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### Gamma detectors and flux to dose

Posted: Thu Mar 31, 2016 2:11 pm
I'm a little confused about detectors, perhaps someone can help.

From here:
http://ttuki.vtt.fi/serpent/viewtopic.p ... f4f3#p5884

I infer that everything is done via reaction rate multipliers (MT-201-248).
So I can see one could get absorbed dose / heating to specified material.
and MT -200 is for absorbed dose in actual material.

The C-lite fusion example for decay gamma shutdown dose rate uses a mesh tally with 'dr -200 void.'
What does that equate to? Absorbed dose (Gy/cm3/h) in 'actual material' for all regions that are void? (but in void there are no collisions?).

Is there a way to get Sv/h from a detector?
Is it possible to specify flux to dose factors in a detector? e.g. if one was required to use ICRP-74 flux to dose factors, could one do this? e.g. like the DE-DF cards in MCNP. If not I suppose tallying flux in bins, and doing some post processing with bin-averaged flux to dose factors would work but not very elegant.

### Re: Gamma detectors and flux to dose

Posted: Thu Mar 31, 2016 9:09 pm
The void entry in the detector response means that the material is taken from the collision point. For response function -200 this means that if the collision occurs in steel, Serpent looks for the composition of steel and calculates the mass attenuation coefficient accordingly based on the elemental data. The coefficients are used as they are provided at the linked website.

I'm not very familiar with the procedure of converting Sv to Gy, but I believe it requires some additional data. Serpent allows using user-defined response functions with reaction identifier -100:

Code: Select all

``````dr -100 <int> <np> <E1> <f1> <E2> <f2> ...
``````
where <int> is the interpolation type used between the points (1 = histogram, 2 = lin-lin, 3 = lin-log, 4 = log-lin, 5 = log-log), <np> is the number of points and <En> <fn> are the energy-value pairs.

### Re: Gamma detectors and flux to dose

Posted: Fri Apr 01, 2016 10:13 am
Jaakko Leppänen wrote:Serpent allows using user-defined response functions with reaction identifier -100:

Code: Select all

``````dr -100 <int> <np> <E1> <f1> <E2> <f2> ...
``````
where <int> is the interpolation type used between the points (1 = histogram, 2 = lin-lin, 3 = lin-log, 4 = log-lin, 5 = log-log), <np> is the number of points and <En> <fn> are the energy-value pairs.
Perfect yes that's what I meant. User-defined response function with interpolation is exactly what MCNP does with DE-DF cards.
Is it possible to define the response function as a name and then use 'dr my_dose_response' ? (to re-use this on multiple detectors)

### Re: Gamma detectors and flux to dose

Posted: Fri Apr 01, 2016 11:25 am
That would actually be a better option. In the current version you'll have to copy-paste the response to all detectors.

### Re: Gamma detectors and flux to dose

Posted: Mon Jan 30, 2017 12:58 pm
Using
dr -203 (photon dose in air, Dry (Near Sea Level))

on a detector volume,

What is the unit in the detector output and what is it normalized to?

### Re: Gamma detectors and flux to dose

Posted: Mon Jan 30, 2017 2:07 pm
See the reference at:

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

The units are in Gy, normalization depends on how the source is normalized.

### Re: Gamma detectors and flux to dose

Posted: Mon Jan 30, 2017 2:51 pm
Jaakko Leppänen wrote:See the reference at:

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

The units are in Gy, normalization depends on how the source is normalized.
Okay thanks..

A.)
If I have a body composed of different radioactive nuclides and multiple cells, and
where my source sampling points are covered by:

src 1 g sg -1 1 sx -0.525 0.525 sy -0.525 0.525 sz -10 88.1184
set nps 10000000
set rfr -1000 petermat.dat %material from burn-up calculation

Which covers a larger volume than the radioactive body.
I then put a detector volume (air) next to the body
det 1d dc 200 dr -203 air2 %photon dose in air, Dry (Near Sea Level)

I dont have any other options in my input file for normalization. How do I get Gy/hour? I see in the output file that my materials have photon emission rates.

Example:
Material "insulator":

- Material is included in majorant
- Material is included in geometry
- Atom density 1.1663E-01 1/barn*cm
- Mass density 3.9500E+00 g/cm3
- Volume 3.2041E-01 cm3
- Mass 1.2656E+00 g
- Photon emission rate 1.6319E+09 1/s
- 16 nuclides in composition
- No nuclides associated with S(a,b) data
- 344.30 kb of memory allocated for data

Isotopic composition (non-zero densities):

-------------------------------------------------------------------
Nuclide a. weight temp a. dens a. frac m. frac
-------------------------------------------------------------------
1000.84p 1.00798 0.0 8.36315E-09 7.17092E-08 3.54388E-09
2000.84p 4.00260 0.0 3.56230E-09 3.05447E-08 5.99421E-09
5000.84p 10.81103 0.0 4.94021E-19 4.23594E-18 2.24528E-18
6000.84p 12.01104 0.0 3.56220E-09 3.05438E-08 1.79870E-08
7000.84p 14.00672 0.0 9.57285E-14 8.20817E-13 5.63685E-13
8000.84p 15.99930 0.0 6.99092E-02 5.99431E-01 4.70213E-01
10000.84p 20.18005 0.0 1.12574E-21 9.65256E-21 9.55033E-21
11000.84p 22.98977 0.0 1.12961E-23 9.68572E-23 1.09174E-22
12000.84p 24.30505 0.0 1.62303E-13 1.39165E-12 1.65837E-12
13000.84p 26.98154 0.0 4.67160E-02 4.00563E-01 5.29897E-01
14000.84p 28.08551 0.0 6.07075E-07 5.20532E-06 7.16777E-06
15000.84p 30.97376 0.0 9.50789E-24 8.15247E-23 1.23805E-22
16000.84p 32.06439 0.0 1.98423E-29 1.70137E-28 2.67470E-28
sum 1.16626E-01 1.00000E+00 1.00012E+00
-------------------------------------------------------------------

B.)
I also want to take a pulse height spectrum in a germanium crystal in some position so I create

det 1c dc 201 dr -27 ge de peterbins3
ene peterbins3 2 2048 1e-3 10

mat ge -5.323 rgb 0 150 0
32070.70c 0.2123
32072.70c 0.2766
32073.70c 0.0773
32074.70c 0.3594
cell 201 0 ge -251
surf 251 cylx 0 0 15 2.5 20

Is this a good way to take a pulse height spectrum?

c.)

In my initial burn up calculation (where I create my petermat.dat material file) I use .70c cross sections.
Why are the ending .84p in the output of the photon source problem where material inputs are from petermat.dat?

Thanks!

### Re: Gamma detectors and flux to dose

Posted: Tue Jan 31, 2017 10:29 am
In radioactive decay source mode Serpent normalizes the source rate to total emission rate from radioactive decay. If your volumes are correct, you should get the output automatically in correct units. When you use a neutron transport input for photon transport calculation Serpent converts automatically all isotopic compositions to elemental and uses the first photon data library in the directory file.

### Re: Gamma detectors and flux to dose

Posted: Tue Jan 31, 2017 11:40 am
Thanks!

Since correct cell volumes are important: If I use a negative universe for a detector, does mcvol also include such cells? I know that it doesnt calculate negative universe volumes when using checkvolumes.

Thanks, Peter

EDIT: Sorry it does indeed calculate negative universes...
Jaakko Leppänen wrote:In radioactive decay source mode Serpent normalizes the source rate to total emission rate from radioactive decay. If your volumes are correct, you should get the output automatically in correct units. When you use a neutron transport input for photon transport calculation Serpent converts automatically all isotopic compositions to elemental and uses the first photon data library in the directory file.

### Re: Gamma detectors and flux to dose

Posted: Tue Jan 31, 2017 2:44 pm
Which of the ENDF reaction rates ( http://serpent.vtt.fi/mediawiki/index.p ... on_numbers ) would be equivalent to typical dose rate counters carried/used at facilities?
or is this a better conversion table (ICRP 21), taken from the manual of another code.. :

dr -100 1 100 1.00E-02 7.72E-12 1.50E-02 3.08E-12 2.00E-02 1.63E-12 3.00E-02 7.11E-13 4.00E-02 4.33E-13 5.00E-02 3.33E-13
6.00E-02 3.08E-13 8.00E-02 3.33E-13 1.00E-01 4.08E-13 1.50E-01 6.61E-13 2.00E-01 9.58E-13 3.00E-01 1.54E-12
4.00E-01 2.14E-12 5.00E-01 2.53E-12 6.00E-01 3.17E-12 8.00E-01 4.08E-12 1.00E+00 4.97E-12 1.50E+00 6.78E-12
2.00E+00 8.42E-12 3.00E+00 1.11E-11 4.00E+00 1.32E-11 5.00E+00 1.54E-11 6.00E+00 1.74E-11 8.00E+00 2.14E-11
1.00E+01 2.53E-11 air2

(The above table in mcnp format: )

de24 0.100E-01 0.150E-01 0.200E-01 0.300E-01 0.400E-01 0.500E-01
0.600E-01 0.800E-01 0.100E+00 0.150E+00 0.200E+00 0.300E+00
0.400E+00 0.500E+00 0.600E+00 0.800E+00 0.100E+01 0.150E+01
0.200E+01 0.300E+01 0.400E+01 0.500E+01 0.600E+01 0.800E+01
0.100E+02
df24 7.72222E-12 3.08333E-12 1.63333E-12 7.11111E-13 4.33333E-13
3.33333E-13 3.08333E-13 3.33333E-13 4.08333E-13 6.61111E-13
9.58333E-13 1.54444E-12 2.13611E-12 2.525E-12 3.16667E-12
4.08333E-12 4.97222E-12 6.77778E-12 8.41667E-12 1.11111E-11
1.32222E-11 1.54444E-11 1.73611E-11 2.13611E-11 2.525E-11