問題一覧
1
Calculates the effectiveness of control rods and moderators.
Neutron Flux
2
Provides a theoretical basis for neutron - physical computing of reactor core. Used diffusion equation for spatial flux distribution. Fick's law
Neutron Diffusion Theory
3
LANL
Los Alamos National Library
4
Fission event releases 200 mev. Some energy is carried away by neutrons & gamma rays. E=mc².
Fission Energy Calculation
5
Virtual Experiments Running a model to prodect how the reactor will behave and respond to different situations.
Simulation
6
Heat is conducted first through the fuel material to cladding.
Conduction
7
3D continuous energy Monte Carlo Neutron and photon transport code Burnup calculation capability
SERPENT
8
Sum of energy generated by fissions per unit time.
Thermal Power Calculation
9
Determines the overall neutron population.
Neutron Balance
10
Modeling the behavior of neutrons as they move through reactor core.
Transport Theory
11
ensures that reactor doesn't overheat.
Coolant flow & Heat transfer
12
Amount of power produced per unit volume of fuel.
Power Density
13
solute diffuses from high to low concentration
Fick's Law
14
Neutrons produced from fission are slowed down to thermal (low-energy) neutrons.
Thermalization
15
Quantifies the most important neutron - physical processes.
Neutron Life Cycle
16
Creating mathematical virtual representation of reactor and its components (fuel rods, control rods, coolant, etc.)
Reactor Modeling
17
Key Factors Affecting Shut Down Margin
Control Rods Burn Up Coolant temperature Xenon Poisoning Borated Water
18
it is how far the reactor is from criticality.
Reactivity (p)
19
Determines the response of reactor Insert reactivity and cool the core through passive safety systems.
Loss of Coolant Accident (LOCA)
20
State of the art lattice physics code Calculate fuel depeletion & cross-section data Use for PWR and BWR fuel assemblies
CASMO 5
21
Incorporates prompt and delayed neutrons.
Point Kinetic Equation
22
Heat is transferred from cladding surface to coolant.
Convection
23
Reactor Protection System
Coolant Temperature (CT: Monitor Coolant temp.) Pressure (Prevent the coolant from boiling) Neutron Flux Monitoring (Measure Neutron Flux)
24
k<1
Subcritical (Power decreases)
25
Point at which coolant can no longer remove heat from core
Critical Heat Flux
26
Amount of negative reactivity that can be inserted into the reactor core.
Safety Shutdown Margin (SDM)
27
Ratio of Neutrons
Multiplication factor (k)
28
Simulates the transport of neutrons, photons, electrons, ions, and other elementary particles
Monte Carlo N - Principle (MCNP)
29
k=1
Critical (Steady power)
30
Study of time-dependent behavior of a nuclear reactor in relation to changes in neutron population over time.
Reactor Kinetics
31
Prevent hot spots and ensure uniform cooling.
Thermal hydraulics
32
Ensures safe and efficient nuclear operations
Reactor control and safety calculation
33
rate at which coolant flows through the core affects how much heat is carried away.
Coolant flow rate calculations
34
Group of delayed neutron precursors.
Precursor Concentration Equation
35
k>1
Supercritical (Power increases)