Volume neutron source information

Dear fluka experts,

I want to record the neutron source information, as mention here and I appreciate for selfless information. After several times try to use usryield to record the neutron source, I always got the ring source shape due to the setting of region to region. I might wrong for using usryield as an estimator to recording the information because the expected result of neutron source information is to cover all over my geometry by volume, not just surface. Will using USRTRACK help to record the neutron in the volume region or I have to set emerging in USRYIELD card? but USRTRACK will not have direction info right? Please correct me.

Question 2:
According to this paper,
My working model is similar to Fig 1. I want to record the neutron source same as this paper. Does it assign many USRYIELD cards at every intersect boundary? Is there any suggestion on how to do the same as this paper?

Hello @thanapong,

I will need you to further clarify what you want to obtain to try to provide an optimal solution.

In your previous question you attached an image with a fraction of a paper saying “ …The information on spallation neutrons passing through the target surface into the core is recorded … ” and you mentioned that you were following that paper. If the neutron positions are recorded when passing through the surface, they will be at the surface by definition, so you cannot get them all over your volume.

Neutrons will be moving through your geometry, and you need to choose when to obtain their properties (e.g. position), since they will obviously be in constant change.

Maybe what you want to do is to score their properties at the moment when the neutrons are created, this way you will obtain them distributed in the volume where they are created.

A completely different thing is what is shown in Fig. 2 and 3. of the paper you mention in this post. As stated in the paper they are fluxes shown as function of neutron energy and solid angle. They are most probably obtained using a simple USRBDX scoring. Below a presentation that may help you to understand how to use this built-in scoring:

I would appreciate if you can comment further on what do you want to obtain, but focussing on the physical meaning rather than if it should look as a volume or a surface.

Kind regards,

Dear @fogallar,
Thank you for replying, sorry for the unclear explanation. My work on fluka is the simulation of impinging the PbBi cylindrical target with a 60MeV proton beam and obtain the neutron source distribution that incident from the proton. My Geometry is the same as refer to the paper’s, with the target zone and subcritical core.

From your reply paragraph 3, “Maybe what you want to do is to score their properties at the moment when the neutrons are created, this way you will obtain them distributed in the volume where they are created.” is totally corrects the idea I want to try. Because I have tried using usryield with target geometry (attach below) and surround it with “RCC” and assign “vacuum” material as your advice here. After I used it coupling to openmc, it shows the fission rate become zeroes, so I think it might be something wrong, and I want to try creating more complex geometry on fluka and scoring by the neutron in volume of geometry instead of surface.

this picture is before surround with “RCC”

Thank you in advance,

Hi there,

I will assume then that you want to score your neutrons at the moment they are created but please note that this is not what is explained in the paper you were referring to in your initial question.

To score the neutrons at the moment they are created I would use the mgdraw.f routine, in particular the USDRAW entry. There you can implement your logic to score the neutrons just after the reaction that generates them, maybe this post could give you some hints.

But, if you plan to use these neutrons as source for a second simulation you should be careful or you could end up counting some of them more than once. For instance, imaging you score the properties of one neutron, but you let it continue running in your initial simulation, it may give rise to the creation of more neutrons, and you would also count them, which would be wrong if then you use all of them in your second simulation (you would simulate twice the “life” and products of the first neutron). To avoid this, you should discard the neutrons after you “score” them, for instance by zeroing their statistical weight.

Note that if you do that then you shouldn’t score anything else during your simulation, at least not blindly because you would be discarding an important part of it, the neutrons.

Kind regards,