Radioactive decay source

Separate section for discussion on gamma transport
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Jaakko Leppänen
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Radioactive decay source

Post by Jaakko Leppänen » Fri Jun 26, 2015 2:27 pm

Photon transport simulations can be run with a radioactive decay source, which combines material compositions (isotopic) to radioactive decay data and photon line spectra read from ENDF decay files. The syntax for the decay source is:

Code: Select all

src <name> g sg <mat> <mode>
where <mat> is the material in which the decay gammas are emitted or "-1" if all materials are included. The <mode> parameter has two options:

1 - Analog source sampling, which means that the source point is sampled randomly, and accepted with probability given by the ratio of local to maximum emission rate
2 - Implicit source biasing, which means that the source point is sampled randomly, all points in radioactive materials are accepted, and the weight of the emitted photon is scaled based on the ratio of local to average emission rate

Mode 2 can be much faster, but may lead to statistical problems if source photons with very low and very high weight can access the same region of the geometry. The decay source can be used in combination with the "sx", "sy", and "sz" options, which define the boundaries for the sampling volume, and may significantly improve the efficiency if the source material covers only a small fraction of the geometry.

The decay source mode also requires a decay data library to be defined ("set declib"), and that the volumes of radioactive materials are calculated automatically or provided in the input. The results are automatically normalized to total emission rate calculated from the material and decay data. The calculation also prints the source intensity spectra in a separate file "<input>_gsrc.m" (at the moment mainly for debugging purposes).

There are two ways in which I've found the use of this source mode convenient...

In photon transport mode it is possible to define isotopic material compositions and let Serpent calculate the elemental compositions automatically (see related post). The compositions can be defined without neutron data library ID's using the ZAI's, for example:

Code: Select all

mat uranium -19.0 
922340      0.00540
922350      0.72000
922380     99.27000
This material will be converted into elemental uranium for the transport simulation, but since it has isotopic composition, in can also be used with the radioactive decay source, for example:

Code: Select all

src 1 g sg uranium   2
If there were more radioactive materials in the geometry, replacing material name "uranium" with "-1" would extend the source to cover all of them.

Another way to use the decay source is to combine the results of an activation or burnup calculation with a photon transport simulation. This can be done using binary restart files described in a related thread. For example, if material "uranium" was irradiated for 100 days in a previous neutron transport calculation, and the compositions written in restart file "U.comp", this irradiated material could be used for decay source by defining an additional entry:

Code: Select all

set rfr -100  "U.comp"
Serpent would then read the composition for material "uranium" from file "U.comp" at 100 days irradiation, overriding the composition provided in the material card, and use that for the radiactive decay source.

An complete example of a decay source with irradiated fuel is found at:

http://virtual.vtt.fi/virtual/montecarl ... emos/STL3/

The example is a bit complicated, because the geometry is defined using the new CAD-based geometry type.
- Jaakko

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Re: Radioactive decay source

Post by Andy_Turner » Tue Jan 12, 2016 3:44 pm

Hi Jaakko

When I create myself a case to run, I get the following error on the gamma run:
"Input error in parameter "mat" on line 74 in file "iter_pp_act.comp":
Nuclide 160000 not found in data libraries"

I am able to direct Serpent to the MCPLIB84 photon libraries (it works with ZZZ000.84p materials) and also providing it two libraries under 'set acelib' and it makes ZZZ000 entries automatically - so the photon data is linked okay. 'Set pdatadir' is also set. But when I try to feed it a binary from an activation step to use, it can't find this nuclide. At Z=160, so I'm not surprised it isn't in the libraries, since it appears the neutron burn is doing something rather exotic.
Whatever has gone wrong with the burn step, there is always the possibility that photon production data won't be available for a nuclide created in the input, so this should be handled by serpent (e.g. just warning the user that no data is available for that nuclide).
Is there a way to ask Serpent to ignore the error?

Kind regards
- Andy
- Andy Turner, CCFE

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Re: Radioactive decay source

Post by Jaakko Leppänen » Tue Jan 12, 2016 4:03 pm

So I guess the problem is that the activation calculation produces isotopes for which there is no photon data? Could you post your input here (if it's simple enough)?
- Jaakko

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Re: Radioactive decay source

Post by Andy_Turner » Tue Jan 12, 2016 4:38 pm

----Neutron calculation--------:
surf 1 sph 0 0 0 10
surf 2 sph 0 0 0 20

cell 1 0 steel -1
cell 2 0 void 1 -2
cell 3 0 outside 2

set nps 100000 1
set acelib "/home/serpent/serpent1/xsdata/xs/jeff31.serp.xsdir"
set declib "/home/serpent/serpent1/xsdata/xs/sss_jeff311.dec"
src my_ptsrc se 14.0 sp 0 0 0

mat steel -7.93 burn 1
5010.03c 1.0227E-05
5011.03c 4.1166E-05
6012.03c 1.0399E-03
7014.03c 2.7695E-03
7015.03c 1.0118E-05
8016.03c 6.9490E-05
13027.03c 1.0299E-03
14028.03c 9.1196E-03
14029.03c 4.6328E-04
14030.03c 3.0576E-04
15031.03c 4.4794E-04
16000.03c 1.2998E-04
19000.03c 7.0990E-06
22046.03c 1.4353E-04
22047.03c 1.2944E-04
22048.03c 1.2825E-03
22049.03c 9.4121E-05
22050.03c 9.0119E-05
23000.03c 4.3594E-05
24050.03c 8.1240E-03
24052.03c 1.5666E-01
24053.03c 1.7764E-02
24054.03c 4.4219E-03
25055.03c 1.8197E-02
26054.03c 3.7637E-02
26056.03c 5.9081E-01
26057.03c 1.3644E-02
26058.03c 1.8158E-03
27059.03c 4.7093E-04
28058.03c 7.8958E-02
28060.03c 3.0415E-02
28061.03c 1.3221E-03
28062.03c 4.2154E-03
28064.03c 1.0735E-03
29063.03c 1.8115E-03
29065.03c 8.0816E-04
40000.03c 1.2198E-05
41093.03c 5.9792E-05
42092.03c 2.1414E-03
42094.03c 1.3382E-03
42095.03c 2.3052E-03
42096.03c 2.4183E-03
42097.03c 1.3860E-03
42098.03c 3.5071E-03
42100.03c 1.4020E-03
50000.03c 9.3587E-06
73181.03c 3.0696E-05
74182.03c 8.0115E-07
74183.03c 4.3262E-07
74184.03c 9.2631E-07
74186.03c 8.5950E-07
82206.03c 5.2299E-07
82207.03c 4.7959E-07
82208.03c 1.1371E-06
83209.03c 2.1297E-06

% --- Switch group constant generation off:
set gcu -1
% --- Full delta-tracking mode:
set dt 1.0
% --- Write material compositions in binary restart file:
set rfw 1 "test.comp"
% --- No predictor-corrector:
set pcc 0
% --- Depletion history:
set srcrate 1.0714e17
dep daystep 500

----- gamma run -----

surf 1 sph 0 0 0 10
surf 2 sph 0 0 0 20

cell 1 0 steel -1
cell 2 0 void 1 -2
cell 3 0 outside 2

set nps 100000 1
set acelib "/home/serpent/serpent1/xsdata/xs/jeff31.serp.xsdir" "/home/mcnp/xs/xsdir_serp_test"
set declib "/home/serpent/serpent1/xsdata/xs/sss_jeff311.dec"
set pdatadir "/home/mcnp/xs/photon_data"
% material is over-written
mat steel -7.93
5010.03c 1.0227E-05
5011.03c 4.1166E-05
6012.03c 1.0399E-03
7014.03c 2.7695E-03
7015.03c 1.0118E-05
8016.03c 6.9490E-05
13027.03c 1.0299E-03
14028.03c 9.1196E-03
14029.03c 4.6328E-04
14030.03c 3.0576E-04
15031.03c 4.4794E-04
16000.03c 1.2998E-04
19000.03c 7.0990E-06
22046.03c 1.4353E-04
22047.03c 1.2944E-04
22048.03c 1.2825E-03
22049.03c 9.4121E-05
22050.03c 9.0119E-05
23000.03c 4.3594E-05
24050.03c 8.1240E-03
24052.03c 1.5666E-01
24053.03c 1.7764E-02
24054.03c 4.4219E-03
25055.03c 1.8197E-02
26054.03c 3.7637E-02
26056.03c 5.9081E-01
26057.03c 1.3644E-02
26058.03c 1.8158E-03
27059.03c 4.7093E-04
28058.03c 7.8958E-02
28060.03c 3.0415E-02
28061.03c 1.3221E-03
28062.03c 4.2154E-03
28064.03c 1.0735E-03
29063.03c 1.8115E-03
29065.03c 8.0816E-04
40000.03c 1.2198E-05
41093.03c 5.9792E-05
42092.03c 2.1414E-03
42094.03c 1.3382E-03
42095.03c 2.3052E-03
42096.03c 2.4183E-03
42097.03c 1.3860E-03
42098.03c 3.5071E-03
42100.03c 1.4020E-03
50000.03c 9.3587E-06
73181.03c 3.0696E-05
74182.03c 8.0115E-07
74183.03c 4.3262E-07
74184.03c 9.2631E-07
74186.03c 8.5950E-07
82206.03c 5.2299E-07
82207.03c 4.7959E-07
82208.03c 1.1371E-06
83209.03c 2.1297E-06

% source
% radioactive decay source -1 = in all active materials. Last entry, 1 = analogue, 2 = implicit bias.
src 1 g sg -1 1
% overwrite composition with result at N days.
set rfr -500 "test.comp"
-----------------------------------

If I repeat the exercise with the neutron run material being 100% 26056.03c then it works fine.

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Re: Radioactive decay source

Post by Jaakko Leppänen » Tue Jan 12, 2016 5:24 pm

I got the same problem. It's not related to missing data, but something else. I'll look into it and get back to you...
- Jaakko

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Re: Radioactive decay source

Post by Jaakko Leppänen » Tue Jan 12, 2016 5:32 pm

The problem is more generally related to using natural element compositions in burnable (activated) materials. When you run the neutron transport calculation, Serpent doesn't know what to do with nuclides such as 16000.03c. This is not really the main cause of this error, but replacing the elemental nuclides with with their corresponding isotopic composition in the activation calculation should remove this error as well.
- Jaakko

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Re: Radioactive decay source

Post by Andy_Turner » Tue Jan 12, 2016 6:31 pm

Many thanks, yes it runs when I avoid natural element ZAIDs.
- Andy Turner, CCFE

Peter Wolniewicz
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Re: Radioactive decay source

Post by Peter Wolniewicz » Wed Jan 25, 2017 12:00 pm

Is it possible to use "negative" compositions (by mass) instead of isotopic in this type of source definitions?

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Re: Radioactive decay source

Post by Jaakko Leppänen » Wed Jan 25, 2017 10:29 pm

Yes. It's a standard material card.
- Jaakko

Peter Wolniewicz
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Re: Radioactive decay source

Post by Peter Wolniewicz » Thu Jan 26, 2017 1:09 pm

Jaakko Leppänen wrote:The problem is more generally related to using natural element compositions in burnable (activated) materials. When you run the neutron transport calculation, Serpent doesn't know what to do with nuclides such as 16000.03c. This is not really the main cause of this error, but replacing the elemental nuclides with with their corresponding isotopic composition in the activation calculation should remove this error as well.
I dont have any natural compositions in my material card :

mat cfue22 -19 rgb 230 0 0 vol 8.4211e-4 burn 1
92234.70c 1.3
92235.70c 92
92236.70c 0.02
92238.70c 6.6

the bumat file produced contains nuclides such as:
310860 2.99780233440427E-10

When I then run a SRC calculation using:
src 1 g sg -1 1

and using the materials produced I get:

Input error in parameter "mat" on line 279 in file "FC1a":
Nuclide 310860 not found in data libraries


Im lost here ... What can I do?

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