Publications

MX precipitate behavior in an irradiated advanced Fe-9Cr steel: Self-ion irradiation effects on phase stability

Extending damage accumulation of commercial reactor irradiated 316 stainless steel with ion irradiation

Accelerating Domain-Aware Electron Microscopy Analysis Using Deep Learning Models with Synthetic Data and Image-Wide Confidence Scoring

3D printed features on nuclear fuel cladding for optimized heat transfer

Systematic study of radiation-induced segregation in neutron-irradiated FeCrAl alloys

System and method for multi-modal microscopy

Materials swelling revealed through automated semantic segmentation of cavities in electron microscopy images

FeCrAl fuel/clad chemical interaction in light water reactor environments

Evaluation of Human-Bias in Labeling of Ambiguous Features in Electron Microscopy Machine Learning Models

Towards Optimized Characterization of Dislocation Loops in Irradiated FCC Alloys Using On-Zone STEM Techniques

The Development of Castable Nanostructured Alloys for Fusion Reactor First-Wall/Blanket Applications.

Real-time, On-Microscope Automated Quantification of Features in Microcopy Experiments Using Machine Learning and Edge Computing

Post-Irradiation Fracture Toughness Characterization of Generation II FeCrAl Alloys

Performance and limitations of deep learning semantic segmentation of multiple defects in transmission electron micrographs

Micromechanical aspects of deformation and failure of advanced Iron-Chromium-Aluminum alloys

Microchemical evolution of irradiated additive-manufactured HT9

Investigation of deformation mechanisms in an advanced FeCrAl alloy using in-situ SEM-EBSD testing

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

Effect of N2-and CO2-containing shielding gases on composition modification and carbonitride precipitation in wire arc additive manufactured grade 91 steel

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation (vol 549, 152804, 2021)

Creep behavior of an additively manufactured 9Cr steel in the as-built condition

Corrigendum to" Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

A deep learning based automatic defect analysis framework for In-situ TEM ion irradiations

Videos captured using Transmission Electron Microscopy (TEM) can encode details regarding the morphological and temporal evolution of a …
DOI

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in …
DOI

STEM characterization of dislocation loops in irradiated FCC alloys

Segregation behavior and phase instability of Eurofer97 after neutron irradiation to 72 dpa

Performance, successes and limitations of deep learning semantic segmentation of multiple defects in transmission electron micrographs

Multi defect detection and analysis of electron microscopy images with deep learning

Microstructures and mechanical properties of a modified 9Cr ferritic-martensitic steel in the as-built condition after additive manufacturing

Microstructure and mechanical properties of high Mn-containing ferritic-martensitic alloys exposed to cyclical thermal treatment

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Fracture Toughness Characterization of Generation II FeCrAl Alloys after\~ 18 dpa Irradiation

FeCrAl Accident Tolerant Fuel Cladding: Insights from a Decade of Development

Editorial for proceedings of ICFRM-19

Dislocation loops in proton irradiated uranium-nitrogen-oxygen system

Development and deployment of automated machine learning detection in electron microcopy experiments

Deconvoluting the effect of chromium and aluminum on the radiation response of wrought FeCrAl alloys after low-dose neutron irradiation

An Open Analysis Platform with In-Situ Machine Vision for Electron Microscopy Final Technical Report

A deep learning based automatic defect analysis framework for In-situ TEM ion irradiations

Microstructures and mechanical properties of a modified 9Cr ferritic-martensitic steel in the as-built condition after additive manufacturing

A newly developed nano-structured high-Mn 9Cr ferritic-martensitic (FM) steel designed for additive manufacturing …

STEM Characterization of Dislocation Loops in Irradiated FCC Alloys

In this study, we demonstrate the methodology systematically developed for dislocation loop (perfect and faulted loops) imaging and …

Microstructural characterization of cold-worked 316 stainless steel flux thimble tubes irradiated up to 100 dpa in a commercial Pressurized Water Reactor

Two flux thimble tubes (FTTs) made of 15% cold-worked 316 stainless steel (SS) were harvested from Ringhals Pressurized Water Reactor …

Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature

The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between …

Rapid Characterization Methods for Accelerated Innovation for Nuclear Fuel Cladding

Radiation Effects in FeCrAl Alloys for Nuclear Power Applications

Iron–chromium–aluminum (FeCrAl) alloys are of interest in the nuclear industry for their excellent corrosion and oxidation resistance. …

Perspectives on the FESAC transformative enabling capabilities: Priorities, plans, and Status

In early 2017, the Fusion Energy Sciences Advisory Committee (FESAC), an advisory committee to the United States Department of Energy, …

Nanoscale analysis of ion irradiated ODS 14YWT ferritic alloy

In this work, the nanoscale microstructure of an advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy (SM13 heat) with …

Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature

The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between …

In-situ Micromechanical Testing of Neutron Irradiated FeCrAl Alloys

Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding

As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at …

High-Efficiency Three-Dimensional Visualization of Complex Microstructures via Multidimensional STEM Acquisition and Reconstruction

Complex material systems in which microstructure and microchemistry are nonuniformly dispersed require three-dimensional (3D) …

Post-Irradiation Fracture Toughness Characterization of Generation II FeCrAl Alloys

Post irradiation examination of nanoprecipitate stability and $\alpha$′ precipitation in an oxide dispersion strengthened Fe-12Cr-5Al alloy

Oxide dispersion strengthened (ODS) FeCrAl alloy (Fe-12Cr-5Al wt%) was neutron irradiated to 1.8 displacements per atom (dpa) at 215, …

Performance of a ferritic/martensitic steel for nuclear reactor applications fabricated using additive manufacturing

Ferritic/martensitic (FM) steels are being targeted for use in a range of advanced reactor concepts as cladding and structural …

Mechanical behavior and structure of advanced Fe-Cr-Al alloy weldments

FeCrAl alloys are promising for developing accident tolerant nuclear fuel claddings. These alloys showed good environmental …

Influence of welding and neutron irradiation on dislocation loop formation and $\alpha$′ precipitation in a FeCrAl alloy

An advanced accident-tolerant FeCrAl alloy, C35 M (Fe–13Cr–10Al–1Mo, at %), and its laser-fusion weldments were studied after neutron …

Hydrothermal corrosion of coatings on silicon carbide in boiling water reactor conditions

SiC/SiC ceramic matrix composites are an attractive material for use as accident tolerant fuel cladding. However, despite its …

Evaluation of the continuous dilatometer method of silicon carbide thermometry for passive irradiation temperature determination

Silicon carbide (SiC) is a primary candidate for passive irradiation temperature monitoring, and continuous dilatometry (CD) has been …

Evaluation of post-weld heat treatments applied to FeCrAl alloy weldments

The nuclear incident at the Fukushima Daiichi nuclear power plant has created a strong push for accident-tolerant fuel cladding to …

Emulation of fast reactor irradiated T91 using dual ion beam irradiation

Dual ion irradiations using 5 MeV defocused Fe2+ ions and co-injected He2+ ions were conducted on a ferritic-martensitic steel alloy, …

Distinct effects of irradiation on the structure and chemical reactivity of silicates and carbonates

Concrete in nuclear power plants (NPP) is subject to sustained exposure to neutron irradiation. The interaction of neutrons with …

Assessment of deformation mechanisms in neutron-irradiated accident-tolerant fecral alloys via in situ mechanical testing and TEM analysis

Iron-chromium-aluminum (FeCrAl) alloys are promising as an accident-tolerant fuel cladding for loss-of-coolant accident scenarios; …

Accident tolerant fecral fuel cladding: current status towards commercialization

FeCrAl alloys are rapidly becoming mature candidate alloys for accident tolerant fuel applications. The FeCrAl material class has shown …

A Road Map for the Advanced Manufacturing of Ferritic-Martensitic Steels

Advanced manufacturing (AM) is a disruptive manufacturing process often referred to as “the next industrial revolution” because of its …

Role of refractory inclusions in the radiation-induced microstructure of APMT

Kanthal APMT is a promising FeCrAl-based alloy for accident-tolerant fuel cladding because of its excellent high-temperature oxidation …

Radiation-induced degradation of concrete in NPPs

Nuclear power plant life extensions to 60 and potentially 80 years of operation have renewed interest in long-term material …

Preliminary Characterization and Mechanical Performance of Additively Manufactured HT9

Laser-blown powder deposition of HT9 was performed to evaluate the feasibility of using advanced manufacturing to fabricate creep …

Precipitation of $\alpha$′ in neutron irradiated commercial FeCrAl alloys

Alkrothal 720 and Kanthal APMT™, two commercial FeCrAl alloys, were neutron irradiated up to damage doses of 7.0 displacements per atom …

Multi-modal STEM-based tomography of HT-9 irradiated in FFTF

Microstructure evolution of T91 irradiated in the BOR60 fast reactor

Microstructures of T91 neutron irradiated in the BOR60 reactor at five temperatures between 376 °C and 524 °C to doses between 15.4 and …

Mechanical performance of neutron-irradiated dissimilar transition joints of aluminum alloy 6061-T6 and 304L stainless steel

Bimetallic transition joints using aluminum alloy 6061-T6 and stainless steel 304 L are useful in providing reliable stainless steel …

Machine learning applications and opportunities in nuclear materials

Investigation of laser direct energy deposition for production of ODS alloys

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with …

High Temperature Mechanical Properties of HFIR Irradiated FeCrAl Alloys

Handbook on the Material Properties of FeCrAl Alloys for Nuclear Power Production Applications (FY18 Version: Revision 1.1) Prepared for U.S. Department of Energy Nuclear Technology R\&D Advanced Fuels Campaign FY18 version author(s): Kevin G. Field 1 FY17 version author(s)

Ex-situ and in-situ determination of alpha prime phase formation/dissolution in high-cr ferritic alloys using small-angle neutron scattering.

Detection of Open Loop Defects in STEM Images of Irradiation-Damaged Alloys--Source Code for Detection and Image Dataset

Designing for radiation tolerance in fecral alloys

Design and Thermal Analysis for Irradiation of Tensile Specimens from Wrought, Powder Metallurgy, and Additive Processed Alloys in the HFIR

Characterization of dislocation loops and $\alpha$ precipitates in neutron-irradiated fecral weldments

Automated defect analysis in electron microscopic images

Electron microscopy and defect analysis are a cornerstone of materials science, as they offer detailed insights on the microstructure …

Advantages of SEM-EBSD in-situ mechanical testing for investigating radiation effects on materials

Accelerated irradiation testing of miniature fuel specimens

9. INTERNATIONAL COLLABORATIONS

2.6 MICROSTRUCTURAL CHARACTERIZATION OF RAFM STEELS IN HFIR JP28 AND JP29 IRRADIATION CAPSULES

Synergies Between $\alpha$' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

Synergies Between $\alpha$' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

Sub-size tensile specimen design for in-reactor irradiation and post-irradiation testing

The present work aims to provide a complete engineering solution, an appropriate experimental database, and a brief physical background …

Status of Wrought FeCrAl-UO2 Capsules Irradiated in the Advanced Test Reactor

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial …

Radiation Induced Segregation at Low Angle Grain Boundaries in Steels NSUF 2017 Milestone Report

Radiation Induced Segregation at Low Angle Grain Boundaries in Steels NSUF 2017 Milestone Report

ORNL/TM-2017/186 Rev.1 Handbook of FeCrAl Alloys for Nuclear Power Production Applications

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including accident tolerant fuel …

ORNL/TM-2017/186 Rev.1 Handbook of FeCrAl Alloys for Nuclear Power Production Applications

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including accident tolerant fuel …

Microstructural stability of irradiated Fe-Cr-Al alloys for fuel cladding applications.

Microstructural evolution of neutron-irradiated T91 and NF616 to ∼4.3 dpa at 469 °C

Ferritic-martensitic steels such as T91 and NF616 are candidate materials for several nuclear applications. This study evaluates …

Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as …

Mechanical performance of RAFM and ODS steels in HFIR JP28 and JP29 irradiations

The U.S. Department of Energy’s Office of Scientific and Technical Information

Irradiation Testing of Accident Tolerant Fuel FeCrAl Cladding

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Impact of neutron irradiation on thermal helium desorption from iron

The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of …

Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental …

Dislocation loop formation in model FeCrAl alloys after neutron irradiation below 1 dpa

FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited …

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions …

Design, properties, and weldability of advanced oxidation-resistant FeCrAl alloys

Iron-chromium-aluminum (FeCrAl) alloys are promising as corrosion- and oxidation-resistant materials for high-temperature applications. …

Deformation Behavior in Neutron Irradiated Generation II FeCrAl Alloys for Accident Tolerant Fuel Cladding

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years …

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years …

Application of NSUF capabilities towards understanding the emulation of high dose neutron irradiations with ion beams

A combined APT and SANS investigation of $\alpha$′ phase precipitation in neutron-irradiated model FeCrAl alloys

FeCrAl alloys are currently under consideration for accident-tolerant fuel cladding applications in light water reactors owing to their …

10. EXPERIMENTAL TECHNIQUES AND LABORATORY SYSTEMS

The Analysis of the General Performance and Mechanical Behavior of Unirradiated FeCrAl Alloys Before and After Welding

The Analysis of the General Performance and Mechanical Behavior of Unirradiated FeCrAl Alloys Before and After Welding

Submission of FeCrAl Feedstock for Support of AFC ATR- ­ 2 Irradiations

Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO2 Interactions

Status Report on Irradiation Capsules Containing Welded FeCrAl Specimens for Radiation Tolerance Evaluation

Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for …

Roles of vacancy/interstitial diffusion and segregation in the microchemistry at grain boundaries of irradiated Fe-Cr-Ni alloys

This work presents a detailed analysis of the diffusion fluxes near and at grain boundaries of irradiated Fe-Cr-Ni alloys, induced by …

Recent Advances in Analytical Electron Microscopy for Irradiated Materials

Recent Advances in Analytical Electron Microscopy for Irradiated Materials

Post-deformation examination of specimens subjected to SCC testing

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl alloys

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident …

Microstructural Evolution of Type 304 and 316 Stainless Steels Under Neutron Irradiation at LWR Relevant Conditions

Life extension of light water reactors will expose austenitic internal core components to irradiation damage levels beyond 100 …

Microstructural Evolution of Type 304 and 316 Stainless Steels Under Neutron Irradiation at LWR Relevant Conditions

Life extension of light water reactors will expose austenitic internal core components to irradiation damage levels beyond 100 …

Microscopy of Plasma-Materials Interactions in Tungsten for Fusion Power

Irradiation-enhanced $\alpha$′ precipitation in model FeCrAl alloys

Model FeCrAl alloys with varying compositions (Fe(10-18)Cr(10-6)Al at.%) have been neutron irradiated at ∼ 320 to damage levels of ∼ 7 …

FeCrAl alloys for accident tolerant fuel cladding in light water reactors

The goal of the U.S. Department of Energy (DOE) Accident Tolerant Fuel Program (ATF) for light water reactors (LWR) is to identify …

Effect of Alloying Additions on Wrought FeCrAl Alloys for Accident-Tolerant Fuel Cladding

Direct Experimental Evidence for Differing Reactivity Alterations of Minerals following Irradiation: The Case of Calcite and Quartz

Concrete, used in the construction of nuclear power plants (NPPs), may be exposed to radiation emanating from the reactor core. Until …

Dependencies of $\alpha$' embrittlement in neutron-irradiated model Fe-Cr-Al alloys

Database on Performance of Neutron Irradiated FeCrAl Alloys

The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl …

Correlative microscopy of neutron-irradiated materials

Development of new, radiation-tolerant materials that maintain the structural integrity and safety margins over the course of a nuclear …

Complementary Techniques for Quantification of $\alpha$' Phase Precipitation in Neutron-Irradiated Fe-Cr-Al Model Alloys

The substandard performance of Zircaloy LWR cladding materials under loss-of-coolant accident (LOCA) conditions has prompted the search …

Complementary Techniques for Quantification of $\alpha$' Phase Precipitation in Neutron-Irradiated Fe-Cr-Al Model Alloys

The substandard performance of Zircaloy LWR cladding materials under loss-of-coolant accident (LOCA) conditions has prompted the search …

Technology Implementation Plan: ATF FeCrAl Cladding for LWR Application

Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

Role of scale factor during tensile testing of small specimens

The influence of scale factor (tensile specimen geometry and dimensions) on mechanical test results was investigated for different …

Recent Advances in Understanding Radiation Damage In Reactor Cavity Concrete

License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has resulted in a renewed focus on …

Radiation Tolerance of Neutron-Irradiated Model Fe-Cr-Al Alloys

Radiation effects in concrete for nuclear power plants, Part II: Perspective from micromechanical modeling

The need to understand and characterize the effects of neutron irradiation on concrete has become urgent because of the possible …

Radiation effects in concrete for nuclear power plants, Part II: Perspective from micromechanical modeling

The need to understand and characterize the effects of neutron irradiation on concrete has become urgent because of the possible …

Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

Preliminary Results on FeCrAl Alloys in the As-received and Welded State Designed to Have Enhanced Weldability and Radiation Tolerance

The present report summarizes and discusses the recent results on developing a modern, nuclear grade FeCrAl alloy designed to have …

Preliminary Results on FeCrAl Alloys in the As-received and Welded State Designed to Have Enhanced Weldability and Radiation Tolerance

The present report summarizes and discusses the recent results on developing a modern, nuclear grade FeCrAl alloy designed to have …

Optimization of Nuclear Grade FeCrAl Fuel Cladding for Light Water Reactors

LAMDA: Irradiated-Materials Microscopy at Oak Ridge National Laboratory

Oak Ridge National Laboratory’s Low Activation Materials Development and Analysis (LAMDA) laboratory is a dedicated facility …

LAMDA: A facility for advanced characterization of irradiated materials at Oak Ridge national laboratory

First Annual Progress Report on Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

First Annual Progress Report on Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

First Annual Progress Report on Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

First Annual Progress Report on Radiation Tolerance of Controlled Fusion Welds in High Temperature Oxidation Resistant FeCrAl Alloys

Fabrication Control Plan for Mo-L ATF Test Speicmens to be Irradiated in the ATR

Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

Evaluation of the Effect of Composition on Radiation Hardening and Embrittlement in Model FeCrAl Alloys, ORNL/TM-2015/518 M2FT-15OR02302243

Effect of exposure environment on surface decomposition of SiC-silver ion implantation diffusion couples

SiC is a promising material for nuclear applications and is a critical component in the construction of tristructural isotropic (TRISO) …

Dommages d'irradiation dans les cavites en beton des reacteurs aux Etats-Unis

Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve …

Design of Experiment for Irradiation of Welded Candidate Fe-Cr-Al Alloys

Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels

The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 …

Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including …

Current status of FeCrAl alloys as an accident tolerant cladding alloy class for commercial light water reactors

Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced …

Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys

This paper provides an overview of advanced scanning transmission electron microscopy (STEM) techniques used for characterization of …

Strain-induced phase transformation at the surface of an AISI-304 stainless steel irradiated to 4.4 dpa and deformed to 0.8\% strain

Surface relief due to localized deformation in a 4.4-dpa neutron-irradiated AISI 304 stainless steel was investigated using scanning …

Stability of Model Fe-Cr-Al Alloys Under The Presence of Neutron Radiation

Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9 wt.\% Cr model ferritic/martensitic steel

Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for …

Radiation Damage In Reactor Cavity Concrete

Phase stability and mechanical properties of nuclear grade FeCrAl Under LWR-relevant neutron irradiation

This study will be utilized to investigate the formation of a' phases and quantify the number density and average diameter if a' …

Microstructural characterization of deformation localization at small strains in a neutron-irradiated 304 stainless steel

A specific phenomenon - highly localized regions of deformation - was found and investigated at the free surface and near-surface layer …

FY-14 FCRD Milestone: M3FT-14OR0202232 FCRD Advanced Fuels Campaign (AFC) Level 3 Milestone Report Letter Report Documenting Progress of Second Generation ATF FeCrAl Alloy Fabrication

Development of the 2 nd generation ATF FeCrAl alloy has been initiated, and a candidate alloy was selected for trial tube fabrication …

Development of advanced oxidation resistant steel for ATF clad application

Degradation of concrete for nuclear structures: Identified mechanisms and knowledge gaps

Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor …

Analysis of Phase Transformation Studies in Solute Addition Alloys

Alloying and grain boundary structure effects on the radiation induced segregation response in Type 304 variants under neutron irradiation

University of Wisconsin Ion Beam Laboratory: A facility for irradiated materials and ion beam analysis

The University of Wisconsin Ion Beam Laboratory (UW-IBL) has recently undergone significant infrastructure upgrades to facilitate …

Optimal conditions for high current proton irradiations at the university of Wisconsin's ion beam laboratory

The National Electrostatics Corporation’s (NEC) Toroidal Volume Ion Source (TORVIS) source is known for exceptionally high proton …

Letter report documenting identifying billets and alloys fabricated for distribution to program, FY13 FCRD milestone report, M3FT-13OR0202291, ORNL/LTR-2013/322

FY-13 FCRD Milestone M3FT-13OR0202311 Weldability of ORNL Accident Tolerant Fuel Cladding Model Alloys For Thin Walled Tubes

Evaluating the Effects of Irradiation on Concrete at Extended Lifetimes

Dependence on grain boundary structure of radiation induced segregation in a 9 wt.\% Cr model ferritic/martensitic steel

Ferritic/Martensitic (F/M) steels containing 9 wt.% Cr are candidates for structural and cladding components in the next generation of …

Dependence on grain boundary structure of radiation induced segregation in a 9 wt.\% Cr model ferritic/martensitic steel

Ferritic/Martensitic (F/M) steels containing 9 wt.% Cr are candidates for structural and cladding components in the next generation of …

Concrete aging and degradation in NPPs: LWRS program R\&D progress report

Relationship between grain boundary structure and radiation induced segregation in ferritic/martensitic steels

Radiation Induced Segregation in High Chromium Ferritic / Martensitic Steels By

Variability in Chromium Segregation at Lath Boundaries in Proton-Irradiated 9 wt.\% Cr Model Steel Determined by Quantitative X-Ray Mapping

Extended abstract of a paper presented at Microscopy and Microanalysis 2010 in Portland, Oregon, USA, August 1 – August 5, 2010.

Response of nanoclusters in nanostructured ferritic alloys to low-dose proton irradiation

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °c proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and …

Radiation induced segregation in ferritic-martensitic steels

Microchemical variations in proton irradiated 9 wt.\% chromium BCC steels

Microchemical variations in proton irradiated 9 wt.\% chromium BCC steels

Formation of biomimetic porous calcium phosphate coatings on surfaces of polyethylene/zinc stearate Blends

Studies were undertaken investigating improvements to the biological interaction of polymeric implant materials through their coating …