Current Research Projects

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).