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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mechanistic Multiscale Uncertainty Propagation in Support of Accelerated Fuel Qualification

Taking a nuclear fuel concept through the research, development, and qualification stages has historically taken on the order of 20 to 25 years because of extensive irradiation tests required for a variety of conditions. The concept of accelerated fuel qualification (AFQ) has been proposed to increase the innovation pace for nuclear fuels. The goal of AFQ is not to replace the traditional qualification approach but rather to reduce the total number of experiments required to ensure approval from the regulatory authority. Of the many AFQ approaches being explored, advanced modeling—and, in particular, mechanistic modeling—is in a uniquely cross-cutting position to reduce the number of required integral tests through the inclusion of separate-effects testing, while helping to extrapolate reactor performance during rare events. We make the case that propagation of uncertainty through various computational length scales helps contextualize mechanistic modeling. We will utilize UO 2 fission gas diffusion predictions from the atomistically informed cluster dynamics code Centipede to inform fuel performance rodlet simulations using the BISON finite element code as the metric for showing how multiscale mechanistic uncertainty quantification can help reduce uncertainty in fuel performance. In conclusion, by quantifying uncertainty and its reduction through multiscale modeling, the qualification process may be accelerated through the reduction of costly irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Refabriation of HERA-HBU-2, LOC-HBU-1, and LOC-HBU-2

The development of advanced fuels for Light Water Reactors (LWRs) is essential for improved fuel performance and safety, including performance margins, uprates, and higher burnup. Because available commercial LWR fuels often cannot be used in research reactors, the refabrication of preirradiated fuel is necessary to enable follow-on testing of these fuels for characterizing fuel behavior during steady and transient irradiations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2016

Economical Production of Pu-238: NIAC Phase I Final Report

All space exploration missions traveling beyond Jupiter must use radioisotopic power sources for electrical power. The best isotope to power these sources is plutonium-238 (Pu-238). The US supply of Pu-238 is almost exhausted and will be gone within the next decade. The Department of Energy has initiated a production program with a $10M allocation from NASA but the cost is estimated at over $100M to get to production levels. The Center for Space Nuclear Research (CSNR) has conceived of a potentially better process to produce Pu-238 earlier and for significantly less cost. Potentially, the front end capital costs could be provided by private industry such that the government only had to pay for the product produced. In the Phase I NIAC (NASA Innovative Advanced Concepts) grant, the CSNR has evaluated the feasibility of using a low power, commercially available nuclear reactor to produce 1.5 kg of Pu-238 per year. The impact on the neutronics of the reactor have been assessed, the amount of Neptunium target material estimated, and the production rates calculated. In addition, the size of the post-irradiation processing facility has been established. Finally, as the study progressed, a new method for fabricating the Pu-238 product into the form used for power sources has been identified to reduce the cost of the final product. In short, the concept appears to be viable, can produce the amount of Pu-238 needed to support the NASA missions, can be available within a few years, and will cost significantly less than the current DOE program.

Plutonium↗
DOE OSTI2021

ECAR-3386 AS-RUN THERMAL ANALYSIS OF THE AGC-3 EXPERIMENT

The third Advanced Graphite Creep (AGC-3) experiment was designed to irradiate various types of graphite specimens at a temperature of 900°C. The specimens were irradiated in an instrumented leadout capsule experiment in the east flux trap of the ATR during cycles 152B, 154B, 155A, and 155B. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Micro-Tensile Properties of Irradiated AGR-2 TRISO Fuel Pyrolytic Carbon (PyC) and Silicon Carbide (SiC) Coatings

Tristructural isotropic (TRISO) coated nuclear fuel particles are emerging as a versatile option for new reactor designs, with the silicon carbide (SiC) layer crucial for retaining fission products. However, the mechanical properties of TRISO coating layers, particularly after irradiation, are not fully understood due to their small size and high radioactivity. Recent in situ micro-tensile testing of various TRISO layers aims to better understand the SiC layer's failure mechanisms, advancing TRISO fuel qualification. These micro-tensile results will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

GRAPHITE–MOLTEN SALT CONSIDERATIONS FOR COMPONENTS IN NUCLEAR APPLICATIONS

The new High Temperature Reactor (HTR) designs being considered for future Gen IV nuclear reactor deployment include designs utilizing molten salt as the primary coolant. These molten-salt cooled, graphite core designs pose new material compatibility challenges that are not considered within the gas-cooled HTR designs that have been previously built and operated. While the Molten Salt Reactor Experiment (MSRE) demonstrated that the molten salt can be considered chemically inert to graphite the novel physical and thermal interactions that the molten salt poses may be just as impactful as the chemical reactivity. Specifically, molten salt intrusion into the open pore structure of nuclear graphite grades can provide additional internal stresses within the microstructure exacerbating the stress buildup from irradiation induced dimensional change. Additionally, designs using a molten salt containing liquid fuel could provide “hot spots” within graphite structural components causing local thermal stresses. Abrasion and erosion concerns are magnified with molten salt due to the extremely high density of the salts (some have higher densities than the structural graphite components). Finally, the graphite-graphite and fuel pebble-graphite tribological behavior are distinctly difference within the molten salt from the inert gas environments and must be investigated. These topics and others are currently under investigation within the DOE Advanced Reactor Technologies (ART) graphite program and will be discussed in depth.

Gallego, Nidia↗
DOE OSTI2021

ECAR-2322 As-Run Thermal Analysis Of The AGC-2 Experiment

The second Advanced Graphite Creep (AGC-2) experiment was designed to irradiate various types of graphite specimens at a temperature of 600°C. The specimens were irradiated in an instrumented leadout capsule experiment in the south flux trap of the ATR during cycles 149A, 1498, 1508, 151A, and 1518. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT

The U.S. nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. Here, this work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles' silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗
DOE OSTI2023

Development of hydrothermal corrosion model and BWR metal coating for CVD SiC in light water reactors

SiC/SiC fiber composites with CVD SiC overcoats are potential candidates for light water reactor advanced accident tolerant cladding materials. Understanding its corrosion kinetics in Light Water Reactor (LWR) conditions is essential to evaluate the concept's viability. Existing models only account for the temperature and oxygen concentration effect on the hydrothermal corrosion behavior applicable to LWR operating conditions. However, the development of a general corrosion rate for CVD SiC that accounts for the impact of irradiated microstructure, flow rate, electrical resistivity, pH, and surface roughness is critical for the practical realization of SiC/SiC-based cladding concepts. After a rigorous experimental campaign, this work updates the existing hydrothermal corrosion model to predict hydrothermal corrosion in LWRs. Numerical radiation and coolant chemistry analysis for LWRs conducted based on the updated corrosion kinetic models suggests that CVD SiC is likely a viable environmental barrier coating for Pressurized Water Reactors while questionable for Boiling Water Reactors (BWR) when the effect of irradiation damage on SiC corrosion is considered. An effective mitigation strategy for the double-layer metal coating is proposed for BWR applications. The double-layer metal coating comprising a FeCrAl overcoat with an intermediate Cr bond coating was observed to provide a stable protective barrier against SiC dissolution in BWR conditions. Finally, the proposed metal coating was also fully adherent following quench and burst tests.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ~400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α'), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Assessing High Burnup U-19Pu-10Zr Fuel Performance against Historical and Modeled Behavior

Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Advancement of LANSCE accelerator facility as a 1-MW Fusion Prototypic Neutron Source

The Fusion Prototypic Neutron Source (FPNS) is considered to be a testbed for scientific understanding of material degradation in future nuclear fusion reactors (Zinkle and Moeslang, 2013; Summary Report on the FPNS Workshop, 2018; Pitcher et al., 2019). The primary mission of FPNS is to provide a damage rate in iron samples of 8-11 dpa/calendar year with He/dpa ratio of 10 appm in irradiation volume of 50 cm 3 or larger with irradiation temperature 300–1000 °C and flux gradient less than 20%/cm in the plane of the sample. The Los Alamos Neutron Science Center (LANSCE) is an attractive candidate for the FPNS project. The Accelerator Facility was designed and operated for an extended period as a 0.8-MW Meson Factory. The existing setup of the LANSCE accelerator complex can nearly fulfill requirements of the fusion neutron source station. The primary function of the upgraded accelerator systems is the safe and reliable delivery of a 1.25-mA continuous proton beam current at 800-MeV beam energy from the switchyard to the target assembly to create 1 MW power of proton beam interacting with a solid tungsten target. The present study describes existing accelerator setup and further development required to meet the needs of FPNS project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

A view on the current and future impact of research reactors

Full text of publication follows. The current fleet of nuclear research reactors worldwide is nearly 70 years old. These reactors have proven to be extremely valuable tools of nuclear science and engineering with a broad and interdisciplinary impact. To date, research reactors are utilized as tools for understanding the physics, operations, and safety of nuclear fission systems. In addition, they are used as intense sources of radiation in support of irradiation testing and nondestructive examination of materials. As this fleet of reactors ages, an urgent need exists to establish new facilities that can propel the benefit of these reactors into the 21. century. In fact, an opportunity exists to build research reactors based on technology concepts that are being considered for nuclear energy reactors. This may include high temperature gas cooled and/or molten salt based advanced and micro reactor concepts. Such future reactors should be designed to maintain the broad utility of current reactors in research and education. However, modern research reactors can be purposefully designed and instrumented to access neutronic and thermal hydraulic information that would support the development and validation of reactor multi-physics modeling and simulation techniques. In this case, the entire phenomenological paradigm of the reactor may be captured to understand the neutronic multiscale and its impact on operations and safety. Moreover, the generated data can be channeled to drive anticipatory examination of the state of the reactor. In general, a symbiotic relation may be envisioned between the modern research reactor and power reactor fleets, which could facilitate the safe and efficient implementation of clean nuclear energy. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Overcoming challenges to support US resumption of high specific activity cobalt-60

Domestic production of high specific activity 60 Co was halted after a target rupture in 2012 at the Advanced Test Reactor (ATR). The Isotope Program (IP) within the US Department of Energy (DOE) Office of Science tasked a multilaboratory team of researchers and managers from Oak Ridge and Idaho National Laboratories with the redesign the radioisotope capsule. The objective of this effort was to create a more robust and reliable design, compared to the pre-2012 target. The team successfully completed this task to produce the DOE-IP cobalt (Co) production capsule design. Furthermore, 66 capsules were successfully fabricated by Oak Ridge National Laboratory (ORNL) and delivered to Idaho National Laboratory (INL) for irradiation in the ATR between January 2014 and October 2016. This paper describes the efforts of the team to prepare and disposition the two initial DOE-IP Co production capsules that were processed in March 2020. These efforts include performing accurate production predictions, experimentally validating predictions with assay measurements, shipping with the Orano-furnished Battelle Energy Alliance Research Reactor shipping package, and disassembling capsules at the isotope vendor site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗