Additive manufacturing (AM) has emerged as a global disruptive technology in
multiple industries for manufacturing complex three-dimensional components by
the deposition of ceramics, alloys, or metal precursors within a variety of
dimensional space. AM techniques provide a unique advantage for the energy
industry due to the shortened development and fabrication times, quality of the
product and repeatability of the process. Not yet commonplace in the energy
sector, AM provides new opportunities in the design space during inception of
new products (both structural component and material design) due to less
limitations on localized design features that could not generally be performed
using conventional fabrication processes (e.g., casting, extrusion, etc.) and
subtractive fabrication. The advantage of using AM in energy applications will
include, but not limited to, advancement in alloying, design and efficiency.
This symposium will integrate invited and contributed talks on use of AM in
various energy industries and includes the following topics based on
experimental and computational approaches:
• Property-microstructure-processing relationships of AM fabricated materials
for structural components (e.g. 316 stainless steel) and fuel systems (e.g.
U-Zr) in nuclear and other energy industries.
• The scope of AM in nuclear energy enabling advanced sensor and
• Advances in AM design concepts such as graded structures and post-processing
• Modeling and simulations for design of high performance AM fabricated
materials and reducing research time and costs.
• Development and Qualification approaches.
Nuclear energy is an essential element of a clean energy strategy, avoiding
greenhouse gas emissions of over two billion tons per year. Ceramic materials
play a critical role in nuclear energy research and applications. Nuclear
fuels, such as uranium dioxide (UO2) and mixed oxide (MOX) fuels, have been
widely used in current light water reactors (LWRs) to produce about 15% of the
electricity in the world. Silicon carbide (SiC) is a promising
accident-tolerant cladding material and is under active research studies. Some
oxide ceramics have been proposed for novel inert matrix fuels or have been
extensively studied as waste forms for the immobilization of nuclear waste.
Moreover, ceramics are under active studies for fusion reactor research.
This symposium focuses on experimental and computational studies of ceramics
for nuclear energy research and applications. Both practical reactor materials
and surrogate materials are of interest. Topics of interest include: defect
production and evolution; mobility, dissolution, and precipitation of solid,
volatile, and gaseous fission products; changes in various properties (e.g.,
thermal conductivity, volume swelling, mechanical properties) induced by
microstructural evolution; and radiation-induced phase changes. Experimental
studies using various advanced characterization techniques for characterizing
radiation effects in ceramics are of particular interest. The irradiation
techniques such as laboratory ion beam accelerators, research and test
reactors, as well as commercial nuclear power reactors are all of interest.
Computational studies across different scales from atomistic to the continuum
are all welcome. Contributions focused on novel fuels such as doped UO2,
high-density uranium fuels like uranium nitrides and silicides, and coatings
for accident-tolerant fuel claddings are also encouraged. This symposium is
intended to bring together national laboratory, university, and nuclear
industry researchers from around the world to discuss the current understanding
of the radiation response of ceramics through experiment, theory, and
Some focused topic areas will be:
" Experimental characterization of non-irradiated and irradiated oxide ceramics
" Multi-scale modeling on microstructure evolution and physical properties in
" Thermal-mechanical properties of oxides for nuclear energy
" Non-oxide ceramics for nuclear energy
The use of molten salts as a coolant in molten salt reactors (MSR) offers many
advantages including low operating pressures, high temperatures, and favorable
heat transfer. For concentrating solar power systems (CSP), the use of molten
salts as a heat transfer medium enables efficient thermal energy storage.
Despite the advantages, the highly aggressive molten salts present a
challenging environment for salt facing materials. Further, the high
temperatures presented by these systems require exceptional mechanical
This symposium covers structural and moderator materials in molten salt for
nuclear power and concentrating solar power.
Abstracts are solicited in the following topics:
- Corrosion of salt-facing materials
- Salt effects in graphite and moderator materials
- Fission product embrittlement
- Alloy selection and design for molten salt applications
- Interaction of fission products with materials
- Mechanical and creep properties
- Salt chemistry effects on materials including radiolysis
- Heat exchanger design
- Welding and cladding issues
- Waste handling and actinide recovery
Current and future generation nuclear reactors require improved structural
materials that improve efficiency during in-service conditions, allow for long
reactor lifetimes, and increase safety during accidents. Given the increasingly
large number of reactor design being considered (e.g. fusion, molten salt,
LWRs, etc.), a series of distinct material concepts have been proposed to
address these needs. Effects of reactor environments on mechanical behavior
will be a key component to predicting strength and performance of materials in
the aforementioned circumstances.
This symposium aims to take a closer look at the mechanical behavior of reactor
components across length scales. With recent advancements and increased use of
in-situ techniques, more is known about irradiation effects on strength than
ever before. Simultaneously, ex-situ techniques are critical to probe
component-sized parts, and validate the use of a material for inclusion within
a reactor. Furthermore, synergy with materials modeling is advancing the
prediction of material performance under normal and accident conditions, as
well as reactor lifetimes.
Topics of interest include, but are not limited to:
• Mechanical behavior testing, including tension, compression, bend, bulge,
creep, fatigue, and fracture
• Effects of environment on strength, including dose, dose rate, temperature,
• Hardness testing, including nanohardness and microhardness
• Development of microstructure sensitive material strength models
• Modeling and simulation of irradiation defect interactions during mechanical
• Macroscopic component modeling for full scale performance
• In-situ mechanical testing, including micromechanical and nanomechanical
compression and tension
• Novel techniques to probe material strength under reactor conditions
The response of fuels and materials to radiation is critical to the performance
of advanced nuclear systems. Key to understanding material performance in a
nuclear environment is the analysis of materials irradiated using test reactors
and ion beam facilities. This symposium will focus on recent results produced
from irradiation programs around the world and will cover fundamental and
applied science aspects of accelerated nuclear materials testing for fission
and fusion reactors. Presentations combining experiment with theory, modeling
and simulation to enhance our understanding of radiation induced degradation in
materials are especially encouraged.
Abstracts are solicited for (but not limited to) the following irradiation
- Fundamental science of radiation damage and defect processes
- Mechanical and fracture behavior of irradiated materials
- Current and advanced nuclear fuels
- Current and advanced structural materials
- Fluence effects in materials
This symposium is focused on nuclear fuels with enhanced accident tolerance for
Light Water Reactor (LWR) with an emphasis on assessment of their performance.
Topics related to design, fabrication, characterization, irradiation,
post-irradiation examination, testing simulating accident conditions, and
modeling/simulation of accident tolerant fuels are within the scope of this
symposium. This symposium will provide a platform for exhibiting and discussing
recent experimental and computational progress in this area.
Abstracts are solicited in (but not limited to) the following topics:
• Accident tolerant fuel and cladding materials development
• Microstructure, mechanical, thermodynamic and physical property
• Radiation damage and post-irradiation examination on accident tolerant fuel
and cladding materials
• Fuel rod, fuel cladding and component materials behaviors under accident
conditions (corrosion, steam corrosion, chemical interaction, etc.)
• Multi-scale modeling/simulation of materials behavior under normal and
Atom probe tomography (APT), is an emergent characterization technique that is
capable of determining the chemical identity of each individual atom and
generating 3D chemical maps imaging the distribution of individual atoms. The
technique offers high spatial resolution (better than 0.3 nm achievable in all
directions) and high analytical sensitivity (as good as 1 appm). APT provides
information on elemental composition of the specimen, 3D visualization of
distribution of atoms, composition of phases, morphology and size of
precipitates, and solute distribution across interfaces, at grain boundaries
and along dislocations. In many APT analyses, crystallographic information has
been retained within the data, with the potential to directly relate the
composition of specific microstructural features to their crystallography with
unprecedented sensitivity and resolution. APT can be utilized in many different
fields for advanced imaging and analysis of metals, minerals and materials,
despite some limitations.
This symposium is designed to bring together scientists, engineers and
technicians from across disciplines to discuss the technique of APT, its
applications and limitations. The symposium will encompass research and
applications spanning a wide variety of topics. Presentations on experimental,
theoretical, and modeling research are solicited. Topics for this symposium
include, but are not limited to:
Applications of APT in advanced characterization of metals, minerals and
3D reconstruction and data analysis
Impact of specimen and instrument parameters and optimization of acquisition
Specimen preparation techniques
Limitations of APT
Progress in APT technique
This symposium will highlight current computational materials science and
engineering efforts for nuclear reactors in the United States and abroad. High
neutron flux, thermal and chemical gradients, and corrosive environments cause
significant degradation in the chemical and mechanical properties of materials.
Enhanced radiation resistance of structural materials and nuclear fuels are
needed to overcome technological challenges necessary for future nuclear
systems. This symposium seeks abstracts that apply�atomistic and mesoscale
simulations to discover, understand, and engineer the macroscale performance of
fission/fusion reactor materials, including fuel, cladding, and structural
This symposium will also consider multiscale modeling efforts that bridge
length and time scales in order to better connect simulation results with
experimental data for predictive model validation. It will also highlight
validation of all relevant models, as well as uncertainty quantification.
Finally, the application of ICME approaches to use modeling and simulation to
better understand structure-property relationships, their associated links with
performance, and their application to designing future reactor concepts and
materials is also desired.
Some examples include:
• Modeling and simulation of materials behavior under extreme environments –
radiation, corrosion, stress and temperature, including radiation effects,
phase stability, fuel-clad interactions, fission product behavior.
• Modeling and simulation of model materials to uncover fundamental
behavior�that affects material performance in radiative environments.
• Developing improved material models for LWR fuel and cladding.
• Modeling and simulation of new fuel materials including metal, silicide, and
• Modeling and simulation of new cladding materials, such as silicon carbide,
coated zirconium alloys, or FeCrAl.
• Development and integration of computational tools, methods, and databases
for reactor structural material design.
Uncertainty quantification and validation of all the applications listed above.
Globally, significant efforts are ongoing to meet the growing energy demand
with the increased use of nuclear energy. Extensive work is being performed to
develop materials and fuels for the advanced nuclear reactors. In addition,
efforts are also ongoing to extend the life of existing nuclear power plants.
Scientists, engineers, and students at various national laboratories,
universities, and industries are working on a number of materials challenges
for the nuclear energy systems. The objective of this symposium is to provide a
platform for these researchers to congregate, exhibit and discuss their current
research work, in addition to sharing the challenges and solutions with the
professional community and thus, shape the future of nuclear energy.
Abstracts are solicited in (but not limited to) the following topics:
- Nuclear reactor systems
- Advanced nuclear fuels - fabrication, performance, and design
- Advanced nuclear fuels - properties and modeling
- Advanced structural materials - fabrication, joining, properties, and
- Lifetime extension of reactors - nuclear materials aging, degradation, and
- Experimental, modeling, and simulation studies
- Fundamental science of radiation-material interactions
- Irradiation effects in nuclear materials
- Materials degradation issues - stress corrosion cracking, corrosion, creep,
fatigue, and others
- Design of materials for extreme radiation environments
- Radiation measurement techniques and modeling studies
- Nuclear waste - disposal, transmutation, spent nuclear fuel reprocessing