Shielded and unshielded dose from a Co-60 source are the same?

[Joao Pedro]

TL;DR

Ambient dose equivalent from a Co-60 source shielded and unshielded dose are basically the same

I was tasked with designing a shield for a Co-60 source, I decided to use ambient dose equivalent, tally 5. I got my results, followed "Radiation Problems: From Analytical to Monte-Carlo Solutions" example and arrived at a dose of X. I performed some shielding equations and arrived on a required shield of ~9.5 cm of lead to bring the dose down to 1 mSv/year, but when I went to double check my answer performing a MCNP test I am getting basically the same pSv/gamma from the tally 5 output as my unshielded input file, am I doing something wrong here? Was my initial input incorrect? (the input file is a bit disorganized since I was tinkering if it)

 

[Joao Pedro]

I couldn't attach files when making the tread.

 

[Alex A]

Hello @Joao Pedro,

Welcome to PhysicsForums. If you run mcnp with options ip it will plot the geometry. Dotted lines tend to indicate geometry errors.

As a shorthand for reading geometry +surface is usually above or outside, -surface is usually below or inside. You've added two nested boxes 20 and 21, cell 16 then needs to have that volume removed. I think you've tried to do that and called it cell 19, but the definition includes -20 21. Box 20 is inside box 21, so the volume inside 20 but outside 21 doesn't make sense. Likewise for cell 17 being +8 -20. That is the volume inside 20 but outside the whole experiment. I think you mean +5. So I've changed 3 lines, two fixed and one commented out.

c 16  2 -0.000124      5 6 7 -8  IMP:N=0 IMP:P=1 $ See if material for cell 15 and 16 are correct
17  304   -0.128     5 -20     IMP:N=0 IMP:P=1
19  304 -0.000124    -8 7 6 5 21       IMP:N=0 IMP:P=1

See if this solves the problem.