Create a depletion chain .xml file.

User Support
1

Hello everyone!

I have recently started using OpenMC. I have installed the 11-dev version, for running depletion calculations and I have used the simple-chain.xml file that came with the code. I would like to include some additional nuclides in the depletion chain, but I have not been able to understand how to create a new depletion chain using the python API. The documentation online is a bit too. Does anyone have some example code on how to do this?

Best Regards,
Sindre

2

Hi Sindre,

There are two scripts in the OpenMC repository for generating depletion chains. One is called openmc-make-depletion-chain and will construct a full depletion chain based on all available data in ENDF/B-VII.1. That is probably overkill for most applications since it will result in the chain having 1000s of nuclides. A more reasonable option is to use openmc-make-depletion-chain-casl which generates a depletion chain with 200-300 nuclides and is based on the chain that was developed in the CASL/VERA project. A few special notes that you should consider in getting accurate results:

  • By default, OpenMC will use the reaction Q values in the chain file for determining energy deposition. Since these don’t account for indirect energy deposition (from things like capture reactions), the Q values tend to be too small. This would result in a core depleting too fast since a higher reaction rate is needed to compensate for the lower Q value. You can set the Q values directly with the fission_q argument to Operator.
  • The CASL chain doesn’t include information about capture branching ratios. These can be modified by using the Chain.set_capture_branches method that was recently added.

For the next release, we will likely post a pre-generated depletion chain at openmc.org so that users don’t have to worry about this. Until then, let us know if you have questions.

Best regards,
Paul

3

Hi everyone,
I wish to understand how to generate a depletion chain. For this purpose, I downloaded from the National Nuclear Data Center the sub-libraries: Neutron Reaction, Neutron Induced Fission Product Yields, and Decay Reaction. I have them in three different folders along with the attached script to try to generate the depletion chain. When executing the script (not sure if it is correct), I am getting errors:

File “/home/javier/Documents/OpenMC/NuclearData/ENDFB7.1_Chain/create_dep_chain.py”, line 37, in
chain = openmc.deplete.Chain.from_endf(decay_files,fpy_files,neutron_files,progress=True)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/deplete/chain.py”, line 285, in from_endf
evaluation = openmc.data.endf.Evaluation(f)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/data/endf.py”, line 393, in init
fh = open(str(filename_or_obj), ‘r’)
FileNotFoundError: [Errno 2] No such file or directory: ‘n-095_Am_242.endf’

Please, any recommendations?
Thanks,
Javier
create_chain.py (1.4 KB)

4

Hi @Javier_Gonzalez

Hope you’re doing well!

Because Python method listdir() returns a list containing items in the directory not the path of those endf files.

import glob
decay_files = glob.glob('/path/to/ENDF-B-VII.1-decay/*endf')
fpy_files = glob.glob('/path/to/ENDF-B-VII.1-nfy/*endf')
neutron_files = glob.glob('/path/to/ENDF-B-VII.1-neutrons/*endf')

create_chain.py (1.4 KB)

5

Hey @Pranto, you are completely right, I was not indicating the specific ENDF files.

Another question. For instance, in my model, I have materials with nuclides such as U235, U238, O16, Si28, C0, Li6, Be9, F19, etc., and based on those nuclides I wish to generate a depletion chain. As above-mentioned, I have three folders that contain the sub-libraries: in the Neutron Reaction folder there are n files, in the FPY folder there are m files (m<n), and in the Decay folder there are k files (m<n<k).

There are nuclides for which I have files in the three folders (e.g., U235), however for other nuclides (e.g., O16) I do not have FPY files. My question is, what should I do in the latter case?

I imagine there will be an error when

openmc.deplete.Chain.from_endf(decay_file,fpy_file,neutron_file,progress=True)

is called because the fpy_file has not been indicated.

Thanks,
Javier

6

@Javier_Gonzalez You can use chain.reduce() method to reduce the size of the depletion chain.

import openmc.deplete
chain = openmc.deplete.Chain.from_xml('chain_casl_pwr.xml')
red = chain.reduce(['U235', 'U238', 'Si28'], 0) # you'll get a chain just containing U235, U238, Si28 only
red.export_to_xml('simple.xml')

You can use level to control your search.

red = chain.reduce(['U235', 'U238', 'Si28'], 2)

You can find more details here.

Hope this will help!

7

Hi @Pranto,

Thanks for the recommendation, unfortunately that is not answering my question. My intention is to use ENDF files from the TENDL or JEFF libraries, that’s why I wish to know how to use openmc.deplete.Chain.from_endf.

Thank you again,
Javier

8

You should not expect any errors because you are missing fission product yield (FPY) data for non-fissile isotopes like oxygen. The depletion chains built and provided via https://openmc.org/depletion-chains/ are built using a similar approach

9

Understood @andrewjohnson. I am trying with the script and ENDF files in this folder, and am getting the following error:

===============
Traceback (most recent call last):
File “/home/javier/Documents/OpenMC/NuclearData/ENDFB7.1_Chain/create_chain.py”, line 41, in
chain = openmc.deplete.Chain.from_endf(decay_file,fpy_file,neutron_file,progress=True)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/deplete/chain.py”, line 285, in from_endf
evaluation = openmc.data.endf.Evaluation(f)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/data/endf.py”, line 393, in init
fh = open(str(filename_or_obj), ‘r’)
FileNotFoundError: [Errno 2] No such file or directory: ‘n’

===============

Please, any recommendations?

10

Looking at that folder, I only see a single decay, FPY, and neutron file for U235. In general, you’ll need to have multiple ENDF files for each category (decay, FPY, neutron). Each of the decay_file, fpy_file, and neutron_file arguments should be a list of strings (indicating file paths of each ENDF file).

11

Hi @paulromano, I had included only 3 files in case someone wanted to try and I thought it was possible to do it with few files. Now I have uploaded the 3 folders (decay, FPY, and neutron), each with multiple ENDF files, and the updated script. Still getting this error:

===========================
File “/home/javier/Documents/OpenMC/NuclearData/ENDFB7.1_Chain/create_chain.py”, line 60, in
chain = openmc.deplete.Chain.from_endf(decay_file,fpy_file,neutron_file,progress=True)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/deplete/chain.py”, line 285, in from_endf
evaluation = openmc.data.endf.Evaluation(f)
File “/home/javier/anaconda3/lib/python3.8/site-packages/openmc/data/endf.py”, line 393, in init
fh = open(str(filename_or_obj), ‘r’)
FileNotFoundError: [Errno 2] No such file or directory: ‘ENDF-B-VII.1-neutron/n-001_H_001.endf’

===========================

Any recommendation is appreciated. Thanks.

12

The error mentions a ENDF-B-VII.1-neutron/ directory but it looks like you have a ENDF-B-VII.1-neutrons/ directory.

13

Yes @paulromano, you are completely right. Now it is working. Thanks for noticing it!