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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding

Enhancing the accident tolerance of PWRs came under serious consideration following the 2011 Fukushima Daiichi nuclear power plant events. While the current nuclear fuels are based on mature technology and have an excellent operational record, further enhancing their strength through accident tolerant fuel (ATF) research and development continues to be of interest. Particularly, the chromium coated clad ATF technology represents an evolutionary change to improve the high temperature performance of existing Zircaloy claddings. Additionally, despite several years of investigation of coated cladding, the integral loss of coolant accident (LOCA) performance of chromium coated cladding has had little attention. In this work, cold-spray chromium coated zircaloy-4 claddings are evaluated using EDF's semi-integral LOCA test facility. Results show that the chromium coating induces a reduced ballooning and burst and an increased burst temperature in the investigated conditions. However, these tests exhibit limited benefits concerning the post-quench behavior of the coated specimens which occurs in the last stage of a prototypical LOCA scenario.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Zirconium nanoparticle coating development for FCCI diffusion barrier in nuclear cladding

Fuel clad chemical interaction (FCCI) is a pervasive issue for metallic nuclear fuels. FCCI can damage cladding, reduce thermal efficiency, and compromise the structural integrity of the fuel rod. In this study, we evaluated the FCCI mitigation capabilities of electrophoretically-deposited Zr nanoparticle coatings on ferritic-martensitic stainless-steel cladding (HT9). Zr was deposited on the internal surface of 6 mm outer diameter cladding using electrophoretic deposition (EPD). Thickness varied among samples with similar EPD conditions. A uniformly coated sample was selected and sent to the Argonne Tandem Linac Accelerator System (ATLAS) to simulate the coating's behavior under irradiation conditions. Furthermore, the coating prevented FCCI in diffusion couple studies. Results of this study confirm that Zr could potentially make an effective FCCI barrier in metallic fuel rods but the deposition process needs improvement.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Method of manufacturing a reinforced nuclear fuel cladding using an intermediate thermal deposition layer

A method is described that includes the steps of making a thin walled Zr alloy tube, loading nuclear fuel pellets into the tube, compressing the tube onto the fuel pellets to substantially reduce free space around the fuel pellets, positioning end plugs at each of two ends of the tube, filling the tube with a heat transferring gas, and coating the compressed tube with a corrosion resistant material using a thermal deposition process, such as cold spray, before inserting the tube into a pre-formed SiC composite cover having at least one closed end.

Lahoda, Edward J.↗
DOE OSTI2023

Effects of Cr/Zircaloy-4 coating qualities for enhanced accident tolerant fuel cladding

Cr-coated zirconium alloys represent a modern approach to enhance cladding safety during accident scenarios. Two high-power impulse magnetron sputtered Cr-coated Zry-4 systems were subjected to simulated loss-of-coolant accident conditions to investigate cladding performance. The first Cr-coating (4.8 µm thick) was deposited onto Zry-4 cladding and exhibited through-thickness cracking while the second Cr-coating (6.8 µm thick) was deposited with improved deposition parameters onto polished Zry-4 and exhibited no cracking. During burst testing, the coating with a higher density of defects failed to consistently reduce oxidation and exhibited similar burst behavior as Zry-4. In contrast, the second Cr-coating reduced ZrO 2 formation through formation of Cr 2 O 3 and displayed enhanced burst temperatures by ~80 °C compared to Zry-4. Utilizing an empirical relation for burst behavior of zirconium alloys, the 6.8 µm Cr/Zry-4 system displayed enhanced burst temperatures equivalent to an effective 0.464 mm increase in Zry-4 wall thickness, highlighting the value of continuous Cr coatings for accident scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Monitoring the Possible Effects of Corroding Fuel Cladding in Nuclear Ponds: a Modified ROV for Storage Ponds - 20144

Nuclear ponds are used worldwide to store radioactive nuclear material, often for extended periods of time. The pond water conditions need to be closely monitored so that mitigation strategies can be put in place should any leakage of radioactive material occur. This paper describes the selection, development and on-site deployment of a bespoke system whose design was based on of-the-shelf components. The system is able to identify, locate and differentiate the leakage of radioactive material stored in nuclear ponds from the background activity of the water in the pond. The selection of an appropriate detector and monitoring geometry was supported by MCNP modelling which allowed a high degree of confidence in the proposed design. The development of the system allowed a real-time indication of the canister integrity, without incurring the increased operator doses from manual sampling operations that were previously used on site. The benefits of the developed system include the fact that the system is lightweight, easily deployable, retrievable, decontaminable and results in negligible waste products. Moreover, it is able to analyse the samples in-situ, in this way reducing the time needed to manually collect and send the samples to a laboratory for lengthy analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions

A novel experiment to simulate cyclic dryout in boiling water reactors has been developed to better understand the performance of nuclear grade FeCrAl cladding in a BWR during dryout conditions caused by an Anticipated Operational Occurrence or Anticipated Transient Without SCRAM - both of which are Design Basis Accidents. Internally pressurized C26 M FeCrAl alloy cladding and Zircaloy-2 cladding were subjected to rapid 300°-650 °C thermal cycling in a steam environment; actual maximum temperatures were found to vary between materials but were always above 650 °C. In the range of 32–55 MPa hoop stress, Zircaloy-2 cladding burst within 1–16 cycles (about 100 s of dryout duration above 600 °C), while at 76 MPa hoop stress, C26 M cladding remained virtually undeformed after completing 54 cycles (over 1000 s of dryout duration above 600 °C). Higher temperature 300°-700 °C and 300°C–800 °C cycling experiments had to be performed to induce C26 M burst – failure occurred after 20 cycles in the former and during the first cycle in the latter. Zircaloy-2 and C26 M failure criteria were used to generate hoop stress specific dryout lifetimes. Overall, the simulated cyclic dryout experiments show that nuclear grade C26 M cladding has significantly enhanced survivability under dryout conditions relative to Zircaloy-2.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Hydrothermal Corrosion of First-Generation Dual-Purpose Coatings on Silicon Carbide for Accident-Tolerant Fuel Cladding

In this study, four different commercial coatings, TiN, Cr, CrN, and Cr/CrN (multilayer) were deposited by physical vapor deposition (PVD) on high resistivity chemical vapor deposited (CVD) SiC for the dual purpose of corrosion mitigation and fission gas hermeticity. The coatings were exposed for up to 2600 hours in flowing high temperature (288-350°C), high purity, liquid water with dissolved hydrogen (0.15-3ppm) or dissolved oxygen (1-2ppm). The Cr/CrN coatings did not adhere well and spalled quickly during exposure. In oxygenated conditions, none of the coatings were protective . CrN coatings spalled preferentially at defect sites, but reacted slowly enough in the absence of defects to potentially be chemically protective. TiN coatings fully oxidized within a few hundred hours and spalled. In hydrogen, all the coatings were chemically stable. Coating failures were identified and attributed to localized spallation, potentially related to coating defects. Coating improvements are recommended to reduce spallation and improve coating protection for future generations of dual-purpose coatings.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Correlation between Interface Microstructure and Bond Strength of Al6061/Al6061 HIP-Bonded Plates for Fuel Cladding Application

The effects of hot isostatic pressing (HIP) processing parameters on microstructure and bond strength of Al6061/Al6061 plate interfaces were systemically investigated at the micro- and nanoscale. The HIP bonding of the two aluminum plates was performed at four temperatures: 350°C, 400°C, 450°C, and 560°C. The results reveal the formation of micron scaled Mg 2 Si precipitate and uniform nano scaled Mg 2 Al 2 O 5 oxide at the interface. The formation of Mg 2 Al 2 O 5 is closely associated with the fast bulk diffusion of Mg from the Al matrix to the amorphous Al 2 O 3 /Al interface and subsequent reaction between Mg and Al 2 O 3 . Finally, the highest and lowest bond strengths were observed in samples HIP bonded at 450°C and 350°C, respectively.

36 MATERIALS SCIENCE↗