Methodology for modeling the fuel rod power distribution within a nuclear reactor core
View Patent ↗A method for modeling a nuclear reactor core that follows the history of each fuel pin and employs fuel pin flux form factors to explicitly track each fuel pin's fluence and burnup along its axial length and uses this information to obtain fundamental data for each fuel rod, i.e. fuel rod cross-sections, for each fuel pin segment. The data obtained for the fuel pins segments are employed to adjust the fuel pin flux form factors to match the real fuel pins' history so that the fuel rod power distribution can be precisely calculated based on the fuel rod cross-sections and the flux form factors.
1. A method of modeling an axial and radial fuel rod by fuel rod power distribution of a nuclear fuel assembly by determining a neutronic cross section for each fuel rod within the fuel assembly individually over a number of axial increments, comprising the steps of:
a) defining 2×2 radial nodes per assembly for each of the axial increments and calculating a three dimensional node average and surface flux distribution for each 2×2 node;
b) calculating a reference kappa-fission, absorption, and neutron flux form factors for each fuel rod based on a number of reference values and local conditions based upon a reference history;
c) calculating actual pin neutronic cross-sections (taking into account kappa-fission and absorption) for each of the axial increments by applying a fuel rod fast fluence correction to the results obtained in step b), and adjusting the fuel rod flux form factors using the following equation:
f
g
act
(
x
,
y
)
=
f
g
ref
(
x
,
y
)
·
∏
g
′
=
1
,
g
∑
a
g
′
,
ref
·
∑
g
′
->
g
act
∑
a
g
′
,
act
·
∑
g
′
->
g
ref
;
wherein
f g ref (x, y) is the reference neutron flux form factor of a fuel rod at location (x, y) for energy group “g”, pre-generated at hot full power condition, Σ a g,ref and Σ a g,act are respectively the reference and actual absorption cross-sections of the fuel rod for energy group “g”, while Σ g′→g ref and Σ g′→g act are the neutron scattering cross-sections from energy group “g” to “g”, according to a pin neutronic cross-section change so that both pin neutronic cross-sections and pin flux form factors correspond to an actual fuel rod and fuel assembly history;
d) generating homogeneous neutron flux values for an adjacent group of fuel rods in the fuel assembly;
e) calculating a number of heterogeneous neutron fluxes of each of the adjacent group of fuel rods respectively from the homogeneous neutron flux values obtained in step d) and the neutron flux form factors;
f) using a fuel rod axial interaction correction model to adjust the heterogeneous neutron fluxes to take into account an axial interaction of each fuel rod between different axial levels; and
g) determining the power for each fuel rod from a constant times the sum of a product of the calculated kappa-fissions and the heterogeneous neutron fluxes over a number of energy groups.
2. The method of claim 1 wherein the neutron flux form factor is a function of the composition and enrichment of the fuel rod.
3. The method of claim 1 wherein each of the flux form factors initially determined for each fuel rod to simulate the fuel rods history is a reference flux form factor, which does not directly take into account the in-core placement history of the fuel assembly.
4. The method of claim 3 including the step of adjusting the reference flux form factors to take into account the affect on fuel rod from the actual history of control rod insertion into the fuel assembly.
5. The method of claim 4 wherein the control rods are inserted within the fuel assembly for a preselected percentage of an operating cycle of the fuel assembly.
6. The method of claim 3 wherein the adjustment to the flux form factor reflects the fuel assemblies previous history including placement, local power and burnup experienced by each fuel assembly over previous loading cycles within the core.