Setting temperature for a specific nuclide in OpenMC?

User Support
1

Hi everyone,
I need to set the temperature for a specific nuclide within the material. Any quick tips on how to do this, or do I need to prepare cross-sections for each nuclide at the desired temperature?

Thanks!

2

Hello.
My recollection is that the C++ source code forces all isotopes within a material to a common temperature. To implement a difference, you would need to prepare a dedicated cross section data table or fork the code and implement your change directly in the C++ code.

3

@AlexandreTrottier is correct – OpenMC does not allow you to set the temperature on a single nuclide. It is a property of a cell/material.

4

@AlexandreTrottier Thanks for the tip! Here’s what I did: got the cross-sections in ACE format for a nuclide at the desired temperature using NJOY, then manually changed the temperature in the ACE file. Even though they’re prepared at 500 K, the file specifies 300 K. After that, I converted them to H5 format. It’s a workaround, but it’s getting the job done for now.

@paulromano I believe having this feature in upcoming versions could open up possibilities for studying various phenomena, especially reactivity temperature coefficients.

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5

OK, that sounds like a good work around. @paulromano, I am not following your suggestion, since it seems to me that any coefficient study needs to have all isotopes in a cell go to the same temperature. Am I missing something?

6

Normally, setting the material temperature works for calculating any coefficient. However, if we aim to delve into each isotope’s contribution, it becomes essential to specify the temperature of each isotope in that material.

7

So the intent is to support a sensitivity analysis of the temperature coefficients. Interesting. I would appreciate if you could let me know how such a study progresses.

8

Absolutely! I’ll keep you posted. If you have any specific aspects you’re curious about feel free to let me know!

9

Thank you. I’m mostly curious about the design of numerical experiment side of such a sensitivity study. Are you interested in cross-term effects, types of sensitivity measures etc…

10

I’d be happy to discuss the details with you. If you prefer, we can continue this conversation privately. Please feel free to message me at your convenience.

11

I want to add a thought here. In addition to nuclide-level sensitivity study, there are some other situations in which it is beneficial to allow modeling of the same nuclide at different temperatures. Here are two examples:

  1. What if we like to model a complicated geometry like porous media, which can be considered as a mixture of different materials at different temperatures? Currently, I have not found a way to differentiate temperatures in a mixture like this. I appreciate it if you let me know such a capability in OpenMC.
  2. It is very common to do neutronics calculations with homogenized models, in which homogenized materials at different temperatures should be differentiated. Even though it is not commonly done in Monte Carlo simulation, it is also useful to quantify the homogenization effects with Monte Carlo methods directly to guide deterministic method development.

Such a capability could be realized in OpenMC without change to the data library by defining a new type of universe/material that holds traditional material components of individual volume fractions. During particle transport, the macroscopic interaction properties can be obtained by the statistical average of individual materials.

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12

I’m currently trying to perform depletion simulations of simple homogenized models where it’s an issue for me that the mix_materials doesn’t allow the specific elements/isotopes inheriting the temperatures of the constituent materials:

U10Zr_mat = openmc.Material(name=‘U10Zr_mat’,temperature = 900)
HT9_mat = openmc.Material(name=‘HT9_mat’,temperature = 600)
MgO_mat = openmc.Material(name=‘MgO_mat’,temperature = 600)

core_mat = openmc.Material.mix_materials([U10Zr_mat, HT9_mat, NaK_mat],[.48, .142, .378], name=‘core_mat’, percent_type = ‘vo’)

In the xml file, core_mat simply has no temperature assigned which is unfortunate.

13

Hi everyone,

We’ve just published a paper related to this topic:

Doppler effect evaluation of the TRU-based Moroccan TRIGA Mark II reactor: Applying novel methodologies

This paper might be interesting for those looking into new fuel cycle options or doing a sensitivity analysis. We introduced two methods: one to break down the fuel temperature coefficient using a four-factor formula, and another to see how each isotope and upscattering contribute to it.

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