Search found 21 matches

by Volodymyr Gulik
Thu Jul 08, 2021 9:22 am
Forum: Users
Topic: 'Sum' for Automated burnup sequence
Replies: 5
Views: 426

Re: 'Sum' for Automated burnup sequence

Thanks for explanation. Do you have any ideas how we can fix it?

Volodymyr
by Volodymyr Gulik
Wed Jul 07, 2021 10:02 pm
Forum: Users
Topic: 'Sum' for Automated burnup sequence
Replies: 5
Views: 426

Re: 'Sum' for Automated burnup sequence

Ana, thank you for your explanation. We have tried to use your suggestion. But it seems that we still have same error. There is very simple input file for our task: % ============================================================================ % --- Normalization to total fission power: [Wt] = 3000 ...
by Volodymyr Gulik
Thu Jul 01, 2021 4:02 pm
Forum: Users
Topic: 'Sum' for Automated burnup sequence
Replies: 5
Views: 426

'Sum' for Automated burnup sequence

Hello. It seems that in versions 2.1.31 and 2.1.32, we cannot use the 'sum' input when setting <dens> stp for the branch card (there is error: sum over composition cannot be used with adjustment), despite the fact that this was possible in previous 2.1.30 version. At the same time, it is written on ...
by Volodymyr Gulik
Fri Nov 20, 2020 10:44 am
Forum: Gamma transport
Topic: Comparison of Serpent and XCom results for photon transport
Replies: 0
Views: 341

Comparison of Serpent and XCom results for photon transport

Good day. Our Serpent calculations (version 2.1.31) for several types of concrete provide mass attenuation coefficients that are slightly but systematically smaller then that determined from XCOM, within 0,7%. Such deviations were observed for the energies 1.33 МеV, 2.75 МеV and 3.5 МеV; for 0.66 Ме...
by Volodymyr Gulik
Wed Jan 31, 2018 2:08 pm
Forum: Users
Topic: Use primary neutrons for external source mode
Replies: 2
Views: 1317

Re: Use primary neutrons for external source mode

Jaakko.

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

Volodymyr
by Volodymyr Gulik
Tue Jan 30, 2018 5:56 pm
Forum: Users
Topic: Use primary neutrons for external source mode
Replies: 2
Views: 1317

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 ...
by Volodymyr Gulik
Mon Aug 28, 2017 4:25 pm
Forum: Development
Topic: Equilibrium xenon in decay steps
Replies: 12
Views: 7648

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...
by Volodymyr Gulik
Tue Jul 14, 2015 9:34 am
Forum: Users
Topic: One-group data for burnup calculations
Replies: 26
Views: 16898

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?
by Volodymyr Gulik
Thu Jul 09, 2015 2:28 pm
Forum: Users
Topic: One-group data for burnup calculations
Replies: 26
Views: 16898

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...
by Volodymyr Gulik
Mon Jun 29, 2015 2:26 pm
Forum: Users
Topic: One-group data for burnup calculations
Replies: 26
Views: 16898

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 ...