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Photon detectors

Posted: Fri Jun 26, 2015 1:31 pm
by Jaakko Leppänen
Detectors in photon transport simulation work similar to neutrons, only the response functions are different. The microscopic reaction MT's for photon reactions are:

MT 504 : Compton scattering
MT 502 : Rayleigh scattering
MT 522 : Photoelectric effect
MT 516 : Pair production
MT 301 : Average heating number

The macroscopic total cross section is MT -25 and macroscopic heat production -26. In addition there are built-in mass-energy absorption coefficients for dose-rate calculations obtained from:

Hubbell, J. H. and Seltzer, S.M. (2004), "Tables of X-Ray Mass Attenuation Coefficients and Mass Energy-Absorption Coefficients." (version 1.4).
http://www.nist.gov/pml/data/xraycoef/

The MT's for different materials are:

MT -201 : A-150 Tissue-Equivalent Plastic
MT -202 : Adipose Tissue (ICRU-44)
MT -203 : Air, Dry (Near Sea Level)
MT -204 : Alanine
MT -205 : B-100 Bone-Equivalent Plastic
MT -206 : Bakelite
MT -207 : Blood, Whole (ICRU-44)
MT -208 : Bone, Cortical (ICRU-44)
MT -209 : Brain, Grey/White Matter (ICRU-44)
MT -210 : Breast Tissue (ICRU-44)
MT -211 : C-552 Air-equivalent Plastic
MT -212 : Calcium Sulfate
MT -213 : 15 mmol/l Ceric Ammonium Sulfate Solution
MT -214 : Cesium Iodide
MT -215 : Concrete, Barite (Type BA)
MT -216 : Concrete, Ordinary
MT -217 : Eye Lens (ICRU-44)
MT -218 : Calcium Fluoride
MT -219 : Ferrous Sulfate (Standard Fricke)
MT -220 : Gadolinium Oxysulfide
MT -221 : Gafchromic Sensor
MT -222 : Gallium Arsenide
MT -223 : Glass, Lead
MT -224 : Photographic Emulsion (Kodak Type AA)
MT -225 : Lithium Fluride
MT -226 : Lithium Tetraborate
MT -227 : Lung Tissue (ICRU-44)
MT -228 : Magnesium Tetroborate
MT -229 : Mercuric Iodide
MT -230 : Muscle, Skeletal
MT -231 : Polyethylene Terephthalate (Mylar)
MT -232 : Radiochromic Dye Film (Nylon Base)
MT -233 : Ovary (ICRU-44)
MT -234 : Photographic Emulsion (Standard Nuclear)
MT -235 : Polymethyl Methacrylate
MT -236 : Polyethylene
MT -237 : Polystyrene
MT -238 : Polyvinyl Chloride
MT -239 : Glass, Borosilicate (Pyrex)
MT -240 : Polytetrafluoroethylene (Teflon)
MT -241 : Cadmium Telluride
MT -242 : Tissue-Equivalent Gas (Methane Based)
MT -243 : Tissue-Equivalent Gas (Propane Based)
MT -244 : Testis (ICRU-44)
MT -245 : Tissue, Soft (ICRU Four-Component)
MT -246 : Tissue, Soft (ICRU-44)
MT -247 : Plastic Scintillator (Vinyltoluene)
MT -248 : Water, Liquid

For example, the response function to calculate photon dose rate in air is:

Code: Select all

dr -203 void
In addition, MT -200 calculates the coefficients based on the material at the collision site.

Note that similar to neutron transport calculation, the material entry in the response function (in this case "void") should not be confused with the detector material entry "dm", which defines the material over which the detector is scored.

The results for these dose rates should be in units of Gy*cm3/h, so division by detector volume should yield the dose rate in standard units, provided that the source is appropriately normalized. Note, however, that all this methodology is still new and not very well tested, so let us know if you disagree with the results.

Re: Photon detectors

Posted: Fri Apr 08, 2016 11:37 am
by Andy_Turner
Can you confirm the units that come out of photon heating using dr -26 <mat> ?

Having performed simple 'thin cell' comparisons with MCNP F4 FM -4 1 / -5 -6 and F6 tallies, I get agreement with the '-200 void' response function which is indeed Gy-cm3/hr per source particle.
For neutrons using dr -4 <mat>, I get excellent agreement with MCNP heating data if I assume the detector result is in units of J in cell per source particle. i.e. Gy/h per source particle = result x 1000 g/kg x 3600 s/hr / [vol of detector cm3 x density of actual or virtual material g/cm3].

But for photons (dr -26), the result is 11 orders of magnitude larger than it was for neutrons (dr -4) before I apply any factors. I can only assume the units are different. Can you confirm what they are?

Re: Photon detectors

Posted: Mon Apr 11, 2016 1:48 pm
by Jaakko Leppänen
This problem was traced back to the processing of photon cross sections (heating numbers were not multiplied by total xs), and it will be fixed in the next update.

Re: Photon detectors

Posted: Thu Jan 26, 2017 4:41 pm
by Peter Wolniewicz
Is there a way to get a pulse tally just like F8 (I think it is) in MCNP?

Lets say I put some radioactive material in my simulation and construct a germanium detector in which I tally "pulses" :)

best regards,
Peter

Re: Photon detectors

Posted: Mon Jan 30, 2017 11:06 am
by Jaakko Leppänen
Response function -27 should be equivalent with the pulse-height tally in MCNP. For the complete list of detector responses, see:

http://serpent.vtt.fi/mediawiki/index.p ... on_numbers

Re: Photon detectors

Posted: Thu Dec 19, 2019 8:09 pm
by mmcd
Question regarding photon detectors (actually more generally just about detectors). Apologies if this is answered elsewhere, still new and having some confusion.

I am looking to calculate the gamma dose in Gy/hr at a point some distance from a source and would like to know the proper way to define my detector to do this.

As an example, is it correct to create a volumetric region at the specified distance from the source (say Cell 10) such as:
det 1 dc 10 dr -200 void

Then, do I still need to use dv to specify the volume of this region, or is the volume of cell 200 automatically used? Further in this example, would the resultant dose be in Gy/hr as required?

Thanks!

Re: Photon detectors

Posted: Wed Jan 01, 2020 11:42 pm
by Jaakko Leppänen
Using a cell detector (dc option in the cell card) is one way to do it. Serpent doesn't divide the result with volume, so you have to do this manually, or by setting the dv option.

If you are looking at dose rates in discrete locations, I recommend you use the super-imposed track-length detector instead. This will give you much better statistics if the region is small.