## Triton depletion code

Hi

I'm wondering if anyone has experience working with the triton depletion code for SCALE? I'm having some issues when I'm trying to calculate the activity in radioisotope targets surrounding my reactor core.
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 What is your issue? Also, have you tried posting a question on the SCALE notebook.
 I have visited the notebook forum, but the Triton section has not had a new post in over two years, so I'm not sure if is still active. I want to determine the activity of mo-99 in uranium targets surrounding a reactor core. But I'm unsure what input parameters to choose for the burn-up scheme. I'm not sure what to choose power depletion or flux depletion, or if they are normalised or not. The second issue how to determine the neutron flux in the core. Triton gives the average flux of the core, and is thus not that helpful in determining the best positions for the uranium targets.

## Triton depletion code

 Quote by vifteovn I have visited the notebook forum, but the Triton section has not had a new post in over two years, so I'm not sure if is still active. I want to determine the activity of mo-99 in uranium targets surrounding a reactor core. But I'm unsure what input parameters to choose for the burn-up scheme. I'm not sure what to choose power depletion or flux depletion, or if they are normalised or not. The second issue how to determine the neutron flux in the core. Triton gives the average flux of the core, and is thus not that helpful in determining the best positions for the uranium targets.
TRITON calculates spatially dependent fluxes using KENO or NEWT to pass to ORIGEN-S. To determine the best location to place the target, it would be simpler to create a plain KENO or NEWT model first since the TRITON output can become incomprehensibly lengthy.

The difference between the depletion modes is described in the Depletion Block section of the SCALE TRITON manual. The part that is normalized is the determination of total power which is used to calculate local fluxes. If you specify an average assembly power, the local flux is normalized. If you want to specify the flux or power of an individual pin, then the assembly power is normalized to the flux you specify.

 Quote by QuantumPion TRITON calculates spatially dependent fluxes using KENO or NEWT to pass to ORIGEN-S. To determine the best location to place the target, it would be simpler to create a plain KENO or NEWT model first since the TRITON output can become incomprehensibly lengthy. The difference between the depletion modes is described in the Depletion Block section of the SCALE TRITON manual. The part that is normalized is the determination of total power which is used to calculate local fluxes. If you specify an average assembly power, the local flux is normalized. If you want to specify the flux or power of an individual pin, then the assembly power is normalized to the flux you specify.
When calculating the flux in KENO do you mean by the help of KMART?

I tried to use NEWT, but it was not happy when I tried to run the model with 150 fuel assemblies, and it crashed.. Seems I can only make NEWT work with one-few fuel bundles, and that does not exactly help me place targets..

 Quote by vifteovn When calculating the flux in KENO do you mean by the help of KMART? I tried to use NEWT, but it was not happy when I tried to run the model with 150 fuel assemblies, and it crashed.. Seems I can only make NEWT work with one-few fuel bundles, and that does not exactly help me place targets..
You don't need to model the whole core with every single fuel pin explicitly, no personal computer will handle that in a reasonable time. You need to generalize the problem to meet your goals. If you only care about the flux outside the core you probably only need to model one or two rows of fuel on the periphery adjacent to the target.

KMART is one way to graphically show the results but you can just look at the relevant output edits for NEWT or KENO as well.
 Don't you need the whole core to get the correct flux? Is there a way to get the outputfile to read neutrons/cm^2 s or will it only give up the relative flux?

 Quote by vifteovn Don't you need the whole core to get the correct flux? Is there a way to get the outputfile to read neutrons/cm^2 s or will it only give up the relative flux?
Only if you didn't know the local assembly powers and had to model it. In which case you should use a cell-mixing function to approximate the core. Otherwise you can specify local assembly power and boundary conditions.

You'll have to look at the examples in the manual for output options, I don't know them off the top of my head.
 I can determine the activity of mo-99 in uranium targets for you by MCNPX2.6 or COMB (Coupled MCNP-ORIGEN burnup code system).

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