diff --git a/examples/python/pincell/build-xml.py b/examples/python/pincell/build-xml.py
new file mode 100644
index 000000000..6916bef9c
--- /dev/null
+++ b/examples/python/pincell/build-xml.py
@@ -0,0 +1,214 @@
+import openmc
+
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 100
+inactive = 10
+particles = 1000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml File
+###############################################################################
+
+# Instantiate some Nuclides
+h1 = openmc.Nuclide('H-1')
+h2 = openmc.Nuclide('H-2')
+he4 = openmc.Nuclide('He-4')
+b10 = openmc.Nuclide('B-10')
+b11 = openmc.Nuclide('B-11')
+o16 = openmc.Nuclide('O-16')
+o17 = openmc.Nuclide('O-17')
+cr50 = openmc.Nuclide('Cr-50')
+cr52 = openmc.Nuclide('Cr-52')
+cr53 = openmc.Nuclide('Cr-53')
+cr54 = openmc.Nuclide('Cr-54')
+fe54 = openmc.Nuclide('Fe-54')
+fe56 = openmc.Nuclide('Fe-56')
+fe57 = openmc.Nuclide('Fe-57')
+fe58 = openmc.Nuclide('Fe-58')
+zr90 = openmc.Nuclide('Zr-90')
+zr91 = openmc.Nuclide('Zr-91')
+zr92 = openmc.Nuclide('Zr-92')
+zr94 = openmc.Nuclide('Zr-94')
+zr96 = openmc.Nuclide('Zr-96')
+sn112 = openmc.Nuclide('Sn-112')
+sn114 = openmc.Nuclide('Sn-114')
+sn115 = openmc.Nuclide('Sn-115')
+sn116 = openmc.Nuclide('Sn-116')
+sn117 = openmc.Nuclide('Sn-117')
+sn118 = openmc.Nuclide('Sn-118')
+sn119 = openmc.Nuclide('Sn-119')
+sn120 = openmc.Nuclide('Sn-120')
+sn122 = openmc.Nuclide('Sn-122')
+sn124 = openmc.Nuclide('Sn-124')
+u234 = openmc.Nuclide('U-234')
+u235 = openmc.Nuclide('U-235')
+u238 = openmc.Nuclide('U-238')
+
+# Instantiate some Materials and register the appropriate Nuclides
+uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
+uo2.setDensity('g/cm3', 10.29769)
+uo2.addNuclide(u234, 4.4843e-6)
+uo2.addNuclide(u235, 5.5815e-4)
+uo2.addNuclide(u238, 2.2408e-2)
+uo2.addNuclide(o16, 4.5829e-2)
+uo2.addNuclide(o17, 1.1164e-4)
+
+helium = openmc.Material(material_id=2, name='Helium for gap')
+helium.setDensity('g/cm3', 0.001598)
+helium.addNuclide(he4, 2.4044e-4)
+
+zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
+zircaloy.setDensity('g/cm3', 6.55)
+zircaloy.addNuclide(o16, 3.0743e-4)
+zircaloy.addNuclide(o17, 7.4887e-7)
+zircaloy.addNuclide(cr50, 3.2962e-6)
+zircaloy.addNuclide(cr52, 6.3564e-5)
+zircaloy.addNuclide(cr53, 7.2076e-6)
+zircaloy.addNuclide(cr54, 1.7941e-6)
+zircaloy.addNuclide(fe54, 8.6699e-6)
+zircaloy.addNuclide(fe56, 1.3610e-4)
+zircaloy.addNuclide(fe57, 3.1431e-6)
+zircaloy.addNuclide(fe58, 4.1829e-7)
+zircaloy.addNuclide(zr90, 2.1827e-2)
+zircaloy.addNuclide(zr91, 4.7600e-3)
+zircaloy.addNuclide(zr92, 7.2758e-3)
+zircaloy.addNuclide(zr94, 7.3734e-3)
+zircaloy.addNuclide(zr96, 1.1879e-3)
+zircaloy.addNuclide(sn112, 4.6735e-6)
+zircaloy.addNuclide(sn114, 3.1799e-6)
+zircaloy.addNuclide(sn115, 1.6381e-6)
+zircaloy.addNuclide(sn116, 7.0055e-5)
+zircaloy.addNuclide(sn117, 3.7003e-5)
+zircaloy.addNuclide(sn118, 1.1669e-4)
+zircaloy.addNuclide(sn119, 4.1387e-5)
+zircaloy.addNuclide(sn120, 1.5697e-4)
+zircaloy.addNuclide(sn122, 2.2308e-5)
+zircaloy.addNuclide(sn124, 2.7897e-5)
+
+borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
+borated_water.setDensity('g/cm3', 0.740582)
+borated_water.addNuclide(b10, 8.0042e-6)
+borated_water.addNuclide(b11, 3.2218e-5)
+borated_water.addNuclide(h1, 4.9457e-2)
+borated_water.addNuclide(h2, 7.4196e-6)
+borated_water.addNuclide(o16, 2.4672e-2)
+borated_water.addNuclide(o17, 6.0099e-5)
+borated_water.addSAlphaBeta('HH2O', '71t')
+
+# Instantiate a MaterialsFile, register all Materials, and export to XML
+materials_file = openmc.MaterialsFile()
+materials_file.setDefaultXS('71c')
+materials_file.addMaterials([uo2, helium, zircaloy, borated_water])
+materials_file.exportToXML()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml File
+###############################################################################
+
+# Instantiate ZCylinder surfaces
+fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
+clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
+clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
+left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
+right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
+bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
+top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
+
+left.setBoundaryType('reflective')
+right.setBoundaryType('reflective')
+top.setBoundaryType('reflective')
+bottom.setBoundaryType('reflective')
+
+# Instantiate Cells
+fuel = openmc.Cell(cell_id=1, name='cell 1')
+gap = openmc.Cell(cell_id=2, name='cell 2')
+clad = openmc.Cell(cell_id=3, name='cell 3')
+water = openmc.Cell(cell_id=4, name='cell 4')
+
+# Register Surfaces with Cells
+fuel.addSurface(fuel_or, halfspace=-1)
+gap.addSurface(fuel_or, halfspace=+1)
+gap.addSurface(clad_ir, halfspace=-1)
+clad.addSurface(clad_ir, halfspace=+1)
+clad.addSurface(clad_or, halfspace=-1)
+water.addSurface(clad_or, halfspace=+1)
+water.addSurface(left, halfspace=+1)
+water.addSurface(right, halfspace=-1)
+water.addSurface(bottom, halfspace=+1)
+water.addSurface(top, halfspace=-1)
+
+# Register Materials with Cells
+fuel.setFill(uo2)
+gap.setFill(helium)
+clad.setFill(zircaloy)
+water.setFill(borated_water)
+
+# Instantiate Universe
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cells with Universe
+root.addCells([fuel, gap, clad, water])
+
+# Instantiate a Geometry and register the root Universe
+geometry = openmc.Geometry()
+geometry.setRootUniverse(root)
+
+# Instantiate a GeometryFile, register Geometry, and export to XML
+geometry_file = openmc.GeometryFile()
+geometry_file.setGeometry(geometry)
+geometry_file.exportToXML()
+
+
+###############################################################################
+#                   Exporting to OpenMC settings.xml File
+###############################################################################
+
+# Instantiate a SettingsFile, set all runtime parameters, and export to XML
+settings_file = openmc.SettingsFile()
+settings_file.setBatches(batches)
+settings_file.setInactive(inactive)
+settings_file.setParticles(particles)
+settings_file.setSourceSpace('box', [-0.62992, -0.62992, -1, \
+                                      0.62992, 0.62992, 1])
+settings_file.setEntropyLowerLeft([-0.39218, -0.39218, -1.e50])
+settings_file.setEntropyUpperRight([0.39218, 0.39218, 1.e50])
+settings_file.setEntropyDimension([10, 10, 1])
+settings_file.exportToXML()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml File
+###############################################################################
+
+# Instantiate a tally mesh
+mesh = openmc.Mesh(mesh_id=1)
+mesh.setType('rectangular')
+mesh.setDimension([100, 100, 1])
+mesh.setLowerLeft([-0.62992, -0.62992, -1.e50])
+mesh.setUpperRight([0.62992, 0.62992, 1.e50])
+
+# Instantiate some tally Filters
+energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.])
+mesh_filter = openmc.Filter()
+mesh_filter.setMesh(mesh)
+
+# Instantiate the Tally
+tally = openmc.Tally(tally_id=1)
+tally.addFilter(energy_filter)
+tally.addFilter(mesh_filter)
+tally.addScore('flux')
+tally.addScore('fission')
+tally.addScore('nu-fission')
+
+# Instantiate a TalliesFile, register all Tallies, and export to XML
+tallies_file = openmc.TalliesFile()
+tallies_file.addMesh(mesh)
+tallies_file.addTally(tally)
+tallies_file.exportToXML()
diff --git a/examples/python/pincell/geometry.xml b/examples/python/pincell/geometry.xml
deleted file mode 100644
index 53cb2f15d..000000000
--- a/examples/python/pincell/geometry.xml
+++ /dev/null
@@ -1,27 +0,0 @@
-<?xml version="1.0"?>
-<geometry>
-
-  <!--
-      This is a simple pin cell model based on dimensions from the MIT BEAVRS
-      (Benchmarking for Evaluation and Validation of Reactor Simulations)
-      benchmark.
-  -->
-
-  <!-- Surfaces for fuel, gap, cladding. Dimensions from Figure 2 in BEAVRS -->
-  <surface id="1" type="z-cylinder" coeffs="0. 0. 0.39218" /> <!-- Fuel OR -->
-  <surface id="2" type="z-cylinder" coeffs="0. 0. 0.40005" /> <!-- Clad IR -->
-  <surface id="3" type="z-cylinder" coeffs="0. 0. 0.45720" /> <!-- Clad OR -->
-
-  <!-- Reflective surfaces on outside of pin-cell. The lattice pitch is 1.25984
-       cm (taken from Table 2 in BEAVRS). -->
-  <surface id="4" type="x-plane" coeffs="-0.62992" boundary="reflective" />
-  <surface id="5" type="x-plane" coeffs=" 0.62992" boundary="reflective" />
-  <surface id="6" type="y-plane" coeffs="-0.62992" boundary="reflective" />
-  <surface id="7" type="y-plane" coeffs=" 0.62992" boundary="reflective" />
-
-  <cell id="1" material="1" surfaces="  -1" />        <!-- UO2 Fuel -->
-  <cell id="2" material="2" surfaces="1 -2" />        <!-- Helium gap -->
-  <cell id="3" material="3" surfaces="2 -3" />        <!-- Zircaloy cladding -->
-  <cell id="4" material="4" surfaces="3 4 -5 6 -7" /> <!-- Borated water -->
-
-</geometry>
diff --git a/examples/python/pincell/materials.xml b/examples/python/pincell/materials.xml
deleted file mode 100644
index 427fc175d..000000000
--- a/examples/python/pincell/materials.xml
+++ /dev/null
@@ -1,70 +0,0 @@
-<?xml version="1.0"?>
-<materials>
-
-  <!-- By default, use 300K cross sections -->
-  <default_xs>71c</default_xs>
-
-  <!--
-      Since O-18 is not present in ENDF/B-VII, it was necessary to combine the
-      atom densities for O-17 and O-18 in any materials containing Oxygen.
-  -->
-
-  <!-- UO2 fuel at 2.4 wt% enrichment -->
-  <material id="1">
-    <density value="10.29769" units="g/cm3" />
-    <nuclide name="U-234"  ao="4.4843e-06" />
-    <nuclide name="U-235"  ao="5.5815e-04" />
-    <nuclide name="U-238"  ao="2.2408e-02" />
-    <nuclide name="O-16"   ao="4.5829e-02" />
-    <nuclide name="O-17"   ao="1.1164e-04" />
-  </material>
-
-  <!-- Helium for gap -->
-  <material id="2">
-    <density value="0.001598" units="g/cm3" />
-    <nuclide name="He-4"   ao="2.4044e-04" />
-  </material>
-
-  <!-- Zircaloy 4 -->
-  <material id="3">
-    <density value="6.55" units="g/cm3" />
-    <nuclide name="O-16"   ao="3.0743e-04" />
-    <nuclide name="O-17"   ao="7.4887e-07" />
-    <nuclide name="Cr-50"  ao="3.2962e-06" />
-    <nuclide name="Cr-52"  ao="6.3564e-05" />
-    <nuclide name="Cr-53"  ao="7.2076e-06" />
-    <nuclide name="Cr-54"  ao="1.7941e-06" />
-    <nuclide name="Fe-54"  ao="8.6699e-06" />
-    <nuclide name="Fe-56"  ao="1.3610e-04" />
-    <nuclide name="Fe-57"  ao="3.1431e-06" />
-    <nuclide name="Fe-58"  ao="4.1829e-07" />
-    <nuclide name="Zr-90"  ao="2.1827e-02" />
-    <nuclide name="Zr-91"  ao="4.7600e-03" />
-    <nuclide name="Zr-92"  ao="7.2758e-03" />
-    <nuclide name="Zr-94"  ao="7.3734e-03" />
-    <nuclide name="Zr-96"  ao="1.1879e-03" />
-    <nuclide name="Sn-112" ao="4.6735e-06" />
-    <nuclide name="Sn-114" ao="3.1799e-06" />
-    <nuclide name="Sn-115" ao="1.6381e-06" />
-    <nuclide name="Sn-116" ao="7.0055e-05" />
-    <nuclide name="Sn-117" ao="3.7003e-05" />
-    <nuclide name="Sn-118" ao="1.1669e-04" />
-    <nuclide name="Sn-119" ao="4.1387e-05" />
-    <nuclide name="Sn-120" ao="1.5697e-04" />
-    <nuclide name="Sn-122" ao="2.2308e-05" />
-    <nuclide name="Sn-124" ao="2.7897e-05" />
-  </material>
-    
-  <!-- Borated water at 975 ppm -->
-  <material id="4">
-    <density value="0.740582" units="g/cm3" />
-    <nuclide name="B-10"  ao="8.0042e-06" />
-    <nuclide name="B-11"  ao="3.2218e-05" />
-    <nuclide name="H-1"   ao="4.9457e-02" />
-    <nuclide name="H-2"   ao="7.4196e-06" />
-    <nuclide name="O-16"  ao="2.4672e-02" />
-    <nuclide name="O-17"  ao="6.0099e-05" />
-    <sab name="HH2O" xs="71t" />
-  </material>
-
-</materials>
diff --git a/examples/python/pincell/settings.xml b/examples/python/pincell/settings.xml
deleted file mode 100644
index 443af9cde..000000000
--- a/examples/python/pincell/settings.xml
+++ /dev/null
@@ -1,32 +0,0 @@
-<?xml version="1.0"?>
-<settings>
-
-  <!-- Define how many particles to run and for how many batches -->
-  <eigenvalue>
-    <batches>100</batches>
-    <inactive>10</inactive>
-    <particles>1000</particles>
-  </eigenvalue>
-
-  <!-- The starting source is a uniform distribution over the entire pin
-       cell. Note that since this is effectively a 2D model, the z coordinates
-       are inconsequential -->
-  <source>
-    <space type="box">
-      <parameters>
-        -0.62992 -0.62992 -1.
-         0.62992  0.62992  1.
-      </parameters>
-    </space>
-  </source>
-
-  <!-- To assess convergence of the source distribution, we need to define the
-       bounds for a mesh over which the Shannon entropy should be
-       calculated. The extent in the z direction is made arbitrarily large. -->
-  <entropy>
-    <lower_left>-0.39218 -0.39218 -1.e50</lower_left>
-    <upper_right>0.39218  0.39218  1.e50</upper_right>
-    <dimension>10 10 1</dimension>
-  </entropy>
-
-</settings>
\ No newline at end of file
diff --git a/examples/python/pincell/tallies.xml b/examples/python/pincell/tallies.xml
deleted file mode 100644
index bbfd58836..000000000
--- a/examples/python/pincell/tallies.xml
+++ /dev/null
@@ -1,16 +0,0 @@
-<?xml version="1.0"?>
-<tallies>
-
-  <mesh id="1" type="rectangular">
-    <dimension>100 100 1</dimension>
-    <lower_left>-0.62992 -0.62992 -1.e50</lower_left>
-    <upper_right>0.62992  0.62992  1.e50</upper_right>
-  </mesh>
-
-  <tally id="1">
-    <filter type="mesh" bins="1" />
-    <filter type="energy" bins="0. 4.e-6 20.0" />
-    <scores>flux fission nu-fission</scores>
-  </tally>
-
-</tallies>
\ No newline at end of file