MCNP - Measuring Neutron Absorption in a Moderator

  • Thread starter Qianlong
  • Start date
2
0
Hello all. I'm am a first time poster but a long time visitor. I am having a little trouble that I was hoping someone far wiser and more knowledgeable than myself might be able to help with.

I've been using MCNP to investigate criticality in a simple geometry consisting of a central natural uranium sphere in a cylindrical container (i.e. a barrel). Inside the container and surrounding the sphere is a moderator, either water of graphite. I am attempting to investigate the neutron economy of the moderators. I have been told that the best way to do this is to have MCNP record how many neutrons are absorbed in the moderator. I have also been told that there should be some sort of tally to do this.

However, despite scouring over various primers and manuals until my eyes bled, I have been unable to find any such tally (or at least I have not found any tally I have thought capable of making this measurement). I would appreciate it if anyone could let me know what the code would be? I would also welcome any suggestions of better ways to investigate neutron economy of moderators, should there be one.

Thank you very much for your assistance.
 
2
0
Hi...
one way to do this is to calculate flux (f4) in the cell you are interested:
for example:
f4:n 101 (were 101 is the cell number where you want to calculate flux)
If you would like to know the absorption reaction rates in that cell you then write the following tally multiplier, for example:
f4m:n -1 10 (-2:-6) this way you will calculate total absorption reaction rates (absorption+fission, since absorption (-2) in MCNP is just the capture+fissin (-6)), 10 is the number of the material in that cell and -1 is the multiplier (atom density of thet material)...you can use -1 or enter the atom density of that material...

be careful using tallys, when you calculate flux there must always be 4 the last number...
you can use the following numbers...
f04, f14, f24, f34,...

I hope this will help...

best regards
 
Hello every body
i want rebate structure for cylinder as
110 5 -2.42 (121 -124 -123 ) u=1
111 0 (124 -122 -123 )#110 u=1
112 2 -2.699 (125 -126 -127 ) fill=1
113 like 112 but trcl=(0 5.7 0 )
c extra
121 pz -17.088
122 pz 3.3
123 cz 1
124 pz 2
125 pz -17.508
126 pz 3.8
127 cz 1.5
 

Related Threads for: MCNP - Measuring Neutron Absorption in a Moderator

  • Posted
Replies
3
Views
4K
  • Posted
Replies
2
Views
1K
  • Posted
Replies
22
Views
8K
Replies
1
Views
784
  • Posted
Replies
4
Views
1K
Replies
1
Views
754
Replies
3
Views
3K
Replies
15
Views
2K

Physics Forums Values

We Value Quality
• Topics based on mainstream science
• Proper English grammar and spelling
We Value Civility
• Positive and compassionate attitudes
• Patience while debating
We Value Productivity
• Disciplined to remain on-topic
• Recognition of own weaknesses
• Solo and co-op problem solving
Top