Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Surrogates”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Milestone 1.2.14: Surface Characterization of Irradiated Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Specimens

The United States Department of Energy is seeking to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen generation from corrosion layers that would be present on the fuel assembly surfaces. However, to date, no evidence of radiation-induced changes in the microstructure of the corroded aluminum surfaces have been reported. This report provides surface characterization, in terms of corrosion layer composition and morphology, for aluminum alloy 6061-T6 “plume” samples subjected to one of three potential ASNF drying techniques—“vacuum only”, “vacuum + 4 hours at 100 oC”, and “vacuum + 4 hours at 220 oC”—and gamma irradiated (= 53 MGy) in the presence of helium gas with ~0% added relative humidity. At these high absorbed gamma doses, for the first time ever, radiation-induced circular defects were found on the sample surface, regardless of the drying regime employed. Additionally, large cracks that penetrated through the corrosion layer to expose bare aluminum metal were observed in irradiated “plume” specimens subjected to drying conditions of “vacuum + 4 hours at 220 oC”. The nature and implications of these defects on the extended dry storage of ASNF is unclear, requiring further study.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Scaling of thermal stratification in outlet plena of SFRs with gallium as a surrogate fluid

Due to its high thermal diffusivity, liquid sodium can develop thermally stratified regions in the hot and cold pools of SFRs under certain off normal operating conditions. The temperature gradients in the fluid give rise to mixed convection flow in which flow velocities result both from pressure gradients and from volume forces induced by density gradients. High thermal diffusivity and low momentum diffusivity leads to reduced thermal mixing and enhanced stratification. Thermal stratification, if not looked carefully in the design, can lead to severe consequences in normal operation with forced coolant circulation and during off-normal natural circulation. The resulting hydraulic and thermal loads can cause the failure modes in structures and components critical for reactor safety. Thermal stratification in the SFR pools can lead to thermal fatigue, which can lead to structural and material failure, or pose serious concerns for the reactor safety. In addition to thermo-structural issues, the reactivity changes are significantly impacted due to density or temperature driven movement of liquid sodium. Temperature changes or gradients in the Upper Instrumentation Structures (UIS) of SFRs can lead to thermal contraction, a case that can lead to positive reactivity insertion after core shut-down. The motivation of this work is to design a scaled down experimental facility which can generate high fidelity data for thermal stratification physics to validate 3-dimensional (3D) Computational Fluid Dynamics (CFD) models. These validated CFD codes can be used to improve the 1-dimensional (1D) system level codes such as SAS4A/SASSYS-1 which are currently not capable of simulating thermal stratification accurately. These improvements in 1D system level SAS4A/SASSYS-1 codes can significantly improve the safety analyses and design of SFRs. The GaTE facility outlet plenum model is designed based on physics based similarities. Thermal stratification physics is governed primarily by the Richardson numbers (Ri). The expected low flow rates during off-normal conditions can lead to a high degree of thermal stratification in the outlet plenum with a corresponding Ri = 530. Therefore, the inlet flow velocities and inlet hydraulic diameter for the gallium coolant entering the test vessel are required to achieve same Ri. The design of test vessel in GaTE facility is finalized to a 1/20. scale outlet plenum height of the ABTR. This design of the test facility will allow the decoupling of the thermal stratification physics at low flow rates with the transient natural circulation. In other words, it will enable effective control to study thermal stratification at very low flow rates. Numerical studies are conducted with a 3D CFD code to examine the thermal behavior of the selected test design. Simulation results show that thermal gradients are present only in the vertical or axial direction which are similar to the behavior reported in previous studies on full scale SFR models. Future work will involve the experimental studies using the GaTE facility with the high fidelity instrumentation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗