MCNPX pwr pin depletion input file running error

[emilmammadzada]

TL;DR Summary

Help with MCNPX Error: "You need a source subroutine" and Other Warnings

Hello everyone,

I'm working on a PWR fuel pin depletion simulation in MCNPX, but I'm encountering several warnings and an error that stops my simulation. Here’s my input setup:
[c *** PWR pincell ***
c
c --- cell cards ---
1 1 -10.4 -1 imp:n=1 vol=192.29 $ fuel
2 2 -6.55 1 -2 imp:n=1 vol=66.53 $ clad
3 3 -1.0 2 -3 imp:n=1 vol=374.27 $ water
4 0 3 imp:n=0 $ outside

c --- surface cards ---
1 RCC 0 0 0 0 0 365.0 0.4095 $ fuel
2 RCC 0 0 -0.0655 0 0 365.131 0.4750 $ clad
3 BOX -0.65665 -0.65665 -0.1 1.3133 0 0 0 1.3133 0 0 0 367.0 $ water box

c --- material cards ---
m1 8016.60c 0.9651 92235.60c 0.0030 92238.60c 0.0319 $ fuel
m2 40000.60c 1.0 $ clad
m3 1001.60c 0.667 8016.60c 0.333 $ water

c --- Tally cards---
F4:n 1

c --- depletion ---
BURN TIME=100,200,300
MAT=1
POWER=0.066956
PFRAC=1.0,0.8,0.5
OMIT=1,2,92235,92238

c --- KCODE cards ---
kcode 1000 1.0 15 50
ksrc 0.65665 0.65665 150.0
]
running error
[ warning. total nu is now the default for fixed-source problems.
warning. there are no tallies in this problem.
warning. cross-section file bertin does not exist.
imcn is done
runtpe already exists. runtpf is created instead.
warning. 92235.60c lacks delayed neutron cross sections.
warning. 92238.60c lacks delayed neutron cross sections.
dump 1 on file runtpf nps = 0 coll = 0
ctm = 0.00 nrn = 0

xact is done
dynamic storage = 0 words, 0 bytes. cp0 = 0.01

bad trouble in mcrun in routine source
you need a source subroutine.]

 

[Alex A]

Hi @emilmammadzada,

You should check the location of your data files. I don't know if bertin would be used in a burn problem, if in doubt find the file and drop it into the same directory X is in.

You have extra empty lines in your input file, you should have two, one after the cell definitions and one after the surface definitions. Some of those errors are just that the program isn't reading beyond the third empty line. Other than that it looks good.

If you fix those and still get errors, change the input file to add .txt to the end of the name and then attach it to a post.

 

[emilmammadzada]

Hi @emilmammadzada,

You should check the location of your data files. I don't know if bertin would be used in a burn problem, if in doubt find the file and drop it into the same directory X is in.

You have extra empty lines in your input file, you should have two, one after the cell definitions and one after the surface definitions. Some of those errors are just that the program isn't reading beyond the third empty line. Other than that it looks good.

If you fix those and still get errors, change the input file to add .txt to the end of the name and then attach it to a post.

Is the xsdir file you specified as x?

 

[emilmammadzada]

Hi @emilmammadzada,

You should check the location of your data files. I don't know if bertin would be used in a burn problem, if in doubt find the file and drop it into the same directory X is in.

You have extra empty lines in your input file, you should have two, one after the cell definitions and one after the surface definitions. Some of those errors are just that the program isn't reading beyond the third empty line. Other than that it looks good.

If you fix those and still get errors, change the input file to add .txt to the end of the name and then attach it to a post.

now i am getting this error.
warning. total nu is now the default for fixed-source problems.
warning. there are no tallies in this problem.
warning. cross-section file bertin does not exist.
imcn is done
warning. 92235.60c lacks delayed neutron cross sections.
warning. 92238.60c lacks delayed neutron cross sections.
dump 1 on file runtpe nps = 0 coll = 0
ctm = 0.00 nrn = 0

xact is done
dynamic storage = 0 words, 0 bytes. cp0 = 0.01

bad trouble in mcrun in routine source
you need a source subroutine.
my input file is this

 

[Alex A]

Sorry, x meaning wherever mcnpx or mcnpx.exe is but this is often where xsdir is anyway. Try it in the data directory, MCNP_DATA or whatever directory is pointed to by the top line in your xsdir file (if it is in it), or if everything is in the same directory just drop it into that. So long as the file is not damaged it will work somewhere.

 

[emilmammadzada]

Sorry, x meaning wherever mcnpx or mcnpx.exe is but this is often where xsdir is anyway. Try it in the data directory, MCNP_DATA or whatever directory is pointed to by the top line in your xsdir file (if it is in it), or if everything is in the same directory just drop it into that. So long as the file is not damaged it will work somewhere.

I did what you said but I still got this error.
warning. total nu is now the default for fixed-source problems.
warning. there are no tallies in this problem.
warning. cross-section file bertin does not exist.
imcn is done
warning. 92235.60c lacks delayed neutron cross sections.
warning. 92238.60c lacks delayed neutron cross sections.
dump 1 on file runtpe nps = 0 coll = 0
ctm = 0.00 nrn = 0

xact is done
dynamic storage = 0 words, 0 bytes. cp0 = 0.01

bad trouble in mcrun in routine source
you need a source subroutine.

 

[Alex A]

If you still get "you need a source subroutine" you have not removed a blank line. You should only have two in the file. There is more to this than I thought. Your ksrc isn't in your fuel, I've put it in the fuel but not in the middle so you should change this. Your k for the system - a single rod - is very very low. I've turned it into an infinite lattice by making the water box reflecting just for this test.

The BURN card is in the wrong place, the material cards must be below it and it should be

BURN TIME=100,200,300
     MAT=1
     POWER=0.066956
     PFRAC=1.0,0.8,0.5
     OMIT=blah

Because these are variables on the BURN card.

The result then runs into what I hope will be the last problem - missing data tables for fleeting nuclei. So add what it needs and make an OMIT line that works for you. The one with U235 and U238 in it was a strange choice.

See how far you can get!

 

[emilmammadzada]

If you still get "you need a source subroutine" you have not removed a blank line. You should only have two in the file. There is more to this than I thought. Your ksrc isn't in your fuel, I've put it in the fuel but not in the middle so you should change this. Your k for the system - a single rod - is very very low. I've turned it into an infinite lattice by making the water box reflecting just for this test.

The BURN card is in the wrong place, the material cards must be below it and it should be

BURN TIME=100,200,300
     MAT=1
     POWER=0.066956
     PFRAC=1.0,0.8,0.5
     OMIT=blah

Because these are variables on the BURN card.

The result then runs into what I hope will be the last problem - missing data tables for fleeting nuclei. So add what it needs and make an OMIT line that works for you. The one with U235 and U238 in it was a strange choice.

See how far you can get!

Thank you so much for taking the time to answer my question! Your insights were really helpful.