diff --git a/examples/python/pincell_depletion/restart_depletion.py b/examples/python/pincell_depletion/restart_depletion.py
new file mode 100644
index 000000000..6f30c68ea
--- /dev/null
+++ b/examples/python/pincell_depletion/restart_depletion.py
@@ -0,0 +1,78 @@
+import openmc
+import openmc.deplete
+import numpy as np
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 100
+inactive = 10
+particles = 1000
+
+# Depletion simulation parameters
+time_step = 1*24*60*60 # s
+final_time = 15*24*60*60 # s
+time_steps = np.full(final_time // time_step, time_step)
+
+chain_file = './chain_simple.xml'
+power = 174 # W/cm, for 2D simulations only (use W for 3D)
+
+###############################################################################
+#                      Load previous simulation results
+###############################################################################
+
+# Load geometry from statepoint
+statepoint = 'statepoint.100.h5'
+sp = openmc.StatePoint(statepoint)
+geometry = sp.summary.geometry
+
+# Load previous delpletion results
+previous_results = openmc.deplete.ResultsList("depletion_results.h5")
+
+# Reload volumes into geometry
+previous_results.transfer_volumes(geometry)
+
+###############################################################################
+#                   Set transport calculation settings
+###############################################################################
+
+# 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 = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+entropy_mesh = openmc.Mesh()
+entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
+entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
+entropy_mesh.dimension = [10, 10, 1]
+settings_file.entropy_mesh = entropy_mesh
+
+###############################################################################
+#                   Initialize and run depletion calculation
+###############################################################################
+
+op = openmc.deplete.Operator(geometry, settings_file, chain_file)
+
+# Perform simulation using the predictor algorithm
+openmc.deplete.integrator.predictor(op, time_steps, power)
+
+###############################################################################
+#                    Read depletion calculation results
+###############################################################################
+
+# Open results file
+results = openmc.deplete.ResultsList("depletion_results.h5")
+
+# Obtain K_eff as a function of time
+time, keff = results.get_eigenvalue()
+                                                                                                                                                                                                                                                                                                                                                                                                                                                        
+# Obtain U235 concentration as a function of time
+time, n_U235 = results.get_atoms('1', 'U235')