Current Research Projects

“Fundamental Understanding of Transport Under Reactor Extremes 2.0” (New Award, 2022-2026)

(Sponsor: Department of Energy – Energy Frontier Research Centers) (PI for NCSU)

Participants: Los Alamos National Laboratory (Lead), University of California, Berkeley, Bowling Green State University, North Carolina State University, University of Virginia, University of New Mexico, Pacific Northwest National Laboratory

The center works on “Fundamental Understanding of Transport Under Reactor Extremes (FUTURE)” was just renewed for another 4 years (2022-2026), with the goal to understand the coupling between radiation damage and corrosion and predict irradiation-assisted corrosion in passivating and non-passivating environments for materials in nuclear energy systems with more emphasis on effects of multiphase on the corrosion processes.

More info on the dedicated website

Nuclear Science & Security Consortium (2021 – 2026)

Funding Agency: US Department of Energy’s National Nuclear Security Administration (DOE NNSA);

Consortium led by University of California, Berkeley; Collaborators: Air Force Institute of Technology; George Washington University; Michigan State University; Texas A&M University; University of California, Davis; University of Illinois, Urbana-Champaign; University of Nevada, Las Vegas; University of New Mexico; University of Tennessee, Knoxville; Los Alamos National Laboratory; Lawrence Berkeley National Laboratory; Lawrence Livermore National Laboratory; Oak Ridge National Laboratory; Sandia National Laboratories; Pacific Northwest National Laboratory;

This consortium will carry out R&D in five research focus areas: nuclear physics and nuclear data; radiochemistry and nuclear chemistry; nuclear material science; radiation detection; nuclear chemical engineering and nuclear engineering.  Linking these research areas are two cross-cutting efforts: computing and optimization in nuclear applications; and education in nuclear science, technology, and policy.

Further details on this Consortium is available here.

Microstructure Optimization and Novel Processing Development of ODS Steels for Fusion Environments  (2020 – 2023)

Funding Agency: US Department of Energy through the Advanced Research Projects Agency – Energy; Consortium led by Pacific Northwest National Laboratory; Collaborators: NCSU, Ames Laboratory, Pacific Northwest National Laboratory.

This project’s objective is scalable, cost-effective fabrication of high-performance, oxide-dispersion strengthened (ODS) steel with advanced-manufacturing methods for fusion blanket-breeding applications. Gas atomization reaction synthesis (GARS) enables the synthesis of precursor ODS steel powders without prolonged mechanical alloying. This process creates a chromium (Cr)-enriched surface oxide with yttrium/titanium (Y/Ti)-enriched intermetallics in powder interiors. The goal will be to consolidate and extrude GARS powder in one step using first-of-a-kind shear assisted processing and extrusion (ShAPE) and laser-based AM processes. Such scalable, cost-effective fabrication of ODS steels may enable efficient power conversion cycles (≥40%) at operating temperatures beyond 900 K in future fusion power plants.

Passive Strain Measurements for Experiments in Radiation Environments (2020-2023)

Funding Agency: US Department of Energy through LDRD led by Idaho National Laboratory; Collaborators: Idaho National Laboratory, Massachusetts Institute of Technology

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

Bridging the atomic scale and the mesoscale in the characterization of defect production and evolution in high entropy alloys (2020-2023)

Funding Agency: National Science Fondation (NSF); Collaborators: Bowling Green State University (BGSU)

High entropy alloys (HEAs) are emerging as an outstanding class of materials due to their excellent mechanical properties and high radiation tolerance as a result of their unique electronic structure. Chemical disorder and compositional fluctuations in these alloys have large effects on energy dissipation and response to irradiation. Thus this project focuses on understanding defect formation and buildup in these alloys through experimental characterization. The proposed research is expected to reveal the effects 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 through the study of defect production from collision cascades on an atomic and mesoscale level in alloys with increasing chemical complexity from one to five constituents.

Corrosion Sensitivity of Stainless Steels in Pressurized Water Reactor Water Chemistry: Can KOH replace LiOH in PWRs? (2020-2023)

Funding Agency: DOE NEUP; Collaborators: University of California Berkeley (UCB); Pacific Northwest National Laboratory (PNNL); Electric Power Research Institute (EPRI)

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.

Synergy of radiation damage with corrosion processes through a separate effect investigation approach (2020-2023)

Funding Agency: DOE; Collaborators: Los Alamos National Laboratory (LANL)

This project focuses on investigating the synergy of radiation damage with corrosion processes through a series of separate effect experiments which will look at the effect of irradiation on Fe-based systems and Ni-Based systems and how radiation damage affect corrosion processes. Fe-based systems are of interest for Liquid Metal Cooled Reactors and Ni-Based systems are of interest for Molten Salt Reactors. We will investigate the effects of ion irradiation on metal and effects of ion irradiation on Metal/Oxide interfaces and post irradiation Atom Probe Tomography.

Simultaneous Corrosion/Irradiation Testing in Lead and Lead-Bismuth Eutectic: The Radiation Decelerated Corrosion Hypothesis (2020 – 2023)

Funding Agency: DOE NEUP; Collaborators: Massachusetts Institute of Technology (MIT); Oxford University (UK)

Liquid lead and lead-bismuth eutectic (LBE) cooled fast reactors promise the best power density and economics for fission reactors, should they actually be deployed. For decades, the issues of corrosion and how it will change with irradiation have been the bottleneck in lead fast reactor (LFR) and LBE fast reactor (LBEFR) deployment, restricting outlet temperatures below 550°C1 . Without precise knowledge of corrosion and irradiation performance of LFR/LBEFR materials, these reactors will never be deployed, stuck forever in a Catch 22. A far faster, yet reactor-accurate, method of combined corrosion/irradiation testing is required. To break this bottleneck, we will test candidate materials from previous studies in a new, simultaneous corrosion/radiation facility2 . Rather than rely on separate long-term corrosion and neutron irradiation, simultaneous exposure followed by microstructural characterization, mechanical testing, and comparison to existing data will rapidly down-select potential alloy candidates and assess how irradiation affects corrosion.  In addition, we will explore a controversial, yet massively impactful scientific hypothesis, that radiation slows corrosion in the LFR/LBEFR based on molten salt testing.

Ni-based ODS alloys for Molten Salt Reactors (2019-2023)

Funding Agency: DOE NEUP; Collaborators: University of California, University of Idaho, Idaho National Laboratory, Oxford University (United Kingdom)

The objective of this work is to (i) propose and develop a new Ni-based ODS alloy that can be used for structural applications in Molten Salt Reactors as the primary material facing the fuel (ii) demonstrate that its high temperature mechanical properties are adequate for MSR operating temperatures, (iii) Demonstrate its enhanced resistance to radiation damage compared to regular nickel alloys as a result of its inherent multi-interface character through “rapid” ion irradiation testing and (iv) demonstrate its improved corrosion resistance in MSR environment through appropriate experiments.

Understanding of degradation of SiC/SiC materials in nuclear systems and development of mitigation strategies (2018-2021)

Funding Agency: DOE; Collaborators: University of California, General Atomics

The objective of this work is to develop the best possible coating composition via rapid throughput processing and testing techniques and conduct a thorough material analysis of the resulting SiC-SiC/coating structure and transfer the obtained knowledge to engineering scalable coating systems. Furthermore, this proposal targets to develop a comprehensive understanding of the effect of true reactor conditions (in terms of chemistry and radiation) and aims to evaluate the best structure/coating system in this environment. We will deliver the best coating possible for deployment on SiC-SiC substrates for LWR environment while providing an avenue for cost effective mass production of cladding tubes.

Fundamental Understanding of Transport Under Reactor Extremes

(Sponsor: Department of Energy – Energy Frontier Research Centers)

Participants: Los Alamos National Laboratory (Lead), University of California, Berkeley, Bowling Green State University, North Carolina State University, Pacific Northwest National Laboratory, University of California, Berkeley, University of Virginia, and University of Wisconsin, Madison.

The center will work on “Fundamental Understanding of Transport Under Reactor Extremes (FUTURE)”  with the goal to understand the coupling between radiation damage and corrosion and predict irradiation-assisted corrosion in passivating and non-passivating environments for materials in nuclear energy systems.

“High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation”

(Sponsor: Department of Energy – Nuclear Energy University Program)

Participants: U. of Michigan, (lead university), U. Tennessee; Penn State; U. of Wisconsin; U. C. Berkeley; U. C. Santa Barbara; ORNL; LANL; LLNL; ANL; INL; TerraPower LLC; EPRI; U. Manchester; U. Oxford; Areva; U. Queens; CEA.

The objective of this collaborative effort is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations.

The promise for developing new, advanced nuclear reactor concepts that significantly improve on the safety, economics, waste generation and non-proliferation security of commercial nuclear power reactors, and the extension of life of existing light water nuclear reactors rests heavily on understanding how radiation degrades materials that serve as the structural components in reactor cores.

Traditionally, research to understand radiation-induced changes in materials is conducted via radiation effects experiments in test reactors (both fast and thermal), followed by a comprehensive post-irradiation characterization plan. This is a very time consuming process because of the low damage rates that even the highest flux reactors exhibit. In addition the high cost of research on irradiated materials in the face of shrinking budgets put additional constraints on this approach.

A promising solution to the problem is to use ion irradiation to irradiate materials to very high doses. The advantages of ion irradiation are many. Dose rates (typically 10-3 to 10-4 dpa/s) are much higher than under neutron irradiation (10-7 to 10-8 dpa/s), which means that 100s of dpa can be reached in days or weeks instead of years. Because there is little activation the samples are not radioactive. Control of ion irradiation experiments is much better than experiments in reactor.

Challenges to the implementation of ion irradiation as a surrogate for neutron irradiation include rate effects, small irradiation volumes, accounting for transmutation and the lack of data to establish the equivalence. Addressing these challenges constitutes the main focus of this program. This project will demonstrate the capability to evaluate the behavior of reactor materials at high irradiation doses. This effort includes a benchmarking of the microstructures formed under ion irradiation and neutron irradiation and the resulting mechanical properties by a combined experimental and analytical approach. The outcome of this program will be the establishment of the conditions by which ion irradiation can be used as a surrogate for neutron irradiation in reactor.

The project involves characterization of alloys irradiated by single ion irradiation, dual beam ion irradiation and neutron irradiation involves TEM, chemiSTEM, and APT techniques.

Correlation of ChemiSTEM characterization and conventional TEM observation at the same area of ion irradiated HT9 showing radiation-induced precipitation of Ni/Si/Mn-rich precipitates.

“Mechanical behavior of advanced alloys for high temperature applications”

The main objective is to understand deformation and fracture mechanisms in high-temperature structural materials to ensure structural integrity and lifetime prediction of the components (advanced steels, Ni-based alloys, SiC). The lack of understanding of the deformation mechanisms in such marerials has limited the development of predictive capabilities. Of interest are the study of dynamic strain ageing, PLC effect observed at intermediate temperature ranges, and dynamic recrystallization observed at high temperatures. We use an Environmental Mechanical Testing Machine with Digital Image Correlation Capability.

The goal is to elucidate the mechanisms of deformation and creep resistance in these high temperature advanced materials using a mechanistic approach which will allow for predictive mechanistic models to be derived.

The study includes in-situ straining experiments to evidence the mechanisms of dislocation dynamics.

Dislocation cross slip and annihilation in Alloy 617


Related publications:

– Kaoumi D., K. Hrutkay, “Tensile deformation behavior and microstructure evolution of Ni-based superalloy 617”, Journal of Nuclear Materials, 454, 2014, p 265-273.

– Hrutkay K., D. Kaoumi, “Tensile deformation behavior of a nickel based superalloy (Haynes 230) at different temperatures”, Materials Science & Engineering A, 599, 2014, p. 196–203


“Mechanistic and Validated Creep/Fatigue Predictions for Alloy 709 from Accelerated Experiments and Simulations”

(Sponsor: Department of Energy – Nuclear Energy University Program)
Collaborators: NCSU, ANL

As a promising candidate for fast reactor program, Alloy 709 possesses excellent high temperature thermomechanical properties. To support its qualification in the ASME code for Class 1 Components in Elevated Temperature Service (Section 3, Division 1, Subsection NH), we propose mechanistic methods for predicting creep and creep-fatigue deformation rates based on accelerated in-situ and ex-situ tests, and mesoscale dislocation dynamics (DD) simulations. The research work performed in this project will aim at obtaining: (i) creep and creep-fatigue data (ii) microstructure evaluation from (a) in-situ/ex-situ TEM, (b) in-situ XRD using synchrotron radiation at APS/ANL and (c) mesoscale dislocation dynamics simulations informing creep damage mechanics (CDM) model; (iii) a rational framework (CDM) of generalized viscoplastic constitutive equations to reliably predict and extrapolate the results of accelerated tests to reactor operating conditions; (iv) validations of CDM performed through predictions that can be crosschecked and benchmarked against experimental data; and (v) extrapolated creep and creep-fatigue data delivered for use in ASME code development.

In-situ Synchrotron experiments are done where the sample is stressed at temperature while the Diffraction Patterns are continuously collected.


“Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage”

(Sponsor: Department of Energy – Nuclear Energy University Program, IRP)
Collaborators: Colorado School of Mines (leading university), NCSU, ANL, LANL, SNL, CB&I

Chloride-initiated stress corrosion cracking (CISCC) of spent fuel canister (primarily in welds or heat affected zones) is one of the safety concerns during the dry storage of used nuclear fuel at an Independent Spent Fuel Storage Installations (ISFSIs). Deterioration by CISCC can lead to canister penetration, potentially releasing helium and radioactive gases, and permitting air ingress which could pose a threat to fuel rod integrity.

This study will result in enhanced understanding of conditions which could be conducive to CISCC initiation (such as pitting) or CISCC propagation rate, and will develop methods that could be used to identify the occurrence of CISCC in its early stages in the field. The model and methodology developed in the proposed project with quantified uncertainty can be used to inform recommendations for periodic NDE examinations to monitor the extent of any cracking.

Laboratory CISCC studies envisioned in the work frame of this IRP include: Testing to quantify the effect of environmental and metallurgical factors that have an impact on SCC initiation and growth rate, such as salt concentrations, temperature, most susceptible heat affected zone microstructure, metal stresses, pH, and relative humidity; Experimental testing to determine the most susceptible zone within the weld heat affected zones; Controlled and instrumented pitting initiation and crack propagation rate studies with specimens representing the most susceptible microstructure, varying environmental parameters and specimen tensile stress conditions to cover the range expected on the canister surfaces.

Synchrotron micro-tomography of stress corrosion cracks is used to investigate the path and mechanisms of crack propagation. As an illustration of the method, below is the 3D rendering of SCC crack and crack branches in a 304H washer (video)

Past Projects

“Developing Ultra-Small Mechanical Testing Methods and Micro Developing Ultra-Small Mechanical Testing Methods and Microstructural Investigation Procedures for Irradiated Materials”

Participants: University of California, Berkeley; Los Alamos National Laboratory.

Both light water and advanced reactor concepts call for advanced materials understanding and research since both rely on high performance materials in this harsh environment. High temperatures, long deployments, high radiation doses and corrosion make a materials selection in these environments particularly difficult. For new alloying ideas, it is highly desirable to only perform small experimental heats using smaller and smaller materials testing in order to avoid the costs of manufacturing large quantities. Accelerated materials testing, is important in order to achieve high doses quickly to enable new materials concepts under radiation and lead the way towards their qualification. Most accelerated materials testing approaches involve ion beam irradiation or high dose neutron irradiation. Ion beam accelerators only have a limited penetration depth into a material (allowing only μm of irradiated materials on a given sample). On the other hand, neutron irradiated materials are difficult to deal with due to activation concerns and there is often only a limited amount of material available. Regardless, both approaches call for the development of small-scale materials testing techniques and the need to link these techniques to bulk properties.

Therefore, the development of novel small-scale mechanical testing in combination with microstructural investigation and modeling is of great interest to the nuclear materials community for both materials development as well as monitoring applications. In this work the combination of modeling and experiments on multiple length scales is used to evaluate and improve existing small scale mechanical testing techniques in order to help make them relevant to macroscopic properties and useful nuclear engineers, inspectors and designers.

The goal of this project is to develop new small-scale mechanical testing techniques to allow for the estimation or direct measurement of bulk properties. The outcome of combined experiments and modeling will significantly enhance the statistics and information that can be obtained on small radioactive archived samples as well as new ion beam irradiated specimens. As part of this effort, in situ experiments allows us to understand mechanisms of materials deformation.

“Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Simulation”

Participants: University of Tennessee, University of Wisconsin, Penn State.

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Predictive modeling relies on an understanding of the development of microstructure and microchemical evolution under irradiation. This project focused on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ TEM irradiation experiments that can provide validation and benchmarking to the computer codes.

Denuded zones developing at grain boundaries during an in-situ ion irradiation irradiation in a TEM (D. Kaoumi, J. Adamson, M. Kirk, Journal of Nuclear Materials, 445 (1–3): p. 12–19, 2014).