Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “ORNL”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 433 records · Page 24

Synthesis and Single Crystals of Refractory Oxides of Lanthanides and Thorium

At the completion of this program, we can report that we developed a considerable degree of technical improvements in our ability to perform hydrothermal reactions at high temperatures and pressures. We can now routinely perform reactions at 700-750°C and 200 MPa. Currently we are in the process of exploiting this new technology synthesizing a range of exotic new materials investigating relatively poorly understood materials. Our initial efforts focused on the chemistry of rare earth oxides with tetravalent and pentavalent oxides. We recently published a study of the lanthanides with Nb 5+ and Ta 5+ ions, where we grew oxides such as RENdO 4 and RETaO 4 as high quality single crystals. These compounds were targeted as potential hosts for luminescent and scintillation materials, particularly given that they are among the densest oxide hosts and hence have good potential as absorbers for high energy radiation like X-rays and gamma rays. We also isolated a range of unusual new rare earth tantalates with very complex structures. indicating that the chemistry is very sensitive to conditions. We performed some fairly comprehensive examinations of the solid-state chemistry of rare earth ions with various tetravalent metal ions especially Si 4+ , Ge 4+ , Sn 4+ and Ti 4+ . Given the potential role of rare earth silicates in immobilizing radioactive waste elements in long-term storage, and the similarity of our hydrothermal fluids with known geological conditions, this chemistry continues to be relevant. We prepared an extensive series of new lanthanide germanates (e.g. RE 13 Ge 6 O 31 (OH), BaRE 10 (GeO 4 ) 4 O 8 ). and found that there is there is almost no overlap between the chemistry of the rare earth silicates. Stannic oxide (SnO 2 ) is much more refractory and requires higher temperatures and of mineralizer concentrations. One significant result is the growth of RE 2 Sn 2 O 7 pyrochlore single crystals. These are of interest because the rare earth stannate pyrochlores are known to display a wide range of magnetic frustration such as spin ice behavior. We grew high quality single crystals of rare earth germanate and stannate pyrochlores and this led to a collaboration with Professor Kate Ross at Colorado State. Preliminary measurements, indicate that the crystals contain no detectable defects or site disorder. Initial neutron diffraction on single crystals was performed at Oak Ridge, and more detailed experiments involving the Ross group are underway at both NIST and ORNL. This particular chemistry has turned out to be the most potentially significant work on this project and the collaborative effort with Prof. Ross is the topic of a DoE renewal project on quantum materials. Our initial foray into the hydrothermal chemistry of rare earth titanates has also been very promising and a range of cubic and polar ferroic phases of the light rare earths RE 2 Ti 2 O 7 (RE = La - Pr) in the P2 1 phase. We also discovered an interesting new phase Ce 2 Ti 4 O 11 that can have implications in heavy metal immobilization and storage. along with a series of new rare earth titanates (La 5 Ti 4 O 15 (OH) Sm 3 TiO 5 (OH) 3 and Lu 5 Ti 2 O 11 (OH) with exceptionally complex structures. One interesting sidelight has been high temperature hydrothermal chemistry terbium, including the growth of large crystals of TbO(OH). This is not a new compound but it is the first time it has been grown as large single crystals. The Tb atom density is almost as high as that in Tb 2 O 3 and has a very high Verdet constant (ca. 70), making it a very attractive candidate as a Faraday rotator. Unfortunately it is not in a cubic structure but he material is hard, stable, pure and inexpensive, so should still be an attractive Faraday oscillator. We recently received a patent on this material. We also synthesized K 2 Tb(Ge 2 O 7 ) containing stable octahedral Tb 4+ ions, which appears to be the first example of a well-characterized Tb 4+ complex. Given that Tb 4+ has been proposed as a benign surrogate for more treacherous tetravalent ions such as Cf 4+ and Bk 4+ , we think that Tb 4+ silicates can be a particularly useful study for actinide immobilization and related work. We also began reaction studies with rare earths and both ReO 2 and RuO 2 . These resulted in large single crystals of species like RE 5 Ru 2 O 12 , RE 4 Re 2 O 11 , REReO 4 and RE 2 ReO 5 . Several of these samples have already been sent to ORNL for magnetic and neutron diffraction studies.

36 MATERIALS SCIENCE↗

Molten Salt Reactor Signatures and Modeling Study

Molten salt reactor (MSR) technologies, either liquid fueled and cooled or only liquid cooled, pose specific, unique challenges for safeguards of the special nuclear material during the operation, fueling, and maintenance of the reactor. MSRs are one type of Generation IV technologies being invested in and considered for U.S. domestic fabrication primarily for electricity and process heat production. These designs have generated growing commercial interest for several reasons, including high (≈40%) thermal efficiency, ease of fueling, improved use of uranium fuel, potential utilization of thorium fuel, and proposed inherent safety features. U.S. companies have several planned designs that differ in fuel, cooling, and neutron energy spectrum. Driven by commercial interest and the intent of licensing MSRs, the U.S. Nuclear Regulatory Commission (NRC) has developed a vision and strategy to accommodate non-light water reactors (LWR), which include MSRs (U.S. Nuclear Regulatory Commission 2019). In addition, the Department of Energy (DOE) through the Office of Nuclear Energy (NE) Office of Advanced Reactor Technologies (ART), …sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. Reactor types considering the use of salts, liquid metals, or gases for coolant fall under both ARC and aSMR. Therefore, Research Design & Development is being pursued by DOE-NE through national laboratories, universities, and international and industrial collaborations. Additionally, the U.S. is a member of the Gen IV International Forum (GIF). The GIF is a cooperative, multinational organization to guide and carry out research and development needed for the GEN IV reactor systems (Forum 2018). GIF evaluated numerous reactor concepts and down-selected to the six most feasible advanced reactor technologies: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), MSR, supercritical watercooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). In support of the growing interest domestically and internationally, the Materials Protection, Accounting, and Control Technologies (MPACT) campaign, under the DOE-NE Fuel Cycle Technologies (FCT) program, engages in R&D activities by developing advanced instrumentation and analysis for safeguards and security of modern, advanced nuclear fuel cycle (non-LWR) facilities. Because of the historic experience in the operation of the Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) (Robertson, MSRE Design and Operations Report Part I 1965), Oak Ridge National Laboratory (ORNL) is heavily engaged in the various R&D activities through the DOE complex related to MSRs including national technical leadership of the DOE-NE MSR campaign. This report discusses and presents the outcomes of the FY19 MPACT MSR Safeguards task. The challenges presented by MSRs for nuclear material accountancy and control (NMAC) and associated safeguards will be investigated. The objective of this research is to explore and compile the safeguards requirements and identify measurement signatures through an initial high-level MSR design and develop complementary advanced simulation and modeling capabilities. A high-level ORNL-developed MSR design called the Molten Salt Demonstration Reactor (MSDR) (Bettis, Alexander and Watts 1972) was used as the target reactor design for this research. The MSDR model incorporates technology from the MSRE and the Molten Salt Breeder Reactor (Robertson, Conceptual Design of a Single-Fluid Molten-Salt Breeder Reactor 1971). But the MSDR is a 750 MWth graphite moderated liquid fueled (low-enriched uranium) MSR compared to the MSRE’s of 7.5 MWth. The focus of this report is to discuss the evaluation of novel signatures, correlations, and indicators to understand the applicability of current safeguards instrumentation to MSRs using the modeling results from the MSDR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Report on FY 2020 testing of Alloy 709 in support of EPP analysis

Research activities in support of the development of a 500,000 hr, 760ºC ASME Code Case (CC) for Alloy 709 are ongoing at Oak Ridge National Laboratory (ORNL), Argonne National Laboratory, and Idaho National Laboratory. Integration of the simplified Elastic-Perfectly Plastic (EPP) methods for the design of Alloy 709 reactor components is a requirement for the Alloy 709 CC. This report documents the research plans and initiation of testing in support of integrating EPP analysis methods for Alloy 709 at ORNL in FY 2020. Tests are designed and initiated to evaluate the effect of cyclic loading on the creep properties of Alloy 709. The preliminary results show that cyclic loading has an insignificant effect on the minimum creep rate, although additional tests are recommended to confirm this conclusion. A preliminary testing plan for multi-axial creep rupture has been established and the specimens are machined. Execution of the testing plan will continue in FY 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Sister Rod Destructive Examinations (FY20)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sister Rod Destructive Examinations (FY20) Appendix A: Full Length Rod Heat Treatments (FHT)

As a part of the US DOE-NE High Burnup Spent Fuel Data Project, ORNL is performing destructive examinations (DEs) of high burnup (HBU) (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO®, and M5®. The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding are being tested. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data will support SNF dry storage licensing and relicensing activities.This report documents the status of the ORNL Phase 1 DE activities related to full length rod heat treatments (FHT) applied to selected sister rods in Phase 1 of the sister rod test program. The results of this work and detailed examinations are summarized in the Sister Rod DE Status Report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sister Rod Destructive Examinations (FY20) Appendix B: Segmentation, Defueling, Metallograhic Data and Total Cladding Hydrogen

As a part of the US DOE-NE High Burnup Spent Fuel Data Project, ORNL is performing destructive examinations (DEs) of high burnup (>45 gigawatt days per metric ton uranium) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called the “sister rods,” are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO®, and M5®. The DEs are being conducted to obtain a baseline of the HBU rod’s condition prior to dry storage and are focused on understanding overall SNF rod strength and durability. Both composite fuel and empty cladding are being tested. While the data generated can be used for multiple purposes, a primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities.This report documents the status of the ORNL Phase 1 DE activities related to: rough segmentation, defueling, optical microscopy, and cladding total hydrogen measurements for 7 Phase 1 sister rods and outlines the DE tasks performed and the data collected to date. The results of these detailed examinations and others are summarized in the Sister Rod DE Status Report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Predicting Deformation of Parts from HP Printers due to Sintering with Modeling and Simulation (CRADA NFE-18-07363)

This work proposes to leverage Oak Ridge National Laboratory’s (ORNL) sintering simulation software developed at the Manufacturing Demonstration Facility to 1) investigate the deformation of sintered parts and 2) to develop fast surrogate-based optimization methods for HP Inc. This includes the development of ORNL’s software on HP systems, integration and testing with HP part geometries and material properties, the development of surrogate models based on the results of the simulation, and modifications to the software based on feedback from HP. The expected impacts of this work are a drastic reduction in print costs, an increase in printer throughput, and the identification of novel, possibly exotic materials for 3D printing. This work will also decrease the time to market for HP’s print technology by decreasing the amount of resources and type dedicated to physical testing.

42 ENGINEERING↗

TCR Postulated Accident and MHA Dose Assessment

Oak Ridge National Laboratory (ORNL) is engaged in regulatory activities to gain US Department of Energy (DOE) approval to build and operate a small high-temperature gas reactor, the Transformational Challenge Reactor (TCR). ORNL uses advanced manufacturing techniques to produce select reactor components. The TCR is unique in its mission: the reactor’s at-power operating lifetime will be less than one effective full-power day to limit the radiological source term to very low values, thus facilitating approval to operate. The primary regulatory documents governing construction and operation of the facility are NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors” and DOE-STD-3009, “Preparation of Nonreactor Nuclear Facility Documented Safety Analysis.” Although this document is not a safety basis document, it provides a conservative indication of the magnitude of onsite and offsite radiological doses for specific scenarios, thus informing future design and regulatory activities. As required by NUREG-1537, scenarios included in this analysis are (1) a worstcase depressurized loss-of-forced-circulation accident, and (2) a maximum hypothetical accident. Details on the analysis of dose consequences with various system configurations are also provided.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High Temperature Steam Oxidation of Irradiated FeCrAl in the Severe Accident Test Station

FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary results from Low Pressure Steam Oxidation Testing of ALD ZrN and ZrO2 Coating Deposited over UCN Fuel Kernels

Steam oxidation testing was used to investigate as-developed 100 nm ALD ZrN and 1000 nm ZrO 2 coatings deposited over UC1-xNx fuel kernels. The results of these tests were compared against oxidation of uncoated kernels. This work was performed to support development of coatings which both provide high temperature Zr metal diffusion barrier and also provide resistance against oxidation, especially against high temperature steam and/or air. This work was performed in collaboration with Oak Ridge National Laboratory (ORNL), using ORNL-provided samples. The oxidation tests yielded valuable information, including relative differences in performance of different coating materials and coating thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Binder Jet Printing of H13 Tool Steel

ORNL and Carpenter worked on evaluating H13 powders with different powder size distributions for deposition via binder jet additive manufacturing (AM). Binder jet AM has the potential to be a costeffective technique for fabrication of near net shaped tools with complex geometries and internal features such as cooling channels that can improve the production efficiency of injection/compression molding tools. However, a critical challenge in binder jet AM is the densification of the said tool for it to be usable. Powder size distribution can affect the process settings required to deposit the material and the subsequent sintering behavior. In this collaborative effort ORNL and Carpenter demonstrated that the process variables depend on the powder size distribution while the carbon pickup during sintering depends on the binder burnout atmosphere and temperature. On the other hand, there was no noticeable difference in the microstructural evolution of H13 deposited using the different powder size distributions.

42 ENGINEERING↗

Assembly of MiniFuel Targets for Irradiation of TRISO Fuel Compacts in the High Flux Isotope Reactor

To support the development of Kairos Power’s fluoride-salt-cooled high-temperature reactor, irradiation testing of tristructural isotropic (TRISO) fuel compacts was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental data on TRISO fuel during high particle-power operation and validate fuel performance models. Fuel compacts containing enriched uranium oxycarbide (UCO), natural UCO (NUCO), or uranium dioxide (UO 2 ) TRISO particles were fabricated at ORNL and inserted into MiniFuel targets for HFIR irradiation. Five MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six fuel compacts and will be irradiated in HFIR for four cycles, with target temperatures of 500, 700, and 900°C. This report summarizes the experiment design, test matrix, and fabrication. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Supervisory control and health monitoring framework for large-scale additive manufacturing systems. Final CRADA report

ORNL worked with National Instruments (NI) to develop a large-scale, complex additive manufacturing (AM) systems framework for remote health monitoring and supervisory control. We found the framework, based on the Lincoln Electric Metal AM system located at ORNL’s MDF, capable of controlling the process, leading to improved part quality and digital twin creation.

42 ENGINEERING↗

US Muon Workshop 2021: A road map for a future Muon Facility

The workshop title “US Muon Workshop 2021: A road map for a future Muon Facility” was held virtually on February 1-2, 2021. The workshop aimed to bring together world experts in muon spectroscopy (µSR) and other techniques along with interested stakeholders to evaluate the scientific need to construct a new µSR facility in the United States (US). The more than 200 participants highlighted several key scientific areas for µSR research, including quantum materials, hydrogen chemistry, and battery materials, and how each area could benefit from a new, high flux pulsed muon source. Experts also discussed aspects of the µSR technique, such as low-energy µSR, novel software developments, alongside beam and detector technologies, that could enable revolutionary advances in µSR at a next-generation facility. The workshop concluded with discussion of a concept being developed for a new µSR facility at the Spallation Neutron Source (SNS) of Oak Ridge National Laboratory (ORNL). That novel design concept was first envisioned by many of the same µSR experts at a workshop held previously at ORNL in 2016. The participants expressed that the current design had the potential to be a world-leading µSR facility, and strongly encouraged the principal investigators to continue their work in order to refine the concept and determine instrument parameters that would enable new scientific opportunities.

36 MATERIALS SCIENCE↗

Fracture Toughness Characterization of Generation II FeCrAl Alloys after ~18 dpa Irradiation

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

36 MATERIALS SCIENCE↗

Draft Experimental Plan for Commercial SNF Degradation in Repository Environments with a Focus on Fracture Matrix Degradation

This report documents work performed supporting US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.04, “Inventory and Waste Form Characteristics and Performance.” In particular, this report fulfills the M3 milestone, M3SF-21OR010309072 "Draft Experimental Plan for Commercial SNF Degradation in Repository Environments” as described in work package SF-21OR01030907 "SNF Degradation Testing-ORNL".This experimental plan is a collaboration between ORNL, SNL, ANL, and PNNL. It is intended to support source term model development and spent fuel degradation testing in a range of generic repository conditions. Potential testing and experiments to support validation of the Fuel Matrix Dissolution Model (FMDM) and the overall development/validation of repository source term modeling are discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

TCR Data Management Plan

The Oak Ridge National Laboratory (ORNL) Transformation Challenge Reactor (TCR) program is developing additive manufacturing and artificial intelligence (AI) to deliver enabling technologies for advanced reactors. Through the application of these advanced technologies, the program targets delivering solutions to the high costs and lengthy deployment timelines that threaten the future of nuclear energy—the country’s largest source of carbon-free energy. This document describes the plan for integrating and managing TCR data from multiple sources throughout ORNL and external sources, including novel data sources associated with advanced manufacturing, characterization systems, and the TCR Digital Platform that incorporates various levels of AI and data analytics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Rapid Simulation of Irradiation Damage in PWR Internals (NSUF 2019 Milestone Report)

This document summarizes the progress made by Oak Ridge National Laboratory (ORNL) in collaboration with the University of Michigan on “Rapid Simulation of Irradiation Damage in PWR Internals,” a research project led by the University of Michigan and supported by US Department of Energy Office of Nuclear Energy Nuclear Science User Facilities Consolidated Innovative Nuclear Research program. In this reporting period, samples that were exposed to combined neutron and ion irradiation were prepared and shipped to the University of Michigan for further analysis. This report summarizes the specimen preparation performed at ORNL’s Low Activation Materials Development and Analysis facility. The benchmarking work of microstructural characterization on reactor-irradiated samples with different dose levels was completed.

36 MATERIALS SCIENCE↗