Engineering topics

Widgeon Paisner, Scarlett

Publications and source records attributed to Widgeon Paisner, Scarlett.

4 records

DOE OSTI2024

Microstructural and Oxidation Effects of Nb Additions to U3Si2

U3Si2 is a long term, accident-tolerant nuclear fuel candidate for light-water reactors because of its superior thermal conductivity and increased uranium density when compared to traditional uranium dioxide (UO2). While reducing internal thermal stresses and increasing efficiency, U3Si2 exhibits energetic oxidation during certain off-normal and accident scenarios, which include coolant or steam exposure. To mitigate this, Nb is investigated as an alloy constituent to enhance corrosion resistance and increase mechanical strength. The work presented investigates the response of Nb-alloyed U3Si2 to steam atmospheres. A thermogravimetric analysis is conducted in flowing steam to T > 1000 °C to assess oxidation resistance. The phase characterization of as-melted, thermally annealed and post-oxidation compositions with up to 12 vol% Nb by powder X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Accelerated Irradiation and Qualification of Ceramic Nuclear Fuels

Accelerated neutron irradiation testing is an component of accelerated qualification of new nuclear fuels for light water reactor (LWRs), microreactors, and other special purpose reactors. The qualification and licensing of nuclear fuel is a lengthy process that can take 20-25 years to bring a new fuel into service. Accelerated fuel qualification combines both experimental and modeling work to expedite the total qualification time to 5-10 years timeframe. The experimental aspect of this is accelerated irradiation aims to reduce the total time needed for neutron irradiation to achieve targeted burnup, which can take years using conventional irradiation profiles. The data that results from this irradiation testing can then be entered into BISON models to develop robust and reliable performance simulations to ensure safe operation under normal and off normal conditions. This milestone focused on the fabrication of test articles for accelerated irradiation testing at the Advance Test Reactor (ATR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

DISCOVARY PROJECT: Material Background Report

UC/UO 2 composites have been proposed as a next generation fuel for light water reactors (LWRs). Accident tolerant fuels (ATF) have been a focus in the Advanced Fuels Campaign (AFC) to improve the safety and performance of LWRs and includes research and development efforts on the cladding and fuel. The 10 wt.% UC/UO 2 composite fuel was selected as a result of an extensive literature review and was selected due to the improvement of the fuel cycle cost. The inclusion of a UC phase in the composite material improves on the properties of standard UO 2 by increasing the uranium density of the fuel and increasing the thermal conductivity. Significant development has been carried out to refine the processing and sintering parameters and has led to a dense composite without ternary phases present. Characterization of the chemistry and microstructure has been carried out to send ahead of neutron irradiations in the Belgium Research Reactor (BR2) at SCK-CEN.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Accelerated Irradiation and Qualification of Ceramic Nuclear Fuels

Accelerated irradiation efforts are an important topic in accelerated qualification of new nuclear fuel concepts for light water reactor (LWRs), microreactors, and other special purpose reactors. Accelerated fuel qualification combines both experimental and modeling work to expedite the total qualification time from approximately 20-25 years to 5-10 years timeframe. The scope covered under the current work package focuses on the experimental contributions for the neutron irradiations. Accelerated irradiation aims to reduce the total time needed for neutron irradiation to achieve targeted burnup, which is timely using conventional irradiation profiles. The purpose of this report is to detail the samples preparation for such irradiations, which contain standard, medium, and large grain undoped UO 2 , as well as large grain, Cr 2 O 3 -doped UO 2 . The data that will be generated from the irradiation of these samples will yield important information regarding fission gas retention as a function of grain size and the effect that Cr dopants have on diffusivity of fission gases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗