# How to simulate the neutron flux in the reactor core due to external neutron sources

Dear FLUKA experts,
I want to simulate the magnitude and distribution of neutron fluxes produced in the reactor core by several different external neutron sources.

The reactor uses UN as the fuel and BeO as the reflector layer.
The external neutron sources include: 1). spontaneous fission neutron sources, californium-252 (Cf-252).
2). antimony(Sb-124)-beryllium source: Sb-124 is a source of strong gamma rays emitter, while Be-9 is a neutron emitter element.

For 1), I tried BEAM card with an ISOTOPE particle, the HI-PROPE card defining Cf-252 (even the mass number in HI-PROPE card is limited to 247.) and the RADDECAY card to define the semi-analoge mode. However, the results from the USRBIN and USRTRACK card is still zero. For 2), the same zero results are produced.

So, I don’t know how to use FLUKA to simulate this process. Thanks in advance for further explanation.

Dear @chenchcc

Welcome to the forum!

Without knowing the details of your simulation and your objectives, I will try to address some points that might be relevant.

First of all, if your aim is to perform a reactor criticality study, then already the choice of code is questionable, since FLUKA does not run criticality problems. This is further compounded by the multi-group treatment of neutrons <20 MeV, although point-wise transport will become available with the next major release later this year. It is up to you to judge whether your objectives are compatible with the above limitations.

About 252Cf, indeed FLUKA does not generate the prompt fission neutrons, as far as I am aware. This could be addressed by directly generating neutrons from an appropriate spectrum via a SOURCE routine. The 252Cf neutron spectrum is well-documented and could be implemented without excessive effort. (As a side note, the A<247 limit in HI-PROPE is applied by Flair as a check on the input, but it is not actually enforced by FLUKA.)

Now to the Sb-Be source. I would suggest that also in this case the best way to go would be to directly generate an appropriate neutron spectrum via BEAM card or SOURCE routine, since the spectrum is almost monoenergetic around 23 keV, with a small component around 378 keV. As a second option, you could generate the principal 124Sb photons (1691 and 2091 keV, or just the 1691 keV), activating photonuclear reactions with a PHOTONUC card and biasing the photon interaction length in the BeO via a LAM-BIAS card. My guess is that you would not get enough neutrons inside the reactor for a study of reasonable duration, and little would be gained anyway, but you can try. Things will be even more dire if you use a 124Sb ISOTOPE source as primary (always with PHOTONUC and LAM-BIAS as above), given the low probabilities of the interactions (i.e. the appropriate decay gammas of 124Sb inducing a (γ,n) reaction in 9Be).

Finally, keep in mind that in order for any scoring to provide results when running a decay calculation (RADDECAY card in semi-analogue mode) you need one or more DCYSCORE cards properly associated to your scoring cards. Otherwise getting all zeros is the expected outcome.

Dear @atsingan ,

Thank you for the detailed explanation

I was trying to compare the flux of secondary neutrons produced in the core by several different external neutron sources, and to assess their suitability as a start-up neutron source for the reactor. I think it does not need to perform a criticality calculation.

Previously, I have evaluated the secondary neutron flux produced in the core using protons as external source. I think it is better to use the same software to calculate the results, so as to introduce unnecessary uncertainties.

I will try to learn to use the SOURCE routine to define the neutron source (neutron spectrum) produced by spontaneous fission of Cf-252 as well as the neutron spectrum produced by Sb-Be, as you suggested.

Also, can you say a little more about this statement? Is it already built in FLUKA? Where to find it?

Dear @chenchcc

What I meant concerning the 252Cf neutron spectrum is that there is plenty of literature on it, as a simple online search will reveal. See for example Phys. Rev. 157, 1076 (1967) - $^{252}\mathrm{Cf}$ Fission Neutron Spectrum from 0.003 to 15.0 MeV where the spectrum is fitted reasonably well with a Maxwellian. But there are many more parametrisations and measured spectra than you can use.

If you search this forum for e.g. “sampling” or “source routine” you can find explanations on how to set up a source routine to sample from a spectrum or a continuous distribution.

Dear @atsingan

Thank you for your quick response.

I am reading the docs on SOURCE routine and the examples given by other experts and users in this forum. Then I will try to write out these two external neutron sources.