microscopic cross section calculation

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microscopic cross section calculation

Postby motalab » Mon Aug 03, 2015 12:11 pm

In a fuel pin there are different kinds of nuclides. I want to calculate the microscopic cross section of a particular nuclide. How can I write the detector input line? Suppose I want to calculate the microscopic absorption cross section of Xenon in the CANDU fuel pin.

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Re: microscopic cross section calculation

Postby Jaakko Leppänen » Tue Aug 04, 2015 9:50 am

See the examples in Sec. 7.1.1 of Serpent Manual.
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Re: microscopic cross section calculation

Postby aki » Fri Aug 21, 2015 10:04 pm

Sec. 7.1.1 of Serpent Manual. talks about the macroscopic cross section.
It gives an example where you actually calculate the macroscopic fission and capture cross section of U-235 and U-238 by dividing the reaction rate with the flux.

I am curious if there is a way to actually get the microscopic cross section ?
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Re: microscopic cross section calculation

Postby Jaakko Leppänen » Sun Aug 23, 2015 11:53 am

You can calculate microscopic cross section in a similar way, using microscopic reaction rates (positive reaction numbers in the response function).
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Re: microscopic cross section calculation

Postby aki » Mon Aug 24, 2015 5:55 pm

Thanks a lot.
Can I actually get the microscopic scattering cross section matrix ??
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Re: microscopic cross section calculation

Postby Jaakko Leppänen » Mon Aug 24, 2015 7:56 pm

Unfortunately, no.
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Re: microscopic cross section calculation

Postby motalab » Wed Aug 26, 2015 11:48 am

set poi "<opt>", using this option we can calculate fission product poison cross section(I-135, Xe-135, Pm-149 and Sm-149 and absorption of Xe-135 and Sm-149). My question is that the microscopic absorption cross section of Xe-135 calculated by Serpent2, is it homogenized or not?

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Re: microscopic cross section calculation

Postby Jaakko Leppänen » Wed Aug 26, 2015 12:37 pm

The cross section is averaged over all fissile materials, so I guess you can call it homogenized.
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Scattering matrixes order

Postby motalab » Fri Sep 18, 2015 5:20 am

Hi Jaakko

B1_S0 (idx, [1: 8]) = [ 3.08714420E-01 1.4E-05 8.64302225E-03 6.8E-05 6.53812289E-05 0.00104 4.34455707E-01 9.2E-06 ];

the above line is scattering matrix for 2 group calculation. I arranged them like:Self scattering in fast group, up-scattering, down-scattering, and Self scattering in thermal group. All are associated with error. Am I correct in order? If I am correct then why up-scattering is higher than down-scattering in CANDU lattice in the above result?

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Re: microscopic cross section calculation

Postby Jaakko Leppänen » Fri Sep 18, 2015 4:25 pm

The indexing is different in Serpent 2. With 2 energy groups down-scattering comes before up-scattering. You can get the data into correct matrix shape in Matlab with:

Code: Select all
octave:2> reshape(B1_S0(1:2:end), 2, 2)
ans =

   3.0871e-01   6.5381e-05
   8.6430e-03   4.3446e-01
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