MCNP4c2: Fission Reactions in a Spherical Subcritical Reactor

In summary, you should be able to use MCNP4 to calculate the fission reactions per second in a geometry of a spherical sub critical reactor of Uranium with low percentage of U 235 with external neutron source.
  • #1
Dimitris Catzis
22
0
TL;DR Summary
I want some help to make a simple simulation of a sub critical reactor with external source
Hi, i am new to simulation and for my thesis i have to make a simple simulation by using mcnp4c2. Is anybody familiar with this version of MCNP?

I need to calculate the fission reactions per second in a geometry of a spherical sub critical reactor of Uranium with low percentage of U 235 with external neutron source.
Thanks a lot.
 
Engineering news on Phys.org
  • #2
Did you have specific questions?

MCNP version 4 is pretty old, but usable. Do you have the full install? Do you have the documents?

What do you want to get out of your simulation?
 
  • #3
Hi,
thanks for your response.

I think i have the full installation( about 1.4 GB).

I want to calculate the energy released from fission reactions from a spherical geometry of Uranium with the neutron source at the center and Beryllium as shielding, I have the spectrum of the neutron source. My thesis is about sub critical reactor for space applications. Is this version capable?
 
  • #4
You should be able to do that with MCNP4.

You need to find the documents directory in your install. It should include a user manual. Also there should be a theory manual. If you've got source code there should be a developer's manual. IIRC, back in MCNP4 days they routinely included the source code so you might have it. Read up on how to set up the cells with the materials. Read up about tallies of type F6 (heat deposition) and F7 (fission energy) particularly. Then you will need to know the number of neutrons per second the source releases and use that to normalize everything. The tallies report their results in "per particle started." So you convert to "per second" by using the neutrons per second the source releases.

Also, carefully read about the SDEF card. This is the source definition card. It allows you to specify the location and energy of source particles.

Other possible things you might be interested in are F4 tallies. These give you the particle flux. You could also investigate things like the total number of neutrons generated from any given source neutron. This gives you an estimate of how close to criticality your reactor is. (Hmm... The spell checker on Physics Forums does not know criticality. Hmm...) You could compare that to what you get from a KCODE calculation, which you should also read up on.

You may want to play around with your geometry. For example, just because the material in a sphere is all the same you don't automatically want just a single sphere. You might want some spherical shells to allow you to refine your tallies. Maybe you want to figure out how much heat gets deposited in each layer.

Probably as a first pass through you want to use neutrons only. Once you get your geometry and materials correct, and you are happy with your SDEF and tallies, then you might want to add photons. This means you will need to read up on the MODE card, and possibly about the PHYS card. Some things you will be limited by the available cross section libraries.
 

Related to MCNP4c2: Fission Reactions in a Spherical Subcritical Reactor

1. What is MCNP4c2?

MCNP4c2 is a computer code used for simulating and analyzing neutron and photon transport in complex systems. It is commonly used in nuclear engineering and radiation physics research.

2. How does MCNP4c2 model fission reactions in a spherical subcritical reactor?

MCNP4c2 uses Monte Carlo methods to simulate the behavior of individual neutrons and photons in a system. It takes into account the physical properties of the materials in the reactor, such as their density and composition, and calculates the probability of interactions and fission events occurring.

3. What is a subcritical reactor?

A subcritical reactor is a nuclear reactor that is not self-sustaining and requires an external source of neutrons to maintain a chain reaction. This is in contrast to a critical reactor, which is self-sustaining and does not require an external source of neutrons.

4. What are fission reactions?

Fission reactions are nuclear reactions in which a heavy nucleus splits into two or more smaller nuclei, releasing a large amount of energy. These reactions are the basis for nuclear power and weapons.

5. How is MCNP4c2 used in nuclear engineering research?

MCNP4c2 is used to simulate and analyze the behavior of neutrons and photons in nuclear systems, such as reactors, fuel cycles, and radiation shielding. It can help researchers understand the effects of different materials and geometries on the behavior of these particles and inform the design and operation of nuclear systems.

Similar threads

Replies
3
Views
2K
Replies
16
Views
3K
  • Nuclear Engineering
Replies
9
Views
2K
  • Nuclear Engineering
Replies
15
Views
2K
  • Nuclear Engineering
Replies
2
Views
1K
Replies
11
Views
2K
  • Nuclear Engineering
2
Replies
46
Views
12K
  • Nuclear Engineering
Replies
9
Views
4K
  • Nuclear Engineering
Replies
1
Views
943
  • Nuclear Engineering
Replies
13
Views
8K
Back
Top