Reactor Pressure Vessel Steels Research Articles

A study on initial formation stage of proton irradiation defects in reactor pressure vessel steel

A study on initial formation stage of proton irradiation defects in reactor pressure vessel steel

Specimen reconstitution method for the assessment of fracture properties of reactor pressure vessel steels

Routine testing of the fracture properties of irradiated surveillance specimens is a critical method for evaluating irradiation embrittlement in reactor pressure vessel (RPV) steels. Given the limited space in irradiation surveillance tubes, specimen reconstitution technology is expected to provide a viable solution to the shortage of irradiated specimens. This study investigates the specimen reconstitution method for 16MND5 steel 0.5 T CT specimens used in the RPV. By studying the local mechanical properties of the residual specimen and performing finite element analysis of the crack tip plastic zone, the appropriate size for specimen reconstitution is determined. The electron beam welding method is used to join the residual specimen and the base material, and the J-integral formula for the reconstituted specimen is determined based on the finite element analysis, taking into account the influence of the weld seam and the degradation of the mechanical properties of the residual specimen. The feasibility of the specimen reconstitution method is validated through fracture tests on base material specimens and reconstituted specimens at different temperature.

Development of a new irradiation-embrittlement prediction model for reactor pressure-vessel steels

Development of a new irradiation-embrittlement prediction model for reactor pressure-vessel steels

Fracture studies on cruciform bend specimens of pressure vessel steel subjected to thermo-mechanical loading

The safety of a nuclear power plant during incidents, such as a Loss of Coolant Accident (LOCA), heavily relies on conducting a comprehensive structural integrity assessment of both the Reactor Pressure Vessel (RPV) and its components, specifically to withstand Pressurized Thermal Shock (PTS). PTS is characterized by a combination of steep temperature gradient, resulting from the injected emergency coolant during a LOCA, and internal pressure within the RPV. Majority of the reported fracture assessment studies on RPV steel, whether experimental or numerical investigations, have predominantly focused on standard uniaxial specimens at iso-thermal loading conditions. To better simulate the thermal shock scenario in RPVs, the present work aims to assess the impact of a biaxial stress field on both fracture parameters (crack mouth opening displacement and J-integral) and plastic collapse load. This assessment is conducted through experimental and numerical investigations, both with and without prior transient thermal load. Fracture experiments are performed on cruciform bend specimens, featuring two different biaxiality ratios (1:1 and 2:1). Moreover, numerical studies on cruciform specimens are conducted using finite element analysis to validate and corroborate the observations derived from the fracture experiments.

How precisely are solute clusters in RPV steels characterized by atom probe experiments?

Atom probe tomography (APT) is a powerful microscopy technique to characterize nano-sized clusters of the alloying elements in the bulk of reactor pressure vessel (RPV) steels. These clusters are known to dominantly determine the evolution of mechanical properties under irradiation. The results are conventionally summarized as the overall number density N and the average diameter D of the solute clusters identified in the material. Here, we illustrate that these descriptors are intrinsically imprecise because they are steered by the parameters involved in the measurement and data processing, some of which are directly under the control of the operators, but some others not. Consequently, a direct comparison between data derived at different laboratories is compromised, and key trends such as the evolution with dose, are masked. This study relies on a state-of-the-art physical model for neutron irradiation in steels to make reliable estimates of the microstructure before the measurement is performed, which allows the prediction of the population of solute clusters that are not seen by APT. We mimic APT measurements from simulated microstructures, performing a detailed study of the effects of the parameters of the analysis. We show that the values of N and D reported in the scientific literature can be matched by the predictions of our theoretical model only if specific sets of parameters are used for each laboratory that issued the measurements. We also show that if, on the contrary, all studied cases are analyzed in a consistent way, the scatter of N and D values is reduced. Specifically, we find that the average diameter D is nearly a constant value with dose, independently of the material's chemical compositions, while N increases with dose, but is also influenced by other variables. The approach we developed and used proves to have added value as complement to APT experiments in reactor pressure vessel steels.

On the iron content of Mn-Ni-Si-rich clusters that form in reactor pressure vessel steels during exposure to neutron irradiation

Solute cluster formation during service is the primary cause of embrittlement in reactor pressure vessel (RPV) steels. Atom probe tomography (APT) has consistently shown that the solute clusters are enriched in Cu, P, Mn, Ni, and Si compared to the matrix. However, there is pronounced disagreement within the literature as to the Fe content of the solute clusters in low-Cu RPV steels; some authors report Fe contents in excess of 80 at.% whilst others attribute all measured Fe in the clusters to aberrations from APT and report Fe-free features. This discrepancy has profound implications for determining whether cluster formation is radiation-induced or radiation-enhanced.In this article, we perform a systematic analysis of the published literature to demonstrate the variance present in the measured Fe content of solute clusters characterised by APT. We investigate potential explanatory variables for this variation and discuss the difficulties in performing statistical analyses on these data. To determine the impact of APT aberrations on the measured Fe content of solute clusters, we combine results from: simulations that account for the limited spatial precision in APT; models that predict ion trajectories in the presence of non-uniform electrostatic fields; and experimental data.Our results show that APT aberrations can introduce large amounts of artificial Fe into solute clusters, especially at the small sizes (radius <1.5 nm) typically observed in RPV steels. We also show, however, that some of the variance in Fe content in the literature may be explained by real differences in the clusters’ Fe contents.

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.

On the grain level deformation of BCC metals with crystal plasticity modeling: Application to an RPV steel and the effect of irradiation

Capturing realistic deformation behavior in BCC metals at the polycrystal scale is an important aspect of predicting the material’s strength and failure. Furthermore, local deformation/strength heterogeneity also influences the lifetime and its assessments in reactor pressure vessel (RPV) steels, especially with accumulating irradiation doses during long-term operational conditions. This work utilizes a micromorphic crystal plasticity (CP) model for BCC materials with the capability to address temperature-dependent stress–strain response and irradiation effects relevant to RPV materials. The microstructure of the investigated material was characterized prior to and during testing using electron microscopy experiments, which are used for finite element model generation and simulation result validation. To analyze the validity of the model to predict strain localization under monotonic tensile loading, micro digital image correlation (DIC) was employed jointly with the CP simulations. Different modeling choices, such as the use Schmid/non-Schmid laws and gradient based CP modeling, greatly affected the capability of the model to represent similar magnitudes in strain localization and grain reorientation. The requirement for experimental measurements, including image based analyses, on a microstructural level as a validation tool for CP modeling is clearly demonstrated.

Investigation of Irradiation Hardening and Effectiveness of Post-Irradiation Annealing on the Recovery of Tensile Properties of VVER-1000 Realistic Welds Irradiated in the LYRA-10 Experiment

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 1020 n·cm−2, and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties.

Cluster dynamics study on nano damage of RPV steels under proton irradiation at 290°C

Irradiation-induced defects such as dislocation loops, cavities or solute clusters and chemical composition segregation of reactor pressure vessel (RPV) steel are the root causes of irradiation embrittlement. Combining two nucleation mechanisms, namely, the uniform nucleation and non-uniform nucleation of solute clusters (such as Cu-rich phase), a cluster kinetic simulation was established based on the reaction rate theory, and the co-evolution of matrix damage and Cu-rich phase in low-copper RPV steel was simulated under irradiation. And the average size and number density of defective clusters and solute clusters were established with irradiation dose. Compared with the average size and number density of dislocation loops observed by transmission electron microscopy (TEM) of proton irradiated RPV steel at 290°C, the verification results show that the cluster dynamics model considering both the nucleation mechanism of interstitial dislocation loops and vacancy clusters can well simulate the irradiation damage behavior of materials.

Sensor degradation in nuclear reactor pressure vessels: the overlooked factor in remaining useful life prediction

Sensor degradation poses a critical yet ‘often overlooked’ challenge in accurately predicting the remaining useful life (RUL) of nuclear reactor pressure vessels (RPVs), hindering safe and efficient plant operation. This paper introduces an approach to RUL estimation that explicitly addresses sensor degradation, a significant departure from conventional methods. We model neutron embrittlement, a dominant degradation process in RPV steel, as a Wiener process and leverage real-world surveillance capsule data for insightful parameterization. Maximum likelihood estimation is utilized to characterize the degradation dynamics in the model. A Kalman filter then seamlessly integrates the degradation model with sensor measurements, effectively compensating for degradation-induced errors and providing refined state estimates. These estimates power a robust RUL prediction framework. Our results expose the profound impact of sensor degradation on conventional RUL predictions. By directly confronting sensor degradation, our method yields substantially more accurate and reliable RUL estimates. This work marks a significant advancement in the field of materials degradation, offering a powerful tool to optimize nuclear power plant safety and longevity.

Research progress on thermal aging of reactor pressure vessel steel

Nuclear energy is highly valued in many countries for its reliability, cleanliness and cost-effectiveness. Ensuring the safe development and utilization of nuclear energy has become crucial for addressing the global energy shortage and mitigating climate change. The reactor pressure vessel (RPV) is exposed to a high-temperature, high-corrosion and high-radiation environment, the RPV materials have a significant impact on the stability and safety of a nuclear power station. In this paper, the effects of long-term high-temperature hot aging on microstructure and mechanical properties of RPV steel, as well as the research progress in embrittlement mechanism and service life prediction, are reviewed. At the same time, the existing problems and further research directions of RPV steel are pointed out.

Simulation Study on Defect Damage Behavior of Fe under Irradiation Environment

Abstract As a commonly used structural material in nuclear power plants, Reactor pressure vessel (RPV) steel is subjected to high-energy neutron irradiation from the reactor core for a long time, and the service environment is harsh. The accumulation of point defects (interstitial atoms and vacancies) generated by irradiation over a long period can seriously affect the microstructure of the material. The macroscopic properties of the material changed until the material failed, which threatened the safety and operation stability of nuclear power plants. This paper used the method of molecular dynamics to simulate the cascade collision process of Fe in the irradiation environment. The relationship between different factors and radiation damage defects was studied. Firstly, the quantity of the Frenkel pairs increases rapidly in the irradiation environment, and then recombination occurs after reaching the peak, which makes the quantity of the Frenkel pairs decrease rapidly. Finally, the quantity of Frenkel pairs is in a stable trend. When PKA energy and temperature increase, the higher the quantity of Frenkel pairs at the peak is, the higher the recombination rate of defects is. Meanwhile, the larger the cluster size of interstitial atoms and vacancies is, the greater the corresponding quantity of clusters is. As PKA energy increases, the quantity of residual Frenkel pairs at stability increases, while the number of residual Frenkel pairs at stability decreases with temperature. This is attributed to the intense thermal motion of molecules at high temperatures, resulting in the quantity of Frenkel pairs decreasing at the stable stage. Therefore, by simulating the irradiation damage process of Fe, the prediction of material irradiation damage under a neutron irradiation environment is provided, and the service life of materials can be provided with theoretical guidance.

Analysis of Shift in Nil-Ductility Transition Reference Temperature for RPV Steels Due to Irradiation Embrittlement Using Probability Distributions and Gamma Process

Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials.

Experimental and modeling study on irradiation effect of A508-Ⅲ steel

Irradiation can induce microscopic defects in reactor pressure vessel (RPV) steel, resulting in macroscopic hardening and embrittlement. This paper proposes a micro‑meso-macroscale calculation framework for the mechanical properties of irradiated RPV steel based on experimental and numerical investigations. The mean size and number density of Fe-ion-irradiation-induced dislocation loops were measured using transmission electron microscopy (TEM). These parameters were described by a power-function relationship with the irradiation dose at various temperatures and were used as input parameters for the crystal plasticity finite element model (CPFEM). Nano-indentation tests of A508-Ⅲ steel under different irradiation conditions were successfully simulated. In addition, microstructural characterization and molecular dynamics (MD) simulations revealed the phenomenon of dislocation pinning and the sweeping of irradiation defects during dislocation migration, implying that the dislocation-loop interactions could be responsible for the formation of dislocation channels and curved dislocation lines. Furthermore, increase in yield stress and the ductile-brittle transition temperature (DBTT) at various temperatures were investigated. Results indicated that yield stress, DBTT, and the degree of irradiation hardening and embrittlement increased with the irradiation dose and decreased with irradiation temperature. Finally, the adjusted reference temperature (ART), a significant design parameter for nuclear reactors, was calculated and evaluated using the multiscale framework. The framework can serve as a reference text for extending the life of nuclear reactors.

Understanding the effect of phosphorous on the ion-irradiation behaviour of RPV model steels using atom probe tomography and nanoindentation

Understanding the formation of the embrittling Mn-Ni-Si (MNS)-rich clusters in reactor pressure vessel (RPV) steels is of economic, environmental, and safety importance. Hence we investigated the influence of phosphorous (P) on the formation of MNS-rich clusters in RPV model steels employing atom probe tomography and nanoindentation tests. The atom probe tomography results show that the average number density and volume fraction of clusters decrease slightly with an increase in the bulk P content; however, higher bulk P led to a slight increase in the average diameter of the clusters. A higher amount of bulk P led to higher Cu in the clusters; suggesting synergy between Cu and P. An increase in the irradiation hardening values was observed due to higher bulk P content. This is attributed to the stabilisation of the self-interstitial atoms (SIA) clusters by P. A higher recovery for the sample containing higher bulk P indicated that the SIA clusters dissolved after post-irradiation annealing.

Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ∼0.5–1.0 displacements per atom (dpa) at ∼270 °C and ∼370 °C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs.

A multiscale study on the microstructure and hardening models of the irradiation defects on reactor pressure vessel steels: Modelling and experiment

Irradiation defects are known to impede the dislocation motion in irradiated reactor pressure vessel (RPV) steels. However, consistent experimental studies of the respective irradiation defects and relevant theoretical studies of the interaction mechanism and hardening model are still lacking. In this study, the microstructure and irradiation-induced hardening of RPV steels are investigated using transmission electron microscopy (TEM)/energy-dispersive X-ray spectrometry (EDS) and molecular dynamics (MD) simulations. In A508-Ⅲ steel, 9R, twinned fcc, and other typical Cu-rich cluster structures are observed using TEM after the irradiation experiment. Based on the experimental results, MD simulations are performed to study the interactions between the dislocations and Cu clusters or voids. A mechanism diagram of the interaction mechanism coupling the effects of temperature and defect size is shown for both the Cu clusters and voids. A relevant hardening model is proposed based on the MD data, which is further discussed in relation to the macroscopic mechanical properties. Furthermore, the irradiation-induced hardening of the RPV steel is successfully predicted, which further demonstrates the reliability of the proposed unified hardening model.

Effect of non-clustering element C and Mo on irradiation-induced clustering in reactor pressure vessel steels

Commercial reactor pressure vessel steels in nuclear power plant applications contain C and other minor non-clustering elements in addition to the clustering elements of Mn, Ni and Si. However, the effect of such non-clustering elements on irradiation-induced solute clustering in ferritic steels has seldom been addressed. Herein, we reveal that a commercial RPV steel and its Fe-Mn-Ni-Si model alloy with coarse/fine grains subjected to ion irradiation exhibits different irradiation-induced clustering, particularly in chemistry, mean size, and number density of irradiation-induced MnNiSi-clusters. The different chemical compositions of the clusters indicate different thermodynamics in driving solute clustering in RPV steel and its model alloy. The different number density and size evolutions of the clusters indicate different kinetics in governing irradiation-induced clustering. The presence of C and Mo, as non-clustering elements, is primarily responsible for the different clustering behaviors of Ni, Mn and Si in the two materials exposed to ion irradiation. The research finding is of significant implication for modelling and fundamental understanding irradiation-induced clustering in commercial RPV steels.

Study on the Re-Aging Behavior of Cu-Rich Precipitates in a FeCu Alloy under Electropulsing.

The nanoscale Cu-rich precipitates (CRPs) are one of the most critical microstructural features responsible for degrading the mechanical properties of reactor pressure vessel (RPV) steels. The prospect of the rapid regeneration of the service performance of degraded materials through electropulsing is attractive, and electropulsing has been proven to have the application potential to eliminate the CRPs and recover the mechanical properties of RPV materials. However, few studies have investigated the secondary service issue of electropulsing. This paper provides experimental findings from microstructural investigations and property evaluations of a FeCu RPV model alloy subjected to re-aging following recovery electropulsing and annealing treatments. The evolution behavior of CRPs and the changes in the hardness of the alloy during the re-aging process after electropulsing treatment were consistent with the initial aging process, while the re-aging process of the annealing treatment group was quite different from the initial aging. The difference between the electropulsing and annealing treatment groups was that the annealing treatment failed to eliminate the precipitates completely, leaving behind some large precipitates. This work demonstrates the potential application of EPT in this field.