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A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles

License: MIT License

Python 100.00%

openmc source particle plot coordinates energy direction gamma

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nucg1gb khurrumsaleem allentro

openmc_source_plotter's Issues

Add images to the readme

would be good to get an image of for plotly

source locations for a plasma (not currently possible due to version conflict)
source location for ring source (not currently possible due to version conflict)
source energy of muir spectrum
source direction of a point source

openmc version 0.11.0 is used as it writes the initial_source.h5 but it doesn't include the cylindrical source.

Invalid spatial distribution for external source: cylindrical

matplotlib stem plots for gamma

plt.stem plots lokk quite nice compared to the line plot

fix direction plot aspect ratio

because of the size difference in the axis it can often appear as if ring sources are sending particles straight up and down

For example a ring source with a radius of 10 will not show the particles traveling left and right very well

Adding Docker file for CI

To get the CI working this package needs a Dockerfile

Other dockerfiles I've built might not be working well with mpich and this is why I can't make initial_source.h5 using openmc version 0.12
https://openmc.discourse.group/t/obtaining-initial-source-for-fix-source-mode/1463/2

So the options are to

use the official openmc docker
Make a new Docker image with either openmc compiled correctly
Make ne new docier image with openmc 0.13 and openmc 0.11, then use openmc in the exec arg

Allow plots with multiple sources

Perhaps input_filename could accept a list of h5 files

Then multiple plots could be made on a single axis, this would be useful for the energy plotting function

plot_energy_from_initial_source(
    input_filename: List
)

implement monkey patch to overwrite openmc.Source

To make the name space the same as OpenMC we can use a monkey patch. Here is the code examples

package code

import openmc

class Source(openmc.Source):

    def plot_source_energy(self, n_samples: int = 1000, prn_seed: int = None):
        ...


openmc.Source = Source  # overwrites the openmc class with this extended class

user code

import openmc
import openmc_source_plotter

my_source = openmc.Source()
my_source.plot_source_energy()

FileNotFoundError when using plot_source_energy()

MWE:

import openmc_source_plotter as osp
import openmc_plasma_source as ops

import numpy as np
import math


my_source = ops.FusionRingSource(
    fuel="DT",
    radius=1,
    z_placement=1,
    angles=(math.radians(-1), math.radians(1)),
)

plot = osp.plot_source_energy(
    source=my_source,
    number_of_particles=2000,
    energy_bins=np.linspace(0, 20e6, 100)
)

Produces:

(base) root@9cd1860b80ad:/home/fusion-neutronics-workflow# python plot_source.py 
                                %%%%%%%%%%%%%%%
                           %%%%%%%%%%%%%%%%%%%%%%%%
                        %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                      %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                    %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                   %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                                    %%%%%%%%%%%%%%%%%%%%%%%%
                                     %%%%%%%%%%%%%%%%%%%%%%%%
                 ###############      %%%%%%%%%%%%%%%%%%%%%%%%
                ##################     %%%%%%%%%%%%%%%%%%%%%%%
                ###################     %%%%%%%%%%%%%%%%%%%%%%%
                ####################     %%%%%%%%%%%%%%%%%%%%%%
                #####################     %%%%%%%%%%%%%%%%%%%%%
                ######################     %%%%%%%%%%%%%%%%%%%%
                #######################     %%%%%%%%%%%%%%%%%%
                 #######################     %%%%%%%%%%%%%%%%%
                 ######################     %%%%%%%%%%%%%%%%%
                  ####################     %%%%%%%%%%%%%%%%%
                    #################     %%%%%%%%%%%%%%%%%
                     ###############     %%%%%%%%%%%%%%%%
                       ############     %%%%%%%%%%%%%%%
                          ########     %%%%%%%%%%%%%%
                                      %%%%%%%%%%%

                 | The OpenMC Monte Carlo Code
       Copyright | 2011-2022 MIT, UChicago Argonne LLC, and contributors
         License | https://docs.openmc.org/en/latest/license.html
         Version | 0.13.0
        Git SHA1 | cff247e35785e7236d67ccf64a3401f0fc50a469
       Date/Time | 2022-07-08 08:43:41
   MPI Processes | 1
  OpenMP Threads | 4

 Reading settings XML file...
 Reading cross sections XML file...
 Reading materials XML file...
 Reading geometry XML file...
 Minimum neutron data temperature: 1.7976931348623157e+308 K
 Maximum neutron data temperature: 0 K
 Preparing distributed cell instances...
 Reading plot XML file...
 Writing summary.h5 file...

 ===============>     FIXED SOURCE TRANSPORT SIMULATION     <===============

 Simulating batch 1
 Creating state point statepoint.1.h5...

 =======================>     TIMING STATISTICS     <=======================

 Total time for initialization     = 1.2467e-01 seconds
   Reading cross sections          = 7.5700e-05 seconds
 Total time in simulation          = 2.7996e-02 seconds
   Time in transport only          = 3.4304e-03 seconds
   Time in active batches          = 2.7996e-02 seconds
   Time accumulating tallies       = 1.3000e-06 seconds
   Time writing statepoints        = 2.4537e-02 seconds
 Total time for finalization       = 4.5000e-06 seconds
 Total time elapsed                = 1.5276e-01 seconds
 Calculation Rate (active)         = 71439 particles/second

 ============================>     RESULTS     <============================

 WARNING: Could not compute uncertainties -- only one active batch simulated!
 Leakage Fraction           = 1.00000

Traceback (most recent call last):
  File "/opt/conda/lib/python3.9/shutil.py", line 806, in move
    os.rename(src, real_dst)
OSError: [Errno 18] Invalid cross-device link: 'initial_source.h5' -> '/tmp/openmc_source_y85xaz55.h5'

During handling of the above exception, another exception occurred:

Traceback (most recent call last):
  File "/home/fusion-neutronics-workflow/plot_source.py", line 19, in <module>
    plot = osp.plot_source_energy(
  File "/opt/conda/lib/python3.9/site-packages/openmc_source_plotter/core.py", line 38, in plot_source_energy
    create_initial_particles(
  File "/opt/conda/lib/python3.9/site-packages/openmc_source_plotter/utils.py", line 100, in create_initial_particles
    shutil.move("initial_source.h5", output_source_filename)
  File "/opt/conda/lib/python3.9/shutil.py", line 826, in move
    copy_function(src, real_dst)
  File "/opt/conda/lib/python3.9/shutil.py", line 435, in copy2
    copyfile(src, dst, follow_symlinks=follow_symlinks)
  File "/opt/conda/lib/python3.9/shutil.py", line 264, in copyfile
    with open(src, 'rb') as fsrc, open(dst, 'wb') as fdst:
FileNotFoundError: [Errno 2] No such file or directory: 'initial_source.h5'

openmc_source_plotter version: current
openmc version 0.13.0

Should I use openmc 0.11 to have this work?

Import error openmc=0.13.3

Not sure if intended to be compatible with openmc=0.13.3. I upgraded to 0.14.0 and error went away.

import openmc
import openmc_source_plotter

--> 230 openmc.SourceBase.sample_initial_particles = sample_initial_particles
    231 openmc.model.Model.sample_initial_particles = sample_initial_particles
    232 openmc.Model.sample_initial_particles = sample_initial_particles

AttributeError: module 'openmc' has no attribute 'SourceBase'

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