Reactor Pressure Vessel Steels Research Articles

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.

Ultrafast annihilation of irradiation-induced defects using pulsed electric current for damage performance regeneration

As the most important irradiation-induced defects, dislocation loop and copper-rich nanocluster are the major contributors to the embrittlement of the neutron-irradiated reactor pressure vessel steels. In this study, such nano-defects were introduced into the material by 3 MeV Fe ions up to the dose of 1 dpa at high temperature (290 °C) to simulate neutron irradiation. It was found that pulsed electric current can effectively reduce 95 % of irradiation-induced hardening. Correspondingly, the characterization results showed that almost all the dislocation loops disappeared and the quantity of copper-rich nanoclusters also reduced greatly at relatively low temperature (450 °C), and the process took only 20 min. Meanwhile, it was qualitatively proved by positron annihilation spectroscopy that the number of irradiation-induced vacancy-type defects and solute-enriched clusters was significantly decreased after electropulsing. Furthermore, under the pulsed electric field, the rapid annihilation of the dislocation loops due to their accelerated collision with vacancies can remove the nucleation sites of the copper-rich nanoclusters and make them become dispersed, further promoting the nanoclusters that lack nucleation sites dissolving faster. Therefore, this electropulsing treatment provides a practical ‘‘in-situ” performance repair technology to extend the service life of reactor pressure vessel steels by regulating the interaction between vacancies, interstitial atoms and irradiation-induced defects.

Evaluation of the Embrittlement in Reactor Pressure-Vessel Steels Using a Hybrid Nondestructive Electromagnetic Testing and Evaluation Approach.

The nondestructive determination of the neutron-irradiation-induced embrittlement of nuclear reactor pressure-vessel steel is a very important and recent problem. Within the scope of the so-called NOMAD project funded by the Euratom research and training program, novel nondestructive electromagnetic testing and evaluation (NDE) methods were applied to the inspection of irradiated reactor pressure-vessel steel. In this review, the most important results of this project are summarized. Different methods were used and compared with each other. The measurement results were compared with the destructively determined ductile-to-brittle transition temperature (DBTT) values. Three magnetic methods, 3MA (micromagnetic, multiparameter, microstructure and stress analysis), MAT (magnetic adaptive testing), and Barkhausen noise technique (MBN), were found to be the most promising techniques. The results of these methods were in good agreement with each other. A good correlation was found between the magnetic parameters and the DBTT values. The basic idea of the NOMAD project is to use a multi-method/multi-parameter approach and to focus on the synergies that allow us to recognize the side effects, therefore suppressing them at the same time. Different types of machine-learning (ML) algorithms were tested in a competitive manner, and their performances were evaluated. The important outcome of the ML technique is that not only one but several different ML techniques could reach the required precision and reliability, i.e., keeping the DBTT prediction error lower than a ±25 °C threshold, which was previously not possible for any of the NDE methods as single entities. A calibration/training procedure was carried out on the merged outcome of the testing methods with excellent results to predict the transition temperature, yield strength, and mechanical hardness for all investigated materials. Our results, achieved within the NOMAD project, can be useful for the future potential introduction of this (and, in general, any) nondestructive evolution method.

Mixed-mode I&II fatigue crack growth behaviors of 16MND5 steel: The role of crack driving forces and crack closure

Although numerous fatigue crack growth (FCG) tests have been conducted, the conclusions drawn regarding the mixed-mode FCG rate are inconsistent and even contradictory. A reasonable conjecture is that crack driving force and crack closure effects may play a crucial role in causing this situation. In this paper, the 16MND5 steel, commonly used reactor pressure vessel (RPV) steel, was employed to investigate the mixed-mode I&II FCG behaviors through the compact-tension-shear (CTS) experiments. The influence of crack driving force and crack closure effect was analyzed and discussed. The results show that the crack propagation path tends to maximize the tangential stress component at the crack tip, and the effect of ΔKII can be negligible once the Paris region is reached. Moreover, when considering the crack closure corrected ΔKI, named ΔKI eff, all the FCG rate data under different loading conditions can be collapsed into a narrower band of distribution. The premise for arriving at this result is to use the stress intensity factors solution that are applicable to specimens with real slanted propagating cracks, rather than applying equivalent methods based on single-edge straight crack assumptions. Finally, the fatigue fractography further confirms that the FCG mechanism during the Paris region all follow a similar quasi-mode I behavior.

Microstructure-Informed Prediction of Hardening in Ion-Irradiated Reactor Pressure Vessel Steels

Ion irradiation combined with nanoindentation is a promising tool for studying irradiation-induced hardening of nuclear materials, including reactor pressure vessel (RPV) steels. For RPV steels, the major sources of hardening are nm-sized irradiation-induced dislocation loops and solute atom clusters, both representing barriers for dislocation glide. The dispersed barrier hardening (DBH) model provides a link between the irradiation-induced nanofeatures and hardening. However, a number of details of the DBH model still require consideration. These include the role of the unirradiated microstructure, the proper treatment of the indentation size effect (ISE), and the appropriate superposition rule of individual hardening contributions. In the present study, two well-characterized RPV steels, each ion-irradiated up to two different levels of displacement damage, were investigated. Dislocation loops and solute atom clusters were characterized by transmission electron microscopy and atom probe tomography, respectively. Nanoindentation with a Berkovich indenter was used to measure indentation hardness as a function of the contact depth. In the present paper, the measured hardening profiles are compared with predictions based on different DBH models. Conclusions about the appropriate superposition rule and the consideration of the ISE (in terms of geometrically necessary dislocations) are drawn.

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.

Development of pressure-temperature limit curves considering unified constraint for reactor pressure vessel

In this work, the pressure-temperature (P-T) limit curves considering unified constraint for a typical large-sized reactor pressure vessel (RPV) with axial surface cracks have been investigated and developed based on the KJ-Ad* two-parameter and the Master Curve (MC) method. The ASME fracture toughness KIc curve and the MC with and without considering unified constraint for irradiated RPV steel were used in the constructions of the P-T limit curves. The effects of crack sizes (constraint levels) and fracture toughness curves on the P-T limit curves have been analyzed for the RPV under heatup and cooldown conditions. The results further show that the P-T limit curves based on traditional ASME KIc curve has the highest degree of conservatism and the use of the MC method can significantly increase the safety margin of P-T limit curves. With increasing crack sizes (increasing constraint levels), the safety region in P-T limit curves reduces, which may lead to the decrease of operating flexibility for nuclear power plants. The accuracy of the P-T limit curves can be improved by considering unified constraint. For smaller size cracks with lower constraint in the RPV, the consideration of constraint can reduce conservatism of the P-T limit curves, and for larger size cracks with higher constraint, it can avoid the non-conservatism. It is recommended to obtain more accurate P-T limit curves based on the MC with considering unified constraint and finite element analysis for RPVs under various operating conditions.

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.

The impact of dose rate on defect clustering and irradiation hardening in reactor pressure vessel steel under Fe ion irradiation

The impact of dose rate on defect clustering and irradiation hardening in reactor pressure vessel steel under Fe ion irradiation

Investigation of the equilibrium morphology of fcc [formula omitted] in Fe–Cu alloys using a non-local Allen–Cahn model

Cu-precipitation is one of the main concerns of the degradation of mechanical properties in reactor pressure vessel steels. It is especially prominent in the over-aged state, where the Cu-rich phase possesses an equilibrium face-centered cubic crystal structure. In this study, we develop and implement a non-local phase-field model to study the equilibrium morphology of precipitates in the over-aged state. The model reveals that the precipitates should possess oblate lath-like morphology, which is supported by novel experimental data. The orientation of the precipitates agrees well with the available literature. The mechanism of the morphology accommodation is explained using a theoretical framework of the invariant-line method. Also, morphology was analyzed quantitatively in the sense of its elongation. It was found that the equilibrium values deviate significantly from experimental data in this metric, which is assumed to be due to the metastable state of the precipitates during the coarsening process.

The influence of solute elements on grain size, hardness and irradiation-induced defect clusters in dilute Fe-based alloys

The precipitation of solutes is the primary factor that leads to the occurrence of irradiation hardening and embrittlement, thus limiting the service life of reactor pressure vessel (RPV) steels. In this study, four model alloys (Fe-1.35Mn-0.75Ni, Fe-1.35Mn-0.20Si, Fe-0.75Ni-0.20Si and Fe-1.35Mn-0.75Ni-0.20Si) were used to investigate the influence of solute elements on grain size, hardness, irradiation hardening, and defect clustering behavior. The materials were irradiated with 3.5 MeV Fe ions to doses of 0.1, 0.3, 1.0, and 3.0 dpa at 290 °C. Results indicated that the addition of solute Mn resulted in grain refinement while the addition of solutes Ni and Si led to grain coarsening. The strength of the model alloys was affected by the change in grain size. A definite dose rate effect was found in all four model alloys under irradiation, where irradiation with a higher dose rate led to a lower irradiation hardening effect and irradiation with a lower dose rate induced a higher hardening effect. Atom probe tomography (APT) results revealed that the precipitated solutes were always coupled with interstitial defect clusters. Solute Mn had a strong influence on cluster nucleation and a weak influence on cluster size. Solutes Si and Ni significantly promoted the nucleation and growth of clusters, and Ni had a greater influence on the nucleation and growth of clusters than Si. Meanwhile, solute Si consistently exhibited a low fraction in the clusters when the solute Mn element was presented in the alloys.

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.

Effect of material inhomogeneity on constraint loss measured with subsized C(T) fracture specimens of a reactor pressure vessel steel

In this study, the fracture behavior of the low-alloy reactor pressure vessel JRQ steel was investigated in the ductile-to-brittle transition region with two different subsized compact tension fracture specimens: 0.18T C(T) and 0.5T C(T). The specimens were extracted from two thin plates (surface and middle) of the large JRQ steel plate, which is well-known to be macroscopically inhomogeneous in terms of brittleness. In particular, the reference temperature T0 as defined in the ASTM-E1921 standard was calculated for the two plates with the 0.18T C(T) and 0.5T C(T) specimens. A significant difference in the reference temperature T0 of more than 90 °C as determined with the 0.5T C(T) data was found between the two plates indicating a clear macroscopic inhomogeneity between the two depths. The measured toughness of 0.18T C(T) specimens and corresponding T0 evaluation showed that the effect of constraint loss was more pronounced for the middle plate due to its lower strength and that the constraint loss effect can be confounded with inhomogeneity. A local approach to fracture model was applied to account for the constraint loss by calibrating a criterion based on critical stressed volume V* and critical stress σ* obtained by finite element simulations. A big difference in the local fracture stress σ* between the two plates was found and based on the deduced local criterion it was shown that the measured 0.18T toughness can be effectively predicted within the temperature range of the ductile-to-brittle transition.

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.