Number of secondaries generated in a volume MCNP?

In summary, to get the number of secondaries in a certain volume in MCNP, you can use the F4 tally with a tally multiplier card. This will multiply the flux with the atom density and microscopic cross section of the desired reaction, giving you the total number of particles formed in the target. A complete list of reaction types and numbers can be found in Appendix G of the MCNP5 manual Vol I, and an example of a similar problem can be found on Page 4-39 of the MCNP5 manual Vol II.
  • #1
Neo Tran
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Hello everyone,
I am having a problem with MCNP. My question is how to get number of secondaries in a certain volume. For example I have a neutron beam bombarded Pb target, and I want to count all of proton formed in the target. I considered tally F4, but the unit is 1/cm**2. Who can explain the tally to me? More over, help me solve this problem, please.
Thank you very much,Neo.
 
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  • #2
For this, you will need to use the tally multiplier functionality with the F4 tally. F4 gives a flux averaged over the volume of the cell in units of particles/cm**2 per source particle. Using a tally multiplier card, you can get MCNP to multiply this flux with the atom density of the target nuclide and the microscopic cross section of the desired reaction (in this case, (n,p) reaction). If you multiply this value with the source term, you'll get the total number of protons formed in the target.
This is what your MCNP cards with look like (assuming a target cell 1):

F4:N 1
FM4 C M (103) (203)

Where C is the atom density of lead in your target, M is the corresponding material number in your input file for lead (assuming it is pure elemental lead)
103 is the reaction number for (n,p) reaction and 203 is a reaction number for total number of protons produced by all reactions. You can find a complete list of all reaction types and numbers in Appendix G of the MCNP5 manual Vol I
You can also find an example of the similar problem on Page 4-39 of the MCNP5 manual Vol II
 
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Likes Ericdjs and Neo Tran
  • #3
quarkle said:
For this, you will need to use the tally multiplier functionality with the F4 tally. F4 gives a flux averaged over the volume of the cell in units of particles/cm**2 per source particle. Using a tally multiplier card, you can get MCNP to multiply this flux with the atom density of the target nuclide and the microscopic cross section of the desired reaction (in this case, (n,p) reaction). If you multiply this value with the source term, you'll get the total number of protons formed in the target.
This is what your MCNP cards with look like (assuming a target cell 1):

F4:N 1
FM4 C M (103) (203)

Where C is the atom density of lead in your target, M is the corresponding material number in your input file for lead (assuming it is pure elemental lead)
103 is the reaction number for (n,p) reaction and 203 is a reaction number for total number of protons produced by all reactions. You can find a complete list of all reaction types and numbers in Appendix G of the MCNP5 manual Vol I
You can also find an example of the similar problem on Page 4-39 of the MCNP5 manual Vol II
You was right. Thank you very much Quarkle. It is very useful for me in this time.
 

1. How is the number of secondaries generated in a volume calculated in MCNP?

The number of secondaries generated in a volume in MCNP is calculated using the Monte Carlo method, which involves simulating the transport of particles through a specified volume and tracking the number of secondary particles produced as a result of interactions with the volume's materials.

2. What factors affect the number of secondaries generated in a volume in MCNP?

The number of secondaries generated in a volume in MCNP can be affected by several factors, including the type and energy of the incident particle, the properties of the volume's materials, and the size and shape of the volume.

3. Can the number of secondaries generated in a volume in MCNP be controlled?

Yes, the number of secondaries generated in a volume in MCNP can be controlled by adjusting the input parameters, such as the incident particle's energy and the properties of the volume's materials. Additionally, advanced variance reduction techniques can be implemented to optimize the simulation and reduce the number of secondaries generated.

4. How accurate is the number of secondaries generated in a volume calculated in MCNP?

The accuracy of the number of secondaries generated in a volume calculated in MCNP depends on the quality of the input data and the chosen simulation parameters. In general, MCNP is a widely used and validated code, so the results can be considered reliable and accurate if the simulation is set up correctly.

5. Are there any limitations to using MCNP to calculate the number of secondaries generated in a volume?

There are some limitations to using MCNP to calculate the number of secondaries generated in a volume, such as the need for accurate input data and the relatively long simulation times. Additionally, MCNP may not be suitable for simulating certain types of radiation, such as low-energy neutrons or high-energy photons, due to its limitations in modeling certain physical processes.

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