Neil A. Armstrong Test Facility
Quick facts of the test facilities at ATF in a short one page document.
James O Klein↗
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Results for “ATF”
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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Integrated Traction Drive Thermal Management: Keystone Project 3
The presentation was developed for the U.S. Department of Energy (DOE) Vehicle Technologies Program 2020 Annual Merit Review and Peer Evaluation Meeting. It reports on progress and accomplishments of DOE funded Integrated Traction Drive Thermal Management project. Highlighted are NREL's research accomplishments during FY 2020. Challenges associated with thermal management of integrated traction drives for electric vehicles arise due to the limited packaging space, coolant distribution, electrical connections, and ambient temperatures. For this project at NREL, research focuses on enabling thermal management technologies to support a modular, compact, self-contained integrated electric drive unit with motor and power electronics. Integration approaches, based on packaging location and coolant options for power electronics integrated with candidate motor technologies to reduce cables and connectors, are reviewed and compared using finite element analysis (FEA) tools. Automatic transmission fluid (ATF) jet impingement cooling characterization is carried out experimentally (measuring heat transfer coefficients) and using computational fluid dynamics (CFD) modeling.
47 OTHER INSTRUMENTATION↗
FY21 progress report for Advanced Re-fabrication/Re-instrumentation Capability Development
This report will focus on progress and development in establishment of refabrication/instrumentation capability at Idaho National Laboratory (INL). Refabrication/instrumentation is considered an enabling capability that allows access to fuel materials at any point in their life cycle. It is also critical in supporting deployment and qualification of ATF materials as many candidate materials are already undergoing irradiation as lead test rods (LTR) in commercial nuclear power plans (NPP) and irradiation in ATR alone can not produce the quantities of materials necessary to support qualification.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NSUF BRR Cask Shipment Summary Report
On September 16, 2021 the Idaho National Laboratory shipped a Battelle Research Reactor (BRR) cask loaded with five Nuclear Science User Facilities (NSUF) experiments and one Accident Tolerant Fuels (ATF) experiment from the Advanced Test Reactor (ATR) to the Hot Fuel Examination Facility (HFEF) at the INL Materials and Fuels Complex (MFC) for post-irradiation examination.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Application of Dynamical System Scaling for Accelerated Fuel Qualification Efforts
There are considerable ongoing research and testing campaigns to qualify new fuel designs such as metallic fuels for advanced reactor designs and the Accident Tolerant Fuel (ATF) campaign for light water reactors (LWR). The typical research and development lifecycle needed to qualify a nuclear fuel design, under ideal conditions, can take up to 20-25 years which limits the ability for new fuels to make fast deployments into commercial, test, and research reactors. While there are several efforts to accelerate nuclear fuel qualification through advanced modeling coupled with state-of-the-art experiments, there is not yet a framework to methodically quantify and rank distortions occurring in experimental test specimens needed to validate nuclear fuel performance codes. This study uses the Dynamical System Scaling (DSS) methodology to quantify transient scaling distortions that occur over experiment and simulated transients. DSS analysis metrics provide a researcher with several tools and information embedded within test data to identify dominant phenomena, associated timescales, and quantify a transient?s overall dynamic distortion. This is demonstrated with separate analyses of the SETH-C and CHF-SERTTA thermal-hydraulic experiments in the TREAT facility at Idaho National Laboratory. The outcome of this work is a scaling and data analysis approach to account for the influence a scaled fuel test specimen?s geometric and temporal distortions have on its ability to be representative of the full-scale design.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Post-Irradiation Fracture Toughness Characterization of Generation II FeCrAl Alloys
FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding application due to their exceptional resistance to oxidation in elevated temperature steam environments. Currently, limited fracture toughness data are available for the FeCrAl alloys, including the FeCrAl alloys newly developed at Oak Ridge National Laboratory (ORNL) under the U.S. Department of Energy’s Advanced Fuels Campaign (AFC) program. In this study, two Generation II candidate FeCrAl alloys, i.e., C06M (81.8Fe-10Cr-6Al-0.03Y-2Mo-0.2Si) and C36M (78.8Fe-13Cr-6Al-0.03Y-2Mo-0.2Si), were irradiated in the High Flux Isotope Reactor (HFIR) at ORNL to assess the fracture characteristics of these alloys after neutron irradiation. A total of six rabbit capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. Post-irradiation fracture toughness testing was performed following the Master Curve method in the ASTM E1921 standard. The main findings of this study are:1) Both the C06M and C36M alloys exhibited a similar response to irradiation concerning irradiation hardening and embrittlement.2) The irradiation temperature played different roles in terms of irradiation hardening and embrittlement for both C06M and C36M: after irradiation between 166°C and 204°C, both materials exhibited significant irradiation hardening and embrittlement; after irradiation between 315°C and 343°C, both materials showed small irradiation hardening without irradiation embrittlement. After irradiation between 501°C and 507°C, however, the irradiation softening without irradiation embrittlement was observed in both materials.3) Comparing the microhardness and Master Curve reference temperature T0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M steels. This should be mainly due to a flat response of the Master Curve reference temperature T0q to the irradiations at 166–204°C and 315–343°C ranges4) C06M showed a lower T0q, meaning better toughness, than C36M at the unirradiated condition, and such trend was kept even after neutron irradiation except for the 166–204°C irradiation after which both materials had similar T0q.5) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.
Chen, Xiang↗
HERA Project Update: RIA Test Results from TREAT
Presentation on HERA Project Update: RIA Test Results from TREAT at the Annual EPRI/DOE/INL Joint Combined Workshop on ATF and HBU in Dallas, TX April 3-5, 2023.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Post-Transient Examination Results of RIA Commissioning Teats at the Transient Reactor Test Facility
Six reactivity-initiated-accident (RIA) commissioning tests have been performed at Idaho National Laboratory’s (INL) Transient Reactor Test Facility. Five of these tests were performed using fresh fuel rodlets, and the sixth test was performed using a previously irradiated rodlet from the ATF-2 irradiation experiment. These experiments demonstrate the ability to perform RIA testing including the ability to perform experiments using previously irradiated materials. Post-transient examinations revealed pellet cladding interactions and cladding ballooning depending on the initial boundary conditions of the test. Increased thickness of both zirconium oxide and alpha-zirconium were found in the cladding with tests performed at higher levels of total energy deposition. No oxide or alpha-zirconium was found in the test with the greatest amount of ballooning, which was also the test with the highest initial rod internal pressure. The highest energy deposition test (1110 J/g) resulted in failure of the rodlet. Fragmentation of the fuel pellets occurred in this test with the size of the fragments decreasing inversely to the pellet radius.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
MECHANISTIC MULTIPHYSICS MODELING OF CLADDING RUPTURE IN NUCLEAR FUEL RODS DURING LOSS-OF-COOLANT ACCIDENT CONDITIONS
The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to release. Current licensing tools used by industry and the United States Nuclear Regulatory Commission are limited in geometric fidelity and materials that can be analyzed. These simulation tools generally employ a quasi-two-dimensional (1.5D or Layered1D) or 2D-RZ axisymmetric geometric representations exclusively. While a valid approach under some instances, there are times when important physics have an asymmetric behavior in the fuel rod. Examples include fuel fragmentation, thermal-hydraulic boundary conditions, and cladding rupture, all of which are important for LOCA analysis. As industry pursues burnup extensions it must be demonstrated that fuel dispersal can be mitigated or eliminated. To do this, an understanding of the rupture opening after cladding failure is required. This work presents the development of a model for predicting the size and location of the rupture opening in failed fuel rods during LOCA conditions using advanced modeling and simulation tools. In order to supply the rupture model with appropriate boundary conditions, improvements to fuel fragmentation, axial relocation and oxidation modeling were required. First, the eXtended Finite Element Method (XFEM) is used to mechanistically predict the number of fuel fragments that form due to material strength randomization, criteria for strength randomization, mesh density, power ramping rates and irradiation effects. These predictions with associated uncertainty were compared to empirical correlations developed for UO2 verifying that they can be used with increased confidence in subsequent axial relocation analyses. Secondly, a new first-of-its-kind Layered2D computational framework was developed that provides the ability to apply azimuthally varying boundary conditions while providing discrete layers to track fuel movement during the LOCA. An existing fuel axial relocation model developed for Layered1D was extended to work within the Layered2D framework. A large sensitivity study was performed on the initial version of the model to identify modeling parameters of particular importance, with the emissivity used for radiation after blowdown being the primary source of uncertainty. Then, a simplistic approach to incorporate mechanical degradation of the cladding due to oxidation was also developed to investigate the impact of reduced cladding thickness on predictions of the time to failure of cladding tubes. It was found that the cladding will typically rupture prior to a reduction in thickness that is sufficient to impact the rupture behavior. A model was then developed for predicting cladding rupture that transfers the cladding surface temperatures, rod internal and external pressures, fast neutron fluence, and fast neutron flux from a more detailed Layered1D, Layered2D, or 2D-RZ analysis to a 3D cladding only analysis. Comparisons of the rupture model to a few experiments indicate reasonable predictions. The rupture model was then applied to two accident tolerant fuel concepts (FeCrAl and Cr-coated Zircaloy) where it predicted that both ATF concepts would have smaller rupture openings and delayed rupture times than the standard Zircaloy-4 cladding material under identical loading conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Novel, Middle and Long Wave Infrared Laser Sources For Accelerator Applications
The main accomplishment during Years 2020-2024 of the DOE Accelerator Stewardship Program is proof-of-principle demonstration of amplification of an ultrashort 10 µm seed pulse in a multiatmosphere CO 2 laser optically pumped by nanosecond pulses of either a ~4.3 µm Fe:ZnSe laser or a 2.8 µm electro-optically (EO) Q-switched Er: Laser. This has been accomplished at the UCLA Neptune Laboratory by using about 100 mJ, ~100 ns long pulses generated by the pump laser built for this purpose by the UAB team. Implementation of EO Q-switching regime in the solid-state Er:laser system made possible nanosecond synchronization between the pump pulse and the 0.5 ps seed pulse produced by a commercial Ti:sapphire laser based mid-IR OPA/DFG. The injection mode-locking regime realized in a compact Mult atmosphere 3 Hz CO 2 -He regenerative amplifier enabled generation of a train of ultrashort pulses separated by the cavity round-trip time of ~3.2 ns. The measured gain lifetime of ~1 µs at a total pressure of ~12-15 atm indicates that lifetime of the upper laser level is a critical limiting factor in designing high-pressure CO 2 amplifiers. A single picosecond 10 mm pulse with an energy 0.1-0.2 mJ was extracted from the amplifier’s cavity using an ultrafast Ge semiconductor switch whose reflectivity was photo-induced by a femtosecond 800 nm pulse. For direct amplification regime, nonlinear absorption in Ge at 10 mm at GW/cm 2 intensities limited the extracted energy. Therefore, the generated energy could be further increased by minimizing losses in the switch and optimizing the cavity mode size. Based on the obtained data in collaboration with BNL colleagues, we have designed a high-pressure CO 2 amplifier to be used as a preamplifier for the future experiments at the ATF BNL. The latter work is the subject of funding availability.
60 APPLIED LIFE SCIENCES↗
A Combinatorial Approach to Testing and Characterization of Irradiated Fuels and Materials
•Combinatorial approaches to testing and characterization of materials is a very powerful tool to help explore some fundamentals of materials science.? •Three topics will be explored to help illustrate this:? •A fitness-for-service assessment of a CANDU component (this is work from my past)? •The use of small-scale mechanical testing of Cr-coated ATF cladding in support of qualification (this is work developed upon joining INL)? •A multi-scale characterization approach of TRISO fuel to investigate fission gas migration and microstructural features (this is work I now get to participate in on joining MFC)? •?
36 - MATERIALS SCIENCE↗
Critical heat flux on zircaloy and accident tolerant fuel cladding under prototypical conditions of pressurized and boiling water reactors
First of a kind high spatial and temporal resolution temperature measurements of Pressurized Water Reactor and Boiling Water Reactor simulated fuel pins (high heat flux cosine profile heaters) in annular rod type geometry during nucleate boiling, departure from nucleate boiling and re-wetting has been performed. Distributed fiber optic measurements of the cladding axial temperature with spatial resolutions down to 2.5 mm axially at up to three equally spaced azimuthal locations per rod have demonstrated the similarities and differences of the different accident tolerant fuel cladding materials. These experiments show the performance of accident tolerant cladding with a precise measurement of the location and evolution of dryout followed by the re-wetting phenomena (see supplementary time history of a critical heat flux event). Experimental parameters varied coolant mass fluxes from 1695 to 2712 kg/m 2 s, pressures from 10 to 20 MPa, and inlet subcooling from 10 to 55 °C. It was found that the coatings and different material have little effect on the value of the critical heat flux. Any discrepancies between the local critical heat flux verified among the cladding material are attributed to different departure from nucleate boiling phenomena captured during the experiment with the high-resolution optical fiber temperature data. Substantial differences in the performance of the different cladding material to survive a critical heat flux event were however observed by post-test X-ray and optical inspection. External evident damage and internal cracks to the simulated fuel pin having the bare zircaloy cladding were observed in regions associated with the occurrence of the critical heat flux for longer periods. Such damages were associated with zirconium oxidation, and not noticed on the Cr coated zircaloy or FeCrAl claddings. These results indicate that the use of FeCrAl or a Cr coated zircaloy as cladding in Pressurized Water Reactor and Boiling Water Reactor may increase the survivability of the cladding material for short transients.
32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
CALPHAD modeling of uranium nitride (UN) fabrication routes enabled by first-principles calculations
The thermochemical details of fabricating uranium nitride (UN) by ammonolysis of uranium tetraflouride (UF 4 ) were determined using density functional theory (DFT) and CALculation of PHAse Diagrams (CALPHAD) computational methods. The thermochemical data of all binary, ternary, and quaternary U-H-N-F phases were computed using DFT, and the data for the phases that have not been measured experimentally, including UN 2 and NH 4 F(g), were combined with existing experimentally-determined data for CALPHAD modeling. The DFT data were benchmarked using experimental Gibbs energy of reaction and experimental thermochemical data for individual species. Phase diagrams relevant to the ammonolysis reaction are depicted, showing regions of stability for solid U-N, U-F and U-N-F phases. An unidentified phase produced in a previous experiment was identified as UN 0.95 F 1.2 (UNF) by comparing its X-ray diffraction spectrum to the experimental spectrum, and its formation during the fabrication of UN from UF 4 is supported by the simulated phase diagram. Here, it is calculated that UN 2 can be produced by the ammonolysis of UF 4 , but requires elevated temperatures, high NH 3 (g) partial pressure, and large amounts of flowing NH 3 (g) to avoid solid flu oride impurities in the uranium nitride. Likewise, U 2 N 3 can be produced instead at temperatures greater than 980 K. The use of silane (SiH 4 ) gas was investigated as a potential additive in the ammonolysis fabrication route to speed removal of fluorine. The addition of SiH 4 (g) offers little advantage to the removal of fluorine, and adds the complication of Si 3 N 4 formation. The use of DFT to fill in missing data to perform CALPHAD calculations demonstrated here allows for the determination of more comprehensive and trustworthy phase diagrams than the use of existing experimental data alone.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Cr-doped U 3 Si 2 composite fuels under steam corrosion
Dense Cr-doped U 3 Si 2 composite fuels were manufactured by spark plasma sintering, and the effects of Cr addition on mechanical properties and oxidation resistance were investigated. Furthermore, dynamic oxidation testing by thermogravimetric analysis revealed significantly improved oxidation resistance of U 3 Si 2 with minimal doping of 3 wt% Cr. The onset oxidation temperature increased to above 550 °C in air and ~520 °C in steam conditions for the 5 wt% and 10 wt% Cr-doped composites. Steam corrosion testing under 360 °C for 24 hours indicated well-maintained pellet integrity without pulverization for the 10 wt% Cr-doped U 3 Si 2 pellet which only showed minor surface oxidation. The first promising results open up the possibility of designing and manufacturing metal additive-doped U 3 Si 2 composite fuels with significantly-improved corrosion resistance as a potential candidate for accident tolerant fuels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Zinc water chemistry reduces dissolution of FeCrAl for nuclear fuel cladding
Not Available
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Understanding the high-temperature behavior and corrosion resistance of Cr-Nb coated cladding for BWRs
While Cr-coated Accident Tolerant Fuel has proven successful in Pressurized Water Reactors, its dissolution in high dissolved oxygen Boiling Water Reactor environments represents a major technological barrier. Here, this study presents a comprehensive evaluation of novel Cr-Nb alloys as a coating breakthrough solution for Accident Tolerant Fuel cladding in Boiling Water Reactors. We systematically investigated Physical Vapor Deposited Cr-Nb coatings with 13% and 24 at% Nb content on commercial Zr-based cladding through rigorous testing under both normal and accident conditions. Our results demonstrate Cr-Nb coatings remained completely intact during extended Boiling Water Reactor autoclave testing, while pure Cr coatings failed catastrophically due to oxide dissolution. Under Loss-of-Coolant Accident conditions at 1100°C, Cr-Nb coatings maintained protective capability for 45 min. At 800°C, the alloy coatings matched pure Cr's excellent corrosion resistance. Despite microstructural changes during prolonged high-temperature exposure, no coating delamination occurred, ensuring continued protection of the underlying Zr substrate. These findings establish Cr-Nb coatings as a viable Accident Tolerant Fuel solution for Boiling Water Reactors.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Evaluation of the ring tension test (RTT) for robust determination of material strengths
The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC
Chemical vapor deposited (CVD) SiC and nuclear-grade SiC/SiC composite materials were exposed to light water reactor (LWR) conditions at the MIT Nuclear Reactor (MITR) here. Three sets of samples were exposed within flowing PWR-chemistry-coolant under either 1) neutron and ionizing irradiation with resulting radiolysis products, 2) radiolysis products without neutron flux, 3) loop coolant. Post-irradiation examination (PIE) exhibited an increase in corrosion from radiolysis products (particularly H 2 O 2 ), and potentially a further increase from radiation damage. For CVD SiC, regardless of exposure location, the in-situ observed corrosion rate was less than 1 µm per year, potentially within acceptable limits for LWR application.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗