adding cross sections xml

fusion-energy · Jan 17, 2024 · 37bdd58 · 37bdd58
1 parent 7e7bba8
commit 37bdd58
Show file tree

Hide file tree

Showing 3 changed files with 13 additions and 0 deletions.
diff --git a/examples/point_source_example.py b/examples/point_source_example.py
@@ -1,5 +1,9 @@
 import openmc
 from openmc_plasma_source import FusionPointSource
+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')

diff --git a/examples/ring_source_example.py b/examples/ring_source_example.py
@@ -1,6 +1,10 @@
 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')

diff --git a/examples/tokamak_source_example.py b/examples/tokamak_source_example.py
@@ -1,6 +1,11 @@
+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)