Materials science for nuclear security applications is a key component. Ensuring the functionality of components for nuclear security applications, replacement of components for existing systems and enhancing performance while reducing costs are key elements based on smart materials design and a comprehensive understanding of the materials degradation mechanisms in service conditions.

Scientific goal: determine and characterize the radiation-induced defects in the base metal; in the oxides (including nanocrystalline oxides) and at Metal/Oxide interfaces using in-situ TEM characterization.

High entropy alloys (HEAs) are emerging as a candidate for improving mechanical strength and radiation tolerance as a result of their unique electronic structure. This is because chemical disorder and compositional fluctuations have large effects on energy dissipation, defect evolution and their response to radiation. While previous transmission electron microscopy (TEM) and other studies showed that damage accumulation was suppressed by increasing chemical disorder, they could not reveal vacancy clusters below 2 nm leaving critical gap in understanding defect formation and buildup in these alloys. The proposed research aims to experimentally monitor defect formation on atomistic scale and their buildup to large clusters and voids. It will combine in-situ and ex-situ positron annihilation spectroscopy (PAS) with in-situ and ex-situ TEM to capture isolated vacancies, small vacancy clusters, larger clusters and voids, thus bridging the gap between the atomic scale and mesoscale characterization of radiation induced defects in HEAs. The proposed research should reveal the effect of chemical disorder on defect formation, migration and evolution in a radiation environment and reveal the damage and annealing mechanisms in Single -Phase Concentrated Solid Solution alloys (SP-CSAs) and HEAs i through the following research plans: 1- study of defect production from collision cascades on an atomic and mesoscale level in alloys with increasing chemical complexity from one to five. PAS can quantitative information about the formation of single vacancies and small clusters in irradiated samples in individual and overlapped cascades. ���������It will especially measure defect production in individual defect cascades from very low doses before they begin to overlap. 2- Study of defect evolution and annealing mechanism through measuring the number and size of surviving defects by PAS and TEM and their dependence on the degree of chemical disorder. This will include measuring vacancy clusters below 2nm and atomic size defects after cascade overlapping by PAS. 3- Study of the effect of different chemical elements on defect movement and evolution in these alloys. 4- Real time monitoring of defects at low ���������and high dose rates trough In-situ ���������PAS and In-situ TEM ���������measurements with ion irradiation will be carried out to study defect dynamic including production, annihilation and evolution, 5- ���������Measuring electronic structures and construct fermi surfaces of SP-CSAs and HEAs and elucidate how they are affected by increasing chemical complexity. Combined with temperature dependent transport measurements they will reveal how the chemical complexity affects electronic structures and transport properties which are directly related to thermal conductivity and energy dissipation and subsequently defect evolution.

The objective of this work is to determine if switching from LiOH to KOH to control the pH in nuclear reactors is possible without worsening the corrosion behavior of the structural alloys used in PWR core internal components. The impacts of such a change and the consequent water chemistry alterations on the corrosion processes and NPP core-internal component service-life will be assessed and better understood

NCSU will oversee the project and be in charge of the characterization of the processed samples and of the mechanical testing and irradiation testing.

NCSU will be in charge of characterizing the tested samples using techniques including electron microscopy

Scalable and cost-effective fabrication of creep-resistant oxide dispersion strengthened (ODS) steel is critical for constructing blankets in future fusion reactors that operate above 900K for economic power conversion cycles. The current ODS steel fabrication route requires mechanical alloying (MA) of precursor ODS steel powder for over 40 hours and highly skilled thermal-mechanical processing (TMP), which are cost-prohibitive and lack consistent part to part reliable performance for certification. For the first time, Pacific Northwest National Laboratory (PNNL) and its partners will combine MA-free synthesis of precursor ODS steel powder and advanced manufacturing methods, including first-of-its-kind shear assisted processing and extrusion (ShAPE) and additive manufacturing (AM) methods of laser directed energy deposition (DED) and electron beam melting (EBM), to fabricate ODS steel tubes, plates, and complex parts with scalability. This will reduce cost for MA-free ODS steel components by 50% or more and yield reliable performance with creep resistance above 900K that can be certified for reactor use. In addition, PNNL������������������s friction stir welding (FSW) machine, having the highest stiffness in the US, is capable of joining ODS steel components for fusion blanket mock-up structures without compromising oxide distribution.

Understanding the mechanical response of materials under radiation is essential to qualifying any material for use within a nuclear reactor. In a radiation environment, such as in a reactor core, the radiation damage degrades many instruments thus it limiting the instrumentation options. Furthermore, the use of active instrumentation for in situ monitoring increases the cost of experiments. On the other hand, passive instrumentation can be developed and used to evaluate critical parameters for fast and reliable screening tests prior to the extended irradiation campaigns. The accuracy of passive instruments can be significantly improved when coupled with state-of-the-art computational methods. In this project, we propose to develop passive instrumentation for the determination of permanent strains induced by irradiation and extract critical parameters using computational methods. The model used to interpret the experiment will utilize existing crystal plasticity models developed within Nuclear Energy Advanced Modelling and Simulation (NEAMS) as well as machine learning algorithms to be developed at Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT). An experiment will also be designed at INL for irradiation at the MIT test reactor (MITR). The experiment will benefit from engineered anisotropic materials and characterize the directional deformation in response to neutron radiation. The results of the experiment will be incorporated into the model so that the material response can be predicted for future uses as a probe material. This will enable materials research to more quickly and effectively separate radiation and thermal contributions to mechanical deformation

The goals of this Energy Frontiers Research Center are (a) to understand the coupling between radiation damage and corrosion, (b) to predict irradiation-assisted corrosion in passivating and non-passivating modes, and (c) to inform design of materials that better withstand degradation in this environment of coupled extremes. The underlying feature controlling these phenomena is mass transport. While thermodynamics governs what can happen, kinetics dictates what will happen over relevant times. Our goal is thus to develop a Fundamental Understanding of Transport Under Reactor Extremes ������������������ FUTURE.

A novel thermo-mechanical fatigue (TMF) testing system, referred by miniature TMF (MTMF) system has been developed at NCSU for in-situ testing of miniature specimens within Scanning Electron Microscopes (SEM). The MTMF is capable of prescribing axial-torsional loading to solid specimen and axial-torsional-internal pressure loading to tubular specimen of 1 mm diameter at elevated temperatures (up to 1000oC) to investigate deformation of microstructure and failure mechanism in real time. Currently, in-situ SEM testing with the MTMF is performed at the Analytical Instrumentation Facility (AIF) at NCSU. This poses a serious restriction to investigate failure mechanisms of very high temperature reactor (VHTRs) materials primarily because with a user facility, such as AIF, we can only perform short-term tests that span over few days. However, fatigue, creep and creep-fatigue tests for VHTR materials may span from few days to several weeks. Hence, existing SEMs on campus are not available for long-term in-situ testing of VHTR materials. Currently, fatigue, creep and creep-fatigue failure mechanisms of new and existing alloys are mostly investigated through ex-situ testing or short duration in-situ uniaxial testing within SEM. Consequently, initiation and propagation of many failure mechanisms, especially interactions between creep and fatigue mechanisms in reducing high temperature component lives remain unknown. Hence, developing a shared in-situ testing laboratory (ISTL) is essential to allow NCSU researchers to perform novel research on nuclear materials addressing issues of fatigue, creep and creep-fatigue failure mechanisms. The proposed ISTL dedicated to performing long-term fatigue, creep and creep-fatigue tests is in critical need to develop design criteria of VHTR materials for ASME Code Sec III Div 5. However, existing facilities at NCSU or any other universities or national labs in the nation do not have a facility dedicated to perform long term tests representing realistic loading conditions of VHTR. Therefore, a suitable SEM compatible with the MTMF system at NCSU is proposed to be acquired to develop an ISTL to address high temperature nuclear materials and ASME Code issues. With the availability of such a ISTL, uniaxial and multiaxial cyclic experiments prescribing realistic thermo-mechanical fatigue (TMF), creep and creep-fatigue loading can be performed on specimens of VHTR materials, such as Alloy 617, 316H, 800H, Grade 91 steel, for addressing the high temperature component design and development issues. Finally, because of the size of commercially available TMF systems, these cannot be used for in-situ SEM testing, which is essential for investigating existing alloys and developing new alloy for VHTRs. Hence, acquisition of a SEM will give the NCSU research community unprecedented capability to perform fundamental research and educate next generation scientists in studying real-time long-term microstructure evolution of nuclear materials under uniaxial and multiaxial loading. In addition, the proposed equipment will allow training undergraduate and graduate students and postdocs in performing material characterization using advanced techniques and provide hands on experiences to students in various undergraduate and graduate courses.