## microscopic cross section calculation

### microscopic cross section calculation

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.

Best Regards

Best Regards

- Jaakko Leppänen
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### Re: microscopic cross section calculation

See the examples in Sec. 7.1.1 of Serpent Manual.

- Jaakko

### Re: microscopic cross section calculation

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 ?

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 ?

- Jaakko Leppänen
- Site Admin
**Posts:**2440**Joined:**Thu Mar 18, 2010 10:43 pm**Security question 2:**0**Location:**Espoo, Finland-
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### Re: microscopic cross section calculation

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

- Jaakko

### Re: microscopic cross section calculation

Thanks a lot.

Can I actually get the microscopic scattering cross section matrix ??

Can I actually get the microscopic scattering cross section matrix ??

- Jaakko Leppänen
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### Re: microscopic cross section calculation

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?

Best regards

Motalab

Best regards

Motalab

- Jaakko Leppänen
- Site Admin
**Posts:**2440**Joined:**Thu Mar 18, 2010 10:43 pm**Security question 2:**0**Location:**Espoo, Finland-
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### Re: microscopic cross section calculation

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

- Jaakko

### Scattering matrixes order

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?

Thanks

Motalab

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?

Thanks

Motalab

- Jaakko Leppänen
- Site Admin
**Posts:**2440**Joined:**Thu Mar 18, 2010 10:43 pm**Security question 2:**0**Location:**Espoo, Finland-
**Contact:**

### Re: microscopic cross section calculation

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
```

- Jaakko