Unable to load chain_endfb71.xml into depletion code

User Support
1

Hello,

I am trying to run a depletion code example from John Shimwell’s neutronics-workshop repo, but am unable to load the xml file into the depletion operator. I used the generate_endf71_chain.py from the openmc-dev/data repo, and saved the xml file in the same path as my notebook.

#Chain file setup
chain_filename = 'chain_endfb71.xml'
chain = openmc.deplete.Chain.from_xml(chain_filename)

operator = openmc.deplete.Operator(model, chain_filename)

#1e9 neutrons per second for 5 years then 5 years of no neutrons (shut down cooling time)

time_steps = [365*24*60*60] * 5 + [365*24*60*60] * 5
source_rates = [1e9]*5 + [0] * 5

integrator = openmc.deplete.PredictorIntegrator(
    operator=operator, timesteps=time_steps,source_rates=source_rates
)

integrator.integrate()

---------------------------------------------------------------------------
OSError                                   Traceback (most recent call last)
Input In [12], in <cell line: 5>()
      2 chain_filename = 'chain_endfb71.xml'
      3 chain = openmc.deplete.Chain.from_xml(chain_filename)
----> 5 operator = openmc.deplete.Operator(model, chain_filename)
      7 #1e9 neutrons per second for 5 years then 5 years of no neutrons (shut down cooling time)
      9 time_steps = [365*24*60*60] * 5 + [365*24*60*60] * 5

File ~/.mambaforge3/envs/openmc/lib/python3.9/site-packages/openmc/deplete/operator.py:200, in Operator.__init__(self, geometry, settings, chain_file, prev_results, diff_burnable_mats, normalization_mode, fission_q, dilute_initial, fission_yield_mode, fission_yield_opts, reaction_rate_mode, reaction_rate_opts, reduce_chain, reduce_chain_level)
    198         warn("Fission Q dictionary will not be used")
    199         fission_q = None
--> 200 super().__init__(chain_file, fission_q, dilute_initial, prev_results)
    201 self.round_number = False
    202 self.settings = settings

File ~/.mambaforge3/envs/openmc/lib/python3.9/site-packages/openmc/deplete/abc.py:113, in TransportOperator.__init__(self, chain_file, fission_q, dilute_initial, prev_results)
    111             break
    112     else:
--> 113         raise IOError(
    114             "No chain specified, either manually or "
    115             "under depletion_chain in environment variable "
    116             "OPENMC_CROSS_SECTIONS.")
    117     chain_file = lib['path']
    118 else:

OSError: No chain specified, either manually or under depletion_chain in environment variable OPENMC_CROSS_SECTIONS.

The example I looked at initially had chain_filename = '/chain_endfb71.xml' , but I got the following error:

image
image1068×893 137 KB

I also tried chain_filename = './chain_endfb71.xml', but got the following error:

---------------------------------------------------------------------------
OSError                                   Traceback (most recent call last)
Input In [11], in <cell line: 5>()
      2 chain_filename = './chain_endfb71.xml'
      3 chain = openmc.deplete.Chain.from_xml(chain_filename)
----> 5 operator = openmc.deplete.Operator(model, chain_filename)
      7 #1e9 neutrons per second for 5 years then 5 years of no neutrons (shut down cooling time)
      9 time_steps = [365*24*60*60] * 5 + [365*24*60*60] * 5

File ~/.mambaforge3/envs/openmc/lib/python3.9/site-packages/openmc/deplete/operator.py:200, in Operator.__init__(self, geometry, settings, chain_file, prev_results, diff_burnable_mats, normalization_mode, fission_q, dilute_initial, fission_yield_mode, fission_yield_opts, reaction_rate_mode, reaction_rate_opts, reduce_chain, reduce_chain_level)
    198         warn("Fission Q dictionary will not be used")
    199         fission_q = None
--> 200 super().__init__(chain_file, fission_q, dilute_initial, prev_results)
    201 self.round_number = False
    202 self.settings = settings

File ~/.mambaforge3/envs/openmc/lib/python3.9/site-packages/openmc/deplete/abc.py:113, in TransportOperator.__init__(self, chain_file, fission_q, dilute_initial, prev_results)
    111             break
    112     else:
--> 113         raise IOError(
    114             "No chain specified, either manually or "
    115             "under depletion_chain in environment variable "
    116             "OPENMC_CROSS_SECTIONS.")
    117     chain_file = lib['path']
    118 else:

OSError: No chain specified, either manually or under depletion_chain in environment variable OPENMC_CROSS_SECTIONS.

Curious to see where I went wrong here, should the xml file be saved somewhere in the openmc root folder?

Thanks,
Myles

2

Hi Myles,

have you tried running your code using one of the depletion chains you can download from the OpenMC page? (Depletion Chains | OpenMC) You can download the chain_endfb71.xml file from here.

best,

Luiz

3

Hi Luiz,

Thanks for your suggestion! I gave the endf_pwr and endf_sfr chains a try, and got the same error… with and without using . or / in the name.

Best
Myles

4

Myles, could you please send me the model you’re using, so I can run it here to see what will happen.

5

Hi Luiz,

Sorry for the delay, here’s a link to the example that I am following (for a sanity check I have even copied and pasted everything to make sure my implementation wasn’t incorrect, but I still get the same errors)

Best,
Myles

6

@myless I think the problem is that you are using syntax for OpenMC version 0.13.0 (notably, passing a Model object to Operator(...)) but you are actually using version 0.12.2 or earlier (which expects a Geometry and Materials object. I would recommend updating your OpenMC installation to 0.13.0 and then trying again.

7

Hi Paul,

Thanks for your reply. The newest version of OpenMC fixes this!

Best,
Myles

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