Nuclear Pressure Vessel Steels Research Articles

Study on hot deformation behavior and processing maps of SA508-ІV steel for novel nuclear reactor pressure vessels

The hot deformation behavior and microstructural evolution of SA508-ІV steel for a novel nuclear reactor pressure vessels were studied by hot compression tests in the strain rates 0.001–1 s−1 and temperature ranges 950–1250 °C respectively. The results showed that the high temperature (1250 °C) and low strain rate (0.001 s−1) were more likely to induce the abnormal growth of grains, resulting in the decrease of the power dissipation efficiency. The flow instability region covered the temperature and strain rate ranges in 950–1200 °C and 0.1–1 s−1 with inhomogeneous microstructure. The optimal processing window of SA508-ІV steel was obtained in the temperature ranges 1050–1175 °C and strain rates 0.01–0.1 s−1 respectively with a peak efficiency of 45%. By analyzing the microstructure of the steel in this zone mentioned above, fine recrystallized grains were got.

The action mechanism and application of alloying elements in the pressure vessel steel

The research progress of the influence of alloying elements on the microstructure and mechanical properties of pressure vessel steel was summarized. The effects of different alloying elements on the composite mechanical properties of the pressure vessel, such as the strength and toughness and the low temperature fracture resistance, were analysed. The action mechanism of alloying elements in pressure vessel steel was generalized. Combined with the actual production, the adding alloying elements, the research progress and the development prospect of the nuclear power pressure vessel steel were mainly analysed and discussed. The future research should focus on the mechanism of each component element, clarify the interaction relationship between the alloying elements, and develop a complete set of alloy adding scheme. So that the pressure vessel steel, which not only conforms to the actual situation, but also meet the specific requirements should be produced.

Prediction of cleavage crack propagation path in a nuclear pressure vessel steel

A propagation criterion based on maximum principal stress is proposed to predict the cleavage crack propagation in a nuclear vessel steel. Experiments are performed on three specimen geometries: standard Compact Tensile, precracked ring (mixed mode) and Extended Compact Tensile. Their crack paths are respectively directionally stable and straight, directionally stable and curved, directionally unstable and deflected. Numerical computations are performed by eXtended Finite Element Method. The propagation criterion, combined with a deterministic direction criterion based on the maximum hoop stress provides good predictions for directionally stable crack paths whereas predicting directionally unstable crack paths requires a probabilistic direction criterion.

A multi-scale correlative investigation of ductile fracture

The use of novel multi-scale correlative methods, which involve the coordinated characterisation of matter across a range of length scales, are becoming of increasing value to materials scientists. Here, we describe for the first time how a multi-scale correlative approach can be used to investigate the nature of ductile fracture in metals. Specimens of a nuclear pressure vessel steel, SA508 Grade 3, are examined following ductile fracture using medium and high-resolution 3D X-ray computed tomography (CT) analyses, and a site-specific analysis using a dual beam plasma focused ion beam scanning electron microscope (PFIB-SEM). The methods are employed sequentially to characterise damage by void nucleation and growth in one volume of interest, allowing for the imaging of voids that ranged in size from less than 100 nm to over 100 μm. This enables the examination of voids initiated at carbide particles to be detected, as well as the large voids initiated at inclusions. We demonstrate that this multi-scale correlative approach is a powerful tool, which not only enhances our understanding of ductile failure through detailed characterisation of microstructure, but also provides quantitative information about the size, volume fractions and spatial distributions of voids that can be used to inform models of failure. It is found that the vast majority of large voids nucleated at MnS inclusions, and that the volume of a void varied according to the volume of its initiating inclusion raised to the power 3/2. The most severe voiding was concentrated within 500 μm of the fracture surface, but measurable damage was found to extend to a depth of at least 3 mm. Microvoids associated with carbides (carbide-initiated voids) were found to be concentrated around larger inclusion-initiated voids at depths of at least 400 μm. Methods for quantifying X-ray CT void data are discussed, and a procedure for using this data to calibrate parameters in the Gurson-Tvergaard Needleman (GTN) model for ductile failure is also introduced.

Formation and recovery of Cu precipitates in Fe–Cu model alloys under varying heat treatment

Fe–Cu systems, model alloys of nuclear reactor pressure vessel (RPV) steels, heat treatments by annealing and quenching, are studied by the positron annihilation technique and high‐resolution SEM. The different effect of annealing and quenching on the formation of Cu precipitates was discussed. The ultrafine Cu precipitates, which are formed in Fe–x wt.% Cu (x = 1.0 and 1.5) alloys after thermal annealing, are observed clearly and no obvious Cu precipitates were found in the quenched alloys. The annealing temperature has an effect on the concentration of Cu precipitates, which leads to a change of the proportion of positron annihilation in the Cu precipitates, and this can be reflected by the change of the W parameters. Positron annihilation Doppler broadening spectroscopy results show that Cu precipitates formed during annealing at 1173 K still exist when the annealing temperature is higher than 1453 K. SEM measurements were used to observe the size and the distribution of Cu precipitates in the samples, after using different annealing temperature.

Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

Abstract SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t 8/5 (the HAZ cooling time from 800 °C to 500 °C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 °C or higher is recommended.

Atomic-scale simulation study of structural changes of Fe-Cu binary system containing Cu clusters embedded in the Fe matrix during heating

Nano-size Cu precipitates are the main products of irradiation embrittlement of nuclear reactor pressure vessel steels. Molecular dynamics simulation within the framework of embedded atom method is performed to study atomic packing change in Fe-Cu binary system, where the small Cu clusters are embedded in the crystal body centered cubic (BCC) Fe lattices. As the temperature increases, atomic packing change occurs in the Fe-Cu binary system. The mean square displacement of Cu atom, pair distribution function of the Cu atoms, and the atomic density profile along the radial direction are calculated. The atom packing structures in pure Cu region, Fe-Cu interface region, and pure Fe matrix are analyzed. The simulation results show that the packing structures in the Cu cluster and the Fe matrix are greatly affected by the sizes of these clusters and the volume of the Fe matrix containing these clusters. The structural changes present apparent differences, for the Fe matrixes contain these confined Cu clusters with different atom numbers during heating. As the Fe matrix can only provide small space to accommodate the Cu atoms, packing patterns in many Cu atoms are disordered for the Febulk-Cu135 system. In this binary system, strain region in the Fe matrix is adjacent to the Cu cluster. In the meantime, there are a lot of vacancy defects and strain regions in the matrix. For the Febulk-Cu141 system, although the Cu cluster contains more atoms, the Fe matrix can accommodate Cu atoms in a larger space, and the majority of these Cu atoms are located at the BCC crystal lattices. With increasing the temperature, the changes can be observed that the number of the strain regions decrease, whereas the sizes of some strain regions increase.

Hot Deformation Behavior of SA508GR. 4N Steel for Nuclear Reactor Pressure Vessels

A hot compression experiment (1073–1473 K, strain rates of 0.001–10s−1) of SA508GR. 4N low alloy steel was performed using a Gleeble-3800 thermal-mechanical simulator, and the hot deformation behavior of the steel was investigated by analyzing both the true stress-true strain curves and its microstructures. The thermal deformation equation and hot deformation activation energy (Q) of SA508GR. 4N steel were obtained by regression with a classic hyperbolic sine function. The hot processing map of SA508GR. 4N steel was also established. An empirical equation for the stress peak was described for practical applications. The SA508GR. 4N steel showed a critical Zener-Hollomon parameter (lnZc) for dynamic recrystallization (DRX) of 37.44, below which full DRX may occur. The sensitivity of the SA508GR.4N steel increased linearly with test temperature, such that higher temperatures led to enhanced workability.

Quantifying the metallurgical response of a nuclear steel to welding thermal cycles

Solid-state phase transformations can drastically influence the evolution of stress in welds due to the strains associated with the transformations and related changes in mechanical properties. As such, finite-element predictions of welding residual stresses need reliable materials data including, where applicable, information on phase transformation kinetics and phase- and temperature-dependent material properties. Owing to a scarcity of such data, many authors have used uncalibrated empirical modelling approaches for the prediction of welding residual stresses. This paper addresses this critical shortage for an important nuclear pressure vessel (SA508) steel. Austenite formation, grain growth and decomposition data are presented and subsequently used to calibrate transformation models. These models are shown to accurately predict microstructure and residual stresses for experimental test cases.This paper is part of a Themed Issue on Measurement, modelling and mitigation of residual stress.

A statistical model of cleavage fracture in structural steels with power-law distribution of microcrack size

Based on the well-accepted physical understanding that cleavage fracture of structural steels is preceded by plastic deformation, a critical flaw in the Beremin model and its modifications is proposed, and a new statistical model with the power-law distribution of microcracks is obtained to describe the cumulative probability of cleavage fracture. A set of cleavage fracture toughness data of a nuclear pressure vessel steel is used to highlight the difference between the new model and the Beremin model.

Determination of Fatigue Properties Using Miniaturized Specimens

It is very important to determine Paris’ power law constants to know the fatigue crack growth rate (FCGR) of a structure containing crack. These constants are generally determined by FCGR test using conventional specimens as specified by ASTM-E647 standard. This standard FCGR test procedure has placed strict size requirements on the conventional specimens and hence a large amount of material is needed to fabricate the specimens. However, sometimes, like in the case of testing of irradiated materials or nuclear pressure vessel surveillance, the material available for testing is not sufficient to make standard size specimens. Therefore, miniature specimens should be used to minimize the amount of material. In this work, constant load amplitude FCGR behavior of nuclear pressure vessel steel (20MnMoNi55) has been studied using miniature Single Edge Notched Tensile (SENT) specimens. The fatigue tests have been performed according to the procedure described in ASTM-E647 standard. The Paris’ power law coefficient and exponent obtained by FCGR test of miniature SENT specimens are in the range of 10−12 to 10−13 and 2.8–3.5 respectively, when the FCGR (da/dN) is expressed in m/cycle and stress intensity range (ΔK) in MPa√m. These values of Paris’ power law constants of miniature specimens are nearly of same order as that of the standard specimens where the results of standard specimens are taken from the literature.

Modeling of Austenite Grain Growth During Austenitization in a Low Alloy Steel

The main purpose of this work is to develop a pragmatic model to predict austenite grain growth in a nuclear reactor pressure vessel steel. Austenite grain growth kinetics has been investigated under different heating conditions, involving heating temperature, holding time, as well as heating rate. Based on the experimental results, the mathematical model was established by regression analysis. The model predictions present a good agreement with the experimental data. Meanwhile, grain boundary precipitates and pinning effects on grain growth were studied by transmission electron microscopy. It is found that with the increasing of the temperature, the second-phase particles tend to be dissolved and the pinning effects become smaller, which results in a rapid growth of certain large grains with favorable orientation. The results from this study provide the basis for the establishment of large-sized ingot heating specification for SA508-III steel.

Thirty-five years in environmentally assisted cracking in Italy: a point of view

Abstract In this paper, I will attempt to summarize my experience of approximately 35 years on environmental fatigue damage, which I gained from industries outside Milan, Italy, and as a visiting scientist at the Naval Research Laboratory in Washington, DC. I will present a brief summary of my understanding of the topic of environmentally assisted fatigue of metals. About halfway in my career, I moved from the nuclear industry towards the oil and gas industry. This shift allowed me to understand that environmentally assisted cracking, which is an important topic for pipeline steels, is also similar to the nuclear pressure vessel steel community and that the problems are similar. At the same time, I became aware that the scientific communities are not working well to communicate the issues and challenges. People working in one industrial sector do not seem to appreciate the contribution from another industrial sector. This is understandable, as there are industrial rules and regulations. Because of such constraints, a knowledge gap is introduced into the technical community. This lack of cooperation slows down the progress of relevant research, especially nowadays, when we are facing increasingly complex service situations that may lead to failures. While trying to cope with such difficulties, I became interested in knowledge management. I learned that there is a need to build bridges between contemporary science, engineering, and business. We are facing big challenges in analyzing failures, and maybe it is time to do things differently and consolidate different disciplines of engineering with applications that can give us more reliable predictions. Organizations and companies worldwide are currently facing (a) increasing quantity of available information on material usage, (b) increasing number of people with high education in science and engineering, (c) increasing complexity of the failure problems to be solved, and (d) increasing lack of communication between institutions. It is estimated that the material loss in the US due to corrosion alone is approximately $300 billion per year. The necessity of knowing, understanding, and processing such complex data and problems is a very large task. Thus, there is a need for exploiting the technical information and using communication technology tools to help support teamwork, networking, using available experience and expertise, and, hopefully, international cooperation. Knowledge management involves a cultural change to stimulate a better use of all types of resources. Progress will be slow and difficult if we do not learn how to cope with this new situation.

Stretch zone formation in cyclic fracture of 20MnMoNi55 pressure vessel steel

Stretch zone formation during cyclic fracture in a nuclear pressure vessel steel has been examined. Cyclic fracture tests are carried out at different load ratios and incremental plastic displacements. Stretch zone features were manifested in local regions, however, their extent could be correlated to the test parameters employed. The mean stretch zone width decreased with R ratio and increased with incremental plastic displacement. Stretch zone dimension could be empirically correlated to a cyclic stretching coefficient proposed in this investigation. A comparison of cyclic fracture toughness obtained from stretch zone width and that measured experimentally was made, and the systematic difference was explained.

Correlations between Charpy V-notch impact energy and fracture toughness of nuclear reactor pressure vessel (RPV) steels

Correlations between Charpy V-notch (CVN) impact energy and fracture toughness are of great practical significance for reactor pressure vessel (RPV) structural integrity assessment. In this paper, correlations between the two for three commonly used RPV steels, namely Chinese A508-3 steel, USA A533B steel, Euro 20MnMoNi55 steel, are investigated with both a direct conversion and the Master Curve method. It is found that fracture toughness can be well predicted from CVN impact energy through reference temperature T0 of the Master Curve method. Five predicted T0 values from different correlations are close to the T0 from actual test. The mean value of the five T0 values provides a good estimate for the fracture toughness in the transition region.

Comparison of fracture properties in SA508 Gr.3 and Gr.4N high strength low alloy steels for advanced pressure vessel materials

Nuclear power systems are moving to a larger capacity or smaller modular type. In any either case, advanced pressure vessel materials with high strength and toughness are definitely needed for an optimization of the design and construction, as well as the long-term operation. In this paper, two candidate materials, both of which are within the current ASME specifications of SA508 steel forging, are compared from the view point of fracture resistance properties for a nuclear pressure vessel steel. The microstructure and mechanical properties of SA508 Gr.3 Cl.1, Cl.2, and Gr.4N steels were also characterized. The predominant microstructure of SA508 Gr.4N model alloy is tempered martensite, while SA508 Gr.3 Cl.1 and Cl.2 steels show a tempered upper bainitic structure. SA508 Gr. 4N model alloy showed the best strength and transition behavior among the three types of SA508 steel. SA508 Gr.3 Cl.2 steel has good strength and fracture toughness, but there is a decrease in the upper-self energy. The fracture resistance and fatigue crack growth rate of SA508 Gr.3 Cl.2 and Gr.4N steels were comparable to those of SA508 Gr.3 Cl.1 steel. In terms of mechanical properties, SA508 Gr.4N steel is a fascinating material for the pressure vessel application although it still needs verification on the aging behavior such as the irradiation embrittlement resistance.

Microstructure and mechanical characteristics of a laser welded joint in SA508 nuclear pressure vessel steel

Abstract SA508 steels are typically used in civil nuclear reactors for critical components such as the reactor pressure vessel. Nuclear components are commonly joined using arc welding processes, but with design lives for prospective new build projects exceeding 60 years, new welding technologies are being sought. In this exploratory study, for the first time, autogenous laser welding was carried out on 6 mm thick SA508 Cl.3 steel sheets using a 16 kW fiber laser system operating at a power of 4 kW. The microstructure and mechanical properties (including microhardness, tensile strength, elongation, and Charpy impact toughness) were characterized and the microstructures were compared with those produced through arc welding. A three-dimensional transient model based on a moving volumetric heat source model was also developed to simulate the laser welding thermal cycles in order to estimate the cooling rates included by the process. Preliminary results suggest that the laser welding process can produce welds that are free of macroscopic defects, while the strength and toughness of the laser welded joint in this study matched the values that were obtained for the parent material in the as-welded condition.

IRRADIATION EMBRITTLEMENT MECHANISMS AND RELEVANT INFLUENCE FACTORS OF NUCLEAR REACTOR PRESSURE VESSEL STEELS

Nuclear reactor pressure vessel is the irreplaceable component of the nuclear power plant and its integrity is one of the key issues of any nuclear power plant for long term operations. Various nanofeatures, includ- ing solute clusters, matrix damage and grain boundary segregation formed in reactor pressure vessel steels in the face of neutron irradiation. These ultrafine microstructural features lead to an increase in the ductile brittle transi- tion temperature as is the measure used to describe the irradiation embrittlement. The balance of features depends on the composition of the reactor pressure vessel steels and the irradiation conditions. This paper reviews the cur- rent phenomenological knowledge and understanding of the basic mechanisms and relevant influence factors for ir- radiation embrittlement of nuclear reactor pressure vessel steels. To be specific, the formation and evolution pro- cesses of the embrittling features are presented. Also, the influences of material variables, such as copper, nickel and manganese contents on irradiation embrittlement and those of irradiation variables, such as neutron flux and post irradiation annealing are summarized. In addition, fundamental research issues that remain to be addressed are briefly pointed out.

Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σF, of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors.

Continuous Cooling Transformations in Nuclear Pressure Vessel Steels

A class of low-alloy steels often referred to as SA508 represent key materials for the manufacture of nuclear reactor pressure vessels. The alloys have good properties, but the scatter in properties is of prime interest in safe design. Such scatter can arise from microstructural variations but most studies conclude that large components made from such steels are, following heat treatment, fully bainitic. In the present work, we demonstrate with the help of a variety of experimental techniques that the microstructures of three SA508 Gr.3 alloys are far from homogeneous when considered in the context of the cooling rates encountered in practice. In particular, allotriomorphic ferrite that is expected to lead to a deterioration in toughness, is found in the microstructure for realistic combinations of austenite grain size and the cooling rate combination. Parameters are established to identify the domains in which SA508 Gr.3 steels transform only into the fine bainitic microstructures.