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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.

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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

The effect of temperature and burnup on U-10Zr metallic fuel chemical interaction with HT-9: A SEM-EDS study

The fuel cladding chemical interaction (FCCI) between Uranium-Zirconium-based metallic fuel and cladding materials during in-pile service is one of the most constraining phenomena affecting the performance of this fuel system. In this study, we investigated the effect of temperature and burnup on the FCCI development in two U-10 wt.% Zr (U-10Zr) fuel samples with HT-9 cladding irradiated as part of the MFF-3 irradiation test in the Fast Flux Test Facility (FFTF). One sample achieved a burnup of 13.1 at.% and operated with an average inner cladding temperature of 530°C, while the other achieved a burnup of 8.5 at.% and was subjected to an average inner cladding temperature of 615°C. Automated scanning electron microscopy (SEM) back-scattered electron (BSE) imaging of entire fuel cross-sections and SEM energy dispersive x-ray spectroscopy (EDS) analysis on specific fuel-cladding interface regions successfully provided a comprehensive characterization of the depth and type of interaction happening under different irradiation conditions. Further, our analysis shows that FCCI development on both fuel and cladding side is strongly influenced by the inner cladding temperature and, to some extent, the formation and integrity of Zr-rich layers between the fuel and cladding, while the impact of burnup and power is negligible. Measured FCCI thicknesses were compared to BISON simulations using both an empirical model based upon legacy data from the Experimental Breeder Reactor II (EBR II) irradiations and a mechanistic model currently under development for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, showing satisfactory agreement. Nonetheless, this comparison supports the need for additional microstructural characterization in intermediate ranges of temperature, power, and burnups in prototypic-length pins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Completion of Irradiation of Sensors in ATRC Milestone

This set of PowerPoint slides demonstrates how milestone "M2CT-21IN0702023: Perform test of neutron flux sensors in the Advanced Test Reactor Critical Facility". It supports an entry in PICS claiming achievement of the milestone. These slides provide a good summary of the purpose and impact of the milestone for our DOE Headquarters colleagues.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

A simplified approach for creep damage and remaining life calculation for materials surveillance in advanced reactors

High temperature reactor materials will experience a combination of mechanical degradation caused by creep and fatigue and environmental degradation caused by neutron irradiation and, in some cases, exposure to corrosive coolants. A materials surveillance program is one option for ensuring the safe, reliable operation of key reactor components under these conditions. Such a program would monitor the degradation in key material properties over time, using this data to predict how changes in the material properties affect components performance. This could then be used to make plant operational decisions and support eventual plant license extensions. Past work de-scribes such a material surveillance program based on passively actuated mechanical test articles that impose creep-fatigue type loading on test materials driven only by changes in temperature, for example those experienced by components under standard operating cycles. This report focuses on two aspects of the proposed material surveillance program: how to design the test articles to mimic the mechanical response of the corresponding component and how to use ex-situ test data to deter-mine the amount of damaged experienced by the test article (and hence the corresponding compo-nent) in service. Specifically, this report derives and validated simplified methods for both tasks, replacing earlier, more complicated approaches based on simulating the test article response and finding the best specimen design or current damage via complex numerical optimization. The sim-plified methods for each task developed here can be implemented in spreadsheet software and are simple enough for practical use in future operating plants. This report derives the methods (one for sizing and two options for damage inference), verifies the simplified approaches versus more so-phisticated methods, and compares the results of applying the simplified methods to previous re-sults using the complex numerical optimization approach. The report also provides a worked sam-pled problem applying the simplified techniques to a realistic high temperature reactor component.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

ANS Winter 2024 Summary: MCCAFE: The Monte Carlo Constructor for ATR Fuel Elements

The Irradiation Experiment Neutronics Analysis Department at Idaho National Laboratory (INL) has implemented a new analysis workflow for experiments in the Advanced Test Reactor (ATR). One key piece of this workflow is the Monte Carlo Constructor for ATR Fuel Elements, or MCCAFE. For each ATR operating cycle, the Reactor and Nuclear Safety Engineering (RNSE) Department first solves the core in eigenvalue mode and depletes the driver fuel materials. In a separate calculation, neutronics analysts model and deplete the materials of one or more irradiation experiments, usually in a series of fixed-source Monte Carlo N-Particle (MCNP) models of the ATR for neutron transport calculations. It was desirable to use the results of the former calculations to inform the models of the latter. MCCAFE is a Python program developed using American Society of Mechanical Engineers Nuclear Quality Assurance-1 procedures at INL. Its purpose is to take the calculated results from the RNSE depletion solutions and the measured or projected operating parameters from the Nuclear Data Management and Analysis System (NDMAS) to generate fixed-source models of the ATR core at given points in time across one or more cycles.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Results of Neutron Dosimetry Measurements for the Advanced Test Reactor Cycle 172A-1

Results of neutron dosimetry measurements in ATR for cycle 172A-1 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

172A↗
DOE OSTI2025

Results of Neutron Dosimetry Measurements for the Advanced Test Reactor Cycle 173-C

Results of neutron dosimetry measurements in ATR for cycle 173-C are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

DOE AGR fuel development and qualification program overview to include TRISO fuel development and program elements, AGR timeline, Irradiation test results, performance evaluation results, major accomplishments and milestones over 12 months, and university-led TRISO related research projects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AFIP6-MkII and RERTR-12 Porosity Data Collection and Analysis for Modeling and Simulation

Gathering data for the improvement of nuclear fuel modeling and simulation efforts is the primary driver for this work. Mechanistic models allow for a better understanding of the material on a micro- and macrostructural level while saving time and money over traditional experiment efforts. Historically, summarized data and correlations are the inputs for empirical material models and model validation. When improving these models for nuclear fuels with experimental results, there is a lack of reliable data readily available. Experiments - RERTR-12 and AFIP6-MkII - were conducted to understand the irradiation behavior of metallic U-10Mo monolithic fuels for use in extreme reactor environments such as research reactors like the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR). Microstructural characteristics of fission gas pores (FGP) in each experiment are collected using an automated image analysis technique developed at the University of Florida and presented here. A series of statistical tests are performed to explore the reliability of the results, as well as understand where the data is lacking and what future data collection is necessary to provide sufficient information to assist modeling efforts. The focus is on the porosity, pore size, and eccentricity of FGPs formed during irradiation in three AFIP6-MkII samples and one RERTR-12 sample. From the analysis, it is clear there are substantial impacts of fission density on the pore structure, but there also exist also underlying connections between each sample and the behavior observed in the pores. Further analyses of the pre- and post-irradiation microstructure are needed to improve the understanding of these connections. An early method for microstructural data analysis is presented within and is currently being expanded to include other microstructure data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report - 173B (Rev. 1)

Results of neutron dosimetry measurements in ATR for cycle 173B-1&2 are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

173B↗
DOE OSTI2023

PARFUME/BISON Fission Product Release Predictions versus AGR-3/4 Heating Test Measurements

The fuel performance modeling codes PARFUME (PARticle FUel ModEl) and BISON were used to predict the release of fission products silver, cesium, and strontium from as-irradiated fuel compacts containing tristructural isotropic (TRISO) coated particles during heating tests post irradiation. The AGR-3/4 fuel compacts were irradiated as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) program. The heating tests were conducted at temperatures between 1200°C and 1700°C to simulate reactor accident conditions. The measured fission product release fractions from the heating tests were compared to modeling predictions calculated by PARFUME and BISON to evaluate how the codes compare to experimental results. Comparisons between the experimental measured fission product release fractions from silver, cesium and strontium indicate that both modeling codes overpredict the fission product release fractions demonstrating that the diffusivities used in the codes are overestimated. This results in a conservative estimate predicted by the codes when evaluating the fission product release relative to experimental data as it pertains to silver, cesium, and strontium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2023

An examination of the precipitation behavior of proton irradiated dual phase 308L weldment filler materials

Voids, G phase particles, and Ni-Si rich clusters in proton irradiated dual phase 308L groove filler of a SA508–304 L dissimilar metal weldment are analyzed using advanced characterization techniques. These weldments are often used in light water nuclear reactors and are subject to enhanced corrosion and associated stress corrosion cracking (SCC). Radiation damage is known to accelerate SCC. Ni-Si enriched clusters were observed in proton irradiated γ austenite, while G phase M 6 Ni 16 Si 7 (where M transition metal element) precipitates were observed in proton irradiated δ ferrite. Compositional analysis of the G phase precipitates and Ni-Si clusters from STEM-EDS and APT are compared. Unlike G phase particles in proton irradiated δ ferrite, Ni-Si clusters in proton irradiated γ austenite are not rich in Mn. Both STEM-EDS and APT line-scan profiles of the Ni-Si clusters show that the Fe and Cr concentration gradient between matrix γ austenite and the Ni-Si clusters is not as sharp as those between matrix δ ferrite and G phase. Further, HR-STEM imaging indicates that the lattice parameter of the Ni-Si clusters is commensurate with γ austenite and the clusters do not represent the precipitation of a second phase. Finally, our analysis demonstrates the density and volume fraction of G phase particles and the density of voids scales with proton irradiation damage and the energy to recoils.

308L groove filler↗
DOE OSTI2023

Simulation of a TRISO MiniFuel irradiation experiment with data-informed uncertainty quantification

An irradiation experiment using tristructural isotropic (TRISO) fuel particles and the miniature fuel (MiniFuel) irradiation vehicle was performed in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) to support development of the Kairos Power fluoride salt–cooled, high-temperature reactor (KP-FHR). Here, this paper describes modeling predictions of temperatures and fuel burnup for the as-built experiment. An uncertainty quantification (UQ) analysis was performed to determine the effect of TRISO particle volume and position on the temperature predictions at various fuel heat generation rates (HGRs). This UQ study utilized fuel kernel position and volume measurements previously collected using X-ray computed tomography (XCT) techniques and Monte Carlo sampling methods to generate fuel compact cases that were then analyzed using a finite element thermal model. The UQ analysis indicated that uncertainty in calculated temperatures caused by varying TRISO particle arrangement is relatively small, even at high fuel HGR. Final predictions of particle temperatures throughout the irradiation are shown to be relevant to KP-FHR normal and off-normal operating conditions and to previous TRISO irradiation experiments. The combination of XCT with UQ analyses will inform post-irradiation examination (PIE) of the irradiated fuel compacts, and these analyses can be used to develop fuel performance models for coated particle fuel forms. Both PIE of separate-effects irradiation data and enhanced fuel performance modeling support accelerated qualification of TRISO fuels for a broad range of advanced reactor applications. The novel approach demonstrated here of measuring TRISO particle configurations with XCT methods and generating representative fuel compacts for finite element modeling and UQ analysis could be leveraged by the broader particle fuel community in the development of other TRISO fuel experiments in which these variables may have a significant impact on key outcomes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluation of AGR-3/4 In-pile Silver Release Predictions Against Post-Irradiation Examination Measurements

Fuel performance modeling codes that accurately predict the transport of radionuclides in high-temperature gas-cooled reactors that utilize tristructural isotopic (TRISO) fuel particles are an important aspect of reactor safety analyses. One objective of the Advanced Gas Reactor (AGR)-3/4 experiment was to assess the transport of fission products through fuel particles and their subsequent release into the compact matrix and structural graphite materials. This was accomplished by irradiating uranium oxycarbide (UCO) driver fuel particles and designed-to-fail (DTF) particles to serve as known sources of fission products. The fission product of particular interest when it comes to such transport is silver (Ag-110 m), as it has a 250-day half-life and has relatively high mobility in the TRISO coating layers. Furthermore, to assess the current modeling capabilities and diffusion parameters employed in the fuel performance codes PARFUME and BISON, the fractional release of silver release predicted by the two codes were compared against post-irradiation examination measurements from the AGR-3/4 experiment.

AGR-3/4 Experiment↗
DOE OSTI2020

Investigation of the structure and chemical nature of Pd fission product agglomerations in irradiated TRISO particle SiC

Tristructural-isotropic (TRISO)-coated fuel particles are used in high-temperature gas-cooled nuclear reactors. Although the polycrystalline 3C-SiC layer acts as the main barrier to fission product release, post-irradiation examinations have shown that certain metallic fission products are found outside of the TRISO coated particle, with no observable micro-cracks or other mechanical failure feature. In this study, an atomic resolution transmission electron microscopy investigation of a SiC layer of a neutron irradiated (19.38% fissions per initial metal atom average burnup; 1072°C time-averaged temperature) TRISO-coated particle from the first advanced gas reactor experiment at Idaho National Laboratory containing mainly Pd and U was conducted. The fission product Pd was found to be present at dislocation cores associated with twins, stacking faults, and their intersections. The Pd present at the dislocation cores were bonded to Si, suggesting that the diffusion of Pd along the SiC dislocation cores involves the formation of Pd-Si bonds. Evidence of single atom grain boundary migration was also found. Fission product agglomerations at grain boundaries most likely consist of a random solid solution containing Si-FP, Si-C, and C-C type bonding arrangements. No difference in the Si- or C-bonding environment is found for the unoccupied grain boundary adjacent to a fission product agglomeration compared to the bulk SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

User Manual for the Thermal Hydraulic Code SE2RCT

SE2-ANL is a modified version of SUPERENERGY-2 which interfaces the power distribution in the ANL maintained physics codes for the user. SE2RCT is a special version of SE2-ANL developed for specific use on models of the Experimental Breeder Reactor-II (EBR-II). SE2RCT fulfills a specific role in the calculated results as part of the Fuels Irradiation and Physics Database (FIPD) and is built to connect with the Physics and Analysis DataBase (PADB) for metallic fuel research and development in advanced sodium-cooled fast reactors (SFRs) created for EBR-II. SE2RCT is primarily a thermal analysis code that interfaces with the RCT based pin power details instead of the built-in pin power dehomogenization strategy in SE2-ANL. Beyond the thermal hydraulic calculation, the SE2RCT code provides detailed axial distributions of the RCT calculated results. SE2RCT also calculates the the axial distribution of pin wise DPA for the cladding materials which is part of the FIPD calculated results. This report serves as the user manual for SE2RCT and thus eliminates the QA gap of missing documentation on the SE2RCT software. All of the features and limitations of SE2RCT are documented in this report. The input and output specifications of SE2RCT are discussed in detail, and some unique features of SE2RCT, e.g. accurate pin power and temperature results for problems with non-uniform composition within an assembly, are also demonstrated and tested.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗