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DOE OSTI2026

Harvesting Isotopically Pure Ac-225 from Ra-225 Produced in Thorium Target Spallation Reaction

A method for isolating Ra-225 from linear accelerator irradiated Th metal chemical processing waste streams has been established. Separating Ra-225 from the complex matrix following proton irradiation of Th metal provides a source of isotopically pure Ac-225 free of the Ac-227 impurity that is present in accelerator-produced Ac-225. Ra isotopes are first separated from the irradiated Th target via cation exchange methods. Importantly, the Ra-containing fraction is cleaned from residual citrate via dilute nitric acid washes (0.5 M) on AG50W-X8 resin. The mixed Ra isotopes are then further purified from Ba-140 impurity using a Sr resin column. Ra elutes through the resin with successive washes of 2.5 M HNO 3 while Ba-140 is retained. Ac-225 is then harvested by loading the Ra content onto an AG50W-X8 cation exchange column where the Ra-225 parent is eluted with 2.5 M HNO 3 before the Ac-225 daughter is eluted with 8 M HNO 3 .

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2025

Nanoscale clustering and fission product segregation in irradiated annular U-10Zr fuel

Zirconium (Zr) is added to uranium (U) to improve the performance of metallic fuel for fast reactor applications. This study employs transmission electron microscopy (TEM) and atom probe tomography (APT) to investigate nanoscale clustering of U and Zr, as well as segregation of fission products (FPs), in annular U-10Zr (in weight) metallic fuel irradiated at the Advanced Test Reactor (ATR). The results reveal variations in the shape, size, and chemical composition of clusters at different radial locations within the irradiated fuel cross-section. Zr-rich clusters exhibit higher concentration of FPs compared to U-rich clusters, potentially due to the co-precipitation of Zr and FPs in the fuel matrix during cooling at the end-of-life. In conclusion, this work complements the study of fuel constituents and fission product distribution across multiple length scales in irradiated U-10Zr metallic fuel.

Atom probe tomography↗

DOE OSTI2023

Removal of fission product tellurium and iodine from recently irradiated HEU using copper metal

Gamma detection of select fission products or activation products such as 153 Sm and 239 Np in a fission sample is difficult due to gamma interferences from radiotellurium and radioiodine. To aid in the detection of these difficult-to-detect nuclides, the radiotellurium and radioiodine were removed using Cu metal turnings. This method is highly effective at the removal of the interfering radionuclides removing > 98% of the radiotellurium and radioiodine. A mechanistic study was conducted. Determining the deposition mechanism, using a number of techniques including XANES and electrochemistry providing an understanding of the electrochemical path for the Te deposition. Initial work focused on the use of stable Te, further expanding into activated TeO 2 , and a fresh fission sample. Importantly, this method presents a rapid method for the direct detection of select fission products that could have a significant impact on nuclear related fields including nuclear forensics.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2023

Irradiation Effect on Noble Metal Particles in Water Using in situ Liquid Cell STEM Observation

During geologic disposal of spent nuclear fuel (SNF) in an engineered nuclear waste repository, once all other barriers have degraded, oxidizing may occur at the solid-water interface owing to a self-generated radiolytic field. The repository design includes large quantities of iron (Fe), that is anticipated to corrode under an anoxic environment, and generate hydrogen (H 2 ) gas. This H 2 gas is thought to be able to suppress the dissolution of SNF through a catalytic reaction with noble metal particles (NMP) that are pre-existing in the SNF. This interaction leads to the decomposition of the major oxidant, hydrogen peroxide (H 2 O 2 ). In conclusion, these processes are described in the Fuel Matrix Degradation (FMD) model that is being used to predict SNF degradation rates. The NMP, therefore, plays an important role within the FMD model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1969

Production of metals and compounds by radiation chemistry

Preparation of metals and compounds by radiation induced chemical reactions involves irradiation of metal salt solutions with high energy electrons. This technique offers a method for the preparation of high purity metals with minimum contamination from the container material or the cover gas.

Marsik, S. J.↗

DOE OSTI2023

Phase patterning of metallic glasses through superfast quenching of ion irradiation-induced thermal spikes

Amorphous metallic glasses (MGs) convert to crystalline solids upon annealing at a high temperature. Such a phase change, however, does not occur with the local melting caused by damage cascades introduced by ion irradiation, although the resulting thermal spikes can reach temperatures > 1000 K. This is because the quenching rate of the local melting zone is several orders of magnitude higher than the critical cooling rate for MG formation. Thus the amorphous structure is sustained. This mechanism increases the highest temperature at which irradiated MG sustains amorphous phase. More interestingly, if an irradiated MG is pre-annealed to form a polycrystalline structure, ion irradiation can locally convert this crystalline phase to an amorphous phase if the grains are nanometers in size and comparable to the damage cascade volume size. Combining pre-annealing and site selective ion irradiation, patterned crystalline-amorphous heterogeneous structures have been fabricated. This finding opens new doors for various applications.

36 MATERIALS SCIENCE↗

DOE OSTI2025

A methodology for corrosion testing ODS steels in liquid tin under reactor conditions in HFIR

Liquid metal–based divertor concepts have promising attributes for application in fusion reactors. Liquid tin is being considered as one of those liquid metals due to its excellent thermophysical properties. However, tin is extremely corrosive to steels at elevated temperatures, and the coupled effects of corrosion with neutron irradiation for application in fusion reactors have not been quantified. Researchers at Oak Ridge National Laboratory (ORNL) have designed and irradiated a series of capsules in the High Flux Isotope Reactor (HFIR) to help address gaps in the literature with respect to coupled corrosion and neutron irradiation effects. The capsules are filled with solid tin designed to melt due to the gamma heating in HFIR and allow interaction with the specimens. The capsules contain a SiC thermometer for post-irradiation temperature verification. Five capsules were irradiated in HFIR for 10.5 days, accumulating 0.93 dpa in the FeCrAl specimens. Post-irradiation temperature verification was performed on the SiC thermometers, and temperatures were observed to be higher than what was expected from thermal models. An explanation for the model underprediction is given herein. In conclusion, the capsules described within this article are the first to irradiate liquid metal within the HFIR flux trap and demonstrated the safety basis for continued liquid coolant studies in HFIR.

FeCrAl↗

DOE OSTI2023

Advanced characterization-informed machine learning framework and quantitative insight to irradiated annular U-10Zr metallic fuels

Abstract U-10Zr Metal fuel is a promising nuclear fuel candidate for next-generation sodium-cooled fast spectrum reactors. Since the Experimental Breeder Reactor-II in the late 1960s, researchers accumulated a considerable amount of experience and knowledge on fuel performance at the engineering scale. However, a mechanistic understanding of fuel microstructure evolution and property degradation during in-reactor irradiation is still missing due to a lack of appropriate tools for rapid fuel microstructure assessment and property prediction based on post irradiation examination. This paper proposed a machine learning enabled workflow, coupled with domain knowledge and large dataset collected from advanced post-irradiation examination microscopies, to provide rapid and quantified assessments of the microstructure in two reactor irradiated prototypical annular metal fuels. Specifically, this paper revealed the distribution of Zr-bearing secondary phases and constitutional redistribution across different radial locations. Additionally, the ratios of seven different microstructures at various locations along the temperature gradient were quantified. Moreover, the distributions of fission gas pores on two types of U-10Zr annular fuels were quantitatively compared.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the Advanced Test Reactor

The Advanced Test Reactor (ATR) is a light water reactor with aluminum-clad driver fuel. A primary mission of the ATR is to support the next generation of nuclear reactors. This support necessarily requires irradiation of advanced materials such as sodium, fuel salts, and metal eutectics. Irradiation of advanced materials in the ATR environment presents a challenge when completing accident analyses and demonstrating compliance to the Safety Analysis Report (SAR). Many advanced materials have the possibility to react with the ATR protective barriers such as the cladding or primary coolant system (PCS) boundary during postulated accident scenarios. Further, molten fuel experiments fall outside of the standard regulatory framework for dose consequence analyses. ATR is currently developing new safety analysis methods to support irradiation of advanced materials. The primary considerations for this development are 1) experiment containment design requirements, 2) primary coolant system response to an experiment containment failure, and 3) dose analyses for molten fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the ATR

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Recent Metallic Fuel Data Recovery in FIPD

The Metallic Fuels Irradiation and Physics Database (FIPD) [1] is an organized collection of metallic fuel test pin data (U-xPu-yZr, = 0 ~ 28; y = 2 ~ 10) and documentation available to industry. FIPD mainly contains three types of data: (1) Fuel pin fabrication data, including fuel slug diameter, fuel slug length, cladding diameter, smear density, etc. (2) Fuel pin operation conditions, including axial distributions for power, temperatures, fluences, burnup, and isotopic densities, etc. and (3) Fuel pin post-irradiation examination (PIE) data, including fission gas release and gas chemistry, profilometry, and neutron radiography, etc. The operating conditions for pins with PIE data available in FIPD span significant ranges across key parameters. The fuel peak burnup extends from less than 5% up to 20 at%. The cladding peak temperature varies from about 490°C to 660°C. Finally, the cladding peak DPA shows a wide range from less than 5 to 120. These broad ranges reflect the diverse testing conditions and operational parameters captured in the available PIE data. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). As reported in the ref. [3, 4], most of the PIE data generated during the IFR program [5] has been collected, reviewed, processed, and integrated into FIPD. The most recently added PIE data can be found in ref. [4], which shows the collection of over 95% of the PIE data by the time of this paper. The recent improvements to the database are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Legacy Metallic Fuel U(Pu)Zr Data Qualification

The Metallic Fuels Irradiation and Physics Database (FIPD) is an organized collection of legacy metallic fuel U(Pu)Zr data, measurements, and reactor conditions from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The database provides a wealth of information and is easily accessed and utilized by the U.S. nuclear industry and the Nuclear Regulatory Commission (NRC). The database contains three categories of data. The first category covers fuel pin fabrication specifications, including fuel slug diameter and length, cladding diameter, smeared density, etc. The second consists of operational conditions, which include axial distributions of power, temperature, fluence, burnup, isotopic density, etc. The third category contains post-irradiation examination (PIE) results, largely collected at facilities now part of the Idaho National Laboratory (INL), formerly part of Argonne National Laboratory -West (ANL-W) and at the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory, comprising fission gas release and chemistry analysis, profilometry measurements, neutron radiography data, etc. All these data have been organized and preserved in FIPD.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Fuel-Cladding Eutectic Study of Legacy Fast Flux Test Facility (FFTF) MFF HT9/U-10Zr Metallic Fuel

This report presents the first systematic investigation of fuel-cladding eutectic interaction (FCEI) in irradiated HT9/U-10Zr metallic fuel from the Fast Flux Test Facility (FFTF) Materials Fuels Form (MFF) program, using differential scanning calorimetry (DSC) coupled with scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Two irradiated fuel cross-sections, MNT07H (9.5 at% burnup, x/L = 0.78) and MNT08H (7.0 at% burnup, x/L = 0.93), were subjected to three successive isothermal annealing rounds (R1–R3) at 820°C for 20 minutes each, yielding a cumulative transient duration of one hour. This study directly addresses a recognized gap in the existing FCEI database, which previously lacked irradiated HT9/U-10Zr data at burnup levels above 8 at%. Two principal findings emerge from the study. First, for both samples, FCEI remained spatially confined within the pre-existing fuel-cladding chemical interaction (FCCI) zone boundaries after R3, with no measurable eutectic penetration into unaffected cladding beyond the original FCCI layer. This self-limiting behavior is consistent with historical Fuel Behavior Test Apparatus (FBTA) results and is attributed to the near-eutectic phase composition of the FCCI zone, which rapidly absorb the available eutectic-forming constituents and then stall penetration once the FCCI zone is consumed. A comparison with unirradiated surrogate data further supports this mechanism: whereas a U–34 at.% Fe sample would be expected to show ~176 µm of iron penetration under comparable conditions, the irradiated samples exhibited only ~20 µm, a discrepancy attributed to irradiation-induced interfacial porosity and pre-existing FCCI composition gradients. Second, for MNT08H, FCEI was observed exclusively on the half of the cladding circumference where pre-existing steady-state FCCI was present, with no detectable FCEI on the opposite half. Three hypotheses are proposed to explain this asymmetry: the inhibiting role of a zirconium-rich rind at the fuel-cladding interface; the chemical sequestration of iron by redistributed zirconium within the fuel matrix; and the persistence of fuel-cladding gaps on the FCEI-free half that preclude direct contact. All three hypotheses require further experimental investigation. The results extend the empirical FCEI database into higher-burnup territory and demonstrate the viability of DSC-based testing as a substitute for the no-longer-available FBTA apparatus. Future work will include additional cross-section testing, compilation of the full FCEI dataset, model evaluation, and DSC testing of ternary fuel compositions to broaden the experimental basis for safety assessment of sodium-cooled fast reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Synthesis and characterization of electron beam irradiation-induced damage in polycrystalline metal thin films

High-energy physics research, industrial sterilizing, and material processing depend extensively on electron beam accelerators. Exit windows are crucial components of such electron accelerator systems, maintaining vacuum integrity inside the machine while providing mechanical strength, thermal stability, and radiation resistance at the beam-target interface. In this study, thin metallic films of Ni, Ti, Cr, and V were explored for use in electron beam exit windows, and their properties were compared with the properties of their bulk counterparts. Simulation results of metal foils predicted Ti to exhibit less beam power dissipation compared to Ni. However, Ni possesses superior mechanical and structural properties compared to Ti. The performance of these films under electron beam irradiation was examined by depositing thin layers of these films on silicon and metallic substrates using magnetron sputtering and exposing them to e-beam irradiation in a controlled setup. The deposited films were subjected to a dose of approximately 66 kGy at a beam energy of 10 MeV and characterized prior to and postexposure to the beam using field emission scanning electron microscopy, atomic force microscopy, x-ray diffraction (XRD), and nanoindentation. Particular emphasis was given to characteristics like the grain structure, surface morphology, dislocation density, and hardness. XRD patterns revealed irradiation-induced changes in peak intensities, while the crystallinity remained largely unchanged. Nanoindentation results showed that the pristine and irradiated Ti and Ni films were twice as hard when compared to bulk Ti and Ni samples, regardless of the substrate type used (Si or bulk Ti, or Ni). These results emphasize the superior mechanical properties of thin metal films compared to their bulk counterparts. In conclusion, this study advances the optimization of thin film materials for robust and efficient e-beam applications, ensuring improved durability and operational reliability.

AFM↗

NASA NTRS1974

Stress waves generated in thin metallic films by a Q-switched ruby laser

Investigation results on stress waves generated by Q-switched ruby laser irradiated thin metal films under confinement, studied over a wide range of film materials and film thicknesses, are reviewed. The results indicate that the dependence on these parameters is much weaker than is predicted by heat transfer estimations commonly used to describe the interaction of laser irradiation with unconfined bulk-solid surfaces.

Yang, L. C.↗

DOE OSTI2025

Determination of Scale Bar for AGHCF Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) has supported the development of several databases containing information on the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data represents one of the most critical types of post-irradiation examination (PIE) data being collected, organized, and archived in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Metallic Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three principal sets of metallography data. The first set comprises metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins examined in the Hot Fuel Examination Facility (HFEF). The second set consists of metallography data from EBR-II irradiated fuel pins examined in the Alpha-Gamma Hot Cell Facility (AGHCF). The third set includes metallography data from transient-tested fuel pins (including both out-of-pile furnace tests and TREAT tests) examined in AGHCF. Since both the second and third sets were generated in AGHCF, they are governed by identical specifications. The metallography data in the databases consist of digital images scanned from either positive or negative photographic films. To analyze the microstructure of a fuel pin, a series of preparatory steps are required, including sectioning, epoxy mounting, mechanical grinding and polishing, and etching. Following sample preparation, specimens are transferred for metallographic examination. The AGHCF and HFEF metallography data were generated using optical microscopes manufactured by Leitz and Bausch and Lomb (B&L). Images were recorded on Polaroid film at preset magnifications. Magnification verification for the Leitz and B&L metallographs was conducted every two months prior to 1989 and at least every six months from 1989 through the conclusion of the IFR program. Magnifications determined from imaging of microslide standards were compared to the instrument settings for magnifications ranging from 50× to 500×. If the magnifications determined from standards deviated from the instrument settings, adjustments were made to the bellows extension until agreement was achieved. The specifications for AGHCF and HFEF legacy metallography data have been established based on available hard-copy and digital records, most of which have been incorporated into the data repositories associated with FIPD, OPTD, and TREXR. Detailed specifications including hard-copy records, digitized records, cutting diagrams and sectioning schemes, high-magnification photographs, photomosaics (composites), information tags, scale bars, and magnification verification procedures can be found in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Fuels Irradiation and Physics Database (FIPD) Development and Data Qualification FY24 Updates

The DOE-NE Advanced Reactor Technologies program has supported efforts to recover and preserve metallic fuel data generated throughout the past fast reactor R&D programs. Those efforts are currently focused on establishing databases for information from the experiments conducted during the Integral Fast Reactor program including data generated at EBR-II, FFTF, and TREAT reactors, as well as out of pile transient testing data. These databases are essential for future-licensing activities of metallic fuel based advanced fast reactors. This report describes the Metallic Fuels Irradiation & Physics Database (FIPD) development, content updates, data qualification status, and user support in FY24. Future plans are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Impact of Low-Temperature Neutron Irradiation on Tensile Behavior of Base Metal and Electron-Beam Welded 316L Stainless Steel

Here, a study of 316L type stainless steel in its base metal and electron-beam (e-beam) welded conditions was performed to observe the effects of low-temperature (60 °C to 100 °C) neutron irradiation on the tensile behavior of the samples. Fractography was used in understanding the tensile-tested fracture surfaces of the 316L samples in these different forms with the characterization of the both base metal and welded samples using electron microscopy. Irradiation of the tensile specimens made free of defects of cutting and mechanical polishing showed a reduction in their tensile ductility with increased radiation-induced hardening up to 1.40 × 10 19 n/cm 2 (E > 0.1 MeV) fluence that corresponds to 1.1x 10 -2 dpa, even at the low irradiation temperatures. These low-temperature neutron irradiated base metal and e-beam welded 316L specimens also consisted of closely similar fracture surfaces characteristic of ductile rupture.

36 MATERIALS SCIENCE↗