Hello;
I am using MCNP5 to calculate the dose for some luminescent radiation detector such as RPL , TLD-100 or OSL in radiotherapy (Mega-voltage photons and electrons beams) use F6 tally, so it becomes dose/(particle source)
I found a difficult to calculate the total dose and dose rate because i haven't the real total number particle source (N, D(total) = D(per p-s) * N)
Any help would be much appreciated
It is not necessary to have the actual activity or NPS if you know all other relevant factors, in particular, the geometry, materials, beam characteristics, and voltage or photon energy. Once you have solved a nominal case it is a problem of scale. Nevertheless, it is surprising to have all other information without the activity or NPS.
Are you doing an evaluation of dosimeters for generic radiotherapy systems? If you lack relevant factors, an approach would be to develop generic models of the system as near the actual as possible. This would give you a range of nominal cases, but at some point you will need much more information.
Using MCNP to simulate a RT device and dosimeter responses, one can get several tallies calculated at the same run. In your case, it might be more practical to calculate both dose response of your detector (expressed better via analogue *f8 tally rather than fluence-based f6 one, because the latter one implies electron equilibrium and leads to dose overestimation for high-energy photons) as well as fluence/energy fluence/air kerma at the dosimeter location.
In other words, using Monte Carlo simulation of photon transport, it is easier to get relationship between various dosimetric responses for the same geometry or setup. Getting absolute calibration seems to be more complicated as it would require to solve other problems related to conversion of absolute current in an accelerator to energy-dependent photon fluence in an exit window.
MCNP code normalize the results per particle. To calculate the dose, it needs to have the radiation source activity.
Hi..
I want to know how can I use Tally f4 and f6 for calculating dose?..
Sahar Ammari
F4 and *F4 give the average flux in a cell with a unit #/cm2 and MeV/cm2 respectively.
F6 and *F6 give the energy deposition in a cell with a unit of MeV/g and jerks/g respectively.
For more detail see this resume of MCNP manual
https://www.mne.k-state.edu/~jks/MCNPprmr.pdf
You must insert the tally f4 and f6 in Data card section.
F4 tally can be used for dose calculation. For this, By using the DE and DF cards, you can define the flux-to-dose conversion function. You can find values for that in the MCNP manual or in websites and different books.
For e.g.:
C ICRP 21 FLUX TO DOSE CONVERSION FACTORS
DE4 2.5E-8 1.0E-7 1.0E-6 1.0E-5 1.0E-4 1.0E-3 1.0E-2 0.1 0.5 1.0
2 5 10 20
DF4 3.85E-6 4.17E-6 4.55E-6 4.35E-6 4.17E-6 3.70E-6 3.57E-6
2.08E-5 7.14E-5 1.18E-4 1.43E-4 1.47E-4 1.47E-4 1.54E-4
The f6 tally
results are given in units of MeV/g. It can be modified by a
constant multiplier (1 MeV/g = 1.602×10-8 rad) to get dose
values in units of rad-in-air and then converted to Sv/hr
(1rad = 0.01Sv for gamma radiation).
How I can define the geometry for my tally if I want to calculate an effective dose to a worker standing 1m from the shield?
@Snezana Petrovic
With MCNP you can get absorbed dose (MeV / g) or deposited energy (MeV) with F6 or * F8, respectively, then you can convert it to Gy or Sievert (effective dose) at any distance from the shield.
But what do you mean by " geometry for my tally "?
Abdelhai Ben Ali thank you for your answer. I wasn't sure should I use fantom or I can put a simple slab 1m from my shield as my tally?
At the moment I am using FMESH4 tally to calculate the dose outside of my laboratory.
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