Hello All,
I am new to the openmc code and modelling a fixed source simulation of an iron slab using 14.1 MeV neutron source at 10cm from the front of the slab. My tally is a mesh filter and cell filter and scored on flux. I will like to convert my results to units of particles/cm2-s. I have been following the openmc user guide but I don’t know the source rate to use it to multiply. How can I find out what the source rate is?
For example if i get a value of 0.0625961832063 for the above mesh and score. How can I convert it to paticles/cm2-s.
I will appreciate if you can guide me how to normalize this value to get flux values.
Thank you for your help and sorry if the answer to this question is too obvious.
@Pranto thank you for your swift responds. I can calculate the volume manually but I was wondering if it is possible to extract the mesh volume from the statpoint file directly and use it to divide.
Hi all !
Sorry to revive this discussion but I have doubt about the fixed source tally normalization. In 8.3. Normalization of Tally Results it is written:
As described in Scores, all tally scores are normalized per source particle simulated.
In the case of a fixed source of neutrons with 1e5 neutrons per batch, does it mean that the unit if the flux score is [ particle-cm/neutron] or [ particle-cm/(1e5 neutrons)] ?
I wonder after reading the answer of @Will_Boyd in an old discussion flux :
However, you must scale this the flux by the ratio of the # neutrons per batch to the # neutrons emitted per second from your external source in your problem.
Thank you in advance for any clarification,
Sabine.
Hi Sabine – the units of the flux score would be [particle-cm/neutron]. Changing the batch size will not change the expected value of the tallies. Of course, you’ll get better or worse answers depending on how many particles you simulate, but the expected value should always be the same.