deplete.ReactionRates command

User Support
1

Hi,

I want to run the openmc.deplete.ReactionRates, but I get a syntax error as follow:

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As I am still new to openmc and there is no example on the official website, what is the correct way to execute the reaction rate command? Thanks.

2

@brianliu329,

The issue is firstly related to how you are calling the function. In python, keyword arguments must after positional arguments . So something like

funcion(a, b=5, c)

is not allowed but something like

function(a, 5, c)

might be.

It looks like you are passing reactions= as the second argument to the ReactionRates object, and a string results_filename as the final argument.

Are you trying to pull reaction rates from a result file generated from a depletion run? The ReactionRates class might not be what you want. If you’re trying to process a depletion file, the ResultsList class might be better suited.

The pincell depletion example has a much better explanation on how to extract reaction rates using the ResultList class.

Cheers,
Andrew

3

Hi Andrew,

Thanks for the reply. I followed the pincell depletion example to extract reaction rates from the ResultList. As I am simulating the transmutation of aluminium alloy, I guess I need to change the reaction rate to something else instead of having fission and I get the following error:

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May I ask what the third argument (rx) should be in my case?

In addition, since the material I am dealing with is an alloy, how do I add mutiple isotopes into the reaction rates? Thanks.

4

The third argument needs to be the name of a reaction that is tracked during depletion, e.g. (n,gamma), (n,2n), or fission. If you want to add up the reaction rates for all nuclides in a material, you’ll have to write a loop to add it up, something like:

nuclides = list(results[0].nuc_to_ind.keys())
reaction_rate = np.zeros(len(results))
for nuc in nuclides:
    time, rr = results.get_reaction_rate("1", nuc, "fission")
    reaction_rate += rr
5

Thanks Paul. May I ask how you define the “reactions” in the second line? I tried to put reactions = (n,total) and obviously it would not do anything. I then also tried to do openmc.data.Reaction as follow:

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which did not work.

Sorry if these are stupid questions.

6

My fault – that should have been results, not reactions. I’ve fixed it above. Sorry for the confusion!

7

Hi,

Thanks for the response. I now get an error and I am not sure what the reason is.

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I did not have a material identifier so I guess openmc automatically assigns it as 1. The nuclides here, I guess it is referring to the nuclides in my small chain:
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Do you know why I am only getting an ‘Al26’ error? Is it something possibly to do with my chain file?

8

Yes, I believe this is because some nuclides in the chain (such as Al26) don’t have associated neutron cross sections (and hence no reaction rate is tallied). Instead of using results[0].nuc_to_ind, if you use results[0].rates[0].index_nuc, it should limit it to only nuclides with neutron data:

nuclides = list(results[0].rates[0].index_nuc.keys())
9

It is good now. Thanks for all the helps Paul.