Reactor Pressure Vessel Embrittlement Research Articles

Magnetic Indicator for Evaluating Cu Clustering and Hardening Effect in RPV Model Alloy

The reactor pressure vessel (RPV) is a critical barrier in nuclear power plants, but its embrittlement during service poses a significant safety challenge. This study investigated the effects of Cu-enriched clusters on the mechanical and magnetic properties of Fe-0.9 wt.%Cu model alloys through thermal aging. Using Vickers hardness tests, Magnetic Barkhausen Noise (MBN) detection, and Atom Probe Tomography (APT), the study aimed to establish a quantitative correlation between MBN signals, Vickers hardness, and Cu-enriched clusters, facilitating the non-destructive testing of RPV embrittlement. Experimental results showed that the hardness and MBN parameters (RMS and Vpp values) changed significantly with aging time. The hardness increased rapidly in the early stage (under-aged), followed by a plateau and then a decreasing trend (over-aged). In contrast, MBN parameters decreased initially and then increased. APT analysis revealed that Cu-enriched clusters increase in size to 4.60 nm and coalesced during aging, with their number density peaking to 3.76 × 1023 m−3 before declining. An inverse linear correlation was found between MBN signals and the combined factor Nd2Rg (product of the number density squared and the mean radius of Cu-enriched clusters). This correlation was consistent across both under-aged and over-aged states, suggesting that MBN signals can serve as applicable indicators for the non-destructive evaluation of RPV steel embrittlement.

Probabilistic fracture mechanics analyses of a reactor pressure vessel using the irradiation embrittlement evaluation based on the Bayesian nonparametric method

In the structural integrity assessment of an embrittled reactor pressure vessel (RPV) based on probabilistic fracture mechanics (PFM), it is crucial to consider both of the mean value and the uncertainty of the embrittlement prediction. Typically, the embrittlement prediction uncertainty is given by a normal distribution, and its standard deviation is determined from the residuals of the measured and predicted values of all data used to develop the embrittlement prediction method. Therefore, the same uncertainty is assumed regardless of neutron fluence and the available data. To overcome this issue, the Japan Atomic Energy Agency recently developed the embrittlement prediction method based on machine learning and Bayesian statistics, i.e., the Bayesian nonparametric (BNP) method, and introduced it to the PFM analysis code. The BNP method can predict the probability distribution according to data scarcity and provide more reasonable uncertainty because it estimates significant uncertainties where the scatter of actual measured data is large and the number of data is small. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor is conducted using the PFM analysis of structural components in aging light water reactors, where the embrittlement probability distribution calculated using the BNP method is used. Furthermore, the effects of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs are investigated in the PFM analyses, and the results are presented.

The effects of flux on the radiation-induced embrittlement of reactor pressure vessel steels: review of current understanding and application to high fluences

It is necessary to quantify the effects of flux on reactor pressure vessel steel embrittlement under neutron irradiation, if surveillance or high-flux test reactor data is used to predict vessel embrittlement occurring at lower fluxes. This is particularly important when considering embrittlement occurring during extended (60–80 years) operation for which there is no direct experience. Dedicated investigations are time-consuming and expensive even when only small flux-fluence ranges are investigated, so collating data from multiple campaigns is necessary to provide sufficient information to cover the wide range of fluxes required for vessel assessment in the long term. This paper collates and reviews such data. The review finds that flux dependences probably differ in sign and strength in different regimes (low flux and fluence, intermediate flux at low and high fluence, high flux at low and high fluence) with the regime limits affected by composition and temperature. The current understanding of diffusion processes and microstructural development are invaluable in interpreting the trends and limits. Many contradictory data sets were found, however, and not all contradictions could be dismissed as resulting from poor quality data. Suggestions are made for investigations to clarify the uncertainties. One wide-ranging model of flux effects, based on an extensive data set, is used to compare high-fluence data from different sources, to assess whether embrittlement rates accelerate after a high, threshold fluence. The model helps to identify experiments which investigated comparable flux-fluence-temperature regimes. The comparable data are split evenly between data sets supporting acceleration after a particular fluence and data sets contradicting it. The model identifies regimes in which further campaigns would clarify the causes of these contrasting observations.

Predictions and Uncertainty Estimates of Reactor Pressure Vessel Steel Embrittlement Using Machine Learning

An essential aspect of extending safe operation of the world’s active nuclear reactors is understanding and predicting the embrittlement that occurs in the steels that make up the Reactor pressure vessel (RPV). In this work we integrate state of the art machine learning methods using ensembles of neural networks with unprecedented data collection and integration to develop a new model for RPV steel embrittlement. The new model has multiple improvements over previous machine learning and hand-tuned efforts, including greater accuracy (e.g., at high-fluence relevant for extending the life of present reactors), wider domain of applicability (e.g., including a wide-range of compositions), uncertainty quantification, and online accessibility for easy use by the community. These improvements provide a model with significant new capabilities, including the ability to easily and accurately explore compositions, flux, and fluence effects on RPV steel embrittlement for the first time. Furthermore, our detailed comparisons show our approach improves on the leading American Society for Testing and Materials (ASTM) E900-15 standard model for RPV embrittlement on every metric we assessed, demonstrating the efficacy of machine learning approaches for this type of highly demanding materials property prediction.

Dose and dose rate dependence of precipitation in a series of surveillance RPV steels under ion and neutron irradiation

Life extension of light water reactors requires robust models to predict the embrittlement of reactor pressure vessel (RPV) steels under long-term neutron irradiation. Embrittlement is due to nanoscale precipitation hardening by phases containing Cu, Mn, Ni, and Si, resulting in increases of the ductile to brittle transition temperature. Embrittlement data for low flux, high fluence extended life conditions, up to 80 years or more, are largely not available. Thus, higher flux, accelerated test reactor irradiations have been used to help develop models to predict low flux embrittlement at high fluence. Here flux, or dose rate, is expressed in units of displacements per atom per second (dpa/s). The dose rate can significantly influence precipitate evolution in a way that depends on fluence (dpa), temperature, and alloy composition, as well as flux itself. Here, we explore precipitation in surveillance steels with varying compositions at high fluxes (10−5–10−6 dpa/s) produced using 70 MeV Fe2+ ion irradiations. Atom probe tomography was used to characterize the dose dependence of precipitation at very high ion irradiation dose rates, and to evaluate corresponding similarities or differences compared to neutron irradiated steels at a much lower dose rate of ≈ 5 × 10−9 dpa/s. The data show that, on average, the ion irradiation precipitate volume fractions increase by factors of ∼ 1.5 to 2 between 0.3 dpa and 1.2 dpa. Limited data show that the major effect of neutron versus ion irradiation is the former has lower precipitate number densities and larger sizes. Otherwise, the precipitates are remarkably similar in composition and volume fraction.

Microstructural characterisation of brittle fracture initiation sites in reactor pressure vessel steels

Neutron embrittlement of reactor pressure vessel (RPV) steels potentially limits the safe operational time of nuclear reactors. In tested RPV steel samples microcracks are known to originate from cleaved precipitates, inclusions or grain boundaries, which eventually lead to brittle fracture. In most cases, these precipitates are reported as carbides. Latest research indicates the critical fracture event to be the crossing of grain boundary instead of the particle–matrix interface by the microcrack. Very few works have performed detailed fractography, which revealed the size and type of the feature found in the crack nucleation centre, together with metallographic bulk investigations. Additionally, the effect of subsequent neutron irradiation on microcrack nucleation is unknown. In the present work, detailed fractography and metallography is performed on a RPV steel in both its unirradiated as well as neutron irradiated state. A particle in the centre of the initiation site has been identified in a majority of the cases, while no specific type of particles were involved in more than 1/3 of all samples. Those particles are either Mo-rich carbides or Al-rich inclusions. In comparison, the metallographic analysis of the bulk steel reveals Mo-rich and Mn-rich carbides. Mn-rich carbides constitute the majority of all carbides in the steel but are never found to be responsible for brittle fracture initiation. Only a small fraction of observable carbides is Mo-rich which are responsible for fracture. Thus, the present work demonstrates a discrepancy between particles found in brittle fracture initiation sites and particles that are directly observable in bulk steel. Some of the initiating particles are extremely rare in the bulk. Furthermore, it is revealed that the local fracture stress does not depend on the type or size of particle or grain involved at the initiation site. In conclusion, the fracture stress strongly depends on the type of the grain boundary, rather than on a stress based upon a Griffith criterion, which considers particle or grain size alone. The brittle fracture mechanism was found to be unaffected by neutron irradiation. These findings could help refine modelling of the critical brittle fracture stress and fracture toughness using microstructural parameters as input.

Characterization of Fast Neutron Transmission Through an Iron Shield

In this paper we give an analysis of the neutron transmission through an iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for this investigation comes from a well-known deficiency in the iron inelastic data from the older library evaluation (ENDF/B-V), giving a concern for a fast neutron flux underestimation within the reactor pressure vessels. In order to benchmark the next-generation ENDF/B-VI iron data, the U.S. Nuclear Regulatory Commission and the former Czechoslovakian National Research Institute have jointly preformed several experiments in 1990s, addressing neutron leakage spectra obtained for a 252Cf fission source in a centre of an iron sphere. It was shown that the ENDF/B-VI iron cross section, containing several improvements over previous evaluations, will not entirely resolve the neutron spectrum discrepancies observed at high neutron energies. Since safety analyses of reactor pressure vessel embrittlement are often based on neutron transport calculations using specific multigroup cross section libraries, simulation of this benchmark was performed using a hybrid shielding methodology of ADVANTG3.0.3 and MCNP6.1.1b codes. Comparison of calculated and referenced dosimeter activation rates are presented for several "standard" nuclear reactions, often used in reactor pressure vessel dosimetry. For that purpose, the new IRDFF-II special library from the IAEA Nuclear Data Services was used as a reference source of dosimetry cross sections. The MCNP6.1.1b code was used for calculation of reaction rates, which were also compared with previous IRDFF-1.05 special library and general purpose ENDF/B-VII.1 library.

Investigation of Cu-enriched precipitates in thermally aged Fe-Cu and Fe-Cu-Ni model alloys

To better understand the mechanisms of thermally aged and embrittled reactor pressure vessel steels, the hardening behavior and microstructure evolution of different Fe-Cu and Fe-Cu-Ni model alloys were investigated using transmission electron microscopy (TEM) and small-angle neutron scattering (SANS) analysis methods. The effects of different aging conditions and alloy composition on the formation and growth of Cu-enriched precipitates were discussed. The results showed that the hardness of the Fe-Cu model alloys with low copper content continuously increased with aging, increasing to peak hardness and then decreasing for the Fe-Cu model alloys with high copper content. The aging time to reach peak hardness was greatly reduced due to the addition of Ni in the Fe-Cu model alloys. The TEM results showed that with increased aging temperature and copper content, the Cu-enriched precipitates with a B2 structure precipitated in the ferritic matrix following a direct relationship. Therefore, Ni addition led to relatively higher precipitation nucleation and growth, and the size distribution of the Cu-enriched precipitates derived from the SANS was in overall good agreement with the TEM observations for 6–25 nm regions. Therefore, small-angle scattering could provide better details when analyzing smaller, nano-sized precipitates.

The characterization of a nanoscale MnNi cluster in thermally aged reactor pressure vessel steels

The MnNi-rich clusters or precipitates (MNPs) are one of the most important nanofeatures resulting in irradiation-induced embrittlement of reactor pressure vessel (RPV) steels. In this study, the microstructure and mechanical properties of the thermally-aged RPV steel were investigated by microhardness and tensile test, x-ray diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM), and atom probe tomography (APT), revealing the presence of a novel nanoscale B2 MnNi phase. The formation of these B2 MnNi phases was dominated by the thermodynamic of the RPV-steel system, and high-density dislocations could boost their formation and growth. Meanwhile, for samples aged for 2 h at 723 K, these ordered phases result in a yield-strengthening increment of 190 MPa. Although aged-induced ordered phases have a clear difference in terms of the lattice structure, nucleation rate, and thermal stability compared to irradiation-induced MNPs, the discovery of these B2 ordered phases helps us to establish a more comprehensive understanding of the nature and formation mechanism of MNPs produced by neutron irradiation.

Simulation of crack propagation in a thick-walled cylinder using XFEM

The integrity assessment to evaluate the safety margin of reactor pressure vessels (RPV) often considers only the crack initiation and excludes the crack propagation analysis. This contribution focuses on crack propagation analysis in RPV steels. The eXtended Finite Element (XFEM) method implemented in ABAQUS is applied to a thick-walled cylindrical specimen with a circumferential crack at the inner surface in order to simulate crack propagation in embrittled RPV subjected to Pressurized Thermal Shock (PTS). The thick-walled cylinder considered in this study was tested in the FALSIRE project, which results reported the crack opening displacement (COD) and the crack-arrest cycles occurring during the cooling down process. In order to simulate the cylinder with the XFEM, a reduced three-dimensional finite element (FE) model of a small sector (a slice of the cylinder) is used by applying cyclic symmetry boundary conditions. The COD evolution during the PTS transient is calculated and compared against the experimental COD. In the experiment, the COD shows several initiation-arrest-re-initiation cycles and final arrest. However, the results from the simulation show a smooth continuous increase of the COD indicating a progressive crack growth.

Segregation of Ni and Si to coherent bcc Fe-Cu interfaces from density functional theory

The formation of Cu rich precipitates (CRPs) is known to substantially contribute to the embrittlement of reactor pressure vessel steels with high Cu contents. CRPs are commonly observed to possess a Cu core surrounded by a shell of other segregating solute species like Mn, Ni and Si. Here we calculate the segregation energies of substitutional Ni and Si to {100}, {110} and {111} orientated coherent Fe-Cu interfaces. We find Ni is strongly attracted to all orientations of coherent bcc Fe-Cu interfaces whilst Si interacts in a more complex manner, with strong attraction only to the {100} orientation. In calculating these segregation energies we also explore different methods, concluding that for this system using an external reference bulk (ERB) is optimal. By performing additional calculations we were able to decouple the elastic and chemical contributions to segregation energy finding that for Ni the chemical contributions dominate whilst for Si the two contributions are more balanced. In conjunction with previous work we conclude that should Cu nanoprecipitate sizes influence segregation behaviours it is likely this will not be due to the strain state surrounding the Cu nanoprecipitate. It is more probable that this size dependence would originate from variation in the proportions of the specific orientations of coherent Fe-Cu interface that make up the Cu nanoprecipitate’s surface at different sizes.

EXAFS studies for atomic structural change induced by ion irradiation of a reactor pressure vessel steel

Irradiation embrittlement is one of the most important phenomena in the structural integrity assessment for long-term operated reactor pressure vessels (RPVs). The causes of the irradiation embrittlement of RPVs are the formation of solute clusters and matrix defects. To investigate these embrittlement mechanisms, it is necessary to complementarily apply multiple microstructure analyses such as atom probe tomography and transmission electron microscopy. The extended X-ray absorption fine structure (EXAFS) technique can detect changes in the atomic structure by irradiation around selected atoms. In this study, we performed an EXAFS examination for RPV steels subjected to ion irradiation. We investigated microstructures of the crystal structure around specific element such as Cu, Mn, and Ni. The results indicated that the solute atoms cluster is thought to aggregate in the Fe matrix while maintaining the body-centered cubic structure and that a dumbbell pair could be located around both Cu and Ni atoms.

A Continuum Dislocation Dynamics Crystal Plasticity Approach to Irradiated Body-Centered Cubic α-Iron

Abstract Radiation-induced embrittlement of reactor pressure vessel (RPV) steels can potentially limit the operating life of nuclear power plants. Over extended exposure to radiation doses, these body-centered cubic (BCC) irons demonstrate irradiation damage. Here, we present a continuum dislocation density (CDD) crystal plasticity model to capture the interaction among dislocations and self-interstitial atom (SIA) loops in α-iron. We demonstrate the importance of modeling cross slip using a combined stochastic Monte Carlo approach and the role of slip system strength anisotropy in capturing stochastic cross slip interactions. Through these captured interactions, the CDD crystal plasticity model can capture both the stress response and the physical evolution of dislocations on different slip system planes. Single-crystal verification experiments are used to calibrate the CDD crystal plasticity model, and a set of simplified polycrystalline simulations demonstrates the model’s ability to capture the stress response from tensile experiments on α-iron.

The Cause of Intergranular (IG) Fracture by Thermal Embrittlement in SA508 of Reactor Pressure Vessel (RPV) Steel

Intergranular(IG) fracture due to thermal treatment has been reported in a reactor pressure vessel(RPV) steel of Russian light water reactor in last decade. This is attributed to grain boundary segregation of phosphorus (P) or precipitation of carbide, etc.. This is a finding a difference in microstructure before and after IG cracking; this cannot explain the cause of the IG embrittlement. This old paradigm follows only correlation. Recently, a mechanism in which IG embrittlement occurs due to a decrease in entropy of a material has been reported at a temperature where atomic diffusion is possible. It is anticipated that new paradigm can explain the IG embrittlement of RPV based on a causal relationship. Thus, the thermal treatment at 350-420 oC was applied to RPV steel of SA508 and IG cracking was confirmed. DSC analysis was applied to confirm whether a decrease in entropy due to a short range ordering reaction occurs in SA508. It was possible to quantify the entropy change(⊿S= Q/T) through DSC measurement. A lattice changes due to thermal treatment were confirmed using XRD analysis in aged specimens. The results showed that lattice contraction by aging causes a reduction of fracture toughness. The internal stress formed inside the material due to entropy reduction can be calculated by multiplying the exothermic energy per unit mass by the density. This relationship is expressed by a equation of stress(σ) = exothermic heat(⊿Q) x density(ρ).

Grizzly and BlackBear: Structural Component Aging Simulation Codes

The operating environment of nuclear reactors imposes extreme challenges on the materials from which the structures within and surrounding the reactor are constructed. Understanding the effects of exposure to this environment is critical for ensuring the safe long-term operation of these reactors. The Grizzly and BlackBear codes are being developed to model the progression of aging mechanisms and their effects on the integrity of critical structures. These codes take advantage of the capabilities of the MOOSE framework to solve the wide range of coupled physics problems that are encountered in predictive simulation of structural degradation. This paper provides an overview of these codes, with a specific focus on two capabilities relevant for light water reactor applications: reactor pressure vessel embrittlement and concrete degradation.

Bayesian Analysis of Japanese Pressurized Water Reactor Surveillance Data for Irradiation Embrittlement Prediction

Abstract The goal of this study was to identify the chemical component variables that should be used in irradiation embrittlement prediction and to determine the uncertainty of prediction of irradiation embrittlement of reactor pressure vessel (RPV) steels. To this end, statistical analysis using a Bayesian nonparametric (BNP) method was performed for Japanese pressurized water reactor (PWR) surveillance test data whose neutron fluence ranged from 3 × 1018 to 1.2 × 1020 n/cm2 (E > 1 MeV). The BNP method is a machine learning statistical method that takes the complexity and uncertainty of input variables into account. Statistical analysis using an index to select the most suitable combination of input variables revealed that four variables, namely, neutron fluence and Cu, Ni, and Si contents, were the most effective combination for embrittlement prediction. Cu content had the largest effect on the degree of embrittlement, followed by Ni and Si, in that order. The shift in the reference nil-ductility temperature (ΔRTNDT) was also calculated using the probability distribution obtained by the BNP method. The overall standard deviation of the residuals between the calculated and measured values of ΔRTNDT was 8.4 °C, which was comparable to that of the current Japanese embrittlement correlation method (JEAC4201-2013). The 95% credible interval (CI) of the posterior distribution of ΔRTNDT (i.e., the range in which data can exist when the uncertainty of input data is taken into consideration) calculated by the BNP method was identical to or smaller than the margin in the current Japanese embrittlement correlation method described in JEAC4201-2013. This result indicates that an adequate margin is provided in JEAC4201-2013.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

An investigation of nucleation sites for the formation of solute clusters in ferrite Fe

The hardening and embrittlement of reactor pressure vessel steels source from the formation of Cu-enriched precipitates (CRPs) and Mn-Ni-Si precipitates (MNSPs). Recent studies of atom probe tomography indicated that the formation of CRPs and MNSPs may have similar mechanisms, and the major difference may be the nucleation site. Cu clusters and Cu-vacancy complexes may act as the nucleation sites for CRPs, whereas the nucleation sites for MNSPs are unknown. In this work, vacancy clusters of different sizes were analyzed by first-principles calculations for their possibility to act as the nucleation sites for MNSPs. The results were compared with that of Cu clusters and Cu-vacancy complexes in CRPs. Cluster size effects were also discussed. It has been found that Ni was attracted by both Cu cluster and vacancy cluster. Si was not much attracted by Cu cluster but was highly attracted by vacancy cluster, and it was the vacancy in Cu-V complex that attracted Si. In all the configurations calculated, the Mn atoms in antiferromagnetic state would stabilize the configurations and lead to positive binding energy with the clusters. The magnetic state of Mn could be correlated with its atomic volume in bcc-Fe.

Fast Neutron Irradiation Optimization of Thorium-Fueled SCWR Reactor Pressure Vessel

Abstract Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at ∼500 °C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.

Simulation of Cu precipitation in Fe-Cu dilute alloys with cluster mobility

Cu-rich precipitates formation is associated with the precipitation hardening of Fe-Cu based steels and the embrittlement of reactor pressure vessel steels under neutron irradiation. The accurate modeling of the time evolution of Cu-rich precipitates is therefore of fundamental importance for the design of Fe-Cu based steels and the prediction of the irradiation induced shift of the ductile to brittle transition temperature of reactor pressure vessels. This work applies cluster dynamics with mobile Cu monomers and clusters to model Cu precipitation in dilute Fe-Cu alloys at several temperatures. Optimized model parameters can be used to simulate the mean radius, number density, volume fraction, and matrix composition evolution during isothermal annealing with reasonable accuracy. The possible reduction of the mobility of Cu-rich clusters due to additional alloying elements in Fe-Cu based steels is discussed.