Jaakko.

Thank you for fast response. I will try to use this card.

Volodymyr

## Search found 17 matches

- Wed Jan 31, 2018 2:08 pm
- Forum: Users
- Topic: Use primary neutrons for external source mode
- Replies:
**2** - Views:
**903**

- Tue Jan 30, 2018 5:56 pm
- Forum: Users
- Topic: Use primary neutrons for external source mode
- Replies:
**2** - Views:
**903**

### Use primary neutrons for external source mode

Hello. I calculate task for external source mode. There are fuel pins and detector in my geometry. I would like to know how many source neutrons (only neutrons from source) will initiate radiative capture in my detector. Is it possible in Serpent? Or could I obtain the coordinates of neutrons which ...

- Mon Aug 28, 2017 4:25 pm
- Forum: Development
- Topic: Equilibrium xenon in decay steps
- Replies:
**12** - Views:
**6005**

### Re: Equilibrium xenon in decay steps

Hello Jaakko. I have task to calculate the group constant generation for VVER-1000. Also, I do not have much experience in this type of simulations. In my case, I have the list of burnup points (for example, 1, 2, 3, 4, 5, 10, 20, 30). I need to calculate the branch with follow conditions: 1. Coolan...

- Tue Jul 14, 2015 9:34 am
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Thank you very much.

But I did not clearly understand next question:

Is it possible to obtain power in a certain cell for each step of the

burn-up (time-dependent) problem?

But I did not clearly understand next question:

Is it possible to obtain power in a certain cell for each step of the

burn-up (time-dependent) problem?

- Thu Jul 09, 2015 2:28 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Thanks. How can I obtain power in a certain cell for each step of the burn-up problem? Is it possible to use detectors for this purpose? And: If I understood correctly, the code (for example): det _Total_Capture dc _Some_Cell dr -2 void dt 3 _Total_Flux_in_Cell will produce the macroscopic capture x...

- Mon Jun 29, 2015 2:26 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Thank you. Can I get macroscopic cross sections for specific nuclides for burnup problems in certain cell using detectors? (I can find microscopic xsections and multiply them by nuclides concentrations on each burnup step, but I would like to know if it is possible to avoid these operations and get ...

- Fri Jun 19, 2015 5:01 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Thanks.

Can I get cross sections and scattering matrix for multigroup equations from detectors?

Can I get cross sections and scattering matrix for multigroup equations from detectors?

- Wed Jun 17, 2015 2:40 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Dear Jaakko.

Thank you for your answers.

Is it possible to obtain the energy spectrum which is used by Serpent for averaging of the cross sections for burnup calculations? Or can I only use detectors for this purpose?

Volodymyr

Thank you for your answers.

Is it possible to obtain the energy spectrum which is used by Serpent for averaging of the cross sections for burnup calculations? Or can I only use detectors for this purpose?

Volodymyr

- Wed Jun 03, 2015 12:41 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

Dear Jaakko.

Thank you for above information.

I have two additional questions: What is the method for calculating one-group transmutation cross sections for burnup matrix? What spectrum is used for cross section averaging?

Best wishes,

Volodymyr

Thank you for above information.

I have two additional questions: What is the method for calculating one-group transmutation cross sections for burnup matrix? What spectrum is used for cross section averaging?

Best wishes,

Volodymyr

- Fri Mar 20, 2015 2:51 pm
- Forum: Users
- Topic: One-group data for burnup calculations
- Replies:
**26** - Views:
**14428**

### Re: One-group data for burnup calculations

My ZAI vector remains the same during all burnup calculations. So I set pcc 0 and tried to calculate a few steps without using predictor-corrector method. I compared N0 and N1 from previous step and found atomic densities for the same nuclides were identical. As I understood the predictor-corrector ...