MCNP Flux and Power Calculation

During a reactor assembly calculation, I need to determine axial and radial flux distribution over the surface. When I use F2 and F4 tally I get some value with unit 1/cm**2
What does the value means, neutron flux is supposed to be in the 10^14 range but output values are 10^2 range.
Can anyone guide me on how to determine axial and radial flux and also power distribution in a reactor assembly in MCNP6.2?

I'll start with the basic stuff I know. Flux is particles per centimeter squared. In MCNP this is per source particle. Dimensionally particles per cm² per particle cancel out to 1/cm². ** just means to the power of.

Your actual flux number looks a little odd. I would check the tally divisors are correct, surface area if F2, volume if F4. Showing us your input file is helpful, without it this is just a guess.

Your question implies a critical reactor, and in a state of k=1 the neutron flux is whatever you want it to be. Set up the problem, get normalised results from MCNP and then calculate the answer using whatever values you know for the problem.

I had a bit of an idiot moment reading your question and for a short time was wondering why you would want flux resolved parallel to a cylinders surface. My googling found https://www.researchgate.net/publication/349683390_Calculation_of_the_Radial_and_Axial_Flux_and_Power_Distribution_for_a_CANDU_6_Reactor_with_both_the_MCNP6_and_Serpent_Codes_The_MCNP61_and_Serpent_1119_codes_were_used_for_Monte_Carlo_transport_calcul which ended my confusion and seems to address your problem. There is no example sadly, but it goes through the method quite well step by step. Have a read and see if it answers any other of your questions.

From my understanding, the value 1/cm**2 represents the number of neutrons passing through a unit area (1 cm**2) per unit time. This value is typically used to measure the neutron flux, which is the rate at which neutrons are moving through a given area.

In order to determine the axial and radial flux distribution, you can use the F2 and F4 tally results in conjunction with the geometry of your reactor assembly. These tallies will give you the neutron flux at specific points in the assembly, which you can then use to create a distribution map.

To determine the power distribution, you can use the flux results along with the cross-sections of the materials in your assembly to calculate the power generated at each point. This can be done using the formula P = Φ * σ * V, where P is power, Φ is flux, σ is cross-section, and V is volume.

I suggest consulting the MCNP6.2 user manual or seeking guidance from experienced users for more specific instructions on how to perform these calculations.