Reactor Pressure Vessel Materials Research Articles

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.

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.

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.

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.

The investigation of the carbon on irradiation hardening and defect clustering in RPV model alloy using ion irradiation and OKMC simulation

The precipitation of solutes is a major cause of irradiation hardening and embrittlement limiting the service life of reactor pressure vessel (RPV) steels. Impurities play a significant role in the formation of precipitation in RPV materials. In this study, the effects of carbon on cluster formation and irradiation hardening were investigated in an RPV alloy Fe-1.35Mn-0.75Ni using C and Fe ions irradiation at 290 °C. Nanoindentation results showed that C ion irradiation led to less hardening below 1.0 dpa, with hardening continuing to increase gradually at higher doses, while it was saturated under Fe ion irradiation. Atom probe tomography revealed a broad size distribution of Ni–Mn clusters under Fe ion irradiation, contrasting a narrower size distribution of small Ni–Mn clusters under C ion irradiation. Further analysis indicated the influence of carbon on the cluster formation, with solute-precipitated defects dominating under C ion irradiation but interstitial clusters dominating under Fe ion irradiation. Simulations suggested that carbon significantly affected solute nucleation, with defect clusters displaying smaller size and higher density as carbon concentration increased. The higher hardening at doses above 1.0 dpa was attributed to a substantial increase in the number density of defect clusters when carbon was present in the matrix.

Study of mechanical properties of RPV base and weld material using SPT technique

Small Punch Test (SPT) is one of the miniature test techniques used for evaluating tensile properties when limited dimension of material is available for standard mechanical test or handling of large specimen is difficult, especially while dealing with irradiated materials. In order to use SPT for determination of mechanical properties, it is crucial to establish correlation coefficients by comparing results obtained by SPT with that from standard tensile tests. Present work involves developing correlation coefficients and evaluation of tensile properties of reactor pressure vessel (RPV) steel using SPT technique. Specimen dimensions used for SPT was 10x10x0.5 mm3. These specimens were prepared from Charpy V-Notch specimens, which were earlier subjected to standard impact testing. Correlation coefficient for RPV material was established by comparison of SPT data with the tensile properties obtained using standard tensile test. Subsequently, SPT data was acquired on RPV thermal material and tensile properties were obtained using the correlation coefficient for validation. This paper describes the SPT set-up used during the current studies and methodology used for establishing the correlation coefficient on RPV base material.

Assessing the impact of irradiation-induced defects on the hardening of reactor pressure vessel materials for sustainable development of nuclear energy

The operational longevity of a nuclear power plant, specifically of the light water reactor type, hinges on the assessment of component degradation, with a particular focus on the reactor pressure vessel (RPV) materials. Among the most prevalent defects responsible for irradiation-induced hardening in RPV materials are point defects (PDs) and dislocation loops (DLs). However, their respective contributions to the overall hardening of RPV materials have not been clearly determined. To address this issue, this study was conducted using iron alloys with equivalent Mn and Ni content to those used in nuclear RPV in Japan. The study employed transmission electron microscopy (TEM), atom probe tomography (APT), and nanoindentation to assess the roles of PDs and DLs in contributing to the hardening behaviour of RPV materials. The results indicated that both solute clusters (SCs), which are formed by the aggregation of PDs, and DLs play equally important roles in the hardening behaviour of RPV materials. However, an in-depth analysis of three-dimensional images using the APT method revealed that there may have been errors in evaluating SCs and DLs, leading to double counting when assessing the impact of irradiation-induced defects on the hardening behaviour of RPV materials. These findings highlight the need for improved accuracy in evaluating these defects to better understand the behaviour of RPV materials under irradiation.

Microstructural Characterization of Reactor Pressure Vessel Steels

Ion irradiation is a promising tool to emulate neutron-irradiation effects on reactor pressure vessel (RPV) steels, especially in the situation of limited availability of suitable neutron-irradiated material. This approach requires the consideration of ion-neutron transferability issues, which are addressed in the present study by comparing the effect of ions with neutron-irradiation effects reported for the same materials. The first part of the study covers a comprehensive characterization, based on dedicated electron microscopy techniques, of the selected unirradiated RPV materials, namely a base metal and a weld. The results obtained for the grain size, dislocation density, and precipitates are put in context in terms of hardening contributions and sink strength. The second part is focused on the depth-dependent characterization of the dislocation loops formed in ion-irradiated samples. This work is based on scanning transmission electron microscopy applied to cross-sectional samples prepared by the focused ion beam technique. A band-like arrangement of loops is observed in the depth range close to the peak of injected interstitials. Two levels of displacement damage, 0.1 and 1 dpa (displacements per atom), as well as post-irradiation annealed conditions, are included for both RPV materials. Compared with neutron irradiation, ion irradiation creates a similar average size but a higher number density of loops presumably due to the higher dose rate during ion irradiation.

Creep behavior and life prediction of a reactor pressure vessel steel above phase‐transformation temperature via a deformation mechanism‐based creep model

AbstractFor nuclear power generation as a carbon‐neutral energy source, in‐vessel retention (IVR) must be implemented to maintain the structural integrity of nuclear reactor pressure vessel (RPV) for more than 72 h under severe accidental conditions. This technology requires accurate prediction of creep deformation and life of RPV material being operated under pressure and extremely high‐temperature gradient. The current work develops a simplified deformation‐mechanism‐based true‐stress (DMTS) model for creep behavior/life‐prediction of SA508 Gr.3 steel, a typical RPV material, above the phase transformation temperatures (800–1000°C). This model is used to evaluate the time to specific creep strain (t3% and t5%) and rupture (tr), in comparison with popular empirical methods such as Orr–Sherby–Dorn (OSD) and Larson–Miller (LM). The simplified DMTS model achieves an excellent agreement with the experimental observations. The controlling deformation mechanisms are also discussed by metallurgical examinations, which provide the physical premise for the model development and application.

Study on irradiation embrittlement behavior of reactor pressure vessels by machine learning methods

A reliable irradiation embrittlement model of reactor pressure vessel (RPV) is critical to the safe operation and life extension of nuclear power plants (NPP). In this study, 474 datasets from NUREG/CR-6551 and 6 machine learning (ML) methods were adopted to accurately predict the irradiation embrittlement behavior (ΔRTNDT) of RPV with comprehensive input factors involving irradiation conditions and metal composition. The GBDT model has the best prediction performance. The calculated result by GBDT model is close to those of JEAC-4201 and ASTM E900-15. Cu content is the most sensitive factor for ΔRTNDT, followed by neutron fluence. This work demonstrates a successful application of ML in promoting NPP safety and RPV materials design.

Analysis of Ductile-to-Brittle Transition Characteristics of Reactor Pressure Vessel Steels

Ductile-to-brittle transition (DBT) characteristics of three steels for reactor pressure vessel (RPV) belt line application are analyzed from new parameters based on model functions describing the strength and toughness characteristics of the materials. In order to estimate nil-ductility temperature (NDT) from strength property, a strain rate–compensated temperature parameter based on the thermally activated deformation of materials is adopted. A measure of NDT is determined from tensile strength properties for the first time assuming an estimated notch tip strain rate at the lower shelf. It is estimated to be 110, 42, and 106 K for the Cr-Mo-V-Ni, 20MnMoNi55, and A533B steels, respectively. The measure of ductile-to-brittle transition temperature (DBTT) in steels using 41-J Charpy impact-absorbed energy on the basis of a logistic class of functions is compared and shown to be equivalent with those obtained from fitting the tanh model equation. A bi-logistic function based on the concept of separable parameters representing the fracture of ductile and brittle zones in steels within the DBTT regime was applied to model the Charpy impact energy behavior of the three steels. The bi-logistic function-fitting parameters yielded a new measure of brittleness as a DBT characteristic of steels that correlated well with other measures of transition temperature of the selected RPV steels. The parameters from the hyperbolic and logistic fitting were used to develop a model relationship suitable for the generation of a master curve based on Charpy energy in exponential form that unifies the transition temperature behavior of the selected western and eastern RPV materials. The model relationship is also found to closely predict ~5 K of the reference temperature To determined as per American Society for Testing and Materials standard E1921 of the selected RPV steels.

Using Mini-CT Specimens for the Fracture Characterization of Ferritic Steels within the Ductile to Brittle Transition Range: A Review

The use of mini-CT specimens for the fracture characterization of structural steels is currently a topic of great interest from both scientific and technical points of view, mainly driven by the needs and requirements of the nuclear industry. In fact, the long-term operation of nuclear plants requires accurate characterization of the reactor pressure vessel materials and evaluation of the embrittlement caused by neutron irradiation without applying excessive conservatism. However, the amount of material placed inside the surveillance capsules used to characterize the resulting degradation is generally small. Consequently, in order to increase the reliability of fracture toughness measurements and reduce the volume of material needed for the tests, it is necessary to develop innovative characterization techniques, among which the use of mini-CT specimens stands out. In this context, this paper provides a review of the use of mini-CT specimens for the fracture characterization of ferritic steels, with particular emphasis on those used by the nuclear industry. The main results obtained so far, revealing the potential of this technique, together with the main scientific and technical issues will be thoroughly discussed. Recommendations for several key topics for future research are also provided.

On the use of mini-CT specimens to define the Master Curve of unirradiated reactor pressure vessel steels with relatively high reference temperatures

The safe operation of nuclear plants requires an accurate characterization of the fracture toughness of reactor pressure vessel materials, which can be reduced over time due to irradiation or thermal aging processes. This necessity is a challenge itself since the availability of specimens inside the surveillance capsules of the vessels is generally scarce. Therefore, innovative techniques have to be applied, in order to increase the reliability of fracture toughness measurements and at the same time to reduce the volume of material needed for the tests. In this paper, the Master Curve (MC) approach has been employed, combined with the use of mini-CT specimens made from two different reactor pressure vessel (RPV) steels (ANP-5 and A533B LUS). The MC methodology allows the fracture toughness of the material to be evaluated by using a single parameter: the reference temperature, T0. This parameter has been previously estimated using mini-CT specimens in a number of unirradiated steels, most of them with relatively low T0 values (-120 °C to −60 °C), providing satisfactory results. This paper adds further validation of the use of mini-CT specimens to define the T0 of unirradiated RPV steels with relatively high values of this parameter. Additionally, the analysis of the fracture surfaces confirms the existence of cleavage fracture following the weakest link theory (i.e., one single initiation point), as required by the Master Curve approach.

Verification Study of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel During Pressurized Thermal Shock

This investigation is to present a verification study of probabilistic fracture mechanics (PFM) analysis code for a reactor pressure vessel (RPV) during pressurized thermal shock (PTS). The probabilistic fracture mechanics code FAVOR, developed by Oak Ridge National Laboratory, is used to calculate the conditional probabilities of crack initiation and penetration for welds that are located in the RPV beltline region. The procedure includes deterministic analyses of the temperature and stress distributions through the vessel wall at the PTS, and probabilistic analyses on the vessel failure probability as a result of PTS transients. The RPV geometries, material properties, and properties related to embrittlement are those in taken from previous studies. Two previously suggested hypothetical transients, which may seriously affect RPV integrity, are also taken into account. To verify the results of PFM round robin analysis of RPV during PTS events, several models and Monte Carlo methods for determining PFM performance are used and they agree on the accuracy of the failure assessment is obtained. The present work can be regarded as various important factors about performing PFM that affect in evaluating the structural safety and operational stability of RPVs. The comparisons of the paper also support the finding that the FAVOR code is very practically useful in assessing failure probability.

Increase of nuclear installations safety by better understanding of materials performance and new testing techniques development (MEACTOS, INCEFA-SCALE, and FRACTESUS H2020 projects)

Research to better understand the phenomena influencing materials and components’ performance is important for increasing the safety of Generation II and III nuclear plants. A crucial step for improving nuclear safety is the development of new experimental techniques that can provide the necessary data. The three H2020 projects presented in this paper, MEACTOS (2017–2022), INCEFA-SCALE (2020–2025), and FRACTESUS (2020–2024), cover the steps needed to realize those safety improvements. The goal of the MEACTOS project is to improve the resistance of critical locations, including welds, to environmentally-assisted cracking through optimizing surface machining and treatments. The project is currently in its final stage, and the complete analysis of the data is finished. The objective of INCEFA-SCALE is to improve predictions of component fatigue lifetime when subjected to Environmentally-Assisted Fatigue (EAF). The strategy consists of producing guidance on how to appropriately accommodate variable amplitude and plant-relevant loading in EAF assessments. Increasing the understanding of the EAF mechanism based on substantial testing, characterization, and analysis program will support the INCEFA-SCALE strategy. The FRACTESUS project will validate the use of miniaturized compact tension specimens by comparing the results of master curve-oriented fracture toughness tests performed with small and large specimens. The round-robin exercises will use irradiated and non-irradiated Reactor Pressure Vessel (RPV) materials. The material selection process is complete in time for the project to enter the testing phase. The output of the project will be beneficial from a long-term operation perspective and a saving in the material amount needed for RPV surveillance programs. Even though each project is devoted to different research areas, common aspects are clearly visible. All three projects investigate phenomena that are relevant to the performance and safe operation of the nuclear plant. Moreover, each project will provide valuable databases and analyses of test results for materials relevant to components in the nuclear plant. The output of these projects will be of great value to the nuclear industry. This paper presents the current progress for each project, emphasizing the common research domains between the projects.

Master curve evaluation of ANP-5 steel using mini-CT specimens

The nuclear industry demands analyses that make possible the long-term operation of nuclear power plants (i.e., beyond 40 years). In this sense, one of the main challenges to overcome is the restricted amount of material available to extend the surveillance programs. To mitigate this issue, mini-CT specimens have been proposed for the evaluation of the fracture properties of reactor pressure vessel (RPV) materials, and particularly, the corresponding Master Curve. These specimens can be taken from the broken halves of previously tested Charpy specimens. In this work, mini-CT specimens have been employed to evaluate the reference temperature of the RPV ANP-5 steel in non-irradiated conditions. The results were compared with those obtained by means of conventional fracture tests.

A Versatile Methodology for Reactor Pressure Vessel Aging Assessments

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime threshold. From the perspective of possible further lifetime extension, satisfying safety requirements is a priority. Characterization of the structural integrity of the reactor pressure vessel (RPV) is an important issue as it is a guiding parameter that influences the reactor lifetime. Embrittlement of RPV material is primarily induced by the bombardment of fast neutrons (with energies greater than 1 MeV). Consequently, fast neutron fluence is one of the quantities used by safety authorities to characterize the structural integrity of RPV. However, future RPV aging assessments might lean on new variables with respect to current laws, such as neutron fluence considering the whole neutron spectrum or displacements per atom (dpa) since the latter is more representative of overall damage generated in the RPV. In order to meet these challenges, a versatile calculation scheme for RPV aging assessments is proposed in this paper. The developed methodology allows one to compute (fast and non-fast) neutron fluence as well as dpa rate, using the Norgett-Robinson-Torrens dpa model and the Athermal Recombination Corrected dpa model, for a wide azimuthal and axial range on the RPV and in the capsules of the aging monitoring program (which contain dosimeters and vessel material samples). This methodology is based on a coupling between deterministic (CASMO5 and SIMULATE5) and Monte Carlo (MCNP6) numerical approaches. First, the deterministic approach is used to evaluate the full-core fission neutron source term. Second, Monte Carlo modeling is used to perform the neutron attenuation from the core to sites of interest, such as the RPV. The computational efficiency, accuracy, and potential benefits of the methodology are presented. Moreover, the frequency at which neutron transport calculations should be performed in order to obtain sufficiently accurate time-integrated data over a reactor cycle is discussed. Finally, the validity of the fast neutron fluence as an indicator of RPV aging is compared against the use of dpa.

A Comparative Study of Fracture Toughness Distribution of Reactor Pressure Vessel Material with Generalized Extreme Value Distribution and Weibull Distribution in the Lower Self of Ductile to Brittle Transition Region

A Comparative Study of Fracture Toughness Distribution of Reactor Pressure Vessel Material with Generalized Extreme Value Distribution and Weibull Distribution in the Lower Self of Ductile to Brittle Transition Region

Irradiation temperature monitoring with SiC for RPV steel at low fluence

Accurate knowledge of the irradiation temperature is a key concern in irradiations of reactor pressure vessel (RPV) steel. We report results of passive temperature monitoring of RPV steel with SiC. Two un-instrumented capsules containing RPV steel blocks were irradiated in Belgian Reactor 2, at SCK CEN in Belgium. Because un-instrumented capsules were used, the irradiation conditions (gamma heating and irradiation temperature) were calculated. To have experimental verification of the irradiation conditions each capsule contained a passive silicon carbide (SiC) temperature monitor. After irradiation the resistivity and the radiation-induced swelling (lattice parameter) is measured using x-ray diffraction (XRD). These measurements were repeated after annealing treatments to determine the peak irradiation temperature. The best method to determine the peak irradiation temperature with the lowest uncertainty in this study was the resistivity technique. Although lattice parameter measurements could be improved by using material without preferential orientation. Tensile tests of the RPV blocks were compared to a large available database of RPV material with the proper sensitivity towards neutron fluence and irradiation temperature to find the irradiation temperature of the RPV steel. The data showed a discrepancy between the temperature obtained from SiC and the temperature obtained from tensile tests. The temperature differences were discussed and rationalized by finite element modelling with respect to uncertainties in the helium gap between the capsule and the RPV steel block. The aluminium holder accounted for the temperature difference and the use of a steel holder is recommended for future similar irradiations to minimize the temperature difference between SiC and tensile specimens . In-depth analysis showed that when the irradiation temperature dropped at the end of the irradiation for a considerable time, the SiC temperature monitor had a memory effect. In this case the post-irradiation resistivity analysis showed two peak irradiation temperatures with a region of constant resistivity between the two.

Fatigue reliability design and assessment of reactor pressure vessel structures: Concepts and validation

Traditional design approaches using safety factors are intended to account for uncertainties due to the lack of available data in engineering practice. However, the direct use of safety factor to represent uncertainties often causes conservative results, and it is not intuitive to describe the reliability of these structures through safety factor. To deal with the issue, this research revisits the initial consideration and intent of the safety factor for static strength and adjustment factor for fatigue design curve in the ASME code. Three conversion models applying to reactor pressure vessel (RPV) materials are proposed on basis of the stress–strength, load–life and strength–damage interference theories, which elaborate the relationship between safety/adjustment factors and structural reliability. The dispersion of strength and fatigue life is quantified, and then implemented to elaborate safety and adjustment factors with reliability respectively, which verifies the validity of safety assessment methods in the ASME code. Moreover, the safety/adjustment factor design under required reliability level is achieved, has the advantages of intuitiveness and convenience, which provides a reference for the reliability and safety design of RPV.