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The Power and Transportation Future
Published in Michael Frank Hordeski, Hydrogen & Fuel Cells: Advances in Transportation and Power, 2020
The nuclear plants now operating in the U.S. are light water reactors, which use water as both a moderator and coolant. These are sometimes called Generation II reactors. In these Generation II Pressurized Water Reactors, the water circulates through the core where it is heated by the nuclear chain reaction. The hot water is turned into steam at a steam generator and the steam is used by a turbine generator to produce electric power.
Energy Resources
Published in Dexter Perkins, Kevin R. Henke, Adam C. Simon, Lance D. Yarbrough, Earth Materials, 2019
Dexter Perkins, Kevin R. Henke, Adam C. Simon, Lance D. Yarbrough
The first commercial light-water power plant, built in Pennsylvania in 1957, operated until 1982. Commercial graphite reactors, based on somewhat different technology, make up 5% of the world’s electrical generation. They first were built in the United Kingdom and Russia in the late 1950s. And, in 1960, Canadian engineers designed a Canada Deuterium Uranium (CANDU) reactor that accounts for about 20% of the world’s nuclear electrical generation. Unlike light-water reactors, graphite and CANDU reactors can use unenriched uranium as fuel. The main difference between the three kinds of reactors involves neutron moderators that slow neutrons so the neutrons are more likely to interact with U-235, ultimately leading to fission decay that generates heat. Light-water reactors—the kind of reactors used in the United States—use normal water as both a coolant and as a neutron moderator. In CANDU and graphite reactors, respectively, heavy water (water in which some or all of the hydrogen, H-1, has been replaced by deuterium, H-2) and graphite are the moderators.
Modular Nuclear Reactors
Published in Yatish T. Shah, Modular Systems for Energy and Fuel Recovery and Conversion, 2019
The reactor is designed to use a heterogeneous metal alloy core with 192 fuel assemblies in two fuel zones. In the version designed for used LWR fuel recycle, all these are fuel, giving peak burn-up of 122 GWd/t. For the LWR fuel recycle version, fuel stays in the reactor 4 years, with one-quarter removed annually, and 72 kg/year net of fissile plutonium consumed. Used PRISM fuel is recycled after the removal of fission products, though not necessarily into PRISM units [1–63].
Annular Flow Simulation Supported by Iterative In-Memory Mesh Adaptation
Published in Nuclear Science and Engineering, 2020
Jun Fang, Meredith K. Purser, Cameron Smith, Ramesh Balakrishnan, Igor A. Bolotnov, Kenneth E. Jansen
Light water reactors are the dominant type of reactor among the modern-day commercial nuclear reactor fleet. Because water is adopted as the working fluid, steam-water two-phase flow becomes an essential thermal-hydraulic phenomenon to consider in reactor designs. Such examples can be found from subcooled boiling in a pressurized water reactor1 (PWR), to the churn-turbulent and annular flows in a boiling water reactor (BWR) core.2 Under normal operation conditions, complex two-phase flow patterns/regimes are to be observed in a BWR because the coolant is continuously heated along the rising passage with increasing steam quality. A considerable amount of literature has been published on classifying the flow regimes and transition criteria in fuel rod bundle geometries.2–4 As the final stage of two-phase flow regime transition, the annular flow is especially challenging to study both experimentally and computationally. Annular two-phase flow distinguishes itself by the presence of a liquid film flowing along channel walls and a gas core occupying central regions of the channel. The gas core contains droplets, which participate in mass, momentum, and energy transfer with the liquid film through two primary mechanisms: (1) entrainment of droplets from the liquid film and (2) deposition of droplets from the gas core to the liquid film.5 The dynamics occurring on the agitated wavy interface between the liquid film and gas core plays a crucial role in affecting the thermohydraulic processes in the reactor core.6 For instance, the amount of droplet entrainment as well as the rate of entrainment significantly affects the occurrences of the dryout, whereas the post critical heat flux (CHF) heat transfer depends strongly on the entrainment and droplet sizes.7