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