MCNP6 gamma dose tally problems

[skoch]

Hello, I am a very new MCNP user and have been doing my best to learn on my own. I am struggling to get this problem I am working on to be even somewhat correct. I am trying to determine the dose rate for a person outside a lead-lined room(box) with a 20 TBq Yb-177 source inside. I'm using an F6:P tally with the standard dose function to convert to rem/hr. However, my result is around 3e-14 rem/hr and I know it should be above 2e-3 rem/hr. I am lost and don't know what I'm doing wrong in my input cards that is making this result so inaccurate. Please help.

 

[Alex A]

Hi skoch,

Welcome to physicsforums. Tally outputs are per source particle, so to convert to a real world dose you need to multiply by the activity of the source. 2.7e-14 * 20 e12 = 0.54

I've not checked anything else yet.

 

[skoch]

Hi skoch,

Welcome to physicsforums. Tally outputs are per source particle, so to convert to a real world dose you need to multiply by the activity of the source. 2.7e-14 * 20 e12 = 0.54

I've not checked anything else yet.

That is incredibly helpful, thank you. I have a few other questions to help me understand my input a little better: Does the activity of the source cell as calculated by MCNP affect the tally? Am I correct in using PAR=SD in my SDEF to include the gammas from Yb-177's daughter Lu-177? Is using function 7 (decay) in SP1 with the half life (in shakes) correct for this problem as well?

Most of the calculations I will be using MCNP for are shielding design verifications and I want to better understand MCNP so I am more confident in my reports.

 

[Alex A]

PAR=SD I've never used. MCNP is usually verbose and unclever. This is really enlightened. I don't trust it at all. I think you are using it incorrectly but I can't be certain. The daughter has a longer half life than the parent, so it is not in equilibrium. I also strongly suspect SD will not calculate the activity of the daughter unless you specify how much of it there is in the material card. You have given ERG an exponential function normally used for decay, which is weird. I also suspect the quantity of source material may be wrong, but this is a small thing.

You are also doing a lot of photon simulation when your source and daughter are beta emitters.

And you have a forced collision card!

I would strip all of this back, and start with MODE E, and a single energy beta source at 1.4 MeV or a two bin source with 0.5 MeV as well according to how you are supposed to approach the problem. You can always extend a working input file to make it better.

You are using a ball of flesh (!) to calculate absorbed dose which should give the right answer but the same full cell to work out flux to convert to rem/hr and I think you should be using an empty cell for this.

 

[skoch]

First off, I want to thank you for how quickly you are responding to my thread. I'll quote your response with my reasonings below:

I also suspect the quantity of source material may be wrong, but this is a small thing.

While I was running a significantly shorter version of this problem (2000 nps) to check my inputs, I was printing all of the tables for my output file. One of the tables calculated the activity of the source cell. I kept adjusting the volume of the source cell until the output table showed an activity as close to 20 TBq as I was able to get. I think I ended at 19.x TBq.

You are also doing a lot of photon simulation when your source and daughter are beta emitters.

I am aware that the isotopes are primarily beta emitters, but we are more concerned with the shielding of the gammas that are also produced (I work for a rad safety consultant company).

And you have a forced collision card!

This was out of a little desperation and taken from a slightly similar problem I found online. It can be easily removed.

PAR=SD I've never used. MCNP is usually verbose and unclever. This is really enlightened. I don't trust it at all. I think you are using it incorrectly but I can't be certain.

I picked par=sd because of this particular part of the MCNP manual which is in the list of added features for 6.2:
"The spontaneous-decay source option was extended to provide an all-particle decay option
(PAR=sd on the SDEF card) [8], which automatically generates the correct number of all types of
source decay particles if those particle types are included on the MODE card. Previously, a user
had to specify a distribution of such particle types and adjust the source normalization accordingly."

If I am reading this correctly, and I'm not sure I am, it reads as though MCNP will generate both the beta and gamma particles that come from the decay of Yb-177.

There is also this section: "The PAR=SD, SN, SP, SB, ST, SA, and ZZZAAA (with ERG=0) options require
time integration of daughter production at each level within a decay chain. This is facilitated by
setting all decay constants to unity and uniformly spacing all time bins within 20 s (or ~20 decay
levels), which will include all decay particle production within most decay chains (i.e., an
equilibrium production)."

Unfortunately, I don't know if I need to specify the time integration for the daughters, or if MCNP handles that with "setting all decay constants to unity...."

You are using a ball of flesh (!) to calculate absorbed dose which should give the right answer but the same full cell to work out flux to convert to rem/hr and I think you should be using an empty cell for this.

This was again taken from the same similar problem I mentioned above. It is a masters thesis in which they are determining dose from a neutron source, which although I'm using a primarily beta emitter, does coincide to finding absorbed dose. My F4:P tally I haven't been using the data from because I don't fully understand flux anyways, so if that is not needed it can be dropped as I am only interested in the absorbed dose taken from the F6:P tally.

Thank you again for your help so far. It has been much easier to parse than the MCNP manual and the Primers I have found online.

 

[Alex A]

You are welcome, physicsforums has people with a lot more experience than I do, and they may chip in with things I don't know, they may even suggest a totally different course of action.

This feels like a very artificial problem, and that's fine, a lot of simulation problems are. It's just a bit weird to wrap my head around. You have explained a lot of what you are doing but my thoughts are still to simplify and then add things when you know the right answer. In this case there are four ~1 MeV gammas from Lu-177m* and I would start with a single energy gamma source at 1.09 MeV. If you have the absolute efficiency of this block of lines you can then work out how well the shielding works very easily. I would personally remove the body tissue and take the flux reading from an empty cell converted with the weighting function. This would produce a correct answer and would take less computer time.

Whatever you decide to do, best of luck and I'm here if you have any more questions.