Reactor Pressure Vessel Steels Research Articles

The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels

Embrittlement of light water reactor pressure vessel (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. The steels were irradiated to a fluence of ∼ 1.38×1020 n/cm2 at ∼ 292 °C in the UCSB ATR-2 experiment. In nominally Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promote precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.

Contributions of Ni-content and irradiation temperature to the kinetic of solute cluster formation and consequences on the hardening of VVER materials

The study of solute clusters induced by neutron irradiation is fundamental to the understanding of the embrittlement phenomena of Reactor Pressure Vessel (RPV) steels. In this paper, the influence of Ni-content and irradiation temperature on these cluster formation are investigated using atom probe tomography.VVER 1000-BM (base metal) and 1000-W (weld) steels were chosen for their high Ni content of 1.2 and 1.7 at%, respectively. They were irradiated up to doses of 0.28 dpa at two temperatures (265 and 300 °C). The irradiations were performed at SCK CEN in the BR2 reactor.After irradiation, Mn, Ni and Si rich clusters were observed. A careful study of their composition indicates that these clusters actually contain iron. Moreover, the evolution of Fe cluster concentration with the dose indicates a constant supply of solutes throughout the irradiation thanks to the flux coupling. With increasing dose, both the cluster size and number density increase.The effect of temperature has to be dissociated from that of Ni content. Both of them result in an increase of the cluster number density at a given dose. The irradiation temperature has a more pronounced effect than chemical composition (mainly Ni effect). This can be explained by the fact that more matrix damage is produced at lower irradiation temperature. The influence of Ni on cluster number density is significantly higher at the lower temperature. All these observations support the hypothesis of radiation-induced segregation, even if a thermodynamic contribution to solute clustering cannot be excluded, in particular in the higher Ni steel irradiated at low temperature.In all cases, the yield strength increases proportionally to the volume fraction of radiation-induced defects (√(Vf)).

Characterizing the flux effect on the irradiation embrittlement of reactor pressure vessel steels using machine learning

In-service exposure to high-energy neutrons embrittles reactor pressure vessel (RPV) steels. An increase in the yield stress (Δσy) results in a corresponding increase in the brittle to ductile transition temperature (ΔTc). Most existing models underpredict ΔTc at higher fluence following accelerated irradiations in test reactors. High fluence, up to 1020 n/cm2 in some cases, will be reached over extended RPV vessel operation of 80 years, or more, at low service flux. Embrittlement has been extensively studied in accelerated, higher flux test reaction irradiations. However, the use of test reactor data naturally raises the question of flux effects. This study used a machine learning approach trained on a set of hardening data, covering a wide range of flux, fluence, and steel compositions to determine the interactive effects of both irradiation and material variables on Δσy. The analysis included machine learning-based cross-plots of the variable dependence of Δσy for six core steels (i.e., CM6, LC, LD, LG, LH, and LI), with controlled differences in their Cu and Ni contents. A primary objective is to evaluate an effective to actual fluence (ɸte/ɸt) ratio, as a function of flux, fluence, and steel composition. This is information critical to properly use intermediate flux-high fluence data in calibrating a low flux-high fluence embrittlement model. The predicted ɸte/ɸt is reasonably consistent with estimates previously derived from a physics-based solute recombination trap model.

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.

The role of nickel in forming a structure providing increased service properties of reactor structural materials

Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.

Experimental studies of the influence of biaxial bending on the crack resistance of reactor pressure vessel steel

Background: to assess the residual life of nuclear reactor hulls in Ukraine, normative temperature dependences of the crack resistance of hull steels obtained on standard samples with a deep through crack under plane strain conditions are used. However, it is known that the thickness of the wall of the reactor body is subject to biaxial loading, which affects the value of the crack resistance characteristics. Therefore, in order to substantiate the possibility of lengthening beyond the design resource of VVER-1000 reactor hulls, an experimental study of the effect of biaxial load on the crack resistance characteristics of hull reactor steel 15X2NMFAA is relevant. Objective: development of an experimental technique and investigation of crack resistance under biaxial loading on small-sized samples in the the brittle-ductile transition temperature range of reactor steels. Methods: on the basis of numerical and engineering calculations, a specimen was developed, the corresponding additional equipment for its tests under biaxial loading on a uniaxial servo-hydraulic machine. The influence of biaxial bending on the fracture toughness of reactor steel 15X2NMFAA was experimentally investigated. Results: the results obtained on cruciform specimen are calculated according to the ASTM 1921 standard and plotted on the Master curve (temperature dependence of fracture toughness). The analysis showed that the data on crack resistance obtained on samples with "short" (a/W=0.1...0.2) linear and surface semi-elliptical cracks exceed the upper limit of the confidence interval of the Master curve, and biaxial loading of cruciform specimen reduces crack resistance compared to uniaxial loading.

Statistical Distribution Model of Charpy Absorbed Energy in Transition Temperature Range for Reactor Pressure Vessel Steel

Abstract A statistical model to estimate the distribution characteristics of Charpy absorbed energy in the transition temperature range was proposed based on the variation in fracture toughness addressed by the Master Curve and the relationship between fracture toughness and Charpy absorbed energy associated with the Charpy Master Curve. Charpy absorbed energy in the transition temperature range can be well approximated by both the Weibull and normal distributions, and the parameters to determine the shape of the distributions (scale and shape parameters of the Weibull distribution, mean and standard deviation of the normal distribution) can be deductively defined as the functions of two independent variables, median of the fracture toughness and the difference between reference temperature and Charpy reference temperature. A series of Charpy impact tests were conducted under the same conditions to obtain the characteristics of the variation in Charpy absorbed energy for a reactor pressure vessel steel SQV2A base metal, and the experimental variation was reasonably predicted by the proposed model.

The development of prediction model on irradiation embitterment for low Cu RPV steels

The development of prediction model on irradiation embitterment (PMIE) of reactor pressure vessel (RPV) is an important method for nuclear reactor long term operation. Based on the physical mechanism of RPV irradiation embrittlement, a preliminary model is determined and the critical threshold of Cu content of 0.072% is obtained according to this preliminary model. Then a prediction model named PMIE-2020 for low Cu RPV steels is developed. At last the residual, standard deviation and predicted values and test values distribution analysis are given. Simultaneously, a comparison between PMIE-2020 and other prediction model and irradiation data is provided. Results indicate that the predicted results of PMIE-2020 has no tendency with influence factors such as neutron fluence, flux, irradiation temperature, chemical elements Cu, P, Mn, Ni, Si. The residual standard deviation is 10.76 °C, which is lower than present prediction model. The distribution between predicted values of PMIE-2020 and test values are located the area near the 45° line. These results prove that the PMIE-2020 have high accuracy on irradiation embrittlement prediction.

Mechanical properties of structural metallic alloys for nuclear applications deduced by small punch test

Small Punch Test (SPT) techniques are becoming a popular tool for the characterization of the irradiation damage effects in nuclear materials thanks to a number of advantages as compared to conventional test methods. As of today, there are two standards available (ASTM and EN) and several methods are developed to extract the tensile properties from the SPT force-displacement relationship. Here, we carry a set of benchmark tests using SPT and conventional uniaxial tensile tests on a number of metallic materials commonly applied in nuclear facilities. The investigation includes reactor pressure vessel steel, high-Cr ferritic martensitic steel, copper-chromium-zirconium alloy, nickel based super alloy, austenitic stainless steel and titanium. The SPT sample geometry and setup is designed according to the requirements of the EN10371 norm. As a result, we have deduced the tensile properties using both techniques and thus have revealed the accuracy/reliability of the currently available standards for the determination of the yield stress and ultimate tensile strength by the SPT.The study has revealed that in an overwhelming number of tests, the yield stress extracted from the SPT (for both EN and ASTM standards) is systematically overestimated as compared to the value measured from the uniaxial tensile test. However, the prediction for the ferritic and austenitic steels is close to the uniaxial stress, for both standards. In the case of Inconel material, the yield stress is underestimated. The ultimate tensile strength extracted by either of the existing methods is in a good agreement with the value otherwise measured from the uniaxial tests for all the materials except the titanium. Based on the obtained results and discussion, a number of recommendations for the evaluation of yield and ultimate tensile stress from the small punch test is given.

Small-angle neutron scattering study of neutron-irradiated and post-irradiation annealed VVER-1000 reactor pressure vessel weld material

Post-irradiation annealing of neutron-irradiated reactor pressure vessel steels is a matter of both technical and scientific interest. Small-angle neutron scattering (SANS), while being sensitive to nm-sized irradiation-induced solute-atom clusters, provides macroscopically representative and statistically reliable measures of cluster volume fraction, number density and size. In the present study, SANS was applied to uncover the size distribution of clusters in as-irradiated samples of a VVER-1000 weld and their gradual dissolution as function of the post-irradiation annealing temperature. The same samples were used to measure Vickers hardness. The results are consistent with Mn-Ni-Si-rich clusters of less than 2 nm radius to be the dominant source of both scattering and hardening. Annealing gave rise to small but significant partial recovery at 350°C and almost complete recovery at 475°C. The dispersed-barrier hardening model was applied to bridge the gap between the characteristics of nano-features and macro-hardness.

Phase Formation Features of Reactor Pressure Vessel Steels with Various Ni and Mn Content under Conditions of Neutron Irradiation at Increased Temperature

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 1022 n/m2 at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ2-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C).

Size effects on tensile properties of Chinese reactor pressure vessel steels and its semi-empirical normalization

As one of major barriers of nuclear reactors, reactor pressure vessels (RPVs) are subjected to neutron irradiation and faced with degradation during operation, therefore their mechanical properties are vital to the safety evaluation of nuclear power plants. Due to the radioactivity of specimens and limited space in the reactors, small tensile specimens are expected to be utilized for tensile property evaluation of RPV steels. However, the size effect would influence the tested tensile results and precise evaluation of mechanical properties of materials. This research is aimed to study the size effects of Chinese RPV steels in detail. It is found that size parameters of tensile specimen will affect the yield behavior, fracture behavior, tensile strength and the elongation, which causes are discussed as well. Moreover, a semi-empirical normalization model is built to normalize the total elongation after fracture of small specimens to the standard industrial specimens, which could make contributes to the normalization of results from different size specimens and serve for the design of small specimen of Chinese RPV steels in industry.

Defect Analysis of Matrix Damage in Reactor Pressure Vessel Steel Using WB-STEM

Defect Analysis of Matrix Damage in Reactor Pressure Vessel Steel Using WB-STEM

Influence analysis of alloy elements on irradiation embrittlement of RPV steel based on deep neural network

Reactor pressure vessel (RPV) is the most important core equipment in PWR. Its service life determines the service life of nuclear power plant and directly affects the economy and safety of nuclear power plant. Because RPV is serviced at high temperature, high pressure and high energy neutrons for a long time, the properties of RPV steel will significantly degrade, in which irradiation embrittlement is the most important factor for the structural integrity of RPV. In this work, about 700 groups of data such as composition, irradiation conditions and ductile brittle transition temperature of RPV steel are collected, and the data are cleaned and screened for modelling by machine learning. The deep neural network is used for establishing the correlation between key component and irradiation embrittlement of RPV steel. The results show that the lower flux of neutron irradiation will make the radiation embrittlement effect of RPV steel more obvious at the same neutron fluence. Cu, P and Ni are the key factors to influence the △DBTT of RPV steel. The synergistic effect of Cu and Ni on irradiation embrittlement is greater than that of Cu and Mn. These results will help to promote the optimization design of new RPV steel.

Statistical analysis and recovery behaviors of dislocations in cold‐rolled and annealed reactor pressure vessel steels

AbstractThe statistical analysis and recovery behavior of dislocations in cold‐rolled and annealed reactor pressure vessel steels with a thickness reduction of 50 % were investigated by using x‐ray diffraction and electron backscattered diffraction analysis in this study. A new modified Kubin & Mortensen's method was proposed to evaluate the geometrically necessary dislocation density, and this method could effectively improve the accuracy of geometrically necessary dislocation evaluation by eliminating the effect of step size and measurement noises. Comparison analysis between modified Williamson‐Hall and modified Kubin & Mortensen's methods was also performed in this study, which shows different recovery behavior of dislocations. The electron backscattered diffraction results show that geometrically necessary dislocation density remained unchangeable after annealing treatment below and at 550 °C, and it occurred to decrease above 550 °C due to the occurrence of recrystallization. However, dislocations started to migrate and annihilate at 340 °C according to x‐ray diffraction results. A possible explanation is a difference in the effect of pairwise dislocation annihilation on statistically stored dislocations and geometrically necessary dislocations.

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.

Micromechanical aspects of the effect of temperature and local plastic strain magnitude on the fracture toughness of ferrite steels

The paper shows that the influence of plastic strain and temperature on the rate of crack nuclei (CN) formation is a crucial factor controlling the shape of the temperature dependence of fracture toughness and its scatter limits. Within the framework of the microscopic model described, it is explained that the dependence of the incompatibility of microplastic deformation at grain boundaries or interfaces on the value of plastic strain and temperature is the reason for the effect of these factors on the rate of CN formation. The dependencies of the CN bulk density on temperature and plastic strain are given. In the latter case, a non-monotonic change in CN density is observed. The maximum intensity of CN formation is observed when the critical value of plastic strain is reached. For ferritic structural steels, this strain is about 2%. Using reactor pressure vessel steel and cast manganese steel as examples, it is shown that not taking these effects into account in the local fracture approach leads to considerable errors in the prediction of the temperature dependence of the fracture toughness and its scatter limits.

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 of the master curve concept (ASTM E1921) and inhomogeneity analysis of a German RPV weld for various loading rates

Fracture-mechanical investigations were carried out on the weld metal of a German reactor pressure vessel steel 22NiMoCr3-7. The crack tip loading rate and test temperature were varied in the brittle-ductile transition region, while the results were evaluated using the Master Curve methodology (ASTME1921), including inhomogeneity analyses in Appendix X5. The highly inhomogeneous datasets could be uniformly described with a bimodal distribution function with good accuracy compared to the standard procedure. However, this agreement decreases with increasing crack tip loading rate. In contrast, a modified Master Curve with a shape factor of p = 0.03/°C progressively matches the data better with increasing crack tip loading rate. The weld metal has a significantly higher toughness than the base material. SEM analysis suggests that the inhomogeneity is due to the stochastic distribution of sharp microcracks, while the observed inhomogeneity is additionally superimposed by dynamic effects (adiabatic heating and local crack arrest) at increased crack tip loading rates. It is argued that the microstructural inhomogeneity is hereby »neutralized« and could instead be interpreted as a possible artifact of the standard Master Curve under dynamic conditions.

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.