INVESTIGATION OF EROSION/CORROSION AND THERMAL-HYDRAULIC BEHAVIOR OF PROTYPIC SODIUM FAST REACTOR COMPONTENTS

Abstract

The global search pivot away from fossil fuels has driven a resurgence of interest in the sodium fast reactor, a technology that was used in the original implementation of nuclear energy for power production but has since been overtaken by other reactor designs. Now, sodium fast reactors are one of the six technologies identified as the next generation of nuclear energy. While much research was conducted in the past century, further refinement of our knowledge and experience using liquid sodium as a primary coolant and working fluid within a nuclear reactor will serve to push the industry forward and further the promise of clean, abundant energy. This thesis describes three areas of research relevant to liquid sodium thermal-hydraulics and material interactions. The erosion/corrosion of stainless steel due to liquid sodium, particularly under high velocity flow conditions, is an area of the existing literature without remaining questions to be answered. Previously conducted experiments indicated that velocity-induced erosion/corrosion has an upper limit, past which increased flow velocity may not have increased effects. To investigate this, a 5000-hour test of erosion/corrosion of six orifices (as would be used for core flow distribution) in prototypical reactor flow conditions was conducted. The results of this test informed the continuation of testing, increasing the oxygen concentration of the liquid sodium flow. In parallel with corrosion testing of orifices, the hydraulic behavior of the same orifice design was directly compared in sodium and water flows. Often cited as a stand-in for sodium hydraulic behavior, water offers a much more accessible means of testing flow characteristics in design work, since its density and viscosity are similar to liquid sodium and because of the ease of water testing compared to liquid sodium. The loop utilized for corrosion testing was used to perform pressure drop testing of orifices similar to the corrosion test specimens, which were then placed into a hydraulically identical water test section. This allowed the loss coefficient to be assessed between the two fluids, to confirm that water can effectively be used as a surrogate fluid for predicting sodium flow characteristics where the dimensionless flow characteristic numbers (i.e. Reynolds number) are matched . Finally, this thesis details thermal-hydraulic performance characterization of a printed circuit heat exchange. These devices are a form of compact heat exchanger offering high surface area per volume and are widely considered for nuclear power applications. There is limited existing research on liquid sodium flow within these devices, and the most comparable study noted uncertainty regarding the heat transfer effects of fluid header regions within the PCHE. To investigate this, a custom designed printed circuit heat exchanger was instrumented with a distributed temperature sensor at multiple points within the device, measuring wall temperature. Operated within a liquid sodium loop and using compressed nitrogen as the other working fluid, this test shed insight into the performance of liquid sodium to nitrogen heat transfer through use of a compact printed circuit heat exchanger. The air-Brayton cycle is considered by several designs for next generation nuclear power, including microreactors. As such, there could be direction application of a sodium-to-nitrogen heat transfer, particularly through a compact heat exchanger.Financial support for several areas of this work was provided by TerraPower LLC, under part of a Department of Energy Advanced Reactor Demonstration Program issued under work order TP-PO-001860, in addition to the University of Wisconsin Thermal Hydraulics Laboratory