question about fixed source

User Support
1

Niuoba sent a message to the user’s group (reposted below) but I fear many people (including myself) may not have noticed it because the attachment contained an exe file leading it to be blocked. I’ve reposted it below along with my response.

  1. Fixed source simulations with a source from a file will only work with the current developmental branch of OpenMC. When the next version 0.7.1 is released, this capability will be included.
  2. The following should get you the scores you need:
    nu-fission n2n n3n n4n absorption
    Note that leakage is automatically given as a global tally.
  3. nu-fission gives you total neutron production (prompt and delayed). If you want delayed only, you should use the new delayed-nu-fission score which will appear in version 0.7.1 (or in the current developmental branch).
  4. There are plans for burnup but there is no specific timeline at this point. Many of the OpenMC developers will be meeting next week and hopefully we will have a better idea then.
    Hope this helps; let me know if you have further questions.

Best,
Paul

2

Thank you Dear Prof.Paul,
I got a lot of message from you.
Many thanks for your kindly help and information.
I will do some practice, and absorb these messages.
Your advise always helps me a lot :slight_smile: .

3

Just wondering

If nu-fission gives you total neutron production is this then the same as Endf reaction MT 201 (Total neutron production) as used in derived files

Thanks Jon

4

If I understand it correctly, the answer is no. MT=201 would be the sum of (cross section * neutron yield) over all reactions that produce a neutron except for elastic scattering. So, it would include contributions from reactions like (n,2n). nu-fission specifically refers to neutron production only from fission.

Best,
Paul

5

Thanks for that Paul

I will try to add mt 201 to the plotter.py at some point

best

Jon