Question on writing tallies.xml

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1

Hi everyone,

I want to tally the number of neutrons escaping from each vacuum boundary (for neutrons born at each cell). But it seems the surface tally is not supported. Then I planned to tally the number of neutrons absorbed at each cell (for neutrons born at each cell) and then tally the total number of neutrons born at each cell. Which ‘scores’ should be used with the filter ‘cellborn’ to get the total number of neutrons born at each cell?

I also want to tally the distribution of the number of fissioned neutrons per absorption at each cell. Is it possible to implement this with tallies.xml without changing the source code?

Thanks!
Jilang

2

Hi Jilang,

  • neutron leakage as function of birth cell – you would have to hardwire it in the code
  • number of neutrons absorbed in each cell per birth cell – if you use a cell filter and a cellborn filter and then tally absorption, that will give you a matrix of absorption rates
  • total number of neutrons born at each cell – this isn’t possible with tallies, but you could write a source file every batch and figure it out that way
  • distribution of number of fission neutrons per absorption at each cell – I’m not sure if I understand this exactly. If you want the average number of fission neutrons per absorption in each cell, that is straightforward with a cell filter and nu-fission and absorption scores.
    If this is for something “research-y”, your best bet is to just bite the bullet and add some global variables to make your own tallies.

Best,

Paul

3

Hi Paul,

Thanks for the quick response and the suggestions.
By "distribution of number of fission neutrons per absorption at each cell", I mean the probabilities P(nu=0),P(nu=1),P(nu=2),…
It seems I have to touch the source code for this tally too.

Thanks!
Jilang

4

Yes, for that you’d definitely need to touch the source, but it would be pretty straightforward. Let me know if you need any particular help.

Best,
Paul