Nuclear Pressure Vessel Steels Research Articles

The effect of microstructure evolution on fatigue property of SA508Gr.4N steel for nuclear reactor pressure vessels

The effect of tempering temperature on the microstructure evolution of SA508Gr.4N steel as well as the subsequent effect on the low cycle fatigue property were studied. The tempering temperature was carried out from 595 °C to 675 °C. The experimental results showed that more M/A islands were decomposed at 630 °C, and carbides were dissolved when the tempering temperature increased to 640 °C. The intercritical tempering temperature of the steel reached at a higher temperature (675 °C). After the intercritical range cooling, the granular bainite transformed into martensite, quasi-polygonal ferrite and a small number of undissolved carbides. Subsequently, the tempered specimens at 595 °C, 630 °C and 675 °C were subjected to low cycle fatigue tests under the strain amplitudes of ±0.45% and ±0.6% at 300 °C, respectively. The fatigue life of the steel increased first and then decreased with the increase of tempering temperature. The fatigue stripes separation exhibited a contrary trend to the fatigue life with the increase of tempering temperature. Moreover, the fatigue crack initiation points were transferred from around the M/A islands to the grain boundaries triple point with the increase of tempering temperature from 595 °C to 630 °C. Subsequently, the fatigue crack initiation points were transferred from the grain boundaries triple point to the ferrite phase and the boundary between martensite and ferrite when the tempering temperature increased to 675 °C. The obtained results can provide meaningful reference for improving the fatigue life of nuclear reactor pressure vessels.

Research Progress of Steels for Nuclear Reactor Pressure Vessels.

The nuclear reactor pressure vessel is an important component of a nuclear power plant. It has been used in harsh environments such as high temperature, high pressure, neutron irradiation, thermal aging, corrosion and fatigue for a long time, which puts forward higher standards for the performance requirements for nuclear pressure vessel steel. Based on the characteristics of large size and wall thickness of the nuclear pressure vessel, combined with its performance requirements, this work studies the problems of forging technology, mechanical properties, irradiation damage, corrosion failure, thermal aging behavior and fatigue properties, and summarizes the research progress of nuclear pressure vessel materials. The influencing factors of microstructures evolution and mechanism of mechanical properties change of nuclear pressure vessel steel are analyzed in this work. The mechanical properties before and after irradiation are compared, and the influence mechanisms of irradiation hardening and embrittlement are also summarized. Although the stainless steel will be surfacing on the inner wall of nuclear pressure vessel to prevent corrosion, long-term operation may cause aging or deterioration of stainless steel, resulting in corrosion caused by the contact between the primary circuit water environment and the nuclear pressure vessel steel. Therefore, the corrosion behavior of nuclear pressure vessels materials is also summarized in detail. Meanwhile, the evolution mechanism of the microstructure of nuclear pressure vessel materials under thermal aging conditions is analyzed, and the mechanisms affecting the mechanical properties are also described. In addition, the influence mechanisms of internal and external factors on the fatigue properties, fatigue crack initiation and fatigue crack propagation of nuclear pressure vessel steel are analyzed in detail from different perspectives. Finally, the development direction and further research contents of nuclear pressure vessel materials are prospected in order to improve the service life and ensure safe service in harsh environment.

Investigation of Potential Material Inhomogeneity in the Magnetically Detected Neutron-Irradiation-Generated Structural Degradation of Nuclear Reactor Pressure Vessel Steel

A novel nondestructive method called magnetic adaptive testing has been previously applied to detect the neutron-irradiation-generated structural changes in reactor pressure vessel steel material. This method has been found to be a useful tool for this purpose, and good correlation—as a tendency—has been found between the estimated ductile-to-brittle transition temperature and magnetic parameters. However, a significant scattering of measured points was also observed for the investigated set of Charpy specimens. The main result of the work was that by magnetic selection of samples, the scatter can be notably reduced. As a conclusion, the magnetically measured parameters seemed to be precise and reliable for the detection of embrittlement of the reactor pressure vessel steel, with lower scattering of points than in the conventionally used destructive mechanical characteristics (ductile-to-brittle transition temperature). This result is surprising and needs further verification. The purpose of the present work is to repeat the measurement on irradiated reactor steel blocks. In this work, instead of the DBTT transition temperature, individually measured Vickers hardness (VH) data were used to help characterize the mechanical properties of the material. The so-called “property transformation” is a known and applied technique in the nuclear industry. The mechanical property characterized by the transition temperature cannot be determined individually for each specimen; instead, it can be obtained only on a set of samples by statistical fitting. Therefore, the individually measured Vickers hardness values can be utilized in order to predict the individual transition temperature values by the help of the property transformation technique. In this paper, however, not these derived transition temperature values, but their origins, the Vickers hardness values, are studied in a direct manner. The same behavior of blocks was observed as in the case of Charpy specimens, which is considered to validate the previously published results. As a possible reason for the scattering of points, large magnetic inhomogeneity of samples cut even from the same block was also proved. The magnetic parameters and Vickers hardness correlate well with each other. This result justifies the potential future application of magnetic techniques in practice aimed at the regular inspection of nuclear reactors.

Investigation of the Influence of Neutron Irradiation on Cladded Nuclear Reactor Pressure Vessel Steel Blocks by Magnetic Adaptive Testing

The influence of neutron irradiation degradation on reactor pressure vessel steel was investigated. Large blocks were irradiated by neutrons in the BR2 reactor at a low irradiation temperature. They were measured by a nondestructive magnetic method, magnetic adaptive testing, before and after the neutron irradiation. It was shown that the modification of the magnetic parameters due to the neutron irradiation was well detectable by this nondestructive method. It was also shown that the influence of neutron irradiation could be detected with a reasonable signal-to-noise ratio through the cladding. The present results are considered as a first step towards the application of magnetic adaptive testing in the nuclear industry.

Effect of heat treatment on the impact toughness and brittle fracture initiation mechanism of a quenched and tempered nuclear Pressure Vessel Steel

The effect of heat treatment conditions on the tensile properties, impact toughness properties and the brittle fracture initiation mechanisms of a low-alloy steel were investigated. Only one type of location for the cleavage initiation sites, common to all cooling rates, was identified. Cleavage crack initiation always occurred in the very vicinity of a high angle boundary. MnS inclusions and Ti(CN) precipitates may play a role, but they did not seem to be necessary for cleavage initiation. The critical stress for cleavage increased with the cooling rate and decreased with the tempering parameter. This is consistent with the observed differences in the ductile-to-brittle transition behavior with the heat treatment conditions.

Analysis of Magnetic Nondestructive Measurement Methods for Determination of the Degradation of Reactor Pressure Vessel Steel.

Nondestructive magnetic measurement methods can be successfully applied to determine the embrittlement of nuclear pressure vessel steel caused by neutron irradiation. It was found in previous works that reasonable correlation could be obtained between the nondestructively measured magnetic parameters and destructively determined ductile-to-brittle transition temperature. However, a large scatter of the measurement points was detected even in the cases of the non-irradiated reference samples. The reason for their scattering was attributed to the local inhomogeneity of material. This conclusion is verified in the present work by applying three different magnetic methods on two sets of Charpy samples made of two different reactor steel materials. It was found that by an optimal magnetic pre-selection of samples, a good, linear correlation can be found between magnetic parameters as well as the ductile-to-brittle transition temperature with low scattering of points. This result shows that neutron irradiation embrittlement depends very much on the local material properties.

Interpretation of Nondestructive Magnetic Measurements on Irradiated Reactor Steel Material

Neutron irradiation-generated embrittlement of nuclear pressure vessel steel was inspected by a nondestructive magnetic method, called magnetic adaptive testing (MAT). This method is based on systematic measurement and evaluation of minor magnetic hysteresis loops. Result of MAT measurement was compared with the result of the traditional Charpy measurement. Good correlation was found between these parameters. One of the main findings of the present work is that the considerable part of scatter of points obtained by magnetic measurement can be attributed to local material inhomogeneity. Another important conclusion is that the embrittlement highly depends on the initial local material conditions, i.e., the initial microstructure, which are very different even within the same block of reactor steel material. By taking this into account, the magnetic descriptors obtain more precise determination of the local embrittlement than the traditionally used destructive mechanical parameters from Charpy data.

Micromagnetic Characterization of Operation-Induced Damage in Charpy Specimens of RPV Steels

The embrittlement of two types of nuclear pressure vessel steel, 15Kh2NMFA and A508 Cl.2, was studied using two different methods of magnetic nondestructive testing: micromagnetic multiparameter microstructure and stress analysis (3MA-X8) and magnetic adaptive testing (MAT). The microstructure and mechanical properties of reactor pressure vessel (RPV) materials are modified due to neutron irradiation; this material degradation can be characterized using magnetic methods. For the first time, the progressive change in material properties due to neutron irradiation was investigated on the same specimens, before and after neutron irradiation. A correlation was found between magnetic characteristics and neutron-irradiation-induced damage, regardless of the type of material or the applied measurement technique. The results of the individual micromagnetic measurements proved their suitability for characterizing the degradation of RPV steel caused by simulated operating conditions. A calibration/training procedure was applied on the merged outcome of both testing methods, producing excellent results in predicting transition temperature, yield strength, and mechanical hardness for both materials.

Weibull Modulus of Cleavage Fracture Toughness of Ferritic Steels

The ordinary Weibull distribution function has been commonly accepted for empirical characterization of cleavage fracture toughness of nuclear reactor and containment pressure vessel steels. However, this method lacks a fundamental basis. This work adopts the standardized Weibull distribution function to analyze cleavage fracture toughness of ferritic steels measured from different sized fracture mechanics specimens at different temperatures to estimate the Weibull modulus. The toughness data of five different nuclear reactor and containment vessel steels are analyzed. The estimations obtained the Weibull modulus (m) in the range of 1.83 to 2.55 and strong temperature dependence of the threshold cleavage fracture toughness Kmin, as opposed to the constant values of mK = 4 and Kmin = 20 MPam1/2 given in ASTM E1921-19. The goodness of fit test by the one-sample Kolmogorov–Smirnov (K–S) test validated Weibull distribution function for describing the toughness distribution.

Analysis of Surface Roughness Influence in Non-Destructive Magnetic Measurements Applied to Reactor Pressure Vessel Steels

The influence of surface roughness on magnetic measurements of Reactor Pressure Vessel Steels was investigated by applying two types of magnetic, non-destructive measurement on nuclear reactor pressure vessel steel samples: magnetic adaptive testing (MAT) and magnetic Barkhausen noise measurement (MBN). The surface roughness was modified by primary and secondary machine cutting forces. Different settings of machine cutting produced different surface conditions. It was found that for both measurements a monotonic correlation was found to exist between magnetic parameters and surface roughness. Results of the MAT measurements found that the correlation depends on the speed (i.e., on the applied slew rate) of the magnetizing current. In a similar fashion, results from the MBN method show good agreement with MAT, where the response diminishes with an increase in surface roughness. The results show the importance of accounting for surface condition in the interpretation of results of non-destructive magnetic testing.

Stress intensity factor for w-shape tension specimens

The evolution of the irradiation embrittlement of nuclear pressure vessel steels is determined experimentally using special specimen geometries. Such sub-sized w-shape eccentrically-loaded single edge cracked tension specimens are employed due to mounting and handling requirements. Stress intensity factor solutions are available in the literature for this geometry; however, they are inaccurate for smaller crack lengths. Therefore, new approximate solutions are presented in this work to determine the stress intensity factor for such a w-shape tension specimen for a wide range of crack lengths (0.05 < a/W < 0.9). Furthermore, the T-stress constraint is investigated for this special specimen design and compared with comparable designs provided in standards.

Radiation Damage of Reactor Pressure Vessel Steels Studied by Positron Annihilation Spectroscopy—A Review

Safe and long term operation of nuclear reactors is one of the most discussed challenges in nuclear power engineering. The radiation degradation of nuclear design materials limits the operational lifetime of all nuclear installations or at least decreases its safety margin. This paper is a review of experimental PALS/PLEPS studies of different nuclear reactor pressure vessel (RPV) steels investigated over last twenty years in our laboratories. Positron annihilation lifetime spectroscopy (PALS) via its characteristics (lifetimes of positrons and their intensities) provides useful information about type and density of radiation induced defects. The new results obtained on neutron-irradiated and hydrogen ions implanted German steels were compared to those from the previous studies with the aim to evaluate different processes (neutron flux/fluence, thermal treatment or content of selected alloying elements) to the microstructural changes of neutron irradiated RPV steel specimens. The possibility of substitution of neutron treatment (connected to new defects creation) via hydrogen ions implantation was analyzed as well. The same materials exposed to comparable displacement damage (dpa) introduced by neutrons and accelerated hydrogen ions shown that in the results interpretation the effect of hydrogen as a vacancy-stabilizing gas must be considered, too. This approach could contribute to future studies of nuclear fission/fusion design steels treated by high levels of neutron irradiation.

In-situ measurements of stress during thermal shock in clad pressure vessel steel using synchrotron X-ray diffraction

Thermal shocks are an important incident in operation of a pressure vessel which can have a significant impact on the structural integrity of the vessel. Often experiments that consider the state of the vessel before and after the thermal shock are used to evaluate the effects of the thermal shock. The studies can be complemented by time-resolved numerical simulations, which may be validated against the final state of the vessel obtained experimentally, to infer the transient response of the material. The transient response is important as the material experiences the highest level of stress in a short period which can induce catastrophic failure. This paper reports time-resolved experimental quantification of strain in reactor pressure vessel material during thermal shock measured by in-situ synchrotron diffraction. Specimens were extracted from a plate of nuclear pressure vessel steel with a nickel alloy cladding deposited by overlay welding. The specimens, with and without cracks, were subjected to thermal loading by heating then rapidly quenching the cladding in cold water. Strains were measured during thermal loading at a point near the crack tip from which the stress state around the crack tip was calculated and compared with a transient finite element model of the experiment. It was found that the peak near-tip stress occurred within the first second after the onset of rapid cooling. It was demonstrated from experimental measurements that the peak stress intensity factor occurred during thermal shock, rather than under steady conditions before or after the thermal shock. It was shown that although the finite element simulation predicts the steady state condition of the material after thermal shock, its transient response dependents significantly on a number of inputs with high uncertainty, making its time-resolved results unreliable for high-fidelity integrity assessments.

Detection of neutron irradiaton induced degradation of reactor steel by magnetic method

A novel nondestructive method, based on step by step measurement of minor magnetic hysteresis loops was used to determine the neutron irradiation induced degradation of nuclear reactor pressure vessel steel material. This method is called Magnetic Adaptive Testing (MAT), and two types of reactor steel material were investigated: base material type 22NiMoCr37 and weld material type 18MND5. Charpy geometry samples were irradiated by neutrons at the BR2 in SCK•CEN. The neutron irradiation generated structural changes in the material. As a result of our measurement, it was shown that this degradation could be detected by MAT measurements. Charpy impact tests were also done on the investigated samples and the results of these measurement were compared with the nondestructively determined magnetic quantities. It was shown that a closely linear correlation existed between the ductile-to-brittle transition temperature (DBTT) and MAT parameters.

Influence of phase transformation on the creep deformation mechanism of SA508 Gr.3 steel for nuclear reactor pressure vessels

This study analyzed the effect of phase transformation on the creep behavior and deformation mechanism of SA508 Gr.3 steel used for nuclear reactor pressure vessels (RPVs). The creep tests were conducted on the steel at the temperatures from 650 to 800 °C at a stress ranging from 15 to 80 MPa. Detailed microstructural examination and theoretical analysis were conducted on the specimens to investigate the creep mechanism in each testing condition. The results showed that phase transformation induced an increase in the minimum creep rate and a decline in creep life. Based on the calculations of stress exponents, activation energies, Larson-Miller and Orr-Sherby-Dorn parameters, the creep deformation and fracture behavior were deemed to be different before and after phase transformation. Before the phase transformation, the predominant creep-deformation mechanism was dislocation gliding whereas it was converted to dislocation climbing after the phase transformation. During the phase transformation, the dislocation gliding coupled with grain-boundary sliding became the rate-controlling creep mechanism. This study provides a comprehensive insight into understanding the creep deformation of the RPV steel, giving a direct guideline to optimize the parameters for in-vessel retention strategies.

Inspection of Reactor Steel Degradation by Magnetic Adaptive Testing.

Degradation of nuclear pressure vessel steel materials, 15Kh2NMFA type and A508 Cl2 type (definition is given in the text) were investigated by a novel magnetic nondestructive testing method, so-called Magnetic Adaptive Testing (MAT), which is based on systematic measurement and evaluation of minor magnetic hysteresis loops. The measured samples were thermally treated by a special step cooling procedure, which generated structural changes in the material. It was found that this type of degradation can be easily followed by magnetic measurements. Charpy impact test were also performed and the results were compared with the magnetic parameters. In case of 15Kh2NMFA steel, a good, reliable and closely linear correlation was found between magnetic descriptors and transition temperature.

Measurement and Prediction of Phase Transformation Kinetics in a Nuclear Steel During Rapid Thermal Cycles

Accurate prediction of the residual stress distributions in steel welds can only be achieved if consideration is given to solid-state phase transformation behavior. In this work, we assess the ability of a model for reaction kinetics to predict the phase transformations, and corresponding evolution of volumetric strain, in a nuclear pressure vessel steel when subjected to rapid weld-like thermal cycles. The cases under consideration involved the rapid heating of SA508 steel to a temperature of either 900 °C or 1200 °C for a period of 10 seconds, and subsequent cooling of the material to room temperature at rates between 0.1 and 100 °C s−1. Predictions for the microconstituent proportions and transformation temperatures for each thermal cycle are compared to those measured through a combination of dilatometry, optical and electron microscopy, and synchrotron X-ray diffraction. In general, there was good agreement between measured and predicted transformation start temperatures and microconstituent fractions for cooling rates relevant to welding ( ge 10 °C s−1). Even in the cases in which discrepancies were found for start temperatures, examination of the corresponding dilatation curves showed a good match between predicted and experimental transformation strain evolution. This is a very positive result in terms of residual stress prediction in welds. At slower cooling rates, significant discrepancies arose owing to the model’s incapacity to predict Widmanstätten ferrite or retained austenite, and its failure to account for the effects of carbon redistribution during transformations involving diffusion. Although not relevant to welding, improvements to the model to rectify these issues would be beneficial in terms of its wider predictive capabilities.

Redistribution of residual stress by thermal shock in reactor pressure vessel steel clad with nickel alloy

This paper characterises the residual stress in nuclear reactor pressure vessel steel clad with nickel-based alloy and investigates the interaction between residual and thermal stresses during thermal shock. Residual stress measurements were made on two plates of SA508 Grade 4N steel, clad with Alloy 82 nickel-based alloy. The techniques used to measure the residual stresses were: deep hole drilling, centre hole drilling, and the contour method. One plate was as-welded, the other post-weld heat-treated. The post-weld heat-treated plate was subjected to thermal shock by heating it up and then spraying the surface of the cladding with cold water. The residual stress was measured again afterwards. A finite element simulation was made to investigate the physical mechanisms causing residual stress redistribution during thermal shock. Thermal shock caused significant residual stress redistribution in the cladding due to elastic-plastic interaction between the thermal stress and the cladding residual stress. The results demonstrate that an assessment of the safety of a reactor pressure vessel during thermal shock could be conservative for small surface defects if it is assumed that residual and thermal stresses combine elastically.

Precise Determination of Trace Hydrogen in SA508-3 Steel for Nuclear Reactor Pressure Vessels

Hydrogen is one of the most detrimental elements in steel, therefore it must be strictly controlled, especially in the ferritic steel SA508-3 used in nuclear power plant, due to its persistent exposure to irradiated environment. However, the tests of hydrogen concentration in current industry have provided usually unstable and sparse results, which seriously perplexes practitioners. In the present study, tested hydrogen contents of a series of specimens with different machining and cleaning conditions were compared. Experimental and theoretical investigations showed that differences in surface contamination can significantly influence hydrogen analysis result, and the measured hydrogen content is the combination of the inherent hydrogen in the steel and the adsorbed hydrogen on the surface. Time-of-Flight Secondary Ion Mass Spectrometry (TOF-SIM) was then used to determine the hydrogen concentration gradient from the surface to the base of the SA508-3 steel, with the aim of confirming the existence of surface hydrogen. The result demonstrated that, if the specimen surface is in a rough and polluted state, the count intensity of hydrogen on the specimen surface, which qualitatively reflects the content of surface hydrogen can be more than 4 times that in the base of the specimen. Finally, on the basis of the analysis of tested results, a pre-treatment method combining a certain machining speed and ultrasonic cleaning is proposed in order to obtain a stable and reliable hydrogen analysis result. Furthermore, a mass of practical industry data is used to confirm the effectiveness of this pre-treatment method.

Atomistic simulation of phosphorus segregation to Σ3 (111) symmetrical tilt grain boundary in α-iron

Irradiation-induced grain boundary (GB) phosphorus segregation is an important factor for estimating the embrittlement of nuclear reactor pressure vessel steels, but the physical process of phosphorus migration to grain boundaries is still unclear. We numerically studied phosphorus migration toward a Σ3(111) symmetrical tilt GB in α-iron using molecular dynamics. We found that, in the vicinity of the GB within ∼1 nm distance, an iron–phosphorus mixed dumbbell and an octahedral interstitial phosphorus atom push a self-interstitial atom into the GB, and the phosphorus atom becomes a substitutional atom. A phosphorus-vacancy complex in the region also becomes dissociated, and the vacancy is absorbed in the GB without dragging phosphorus. The results claim that a novel view of the segregation process is required.