Hi everyone,

I am encountering a 22% difference in neutron flux calculated between OPENMC and MCNP6 for a simple system with a fixed source. The flux was calculated in the volume of the system. From what I have read on the forum, it seems that the issue might be related to flux normalization, particularly when using the track-length estimator.

In MCNP6, I used the F4 tally with the SDEF source card, energy distribution and the total source weight. In OPENMC, I used a fixed source with the same energy distribuition and the total source weight, but the calculated flux was later divided by the cell volume.

```
**Source MCNP:**
sdef pos=0 0 0 axs=0 0 1 vec=1 0 0 erg=d1 cel=1 rad=d2 ext=d3 wgt=1e20
si1 H 2.5119E-08 .... $ value in MeV
sp1 D 0.00 ....
si2 0 30.760
si3 150
nps 10000
**Source OPENMC:**
settings = openmc.Settings()
settings.output = {'tallies': True}
settings.particles = 10000
settings.batches = 50
settings.inactive = 0
settings.run_mode = 'fixed source'
strength = 1E+20
space = openmc.stats.CylindricalIndependent(
r=openmc.stats.Uniform(0, 30.760),
phi=openmc.stats.Uniform(0, 2 * np.pi),
z=openmc.stats.Uniform(150, -150),
origin=(0.0, 0.0, 0.0)
)
angle = openmc.stats.Isotropic()
energy = openmc.stats.Tabular([ 2.5119E-01, .....], [0.00, .....]) #value in eV
source = openmc.IndependentSource(strength=strength,angle=angle, space=space,energy=energy)
settings.source = source
settings.export_to_xml()
```

I’m not sure how to reduce this difference between the results from the two codes. Could someone provide guidance on how to proceed or if there is something specific I should adjust in the normalization process?

Thanks in advance for any help!