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Special Issue "Nuclear Materials 2015"

A special issue of Materials (ISSN 1996-1944).

Deadline for manuscript submissions: closed (31 October 2015)

Special Issue Editors

Guest Editor
Dr. Jie Lian

Department of Mechanical, Aerospace and Nuclear Engineering Rensselaer Polytechnic Institute (RPI) JEC 5048, 110 8th Street, Troy NY 12180, USA
Website | E-Mail
Interests: Design and fabrication of advanced nuclear fuels; Advanced Waste forms for effective nuclear waste management; Radiation-matter interaction; Materials behavior under extreme environments; nanostructured materials for alternative energy applications
Guest Editor
Dr. Di Yun

Fuels and Materials Modeling Section, Nuclear Engineering Division, Argonne National Laboratory, Argonne, IL 60439, USA
E-Mail
Phone: 630-252-4171
Interests: Irradiation effects in materials; Heavy ion irradiation; Nuclear fuel performance modeling and simulation (including oxide fuels and metallic alloy fuels); Materials characterization via multiple experimental techniques including X-ray (Synchrotron), SEM, TEM, and STEM

Special Issue Information

Dear Colleagues,

Materials envisioned for next-generation nuclear systems will encounter extreme environments of intensive radiation, high temperature, and highly corrosive conditions. The current materials for light water reactors are also experiencing great challenges, particularly under accident scenarios. Extensive efforts are ongoing to develop materials and fuels for advanced nuclear systems, including design, fabrication and testing of accident tolerant fuels, structural materials, and materials for effective nuclear waste managements. Fundamental understanding of the microstructural evolution of materials and fuels under irradiations will be critical for the development of materials and modeling the changes of materials properties. Science-based, next-generation fuel performance tools are being developed in order to create a predictive modeling capability for nuclear fuel performance, and to assist in the design and analysis of reactor systems. Experimental validation for modeling and simulation tools is also critical for the development of such tools, and specifically developed separate effect experiments can provide great insight into lower length scale modeling and simulation tools. This Special Issue will focus on the recent developments on these critical topics, and new research articles and review articles are invited.

Dr. Jie Lian
Dr. Di Yun
Guest Editor

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All papers will be peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Materials is an international peer-reviewed open access monthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 1500 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.


Keywords

  • Ÿ   Advanced nuclear fuels – design, fabrication and testing
  • Ÿ   Advanced nuclear fuels – performance evaluation and modeling
  • Ÿ   Advanced structure materials – fabrication, properties and characterization
  • Ÿ   Radiation damage of fission and fusion materials
  • Ÿ   Fundamental science of matter – irradiation interactions
  • Ÿ   Advanced modeling and simulation of materials
  • Ÿ   Materials design under extreme environments
  • Ÿ   Waste form materials for effective nuclear waste management

Published Papers (5 papers)

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Research

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Open AccessFeature PaperArticle The Role of Grain Size on Neutron Irradiation Response of Nanocrystalline Copper
Materials 2016, 9(3), 144; doi:10.3390/ma9030144
Received: 2 December 2015 / Revised: 31 January 2016 / Accepted: 23 February 2016 / Published: 1 March 2016
Cited by 2 | PDF Full-text (4572 KB) | HTML Full-text | XML Full-text
Abstract
The role of grain size on the developed microstructure and mechanical properties of neutron irradiated nanocrystalline copper was investigated by comparing the radiation response of material to the conventional micrograined counterpart. Nanocrystalline (nc) and micrograined (MG) copper samples were subjected to a range
[...] Read more.
The role of grain size on the developed microstructure and mechanical properties of neutron irradiated nanocrystalline copper was investigated by comparing the radiation response of material to the conventional micrograined counterpart. Nanocrystalline (nc) and micrograined (MG) copper samples were subjected to a range of neutron exposure levels from 0.0034 to 2 dpa. At all damage levels, the response of MG-copper was governed by radiation hardening manifested by an increase in strength with accompanying ductility loss. Conversely, the response of nc-copper to neutron irradiation exhibited a dependence on the damage level. At low damage levels, grain growth was the primary response, with radiation hardening and embrittlement becoming the dominant responses with increasing damage levels. Annealing experiments revealed that grain growth in nc-copper is composed of both thermally-activated and irradiation-induced components. Tensile tests revealed minimal change in the source hardening component of the yield stress in MG-copper, while the source hardening component was found to decrease with increasing radiation exposure in nc-copper. Full article
(This article belongs to the Special Issue Nuclear Materials 2015)
Open AccessArticle Energetic Study of Helium Cluster Nucleation and Growth in 14YWT through First Principles
Materials 2016, 9(1), 17; doi:10.3390/ma9010017
Received: 23 November 2015 / Revised: 10 December 2015 / Accepted: 18 December 2015 / Published: 2 January 2016
Cited by 1 | PDF Full-text (1055 KB) | HTML Full-text | XML Full-text
Abstract
First principles calculations have been performed to energetically investigate the helium cluster nucleation, formation and growth behavior in the nano-structured ferritic alloy 14YWT. The helium displays strong affinity to the oxygen:vacancy (O:Vac) pair. By investigating various local environments of the vacancy, we find
[...] Read more.
First principles calculations have been performed to energetically investigate the helium cluster nucleation, formation and growth behavior in the nano-structured ferritic alloy 14YWT. The helium displays strong affinity to the oxygen:vacancy (O:Vac) pair. By investigating various local environments of the vacancy, we find that the energy cost for He cluster growth increases with the appearance of solutes in the reference unit. He atom tends to join the He cluster in the directions away from the solute atoms. Meanwhile, the He cluster tends to expand in the directions away from the solute atoms. A growth criterion is proposed based on the elastic instability strain of the perfect iron lattice in order to determine the maximum number of He atoms at the vacancy site. We find that up to seven He atoms can be trapped at a single vacancy. However, it is reduced to five if the vacancy is pre-occupied by an oxygen atom. Furthermore, the solute atoms within nanoclusters, such as Ti and Y, will greatly limit the growth of the He cluster. A migration energy barrier study is performed to discuss the reduced mobility of the He atom/He cluster in 14YWT. Full article
(This article belongs to the Special Issue Nuclear Materials 2015)
Open AccessArticle Investigation of High-Energy Ion-Irradiated MA957 Using Synchrotron Radiation under In-Situ Tension
Materials 2016, 9(1), 15; doi:10.3390/ma9010015
Received: 31 October 2015 / Revised: 18 December 2015 / Accepted: 24 December 2015 / Published: 2 January 2016
Cited by 4 | PDF Full-text (2943 KB) | HTML Full-text | XML Full-text
Abstract
In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak
[...] Read more.
In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner. Full article
(This article belongs to the Special Issue Nuclear Materials 2015)
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Open AccessArticle Progress in Mirror-Based Fusion Neutron Source Development
Materials 2015, 8(12), 8452-8459; doi:10.3390/ma8125471
Received: 30 October 2015 / Revised: 22 November 2015 / Accepted: 27 November 2015 / Published: 4 December 2015
Cited by 15 | PDF Full-text (969 KB) | HTML Full-text | XML Full-text
Abstract
The Budker Institute of Nuclear Physics in worldwide collaboration has developed a project of a 14 MeV neutron source for fusion material studies and other applications. The projected neutron source of the plasma type is based on the gas dynamic trap (GDT), which
[...] Read more.
The Budker Institute of Nuclear Physics in worldwide collaboration has developed a project of a 14 MeV neutron source for fusion material studies and other applications. The projected neutron source of the plasma type is based on the gas dynamic trap (GDT), which is a special magnetic mirror system for plasma confinement. Essential progress in plasma parameters has been achieved in recent experiments at the GDT facility in the Budker Institute, which is a hydrogen (deuterium) prototype of the source. Stable confinement of hot-ion plasmas with the relative pressure exceeding 0.5 was demonstrated. The electron temperature was increased up to 0.9 keV in the regime with additional electron cyclotron resonance heating (ECRH) of a moderate power. These parameters are the record for axisymmetric open mirror traps. These achievements elevate the projects of a GDT-based neutron source on a higher level of competitive ability and make it possible to construct a source with parameters suitable for materials testing today. The paper presents the progress in experimental studies and numerical simulations of the mirror-based fusion neutron source and its possible applications including a fusion material test facility and a fusion-fission hybrid system. Full article
(This article belongs to the Special Issue Nuclear Materials 2015)

Review

Jump to: Research

Open AccessReview Irradiation Induced Microstructure Evolution in Nanostructured Materials: A Review
Materials 2016, 9(2), 105; doi:10.3390/ma9020105
Received: 30 November 2015 / Accepted: 2 February 2016 / Published: 6 February 2016
Cited by 5 | PDF Full-text (2715 KB) | HTML Full-text | XML Full-text
Abstract
Nanostructured (NS) materials may have different irradiation resistance from their coarse-grained (CG) counterparts. In this review, we focus on the effect of grain boundaries (GBs)/interfaces on irradiation induced microstructure evolution and the irradiation tolerance of NS materials under irradiation. The features of void
[...] Read more.
Nanostructured (NS) materials may have different irradiation resistance from their coarse-grained (CG) counterparts. In this review, we focus on the effect of grain boundaries (GBs)/interfaces on irradiation induced microstructure evolution and the irradiation tolerance of NS materials under irradiation. The features of void denuded zones (VDZs) and the unusual behavior of void formation near GBs/interfaces in metals due to the interactions between GBs/interfaces and irradiation-produced point defects are systematically reviewed. Some experimental results and calculation results show that NS materials have enhanced irradiation resistance, due to their extremely small grain sizes and large volume fractions of GBs/interfaces, which could absorb and annihilate the mobile defects produced during irradiation. However, there is also literature reporting reduced irradiation resistance or even amorphization of NS materials at a lower irradiation dose compared with their bulk counterparts, since the GBs are also characterized by excess energy (compared to that of single crystal materials) which could provide a shift in the total free energy that will lead to the amorphization process. The competition of these two effects leads to the different irradiation tolerance of NS materials. The irradiation-induced grain growth is dominated by irradiation temperature, dose, ion flux, character of GBs/interface and nanoprecipitates, although the decrease of grain sizes under irradiation is also observed in some experiments. Full article
(This article belongs to the Special Issue Nuclear Materials 2015)

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