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Principal Characteristics of Power Reactors
Published in Neil E. Todreas, Mujid S. Kazimi, Nuclear Systems Volume I, 2021
Neil E. Todreas, Mujid S. Kazimi
The Generation II BWR single-loop primary coolant system is illustrated in Figure 1.6, while Figure 1.7 highlights the flow paths within the reactor vessel. The steam–water mixture first enters the steam separators after exiting the core. After subsequent passage through a steam separator and dryer assembly located in the upper portion of the reactor vessel, dry saturated steam flows directly to the turbine. Saturated water which is separated from the steam flows downward in the periphery of the reactor vessel and mixes with the incoming main feed flow from the condenser. This combined flow stream is pumped into the lower plenum through jet pumps mounted around the inside periphery of the reactor vessel. The jet pumps are driven by flow from recirculation pumps located in relatively small-diameter (~50 cm) external recirculation loops, which draw flow from the plenum just above the jet pump discharge location. In the ABWR, all external recirculation loops are eliminated and replaced with recirculation pumps placed internal to the reactor vessel. In the economic simplified boiling water reactor (ESBWR), all jet pumps as well as external recirculation pumps were eliminated by the natural circulation flow design.
Fuel Pins, Fuel Rods, Fuel Assemblies, and Reactor Cores
Published in Robert E. Masterson, Nuclear Engineering Fundamentals, 2017
This makes it easier to control a BWR in the event of a reactor transient because the steam generated cannot move out of one fuel assembly and into another. Sometimes, the cans around BWR fuel assemblies are also called sheaths. Their size and shape have also evolved slightly over the years. Today, there are three common types of control rods used in BWRs. They are the BWR-1, BWR-2, and BWR-3 control rods, the BWR-4 and BWR-5 control rods, and the BWR-6 and BWR-7 control rods. Each successive design generation of BWRs has used a slightly different control rod configuration as the power level of an average BRW has increased. It is likely that another design will go into large-scale production around the year 2020 with the advent of the advanced boiling water reactor (ABWR) and the economic simplified boiling water reactor (ESBWR). You can learn more about their advantages and drawbacks at the following URL:
Risk Assessment and Safety Analysis for Commercial Nuclear Reactors
Published in Kenneth D. Kok, Nuclear Engineering Handbook, 2016
The last commercial nuclear power plant OL was issued in 1993. Since then, the US nuclear industry went into a “dormant period,” and only recently has the momentum for building new nuclear power plants resurged. Specifically, 28 new plants at 19 sites are proposed by the nuclear industry. The new plants are based on five different reactor designs: GE nuclear energy economic simplified boiling water reactor (ESBWR, 1500 MWe): currently under the NRC staff reviewGE Nuclear Energy Advanced Boiling Water Reactor (ABWR, 1350 MWe): design approved in July 1994 and certified in May 1997Westinghouse Advanced Pressurized Water Reactor (APWR): System 80+ standard plant design approved in July 1994 and certified in May 1997Westinghouse standard plant design AP-1000, 1000 MWe: design approved in September 2004; also, AP-600 (600 MWe) standard plant design: design approved in September 1998 and certified in December 1999AREVA’s European Evolutionary Pressurized Reactor (EPR, 1600 MWe): currently under NRC staff review
Severe Accident Phenomena: A Comparison Among the NuScale SMR, Other Advanced LWR Designs, and Operating LWRs
Published in Nuclear Technology, 2020
Scott J. Weber, Etienne M. Mullin
Large ALWRs build off of these strategies while also adding design features to decrease the likelihood of containment failure. For example, the Economic Simplified Boiling Water Reactor (ESBWR) design implements the basemat internal melt arrest and coolability device, which is a passively cooled barrier incorporated into the lower drywell floor that functions to halt thermal attack from core debris. This reduces containment pressure and prevents failure by melt-though. Alternatively, AP1000 credits in-vessel retention for the majority of accident sequences and includes the capability to flood the reactor cavity by gravity draining from the in-containment refueling water storage tank. Both designs have also incorporated additional depressurization systems for the primary vessel, reducing the likelihood of HPME, as well as highly reliable passive containment cooling systems to reduce the likelihood of gradual overpressurization.12,13
Validation and Uncertainty Quantification for Two-Phase Natural Circulation Flows Using TRACE Code
Published in Nuclear Science and Engineering, 2020
Katarzyna Borowiec, Tomasz Kozlowski, Caleb S. Brooks
In recent years, natural circulation has been considered a passive safety feature of decay heat removal in advanced reactor designs. Some of the designs rely on natural circulation for the removal of nominal fission heat, e.g., the Economic Simplified Boiling Water Reactor (ESBWR). Nuclear system codes must be validated against those conditions. Under natural circulation conditions the sensitivity of the void fraction is higher with respect to mass flow rate and power, which makes modeling even more challenging. Under conditions of low pressure and low flow rate, density wave or flow excursion can cause sporadic dryout on the heater surface.3