# Custom source state.Box

**URL:** <https://openmc.discourse.group/t/custom-source-state-box/2566>\
**Category:** User Support\
**Created:** [December 14, 2022, 2:51pm UTC](https://openmc.discourse.group/t/custom-source-state-box/2566 "2022-12-14T14:51:11Z")\
**Posts on this page:** 2\
**Page:** 1

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**Author:** ![Openmc\_beginner](https://avatars.discourse-cdn.com/v4/letter/o/65b543/32.png) [@Openmc\_beginner](https://openmc.discourse.group/u/Openmc_beginner)\
**Post date:** [December 14, 2022, 2:51pm UTC](https://openmc.discourse.group/t/custom-source-state-box/2566/1 "2022-12-14T14:51:11Z")

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hello everyone,  
Recently I am trying to run a simulation with Triso particles and want to calculate the multiplication factor but I had a problem defying the initial neutron source. I tried to insert a point source at specific place in the geometry and state.Box source all over the core volume but from the mesh plots of fission and flux distribution I can see that a lot of places in my geometry doesn’t feel neutron and in every calculation I get different places feeling the flux not the whole geometry.

i’d love to get some suggestion how to maintain my source.  
thanks

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**Author:** ![paulromano](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/paulromano/32/486_2.png) [@paulromano](https://openmc.discourse.group/u/paulromano)\
**Post date:** [December 31, 2022, 5:06am UTC](https://openmc.discourse.group/t/custom-source-state-box/2566/2 "2022-12-31T05:06:10Z")

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When you’re running a k-eigenvalue calculation, the initial neutron source is arbitrary and will be iterated on until the source distribution reaches a converged state (at which point active batches can begin and tallies can be accumulated). In general, it would be better to start with a box source that covers your geometry rather than a point source (which will likely take more batches to converge). However, it sounds like the real issue in your case is perhaps not simulated enough neutrons / active batches, which may lead to poorly converged tally results. The first thing I would try in your case is to increase the number of neutrons per generation and the number of active batches and see if that helps.