## Gamma detectors and flux to dose

Separate section for discussion on gamma transport
Peter Wolniewicz
Posts: 135
Joined: Mon Dec 13, 2010 5:50 pm

### Re: Gamma detectors and flux to dose

Jaakko Leppänen wrote:The detector calculates volume-integrated flux [cm/s]. The -2XX response functions multiply flux by energy [MeV] and the tabulated mass energy-absorption coefficient [cm2/g]. This gives for the units:

[cm/s]*[MeV]*[cm2/g] = [cm3*MeV/g/s]

The result is then multiplied by a factor that converts the energy, mass and time units into J, kg and h --> [cm3*J/kg/h]. When the results is divided by volume you get [J/kg/h], i.e. [Gy/h].

Or am I missing something?

When you use the pre-determined mass energy-absorption coefficients for mixtures and compounds (-201 to -248) the medium in which the dose rate is calculated has no direct effect on the result (there is obviously an indirect effect that comes from the attenuation of photon flux). With response -200 Serpent calculates the mass energy-absorption coefficient for the medium automatically based on tabulated elemental data.
What a great answer! Thank you Jaakko! I'll dive into my calculations tomorrow when I get to work again

Peter Wolniewicz
Posts: 135
Joined: Mon Dec 13, 2010 5:50 pm

### Re: Gamma detectors and flux to dose

I get:
"Total photon source rate is zero in decay source mode",
when I remove material definitions from my input file (although they are pointed to by "set rfr -2019 petermat.dat" ). I thought that material compositions were overidden when I used "set rfr".

The different burn-up steps from the depletion calculation which produced my source file have radioactivities in many nuclides so I dont understand why I get that message. What I am doing wrong?

%Fission chamber
%Fission chamber description. Position A2
cell 170 34 argon -64
cell 171 34 cfue22 64 -65
cell 172 34 inox 65 -66
cell 173 34 nose -67
cell 174 34 insulator -68
cell 175 34 brazing -69
cell 176 34 cableiron 73 -74
cell 177 34 cablecopper 72 -73
cell 178 34 softiron 71 -72
cell 179 34 mgo 70 -71
cell 180 34 cablecopper -70
cell 182 34 air 74 69 68 66 67 76
cell 183 34 air 75 -76

cell 185 34 nose -75 -76

cell 105 0 air2 251 -100
%%cell 104 0 air2 251 -100
%cell 106 0 air2 251 -101
cell 20 0 outside 100
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%5
cell 103 0 fill 34 -253 250 263 264

%Detector cells
cell 201 0 ge 253 -251 -256 %Tube Germanium detector
cell 203 0 air2 253 -251 256 %Tube Germanium detector
cell 202 -1 void -252 %Big detector covering everything

cell 200 0 void -250 %Middle, Dose counters
cell 207 0 void -263 %Above
cell 208 0 void -264 %Below

det 3a dc 207 dr -100 icrp21 dv 6.7693 %above touch
det 3b dc 200 dr -100 icrp21 dv 6.7693 %middle touch
det 3c dc 208 dr -100 icrp21 dv 6.7693 %below touch

det 3d dc 207 dr -201 air dv 6.7693 %above touch
det 3e dc 200 dr -201 air dv 6.7693 %middle touch
det 3f dc 208 dr -201 air dv 6.7693 %below touch

surf 250 cylx 0 0 1.34 0.36 1.56 %r=0.34 %detector dose middle
surf 263 cylx 0 5 1.34 0.36 1.56 %r=0.34 %detector dose above
surf 264 cylx 0 -5 1.34 0.36 1.56 %r=0.34 %detector dose below

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
%Virtual dose cells
cell 204 -1 void -260 %Dose counter in air
cell 205 -1 void -261
cell 206 -1 void -262

mat air -0.001225 rgb 205 205 255
7014.60c 80
8016.60c 20

mat air2 -0.001225 rgb 205 105 215
7014.60c 80
8016.60c 20

mat water -1 rgb 255 0 255
1001.70c 2
8016.70c 1
therm lwtr1 lwtr.01t

set acelib "/home/peter/MY_MCNP/MCNP_DATA/sss.xsdata"
set declib "/home/peter/work/_Serpent/_extralibs/gustav/endf-b-vi-8_decay.dat"
set nfylib "/home/peter/work/_Serpent/_extralibs/gustav/endf-b-vi-8_nfpy.dat"
set sfylib "/home/peter/work/_Serpent/_extralibs/gustav/endf-b-vi-8_sfpy.dat"
set printm 1 0 %Last number is the threshold atomic fraction above which decay nuclides are included in the bumat output.
set inventory
set nps 100000
set mcvol 1000000

src mysource g sg -1 1 sx -0.525 0.525 sy -0.525 0.525 sz -10 88.1184
set rfr -2019 petermat.dat
%set rfr -1101 petermat.dat

surf 64 cyl 0 0 0.2401285253 -2.8 2.8 %fissile coating inner radius
%surf 65 cyl 0 0 0.303000000 -2.8 2.8 %fissile coating outer radius TEMPORARY BECAUSE OPTICALLY THIN ORIGINAL BELOW
surf 65 cyl 0 0 0.2401714748 -2.8 2.8 %fissile coating outer radius
surf 66 cyl 0 0 0.35 -6.8 6.8 %fission chamber outer cyl
surf 67 cyl 0 0 0.525 -10 -6.8 %Nose outer boundary (OB)
surf 68 cyl 0 0 0.3500000000 6.8 7.5894 %Insulator material OB
surf 69 cyl 0 0 0.3500000000 7.5894 7.6184 %brazing material OB
surf 70 cyl 0 0 0.0225 7.6184 88.1184 %cable r5 Approximate length 80.5 cm (+- about 20 cm)
surf 71 cyl 0 0 0.150 7.6184 88.1184 %cable r4
surf 72 cyl 0 0 0.175 7.6184 88.1184 %cable r3
surf 73 cyl 0 0 0.275 7.6184 88.1184 %cable r2
surf 74 cyl 0 0 0.300 7.6184 88.1184 %cable r1
surf 75 cone 0 0 -10 0.525 -0.525 %NOSE TIP
surf 76 pz -10
surf 100 cyl 0 0 35.000000000 -20 88.1184 %space for fission chamber
surf 101 cyl 0 0 40 -20 88.1184
%Photon detectors dv=0.0726

surf 251 cyl 0 0 20 -20 88.1184 %detector cylinder Ro
surf 252 cuboid -30 30 -30 30 -30 30
surf 253 cyl 0 0 4 -20 88.1184 %detector cylinder Ri
surf 254 cylx 0 -5 0.34 -0.56 -0.36 %r=0.34 %detector dose B
surf 255 cylx 0 5 0.34 -0.56 -0.36 %r=0.34 %detector dose C
surf 256 px 0 %cut surf 251 and 253 in half

%side dosimeters
surf 260 sph 10 0 0 2
surf 261 sph 20 0 0 2
surf 262 sph 30 0 0 2

mat ge -5.323 rgb 0 150 0
%32000.63p 1
32070.70c 0.2123
32072.70c 0.2766
32073.70c 0.0773
32074.70c 0.3594

det 1a dc 201 dr -27 ge %dr -27 eq. to pulse height in MCNP6 See serpent wiki on ENDF
det 1b dc 201
det 1c dc 201 dr -27 ge de peterbins3

%Below:
%ICRP 21 MCNP man.
%(Sv/second /p/cm-2) multip. result by 3600 to get Sv/h
det 1d dc 200 dr -100 icrp21
fun icrp21 1 1 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

%virtual dose detectors
det 2a dc 204 dr -100 icrp21 dv 4.1888 %first
det 2b dc 205 dr -100 icrp21 dv 4.1888 %second
det 2c dc 206 dr -100 icrp21 dv 4.1888 %third

plot 2 1000 1000 0 -15 15 -15 15
mesh 8 3 1b 2 1000 1000 0 -15 15 -1 1 -15 15
mesh 8 3 1b 3 1000 1000
mesh 8 3 1b 1 1000 1000

plot 3 3000 3000 0 -10 10 -10 10
plot 2 3000 3000 0 -100 100 -100 100
plot 2 3000 3000 0 -5 5 -16 -6

ene peterbins3 2 2048 1e-3 10

Jaakko Leppänen
Posts: 2388
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Location: Espoo, Finland
Contact:

### Re: Gamma detectors and flux to dose

What are the source materials? You need have them defined using material cards in the same way as in the calculation that produced the restart file. The "set rfr" option only updates the compositions. It doesn't create materials that are not defined.
- Jaakko

Peter Wolniewicz
Posts: 135
Joined: Mon Dec 13, 2010 5:50 pm

### Re: Gamma detectors and flux to dose

Jaakko Leppänen wrote:What are the source materials? You need have them defined using material cards in the same way as in the calculation that produced the restart file. The "set rfr" option only updates the compositions. It doesn't create materials that are not defined.
I have two input files, one of them is a reactor with a fission chamber in it, the other one is the fission chamber model extracted and the reactor part removed. In the second file I also added some detectors and changed the material around the fission chamber. The fission chamber components are the same in both cases - just the environment it is sitting in is changed.

1.) The first input file is run with "set bunorm 1" and power according to total reactor power - I try to calculate the activation, decay and transmutation (of the fissile coating) of the fission chamber. The reactor fuel material has no burn entry but all fission chamber materials has burn 1. All materials during different burn/decay steps are written to a file using the "set rfw 1 petermat.dat".

2.) The second file has a different geometry than the first - it is basically just the fission chamber plus air around it instead of water. I also put in a germanium detector volume to tally the gamma spectrum. All materials of the fission chamber are copied from the first run, and I also added "set rfr -1101 petermat.dat" in order to replace the "copied" materials from the first file by the material compositions during the decay step at 1101 days.

All seems well, but what I did was that since I thought I didn't need the "old materials" (as they are not used), I entered just the material names in the CELL definition and removed all fission chamber material definitons from the input file. The simulation complained about that no radioactivity was found in the source. What is the purpose of having to add material cards that are not used in the input file - they are overwritten anyways? Would a change in the (gamma source) input file of, e.g., the material density or isotopic compositions of materials already present in the rfr-file have any effect on the results?

I have another question regarding this: Since the fissile coating of the fission chamber is very thin, could I in my first run increase the thickness of the fissile coating in order to get better statistics and then in my radioactive source run change the thickness back to its proper value without it causing too much trouble on other reaction rates (I know it will have a slight effect on reaction rates in other parts of the geometry, but I can probably live with such minor errors elsewhere). Or do you here recommend using e.g. "set minxs LN 0.1" or some other small value?

Thanks!

Jaakko Leppänen
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Location: Espoo, Finland
Contact:

### Re: Gamma detectors and flux to dose

Can you send me the input files?
- Jaakko

viksap
Posts: 12
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### Re: Gamma detectors and flux to dose

Peter Wolniewicz wrote:
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!
Sorry to ask this silly question but can you tell me how to plot from the output(from input_det0.m) in the form of pulse height spectrum and which cross section should be used in the case of germanium detector, .70c or .05p for pulse height spectrum calculation.

Thank You

kaartinen
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Location: Loviisa NPP

### Re: Gamma detectors and flux to dose

viksap wrote:
Sorry to ask this silly question but can you tell me how to plot from the output(from input_det0.m) in the form of pulse height spectrum and which cross section should be used in the case of germanium detector, .70c or .05p for pulse height spectrum calculation.

Thank You
For example with an octave script. Some examples are shown in tutorials.

Here is one quick one:

Code: Select all

``````run /path/to/your/file/named/input_det0.m;
plot(DET1E(:,2),DET1(:,11), 'k-');
set (gca, 'FontSize', 16);
set (gca, 'YScale','log');
xlabel("energy");
ylabel("number of pulses");
title("a title");
grid on;
box on;
print -dpng plot.png``````
to run the octave script in terminal:

Code: Select all

``octave nameofyourscript.m``

viksap
Posts: 12
Joined: Thu Nov 15, 2018 2:29 pm
Security question 1: No
Security question 2: 96

### Re: Gamma detectors and flux to dose

Thank You very much!!

viksap
Posts: 12
Joined: Thu Nov 15, 2018 2:29 pm
Security question 1: No
Security question 2: 96

### Re: Gamma detectors and flux to dose

kaartinen wrote:
viksap wrote:
Sorry to ask this silly question but can you tell me how to plot from the output(from input_det0.m) in the form of pulse height spectrum and which cross section should be used in the case of germanium detector, .70c or .05p for pulse height spectrum calculation.

Thank You
For example with an octave script. Some examples are shown in tutorials.

Here is one quick one:

Code: Select all

``````run /path/to/your/file/named/input_det0.m;
plot(DET1E(:,2),DET1(:,11), 'k-');
set (gca, 'FontSize', 16);
set (gca, 'YScale','log');
xlabel("energy");
ylabel("number of pulses");
title("a title");
grid on;
box on;
print -dpng plot.png``````
to run the octave script in terminal:

Code: Select all

``octave nameofyourscript.m``
Hi,

There is a small confusion regarding the detector specifications card, to simulate the pulse height spectrum I have used "dr -27 void" and my detector material is germanium. I am not clear about whether to use void or germanium in "dr -27" entry. In the manual the explanation is given but still it is not clear to me.

Thank you

Ville Valtavirta
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### Re: Gamma detectors and flux to dose

It's generally best to use "void".

void simply uses whatever material(s) is/are actually at the detector location.

-Ville