TAILIEUCHUNG - Pressurized Water Reactor Systems

State the mechanism for steam generation in a PWR. Using simplified diagrams, identify and explain the purpose of the major components and equipment involved with a PWR. Identify the equipment used to control reactor power. | ACADs (08-006) Covered Keywords Chemical and volume control, circulating water, condensate, containment, containment sprat, control rod drive, emergency core cooling, emergency power, feedwater, PWR, BWR. Description The purpose of this class is to familiarize students with the general components of a pressurized water reactor plant. Supporting Material TERMINAL OBJECTIVES Students will understand various PWR systems and the operation of each. Students will understand the differences between PWR and BWR reactor plants. PWR SYSTEMS ENABLING OBJECTIVES State the mechanism for steam generation in a PWR. Using simplified diagrams, identify and explain the purpose of the major components and equipment involved with a PWR. Identify the equipment used to control reactor power. Exhibit a basic understanding of steam generator theory and operation. Exhibit a basic understanding of pressurizer theory and operation. Compare and contrast PWR and BWR reactor plant design with advantages and disadvantages of each. PWR SYSTEMS GENERAL PLANT OVERVIEW PWR SYSTEMS The primary system (also called the Reactor Coolant System) consists of the reactor vessel, the steam generators, the reactor coolant pumps, a pressurizer, and the connecting piping. A reactor coolant loop is a reactor coolant pump, a steam generator, and the piping that connects these components to the reactor vessel. The primary function of the reactor coolant system is to transfer the heat from the fuel to the steam generators. A second function is to contain any fission products that escape the fuel. The following drawings show the layout of the reactor

• Năng lượng
• The reactor vessel
• The steam generators
• The reactor coolant pumps
• Pressurized Water Reactor Systems
• BWR reactor plant design
• Reactor coolant system
• Reactor vessel water level estimation during severe accidents
• Fuzzy neural networks
• Reactor Vessel Water Level
• Repeatedly adding fuzzy neural networkshj
• Estimating the reactor vessel water level
• Nuclear engineering and technology
• Structural assessment of reactor pressure vessel
• Multi layered corium formation conditions
• External reactor vessel cooling
• Representative reactor pressure vessel
• In vessel melt retention
• Severe accident
• Melcor input for vver1000 reactor
• Failure of lower core plate
• Nuclear engineering and design
• Downward facing surface
• High surface heat flux
• Inclination angle
• In vessel retention
• Reactor pressure vessel steel
• Multi nozzle array
• Development of multi cell flows in the three layered configuration
• Oxide layer and their influence on the reactor vessel heating
• The heavy metal layer
• Analytical method
• Station Black Out
• Reactor vessel external cooling strategy
• Comparison of applicability of current transition temperature shift models
• SA533B 1 reactor pressure vessel steel of Korean nuclear reactors
• Korean nuclear reactors
• Aged reactor pressure vessels
• Nuclear reactor vessel water level prediction during severe accidents
• Deep neural networks
• Nuclear power plants
• Maintain nuclear reactor integrity
• Reactor pressure vessel
• Structural integrity
• Probabilistic assessment
• Pressurized thermal shock
• Light water reactor
• Dynamic characteristics assessment of reactor vessel internals
• Fluid structure interaction
• Response Spectrum Analysis
• Identification of nonregular indication according
• Change of grain size surface geometry
• Nuclear power plant (NPP) reactor vessel
• Upper head alloy 690 penetration
• Prediction of the reactor vessel water level
• Severe accident circumstances of NPPS
• Severe accident circumstances
• Fuzzy Neural Network
• Probabilistic Fracture Mechanics
• Beltline Region
• Oak Ridge National Laboratory
• Reactor pressure vessel embrittlement
• DPA rate
• Weight window technique
• Radiation embrittlement
• Displacement per atom
• Monte Carlo code MCNP5
• Heat transfer characteristics
• The one side heated vertical channel
• Inserted porous materials applied as a vessel cooling system
• Vessel cooling system
• High temperature reactor
• Small integral reactor
• Severe accident mitigation
• Station blackout
• Hydrogen control
• Containment integrity
• Containment Bypass
• SBO sequence
• VVER 1000/V 320 reactor
• Nuclear regulatory commission
• Reactor vessel
• Fuel Assemblies
• Reactivity Control
• BWR Generations
• Control rods
• Integrity analysis of an upper guide structure flange
• Upper guide structure flange
• Flange design meets
• Design Acceptance Criteria
• Reactor Vessel Internals
• Nuclear science and technology
• The journal of Nuclear science and technology
• Aquatic environment
• Geochemical parameters
• External reactor vessel cooling method