External fission source of Cf252

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bouchama l
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External fission source of Cf252

Post by bouchama l » Sun May 12, 2019 3:05 pm

Hi users!
I'm novice in serpent, I'm using Serpent to evaluate my code in Geant4, so I make a comparison in different medium (water, concrete and air). At the first time I'm using a punctual source of neutron (2MeV), and I had a good agreement, at second time I'm trying to use a punctual fission source Cf252, in this case I got a discrepancy within two times mach more than the precedent comparison for the same conditions, I'm looking for a long time to resolve this problem but needlessly, I think that the problem is concerned the californium source declaration.
I will be thankful for any proposition.

Thanks in advance.

Her is my code:

set title "External neutron source"
/************************
* Material definitions *
************************/

% --- Water: water density = 0.99823 g/cm3 for temperature = 293.6k

mat water -0.99823 moder lwtr 1001 % 03c = 300k---Adding new temperature by tmp but >>> than 300
1001.03c 2.0
8016.03c 1.0

% --- Thermal scattering data for light water:

therm lwtr lwe7.00t

/************************
* Geometry definitions *
************************/

surf 1 sph 0.0 0.0 0.0 5.0

cell c1 0 water -1
cell c2 0 outside 1

/************************
**** Run parameters ****
************************/

% --- Cross section data

set acelib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7u.xsdata"

% --- Source

src FissionSource sr 98252.03c 18 sp 0.0 0.0 0.0

% --- Particle histories (Numbre of particles + batch size)

set nps 3000000 200

% --- Geometry and mesh plots

plot 3 500 500
mesh 3 500 500

% ---- plot all cross-sections in m-file:

%set xsplot 100 1E-9 10.0 % plots cross sections in given energy range

% --- Detector energy grid (uniform lethargy):

ene 1 3 100 1E-9 10.0 % ene NAME type Emin Emax % type 2 = equal energy-width bins, 3 = equal lethargy-width bins

% --- Detectors

% Flux on surface
% dt : -1= cumulative spectrum; -2= division by energy width; -3= division by lethargy width; -4= sum over cell or material bins
% ds surf param % param = -2 = flux, -1 = inward current, 1 = outward current, 0 = net current)

det flux1 de 1 ds 1 -2

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Jaakko Leppänen
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Re: External fission source of Cf252

Post by Jaakko Leppänen » Tue May 14, 2019 8:44 am

Your source definition actually gives the neutron-induced fission rate in Cf-252. The "sr" entry defines the isotope (98252.03c) and reaction (18).

To get a spontaneous fission source, you'll need to define spectrum manually or use the radioactive decay source option.
- Jaakko

bouchama l
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Re: External fission source of Cf252

Post by bouchama l » Thu May 16, 2019 5:30 pm

Thank for the reply Jakko!

I am trying to define the Californium source as you said, but I get some errors message, that said "Decay library must be defined for radiation source"
I add the following modification:
1)
% Californium
mat mat_cf252 -15.1
98252.03c 1.0
2)
set declib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7.dec"
3)
src FissionSource n sg mat_cf252 2 sp 0.0 0.0 0.0
You said that it is possible to define a Spontaneous fission source by adding manually the source spectrum, in Geant4 I use this way, so can you explain me how can do it?

Kind regards!
Last edited by bouchama l on Mon May 27, 2019 12:20 pm, edited 2 times in total.

bouchama l
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Re: External fission source of Cf252

Post by bouchama l » Fri May 17, 2019 3:54 pm

Hi All;
To configure the Californium source I'm using the two proposition of Jaakko, but unfortinatlly they dont work. When I define the source by using the spectrum fission of cf252 like:

% --- Cross section data
set acelib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7u.xsdata"
set declib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7.dec"

% --- Source
src Cf252_Source sm mat_cf252 sg mat_cf252 1 sp 0.0 0.0 0.0

I get the following Error:

***** Fri May 17 15:28:33 2019
Input error:
No neutron decay source (may be due to lack of data)

For the seconde proposition "radioactive decay source", declared by the way:
src Cf252_Source
se
0.03791
0.08301
0.19830
0.34993
0.60544
0.81070
1.08555
1.55102
1.89664
2.39728
2.82949
3.50389
4.40363
5.04448
5.61632
6.24062
7.02125
8.84360
sw
0.0
32.71076
45.89684
59.07265
68.10622
74.34164
76.17980
73.89820
64.25226
55.40119
45.95365
37.80812
27.66813
17.72709
12.47039
8.97969
6.12068
3.73887
1.12916
sp 0.0 0.0 0.0

I get the following error:

***** Fri May 17 15:46:06 2019
Input error in parameter "src" on line 57 in file "neutron_water":
Invalid source parameter "0.08301"

Any help, explication or proposition are welcom;
Best regards!

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Jaakko Leppänen
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Re: External fission source of Cf252

Post by Jaakko Leppänen » Sun May 19, 2019 8:53 pm

The first error probably means that the decay data library doesn't have data for Cf252 spontaneous fission. Try using JEFF-3.1 instead.

For the the second format you should use the "sb" option of the source card:

http://serpent.vtt.fi/mediawiki/index.p ... ual#src_sb

Note that the format was changed in the latest update distributed last week. The "se" option is used to define a mono-energetic source.
- Jaakko

bouchama l
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Re: External fission source of Cf252

Post by bouchama l » Wed May 22, 2019 4:51 pm

Hi Jaakko!

I'm using the library JEFF31, but it gives the same Error.

% --- Cross section data
%set acelib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7u.xsdata"
%set declib "/home/zazo/Serpent/xsdata/endfb7/sss_endfb7.dec"
set acelib "/home/zazo/Serpent/xsdata/jeff311/sss_jeff311.xsdata"
set declib "/home/zazo/Serpent/xsdata/jeff311/sss_jeff311.dec"

% --- Source
src Cf252_Source sm mat_cf252 sg mat_cf252 1 sp 0.0 0.0 0.0

for the seconde way, I could define the spectrum fission by the way:
% --- Source
src Cf252_Source
sb 18
0.01731 0.0
0.03791 32.71076
0.08301 45.89684
0.19830 59.07265
0.34993 68.10622
0.60544 74.34164
0.81070 76.17980
1.08555 73.89820
1.55102 64.25226
1.89664 55.40119
2.39728 45.95365
2.82949 37.80812
3.50389 27.66813
4.40363 17.72709
5.04448 12.47039
5.61632 8.97969
6.24062 6.12068
7.02125 3.73887
8.84360 1.12916
sp 0.0 0.0 0.0
but by comparing with Geant4 result i get a hight descrepency than when I declare the source by using the command
src Cf252_Source sr 98252.03c 18 sp 0.0 0.0 0.0, in spite of I used the same spectrum like in Geant4.
Best regards.

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Jaakko Leppänen
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Re: External fission source of Cf252

Post by Jaakko Leppänen » Thu May 23, 2019 9:50 am

Which Serpent version are you using? In my 2.1.31 the calculation is running with JEFF-3.1 data and decay source definition :

Code: Select all

src FissionSource 
sg Cf 1
sx -0.1 0.1
sy -0.1 0.1
sz -0.1 0.1
when the geometry is defined as:

Code: Select all

surf 1 sph 0.0 0.0 0.0 5.0
surf 2 sph 0 0 0 0.1

cell c0 0 Cf -2
cell c1 0 water -1 2
cell c2 0 outside 1
The other source definition also depends on which interpolation is used between the points (see the syntax of the sb entry in version 2.1.31).
- Jaakko

bouchama l
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Re: External fission source of Cf252

Post by bouchama l » Wed May 29, 2019 6:06 pm

Hi Jaakko,
The code works well when I tried your proposition to define the radioactive source with the Jeff311 but with the library ENDF/B-VII does not work, my question is: does the radioactive source that you define considered a point-like source or not?
Question 2: get a discrepancy of 10% between Geant4 and Serpent is acceptable, or since they used Monte-Carlo and almost the same libraries the discrepancy will be not more than ~2%?
Best regards!

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Jaakko Leppänen
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Re: External fission source of Cf252

Post by Jaakko Leppänen » Mon Jun 03, 2019 10:36 am

The radioactive decay source "sg" samples source points uniformly throughout the radioactive materials. So it is not a point source, but you can define the source material using a small volume.

There are several factors that may affect the discrepancies, including cross section data used in the calculations. I don't know how large discrepancies to expect between Serpent and Geant4, but in neutron transport calculations using equivalent models and same cross section libraries Serpent results should match MCNP within statistics.
- Jaakko

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