Nuclear engineering researchers develop new resilient oxide dispersion strengthened alloy
The nuclear community has a high need for reliable and durable materials
for core components of nuclear reactors

Date:
March 4, 2021
Source:
Texas A&M University
Summary:
Researchers have recently shown superior performance of a new
oxide dispersion strengthened (ODS) alloy they developed for use
in both fission and fusion reactors.



FULL STORY ========================================================================== Texas A&M University researchers have recently shown superior performance
of a new oxide dispersion strengthened (ODS) alloy they developed for
use in both fission and fusion reactors.


==========================================================================
Dr. Lin Shao, professor in the Department of Nuclear Engineering, worked alongside research scientists at the Los Alamos National Laboratory and Hokkaido University to create the next generation of high-performance
ODS alloys, and so far they are some of the strongest and best-developed
metals in the field.

ODS alloys consist of a combination of metals interspersed with small, nanometer-sized oxide particles and are known for their high creep
resistance.

This means that as temperatures rise, the materials keep their shape
instead of deforming. Many ODS alloys can withstand temperatures up to
1,000 C and are typically used in power generation and engines within
aerospace engineering, as well as cutlery.

The nuclear community has a high need for reliable and durable materials
to make up the core components of nuclear reactors. The material must
be high strength, radiation tolerant and resistant to void swelling
(materials develop cavities when subjected to neutron radiation, leading
to mechanical failures).

Nuclear researchers like Shao are consistently seeking to identify quality creep-resistant and swelling-resistant materials for their use in high- temperature reactors.

"In general, ODS alloys should be resistant to swelling when exposed
to extreme neutron irradiation," said Shao. "However, the majority of commercial ODS alloys are problematic from the beginning." This is
because almost all commercial ODS alloys are based on the ferritic
phase. Ferritic alloys, classified by their crystalline structure
and metallurgical behavior, have good ductility and reasonable
high-temperature strength. However, the ferritic phase is the weakest
phase when judged by its swelling resistance, therefore making the
majority of commercial ODS alloys fail in the first line of defense.

Shao, known internationally for his pioneering work in radiation
materials science, directs the accelerator laboratory for testing
alloys under extreme irradiation conditions. Shao and his research team collaborated with the Japanese research group at Hokkaido University
led by Dr. Shigeharu Ukai to develop various new ODS alloys.

"We decided to explore a new design principle in which oxide particles are embedded in the martensitic phase, which is best to reduce void swelling, rather than the ferritic phase," said Shao.

The resulting ODS alloys are able to survive up to 400 displacements
per atom and are some of the most successful alloys developed in the
field, both in terms of high-temperature strength and superior-swelling resistance.

Details of the complete project were published in the Journal of Nuclear Materials along with the most recent study. The team has since conducted multiple studies and attracted the attention from the U.S. Department
of Energy and nuclear industry. The project resulted in a total of 18
journal papers and two doctoral degree dissertations.

========================================================================== Story Source: Materials provided by Texas_A&M_University. Original
written by Laura Simmons.

Note: Content may be edited for style and length.


========================================================================== Journal Reference:
1. Hyosim Kim, Jonathan G. Gigax, Shigeharu Ukai, Frank A. Garner,
Lin Shao.

Oxide dispersoid coherency of a ferritic-martensitic
12Cr oxide- dispersion-strengthened alloy under self-ion
irradiation. Journal of Nuclear Materials, 2021; 544: 152671 DOI:
10.1016/j.jnucmat.2020.152671 ==========================================================================

Link to news story: https://www.sciencedaily.com/releases/2021/03/210304161114.htm

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