Reactor Pressure Vessel Steels Research Articles

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.

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.

HPC atomic simulations of defect formation and kinetic Monte Carlo simulation of material ageing

Kinetic Monte Carlo (KMC) simulations are developed to simulate microstructure evolution under irradiation of structural materials of nuclear power plants. Methods based on rigid lattice atomic KMC and object KMC, despite some approximations, present the advantage to reach significant time, hence irradiation doses. The physical inputs such as the primary damage and defect cluster properties are the results of intensive atomic simulation on high performance computers. Here, atomic KMC is applied to model microstructure evolution of reactor pressure vessel steels and object KMC is applied to zirconium cladding materials.

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.

Spinodal decomposition and radiation damage of a FeCuMnNi high-entropy alloy

The precipitation of Cu-rich phases during the irradiation process will reduce radiation resistance in Fe-Cu-Mn-Ni reactor pressure vessel steels. Therefore, high-entropy alloys with excellent radiation resistance are expected to replace reactor pressure vessel steels and become structural materials for fusion and generation IV fission reactors. This study investigated the phase formation, spinodal decomposition tendency, and radiation resistance of the FeCuMnNi high entropy alloy using a multi-scale method. The results show that FeCuMnNi alloy has a dual-phase face-centered-cubic structure. Here, for the first time, we found that the precipitation phase of the FeCuMnNi high-entropy alloy during the irradiation process is a Mn-rich phase rather than a Cu-rich phase found in conventional Fe-Cu-Mn-Ni reactor pressure vessel steels. Mn separates from the Cu-Mn-rich phase and forms Fe-Mn-Ni clusters with a Fe-Ni phase during irradiation.

Effects of Intercritical Heat Treatment on the Temper Embrittlement of SA508 Gr.4N Ni-Cr-Mo High Strength Low Alloy Steels for Reactor Pressure Vessels

To analyze the effects of intercritical heat treatment on the temper embrittlement of SA508 Gr.4N steels, two model alloys with different phosphorus (P) contents were fabricated. Each sample was heat treated by applying a conventional heat treatment process (quenching-tempering) with/without an intercritical heat treatment process (IHT) and a step-cooling heat treatment for temper embrittlement. Then their microstructure and mechanical properties were evaluated. The microstructure of the SA508 Gr.4N model alloy was composed of tempered lower bainite and martensite, and nano-sized precipitates formed both inside and at boundaries. The grain size was reduced when IHT was applied. There was a small difference in tensile properties according to the heat-treatment conditions and P contents, but the difference in Charpy impact properties was large. The heat treatment for temper embrittlement (TE) increased the impact transition temperature, and a very significant increase was observed in steels with a high P content. The increase in transition temperature owing to TE was reduced when IHT was applied. The fractograph analysis of Charpy fractured specimens at transition temperatures showed that an increase in intergranular fracture was main reason for the TE, and that IHT reduced the formation of intergranular fracture. The AES results showed that P-Ni was segregated at grain boundaries, and the level of segregation was reduced by applying IHT. This occurred because the formation of prior austenite grain boundaries by IHT dispersed the P at grain boundaries, and reduced the amount of P segregation.

Experimental investigation on creep behaviors and life prediction across phase-transformation of thermal aged 16MND5 steel

Reactor pressure vessel (RPV) is one of the key equipment for nuclear power plant, and long-term exposure to high temperature is an important factor causing its performance deterioration. Under core meltdown conditions, RPV steel experiences high temperature creep due to the In-Vessel Retention (IVR) mitigation measures. This paper aims to investigate high-temperature mechanical properties, especially for the creep behaviors of 16MND5 steel after thermal aging at 650 °C for 2000 h. The results showed that the yield stress and ultimate tensile strength of the 16MND5 steel are decreased, and the creep rupture time is rapidly shortened after thermal aging. The microstructure was analyzed by scanning and transmission electron microscopy (SEM, TEM). 16MND5 steel showed a typical bainitic phase before thermal aging. After 2000 h of thermal aging, the micro-grains were coarsened, and finally changed into a ferritic microstructure. Based on Larson-Miller (LM), Orr-Sherby-Dorn (OSD) parameters methods and Kachanov-Rabotnov (K–R) model to predict uniaxial creep life after thermal aging, the prediction results are satisfactory, and all within the double tolerance band. This work reveals the creep damage mechanism of 16MND5 steel after thermal aging, and provides basic data for the implementation of IVR strategy in nuclear power plants after long-term service.

Assessment of neutron irradiation effects via PKA spectra, displacement damage, and gas production: Application to reactor pressure vessel

Neutron irradiation damage is a major challenge for materials employed in radiation environments, such as Reactor Pressure Vessels (RPV). Based on a determined neutron flux, the present study assesses the neutron irradiation effects via Primary Knock-on Atoms (PKAs) production, the consequent displacement damage, and gas production. The complete PKA spectra including all minor isotopes from nuclear reactions are calculated and used to estimate the displacement damage and gas production. At the inner surface of the RPV, which suffers the strongest neutron irradiation, it is found that some minor PKAs have smaller atomic numbers and higher kinetic energies than dominant isotopes in RPV. These PKAs have much deeper ranges in RPV steel and may thus produce important damage other than point defects. The displacement damage is calculated using SRIM simulations and the corresponding corrections to ensure consistency with the NRT-dpa concept. Notably, the displacement damage in the RPV should be ∼23% higher than the prediction by the NRT-dpa formula, which cannot be directly applied to PKAs different from the target atom. The 4.1 H-appm/dpa and 0.2 He-appm/dpa gas productions at the RPV inner surface are higher than that in the innermost 1/4 thickness by a factor of 2 and 1.3, respectively.

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.