In my simulation, I wanted to get the information regarding fission density. So, I used a SCORE card. But in the output, I got zero result, although I have resnucle output. I have attached here the output file. Can you please guide?

Also, if I want to get activity of radionuclides produced due to fission only, (i.e., excluding the contribution coming from the decay of another fission product, then I have to disable/omit RADDECAY, DECYTIMES and DCYSCORE, right?

What does NEU-BALA refer to? In the manual, it is written as net neutron production in each region. Is it same as scoring neutrons using USRBIN? Can we get information regarding how many fission neutrons are produced?

Noticing that you are using the pointwise treatment of neutrons, please not that unfortunately the scoring for fission processes are not yet fully implemented, and thus, you would be required to customize a user routine (e.g. mgdraw) in order to flag the fission neutrons or use other work-arounds. Would you be interested in just counting them or also to score more information?

As a temporary and simpler solution, you could revert to the group wise treatment of neutrons in FLUKA and user the SCORE card as you already do.

Indeed, if you do not wish to have radioactive decays, you should not use the RADDECAY, DECYTIMES and DCYSCORE cards.

The NEW-BALA according to the user manual is the Neutron balance (algebraic sum of outgoing neutrons minus incoming neutrons for all interactions). As such, it is not the same as scoring neutrons using USRBIN since the USRBIN would score all neutrons, and NEW-BALA would score only the difference (either positive or negative) between outgoing/incomig neutrons. Note that neutrons could be produced not just from fission phenomena (neutron hitting a larger target), but also due to other inelastic interactions.

Using the group wise treatment, I am able to get output in SCORE.
For e.g. considering 0.1 eV neutron and pure U235 target, the SCORE output gives 2.07 fissions/primary neutron.

Is it possible to get number of neutrons/fission as well from SCORE card? In the .out file, it is not showing any value.

Or for this mgdraw is required? If so, can you please mention the variable names for fission neutron?

Is fission star density per beam particle same as the number of neutrons per fission ? My intention was to get this value of nu and total number of fissions:

For a very large number of fissions of U-235 by thermal neutrons it is known that 2.7% of the fissions give no neutrons, 15.8% give one neutron, 33.9% give two, and so on. The average number of neutrons released per fission is denoted by v and has a value of 2.43 for thermal fission in U-235.

The star density denotes the number of fission events per primary. If you are interested in the number of neutrons per fissions, you can find in the output file a table as follows:

where for inelastic neutron you get \nu \approx 2.21 (for a simulation of 10 000 primaries). Would you be interested to obtain the number of neutrons for each fission event?

Thank you @dprelipc for the explanation. I expected that number of fission neutrons will be printed on ‘fission Neutron’ , which is printed as zero here. But as you mentioned nu is printed in inelastic neutron part. Also, as you asked, I think number of neutrons in each fission event can be obtained using = no of inelastic neutrons printed in .out / number of fission obtained from SCORE card. Is it right ?

You may have channels other than fission producing secondary neutrons that would also be counted, so your procedure may not be 100% correct.

My question is whether you be interested to obtain the number of neutrons for each fission event, which is not possible with existing/default scorings and a special user routine would be required.