An investigation of the basic properties of irradiated polyethylene memory materials
Properties of irradiated polyethylene memory materials
Greer, D.↗
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Results for “material irradiation”
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Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4
This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr) 2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr) 2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.
36 MATERIALS SCIENCE↗
In-situ and ex-situ characterization of ion-irradiated AM materials
Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Design Features of TMIST-3 Experiments that Assist with Tritium Management
Design features that assist with tritium management in the TMIST-3 materials irradiation experiments were presented at the 4th DEVICE-MTR workshop at Pacific Northwest National Laboratory May 2022.
99 GENERAL AND MISCELLANEOUS↗
Radiation Effects on Uranyl Tetrachloro Coordination Compounds: Impact of Lattice Water
Nuclear materials, such as uranium-bearing solids, are exposed to high levels of ionizing radiation throughout the nuclear fuel cycle; thus, it is important to develop a molecular-level understanding of how these materials behave and degrade in the presence of gamma (γ) irradiation. In the current study, three U(VI) tetrachloride complexes, M 2 [UO 2 Cl 4 ]·xH 2 O (where M = K + , Rb + , or Cs + and x = 0 or 2), and their respective chloride salts were exposed to 1–50 kGy of γ radiation using a 60 Co source. Irradiated materials were evaluated by using electron paramagnetic resonance (EPR) and Raman spectroscopy and were further explored by using density functional theory (DFT) methods. EPR spectra of the irradiated materials suggest the formation of a Cl-based radical for both the alkali salts and the uranyl tetrachloride compounds, and DFT calculations provide evidence that the Cl 2 –• radical is formed within these materials. The presence of water in the K + and Rb + compounds leads to additional spectroscopic signatures that could be traced back to water radiolysis and the formation of peroxide and superoxide species. DFT results support the formation of HO 2 • in the lattice and potentially the formation of a [UO 2 Cl 3 (O 2 )] 3– species, highlighting the impact of water within the hydrated material to alter U(VI) speciation by radiolysis.
36 MATERIALS SCIENCE↗
Deformation localisation in ion-irradiated Fe and Fe10Cr
Determining the mechanisms for irradiation-induced ductility loss is crucial for the design of reactor structural components. Here, the deformation characteristics around nanoindents in Fe and Fe10Cr irradiated with Fe ions to ~1 displacement-per-atom at 313 K are non-destructively studied. Slip steps surrounding the nanoindents indicate that deformation is localised in the irradiated materials. Lattice rotation and strain fields near the indent site show over 87% confinement of plasticity in the irradiated material. Cr has little effect on the irradiation-induced changes in pile-up topography and deformation fields, suggesting it has limited impact on retaining strain hardening capacity and reducing irradiation-induced embrittlement.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens
Reactor developers continue to recognize opportunities for further enhancing fast spectrum reactor designs with advanced core materials, but all the material test reactors currently available to the United States are thermal spectrum designs. Fortunately, the Advanced Test Reactor and High Flux Isotope Reactor are versatile high flux facilities where spectral modification strategies can be used to reduce undesirable thermal neutron capture transmutation damage and augment fast flux delivered to specimens. New opportunities to leverage high flux regions and specially designed fast flux boosting experiment configurations can be used to achieve meaningful fast fluences on large specimens in ATR. New optimization potentials can be employed to achieve even higher fluences, albeit for smaller specimens, using thermal neutron filters in HFIR test positions. These capabilities, while not true fast reactors, can provide highly relevant environments for researchers needing to study the effects of fast neutron damage in bulk material specimens.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Local measurement of bulk thermal diffusivity using photothermal radiometry
Here, an experimental methodology using photothermal radiometry is developed for the accurate measurement of bulk thermal diffusivity of nuclear fuels and materials irradiated to high doses. Under these conditions, nuclear fuels, such as uranium oxide, and moderator materials, such as graphite, become friable, which requires characterization techniques that can accommodate irregularly shaped fragments. Photothermal radiometry, a good candidate for this application, involves locally heating a sample by using a laser and measuring the temperature field by monitoring blackbody radiation. The interaction volume for this study, less than a millimeter, is carefully chosen to sample a statistically significant number of large-scale structural features, such as pores and gas filled bubbles, and is small enough that the sample fragments can be treated as a thermal half-space. The thermal diffusivity standards considered in this study cover a range of thermal diffusivities representative of both fresh and spent nuclear fuels. We also consider a sample having a porous microstructure representative of large-scale structures found in materials irradiated to high doses. Our measurement methodology circumvents complex thermal wave models that address optical diffraction, nonlinear transfer function associated with blackbody radiation, and finite sample size effects. Consequently, the large measurement uncertainty associated with modeling these effects can be avoided. While the emphasis here is on nuclear fuels and materials, this measurement approach is well suited to measure thermal transport in a variety of technologically important materials associated with advanced synthesis techniques. Examples range from small, exotic single crystals grown using hydrothermal growth techniques to additively manufactured components having complex geometries.
36 MATERIALS SCIENCE↗
Radiological Handling and Containment Considerations in Support of an American Medical Isotope Producer
This work will discuss evaluation and design of new and modified facility level radiological containment, transfer, and handling systems in support of the United States’ American Medical Isotope Production Act. In the United States alone, Molybdenum-99 (Mo-99) is used as a precursor to locally generate its decay product, technetium-99m (Tc-99m), for use in approximately 40,000 medical procedures every day to detect cancer and heart disease. Currently, however, Mo-99 is primarily produced overseas and typically requires highly enriched uranium (HEU) which is classified as weapon -usable. This poses a serious security risk to the United States and, in 2012, Congress passed the American Medical Isotope ACT which aimed to cease all Mo-99 production that requires HEU and support American based companies who have found safer production methods. Through this effort, a system was designed and developed to access irradiated material in processing hot cells for ease of material introduction and removal in the confined area. The proposed solution uses a double-door design that allows for easy access to material while also providing shielding through a mechanically manipulated carousel. A system was also designed to transport irradiated materials from target reactors to a target receipt box using a rail guide system, pneumatic motor, and lead screw drive.
07 ISOTOPE AND RADIATION SOURCES↗
Status on Development of Graphite Analytical Tool (GAT)
The DOE-ART Graphite R&D program has been generating significant amounts of irradiated and unirradiated graphite data since 2006 when the program was part of the DOE NGNP (Next Generation Nuclear Plant) Project. This data includes critical irradiation creep and irradiated material property changes from the Advanced Graphite Creep (AGR) experiment as well as significant amounts of data on unirradiated material property values on several current nuclear graphite grades (Baseline program). Previously, Idaho National Laboratory has developed an internal analysis tool to assist with analysis of the unirradiated and irradiated data. The Graphite Analytical Tool (GAT) is intended to provide easy access to the graphite data in the form of comparing unirradiated and irradiated material property changes, comparison of material property differences between various nuclear graphite grades, and illustrate trends within the irradiated and unirradiated data generated within the DOE-ART Graphite R&D program. This report summarizes the progress to-date on the development of this analytical tool.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FY2022 Progress Report for Advanced Re-fabrication/Re-instrumentation Capability Development
In support of performing follow-on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and re-instrumentation capability. As a result, the United States (U.S.) Department of Energy (DOE) has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory (INL). The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of reinstrumentation and the progress to-date. Halden spent nearly 30 years developing both refabrication and re-instrumentation. Collaboration with Halden is allowing Idaho National Laboratory (INL) to develop this capability much more rapidly. In FY-21 the results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. In FY-22, the results include; Completing set up of the drilling and welding modules procured from Halden, and early experimental trials using that equipment.; Completing set up of an out-of-cell circumferential weld system to allow for further weld development to take place and support fabrication of fuel for fresh fuel experiments.; Evaluation of the Hot Fuel Examination Facility (HFEF) infrastructure to support future installation of advanced re-fabrication/re-instrumentation equipment, specifically related to necessary infrastructure for cryo-drilling.; Developments in dry-drilling alternative. Including experimental studies showing cordierite is the most suitable surrogate for UO 2 for performing drilling studies. That drilling performance is enhanced when a fuel-clad bonding condition is simulated.; Conceptual design completed for attaching surface thermocouples to irradiated fuel in support of TWIST capsule experiments. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Proton irradiation on materials
A computer code is developed by utilizing a radiation transport code developed at NASA Langley Research Center to study the proton radiation effects on materials which have potential application in NASA's future space missions. The code covers the proton energy from 0.01 Mev to 100 Gev and is sufficient for energetic protons encountered in both low earth and geosynchronous orbits. With some modification, the code can be extended for particles heavier than proton as the radiation source. The code is capable of calculating the range, stopping power, exit energy, energy deposition coefficients, dose, and cumulative dose along the path of the proton in a target material. The target material can be any combination of the elements with atomic number ranging from 1 to 92, or any compound with known chemical composition. The generated cross section for a material is stored and is reused in future to save computer time. This information can be utilized to calculate the proton dose a material would receive in an orbit when the radiation environment is known. It can also be used to determine, in the laboratory, the parameters such as beam current of proton and irradiation time to attain the desired dosage for accelerated ground testing of any material. It is hoped that the present work be extended to include polymeric and composite materials which are prime candidates for use as coating, electronic components, and structure building. It is also desirable to determine, for ground testing these materials, the laboratory parameters in order to simulate the dose they would receive in space environments. A sample print-out for water subject to 1.5 Mev proton is included as a reference.
Chang, C. Ken↗
FY21 Progress Report for Advanced Re fabrication/Re-instrumentation Capability Development
In support of performing follow on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and reinstrumentation capability. As a result, the United States Department of Energy has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory. The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of re-instrumentation and the progress to date. Halden spent nearly 30 years developing both refabrication and reinstrumentation. Collaboration with Halden is allowing INL to develop this capability much more rapidly. The results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Report on Inventory of Samples from Six Capsules from BOR-60
As part of the University of Michigan Grand Challenge Integrated Research Project, Oak Ridge National Laboratory (ORNL) received a drum containing six capsules from Pacific Northwest National Laboratory in August 2025. Each capsule originated at the University of Michigan and contained approximately 44 disks of various iron-based alloys (T91, HT9, T92, 800H, and others). These capsules were irradiated at several temperatures in the BOR-60 fast reactor for multiple cycles to accumulate high levels of damage. The drum was sent to the Irradiated Materials Examination and Testing (IMET) hot cell facility, a Class III nuclear facility, located in Building 3025E at ORNL. The IMET hot cell facility at ORNL is designed to receive irradiation capsules from the High Flux Isotope Reactor and perform capsule opening, basic optical examination, and mechanical testing on neutron-irradiated materials. Capsules were unloaded from the drum and placed in cell 6 for disassembly in February 2026. A low-speed saw opened each capsule, and a manipulator moved the disks to a dual microscope setup for disk identification, as shown in Figure 1.
36 MATERIALS SCIENCE↗
Characterization of Irradiated Nuclear Materials At LANSCE
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Post Irradiation Examination, Neutron Diffraction,↗
Capability Needs for Irradiated and Radioactive Materials Research (Ad hoc Committee Summary Report)
It is essential for the U.S. nuclear energy community to have access to world-leading equipment suitable for conducting research on both nuclear fuels and structural materials, including neutron irradiated–and therefore activated–materials. To address future research infrastructure requirements to support the DOE-NE mission, the Office of Reactor Fleet and Advanced Reactor Deployment established an ad hoc committee to gather information on potential capability gaps for radioactive materials and radiation effects research. This document summarizes the discussions of the committee addressing the high-level challenges in irradiated and radioactive materials research and the capabilities needed to address these challenges. After considering the various needs, the committee agreed on the four top-level targets and priority capability gaps.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Nanostructured Solar Irradiation Control Materials for Solar Energy Conversion
Tailoring the solar absorptivity (alpha(sub s)) and thermal emissivity (epsilon(sub T)) of materials constitutes an innovative approach to solar energy control and energy conversion. Numerous ceramic and metallic materials are currently available for solar absorbance/thermal emittance control. However, conventional metal oxides and dielectric/metal/dielectric multi-coatings have limited utility due to residual shear stresses resulting from the different coefficient of thermal expansion of the layered materials. This research presents an alternate approach based on nanoparticle-filled polymers to afford mechanically durable solar-absorptive and thermally-emissive polymer nanocomposites. The alpha(sub s) and epsilon(sub T) were measured with various nano inclusions, such as carbon nanophase particles (CNPs), at different concentrations. Research has shown that adding only 5 wt% CNPs increased the alpha(sub s) and epsilon(sub T) by a factor of about 47 and 2, respectively, compared to the pristine polymer. The effect of solar irradiation control of the nanocomposite on solar energy conversion was studied. The solar irradiation control coatings increased the power generation of solar thermoelectric cells by more than 380% compared to that of a control power cell without solar irradiation control coatings.
Kang, Jinho↗