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NUCLEAR (Module 6)
35問 • 1年前
  • Alexander Naje
    • 通報

    問題一覧

  • 1

    Quantifies the most important neutron - physical processes.

    Neutron Life Cycle

  • 2

    Neutrons produced from fission are slowed down to thermal (low-energy) neutrons.

    Thermalization

  • 3

    Study of time-dependent behavior of a nuclear reactor in relation to changes in neutron population over time.

    Reactor Kinetics

  • 4

    Ratio of Neutrons

    Multiplication factor (k)

  • 5

    k=1

    Critical (Steady power)

  • 6

    k<1

    Subcritical (Power decreases)

  • 7

    k>1

    Supercritical (Power increases)

  • 8

    it is how far the reactor is from criticality.

    Reactivity (p)

  • 9

    Incorporates prompt and delayed neutrons.

    Point Kinetic Equation

  • 10

    Group of delayed neutron precursors.

    Precursor Concentration Equation

  • 11

    Fission event releases 200 mev. Some energy is carried away by neutrons & gamma rays. E=mc².

    Fission Energy Calculation

  • 12

    Amount of power produced per unit volume of fuel.

    Power Density

  • 13

    Sum of energy generated by fissions per unit time.

    Thermal Power Calculation

  • 14

    Heat is conducted first through the fuel material to cladding.

    Conduction

  • 15

    Heat is transferred from cladding surface to coolant.

    Convection

  • 16

    rate at which coolant flows through the core affects how much heat is carried away.

    Coolant flow rate calculations

  • 17

    Prevent hot spots and ensure uniform cooling.

    Thermal hydraulics

  • 18

    Provides a theoretical basis for neutron - physical computing of reactor core. Used diffusion equation for spatial flux distribution. Fick's law

    Neutron Diffusion Theory

  • 19

    Modeling the behavior of neutrons as they move through reactor core.

    Transport Theory

  • 20

    Ensures safe and efficient nuclear operations

    Reactor control and safety calculation

  • 21

    Amount of negative reactivity that can be inserted into the reactor core.

    Safety Shutdown Margin (SDM)

  • 22

    ensures that reactor doesn't overheat.

    Coolant flow & Heat transfer

  • 23

    Reactor Protection System

    Coolant Temperature (CT: Monitor Coolant temp.) Pressure (Prevent the coolant from boiling) Neutron Flux Monitoring (Measure Neutron Flux)

  • 24

    Point at which coolant can no longer remove heat from core

    Critical Heat Flux

  • 25

    Determines the response of reactor Insert reactivity and cool the core through passive safety systems.

    Loss of Coolant Accident (LOCA)

  • 26

    Calculates the effectiveness of control rods and moderators.

    Neutron Flux

  • 27

    Determines the overall neutron population.

    Neutron Balance

  • 28

    Key Factors Affecting Shut Down Margin

    Control Rods Burn Up Coolant temperature Xenon Poisoning Borated Water

  • 29

    Simulates the transport of neutrons, photons, electrons, ions, and other elementary particles

    Monte Carlo N - Principle (MCNP)

  • 30

    State of the art lattice physics code Calculate fuel depeletion & cross-section data Use for PWR and BWR fuel assemblies

    CASMO 5

  • 31

    LANL

    Los Alamos National Library

  • 32

    3D continuous energy Monte Carlo Neutron and photon transport code Burnup calculation capability

    SERPENT

  • 33

    Creating mathematical virtual representation of reactor and its components (fuel rods, control rods, coolant, etc.)

    Reactor Modeling

  • 34

    Virtual Experiments Running a model to prodect how the reactor will behave and respond to different situations.

    Simulation

  • 35

    solute diffuses from high to low concentration

    Fick's Law

    • NUCLEAR (Module 7)

    NUCLEAR (Module 7)

    Alexander Naje · 26問 · 1年前

    NUCLEAR (Module 7)

    NUCLEAR (Module 7)

    26問 • 1年前
    Alexander Naje

    NUCLEAR (Module 8)

    NUCLEAR (Module 8)

    Alexander Naje · 39問 · 1年前

    NUCLEAR (Module 8)

    NUCLEAR (Module 8)

    39問 • 1年前
    Alexander Naje

    問題一覧

  • 1

    Quantifies the most important neutron - physical processes.

    Neutron Life Cycle

  • 2

    Neutrons produced from fission are slowed down to thermal (low-energy) neutrons.

    Thermalization

  • 3

    Study of time-dependent behavior of a nuclear reactor in relation to changes in neutron population over time.

    Reactor Kinetics

  • 4

    Ratio of Neutrons

    Multiplication factor (k)

  • 5

    k=1

    Critical (Steady power)

  • 6

    k<1

    Subcritical (Power decreases)

  • 7

    k>1

    Supercritical (Power increases)

  • 8

    it is how far the reactor is from criticality.

    Reactivity (p)

  • 9

    Incorporates prompt and delayed neutrons.

    Point Kinetic Equation

  • 10

    Group of delayed neutron precursors.

    Precursor Concentration Equation

  • 11

    Fission event releases 200 mev. Some energy is carried away by neutrons & gamma rays. E=mc².

    Fission Energy Calculation

  • 12

    Amount of power produced per unit volume of fuel.

    Power Density

  • 13

    Sum of energy generated by fissions per unit time.

    Thermal Power Calculation

  • 14

    Heat is conducted first through the fuel material to cladding.

    Conduction

  • 15

    Heat is transferred from cladding surface to coolant.

    Convection

  • 16

    rate at which coolant flows through the core affects how much heat is carried away.

    Coolant flow rate calculations

  • 17

    Prevent hot spots and ensure uniform cooling.

    Thermal hydraulics

  • 18

    Provides a theoretical basis for neutron - physical computing of reactor core. Used diffusion equation for spatial flux distribution. Fick's law

    Neutron Diffusion Theory

  • 19

    Modeling the behavior of neutrons as they move through reactor core.

    Transport Theory

  • 20

    Ensures safe and efficient nuclear operations

    Reactor control and safety calculation

  • 21

    Amount of negative reactivity that can be inserted into the reactor core.

    Safety Shutdown Margin (SDM)

  • 22

    ensures that reactor doesn't overheat.

    Coolant flow & Heat transfer

  • 23

    Reactor Protection System

    Coolant Temperature (CT: Monitor Coolant temp.) Pressure (Prevent the coolant from boiling) Neutron Flux Monitoring (Measure Neutron Flux)

  • 24

    Point at which coolant can no longer remove heat from core

    Critical Heat Flux

  • 25

    Determines the response of reactor Insert reactivity and cool the core through passive safety systems.

    Loss of Coolant Accident (LOCA)

  • 26

    Calculates the effectiveness of control rods and moderators.

    Neutron Flux

  • 27

    Determines the overall neutron population.

    Neutron Balance

  • 28

    Key Factors Affecting Shut Down Margin

    Control Rods Burn Up Coolant temperature Xenon Poisoning Borated Water

  • 29

    Simulates the transport of neutrons, photons, electrons, ions, and other elementary particles

    Monte Carlo N - Principle (MCNP)

  • 30

    State of the art lattice physics code Calculate fuel depeletion & cross-section data Use for PWR and BWR fuel assemblies

    CASMO 5

  • 31

    LANL

    Los Alamos National Library

  • 32

    3D continuous energy Monte Carlo Neutron and photon transport code Burnup calculation capability

    SERPENT

  • 33

    Creating mathematical virtual representation of reactor and its components (fuel rods, control rods, coolant, etc.)

    Reactor Modeling

  • 34

    Virtual Experiments Running a model to prodect how the reactor will behave and respond to different situations.

    Simulation

  • 35

    solute diffuses from high to low concentration

    Fick's Law