Merge pull request #30 from fusion-energy/removing_cross_section_down…

…loader

remove testing dependency to easy conda packaging

environment.yml

@@ -14,5 +14,3 @@ dependencies:
   - openmc=0.13.3=dagmc*
   - networkx
   - pip
-  - pip:
-    - openmc_data_downloader

pyproject.toml

@@ -31,8 +31,7 @@ write_to = "src/_version.py"
 
 [project.optional-dependencies]
 tests = [
-    "pytest",
-    "openmc_data_downloader"
+    "pytest"
 ]
 
 [project.urls]

tests/test_brep_to_h5m/test_h5m_in_transport.py

@@ -1,5 +1,4 @@
 import openmc
-import openmc_data_downloader
 from cad_to_dagmc.brep_to_h5m import brep_to_h5m
 import math
 
@@ -25,8 +24,7 @@ def transport_particles_on_h5m_geometry(
 ):
     """A function for testing the geometry file with particle transport in
     DAGMC OpenMC. Requires openmc and either the cross_sections_xml to be
-    specified or openmc_data_downloader installed. Returns the flux on each
-    volume
+    specified. Returns the flux on each volume
 
     Arg:
         h5m_filename: The name of the DAGMC h5m file to test
@@ -47,19 +45,26 @@ def transport_particles_on_h5m_geometry(
         # simplified material definitions have been used to keen this example minimal
         mat_dag_material_tag = openmc.Material(name=material_tag)
         mat_dag_material_tag.add_nuclide(nuclides[i], 1, "ao")
-        mat_dag_material_tag.set_density("g/cm3", 0.1)
+        mat_dag_material_tag.set_density("g/cm3", 0.01)
 
         materials.append(mat_dag_material_tag)
 
     if cross_sections_xml:
-        materials.cross_sections = cross_sections_xml
+        openmc.config["cross_sections"] = cross_sections_xml
+
     else:
-        # downloads the nuclear data and sets the openmc_cross_sections environmental variable
-        materials.download_cross_section_data(
-            libraries=["ENDFB-7.1-NNDC"],
-            set_OPENMC_CROSS_SECTIONS=True,
-            particles=["neutron"],
-        )
+        with open("cross_sections.xml", "w") as file:
+            file.write(
+                """
+            <?xml version='1.0' encoding='UTF-8'?>
+            <cross_sections>
+            <library materials="H1" path="tests/ENDFB-7.1-NNDC_H1.h5" type="neutron"/>
+            <library materials="H2" path="ENDFB-7.1-NNDC_H2.h5" type="neutron"/>
+            </cross_sections>
+            """
+            )
+
+        openmc.config["cross_sections"] = "cross_sections.xml"
 
     dag_univ = openmc.DAGMCUniverse(filename=h5m_filename)
     bound_dag_univ = dag_univ.bounded_universe()
@@ -133,7 +138,9 @@ def test_transport_on_h5m_with_6_volumes():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
 
@@ -153,7 +160,9 @@ def test_transport_on_h5m_with_1_volumes():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
 
@@ -173,7 +182,9 @@ def test_transport_on_h5m_with_2_joined_volumes():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
 
@@ -193,7 +204,9 @@ def test_transport_on_h5m_with_2_sep_volumes():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
 
@@ -213,7 +226,9 @@ def test_transport_result_h5m_with_2_sep_volumes():
     )
 
     new_tally = transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
     brep_to_h5m(
@@ -225,7 +240,9 @@ def test_transport_result_h5m_with_2_sep_volumes():
         mesh_algorithm=1,
     )
     stl_tally = transport_particles_on_h5m_geometry(
-        h5m_filename=h5m_filename, material_tags=material_tags
+        h5m_filename=h5m_filename,
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
     assert math.isclose(new_tally, stl_tally)
@@ -250,7 +267,9 @@ def test_stl_vs_in_memory_1_volume():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename="h5m_from_in_memory_method.h5m", material_tags=material_tags
+        h5m_filename="h5m_from_in_memory_method.h5m",
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )
 
 
@@ -273,5 +292,7 @@ def test_stl_vs_in_memory_2_joined_volume():
     )
 
     transport_particles_on_h5m_geometry(
-        h5m_filename="h5m_from_in_memory_method.h5m", material_tags=material_tags
+        h5m_filename="h5m_from_in_memory_method.h5m",
+        material_tags=material_tags,
+        nuclides=["H1"] * len(material_tags),
     )

tests/test_cad_to_dagmc/test_h5m_in_simulation.py

@@ -1,6 +1,4 @@
 import openmc
-import openmc_data_downloader
-
 
 """
     - h5m files can be used a transport geometry in DAGMC with OpenMC
@@ -13,21 +11,28 @@ def transport_particles_on_h5m_geometry(
 ):
     """A function for testing the geometry file with particle transport in DAGMC OpenMC"""
 
+    with open("cross_sections.xml", "w") as file:
+        file.write(
+            """
+        <?xml version='1.0' encoding='UTF-8'?>
+        <cross_sections>
+        <library materials="H1" path="tests/ENDFB-7.1-NNDC_H1.h5" type="neutron"/>
+        <library materials="H2" path="ENDFB-7.1-NNDC_H2.h5" type="neutron"/>
+        </cross_sections>
+        """
+        )
+
+    openmc.config["cross_sections"] = "cross_sections.xml"
+
     materials = openmc.Materials()
     for material_tag in material_tags:
         # simplified material definitions have been used to keen this example minimal
         mat_dag_material_tag = openmc.Material(name=material_tag)
         mat_dag_material_tag.add_nuclide("H1", 1, "ao")
-        mat_dag_material_tag.set_density("g/cm3", 2)
+        mat_dag_material_tag.set_density("g/cm3", 0.01)
 
         materials.append(mat_dag_material_tag)
 
-    materials.download_cross_section_data(
-        libraries=["ENDFB-7.1-NNDC"],
-        set_OPENMC_CROSS_SECTIONS=True,
-        particles=["neutron"],
-    )
-
     # makes use of the dagmc geometry
     dag_univ = openmc.DAGMCUniverse(h5m_filename)
 

tests/test_vertices_to_h5m/test_package.py

@@ -2,7 +2,6 @@
 
 import numpy as np
 import openmc
-import openmc_data_downloader
 import pymoab as mb
 from pymoab import core, types
 
@@ -58,20 +57,31 @@ def transport_particles_on_h5m_geometry(
     for material_tag in material_tags:
         # simplified material definitions have been used to keen this example minimal
         mat_dag_material_tag = openmc.Material(name=material_tag)
-        mat_dag_material_tag.add_element("H", 1, "ao")
-        mat_dag_material_tag.set_density("g/cm3", 2)
+        mat_dag_material_tag.add_nuclide("H1", 1, "ao")
+        mat_dag_material_tag.set_density("g/cm3", 0.01)
 
         materials.append(mat_dag_material_tag)
 
     if cross_sections_xml:
         materials.cross_sections = cross_sections_xml
     # downloads the nuclear data and sets the openmc_cross_sections environmental variable
 
-    materials.download_cross_section_data(
-        libraries=["ENDFB-7.1-NNDC"],
-        set_OPENMC_CROSS_SECTIONS=True,
-        particles=["neutron"],
-    )
+    if cross_sections_xml:
+        openmc.config["cross_sections"] = cross_sections_xml
+
+    else:
+        with open("cross_sections.xml", "w") as file:
+            file.write(
+                """
+            <?xml version='1.0' encoding='UTF-8'?>
+            <cross_sections>
+            <library materials="H1" path="tests/ENDFB-7.1-NNDC_H1.h5" type="neutron"/>
+            <library materials="H2" path="ENDFB-7.1-NNDC_H2.h5" type="neutron"/>
+            </cross_sections>
+            """
+            )
+
+        openmc.config["cross_sections"] = "cross_sections.xml"
 
     # makes use of the dagmc geometry
     dag_univ = openmc.DAGMCUniverse(h5m_filename)