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Advances in Nuclear Reactor Pressure Vessel Steels

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A special issue of Metals (ISSN 2075-4701). This special issue belongs to the section "Metallic Functional Materials".

Deadline for manuscript submissions: 10 May 2024 | Viewed by 15894

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Special Issue Editor

Dr. Ferenc Gillemot

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Centre for Energy Research, H-1121 Budapest, 29-33 Konkoly Thege Miklos ut, Hungary
Interests: fracture toughness analysis; fracture mechanics; steel; radiation; nuclear reactors; nuclear power plants; mechanical testing

Special Issue Information

Dear Colleagues,

Nuclear Reactor Pressure Vessel Steels are one of the most sophisticated products of the steel industry. Low alloyed, low carbon ferrite perlite steels used for building pressure vessels. Two basic types of them exists: Cr-Mo-Ni and Cr-Mo-V alloyed steels, one subgroup of the latter also contains nickel. The quantity of the alloying elements is in the range of 2–5%. One difficulty in the production of the vessel element is the large size. Another is the strong limits of the polluting elements, and the welding together of them requires special technology. Finally, the insides of the vessels are clad with welded stainless-steel layers. The first generation of the vessels made from plates, but since about 1980, the vessels have been built only from forged rings to avoid vertical welds.

During their long lifetimes (nowadays 60–80 years are the requirement) they are exposed to high neutron and gamma radiation at elevated temperatures, low cycle fatigue, and corrosion. They must maintain the required safety properties (first of all, fracture toughness) during the whole service life. The mechanical properties of thick forged rings are changing in the function of the distance from the surface, since the cooling rate at quenching is much slower at the middle section than at the surface. The main environmental factor determining the safe lifetime is neutron radiation and it contributed with thermal embrittlement, low cycle fatigue, and sometimes with corrosion. Neutron radiation causes the strong embrittlement of the steels. Around 1990, it was discovered that Cu and P pollution increases radiation sensitivity. Few pollution elements can be highly activated and make maintenance difficult. Presently, It is discovered that Ni and Mn also can play major role in ageing. Little information exists on the effect of production technology (e.g., grain size).

Advanced nuclear pressure vessel steel production includes the development of the production technology and material science and aging assessment. Papers on the development of new type or further developed steels for the present and future generation of pressure vessels are welcomed. They can include the effect of alloying elements on radiation toughness, production development, studies enhancing long-term operation, service degradation mitigation, and embrittlement trend curves.

Dr. Ferenc Gillemot
Guest Editor

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Keywords

present and future vessel steel production
effect of alloying and polluting element
effect of production technology
toughness distribution
microstructure and ageing
vessel steel testing methods
welding and cladding
high temperature vessels
mitigation of ageing
trend curves

Published Papers (9 papers)

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Research

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24 pages, 2422 KiB

Open AccessArticle

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

by Libang Lai, Jann-Erik Brandenburg, Paul Chekhonin, Arnaud Duplessi, Fabien Cuvilly, Auriane Etienne, Bertrand Radiguet, David Rafaja and Frank Bergner

Metals 2024, 14(3), 257; https://doi.org/10.3390/met14030257 - 21 Feb 2024

Viewed by 706

Abstract

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. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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23 pages, 15891 KiB

Open AccessArticle

Microstructural Characterization of Reactor Pressure Vessel Steels

by Libang Lai, Paul Chekhonin, Shavkat Akhmadaliev, Jann-Erik Brandenburg and Frank Bergner

Metals 2023, 13(8), 1339; https://doi.org/10.3390/met13081339 - 26 Jul 2023

Cited by 2 | Viewed by 1272

Abstract

Ion irradiation is a promising tool to emulate neutron-irradiation effects on reactor pressure vessel (RPV) steels, especially in the situation of limited availability of suitable neutron-irradiated material. This approach requires the consideration of ion-neutron transferability issues, which are addressed in the present study by comparing the effect of ions with neutron-irradiation effects reported for the same materials. The first part of the study covers a comprehensive characterization, based on dedicated electron microscopy techniques, of the selected unirradiated RPV materials, namely a base metal and a weld. The results obtained for the grain size, dislocation density, and precipitates are put in context in terms of hardening contributions and sink strength. The second part is focused on the depth-dependent characterization of the dislocation loops formed in ion-irradiated samples. This work is based on scanning transmission electron microscopy applied to cross-sectional samples prepared by the focused ion beam technique. A band-like arrangement of loops is observed in the depth range close to the peak of injected interstitials. Two levels of displacement damage, 0.1 and 1 dpa (displacements per atom), as well as post-irradiation annealed conditions, are included for both RPV materials. Compared with neutron irradiation, ion irradiation creates a similar average size but a higher number density of loops presumably due to the higher dose rate during ion irradiation. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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15 pages, 1263 KiB

Open AccessArticle

Electroslag Hollow Ingots for Nuclear and Petrochemical Pressure Vessels and Pipes

by Lev Medovar, Ganna Stovpchenko, Artem Sybir, Jianjun Gao, Liguo Ren and Dmytro Kolomiets

Metals 2023, 13(7), 1290; https://doi.org/10.3390/met13071290 - 18 Jul 2023

Viewed by 1026

Abstract

The paper presents ground reasoning and results of experiments and modeling of heavy hollow ingot manufacturing using advanced electroslag technology. The requirements for ingots for huge diameter reactor pressure vessels include high density, homogeneity, and minimal segregation, which are very difficult to achieve by traditional casting. In the electroslag remelting process (ESR), hollow ingots form in between two copper water-cooled molds under effective heat removal. This improves the solidification pattern due to the shortening of a solidifying volume thickness more than twice compared with a solid ingot of the same diameter. The shallow liquid metal pool and narrow mushy zone at the ESR hollow ingot solidification assure their high metallurgical quality. Due to the dense and low segregation structure, ESR hollow ingots proved to be used for as-cast pipes and heavy wall billets for further forging. The results of a mathematical simulation within the range of simulated dimensions (the outer diameter up to 2900 mm, wall thickness up to 750 mm) also predict the favorable solidification pattern for thick-wall hollow ingots of big diameters. The analysis made and the modeling results provide a framework for scaling up the sizes of hollow ingots produced by ESR and widening their application for manufacturing heavy wall large diameter shells for nuclear and petrochemical industries. The higher reachable productivity of hollow ingot formation and lower capacity of power supply source than that for solid ingots of the same diameter and weight are also preconditions of their energy saving and cost-effective manufacturing. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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26 pages, 12081 KiB

Open AccessArticle

Ductile Fracture Behavior of ASTM A516 Gr.70 Pressure Vessel Steel by ASTM and ISO Fracture Toughness Standards

by Gabriel de Castro Coêlho, Antonio Almeida Silva, Marco Antonio dos Santos, José J. M. Machado and João Manuel R. S. Tavares

Metals 2023, 13(5), 867; https://doi.org/10.3390/met13050867 - 29 Apr 2023

Cited by 1 | Viewed by 1794

Abstract

Fracture toughness determination is crucial for the design phase of pressure vessels, and, although ASTM E1820 and ISO 12135 fracture toughness standards have existed for some time, some differences have been reported in the determination of this property. This study investigates the ductile fracture behavior of ASTM A516 Gr.70 pressure vessel steel and assesses the differences in estimating both standards. The steel’s tensile properties and initiation fracture toughness (J_IC) were evaluated, taking into account the parallel and perpendicular orientations to the rolling direction. The results reveal the properties’ dependence on the rolling direction, mainly attributed to perlite banding. Additionally, as for the J_IC determination, the differences were associated with the different blunting line slope estimations on each standard, reinforcing the necessity of a work-hardening-based blunting line for each material assessed. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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12 pages, 3237 KiB

Open AccessArticle

Effectiveness of Thermal Annealing in Recovery of Tensile Properties of Compositionally Tailored PWR Model Steels Irradiated in LYRA-10

by Mathilde Laot, Kiki Naziris, Theo Bakker, Elio D’Agata, Oliver Martin and Murthy Kolluri

Metals 2022, 12(6), 904; https://doi.org/10.3390/met12060904 - 25 May 2022

Cited by 3 | Viewed by 1344

Abstract

Understanding the mechanical behaviour of reactor pressure vessel (RPV) steels at high fluences has become an important topic in regard to Long-Term Operations (LTO) of existing nuclear power plants (NPP). The effectiveness of thermal annealing treatments to recover the mechanical properties of compositionally tailored pressurised water reactor (PWR) model steels irradiated to high neutron fluences, up to 1.22 × 10²⁰ n·cm⁻², is analysed in this study. Tensile testing of four different PWR RPV steels was performed after irradiation and subsequent recovery annealing treatment at 450 °C for 40 h. Irradiation-induced hardening and the effectiveness of recovery thermal annealing have been assessed by comparing the strength and ductility properties of irradiated and irradiated and subsequently annealed samples with unirradiated reference samples for all four model steel. The annealing treatment resulted in a significant recovery of the yield strength (~75–89%) and the ultimate tensile strength (~78–96%) of all four PWR model steels. This study proves that substantial irradiation-induced hardening (up to ~389 MPa) observed in steels containing high Ni and Mn contents can still be recovered using the thermal annealing treatment. No influence of annealing on ductility properties has been observed for all four model steels. Microscopy analyses of these steels to understand the underlying irradiation damage and recovery mechanisms are planned for the near future. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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13 pages, 1779 KiB

Open AccessArticle

Assessment of the Generalization Ability of the ASTM E900-15 Embrittlement Trend Curve by Means of Monte Carlo Cross-Validation

by Diego Ferreño, Mark Kirk, Marta Serrano and José A. Sainz-Aja

Metals 2022, 12(3), 481; https://doi.org/10.3390/met12030481 - 12 Mar 2022

Cited by 4 | Viewed by 1482

Abstract

The standard ASTM E900-15 provides an analytical expression to determine the transition temperature shift exhibited by Charpy V-notch data at 41-J for irradiated pressure vessel materials as a function of the variables copper, nickel, phosphorus, manganese, irradiation temperature, neutron fluence, and product form. The 26 free parameters included in this embrittlement correlation were fitted through maximum likelihood estimation using the PLOTTER—BASELINE database, which contains 1878 observations from commercial power reactors. The complexity of this model, derived from its high number of free parameters, invites a consideration of the possible existence of overfitting. The undeniable goal of a good predictive model is to generalize well from the training data that was used to fit its free parameters to new data from the problem domain. Overfitting takes place when a model, due to its high complexity, is able to learn not only the signal but also the noise in the training data to the extent that it negatively impacts the performance of the model on new data. This paper proposes the resampling method of Monte Carlo cross-validation to estimate the putative overfitting level of the ASTM E900-15 predictive model. This methodology is general and can be employed with any predictive model. After 5000 iterations of Monte Carlo cross-validation, large training and test datasets (7,035,000 and 2,355,000 instances, respectively) were obtained and compared to measure the amount of overfitting. A slightly lower prediction capacity was observed in the test set, both in terms of R² (0.871 vs. 0.877 in the train set) and the RMSE (13.53 °C vs. 13.22 °C in the train set). Besides, strong statistically significant differences, which contrast with the subtle differences observed in R² and RMSE, were obtained both between the means and the variances of the training and test sets. This result, which may seem paradoxical, can be properly interpreted from a correct understanding of the meaning of the p-value in practical terms. In conclusion, the ASTM E900-15 embrittlement trend curve possess good generalization ability and experiences a limited amount of overfitting. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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16 pages, 6223 KiB

Open AccessArticle

Effect of Neutron Flux on an Irradiation-Induced Microstructure and Hardening of Reactor Pressure Vessel Steels

by Andreas Ulbricht, Mercedes Hernández-Mayoral, Elvira Oñorbe, Auriane Etienne, Bertrand Radiguet, Eric Hirschmann, Andreas Wagner, Hieronymus Hein and Frank Bergner

Metals 2022, 12(3), 369; https://doi.org/10.3390/met12030369 - 22 Feb 2022

Cited by 5 | Viewed by 1946

Abstract

The existing knowledge about the effect of neutron irradiation on the mechanical properties of reactor pressure vessel steels under reactor service conditions relies to a large extent on accelerated irradiations realized by exposing steel samples to a higher neutron flux. A deep understanding of flux effects is, therefore, vital for gaining service-relevant insight into the mechanical property degradation. The existing studies on flux effects often suffer from incomplete descriptions of the irradiation-induced microstructure. Our study aims to give a detailed picture of irradiation-induced nanofeatures by applying complementary methods using atom probe tomography, positron annihilation, small-angle neutron scattering and transmission electron microscopy. The characteristics of the irradiation-induced nanofeatures and the dominant factors responsible for the observed increase of Vickers hardness are identified. Microstructural changes due to high flux conditions are smaller nm-sized solute atom clusters with almost the same volume fraction and a higher concentration of vacancies and sub-nm vacancy clusters compared to low flux conditions. The results rationalize why pronounced flux effects on the nanofeatures, in particular on solute atom clusters, only give rise to small or moderate flux effects on hardening. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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24 pages, 44545 KiB

Open AccessArticle

Prediction of the Transition-Temperature Shift Using Machine Learning Algorithms and the Plotter Database

by Diego Ferreño, Marta Serrano, Mark Kirk and José A. Sainz-Aja

Metals 2022, 12(2), 186; https://doi.org/10.3390/met12020186 - 19 Jan 2022

Cited by 10 | Viewed by 1805

Abstract

The long-term operating strategy of nuclear plants must ensure the integrity of the vessel, which is subjected to neutron irradiation, causing its embrittlement over time. Embrittlement trend curves used to predict the dependence of the Charpy transition-temperature shift, ΔT_41J, with neutron fluence, such as the one adopted in ASTM E900-15, are empirical or semi-empirical formulas based on parameters that characterize irradiation conditions (neutron fluence, flux and temperature), the chemical composition of the steel (copper, nickel, phosphorus and manganese), and the product type (plates, forgings, welds, or so-called standard reference materials (SRMs)). The ASTM (American Society for Testing and Materials) E900-15 trend curve was obtained as a combination of physical and phenomenological models with free parameters fitted using the available surveillance data from nuclear power plants. These data, collected to support ASTM’s E900 effort, open the way to an alternative, purely data-driven approach using machine learning algorithms. In this study, the ASTM PLOTTER database that was used to inform the ASTM E900-15 fit has been employed to train and validate a number of machine learning regression models (multilinear, k-nearest neighbors, decision trees, support vector machines, random forest, AdaBoost, gradient boosting, XGB, and multi-layer perceptron). Optimal results were obtained with gradient boosting, which provided a value of R² = 0.91 and a root mean squared error ≈10.5 °C for the test dataset. These results outperform the prediction ability of existing trend curves, including ASTM E900-15, reducing the prediction uncertainty by ≈20%. In addition, impurity-based and permutation-based feature importance algorithms were used to identify the variables that most influence ΔT_41J (copper, fluence, nickel and temperature, in this order), and individual conditional expectation and interaction plots were used to estimate the specific influence of each of the features. Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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Review

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10 pages, 1902 KiB

Open AccessReview

Review on Steel Enhancement for Nuclear RPVs

by Ferenc Gillemot

Metals 2021, 11(12), 2008; https://doi.org/10.3390/met11122008 - 13 Dec 2021

Cited by 6 | Viewed by 2113

Abstract

The reactor pressure vessel (RPV) is one of the most important elements of a nuclear power plant (NPP). The RPV determines the plant operational lifetime since it is not replaceable economically. The purpose of the RPV steel study and enhancement to increase the NPP’s (Nuclear Power Plants) operation lifetime from the original 30–40 years up to 60–80 years or even beyond. The RPV lifetime limited by ageing of the RPV steels. RPV ageing highly depends on the main environmental effects: fast neutron radiation, thermal effects causing thermal ageing and low-cycle fatigue. Firstly, the chemical composition via aged mechanical properties was studied. Efforts to increase the toughness against the radiation embrittlement was enhanced by the appearance of the modern microstructural testing devices such as APFIM (atom probe field ion microscopy), SANS (small-angle neutron scattering) positron annihilation spectroscopy (PAS), transmission electron microscopy (TEM) and Mössbauer spectroscopy (MS). The information on the effect of alloying and polluting elements for the microstructure allowed us to produce increased ageing toughness of the RPVs, and to enhance the safety and lifetime calculations of them, supporting long-term safe operation (LTO). Full article

(This article belongs to the Special Issue Advances in Nuclear Reactor Pressure Vessel Steels)

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