diff --git a/examples/python/white/build-xml.py b/examples/python/white/build-xml.py
deleted file mode 100644
index 184cbf346b..0000000000
--- a/examples/python/white/build-xml.py
+++ /dev/null
@@ -1,81 +0,0 @@
-import numpy as np
-import openmc
-
-###############################################################################
-#                      Simulation Input File Parameters
-###############################################################################
-
-# OpenMC simulation parameters
-batches = 500
-inactive = 100
-particles = 10000
-
-
-###############################################################################
-#                 Exporting to OpenMC materials.xml file
-###############################################################################
-
-# Instantiate a Material and register the Nuclide
-fuel = openmc.Material(material_id=1, name='fuel')
-fuel.set_density('g/cc', 4.5)
-fuel.add_nuclide('U235', 1.)
-
-# Instantiate a Materials collection and export to XML
-materials_file = openmc.Materials([fuel])
-materials_file.export_to_xml()
-
-
-###############################################################################
-#                 Exporting to OpenMC geometry.xml file
-###############################################################################
-
-# Instantiate Surfaces
-surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
-surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
-surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
-surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
-surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
-surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
-
-surf1.boundary_type = 'vacuum'
-surf2.boundary_type = 'vacuum'
-surf3.boundary_type = 'white'
-surf4.boundary_type = 'white'
-surf5.boundary_type = 'white'
-surf6.boundary_type = 'white'
-
-# Instantiate Cell
-cell = openmc.Cell(cell_id=1, name='cell 1')
-
-# Use surface half-spaces to define region
-cell.region = +surf1 & -surf2 & +surf3 & -surf4 & +surf5 & -surf6
-
-# Register Material with Cell
-cell.fill = fuel
-
-# Instantiate Universes
-root = openmc.Universe(universe_id=0, name='root universe')
-
-# Register Cell with Universe
-root.add_cell(cell)
-
-# Instantiate a Geometry, register the root Universe, and export to XML
-geometry = openmc.Geometry(root)
-geometry.export_to_xml()
-
-
-###############################################################################
-#                   Exporting to OpenMC settings.xml file
-###############################################################################
-
-# Instantiate a Settings object, set all runtime parameters, and export to XML
-settings_file = openmc.Settings()
-settings_file.batches = batches
-settings_file.inactive = inactive
-settings_file.particles = particles
-
-# Create an initial uniform spatial source distribution over fissionable zones
-uniform_dist = openmc.stats.Point(xyz=(0, 0, 0))
-settings_file.source = openmc.source.Source(space=uniform_dist)
-
-settings_file.export_to_xml()