MCNP6.2 BURN Problem Uranium Dioxide 4.2% Enrichment

In summary: It will just ignore them. So it is important to make sure that the BOPTS and libraries are set up properly.In summary, the conversation is about uranium dioxide with 4.2% enrichment and the use of burn cards to omit certain isotopes in calculations. The expert explains that it is important to make sure that the BOPTS and libraries are set up properly in order for the program to not ignore the specified nuclides and to accurately perform the calculations.
  • #1
mhovi
2
0
TL;DR Summary
Whenever I try to burn more than one material in any cell/assembly. I always get a fatal error saying "Models required. Cannot use memory reduction option.
The following nuclides use physics models rather than data tables:
6014. c
7016. c
8018. c
9018. c "and some other nuclides.

But the same code runs perfectly for burning one material. Can anyone suggest me anything? I've attached a random sample cell problem but the same thing also persist in the assembly calculation code.
Sample cell problem:
uranium dioxide with 4.2% enrichment
c Cell Cards
101   2  -0.0003922  -7            -5  6   imp:n=1 vol= 0.26195 tmp= 1.0109E-7
201   1  -10.55       7 -8         -5  6   imp:n=1 vol= 8.84672 tmp= 1.0109E-7
301   2  -0.001598    8 -9         -5  6   imp:n=1 vol= 0.288775 tmp= 1.0109E-7
401   3  -6.55        9 -10        -5  6   imp:n=1 vol= 3.64607 tmp= 1.0109E-7
501   4  -0.740582    10 -1 2 -3 4 -5  6   imp:n=1 tmp= 1.0109E-7
601   0                   1 -2 3 -4 5 -6   imp:n=0

c Surface Cards
7    cz 0.060
8    cz 0.380
9    cz 0.386
10   cz 0.455
*1   px   0.6375
*2   px  -0.6375
*3   py   0.6375
*4   py  -0.6375
*5   pz   10
*6   pz  -10

c Data Cards
BURN TIME =2.348,2.348 
     MAT=1,3
     POWER=19.6
     PFRAC=1.0,1.0
     BOPT=1.0,-24,1.0
     AFMIN=1E-32
     Omit=1,136,6014,7016,8018,9018,53132,53133,53134,56136,56137,60146,
     61146,62148,64150,64151,64153,90234,91232,31069,31071,32070,32073,
     34074,34077,34078,34079,34080,34082,36085,37086,38084,38087,38088,
     39090,39091,40095,41094,41095,41097,42092,42094,42096,42097,42098,
     44096,44098,44099,44100,44102,44104,44105,44106,46107,47111,49113,
     50113,50114,50115,50116,50117,50118,50119,50122,50123,50124,50125,
     51123,51124,51125,51126,52120,52122,52123,52124,52125,52126,52128,
     53130,53131,54123,54133,56130,56132,56133,56134,56135,56140,57138,
     58136,58138,58139,58140,58141,58142,58143,58144,59142,59143,59145,
     60150,61151,62144,62153,62154,63156,63157,65159,65160,66156,66158,
     66162,66163,66164,68162,68164,68166,68167,68168,68170,32074,32076,
     38089,38090,42099,42100,49115,50112,50126,51121,52130,52132,
     57139,57140,60142,60144
c material card
m1       92235  -0.03701989817
         92238  -0.8444062488
         8016   -0.118573853
m2 2004   -1
m3   50000 -0.014
     26056 -0.00165
     24052 -0.001
     40000 -0.98335
m4   1001   -0.1111
     8016   -0.8889
Kcode   1000 1.0 10 60
Ksrc    0.4 0.0 0.0
 
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  • #2
You need to omit the following isotopes in the burn card too and apply the omitting to all isotopes. if they are not important for your calculation.

All will work.

"23049 23052 23053 24051 25052 25053 25054 25056 25057 26053 26055 27057 28057"

BURN TIME =2.348,2.348
MAT=1,3
POWER=19.6
PFRAC=1.0,1.0
BOPT=1.0,-24,1.0
AFMIN=1E-32
OMIT=-1,149,6014,7016,8018,9018,53132,53133,53134,56136,56137,60146,
61146,62148,64150,64151,64153,90234,91232,31069,31071,32070,32073,
34074,34077,34078,34079,34080,34082,36085,37086,38084,38087,38088,
39090,39091,40095,41094,41095,41097,42092,42094,42096,42097,42098,
44096,44098,44099,44100,44102,44104,44105,44106,46107,47111,49113,
50113,50114,50115,50116,50117,50118,50119,50122,50123,50124,50125,
51123,51124,51125,51126,52120,52122,52123,52124,52125,52126,52128,
53130,53131,54123,54133,56130,56132,56133,56134,56135,56140,57138,
58136,58138,58139,58140,58141,58142,58143,58144,59142,59143,59145,
60150,61151,62144,62153,62154,63156,63157,65159,65160,66156,66158,
66162,66163,66164,68162,68164,68166,68167,68168,68170,32074,32076,
38089,38090,42099,42100,49115,50112,50126,51121,52130,52132,
57139,57140,60142,60144,
23049,23052,23053,24051,25052,25053,25054,25056,25057,26053,26055,
27057,28057
 
  • #3
nukecore said:
You need to omit the following isotopes in the burn card too and apply the omitting to all isotopes. if they are not important for your calculation.

All will work.

"23049 23052 23053 24051 25052 25053 25054 25056 25057 26053 26055 27057 28057"

BURN TIME =2.348,2.348
MAT=1,3
POWER=19.6
PFRAC=1.0,1.0
BOPT=1.0,-24,1.0
AFMIN=1E-32
OMIT=-1,149,6014,7016,8018,9018,53132,53133,53134,56136,56137,60146,
61146,62148,64150,64151,64153,90234,91232,31069,31071,32070,32073,
34074,34077,34078,34079,34080,34082,36085,37086,38084,38087,38088,
39090,39091,40095,41094,41095,41097,42092,42094,42096,42097,42098,
44096,44098,44099,44100,44102,44104,44105,44106,46107,47111,49113,
50113,50114,50115,50116,50117,50118,50119,50122,50123,50124,50125,
51123,51124,51125,51126,52120,52122,52123,52124,52125,52126,52128,
53130,53131,54123,54133,56130,56132,56133,56134,56135,56140,57138,
58136,58138,58139,58140,58141,58142,58143,58144,59142,59143,59145,
60150,61151,62144,62153,62154,63156,63157,65159,65160,66156,66158,
66162,66163,66164,68162,68164,68166,68167,68168,68170,32074,32076,
38089,38090,42099,42100,49115,50112,50126,51121,52130,52132,
57139,57140,60142,60144,
23049,23052,23053,24051,25052,25053,25054,25056,25057,26053,26055,
27057,280
I already tried and this is working. can you explain it to me how could it works? bcs i face similar problem..
 
  • #4
It works with that input? Not in my case
 
  • #5
It will depend what libraries you have, and what BOPTS are in the input file. Potentially you can turn the advanced physics on instead. But you can't tell the program to not ignore nuclides and then not give it data tables or allow it to use models for those nuclides.
 

FAQ: MCNP6.2 BURN Problem Uranium Dioxide 4.2% Enrichment

What is MCNP6.2 BURN?

MCNP6.2 BURN is a computer code used for simulating the transport of particles through materials. It is commonly used in the field of nuclear physics and engineering.

What is a "BURN Problem" in MCNP6.2?

A "BURN Problem" in MCNP6.2 refers to a simulation that includes the depletion of materials due to nuclear reactions. This allows for the prediction of changes in material composition over time.

What is Uranium Dioxide 4.2% Enrichment?

Uranium Dioxide 4.2% Enrichment refers to a type of fuel used in nuclear reactors. It is made up of uranium dioxide with 4.2% of the uranium-235 isotope, which is necessary for sustaining a chain reaction.

How does MCNP6.2 handle the BURN Problem for Uranium Dioxide 4.2% Enrichment?

MCNP6.2 uses a depletion calculation method called "transmutation chains" to simulate the depletion of Uranium Dioxide 4.2% Enrichment. This method takes into account the different nuclear reactions that occur and their effects on the material composition.

What are the applications of using MCNP6.2 BURN for Uranium Dioxide 4.2% Enrichment?

MCNP6.2 BURN can be used to predict the behavior of nuclear reactors using Uranium Dioxide 4.2% Enrichment as fuel. It can also be used to study the effects of different operating conditions and to optimize reactor design for safety and efficiency.

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