Mechanical Behavior of Reactor Structural Materials

A special issue of Metals (ISSN 2075-4701).

Deadline for manuscript submissions: closed (31 March 2023) | Viewed by 4528

Special Issue Editors


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Guest Editor
Idaho National Laboratory, Idaho Falls, ID 83402, USA
Interests: mechanical behavior; radiation effects; electron microscopy

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Guest Editor
Idaho National Laboratory, Idaho Falls, ID 83402, USA
Interests: microscopy; mechanical properties; irradiation damage

Special Issue Information

The world faces a critical challenge in terms of meeting the ever-increasing global energy demand for economic prosperity. Meanwhile, we must mitigate the effects of climate change by increasing the share of clean energy production. Thus, nuclear energy has become an essential option for producing carbon-free energy with a potential capacity to ensure global energy security. To produce sustainable, economical, safe, and proliferation-resistant nuclear energy, the life-time extension of current reactor fleets and successful design and deployment of advanced nuclear reactor systems are critical. To this end, we need to identify and develop structural materials that can withstand harsh conditions in the reactor core, such as stress, high temperature, corrosion medium, and radiation. In contrast to present-generation light water reactors (LWR), advanced reactors operate at much higher temperatures (500–1000 °C), produce heavier irradiation damage on materials (up to 200 displacements per atom, dpa), and have extreme corrosive environments. Such conditions alter the physical and chemical characteristics of materials that deteriorate their performance, leading to unwarranted reactor shutdowns, low efficiency, and safety issues. Among the various material selection and design criteria for reactor structural materials, their mechanical behavior under harsh conditions stands out as a critical factor that can affect the integrity of reactor structural components. The rector structural materials are subjected to a combination of embrittlement, fatigue, creep, and stress corrosion cracking degradation modes aggravated by severe radiation damage. This Special Issue aims to disseminate research articles focusing on the latest developments in the mechanical behavior of the reactor structural materials field.

Dr. Boopathy Kombaiah
Dr. Colin Judge
Guest Editors

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Keywords

  • irradiation
  • mechanical behavior
  • deformation mechanisms
  • reactor structural materials
  • microstructure

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Published Papers (2 papers)

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Research

14 pages, 3773 KiB  
Article
Micropillar Compression of Additively Manufactured 316L Stainless Steels after 2 MeV Proton Irradiation: A Comparison Study between Planar and Cross-Sectional Micropillars
by Ching-Heng Shiau, Miguel Pena, Yongchang Li, Sisi Xiang, Cheng Sun, Michael D. McMurtrey and Lin Shao
Metals 2022, 12(11), 1843; https://doi.org/10.3390/met12111843 - 28 Oct 2022
Cited by 1 | Viewed by 1578
Abstract
A micropillar compression study with two different techniques was performed on proton-irradiated additively manufactured (AM) 316L stainless steels. The sample was irradiated at 360 °C using 2 MeV protons to 1.8 average displacement per atom (dpa) in the near-surface region. A comparison study [...] Read more.
A micropillar compression study with two different techniques was performed on proton-irradiated additively manufactured (AM) 316L stainless steels. The sample was irradiated at 360 °C using 2 MeV protons to 1.8 average displacement per atom (dpa) in the near-surface region. A comparison study with mechanical test and microstructure characterization was made between planar and cross-sectional pillars prepared from the irradiated surface. While a 2 MeV proton irradiation creates a relatively flat damage zone up to 12 µm, the dpa gradient by a factor of 2 leads to significant dpa uncertainty along the pillar height direction for the conventional planar technique. Cross-sectional pillars can significantly reduce such dpa uncertainty. From one single sample, three cross-sectional pillars were able to show dpa-dependent hardening. Furthermore, post-compression transmission electron microscopy allows the determination of the deformation mechanism of individual micropillars. Cross-sectional micropillar compression can be used to study radiation-induced mechanical property changes with better resolution and less data fluctuation. Full article
(This article belongs to the Special Issue Mechanical Behavior of Reactor Structural Materials)
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36 pages, 5909 KiB  
Article
Swelling and He-Embrittlement of Austenitic Stainless Steels and Ni-Alloys in Nuclear Reactors
by Malcolm Griffiths, Steven Xu and Juan Eduardo Ramos Nervi
Metals 2022, 12(10), 1692; https://doi.org/10.3390/met12101692 - 10 Oct 2022
Cited by 3 | Viewed by 2398
Abstract
Rate theory models have been developed for the swelling and He-embrittlement of austenitic stainless steels and Ni-alloys in nuclear reactors. The models illustrate how microstructure evolution during irradiation affects the rate of change of mechanical properties and the dimensional stability. He-stabilised cavity accumulation [...] Read more.
Rate theory models have been developed for the swelling and He-embrittlement of austenitic stainless steels and Ni-alloys in nuclear reactors. The models illustrate how microstructure evolution during irradiation affects the rate of change of mechanical properties and the dimensional stability. He-stabilised cavity accumulation on grain boundaries, which causes brittle failure at low stresses and strains known as He-embrittlement, is shown to be strongly dependent on the irradiation temperature and the rate of production of Frenkel pairs and He atoms. The results show that the accumulation of cavities on grain boundaries falls into two regimes: (i) that dictated by matrix bubble swelling at low temperatures; and (ii) that dictated by matrix void swelling at high temperatures. Full article
(This article belongs to the Special Issue Mechanical Behavior of Reactor Structural Materials)
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