MCNP5 tallies conversion and MCNPX

  • Thread starter Thread starter nuclearsneke
  • Start date Start date
  • Tags Tags
    Mcnp Mcnp5 Mcnpx
AI Thread Summary
To convert the flux (F4:N tally) from MCNP units to cm-2s-1, scale the tally by the volume of the detection cell, ensuring it is set to unit volume. Using a detection shape like a coin with a radius of 1 cm and a thickness of 0.318 can help achieve this. Regarding the MCNPX error related to "bertin," ensure that the required library file is installed, as it is necessary for proper functioning. MCNP can calculate volumes for certain shapes, but this feature may not always work as expected. Proper setup and scaling are crucial for accurate results in MCNP simulations.
nuclearsneke
Messages
15
Reaction score
5
TL;DR Summary
tl;dr how to convert flux gotten via F4:N tally to cm-2s-1 units?
Hi there

I want to convert the flux (F4:N tally) from mcnp units to cm-2s-1 units. How to do that?

Also I have some bug in MCNPX: while running the file, I get an error like "
>bad trouble in imcn in routine xin
>Cannot find bertin
"
How to solve it? Database for MCNP5-MCNPX got installed already.
 
Engineering news on Phys.org
Howdy! In general MCNP F4:N tallies need scaling by the volume of the detection cell. They are track length tallies but the units end up the same and the flux ends up right when scaled. I've found if I use a detection 'coin shape', make the radius whatever I need, say 1cm, I can make the thickness 0.318 to give the detection cell unit volume and then the tally is automatically right. I've read that MCNP is capable of working out the volume for some shapes, but I've not seen this work.

You should have a library file called "bertin", it's one of the extra files X needs.
 

Thread 'MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly'

Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information（power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.） for Material 2 is zero, while Material 1...
View full post »

Thread 'Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)'

Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6? If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning. I'd be really...
View full post »

Similar threads

Mesh Tally in MCNPX for Dose Calculation / How to Add & Visualize?
Replies
0
Views
256
Could anyone help me with my problem, my MCNPX simulation is not working
Replies
2
Views
2K
MCNPX pwr pin depletion input file running error
Replies
15
Views
2K
MCNP6: error -- "bad trouble in imcn" is usually tangled with spacing error?
Replies
6
Views
4K
I get this error when I try to run the code MCNPX
Replies
5
Views
3K
How to convert MCNP-generated data
Replies
4
Views
980
How to finally run an input file via VISED/MCNP5 without error?
Replies
1
Views
3K
Geometry error in newcel
Replies
8
Views
1K
An extremely basic question on MCNP
Replies
2
Views
3K
Why am I getting a fatal error in MCNP with xsdir missing cross-section tables?
Replies
1
Views
15K

Hot Threads

Recent Insights

Back
Top