Energy deposited by neutrons

User Support
1

Hi all,

I was wondering if it could be possible to measure the energy deposited in a cell by neutrons. I have been trying with different MT numbers {like 301} and they don’t appear to be included with OpenMC. What would you suggest me to do?

Thank you.

2

At the moment, the only tally for energy deposition we have is for fission energy deposition (through the kappa-fission, fission-q-prompt, and fission-q-recoverable scores). However, adding a more general capability is a relatively high priority (at least for me personally), so I’m hoping we’ll have something implemented by the next release. If you have specific use cases you’d like supported, feel free to let us know!

Thanks,
Paul

3

Any updates regarding this function? Being able to get neutron energy deposited in i.e. scatterings would be a very welcome feature.

onsdag 29. august 2018 16.32.05 UTC+2 skrev Paul Romano følgende:

4

Hi Julian,

Two new scores were added in OpenMC 0.11, “heating” and “heating-local”. They both correspond to total nuclear heating (what is produced by NJOY’s HEATR module) and will include energy deposited from neutron scattering, but also from other reactions as well. The difference between “heating” and “heating-local” is how they treat secondary photons – “heating-local” assumes that the energy from secondary photons is deposited at the collision site (i.e., the energy is not carried away). Thus, “heating-local” may be a good option if you are not running a coupled neutron-photon calculation. In a coupled calculation, the normal “heating” score may be more appropriate because it’s also possible to get gamma heating from the photon transport calculation.

To be clear, it is not currently possible to get heating only from elastic scattering (MT=302 produced by NJOY/HEATR).

One final note – not all HDF5 nuclear data libraries have the data required for these two new scores. If your data is produced using OpenMC’s Python API (IncidentNeutron.from_njoy), it will have the data, but data converted from other sources may not. Our official ENDF/B-VII.1 library (found here) does have the required data.

Best regards,
Paul