Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly May FY-20
Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly May FY-20
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly April FY-21
Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly April FY-21
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
3D printing of additive structures for nuclear fuels
A method for manufacturing a nuclear fuel compact is provided. The method includes forming an additive structure, consolidating a fuel matrix around the additive structure, and thermally processing the fuel matrix to form a fuel compact in which the additive structure is encapsulated therein. The additive structure optionally includes a vertical segment and a plurality of arm segments that extend generally radially from the vertical segment for conducting heat outwardly toward an exterior of the fuel compact. In addition to improving heat transfer, the additive structure may function as burnable absorbers, and may provide fission product trapping.
Terrani, Kurt A.↗
3D printing of additive structures for nuclear fuels
A method for manufacturing a nuclear fuel compact is provided. The method includes forming an additive structure, consolidating a fuel matrix around the additive structure, and thermally processing the fuel matrix to form a fuel compact in which the additive structure is encapsulated therein. The additive structure optionally includes a vertical segment and a plurality of arm segments that extend generally radially from the vertical segment for conducting heat outwardly toward an exterior of the fuel compact. In addition to improving heat transfer, the additive structure may function as burnable absorbers, and may provide fission product trapping.
Terrani, Kurt A.↗
Finite Element Modeling of Spent Nuclear Fuel in Dry Storage under Seismic Excitation
Spent nuclear fuel (SNF) is currently stored at independent spent fuel storage installations (ISFSIs) at nuclear power plant sites across the US. The US Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program is currently working to understand the response of SNF during a seismic event. This investigation is part of a long-term effort by DOE to quantify the external mechanical loads on SNF. The SFWST program has recently completed campaigns evaluating SNF behavior during normal conditions of transport and in 30 cm package drop scenarios; both campaigns included a combination of testing and modeling. Likewise, the investigation into seismic response of SNF will include full-size shake table tests of a vertical storage overpack containing instrumented surrogate fuel assemblies, as well as pre- and post-test finite element modeling. This work focuses on the development of finite element models simulating the planned test configuration. The seismic response is calculated at two scales: at the package level, and at the fuel level. The package level model evaluates the behavior of the storage overpack and its contents in response to the ground motions of a representative earthquake. The motion of the package is input to the fuel level model, which consists of a detailed finite element model of a 17x17 or 16x16 pressurized water reactor (PWR) fuel assembly or of a 10x10 boiling water reactor (BWR) fuel assembly. The stress and strain on individual components such as fuel rods and guide tubes are extracted from this model. This work details the SNF response to various input ground motions, considering representative earthquakes in the Central and Eastern US under various site conditions. Also discussed is the development of a “reduced scope” finite element model of the SNF, which consists of a single fuel rod while retaining the dynamic response of the entire fuel assembly. This reduced scope model addresses the significant computational expense of modeling the SNF under long-duration input ground motions and facilitates simulating additional parameter variations of the model beyond the as-tested configuration. The modeling results provide important insight that will help to inform future testing and modeling efforts by the SFWST program to close the knowledge gap related to external loads on SNF.
Kadooka, Kevin↗
Multi-Modal Characterization of Nuclear Fuels and Materials at the Idaho National Laboratory Materials & Fuels Complex
The Idaho National Laboratory (INL) leads cutting-edge research pertaining to the advancement of nuclear reactor technologies, including nuclear fuels and materials. The INL Irradiated Materials Characterization Laboratory (IMCL), Electron Microscopy Laboratory (EML), and future Sample Preparation Laboratory (SPL) are available to the nuclear research community to assess the behavior of nuclear fuels and materials, efficiently and comprehensively characterizing from the engineering to atomistic scale. The IMCL is a unique, 12,000-square-foot facility located at the INL Materials and Fuels Complex designed for analysis of irradiated materials. The facility operates advanced characterization instruments that are sensitive to vibration, temperature, and electromagnetic interference in modular radiological shielding and confinement systems, granting researchers the ability to assess the microstructural, chemical, mechanical and thermophysical properties of nuclear materials, especially irradiated fuels. The EML is dedicated to advanced characterization of materials with optical and electron microscopy tools, including scanning electron microscopy/focused-ion beam (SEM/FIB) and transmission electron microscopy (TEM). Upon construction, the SPL will be a 3 story, 49,000 sq. ft facility, that is the most modern reactor structural materials testing and analysis facility in the world, designed to investigate reactor structural materials in support of life-extension programs and development of advanced reactor concepts, including mechanical testing and advanced characterization capabilities. This presentation will showcase some of the main capabilities available at both IMCL, EML, and SPL, specifically illustrating how these characterization techniques are incorporated into multi-modal characterization work scopes to elucidate the degradation of nuclear structural materials and irradiated fuels.
36 MATERIALS SCIENCE↗
Accelerating Nuclear Fuels and Materials Qualification-Learning from MeV Summer School 2022
This presentation focused on the general overview of the lecture content that related to nuclear fuels and material qualification. The content of this presentation is taken from the lecture notes of MeV Summer School 2022. The sources are cited as per lecture notes. The focus of this presentation is to review the learning from the Modeling, Experiment, and Validation (MeV) Summer School (July 18-29, 2022) for early career researchers and scientists. This year's school was hosted by Oak Ridge National Laboratory (ORNL) and was focused on Accelerating nuclear fuels, and materials qualification by combining high through-put materials irradiation and testing, advanced PIE, and Multiphysics modeling.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modeling and Analysis for Spent Nuclear Fuel Seismic Testing
Note: This is the final public release version of PNNL-31671 Draft, which was previously released to the sponsor for review. The intent is the public release version will be PNNL-31671. There are no significant changes from the earlier draft version. This unlimited distribution report is the deliverable for M3SF-21PN010202014: Modeling and Analysis for Spent Nuclear Fuel Seismic Testing. This report summarizes the modeling, analysis, and test plan support completed by Pacific Northwest National Laboratory for the spent nuclear fuel dry storage system seismic test plan through May of 2021. Test plan preparation is planned to continue until the seismic test is completed in July of 2022. This report covers preliminary structural dynamic model development and computer-aided design of test hardware. Based on preliminary modeling, the strongest earthquakes under consideration in this test program provide mechanical loading on the spent nuclear fuel that is comparable to the 30 cm cask drop scenario, although the loads do not appear to be strong enough to cause significant permanent deformation of the fuel assembly spacer grids. The potential for grid deformation during the test will be assessed when the proposed shake table motion becomes available. The weakest earthquakes considered in this test are expected to be comparable to the mechanical loads witnessed in the multimodal transportation test of 2017. The horizontal canister system is predicted to provide nearly-uniform loading condition on the fuel assemblies it contains, while the vertical cask system is predicted to cause non-uniform dynamic loads on the fuel assemblies. In the horizontal case, gravity keeps the fuel assemblies in contact with one basket wall surface unless the loads are strong enough to cause a separation. In the vertical case, the fuel assemblies are relatively long and slender, and seismic motion in the anticipated test range is predicted to cause the assemblies to lean, tilt, and impact the basket walls throughout the seismic event. These gap closures are a nonlinear force transmission condition, so the vertical system is expected to have more variation and variability than the horizontal system. While the horizontal system is expected to have a relatively more linear response than the vertical system, there is still the potential for nonlinear behavior in the horizontal system because the fuel assemblies are free to slide, bounce, and impact the basket walls if the seismic loads are strong enough. One major conclusion of this study is that the use of mixed fuel assemblies in the canister will be acceptable. There are differences in overall system response when the mass, center of gravity, or gaps are changed, but the changes in response are within the bounds of a system that contains completely homogeneous fuel assemblies. One important observation is that the loading for each individual fuel assembly within the vertical canister is expected to be different from that of the others because of nonlinearities in the system. The horizontal canister system is expected to have a more uniform and more predictable response than the vertical canister, making variations in fuel assembly characteristics easier to account for. One important recommendation that comes from this modeling work is to repeat some of the strongest shake tests at different angles, particularly for the vertical cask system. Modeling predicts that strong seismic motion will cause the fuel assemblies to close initial gaps and impact the fuel basket walls when the canister is in the vertical cask configuration. Each shake test will impose a pre-defined three-dimensional time history on the cask system.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Spent Nuclear Fuel Mechanical Loads in the General Package Drop Scenario
The US Department of Energy Spent Fuel and Waste Science and Technology program is performing research to determine the mechanical loading conditions applied to spent nuclear fuel (SNF) during normal conditions of transportation to inform mechanical tests of SNF and close an important knowledge gap related to the practical disposition of SNF in the US. A recent multi-national collaborative test campaign measured SNF assembly impact response to the 30 cm horizontal package drop scenario, which is a common regulatory basis test of SNF package design. Researchers at Pacific Northwest National Laboratory (PNNL) are using the test data to validate explicit finite element models to calculate the mechanical loads and structural response of spent nuclear fuel assemblies in the as-tested 30 cm horizontal package drop scenario. Once the as-tested package drop model is validated, the next step is to apply the model to the general 30 cm drop scenario, which includes all impact angles, all fuel assembly types, and all burnup conditions. PNNL is developing a damage model that will incorporate the results of finite element parametric studies to establish trends in SNF mechanical loading to various input parameters, like impact orientation and burnup. The damage model will have the capability to estimate mechanical loads for any single set of input parameters, but it will first be used to describe the upper bounds of potential SNF loading in the 30 cm drop scenario. This paper describes PNNL’s progress toward developing the general solution to the problem of spent nuclear fuel mechanical loads in the 30 cm package drop scenario and it describes the next steps in closing the knowledge gap.
Klymyshyn, Nicholas A.↗
Development of a Nuclear Fuel Dissolution Monitor Based on Raman Spectroscopy
The processing of spent nuclear fuel and other nuclear materials is a critical component of nuclear material management with implications for global security. The first step of fuel processing is dissolution, with several charges of fuel sequentially added to a batch of solvent. The incomplete dissolution of a charge precludes the addition of the next charge. As the dissolution takes place in a heated, highly corrosive and radiological vessel, direct monitoring of the process is not possible. We discuss the use of Raman spectroscopy to indirectly monitor dissolution through an analysis of the gaseous emissions from the dissolver. Challenges associated with the implementation of Raman spectroscopy include the composition and physical characteristics of the offgas stream and the impact of operating conditions. Nonetheless, we observed that NO2 concentrations serve as a reliable indicator of process activity and correlate to the amount of fuel material that remains undissolved. These results demonstrate the promise of the method for monitoring nuclear material dissolution.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding
In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Phase equilibria of advanced technology uranium silicide-based nuclear fuel
The phases in uranium-silicide binary system were evaluated in regards to their stabilities, phase boundaries, crystal structures, and phase transitions. The results from this study were used in combination with a well assessed literature to optimize the U-Si phase diagram using the CALPHAD method. A thermodynamic database was developed, which could be used to guide nuclear fuel fabrication, could be incorporated into other nuclear fuel thermodynamic databases, or could be used to generate data required by fuel performance codes to model fuel behavior in normal or off-normal reactor operations. The U 3 Si 2 and U 3 Si 5 phases were modeled using the Compound Energy Formalism model with 3 sublattices to account for the variation in composition. The crystal structure used for the USi phase was the tetragonal with an I4/mmm space. Above 450°C, the U 3 Si 5 phase was modeled. The composition of the USi 2 phase was adjusted to USi 1.84 . The calculated invariant reactions and the enthalpy of formation for the stoichiometric phases were in agreement with experimental data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Strategy to safely enable X-ray computed tomography examination of highly radioactive tristructural isotropic nuclear fuel
Nondestructive post-irradiation examination of nuclear fuels and materials is useful in collecting data to inform commercial licensing of new nuclear concepts. This work details the planning, shielding design, and workflow of an X-ray computed tomography examination of an irradiated AGR-5/6/7 fuel compact with a dose rate of 1318 R/hr on contact from both β and ɤ-ray radiation with 120 R/hr of the dose rate coming exclusively from ɤ-rays. Post-irradiation examination of highly radioactive samples can help reduce the timeframe from conceptualization of novel nuclear concepts to their commercial implementation as not only is less time spent waiting for experiments to decay away to acceptable levels for examinations, but data from these experiments can more quickly inform model efforts and subsequent experiments. While this work represents the hottest radiological sample that has been openly transported and examined at Idaho National Laboratory’s Irradiated Materials Characterization Laboratory to date, it is anticipated that this work and its lessons learned will facilitate future examinations of similar or even more radioactive specimens.
42 - ENGINEERING↗
SFCOMPO Database of Spent Nuclear Fuel Assay Data – the Next Frontier
SFCOMPO is the world’s largest database for measured spent nuclear fuel assay data. An international effort coordinated by the Nuclear Energy Agency (NEA) resulted in a significant expansion of the database and its release online in 2017 as a downloadable application. The SFCOMPO Technical Review Group (TRG) was recently formed under the direction of NEA’s Nuclear Science Committee/Working Party on Nuclear Criticality Safety and was mandated to maintain and further coordinate the development of SFCOMPO. This TRG is currently focused on (1) critical evaluation of the experimental assay data by independent experts and (2) development of benchmarks and benchmark models that can be applied to validate burnup codes. This will improve the quality and documentation of the experimental datasets and enable their use by the international community to support code validation for design and safety analysis of spent nuclear fuel transportation, storage, and repository applications. It follows the precedent and draws on the experience gained from similar NEA efforts in the International Reactor Physics Experiment Evaluation Project and the International Criticality Safety Benchmark Experiment Project. Ongoing SFCOMPO evaluations have served as a test bed to develop templates for documenting evaluations, develop review guidance, improve approaches for a global uncertainty analysis, and devise a strategy focused on providing practical information of highest value to the user community. The current effort, status, and associated challenges are discussed.
Ilas, Germina↗
THERMAL STRESS ANALYSIS OF A SPENT NUCLEAR FUEL CANISTER
The potential for chloride induced stress corrosion cracking (CISCC) in spent nuclear fuel dry storage canisters is a current topic of research by the US Department of Energy Spent Fuel and Waste Science and Technology program. One of the important prerequisites for CISCC is a tensile residual stress state. This study utilizes computational models to provide an initial analysis of thermal stresses in a generic spent nuclear fuel canister. A STAR-CCM+ thermal fluid model provides a temperature profile analysis of the canister based on four different ambient temperature conditions. An ANSYS APDL finite element model incorporates the temperature profiles to analyze the thermal stresses in the canister. The calculated thermal stress magnitudes are in the range of 10 MPa to 80 MPa, which could be significant for crack propagation through the canister wall. The maximum thermal stress is not a concern for the structural failure of the canister, but it is high enough that it is a significant contribution to the total stress state of the canister wall, which also includes residual stress from fabrication, internal pressure from helium cover gas, and potential transient mechanical loads, such as earthquakes. This paper describes the initial finite element thermal stress analysis that was completed in 2020, reports the initial findings, and identifies areas where model refinement is still needed to complete this work in the future.
Jensen, Ben J.↗
Enhanced steam oxidation resistance of uranium nitride nuclear fuel pellets
Here, the steam oxidation resistance of UN and UN-(20 vol%)ZrN fuel pellets is evaluated to enhance understanding of steam corrosion mechanisms in advanced nuclear fuel materials. In situ neutron diffraction shows the modified UN fuel pellets form a (U 0.77 ,Zr 0.23 )N solid-solution and the sole crystalline oxidation product detected in bulk is (U 0.77 ,Zr 0.23 )O 2 . U 2 N 3 is not detected in significant quantities during the steam oxidation of UN or (U 0.77 ,Zr 0.23 )N and stable lattice parameters show that hydriding does not take place. Steam oxidation rates, obtained via sequential Rietveld refinement show how (U 0.77 ,Zr 0.23 )N has a higher activation energy (79 ± 1 kJmol -1 vs. 50 ± 5 kJmol -1 ), higher onset temperature (430 °C vs. 400 °C) and slower reaction rates for steam oxidation up to 616 °C, than pure UN. Throughout, both UN and (U 0.77 ,Zr 0.23 )N exhibit linear (non-protective) oxidation kinetics, signifying that degradation of the fuel pellets is caused by the evolution of gaseous products at the interface followed by oxide scale spallation. This quantitative and mechanistic understanding of material degradation enables better defined operating regimes and points towards (U,Zr)N solid solutions as a promising strategy for the design of advanced nuclear fuel materials with enhanced steam corrosion resistance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Safely probing the chemistry of Chernobyl nuclear fuel using micro-focus X-ray analysis
Detailed chemical analysis of the solidified molten fuel still residing in the stricken Chernobyl reactor unit 4 are inferred using multi-modal micro-focus X-ray analysis of a low-radioactivity proxy. A fascinating mixture of molten UO 2 , nuclear fuel cladding, concrete, stainless steel and other nuclear reactor components, these materials behaved like lava, solidifying to form a complex, highly radioactive glass-ceramic. Using element-specific chemical probes (micro-X-ray fluorescence and X-ray absorption spectroscopy), coupled with micro-diffraction analysis, the crystalline phase assemblage of simulants of these heterogeneous materials was established, which included “chernobylite” and a range of compositions in the (U 1-x Zr x )O 2 solid solution. Novel insight to nuclear accident fuel chemistry was obtained by establishing the oxidation state and local coordination of uranium not only in these crystalline phases, but uniquely in the amorphous fraction of the material, which varied depending on the history of the nuclear lava as it flowed through the reactor. This study demonstrates that micro-focus X-ray analysis of very small fractions of material can yield rich chemical information, which can be applied to nuclear-melt down materials to aid decommissioning and nuclear fuel management at nuclear accident sites.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Actinide Radiation Chemistry and Used Nuclear Fuel Reprocessing
Although the actinides boast many unique physical and chemical properties, their inherent susceptibility to radioactive decay—and subsequent consequences of radiation-induced chemistry—are what make them truly interesting and challenging elements to understand. From a closed nuclear fuel cycle perspective, the ability to predict and control the effects of actinide-driven radiolysis is critical for the design, development, and deployment of advanced used nuclear fuel reprocessing strategies and technologies. The absorption of ionizing radiation from actinide decay leads to the formation of a variety of transient and steady-state radicals, ions, and molecular radiolysis products that can lead to significant changes in a reprocessing solvent system’s physical and chemical properties, which ultimately limits that process’ efficiency and longevity. Presented here is an overview of recent advances in actinide radiation chemistry as it applied to used nuclear fuel reprocessing.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗