[ RadSafe ] MCNP problem

[Alessandro Facure]

    Hi All,

    I have a question concerning mesh tally plotting using MCNP5. As posted here before by Carol, I got three  files (out, meshtal, and outr). I want to plot colour maps of neutron fluxes inside the room (in the same way as presented in  http://mcnpx.lanl.gov/opendocs/misc/MeshTal.pdf). 


    I attached my input here and was wondering whether someone can take a look and give me some suggestions. 



Alessandro Facure
Radiactive Facilities Coordination
Brazilian Nuclear Energy Commission
  

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c Created on: Tuesday, November 27, 2007 at 13:19

1 0 -1

2 1 -1.22e-3 1 -2 #(-1 )

3 2 -2.3 2 -3 #(1 -2 )

4 1 -1.22e-3 -4 3 #(2 -3 )

5 0 4 #(-4 3 )

1 s 221.25 -45 -55 1 

2 rpp -418.75 418.75 -221.35 221.35 -165 165

3 rpp -538.75 538.75 -341.35 341.35 -265 285 

4 so 800 

c air 

m1 6012. -0.000124 $Air

7014. -0.755267 8016. -0.231781 18000. -0.012827 

nlib = 42c

c concrete 

m2 1001.50c -0.01 $MAT

8016.50c -0.529 6012.50c -0.001 11023.51c -0.016 

12000.51c -0.002 13027.50c -0.034 14000.51c -0.337 

19000.51c -0.013 20000.51c -0.044 26000.55c -0.014 

imp:n 1 2 12 2 1 $ 1, 5

sdef erg=d1 pos 221.25 -45 -55 

c 16 MeV Spectrum 

# si1 sp1 

0 0 

0.05 698 

1 790 

2.0 662 

3.0 573 

4.0 433 

5.0 356 

6.0 241 

7.0 189 

8.0 124 

9.0 95.1 

10.0 61.9 

11.0 27.9 

12.0 9.64 

13.0 4.04 

16.0 0.0 

c Mesh tally for monitoring particle generation in XZ plane 

fmesh4:n origin -800 -800 -800 

imesh=800 iints=25 

jmesh=800 jints=25 

kmesh=800 kints=1 

c Ignore cross section data above 20 MeV 

phys:n 20 0 0 

c Energy cutoff, upper limit at 20 MeV 

cut:n 20 0.0 $ detailed physics treatment 

nps 100000 $ number of particles 

print $ print everything 

mplot freq 1000 fmesh 4