Solving MCNP6 Burn Card Problem

  • Thread starter kslim
  • Start date
  • Tags
    Mcnp6
In summary, MCNP6 is a Monte Carlo N-Particle code used for simulating and analyzing the transport of particles through materials. The "Burn Card Problem" refers to a common issue encountered when using MCNP6, where the code does not accurately account for the depletion of materials over time. There are several methods for solving this problem, including using the ACER module and the depletion capabilities within MCNP6. However, there are limitations to these methods, such as accuracy depending on input data and the difficulty of solving for systems with high burnup or complex geometries. Alternative codes and methods, such as Serpent, SCALE, and ORIGEN, may also be used. To verify the accuracy of the solution, it is important
  • #1
kslim
1
0
Hello Every body,
I hope all is well.
I have a problem with MCNP6 code for using burn card, so please can anybody help me ?
---------------------------------------------------------------------------
particle maximum smallest largest always always
cutoff particle table table use table use model
particle type energy energy maximum maximum below above
1 n neutron 0.0000E+00 1.0000E+36 2.0000E+01 1.5000E+02 0.0000E+00 1.0000E+36

fatal error. Models required. Cannot use memory reduction option.
-----------------------------------------------------------------------------------------


i don't konw why that fatal error occured.

I will give you my simple input.


-------------------------------------------------------------
Example
1 8 -15.973 -1 25 -27 imp:n=1 VOL=21.9982
2 0 #1 imp:n=0
1 cz 0.2837 $ Fuel(F)
25 pz -43.5 $ Lower Reflector(F)
27 pz 43.5 $ Inner Fuel(F), HT9 Follower(C)
BURN TIME = 120
MAT = 8
POWER = 1.0
PFRAC = 1.0
BOPT = 1.0 -24 1.0
m8 92234 3.0959E-05
92235 9.8822E-04
92236 6.0099E-05
92238 6.0904E-01
93237 5.5808E-03
94236 2.7995E-08
94238 2.7918E-03
94239 7.2628E-02
94240 3.3563E-02
94241 9.0784E-03
94242 6.7610E-03
95241 6.5719E-03
95242 1.7994E-04
95243 1.3456E-03
96242 2.2685E-04
96243 7.5349E-06
96244 3.3887E-04
96245 3.0141E-05
96246 2.0876E-06
40090 1.1575E-01
40091 2.5243E-02
40092 3.8584E-02
40094 3.9101E-02
40096 6.2994E-03
42000 2.5799E-02
nlib=66c $ Active Fuel Slug(Middle Core)
kcode 8000 1.0 30 130
ksrc 0 0 0
------------------------------------------------------------------
Regrads
 
Engineering news on Phys.org
  • #2
Make sure the units and magnitudes are correct.

What is 1.0000E+36? That seems like a large number.
 
  • #3
,
Thank you for reaching out for help with your problem using the MCNP6 code for burn card. It seems like you are experiencing a fatal error with the code and are unsure why it is occurring. In order for other users to better assist you, it would be helpful if you could provide more information about the error, such as any error messages or specific lines of code that may be causing the issue.

In the meantime, I took a look at the example input you provided and noticed that there are some missing variables, such as the particle type and energy for the imp:n=0 card. I would suggest double checking all the input parameters to make sure they are correctly defined.

Additionally, you may want to try reaching out to the MCNP6 community for further assistance. There are online forums and user groups dedicated to discussing and troubleshooting issues with the code.

I hope this helps and good luck with your simulations.
 

1. What is MCNP6 and what is the "Burn Card Problem"?

MCNP6 is a Monte Carlo N-Particle code used for simulating and analyzing the transport of particles through materials. The "Burn Card Problem" refers to a common issue encountered when using MCNP6, where the code does not accurately account for the depletion of materials over time.

2. How can I solve the Burn Card Problem in MCNP6?

There are several methods for solving the Burn Card Problem in MCNP6. One approach is to use the ACER module to generate burnup-dependent cross sections, which can then be used in the MCNP6 simulation. Another method is to use the depletion capabilities in MCNP6 to model the depletion of materials over time.

3. Are there any limitations to solving the Burn Card Problem in MCNP6?

Yes, there are limitations to solving the Burn Card Problem in MCNP6. The accuracy of the solution depends on the accuracy of the input data and the chosen method for solving the problem. Additionally, the Burn Card Problem is more difficult to solve for systems with high burnup or complex geometries.

4. Are there any alternative codes or methods for solving the Burn Card Problem?

Yes, there are alternative codes and methods for solving the Burn Card Problem. Some other Monte Carlo codes, such as Serpent and SCALE, have better capabilities for handling depletion calculations. Additionally, there are deterministic codes, such as ORIGEN, that specialize in depletion calculations.

5. How can I verify the accuracy of my solution to the Burn Card Problem in MCNP6?

To verify the accuracy of your solution, you can compare your results to experimental data or to results from other depletion codes. It is also important to carefully review your input data and simulation parameters to ensure they are accurate and appropriate for your system.

Similar threads

  • Nuclear Engineering
Replies
7
Views
507
Replies
1
Views
1K
Replies
58
Views
3K
  • Nuclear Engineering
Replies
4
Views
2K
Replies
7
Views
2K
  • Nuclear Engineering
Replies
3
Views
2K
Replies
3
Views
6K
  • Nuclear Engineering
Replies
9
Views
3K
  • Nuclear Engineering
Replies
0
Views
2K
  • Introductory Physics Homework Help
Replies
3
Views
1K
Back
Top