cleaned readme

.github/workflows/ci_with_install.yml

@@ -1,12 +1,11 @@
-# This CI will lauch a Docker image that contains all the dependencies required
+# This CI will launch a Docker image that contains all the dependencies required
 # within that image the pytest test suite is run 
 
 name: CI with install
 
 on:
   pull_request:
     branches:
-      - develop
       - main
     paths-ignore:
       - 'docs/**'
@@ -20,22 +19,16 @@ jobs:
   testing:
     runs-on: ubuntu-latest
     container:
-      # image: continuumio/miniconda3:4.10.3
-      image: ghcr.io/openmc-data-storage/miniconda3_4.9.2_endfb-7.1_nndc_tendl_2019
+      image: openmc/openmc:develop
     steps:
       - name: Checkout repository
         uses: actions/checkout@v3
 
-      - name: install dependancies
-        run: |
-          conda install -c conda-forge mamba
-          mamba install -c conda-forge openmc
-
       - name: install package
         run: |
           pip install --upgrade pip
           pip install .
-          python -c "import regular_mesh_plotter"
+          python -c "import openmc_regular_mesh_plotter"
 
       - name: install packages for tests
         run: |
@@ -48,4 +41,6 @@ jobs:
       - name: Run examples
         run: |
           cd examples
-          python plot_regular_mesh_tally_with_csg_geometry.py
+          python plot_minimal_example.py
+          python plot_with_custom_color_map.py
+          python plot_sweep_through_slice_indexes.py

CITATION.cff

@@ -7,7 +7,7 @@ authors:
 - family-names: "Delaporte-Mathurin"
   given-names: "Rémi"
   orcid: "https://orcid.org/0000-0003-1064-8882"
-title: "A Python package for plotting regular mesh tally results from neutronics simulations."
-version: 0.1.0
-date-released: 2021-10-30
-url: "https://github.com/fusion-energy/regular_mesh_plotter"
+title: "A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations."
+version: 1.0.0
+date-released: 2022-09-01
+url: "https://github.com/fusion-energy/openmc_regular_mesh_plotter"

README.md

@@ -2,13 +2,12 @@
 
 [![CI with install](https://github.com/fusion-energy/openmc_regular_mesh_plotter/actions/workflows/ci_with_install.yml/badge.svg?branch=develop)](https://github.com/fusion-energy/openmc_regular_mesh_plotter/actions/workflows/ci_with_install.yml)
 
-[![PyPI](https://img.shields.io/pypi/v/regular-mesh-plotter?color=brightgreen&label=pypi&logo=grebrightgreenen&logoColor=green)](https://pypi.org/project/regular-mesh-plotter/)
+[![PyPI](https://img.shields.io/pypi/v/openmc-regular-mesh-plotter?color=brightgreen&label=pypi&logo=grebrightgreenen&logoColor=green)](https://pypi.org/project/openmc-regular-mesh-plotter/)
 
-[![codecov](https://codecov.io/gh/fusion-energy/openmc_regular_mesh_plotter/branch/main/graph/badge.svg)](https://codecov.io/gh/fusion-energy/openmc_regular_mesh_plotter)
 
-## A minimal Python package that extracts 2D mesh tally results for plotting convenience.
+# A minimal Python package that plots slices of OpenMC regular mesh tallies with the model geometry.
 
-This package is deployed on [xsplot.com](https://www.xsplot.com) as part of the ```openmc_plot``` suite of plotting apps
+![example regular mesh tally plot 1](https://user-images.githubusercontent.com/8583900/265032335-27463ee9-8960-4f5e-a662-dab0b6cd9fc5.png)
 
 # Local install
 
@@ -21,3 +20,7 @@ pip install openmc_regular_mesh_plotter
 # Usage
 
 See the [examples folder](https://github.com/fusion-energy/openmc_regular_mesh_plotter/tree/main/examples) for example scripts
+
+# Web App
+
+This package is deployed on [xsplot.com](https://www.xsplot.com) as part of the ```openmc_plot``` suite of plotting apps

examples/plot_minimal_example.py

@@ -29,7 +29,11 @@
 my_settings.particles = 5000
 my_settings.run_mode = "fixed source"
 # Create a DT point source
-source = openmc.Source()
+try:
+    source = openmc.IndependentSource()
+except:
+    # work with older versions of openmc
+    source = openmc.Source()
 source.space = openmc.stats.Point((100, 0, 0))
 source.angle = openmc.stats.Isotropic()
 source.energy = openmc.stats.Discrete([14e6], [1])
@@ -76,9 +80,7 @@
     outline=True,  # enables an outline around the geometry
     geometry=my_geometry,
     outline_by="material",
-    colorbar_kwargs={
-        "label": r"Heating $\\MJcm^{-3}s^{-1}$"
-    },  # labels support latex formatting
+    colorbar_kwargs={"label": "Heating MJ/cm3/s"},
     outline_kwargs={
         "colors": "grey",
         "linewidths": 2,

examples/plot_sweep_through_slice_indexes.py

@@ -0,0 +1,160 @@
+# this example is used to make the animation in the readme.
+# It is also set up to accept any geometry and the mesh tally will adapt to the geometry dimensions
+
+
+import openmc
+import numpy as np
+from matplotlib.colors import LogNorm
+from openmc_regular_mesh_plotter import plot_mesh_tally
+from matplotlib import cm
+
+# MATERIALS
+
+breeder_material = openmc.Material()  # Pb84.2Li15.8
+breeder_material.add_element("Pb", 84.2, percent_type="ao")
+breeder_material.add_element(
+    "Li",
+    15.8,
+    percent_type="ao",
+    enrichment=7.0,
+    enrichment_target="Li6",
+    enrichment_type="ao",
+)  # natural enrichment = 7% Li6
+breeder_material.set_density("atom/b-cm", 3.2720171e-2)  # around 11 g/cm3
+
+copper_material = openmc.Material()
+copper_material.set_density("g/cm3", 8.5)
+copper_material.add_element("Cu", 1.0)
+
+eurofer_material = openmc.Material()
+eurofer_material.set_density("g/cm3", 7.75)
+eurofer_material.add_element("Fe", 89.067, percent_type="wo")
+
+my_materials = openmc.Materials([breeder_material, eurofer_material, copper_material])
+
+
+# GEOMETRY
+
+# surfaces
+central_sol_surface = openmc.ZCylinder(r=100)
+central_shield_outer_surface = openmc.ZCylinder(r=110)
+vessel_inner_surface = openmc.Sphere(r=500)
+first_wall_outer_surface = openmc.Sphere(r=510)
+breeder_blanket_outer_surface = openmc.Sphere(r=610, boundary_type="vacuum")
+# regions
+central_sol_region = -central_sol_surface & -breeder_blanket_outer_surface
+central_shield_region = (
+    +central_sol_surface
+    & -central_shield_outer_surface
+    & -breeder_blanket_outer_surface
+)
+inner_vessel_region = -vessel_inner_surface & +central_shield_outer_surface
+first_wall_region = -first_wall_outer_surface & +vessel_inner_surface
+breeder_blanket_region = (
+    +first_wall_outer_surface
+    & -breeder_blanket_outer_surface
+    & +central_shield_outer_surface
+)
+# cells
+central_sol_cell = openmc.Cell(region=central_sol_region)
+central_shield_cell = openmc.Cell(region=central_shield_region)
+inner_vessel_cell = openmc.Cell(region=inner_vessel_region)
+first_wall_cell = openmc.Cell(region=first_wall_region)
+breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
+# filling cells with materials
+central_sol_cell.fill = copper_material
+first_wall_cell.fill = eurofer_material
+central_shield_cell.fill = eurofer_material
+breeder_blanket_cell.fill = breeder_material
+
+my_geometry = openmc.Geometry(
+    [
+        central_sol_cell,
+        central_shield_cell,
+        inner_vessel_cell,
+        first_wall_cell,
+        breeder_blanket_cell,
+    ]
+)
+
+# source
+# work with older versions of openmc
+try:
+    source = openmc.IndependentSource()
+except:
+    source = openmc.Source()
+source.angle = openmc.stats.Isotropic()
+source.energy = openmc.stats.Discrete([14e6], [1])
+radius = openmc.stats.Discrete([150], [1])
+z_values = openmc.stats.Discrete([300], [1])
+angle = openmc.stats.Uniform(a=0.0, b=2 * 3.14159265359)
+source.space = openmc.stats.CylindricalIndependent(
+    r=radius, phi=angle, z=z_values, origin=(0.0, 0.0, 0.0)
+)
+
+# Instantiate a Settings object
+my_settings = openmc.Settings()
+my_settings.output = {"tallies": False}
+my_settings.batches = 2
+my_settings.particles = (
+    50000  # increase this to 5000000 to make the same plot as the readme
+)
+my_settings.run_mode = "fixed source"
+my_settings.source = source
+
+# tallies
+bb = my_geometry.bounding_box
+wedge_length = 40  # set to 4 to make plot from readme
+# this just makes the number of mesh voxels in each dimension proportional to the size of the geometry in that dimension.
+# this means we get cube shaped voxels instead of stretched out rectangles
+dimension = [
+    int(bb.width[0] / wedge_length),
+    int(bb.width[1] / wedge_length),
+    int(bb.width[2] / wedge_length),
+]
+mesh = openmc.RegularMesh().from_domain(my_geometry, dimension=dimension)
+mesh_filter = openmc.MeshFilter(mesh)
+mesh_tally = openmc.Tally(name="mesh_tally")
+mesh_tally.filters = [mesh_filter]
+mesh_tally.scores = ["(n,Xt)"]  # tritium production
+my_tallies = openmc.Tallies([mesh_tally])
+
+model = openmc.Model(my_geometry, my_materials, my_settings, my_tallies)
+
+statepoint_filename = model.run()
+
+# makes use of a context manager "with" to automatically close the statepoint file
+with openmc.StatePoint(statepoint_filename) as statepoint:
+    my_mesh_tally_result = statepoint.get_tally(name="mesh_tally")
+
+# set the same color range across all the plots
+data = my_mesh_tally_result.mean.flatten()
+lower_limit = np.min(data[np.nonzero(data)])
+upper_limit = np.max(data[np.nonzero(data)])
+
+for slice_index in range(0, mesh.dimension[1]):
+    plot = plot_mesh_tally(
+        tally=my_mesh_tally_result,
+        basis="xz",
+        slice_index=slice_index,
+        outline=True,  # enables an outline around the geometry
+        geometry=my_geometry,
+        outline_by="cell",
+        pixels=80000,  # double the default pixels to get a better resolution on the geometry outline curve
+        outline_kwargs={
+            "colors": "green",
+            "linewidths": 2,
+        },  # setting the outline color and thickness, otherwise this defaults to black and 1
+        norm=LogNorm(vmin=lower_limit, vmax=upper_limit),  # log scale
+        volume_normalization=False,
+        # colorbar=False, removing color bar from plot
+        cmap=cm.get_cmap("gnuplot"),  # color map contrasts with outline color
+    )
+
+    # adding a title to the plot
+    plot.title.set_text(f"Slice {slice_index}.")
+    plot.figure.savefig(f"plot_slice_index_{str(slice_index).zfill(4)}.png")
+
+import os
+
+os.system("convert -delay 20 plot_slice*.png animated_openmc_regular_mesh_tally.gif")

examples/plot_with_custom_color_map.py

@@ -6,6 +6,8 @@
 from matplotlib.colors import LogNorm
 from openmc_regular_mesh_plotter import plot_mesh_tally
 from matplotlib import cm
+import matplotlib.pyplot as plt
+
 
 # materials
 mat_concrete = openmc.Material()
@@ -129,13 +131,17 @@
 # all (100%) of source particles are 2.5MeV energy
 energy = openmc.stats.Discrete([2.5e6], [1.0])
 
-source = openmc.IndependentSource(space=space, angle=angle, energy=energy)
+# work with older versions of openmc
+try:
+    source = openmc.IndependentSource(space=space, angle=angle, energy=energy)
+except:
+    source = openmc.Source(space=space, angle=angle, energy=energy)
 source.particle = "neutron"
 # Instantiate a Settings object
 my_settings = openmc.Settings()
 my_settings.output = {"tallies": False}
 my_settings.batches = 2
-my_settings.particles = 500000000  # set this to 500000000 to get a full image
+my_settings.particles = 50  # set this to 500000000 to get a full image
 my_settings.run_mode = "fixed source"
 my_settings.source = source
 
@@ -186,7 +192,7 @@
         "colors": "darkgrey",
         "linewidths": 2,
     },  # setting the outline color and thickness, otherwise this defaults to black and 1
-    norm=LogNorm(),  # log scale
+    norm=LogNorm(vmin=1e-2, vmax=1e8),  # log scale
     scaling_factor=1e10,  # multiplies the tally result by scaling_factor which is source strength (neutrons per second)
     volume_normalization=True,
     cmap=cmap,

pyproject.toml

@@ -20,12 +20,6 @@ classifiers = [
 dependencies = [
     "numpy>=1.21.1",
     "matplotlib>=3.4.2",
-    "trimesh",
-    "shapely",
-    "scipy",
-    "pandas",
-    "dagmc_geometry_slice_plotter",
-    "openmc_tally_unit_converter",
     "setuptools_scm",
 ]
 dynamic = ["version"]
@@ -38,9 +32,6 @@ write_to = "src/openmc_regular_mesh_plotter/_version.py"
 [project.optional-dependencies]
 tests = [
     "pytest",
-    "pytest-cov>=2.12.1",
-    "openmc_plasma_source",
-    "openmc_regular_mesh_plotter",
 ]
 
 [project.urls]

src/openmc_regular_mesh_plotter/core.py

@@ -8,7 +8,14 @@
 import openmc.checkvalue as cv
 import matplotlib.pyplot as plt
 
-plt.rcParams["text.usetex"] = True
+from packaging import version
+
+if version.parse(openmc.__version__) < version.parse("0.13.3"):
+    msg = (
+        "openmc_regular_mesh_plotter package requires OpenMC version 0.13.4 "
+        f"or newer. You currently have OpenMC version {openmc.__version__}"
+    )
+    raise ValueError(msg)
 
 _BASES = ["xy", "xz", "yz"]
 
@@ -52,7 +59,7 @@ def plot_mesh_tally(
     value : str
         A string for the type of value to return  - 'mean' (default),
         'std_dev', 'rel_err', 'sum', or 'sum_sq' are accepted
-    outline : tuple
+    outline : True
         If set then an outline will be added to the plot. The outline can be
         by cell or by material.
     outline_by : {'cell', 'material'}

tests/test_units.py

@@ -21,11 +21,16 @@ def test_plot_mesh_tally():
     cell2.fill = mat1
     geom = openmc.Geometry([cell1, cell2])
 
-    source = openmc.IndependentSource()
+    # work with older versions of openmc
+    try:
+        source = openmc.IndependentSource()
+    except:
+        source = openmc.Source()
     source.angle = openmc.stats.Isotropic()
     source.energy = openmc.stats.Discrete([14e6], [1])
     # puts the source in the center of the RectangularParallelepiped
-    source.space = openmc.stats.Point(cell2.bounding_box.center)
+    source.space = openmc.stats.Point([-25.0, 25.0, 25.0])
+    # newer versions of openmc support cell2.bounding_box.center
 
     sett = openmc.Settings()
     sett.batches = 2