, this view has recently been challenged by interpretations of atom probe field ion microscopy (APFIM) measurements that describe the dominant nanofeatures as dilute solute atmospheres (DSAs). The practical impact of these differing views is very significant. This work compares and contrasts the CRP versus DSA descriptions to a wide variety of pertinent data. Mechanical property trends as well as small angle neutron scattering (SANS) and field emission scanning transmission electron microscopy (FEGSTEM) measurements support the presence of CRPs. CRPs are also consistent with the fundamental thermodynamic and kinetic laws. However, standard theory cannot provide the atomic level resolution needed to fully understand the nanofeatures. Therefore, a new Lattice Monte Carlo (LMC) atomistic method is used to simulate the complex chemical structures of the CRPs. The LMC method unifies the SANS/FEGSTEM and APFIM data within a well founded physical framework.

Small Angle Neutron Scattering Modeling of Copper-Rich Precipitates in Steel

AbstractThe magnetic-to-nuclear scattering intensity ratio observed in the scattering from copper-rich precipitates in irradiated pressure vessel steels is much smaller than the value of 11.4 expected for a pure copper precipitate in iron. A model for precipitates in pressure vessel steels which matches the observed scattering typically incorporates manganese, nickel, silicon and other elements and it is assumed that the precipitate is non-magnetic. In the present work consideration is given to the effect of composition gradients and ferromagnetic penetration into the precipitate on the small angle scattering cross section for copper-rich clusters as distinquished from conventional precipitates. The calculation is an extension of a scattering model for micelles which consist of shells of varying scattering density. A discrepancy between recent SANS scattering experiments on pressure vessel stells was found to be related to applied magnetic field strength. The assumption of cluster structure and its relation to atom probe FIM findings as well as the effects of insufficient field for magnetic saturation is discussed.

Size effects on the upper shelf energy of a neutron irradiated pressure vessel weld metal

The methodologies published earlier for predicting the upper shelf energy (USE) of full size Charpy specimens based on subsize test data appear to work satisfactorily for either highly materials (USE > 200 J) or relatively brittle materials (USE ≪ 100 J). A methodology is proposed here that works well for pressure vessel weld materials in both unirradiated and irradiated conditions having USE in the intermediate region (100 J < USE < 200 J). The methodology uses partitioning of the USE into two components, USE p and δUSE ( = USE - USE p). USE p is the absorbed energy for a specimen fatigue-precracked to half the width. The predicted value of the USE of full-size specimens is a sum of two terms. The first term is equal to product of the normalized δUSE of the subsize specimen and the full-size normalization factor for δUSE. The second term is equal to the product of the normalized USE p of the subsize specimen and the fracture volume of the full-size precracked specimen. The predicted values were within about 10% of the measured values for both unirradiated and irradiated materials.

Effect of temperature gradient on crack initiation

Compact tension (CT) specimens of pressure vessel steel A533B have been subjected to linear distributions of temperature along the crack line. The temperature dependence of fracture toughness for the A533B steel creates a fracture toughness gradient in such a specimen. The specimens simulated an irradiated pressure vessel wall and a functionally gradient material. The CT specimens subjected to linear distributions of temperature were tested to evaluate the fracture toughness for crack initiation. In all tests, the temperature at the crack tip was kept at −10°C or −55°C. If the value of the temperature gradient ahead of the crack tip exceeded a critical value, the fracture toughness deviated from the toughness obtained under a uniform temperature of −10°C or −55°C.

Positron lifetime characterization of irradiated pressure vessel model alloys

Positron lifetime characterization of irradiated pressure vessel model alloys