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Dear Colleagues,

Nuclear reactors cannot function without excellent structural materials, especially the advanced fourth generation fission and fusion reactor. In recent years, based on the development needs of various advanced reactor types, a series of advanced nuclear structural materials have been developed and a lot of research has been carried out, such as ferritic martensitic steel, austenitic steel, high-temperature nickel-based alloy, tungsten alloy, high-temperature refractory alloy, high-entropy alloy, silicon carbide, carbon-based composites and so on. Structural materials will face huge challenges due to the severe service environment. Among them, the mechanism of irradiation damage is an important topic. Irradiation can lead to material damage, from the evolution of microstructure to the deterioration of mechanical properties. Hence, it is particularly important to reveal the irradiation damage mechanism of structural materials, as well as to grasp their irradiation performance to ensure the safe operation of the reactor.

This Special Issue, “Irradiation Damage Research of Advanced Nuclear Structural Materials”, invites reviews and research articles on new irradiation material research activities that focus on understanding the irradiation effects on microstructure evolution/mechanical properties, as well as on the underlying mechanisms using experimental and/or modelling approaches. Materials include, but are not limited to, those mentioned above. In addition, both neutron and energetic particle irradiation research papers are welcome.

Dr. Hefei Huang
Guest Editor

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Keywords

reactor structural materials
ion irradiation damage
defects
microstructural characterization
mechanical properties

Published Papers (1 paper)

Research

43 pages, 9377 KiB

Open AccessArticle

Microstructural Effects on Irradiation Creep of Reactor Core Materials

by Malcolm Griffiths

Materials 2023, 16(6), 2287; https://0-doi-org.brum.beds.ac.uk/10.3390/ma16062287 - 13 Mar 2023

Cited by 1 | Viewed by 1585

Abstract

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects. Full article

(This article belongs to the Special Issue Irradiation Damage Research of Advanced Nuclear Structural Materials)

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