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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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115 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Select Transient Metallic Fuel Bison Benchmarks

To explore the current capabilities of Bison with respect to transient metallic fuel modeling, four transient experiments are compared with their respective Bison model. The four experiments are the OPT-1 experiment of the Experimental Breeder Reactor II~(EBR-II) and the M5, M6, and M7 experiments of the Transient Reactor Test facility (TREAT). Each benchmark is a 2D-RZ geometry of a single pin employing coupled multiphysics. Bison models of the benchmark agree well and illustrate areas needing development. The areas of suggested development deal with better material models, better coolant channel module for transients, proper treatment of material melting, improved feedback with failure models, and an ability to set initial conditions from post irradiation examination (PIE) data. Two Fast Flux Test Facility (FFTF) pins are also modeled with M7 transient conditions to illustrate Bison being applied to investigate potential experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

HT9 swelling in high burnup fast reactor fuel pin components

The irradiation-induced void swelling of HT9 stainless steel has been observed many times, but there are little data available to quantify the amount of swelling expected as a function of neutron exposure and temperature. This is further complicated by potential heat-to-heat chemical variations or heat treatment conditions during fabrication. Herein, this paper summarizes the results of most of the relevant previously completed studies, then focuses on the swelling behavior of HT9 mixed-oxide fuel cladding and wrapper wire irradiated to high neutron exposures. Note that the hexagonal duct used for the ACO-3 experiment (in the Fast Flux Test Facility, or FFTF) was also made from HT9 and has been studied previously by other researchers. The cladding enables a wider range of operating temperatures. Immersion density and transmission electron microscopy were used to identify a peak in the amount of swelling located axially, just below core centerline, for both the cladding and the wire. The peak roughly coincided with a peak seen in the cladding diameter profile. The wrapper wire showed enhanced swelling compared to the cladding, likely due to over-tempering during wire fabrication heat treatment. The grains showed decreased martensitic structure were larger in size. These data do not provide for the means to fully characterize the swelling of HT9, but do provide further insight into the factors controlling it. The onset of breakaway swelling was not indicated in this data.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Failure mode analysis of the FFTF heating and ventilation system 25

This document was prepared as a supplemental study to Chapter 9.4 of the FFTF Final Safety Analysis Report and is included as a Category A reference in FFTF System Design Description 25, Heating and Ventilation System. It contains the single failure mode analysis on all the heating and ventilation systems that are essential to the safety of the plant, and confirms the adequacy of the system design to accommodate single failures without jeopardizing overall plant safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Fuel-Cladding Eutectic Study of Legacy Fast Flux Test Facility (FFTF) MFF HT9/U-10Zr Metallic Fuel

This report presents the first systematic investigation of fuel-cladding eutectic interaction (FCEI) in irradiated HT9/U-10Zr metallic fuel from the Fast Flux Test Facility (FFTF) Materials Fuels Form (MFF) program, using differential scanning calorimetry (DSC) coupled with scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Two irradiated fuel cross-sections, MNT07H (9.5 at% burnup, x/L = 0.78) and MNT08H (7.0 at% burnup, x/L = 0.93), were subjected to three successive isothermal annealing rounds (R1–R3) at 820°C for 20 minutes each, yielding a cumulative transient duration of one hour. This study directly addresses a recognized gap in the existing FCEI database, which previously lacked irradiated HT9/U-10Zr data at burnup levels above 8 at%. Two principal findings emerge from the study. First, for both samples, FCEI remained spatially confined within the pre-existing fuel-cladding chemical interaction (FCCI) zone boundaries after R3, with no measurable eutectic penetration into unaffected cladding beyond the original FCCI layer. This self-limiting behavior is consistent with historical Fuel Behavior Test Apparatus (FBTA) results and is attributed to the near-eutectic phase composition of the FCCI zone, which rapidly absorb the available eutectic-forming constituents and then stall penetration once the FCCI zone is consumed. A comparison with unirradiated surrogate data further supports this mechanism: whereas a U–34 at.% Fe sample would be expected to show ~176 µm of iron penetration under comparable conditions, the irradiated samples exhibited only ~20 µm, a discrepancy attributed to irradiation-induced interfacial porosity and pre-existing FCCI composition gradients. Second, for MNT08H, FCEI was observed exclusively on the half of the cladding circumference where pre-existing steady-state FCCI was present, with no detectable FCEI on the opposite half. Three hypotheses are proposed to explain this asymmetry: the inhibiting role of a zirconium-rich rind at the fuel-cladding interface; the chemical sequestration of iron by redistributed zirconium within the fuel matrix; and the persistence of fuel-cladding gaps on the FCEI-free half that preclude direct contact. All three hypotheses require further experimental investigation. The results extend the empirical FCEI database into higher-burnup territory and demonstrate the viability of DSC-based testing as a substitute for the no-longer-available FBTA apparatus. Future work will include additional cross-section testing, compilation of the full FCEI dataset, model evaluation, and DSC testing of ternary fuel compositions to broaden the experimental basis for safety assessment of sodium-cooled fast reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2021

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. Here, the mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Irradiation creep and density changes observed in MA957 pressurized tubes irradiated to doses of 40-110 dpa at 400-750°C in FFTF

An irradiation creep and swelling study was performed on tubing constructed from the Y2O3-strengthened ODS ferritic steel MA957. As a result of the reduction operations during manufacture, the grains in the tubing were highly elongated in the direction of the tubing axis, with an aspect ratio of ~10:1. Pressurized creep tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 dpa to 110 dpa at temperatures ranging from 400 to 750°C. The diametral strains produced during irradiation exhibit very strong transient strains that are linearly dependent on stress and increase with irradiation temperature before reaching temperature-independent steady-state creep rates of 0.6-0.7 X 10-6 (MPa dpa)-1. Contributions to transient strains may not arise only from classical thermal creep or irradiation creep considerations, but also may result from an irradiation-stimulated growth process whereby the highly elongated grain structure reduces the aspect ratio to produce fatter grains and thereby increases in the tube diameter. One manifestation of this process is a change in tube diameter that is not accompanied by a density change characteristic of void swelling or precipitation-induced changes in lattice parameter. These results provide the first conclusive demonstration that resistance to irradiation creep can be extended to higher temperatures by dispersoid addition, and most importantly, this resistance is maintained to high radiation damage levels. However, the irradiation creep compliance is not reduced by dispersoid addition, casting some doubt on various proposed climb and glide mechanisms of irradiation creep.

100 dpa↗