Publications

Mar, 2026

Conference Paper

Plastic Deformation Processes Accompanying Stress Corrosion Crack Propagation in Irradiated Austenitic Steels

During stress corrosion crack propagation, stress values in the crack tip vicinity often exceed material yield stress limit. Plastic deformation processes may accompany and influence cracking. Here, stress corrosion crack propagation and deformation mechanisms were investigated using EBSD analysis. The investigated material was 304L Ti-enriched austenitic stainless steel i…

Apr, 2025

Journal

How irradiation promotes intergranular stress corrosion crack initiation

Irradiation assisted stress corrosion cracking (IASCC) is a form of intergranular stress corrosion cracking that occurs in irradiated austenitic alloys. It requires an irradiated microstructure along with high temperature water and stress. The process is ubiquitous in that it occurs in a wide range of austenitic alloys and water chemistries, but only when the alloy is irra…

Sep, 2024

Journal

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Feb, 2024

ORNL Report

Microstructural Characterization of Annealed and Aged Refractory Alloys

Nov, 2023

ORNL Report

Technologies to Reactors: Enabling Accelerated Deployment of Nuclear Energy Systems Workshop Report

Nov, 2023

ORNL Report

Technical Gap Assessment for Materials and Component Integrity Issues for Molten Salt Reactors

Nov, 2023

ORNL Report

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM, MATERIALS RESEARCH PATHWAY, TECHNICAL PROGRAM PLAN

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) Program is charged with performing the research and development (R&D) to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, it is essential to use the mechanistic understanding of deg…

Nov, 2023

Journal

Varied enhancements in mechanical properties and sodium compatibility of Grade 92 by thermomechanical treatments

As a candidate material for advanced nuclear reactors, a ferritic-martensitic (FM) steel Grade 92 in the standard normalized and tempered (NT) condition was subjected to two types of thermomechanical treatments (TMTs) with one having a lower level of deformation (T5A) and the other having a higher level of deformation (T5B). The testing results of the three conditions indi…

Jun, 2017

Book Chapter

Properties of Austenitic Stainless Steels for Nuclear Reactor Applications

Austenitic stainless steels are widely used in current commercial BWR and PWR systems as in-core and surrounding structural materials. These versatile steels have also been the main materials considered and applied for many advanced reactor technologies, including fast-breeder reactors and magnetic fusion reactors. For advanced reactors for higher-temperatures or high life…

Jun, 2017

Journal

Structure and Mechanical Properties of Improved Cast Stainless Steels for Nuclear Applications

Casting of stainless steels is a promising and cost saving way of directly producing large and complex structures, such a shield modules or divertors for the ITER. In the present work a series of modified high-nitrogen cast steels has been developed and characterized. The steels, based on the cast equivalent of 316 composition, have increased N (0.14-0.36%) and Mn (2-5.1%)…

May, 2017

Conference Paper

Characterization of Materials Properties and Crack Propagation Mechanisms in Damaged Alloy 718 Leaf Springs Following Commercial Reactor Exposure

May, 2017

Journal

Characterization of alloy 718 processed at different experimental conditions