Normalization of Tally Results

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1

Dear experts,
I have 2 questions:

  1. When normalizing neutron flux in OpenMC for a mesh-based tally, should the volume used in the normalization formula below:

image

refer to:

  1. The volume of the entire reactor core, or
  2. The volume of a single mesh cell where the flux is being calculated?

Also,
2) In the formula for normalization:

image

if we’re interested in obtaining absolute values of the thermal neutron flux, should we:

  • Use the thermal heating rate specifically for its evaluation,
  • Or rely on the total heating rate (sum of thermal and fast contributions) for the normalization factor?
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2

J_per_eV = 1.602e-19 # [J/eV]
P = 1e7 # System power in watts (10 MW), adjust accordingly

Total heating rate [J/source]

total_heating_rate = J_per_eV * (thermal_heating + fast_heating)

Normalization factor [source/s]

f = P / np.sum(total_heating_rate)

Volume of each mesh cell

mesh_volume = ((52.4 - (-52.4)) / mesh.dimension[0]) ** 2 # [cm^3]

Absolute flux [particles/cm^2/s]

thermal_flux_abs = (f * thermal_flux) / mesh_volume
fast_flux_abs = (f * fast_flux) / mesh_volume

Reshape the data to match mesh dimensions

thermal_flux_abs = thermal_flux_abs.reshape(mesh.dimension)
fast_flux_abs = fast_flux_abs.reshape(mesh.dimension)

when i use above logic, the flux is about 10E+16 n/cm^2/sec, but based on experiments, it should be about 10E+14 n/cm^2/sec

3

Hi，
considering my experience, 1. the volume in the flux calculation is the volume of the region in which you tally the flux, 2. to get the right f factor, the H’ should be the heating of the whole core and the whole energy, for corresponding to the total system power.

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