diff --git a/examples/python/basic/build-xml.py b/examples/python/basic/build-xml.py
deleted file mode 100644
index 5caed2810..000000000
--- a/examples/python/basic/build-xml.py
+++ /dev/null
@@ -1,121 +0,0 @@
-import openmc
-
-
-###############################################################################
-#                      Simulation Input File Parameters
-###############################################################################
-
-# OpenMC simulation parameters
-batches = 15
-inactive = 5
-particles = 10000
-
-
-###############################################################################
-#                 Exporting to OpenMC materials.xml file
-###############################################################################
-
-# Instantiate some Materials and register the appropriate Nuclides
-moderator = openmc.Material(material_id=41, name='moderator')
-moderator.set_density('g/cc', 1.0)
-moderator.add_element('H', 2.)
-moderator.add_element('O', 1.)
-moderator.add_s_alpha_beta('c_H_in_H2O')
-
-fuel = openmc.Material(material_id=40, 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([moderator, fuel])
-materials_file.export_to_xml()
-
-
-###############################################################################
-#                 Exporting to OpenMC geometry.xml file
-###############################################################################
-
-# Instantiate ZCylinder surfaces
-surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=7, name='surf 1')
-surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=9, name='surf 2')
-surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=11, name='surf 3')
-surf3.boundary_type = 'vacuum'
-
-# Instantiate Cells
-cell1 = openmc.Cell(cell_id=1, name='cell 1')
-cell2 = openmc.Cell(cell_id=100, name='cell 2')
-cell3 = openmc.Cell(cell_id=101, name='cell 3')
-cell4 = openmc.Cell(cell_id=2, name='cell 4')
-
-# Use surface half-spaces to define regions
-cell1.region = -surf2
-cell2.region = -surf1
-cell3.region = +surf1
-cell4.region = +surf2 & -surf3
-
-# Register Materials with Cells
-cell2.fill = fuel
-cell3.fill = moderator
-cell4.fill = moderator
-
-# Instantiate Universes
-universe1 = openmc.Universe(universe_id=37)
-root = openmc.Universe(universe_id=0, name='root universe')
-cell1.fill = universe1
-
-# Register Cells with Universes
-universe1.add_cells([cell2, cell3])
-root.add_cells([cell1, cell4])
-
-# 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
-bounds = [-4., -4., -4., 4., 4., 4.]
-uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
-settings_file.source = openmc.source.Source(space=uniform_dist)
-
-settings_file.export_to_xml()
-
-
-###############################################################################
-#                   Exporting to OpenMC tallies.xml file
-###############################################################################
-
-# Instantiate some tally Filters
-cell_filter = openmc.CellFilter(cell2)
-energy_filter = openmc.EnergyFilter([0., 20.e6])
-energyout_filter = openmc.EnergyoutFilter([0., 20.e6])
-
-# Instantiate the first Tally
-first_tally = openmc.Tally(tally_id=1, name='first tally')
-first_tally.filters = [cell_filter]
-scores = ['total', 'scatter', 'nu-scatter',
-          'absorption', 'fission', 'nu-fission']
-first_tally.scores = scores
-
-# Instantiate the second Tally
-second_tally = openmc.Tally(tally_id=2, name='second tally')
-second_tally.filters = [cell_filter, energy_filter]
-second_tally.scores = scores
-
-# Instantiate the third Tally
-third_tally = openmc.Tally(tally_id=3, name='third tally')
-third_tally.filters = [cell_filter, energy_filter, energyout_filter]
-third_tally.scores = ['scatter', 'nu-scatter', 'nu-fission']
-
-# Instantiate a Tallies collection and export to XML
-tallies_file = openmc.Tallies((first_tally, second_tally, third_tally))
-tallies_file.export_to_xml()
diff --git a/examples/python/boxes/build-xml.py b/examples/python/boxes/build-xml.py
deleted file mode 100644
index d04f0f67a..000000000
--- a/examples/python/boxes/build-xml.py
+++ /dev/null
@@ -1,125 +0,0 @@
-import numpy as np
-import openmc
-
-###############################################################################
-#                      Simulation Input File Parameters
-###############################################################################
-
-# OpenMC simulation parameters
-batches = 15
-inactive = 5
-particles = 10000
-
-
-###############################################################################
-#                 Exporting to OpenMC materials.xml File
-###############################################################################
-
-# Instantiate some Materials and register the appropriate Nuclides
-fuel1 = openmc.Material(material_id=1, name='fuel')
-fuel1.set_density('g/cc', 4.5)
-fuel1.add_nuclide('U235', 1.)
-
-fuel2 = openmc.Material(material_id=2, name='depleted fuel')
-fuel2.set_density('g/cc', 4.5)
-fuel2.add_nuclide('U238', 1.)
-
-moderator = openmc.Material(material_id=3, name='moderator')
-moderator.set_density('g/cc', 1.0)
-moderator.add_element('H', 2.)
-moderator.add_element('O', 1.)
-moderator.add_s_alpha_beta('c_H_in_H2O')
-
-# Instantiate a Materials collection and export to XML
-materials_file = openmc.Materials([fuel1, fuel2, moderator])
-materials_file.export_to_xml()
-
-
-###############################################################################
-#                 Exporting to OpenMC geometry.xml file
-###############################################################################
-
-# Instantiate planar surfaces
-x1 = openmc.XPlane(surface_id=1, x0=-10)
-x2 = openmc.XPlane(surface_id=2, x0=-7)
-x3 = openmc.XPlane(surface_id=3, x0=-4)
-x4 = openmc.XPlane(surface_id=4, x0=4)
-x5 = openmc.XPlane(surface_id=5, x0=7)
-x6 = openmc.XPlane(surface_id=6, x0=10)
-y1 = openmc.YPlane(surface_id=11, y0=-10)
-y2 = openmc.YPlane(surface_id=12, y0=-7)
-y3 = openmc.YPlane(surface_id=13, y0=-4)
-y4 = openmc.YPlane(surface_id=14, y0=4)
-y5 = openmc.YPlane(surface_id=15, y0=7)
-y6 = openmc.YPlane(surface_id=16, y0=10)
-z1 = openmc.ZPlane(surface_id=21, z0=-10)
-z2 = openmc.ZPlane(surface_id=22, z0=-7)
-z3 = openmc.ZPlane(surface_id=23, z0=-4)
-z4 = openmc.ZPlane(surface_id=24, z0=4)
-z5 = openmc.ZPlane(surface_id=25, z0=7)
-z6 = openmc.ZPlane(surface_id=26, z0=10)
-
-# Set vacuum boundary conditions on outside
-for surface in [x1, x6, y1, y6, z1, z6]:
-    surface.boundary_type = 'vacuum'
-
-# Instantiate Cells
-inner_box = openmc.Cell(cell_id=1, name='inner box')
-middle_box = openmc.Cell(cell_id=2, name='middle box')
-outer_box = openmc.Cell(cell_id=3, name='outer box')
-
-# Use each set of six planes to create solid cube regions. We can then use these
-# to create cubic shells.
-inner_cube = +x3 & -x4 & +y3 & -y4 & +z3 & -z4
-middle_cube = +x2 & -x5 & +y2 & -y5 & +z2 & -z5
-outer_cube = +x1 & -x6 & +y1 & -y6 & +z1 & -z6
-outside_inner_cube = -x3 | +x4 | -y3 | +y4 | -z3 | +z4
-
-# Use surface half-spaces to define regions
-inner_box.region = inner_cube
-middle_box.region = middle_cube & outside_inner_cube
-outer_box.region = outer_cube & ~middle_cube
-
-# Register Materials with Cells
-inner_box.fill = fuel1
-middle_box.fill = fuel2
-outer_box.fill = moderator
-
-# Instantiate root universe
-root = openmc.Universe(universe_id=0, name='root universe')
-root.add_cells([inner_box, middle_box, outer_box])
-
-# 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.Box(*outer_cube.bounding_box, only_fissionable=True)
-settings_file.source = openmc.source.Source(space=uniform_dist)
-
-settings_file.export_to_xml()
-
-###############################################################################
-#                   Exporting to OpenMC plots.xml File
-###############################################################################
-
-plot = openmc.Plot(plot_id=1)
-plot.origin = [0, 0, 0]
-plot.width = [20, 20]
-plot.pixels = [200, 200]
-plot.color_by = 'cell'
-
-# Instantiate a Plots collection and export to XML
-plot_file = openmc.Plots([plot])
-plot_file.export_to_xml()
diff --git a/examples/xml/basic/geometry.xml b/examples/xml/basic/geometry.xml
deleted file mode 100644
index b30884f8c..000000000
--- a/examples/xml/basic/geometry.xml
+++ /dev/null
@@ -1,15 +0,0 @@
-<?xml version="1.0"?>
-<geometry>
-
-  <!-- Definition of Cells -->
-  <cell id="1" universe="0" fill="37" region="-2" />
-  <cell id="100" universe="37" material="40" region="-1" />
-  <cell id="101" universe="37" material="41" region="1" />
-  <cell id="2" universe="0" material="41" region="2 -3" />
-
-  <!-- Defition of Surfaces -->
-  <surface id="1" type="z-cylinder" coeffs="0 0 7" />
-  <surface id="2" type="z-cylinder" coeffs="0 0 9" />
-  <surface id="3" type="z-cylinder" coeffs="0 0 11" boundary="vacuum" />
-
-</geometry>
diff --git a/examples/xml/basic/materials.xml b/examples/xml/basic/materials.xml
deleted file mode 100644
index 606c676df..000000000
--- a/examples/xml/basic/materials.xml
+++ /dev/null
@@ -1,16 +0,0 @@
-<?xml version="1.0"?>
-<materials>
-
-  <material id="40">
-    <density value="4.5" units="g/cc" />
-    <nuclide name="U235" ao="1.0" />
-  </material>
-
-  <material id="41">
-    <density value="1.0" units="g/cc" />
-    <nuclide name="H1" ao="2.0" />
-    <nuclide name="O16" ao="1.0" />
-    <sab name="c_H_in_H2O"/>
-  </material>
-
-</materials>
diff --git a/examples/xml/basic/settings.xml b/examples/xml/basic/settings.xml
deleted file mode 100644
index 6e622b6ce..000000000
--- a/examples/xml/basic/settings.xml
+++ /dev/null
@@ -1,16 +0,0 @@
-<?xml version="1.0"?>
-<settings>
-
-  <run_mode>eigenvalue</run_mode>
-  <batches>15</batches>
-  <inactive>5</inactive>
-  <particles>10000</particles>
-
-  <!-- Starting source -->
-  <source>
-    <space type="box">
-      <parameters>-4 -4 -4  4  4  4</parameters>
-    </space>
-  </source>
-
-</settings>
diff --git a/examples/xml/basic/tallies.xml b/examples/xml/basic/tallies.xml
deleted file mode 100644
index a125e9ca2..000000000
--- a/examples/xml/basic/tallies.xml
+++ /dev/null
@@ -1,31 +0,0 @@
-<?xml version="1.0"?>
-<tallies>
-
-  <filter id="1" type="cell">
-    <bins>100</bins>
-  </filter>
-
-  <filter id="2" type="energy">
-    <bins>0 20.0e6</bins>
-  </filter>
-
-  <filter id="3" type="energyout">
-    <bins>0 20.0e6</bins>
-  </filter>
-
-  <tally id="1">
-    <filters>1</filters>
-    <scores>total scatter nu-scatter absorption fission nu-fission</scores>
-  </tally>
-
-  <tally id="2">
-    <filters>1 2</filters>
-    <scores>total scatter nu-scatter absorption fission nu-fission</scores>
-  </tally>
-
-  <tally id="3">
-    <filters>1 2 3</filters>
-    <scores>scatter nu-scatter nu-fission</scores>
-  </tally>
-
-</tallies>
diff --git a/examples/xml/boxes/geometry.xml b/examples/xml/boxes/geometry.xml
deleted file mode 100644
index abe4924e6..000000000
--- a/examples/xml/boxes/geometry.xml
+++ /dev/null
@@ -1,39 +0,0 @@
-<?xml version="1.0"?>
-<geometry>
-
-  <!--
-      This example consists of three nested boxes, and is meant to show how to
-      use Boolean operators to construct complex cell regions.
-  -->
-
-  <surface id="1" type="x-plane" coeffs="-10" boundary="vacuum" />
-  <surface id="2" type="x-plane" coeffs="-7" />
-  <surface id="3" type="x-plane" coeffs="-4" />
-  <surface id="4" type="x-plane" coeffs="4" />
-  <surface id="5" type="x-plane" coeffs="7" />
-  <surface id="6" type="x-plane" coeffs="10" boundary="vacuum" />
-
-  <surface id="11" type="y-plane" coeffs="-10" boundary="vacuum" />
-  <surface id="12" type="y-plane" coeffs="-7" />
-  <surface id="13" type="y-plane" coeffs="-4" />
-  <surface id="14" type="y-plane" coeffs="4" />
-  <surface id="15" type="y-plane" coeffs="7" />
-  <surface id="16" type="y-plane" coeffs="10" boundary="vacuum" />
-
-  <surface id="21" type="z-plane" coeffs="-10" boundary="vacuum" />
-  <surface id="22" type="z-plane" coeffs="-7" />
-  <surface id="23" type="z-plane" coeffs="-4" />
-  <surface id="24" type="z-plane" coeffs="4" />
-  <surface id="25" type="z-plane" coeffs="7" />
-  <surface id="26" type="z-plane" coeffs="10" boundary="vacuum" />
-
-  <!-- Innermost cube -->
-  <cell id="1" material="1" region="3 -4 13 -14 23 -24" />
-
-  <!-- Middle cubic shell -->
-  <cell id="2" material="2" region="2 -5 12 -15 22 -25 (-3 | 4 | -13 | 14 | -23 | 24)" />
-
-  <!-- Outermost cubic shell -->
-  <cell id="3" material="3" region="1 -6 11 -16 21 -26 ~(2 -5 12 -15 22 -25)" />
-
-</geometry>
diff --git a/examples/xml/boxes/materials.xml b/examples/xml/boxes/materials.xml
deleted file mode 100644
index 1d0ab4a1c..000000000
--- a/examples/xml/boxes/materials.xml
+++ /dev/null
@@ -1,21 +0,0 @@
-<?xml version="1.0"?>
-<materials>
-
-  <material id="1">
-    <density value="4.5" units="g/cc" />
-    <nuclide name="U235" ao="1.0" />
-  </material>
-
-  <material id="2">
-    <density value="4.5" units="g/cc" />
-    <nuclide name="U238" ao="1.0" />
-  </material>
-
-  <material id="3">
-    <density value="1.0" units="g/cc" />
-    <nuclide name="O16" ao="1.0" />
-    <nuclide name="H1" ao="2.0" />
-    <sab name="c_H_in_H2O" />
-  </material>
-
-</materials>
diff --git a/examples/xml/boxes/plots.xml b/examples/xml/boxes/plots.xml
deleted file mode 100644
index 0a75f574e..000000000
--- a/examples/xml/boxes/plots.xml
+++ /dev/null
@@ -1,9 +0,0 @@
-<?xml version="1.0"?>
-<plots>
-  <plot id="1" type="slice">
-    <color_by>cell</color_by>
-    <origin>0. 0. 0.</origin>
-    <width>20. 20.</width>
-    <pixels>200 200</pixels>
-  </plot>
-</plots>
diff --git a/examples/xml/boxes/settings.xml b/examples/xml/boxes/settings.xml
deleted file mode 100644
index 9007a12c5..000000000
--- a/examples/xml/boxes/settings.xml
+++ /dev/null
@@ -1,15 +0,0 @@
-<?xml version="1.0"?>
-<settings>
-
-  <!-- Parameters for k-eigenvalue calculation -->
-  <run_mode>eigenvalue</run_mode>
-  <batches>15</batches>
-  <inactive>5</inactive>
-  <particles>10000</particles>
-
-  <!-- Starting source -->
-  <source>
-    <space type="box" parameters="-10. -10. -10. 10. 10. 10." />
-  </source>
-
-</settings>