Struggling with MCNP problem

  • Thread starter zincsulphide
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    Mcnp
In summary, the conversation discusses the poster's struggles with writing an input deck for calculating fluence in a zinc sulphide scintillator using MCNP. They are uncertain about their geometry, the cell card for alpha particles, and the ZAID numbers and SDEF card for a volume source. They also mention receiving a response in their inbox, but do not believe it provided new information. The poster eventually decides to delete their files and account.
  • #1
zincsulphide
9
0
I don't have anyone in my university who can help me with MCNP. I'm trying to write an input deck to calculate fluence in a zinc sulphide scintillator. The alpha particles are being emitted from a solution containing plutonium, plutonium nitrate and plutonium nitride.

There are lots of things I don't understand. I am not sure if my geometry is correct. I am not sure what to put in the cell card for alpha particles "imp:?". I am unsure about the ZAID numbers and the SDEF card for a volume source. Any help to get this working would be appreciated :frown:

MCNP file:
https://docdro.id/p1MAXAd

Background information:
https://docdro.id/mFc8LIe
 
Last edited:
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  • #2
Hi,
I think that I I had answered of some of your question in "inbox" but I see that you ask the same question
I will wait the reponse that you will have with this new post
bye
PSR
 
  • #3
Hi,

I don't think you added anything new in the inbox reply?

First I would like to address the cell and surface cards. I think my geometry is correct, but is it confusing?
 
  • #4
Files deleted. Admin please delete my account.

Cheers
 
  • #5
Poster has left the building, so this thread is closed.
 

1. What is MCNP and what is it used for?

MCNP (Monte Carlo N-Particle) is a computer code used for simulating and analyzing the transport of particles through matter. It is commonly used in the fields of nuclear engineering, medical physics, and radiation protection.

2. Why am I having trouble running MCNP?

There could be several reasons for difficulty in running MCNP. Some common issues include incorrect input parameters, missing or incorrect data files, or insufficient computer resources. It is important to carefully review the input file and consult the MCNP user manual for troubleshooting tips.

3. How can I improve the accuracy of my MCNP simulations?

There are several ways to improve the accuracy of MCNP simulations. These include increasing the number of particles simulated, using more detailed and accurate material data, and ensuring that all relevant physical processes are included in the simulation.

4. Can MCNP be used for all types of radiation transport simulations?

No, MCNP is primarily designed for neutron, photon, and electron transport simulations. It may not be suitable for other types of radiation, such as alpha particles or heavy ions.

5. Are there any alternative codes to MCNP for radiation transport simulations?

Yes, there are several alternative codes available for radiation transport simulations, such as Geant4, FLUKA, and PHITS. Each code has its own strengths and limitations, so it is important to carefully evaluate which code is best suited for your specific simulation needs.

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