Help With MCNP Volume definition

In summary, the conversation discusses the issue of defining a tally volume bounded by three surfaces in a code and the problem with the volume appearing in red dashed lines. The expert suggests adding a definition for the rest of the universe and modifying the cell definitions to exclude overlapping areas. The issue was resolved by modifying the definition of cell 21 to exclude the cone from the tally volume.
  • #1
khary23
93
6
I need some help defining a tally volume. I want a volume bounded by three surfaces, but when I do an initial plot the volume is in red dashed lines. I know that each cell needs to be uniquely defined, but I am not seeing how my volume is not unique. The cell in question is cell 200 in the code. Any hints would be great.CODE

Dose in Radial Direction for BEBIG Ir2.A85-2
C Cell Cards
11 1 -22.42 -1 -7 9 IMP:p,e=1 $ cylinder that defines source volume
12 3 -8.02 -8 -3 5 2 IMP:p,e=1 $ volume that defines outer radius of casing
13 3 -8.02 -8 -7 -2 1 IMP:p,e=1 $ volume that defines inner radious of casing
14 0 -4 19 10 IMP:p,e=0 $ creates grave yard
16 3 -8.02 -2 -8 7 6 IMP:p,e=1 $ volume that defins end cap
17 4 -4.81 -10 -5 -4 IMP:p,e=1 $ drive cable
C 20 2 -0.00125 -18 3 IMP:p,e=1
21 2 -0.00125 -19 18 10 IMP:p,e=1
C 22 2 -0.00125 19 -4 IMP:p,e=1
22 2 -0.00125 -18 10 #11 #12 #13 #14 #16 #21 IMP:p,e=1
200 2 -0.00125 -20 -19 18 IMP:p,e=1 $ Tally Volume

C Surface Cards
1 RCC 0 0 -0.175 0 0 0.35 0.03 $ cylinder surface that defines source
2 RCC 0 0 -0.246 0 0 0.492 0.035 $ inner radius of casing
3 RCC 0 0 -0.246 0 0 0.492 0.045 $ outer radius of casing
4 SO 40 $ Graveyard @ 40 cm
5 PZ -0.246 $ Bottom of source casing
6 RCC 0 0 0.35 0 0 0.072 0.03 $ end cap cylinder
7 PZ 0.175 $ plane that defines the top of the source
8 PZ 0.246 $ plane that defines the top of the casing
9 PZ -0.175 $ plane that defines the bottom of the source
10 RCC 0 0 -0.246 0 0 -40 0.043 $ drive cable
18 SO 1.9
19 SO 2.1
20 KY 0 0.0012 1 $ creates a conical surfface

C Data Cards
MODE p e $photon and electron transport
C tally card
*F4:p 200
# DE4 DF4 $ convert MeV/cm^2 to MeV/g
0.0010 4065
0.0015 1372
0.0020 615.2
0.0030 191.7
0.0040 81.91
0.0050 41.88
0.0060 24.05
0.0080 9.915
0.0100 4.944
0.0150 1.374
0.0200 0.5503
0.0300 0.1557
0.0400 0.06947
0.0500 0.04223
0.0600 0.03190
0.0800 0.02597
0.1000 0.02546
0.1500 0.02764
0.2000 0.02967
0.3000 0.03192
0.4000 0.03279
0.5000 0.03299
0.6000 0.03284
0.8000 0.03206
1.0000 0.03103
1.2500 0.02965
1.5000 0.02833
C Source is a cylindrical Isotropic Ir-192 photon emitting source centered at origin.
SDEF CEL= 11 VEC=0 0 1 POS=0 0 0 RAD=D1 EXT=D2 ERG=D3 PAR=2
C Radius of circle centered on the axis of the cylinder
SI1= 0.03
C Distance from the POS to the end of the cylinder
SI2= 0.175
C Energy Spectrum of Ir-192 (MeV)
SI3= 0 0.110093 0.13634348 0.17698 0.2013112 0.20579549
0.28004 0.2832668 0.29595827 0.30845692 0.31650791
0.329312 0.3744852 0.4164714 0.420532 0.46807152
0.4845780 0.48530 0.489039 0.5885845 0.59337
0.59935 0.60441464 0.61246564 0.70398 0.7658
0.8845418 1.06148 1.0897 1.3783
C Yield(%) Ir-192
SP3= 0 0.0126 0.183 0.0043 0.472 3.300
0.023 0.262 28.67 30.00 82.81
0.0185 0.721 0.664 0.0737 47.83
3.184 0.0022 0.443 4.515 0.0426
0.0039 8.23 5.309 0.0053 0.00149
0.2923 0.0528 0.00108 0.00124
C Define Materials
C Ir-192
M1 077192 1
M2 6000 -0.00012 7000 -0.75527 8000 -0.23178 18000 -0.01283 GAS=1
C Stainless Steel
M3 025000 -0.02 014000 -0.01 024000 -0.17 028000 -0.12 026000 -0.68
C M3 8000 -0.888 1000 -.112
m4 025000 -0.02 014000 -0.01 024000 -0.17 028000 -0.12 026000 -0.68
PRINT 50
NPS 1000000
 
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  • #2
For MCNP files, the best thing is to do the "Attach files" thing rather than copy-paste. The forum deletes leading blanks so it's a challenge to figure out where it's a continue card and where it's not. Anyway...

It looks like you did not define the "rest of the universe." Try adding this to the end of your cell defs.
C rest of universe
9999 0 4 imp:p,e=0

That will mean that anything that goes outside surface 4 is killed. That's the sphere of radius 40 centered on the origin. If that's not exactly what you want then modify as required. For example, possibly you want some details outside the 40 cm radius sphere. But you must define the entire universe, even the parts you don't want any particles to travel.

By the way, congrats on doing the right thing and plotting. It's very good practice.
 
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  • #3
I created the end of the universe as you suggested, but cell 200 is still in red dotted lines suggesting a geometry error. I have attached a picture of the cell in question.
 

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  • #4
Cell 200 is everything outside radius 1.9 (surface 18), inside radius 2.1 (surface 19), and inside surface 20, a cone.

But cell 21 overlaps with that, since it's all between those spheres except for what is excluded by the cylinder defined by surface 10. So you could exclude the cone from cell 21.

old
21 2 -0.00125 -19 18 10 IMP:p,e=1

new
21 2 -0.00125 -19 18 10 20 IMP:p,e=1
 
  • #5
That did it! Thank you!
 
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FAQ: Help With MCNP Volume definition

What is MCNP Volume definition?

MCNP (Monte Carlo N-Particle) is a computer code used for simulating the transport of particles through matter. The volume definition in MCNP refers to the process of defining the geometry and materials of the system being simulated.

How do I define a volume in MCNP?

To define a volume in MCNP, you must first create a geometry file using a text editor. This file will contain the necessary information such as the shape, dimensions, and material properties of the volume. Once the geometry file is created, it can be imported into MCNP and used for simulation.

What are the different types of volumes that can be defined in MCNP?

MCNP allows for the definition of various types of volumes, including simple geometric shapes (such as spheres, cylinders, and boxes), complex shapes (such as cones and toroids), and CAD-based geometries (such as those created in AutoCAD or SolidWorks).

What are some tips for defining volumes in MCNP?

When defining volumes in MCNP, it is important to ensure that the geometry is accurately represented and that the materials used are correctly assigned. It is also recommended to use symmetry whenever possible to reduce simulation time. Additionally, it is important to carefully review the input file and make any necessary adjustments before running the simulation.

Are there any resources available for help with MCNP volume definition?

Yes, there are various resources available for help with MCNP volume definition. These include user manuals and tutorials provided by the developers of MCNP, as well as online forums and communities where users can ask for assistance and share tips and tricks for using the software.

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