He3 detector comparison to experimental results

User Support
1

Hello, I’m new to OpenMC, and I’ve got code running. I found this example of a He3 detector. I would like to compare this to experimental results from a neutron point source.

While the total neutron flux is important, the detector has some efficiency which may be unknown. My question is whether it’s possible to use openmc to compare directly to the experimental results? Perhaps with getting the reaction rate with He3 instead of just the neutron flux? Thank you!

2

That is a great example you found there :grin:

I think this PR on pulse height tallies might be of interest to you

3

Hi,
Can we also use these codes for gamma point sources?
Thank u

4

Yes gamma sources can be created and simulated.

Here is a minimal example for the same workshop

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5

Wow I’m getting help straight from the source! I’ve read through the PR, and I see the brief discussion on applying the pulse height tally to neutrons. They mention that ‘absorption’ tally should be sufficient for neutron detectors.

I do agree that pulse height would be very interesting, but since He3 detectors simply provide count and don’t give spectroscopic data, the absorption rate would give the same results as an experiment. Am I thinking about that the right way?

6

I think absorption is a nice improvement to using flux, I’ve just updated the workshop accordingly.

@cp-f might be keen to comment on your last question @nickschw

7

Can I plot an energy-count plot for a gamma source?

8

If I have understood correctly then I think a tally with an EnergyFilter would be the closest thing I know of in the code. I would still recommend that PR with pulse height tally but perhaps an EnergyFilter is part of the way there

Scripts 2, 3 and 4 show use of such a filter to get a spectrum.

neutronics-workshop/tasks/task_07_CSG_cell_tally_spectra at main · fusion-energy/neutronics-workshop · GitHub

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9

And I want to use two sources in my simulation. I tried to add the 2nd source in this way, but it didn’t work. How can I add the 2nd source?
multiplesource.py (791 Bytes)

10

I get this warning (" module ‘openmc’ has no attribute ‘TimeFilter’) when I want to run this code(neutronics-workshop/3_time_and_energy_filter_tally.py at main · fusion-energy/neutronics-workshop · GitHub). How can I solve it?

11

TimeFilter was introduced in openmc version 0.13.0. Perhaps you have an older version?

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12

Hi,

the pulse-height tally is only necessary for the simulation of gamma detectors. For He3 neutron detectors, the absorption tally is sufficient.

It probably makes sense to take a look at this:
https://www.researchgate.net/publication/349455477_Experimental_validation_of_a_He-3_neutron_detector_Monte_Carlo_simulation_model

The F4 tally of MCNP5 was used to obtain the neutron fluence rate (in neutrons per cm2-s), and the reaction rate (in reactions s−1) of the (n,p) reaction (Equation 1) which was estimated to be equal to the experimental neutron counts

I hope this helps a little.

Christopher

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