Hi; How can we add the value of "source intensity’’ in setting in an OpenMC program? Thank you.

The intensity of a source is given as the strength attribute of the Source class. For example, if you want a source with an intensity of 10: src = openmc.Source(..., strength=10.0) or src = openmc.Source(...) src.strength = 10.0 You can think of the strength as a global tally multiplier.

Hi Paul; Thanks a lot, i appreciate your help. Younes.

Hi @paulromano, Is this value in source particles/s? Maria

Hi @Mary; Yeah I think so cuz the source intensity means the emmision rate, so the unit is particle/s. Younes.

The units are actually arbitrary. Normally tallies are given “per source particle”. If you treat your source strength in source particles/sec, then this results in tallies with units of “per sec” since the tallies are just multiplied by the source strength. However, if you consider your source strength to be something else, for example source particles/year, then you would get tally results “per year”.

Hi @paulromano. My question is because I would like to understand how to use the strength attribute to convert the flux units. The documentation says: “As described in Scores, all tally scores are normalized per source particle simulated. However, for analysis of a given system, we usually want tally scores in a more natural unit. For example, neutron flux is often reported in units of particles/cm^2-sec. For a fixed source simulation, it is usually straightforward to convert units if the source rate is known. For example, if the system being modeled includes a source that is emitting 10^4 neutrons per second, the tally results just need to be multiplied by 10^4. This can either be done manually or using the openmc.Source.strength attribute”. I have always used the first option. I mean, I get the flux in ‘particle-cm/source particle’ (flux obtained with strength = 1 – default value), multiply this flux by the source intensity in ‘source particle/sec’ and divide by the volume (cm^3) corresponding to the tally result to obtain the flux in ‘particle/cm^2-sec’ (see post with the equation). If I properly understood the statement in the documentation, in the case that I want to use the second approach, I have to do the following: openmc.Source.strength = 10^4 which will give us the flux in particle-cm/sec. Then dividing the flux tally by the volume [cm^3], we obtain the units ‘particle/cm^2-sec’. In other words, the Source intensity variable in the mentioned equation is not required, is this correct? Thanks, Maria

Maria – yes, that’s correct. If you are running a fixed source simulation and you specify source = openmc.Source() source.strength = 1e4 ... Then all tallies will be multiplied by 1e4.

Hi all, Just want to ask a follow up question about how this applies to problems with multiple sources. For instance, if you had two identical sources, each with source_strength = 1, would you get tallies that are multiplied by two? Or would OpenMC do some sort of normalization? And if there is normalization, how would that extend to problems with many sources of varying strengths? For a problem with multiple sources, would I essentially need to have all my source_strengths add up to a total of 1 to get tallies comparable to a problem with only a single source with source_strength = 1?

Hello, I think the post below can help you with your question. [image] Photon flux with NaN value User Support Hello all, I am having an issue in a shielding simulation. My model consists of different cylinders, each one filled with a different material (see figure). In the center, I have positioned two point sources, one neutron source and the other is a photon source. As I have both energy spectra, I am using openmc.stats.Tabular to inform them (following the recommendations in this post). I wish to tally the fluxes, then I created a mesh (101 x 101 elements) and particle filters. [image] The mesh h… Maria

Source intensity

User Support

MaryAAZ April 20, 2022, 3:18am 11

Hello,

I think the post below can help you with your question.

Maria

Topic		Replies	Views
Source strength for multiple sources environment User Support	4	168	July 31, 2025
Converting OpenMC Flux Unit to Match MCNP Validation and Verification	3	252	March 16, 2025
Source strength for a fixed source User Support	1	160	January 26, 2025
Source Strength User Support	3	537	April 29, 2022
Fixed source tally normalisation User Support	7	1272	January 7, 2022

Source intensity

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