Fission spectrum

[Clara Rojas]

I know that it's possible to calculate fluxes and intensities of neutron and gamma in the final steps of each cycle with ORIGEN. I'm trying to know the fission spectrum during irradiation and decay, but I'm just able to get the delayed and spontaneous fission neutron intensities. In addition, when I plot it with OPUS throught the "typarams=nspectrum" option, it shows me just SP, (alpha,n) and DN. So, would it be possible to know the fission spectrum at each cycle in ORIGEN? If so, how could I get it from ORIGEN?

Thanks!

[Shane Hart]

Hi Clara,

ORIGEN is a one-group code and, as such, has no idea of a "spectrum".  As input you're merely giving it a total flux and a cross section and it can output the reaction rate (which is where it's getting the neutron intensities).

You would need to have a group structure, with associated chi values, in order to get a meaningful energy distribution of the outcoming neutrons.  I think the real only way to do this is to run transport at each step with TRITON or something.

Regards,

Shane

[Ian Gauld]

Just to follow up on Shane's accurate reply. ORIGEN is developed to generate the delayed (decay) contributions to neutron and gamma emission only. Transport codes are generally used to generate prompt sources. However, having said that, ORIGEN tracks the fission rates of all actinides explicitly using data from transport calculations and these rates can be printed. Therefore the neutron emission during operation may be obtained using the fission rates and fission neutron spectra for each actinide and combining it with the delayed contributions (small) from ORIGEN. Some care is needed not to double account for delayed neutrons. If you use the prompt fission spectrum only then the delayed component could come from ORIGEN, however for these purposes it is likely best to use the total fission chi (from ENDF/B) and exclude the rate from ORIGEN since the ENDF/B values will be more accurate.  (A great addition to ORIGEN to add the neutron source from fission!)

[Clara Rojas]

Hi! Thank you for your reply Shane and Ian! So now it's clear that ORIGEN cannot print this info and I will have to look for the total fission chi distribution in the output of TRITON. 

At this point, I'm a bit lost about the unities of values in TRITON. Total fissions are printed out, but they are normalized per source particle, so I'm wondering how I could get the total number of neutrons from fission. 

On the other hand, if I would like to use the fluxes at each unit, they are normalized per source particle too. So, I suppose that I can use the total flux at each step (which is printed by TRITON) and then, get the n/s distribution for the unit. In addition, the printed fluxes are per unit, but when you have an array of these units, fluxes are referred for each unit in this array, so the total has to be the sum of them?

Thank you again.