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Thermodynamic Properties and Equations of State
Published in Robert E. Masterson, Nuclear Reactor Thermal Hydraulics, 2019
The region between the saturated liquid line (on the left) and the saturated vapor line (on the right) is called the vapor dome. If the coolant represented by this dome is ordinary water, then the vapor dome is called a steam dome. The region immediately to the left of the vapor dome (region 1 in Figure 7.7) is called the subcooled liquid region, and the region immediately to the right of the vapor dome (region 3 in Figure 7.7) is called the superheated vapor region. The region under the vapor dome (region 2) is called the two-phase mixture region. Any coolant that undergoes a phase change will pass through all three of these regions. At the top of the vapor dome, the liquid and vapor phases blend together, and when this occurs, it becomes impossible to distinguish them from each other. On T–V, P–T, or T–S diagrams, the point at which they become indistinguishable is called the critical point. The triple point for water at 273.16°K (0°C) occurs at 611.2 Pa. In classical thermodynamics, triple point becomes the basis for the definition of the Kelvin.
Boiling Water Reactors
Published in Robert E. Masterson, Nuclear Engineering Fundamentals, 2017
The water level in the pressure vessel comes up to about the middle of the steam separators. Above this, the pressure vessel is filled with steam, and below this, the pressure vessel is filled with water. The top of the pressure vessel in a BWR is also called the steam dome. The steam line exits from the top of the steam dome and sends the heated steam to the power turbines. A BWR produces saturated steam at about 286°C and at about 1050 PSI (7.15 MPa). This gives a typical BWR a thermal efficiency between 33% and 34%. However, because of the lower operating pressure and the lower power density, the enrichment of the fuel in the core is lower (about half of what it is in a PWR). Thus, if the new fuel assemblies in the core of a PWR have an average enrichment of 4.5%, the new fuel assemblies in the core of a BWR have an average enrichment of about 3%.
Temperature Coupling Analysis Between Nuclear Steam Generators and Heat Exchanger Inside Pressurized Water Reactors
Published in Nuclear Science and Engineering, 2021
Mohamed S. El-Tokhy, Imbaby I. Mahmoud
The coolant mass flow coming from the reactor is directed into the SG at the initial temperature of Tp1, which is the preliminary temperature of the considered model. The exchange of energy through the secondary coolant happens within the U-tube shape. The water from the downcomer is heated by rising part of the U-bundle, which corresponds to the external area of the shroud. The feeder fluid in the economizer area is preheated due to the continuous flow to the lower part of the U-tube. This occurs before departure from the SG at a temperature of Tp2 (Ref. 33). Node one stands for the passing volume via the primary coolant. Energy is transferred to the secondary fluid from the primary coolant through the casing of the U-tube material. Node two corresponds to metal volume. The observable fractional vapor of the secondary water within the shroud is due to the transfer of energy via the U-tube material. The flow is considered in two stages as it arrives at the SG. Stimulated steam comes into the steam dome that is positioned over the steam separator. A combination of regulated valves within the UTSG helps exit the steam. The input values correspond to the steam mass flow.