Dear all,
I have a question concerning the scoring of neutrons produced from a proton beam and later been moderated. I want to score within a cell (e.g. water tank, or ICRU phantom) the dose from neutrons which are produced in (p,n) or (p,2n) reactions by a 68 MeV proton beam (source) in an iron target. The water tank let’s say is situated some distance behind the target. In a second set up, I will shield the water tank with lead and I want to score the neutron dose of the changed energy spectrum (after shielding)
My question is:
Is it sufficient to use F4:N card with proper DE and DF card to obtain the neutron dose after neutron reactions like (n,2n) or (n,3n) within the lead shield or do I have to use e.g. FM cards to score neutrons which are not directly produced by a source particle but from a secondary particle.
Many thanks in advance for any help!
Best regards,
Andreas Weber
If you are interested only in dose from specific reactions I think the FM card aproach is better. I would try to do the DE DF card combined with FM card.
If it not working than an energy tally (E card) (with the same bin set as the DE DF has) and combine it with an FM card than later multiply the obtained flux with the conversion factors of the DF card and integrate it.
Thank you Gábor for your very fast reply. In fact I am interested in the total neutron dose coming from all neutrons: secondaries, secondaries of secondaries and so on. I was only afraid that a MCNPX default would only track the first (unscattered) generation of neutrons and not the next ones. So I understand in default I would score with F4:N all neutrons, right? Thank you again and best wishes, Andreas
But be sure you are useing cross sections with sufficiently high maximum energies and/or model physics (mphys:on). The transport of other possible secondary charged particles (which can produce neutrons later on (D, T, He...)) is swithed on with the mode card and photo-nuclear processes switched on with the phys:P card.
Otherwise you underestimate the neutron flux and therefore the dose as well.
Oh, this is a very good remark. I will check it tomorrow. Actually I have the impression that I have to few thermalized neutrons what might be explained by the mphys:on, which I ignored so far and the other secondaries which I didn't switch on yet. Best regards, Andreas
P. S. The cross section I choose within the material card by the proper ZAID extension?
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