# MCNP source definition question

• llatosz
In summary, the conversation discusses the difficulties in creating a point source of Californium-252 that can emit different particles with their own energy spectra. One solution suggested is to run two separate calculations for neutrons and gammas and then add them together, while another suggests using a Watt Spectrum for Cf252.
llatosz
I'm modelling a scenario for the research that I'm working on, and I got the cells and surfaces all mapped out for the environment finally, but now I'm totally stuck on creating a source.
I'd like a point source of Californium-252, but after hours of looking, I don't see any out of the 1000 examples in the manual that covers SDEF for a source that can emit different particles!

I have it in MODE N P, but I don't see how you would go about defining the source to spit out 2 different particle types, each with their own energy spectrum.

Anybody know how to make a Ca-252 source in MCNP? You would be a true savior! :)

For my part for the neutron Spectrum I use a Watt Spectrum for Cf252 :

"mode n p
sdef par=n pos 0 0 0 erg d1
sp1 -3 1.025 2.926"

I use a "mode p n" to transport neutrons and to generate gamma capture (n,g reactions) and to transport them.
The result is for one neutron emitted.
I never use the possibility to emitte two different particles because the normalisation of the result is difficult

SinanK and llatosz
One idea is to run two calculations - one with a neutron source, and one with a gamma source - then add them together.
You will have to weight the two calculations depending on what the initial sources are.

For example, if you had a source that emitted X neutrons and Y gammas, then run gamma and neutron calculation separately.
The final result would be X times the neutron answer plus Y times the gamma answer.

llatosz
PSRB191921 said:
For my part for the neutron Spectrum I use a Watt Spectrum for Cf252 :

"mode n p
sdef par=n pos 0 0 0 erg d1
sp1 -3 1.025 2.926"

I use a "mode p n" to transport neutrons and to generate gamma capture (n,g reactions) and to transport them.
The result is for one neutron emitted.
I never use the possibility to emitte two different particles because the normalisation of the result is difficult

This worked very well and was actually what my coworker suggested too! Thank you very much

## What is MCNP and how is it used in source definition?

MCNP (Monte Carlo N-Particle) is a computer code used for simulating and analyzing the transport of particles through different materials. In source definition, MCNP is used to specify the type, energy, direction, and location of particles in a simulation.

## What is a source definition in MCNP and why is it important?

A source definition in MCNP is a set of parameters that describe the characteristics of particles entering the simulation. This includes the type of particles, their energy, direction, and location. It is important because it determines the initial conditions of the simulation and can greatly affect the results.

## What are some common types of sources used in MCNP?

Some common types of sources used in MCNP include point sources, isotropic sources, and directional sources. Point sources emit particles from a single location, isotropic sources emit particles equally in all directions, and directional sources emit particles in a specific direction.

## What is a source biasing in MCNP and how does it work?

Source biasing in MCNP is a technique used to increase the efficiency of simulations by focusing on specific regions of interest. It works by assigning a higher weight to particles emitted from certain source regions, so they have a greater chance of being transported through the desired area.

## How can I define a custom source in MCNP?

To define a custom source in MCNP, you can use the SDEF card which allows you to specify the type, energy, direction, and location of particles in a source. You can also use the SPHINX tool to create a customized source distribution based on user-defined parameters.

• Nuclear Engineering
Replies
2
Views
2K
• Nuclear Engineering
Replies
2
Views
2K
• Nuclear Engineering
Replies
6
Views
2K
• Nuclear Engineering
Replies
6
Views
1K
• Nuclear Engineering
Replies
4
Views
2K
• Nuclear Engineering
Replies
2
Views
6K
• Nuclear Engineering
Replies
1
Views
3K
• Nuclear Engineering
Replies
2
Views
2K
• Nuclear Engineering
Replies
7
Views
2K
• Nuclear Engineering
Replies
4
Views
2K