# Error in Destination Material

**URL:** <https://openmc.discourse.group/t/error-in-destination-material/4695>\
**Category:** User Support\
**Created:** [August 16, 2024, 8:41am UTC](https://openmc.discourse.group/t/error-in-destination-material/4695 "2024-08-16T08:41:53Z")\
**Posts on this page:** 3\
**Page:** 1

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**Author:** ![a\_dwijayanto](https://avatars.discourse-cdn.com/v4/letter/a/5fc32e/32.png) [@a\_dwijayanto](https://openmc.discourse.group/u/a_dwijayanto)\
**Post date:** [August 16, 2024, 8:41am UTC](https://openmc.discourse.group/t/error-in-destination-material/4695/1 "2024-08-16T08:41:53Z")

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Hello,

I encountered a problem when attempting to use transfer rate feature in OpenMC. I wanted to removed some fission products from one material (“fuel”) and store it in another material (“fp\_tank”) for tracking. So I wrote the material inputs more or less like this,

```auto
fuel = openmc.Material(name="fuel") 
fuel.add_nuclide('U235', 1.58775E-03)
fuel.add_nuclide('U238', 6.45147E-03)
fuel.add_nuclide('Th232', 4.09804E-02)
fuel.add_nuclide('F19', 6.07843E-01)
fuel.add_nuclide('Be9', 6.86275E-02)
fuel.add_nuclide('Li6', 1.37255E-05)
fuel.add_nuclide('Li7', 2.74496E-01)
fuel.set_density('g/cm3',3.3)
fuel.volume = 24594900.57
fuel.temperature = 900

```

```auto
fp_tank = openmc.Material(name="tank")
fp_tank.add_element('He', 1)
fp_tank.set_density('sum')
fp_tank.temperature = 327

```

Then I determined the power, removed nuclide, etc.

```auto
removed_nuclide = ['H','He','N','O','Ne', 'Ar','Kr','Nb','Mo','Tc','Ru','Rh','Pd', 'Ag','Sb','Te','Xe','Lu','Hf','Ta','W','Re','Os','Ir','Pt','Au','Rn']
dest_mat = 'fp_tank'

integrator = openmc.deplete.PredictorIntegrator(operator, time_steps, power, timestep_units='d')
integrator.add_transfer_rate(fuel,removed_nuclide, 3.333e-2, '1/s', destination_material=dest_mat)

```

However, this didn’t work as it shows error message, "ValueError: Unable to set “Material name” to “fp\_tank” since it is not in “[‘fuel’]” "

Is it because I place the fp\_tank on the outside of the reactor core boundary or something else?

Thanks.

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**Author:** ![wahidluthfi](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/wahidluthfi/32/3375_2.png) [@wahidluthfi](https://openmc.discourse.group/u/wahidluthfi)\
**Post date:** [August 17, 2024, 2:44am UTC](https://openmc.discourse.group/t/error-in-destination-material/4695/2 "2024-08-17T02:44:15Z")

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Hi Dwijayanto, welcome to the community.  
I think you want to declare destination material into the fp\_tank material named ‘tank’, so it should look like `dest_mat = fp_tank # or their name 'tank'` instead of dest\_mat = ‘fp\_tank’

Also, have you set that material into depleteable and set its corresponding cell volume? I think this material should exist on the openmc geometry model since openmc.deplete remake of the xml file used for depletion calculation.

But I think the experts will comment if this error came from the same problem encountered in this discussion

> [@Transfer rates warning for destination material](https://openmc.discourse.group/t/transfer-rates-warning-for-destination-material/3054/2):
>
> Hi @LukeSeifert, thanks for reporting this. The warn message you’re seeing is definitely misleading. What is set to 0.0 is not the transfer rate of the element you’ve defined but for all other elements present in the material, for which, as expected, a null transfer rate gets set. Don’t even know why is there and will be removed. Note however that the transfer rate for Xenon from material 1 to 2 gets set correctly as you’ve defined as the results you’ve plot suggests. Thanks again. Lorenzo

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<div class="post-metadata">

**Author:** ![a\_dwijayanto](https://avatars.discourse-cdn.com/v4/letter/a/5fc32e/32.png) [@a\_dwijayanto](https://openmc.discourse.group/u/a_dwijayanto)\
**Post date:** [August 20, 2024, 2:52am UTC](https://openmc.discourse.group/t/error-in-destination-material/4695/3 "2024-08-20T02:52:59Z")

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Thanks, I think that resolves it.