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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 In CANDU and graphite reactors, respectively, heavy water (water in which some or all of the hydrogen, H-1, has been replaced by deuterium, H-2) and graphite are the moderators.
Radiation protection in the nuclear industry
Published in Alan Martin, Sam Harbison, Karen Beach, Peter Cole, An Introduction to Radiation Protection, 2018
Alan Martin, Sam Harbison, Karen Beach, Peter Cole
The great majority of commercial nuclear power reactors currently operating or under construction worldwide are light water reactors (LWRs), which use water as both a coolant and a neutron moderator (see Section 13.3.2). These are of two types, pressurized water reactors (PWRs) and boiling water reactors (BWRs). In a PWR, a closed water-coolant system transfers heat from the core to heat exchangers, which raise steam in a secondary circuit to drive a turbo-generator (see Figure 13.5). Bulk boiling of the water in a PWR is prevented by maintaining the system at a very high pressure (about 2200 psi or 150 bar). By contrast, in a BWR the pressure of the system is such that the water boils and the resulting steam, after passing through a steam separator, enters the turbines directly. Typical PWR and BWR plants are illustrated in Figures 13.6 and 13.7. Another type of water-cooled reactor is the Canadian CANDU system, which uses heavy water as the coolant and neutron moderator.
Heat Transfer, Thermal Hydraulic, and Safety Analysis
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
Up to 50% of the reactor power, the heat generation rate can be assumed to follow proportional to the neutron flux. However, if the reactor is shut down, the reactor core produces heat at 7% of the reactor power due to delayed neutron fission, fission product decay, and activation products from neutron capture. The heat generated from fission product decay due to high-power operation can be temporarily higher than the heat generated from the fission product decay at low-power operation. Because of the decay heat shutdown cooling has to be provided for the reactor core. Immediately after the shutdown the reactor power decreases exponentially from 7% of the core power with a period of 80 s. The heat generation from fission product decay is a result of beta and gamma emission from fission product with decay half-life ranging from micro seconds to million years. Decay heat is the principal reason of safety concern in light water reactors (LWRs), and it is the source of 60% of radioactive release risk worldwide.
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
Identifying Thermodynamic Mechanisms Affecting Reactor Pressure Vessel Integrity During Severe Nuclear Accidents Simulated by Laser Heating at the Laboratory Scale
Published in Nuclear Science and Engineering, 2023
Michail Athanasakis-Kaklamanakis, Dario Manara, Luka Vlahovic, Davide Robba, Konstantinos Boboridis, Markus Ernstberger, Rachel Eloirdi, Pedro Amador, Rudy J. M. Konings
Melting of the core materials in a nuclear light water–cooled reactor (LWR) is one of the harshest consequences of a loss-of-coolant accident (LOCA). As the temperature in the fuel rises in the absence of sufficient cooling, the structural and fuel elements progressively reach their melting points to form complex molten mixtures that contain an abundance of chemical elements. Because of gravity, these mixtures relocate, eventually finding their way to the lower head of the reactor pressure vessel (RPV) to form a molten pool referred to as in-vessel corium.1