moved to src layout

.github/workflows/ci.yml

@@ -1,35 +1,33 @@
-name: CI
+name: CI testing
 
 on:
-  pull_request:    
-    branches:
-    - develop
-    - main
+  pull_request:
   push:
     branches:
-    - develop
-    - main
-
+      - main
+
 jobs:
   testing:
     runs-on: ubuntu-latest
-    container:
-      image: openmc/openmc:develop
     steps:
-      - name: Checkout repository
-        uses: actions/checkout@v3
-
-      - name: Install
-        run: |
-          python3 -m pip install --upgrade pip
-          python3 -m pip install .[tests]
 
-      - name: Run tests
-        run: |
-          python3 -m pytest -v tests
+      - name: checkout actions
+        uses: actions/checkout@v4
 
-      - name: Run examples
+      - name: install dependencies
+        shell: bash
         run: |
-          python3 examples/point_source_example.py
-          python3 examples/ring_source_example.py
-          python3 examples/tokamak_source_example.py
+
+          wget -O Miniforge3.sh "https://github.com/conda-forge/miniforge/releases/latest/download/Miniforge3-$(uname)-$(uname -m).sh"
+          bash Miniforge3.sh -b -p "${HOME}/conda"
+          source "${HOME}/conda/etc/profile.d/conda.sh"
+          source "${HOME}/conda/etc/profile.d/mamba.sh"
+          mamba activate
+          mamba install -y -c conda-forge openmc
+          pip install .
+          python -c "import openmc_plasma_source"
+          pip install .[tests]
+          pytest -v tests
+          python examples/point_source_example.py
+          python examples/ring_source_example.py
+          python examples/tokamak_source_example.py

CITATION.cff

@@ -14,6 +14,6 @@ authors:
 - family-names: "Faisal"
   given-names: "Mohammed"
 title: "OpenMC Plasma Source"
-version: 0.2.4
-date-released: 2022-01-31
+version: 0.3.0
+date-released: 2023-01-12
 url: "https://github.com/fusion-energy/openmc-plasma-source"

examples/cross_sections.xml

@@ -0,0 +1,4 @@
+<?xml version='1.0' encoding='utf-8'?>
+<cross_sections>
+  <library materials="H1" path="neutron/H1.h5" type="neutron" />
+</cross_sections>

examples/point_source_example.py

@@ -1,23 +1,17 @@
 import openmc
 from openmc_plasma_source import FusionPointSource
+from pathlib import Path
 
-my_source = FusionPointSource(coordinate=(0, 0, 0), temperature=20000.0, fuel="DT")
+# just making use of a local cross section xml file, replace with your own cross sections or comment out
+openmc.config["cross_sections"] = Path(__file__).parent.resolve() / "cross_sections.xml"
 
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type="vacuum")
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
 
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
+# define the source
+my_source = FusionPointSource(coordinate=(0, 0, 0), temperature=20000.0, fuel="DT")
 
 # Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
@@ -26,19 +20,7 @@
 settings.particles = 1000
 settings.source = my_source
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
 
-model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
-)
+model = openmc.model.Model(materials=None, geometry=geometry, settings=settings)
 
 model.run()

examples/ring_source_example.py

@@ -1,49 +1,30 @@
 import openmc
 from openmc_plasma_source import FusionRingSource
 import math
+from pathlib import Path
 
+# just making use of a local cross section xml file, replace with your own cross sections or comment out
+openmc.config["cross_sections"] = Path(__file__).parent.resolve() / "cross_sections.xml"
+
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type="vacuum")
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
+
+# define the source
 my_source = FusionRingSource(
     radius=700,
     angles=(0.0, 2 * math.pi),  # 360deg source
     temperature=20000.0,
 )
 
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
-
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
-
-# Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
 settings.run_mode = "fixed source"
 settings.batches = 10
 settings.particles = 1000
+# tell OpenMC we're going to use our custom source
 settings.source = my_source
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
-
-model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
-)
+model = openmc.model.Model(materials=None, geometry=geometry, settings=settings)
 
 model.run()

examples/tokamak_source_example.py

@@ -1,6 +1,15 @@
+from pathlib import Path
 import openmc
 from openmc_plasma_source import TokamakSource
 
+# just making use of a local cross section xml file, replace with your own cross sections or comment out
+openmc.config["cross_sections"] = Path(__file__).parent.resolve() / "cross_sections.xml"
+
+
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type="vacuum")
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
 
 # create a plasma source
 my_plasma = TokamakSource(
@@ -22,43 +31,14 @@
     ion_temperature_beta=6,
 )
 
-
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
-
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
-
 # Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
 settings.run_mode = "fixed source"
 settings.batches = 10
 settings.particles = 1000
 settings.source = my_plasma.sources
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
 
-model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
-)
+model = openmc.model.Model(materials=None, geometry=geometry, settings=settings)
 
 model.run()

pyproject.toml

@@ -1,11 +1,42 @@
 [build-system]
 requires = [
-    "setuptools >= 45",
-    "wheel",
-    "setuptools_scm[toml] >= 6.2",
-    "setuptools_scm_git_archive",
+    "setuptools>=65.5.0",
+    "setuptools_scm[toml]>=7.0.5"
 ]
 build-backend = "setuptools.build_meta"
 
+[project]
+name = "openmc_plasma_source"
+dynamic = ["version"]
+description = "Creates tokamak and ICF plasma sources for OpenMC"
+readme = "README.md"
+classifiers = [
+    "Programming Language :: Python :: 3",
+    "License :: OSI Approved :: MIT License",
+    "Operating System :: OS Independent",
+]
+authors = [
+  { name="Rémi Delaporte-Mathurin" },
+]
+requires-python = ">=3.8"
+keywords = ["python", "neutron", "fusion", "source", "openmc", "energy", "tokamak"]
+dependencies = [
+    "numpy>=1.9",
+    "matplotlib>=3.2.2",
+]
+
+[project.urls]
+"Homepage" = "https://github.com/fusion-energy/openmc-plasma-source"
+"Bug Tracker" = "https://github.com/fusion-energy/openmc-plasma-source/issues"
+
 [tool.setuptools_scm]
-write_to = "openmc_plasma_source/_version.py"
+write_to = "src/_version.py"
+
+[project.optional-dependencies]
+tests = [
+    "pytest>=5.4.3",
+    "hypothesis"
+]
+
+[tool.setuptools]
+package-dir = {"" = "src"}

tests/test_tokamak_source.py

@@ -264,15 +264,15 @@ def tokamak_source_strategy(draw):
         st.floats(
             min_value=1e-5 * major_radius,
             max_value=np.nextafter(major_radius, major_radius - 1),
-            **finites
+            **finites,
         )
     )
 
     pedestal_radius = draw(
         st.floats(
             min_value=0.8 * minor_radius,
             max_value=np.nextafter(minor_radius, minor_radius - 1),
-            **finites
+            **finites,
         )
     )
 
@@ -288,7 +288,7 @@ def tokamak_source_strategy(draw):
         st.floats(
             min_value=np.nextafter(-0.5 * minor_radius, +1),
             max_value=np.nextafter(0.5 * minor_radius, -1),
-            **finites
+            **finites,
         )
     )