We are currently solving the following issue within our repository. Here is the issue text:
--- BEGIN ISSUE ---
Unable to run examples/python/pincell
Hi,

After generating the xml files and trying to `openmc` I get the following error:
```
Reading tallies XML file...
ERROR: Two or more meshes use the same unique ID: 1
```

--- END ISSUE ---

Below are some code segments, each from a relevant file. One or more of these files may contain bugs.
--- BEGIN FILES ---
Path: `examples/python/pincell/build-xml.py`
Content:
```
1 import openmc
2 
3 ###############################################################################
4 #                      Simulation Input File Parameters
5 ###############################################################################
6 
7 # OpenMC simulation parameters
8 batches = 100
9 inactive = 10
10 particles = 1000
11 
12 
13 ###############################################################################
14 #                 Exporting to OpenMC materials.xml file
15 ###############################################################################
16 
17 
18 # Instantiate some Materials and register the appropriate Nuclides
19 uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
20 uo2.set_density('g/cm3', 10.29769)
21 uo2.add_element('U', 1., enrichment=2.4)
22 uo2.add_element('O', 2.)
23 
24 helium = openmc.Material(material_id=2, name='Helium for gap')
25 helium.set_density('g/cm3', 0.001598)
26 helium.add_element('He', 2.4044e-4)
27 
28 zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
29 zircaloy.set_density('g/cm3', 6.55)
30 zircaloy.add_element('Sn', 0.014  , 'wo')
31 zircaloy.add_element('Fe', 0.00165, 'wo')
32 zircaloy.add_element('Cr', 0.001  , 'wo')
33 zircaloy.add_element('Zr', 0.98335, 'wo')
34 
35 borated_water = openmc.Material(material_id=4, name='Borated water')
36 borated_water.set_density('g/cm3', 0.740582)
37 borated_water.add_element('B', 4.0e-5)
38 borated_water.add_element('H', 5.0e-2)
39 borated_water.add_element('O', 2.4e-2)
40 borated_water.add_s_alpha_beta('c_H_in_H2O')
41 
42 # Instantiate a Materials collection and export to XML
43 materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
44 materials_file.export_to_xml()
45 
46 
47 ###############################################################################
48 #                 Exporting to OpenMC geometry.xml file
49 ###############################################################################
50 
51 # Instantiate ZCylinder surfaces
52 fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
53 clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
54 clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
55 left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
56 right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
57 bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
58 top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
59 
60 left.boundary_type = 'reflective'
61 right.boundary_type = 'reflective'
62 top.boundary_type = 'reflective'
63 bottom.boundary_type = 'reflective'
64 
65 # Instantiate Cells
66 fuel = openmc.Cell(cell_id=1, name='cell 1')
67 gap = openmc.Cell(cell_id=2, name='cell 2')
68 clad = openmc.Cell(cell_id=3, name='cell 3')
69 water = openmc.Cell(cell_id=4, name='cell 4')
70 
71 # Use surface half-spaces to define regions
72 fuel.region = -fuel_or
73 gap.region = +fuel_or & -clad_ir
74 clad.region = +clad_ir & -clad_or
75 water.region = +clad_or & +left & -right & +bottom & -top
76 
77 # Register Materials with Cells
78 fuel.fill = uo2
79 gap.fill = helium
80 clad.fill = zircaloy
81 water.fill = borated_water
82 
83 # Instantiate Universe
84 root = openmc.Universe(universe_id=0, name='root universe')
85 
86 # Register Cells with Universe
87 root.add_cells([fuel, gap, clad, water])
88 
89 # Instantiate a Geometry, register the root Universe, and export to XML
90 geometry = openmc.Geometry(root)
91 geometry.export_to_xml()
92 
93 
94 ###############################################################################
95 #                   Exporting to OpenMC settings.xml file
96 ###############################################################################
97 
98 # Instantiate a Settings object, set all runtime parameters, and export to XML
99 settings_file = openmc.Settings()
100 settings_file.batches = batches
101 settings_file.inactive = inactive
102 settings_file.particles = particles
103 
104 # Create an initial uniform spatial source distribution over fissionable zones
105 bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
106 uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
107 settings_file.source = openmc.source.Source(space=uniform_dist)
108 
109 entropy_mesh = openmc.Mesh()
110 entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
111 entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
112 entropy_mesh.dimension = [10, 10, 1]
113 settings_file.entropy_mesh = entropy_mesh
114 settings_file.export_to_xml()
115 
116 
117 ###############################################################################
118 #                   Exporting to OpenMC tallies.xml file
119 ###############################################################################
120 
121 # Instantiate a tally mesh
122 mesh = openmc.Mesh(mesh_id=1)
123 mesh.type = 'regular'
124 mesh.dimension = [100, 100, 1]
125 mesh.lower_left = [-0.62992, -0.62992, -1.e50]
126 mesh.upper_right = [0.62992, 0.62992, 1.e50]
127 
128 # Instantiate some tally Filters
129 energy_filter = openmc.EnergyFilter([0., 4., 20.e6])
130 mesh_filter = openmc.MeshFilter(mesh)
131 
132 # Instantiate the Tally
133 tally = openmc.Tally(tally_id=1, name='tally 1')
134 tally.filters = [energy_filter, mesh_filter]
135 tally.scores = ['flux', 'fission', 'nu-fission']
136 
137 # Instantiate a Tallies collection and export to XML
138 tallies_file = openmc.Tallies([tally])
139 tallies_file.export_to_xml()
140
```
--- END FILES ---

Please first localize the bug based on the issue statement, and then generate a patch according to the `git diff` format fenced by three backticks.

Here is an example:

```diff
diff --git a/examples/server_async.py b/examples/server_async.py
--- a/examples/server_async.py
+++ b/examples/server_async.py
@@ -313,4 +313,4 @@
 
 
 if __name__ == "__main__":
-    asyncio.run(run_async_server("."), debug=True)
+    asyncio.run(run_async_server(), debug=True)
diff --git a/examples/server_sync.py b/examples/server_sync.py
--- a/examples/server_sync.py
+++ b/examples/server_sync.py
@@ -313,5 +313,5 @@
 
 
 if __name__ == "__main__":
-    server = run_sync_server(".")
+    server = run_sync_server()
     server.shutdown()

```

diff --git a/examples/python/pincell/build-xml.py b/examples/python/pincell/build-xml.py
--- a/examples/python/pincell/build-xml.py
+++ b/examples/python/pincell/build-xml.py
@@ -119,7 +119,7 @@
 ###############################################################################
 
 # Instantiate a tally mesh
-mesh = openmc.Mesh(mesh_id=1)
+mesh = openmc.Mesh()
 mesh.type = 'regular'
 mesh.dimension = [100, 100, 1]
 mesh.lower_left = [-0.62992, -0.62992, -1.e50]

{"golden_diff": "diff --git a/examples/python/pincell/build-xml.py b/examples/python/pincell/build-xml.py\n--- a/examples/python/pincell/build-xml.py\n+++ b/examples/python/pincell/build-xml.py\n@@ -119,7 +119,7 @@\n ###############################################################################\n \n # Instantiate a tally mesh\n-mesh = openmc.Mesh(mesh_id=1)\n+mesh = openmc.Mesh()\n mesh.type = 'regular'\n mesh.dimension = [100, 100, 1]\n mesh.lower_left = [-0.62992, -0.62992, -1.e50]\n", "issue": "Unable to run examples/python/pincell\nHi,\r\n\r\nAfter generating the xml files and trying to `openmc` I get the following error:\r\n```\r\nReading tallies XML file...\r\nERROR: Two or more meshes use the same unique ID: 1\r\n```\n", "before_files": [{"content": "import openmc\n\n###############################################################################\n#                      Simulation Input File Parameters\n###############################################################################\n\n# OpenMC simulation parameters\nbatches = 100\ninactive = 10\nparticles = 1000\n\n\n###############################################################################\n#                 Exporting to OpenMC materials.xml file\n###############################################################################\n\n\n# Instantiate some Materials and register the appropriate Nuclides\nuo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')\nuo2.set_density('g/cm3', 10.29769)\nuo2.add_element('U', 1., enrichment=2.4)\nuo2.add_element('O', 2.)\n\nhelium = openmc.Material(material_id=2, name='Helium for gap')\nhelium.set_density('g/cm3', 0.001598)\nhelium.add_element('He', 2.4044e-4)\n\nzircaloy = openmc.Material(material_id=3, name='Zircaloy 4')\nzircaloy.set_density('g/cm3', 6.55)\nzircaloy.add_element('Sn', 0.014  , 'wo')\nzircaloy.add_element('Fe', 0.00165, 'wo')\nzircaloy.add_element('Cr', 0.001  , 'wo')\nzircaloy.add_element('Zr', 0.98335, 'wo')\n\nborated_water = openmc.Material(material_id=4, name='Borated water')\nborated_water.set_density('g/cm3', 0.740582)\nborated_water.add_element('B', 4.0e-5)\nborated_water.add_element('H', 5.0e-2)\nborated_water.add_element('O', 2.4e-2)\nborated_water.add_s_alpha_beta('c_H_in_H2O')\n\n# Instantiate a Materials collection and export to XML\nmaterials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])\nmaterials_file.export_to_xml()\n\n\n###############################################################################\n#                 Exporting to OpenMC geometry.xml file\n###############################################################################\n\n# Instantiate ZCylinder surfaces\nfuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')\nclad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')\nclad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')\nleft = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')\nright = openmc.XPlane(surface_id=5, x0=0.62992, name='right')\nbottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')\ntop = openmc.YPlane(surface_id=7, y0=0.62992, name='top')\n\nleft.boundary_type = 'reflective'\nright.boundary_type = 'reflective'\ntop.boundary_type = 'reflective'\nbottom.boundary_type = 'reflective'\n\n# Instantiate Cells\nfuel = openmc.Cell(cell_id=1, name='cell 1')\ngap = openmc.Cell(cell_id=2, name='cell 2')\nclad = openmc.Cell(cell_id=3, name='cell 3')\nwater = openmc.Cell(cell_id=4, name='cell 4')\n\n# Use surface half-spaces to define regions\nfuel.region = -fuel_or\ngap.region = +fuel_or & -clad_ir\nclad.region = +clad_ir & -clad_or\nwater.region = +clad_or & +left & -right & +bottom & -top\n\n# Register Materials with Cells\nfuel.fill = uo2\ngap.fill = helium\nclad.fill = zircaloy\nwater.fill = borated_water\n\n# Instantiate Universe\nroot = openmc.Universe(universe_id=0, name='root universe')\n\n# Register Cells with Universe\nroot.add_cells([fuel, gap, clad, water])\n\n# Instantiate a Geometry, register the root Universe, and export to XML\ngeometry = openmc.Geometry(root)\ngeometry.export_to_xml()\n\n\n###############################################################################\n#                   Exporting to OpenMC settings.xml file\n###############################################################################\n\n# Instantiate a Settings object, set all runtime parameters, and export to XML\nsettings_file = openmc.Settings()\nsettings_file.batches = batches\nsettings_file.inactive = inactive\nsettings_file.particles = particles\n\n# Create an initial uniform spatial source distribution over fissionable zones\nbounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]\nuniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\nsettings_file.source = openmc.source.Source(space=uniform_dist)\n\nentropy_mesh = openmc.Mesh()\nentropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]\nentropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]\nentropy_mesh.dimension = [10, 10, 1]\nsettings_file.entropy_mesh = entropy_mesh\nsettings_file.export_to_xml()\n\n\n###############################################################################\n#                   Exporting to OpenMC tallies.xml file\n###############################################################################\n\n# Instantiate a tally mesh\nmesh = openmc.Mesh(mesh_id=1)\nmesh.type = 'regular'\nmesh.dimension = [100, 100, 1]\nmesh.lower_left = [-0.62992, -0.62992, -1.e50]\nmesh.upper_right = [0.62992, 0.62992, 1.e50]\n\n# Instantiate some tally Filters\nenergy_filter = openmc.EnergyFilter([0., 4., 20.e6])\nmesh_filter = openmc.MeshFilter(mesh)\n\n# Instantiate the Tally\ntally = openmc.Tally(tally_id=1, name='tally 1')\ntally.filters = [energy_filter, mesh_filter]\ntally.scores = ['flux', 'fission', 'nu-fission']\n\n# Instantiate a Tallies collection and export to XML\ntallies_file = openmc.Tallies([tally])\ntallies_file.export_to_xml()\n", "path": "examples/python/pincell/build-xml.py"}], "after_files": [{"content": "import openmc\n\n###############################################################################\n#                      Simulation Input File Parameters\n###############################################################################\n\n# OpenMC simulation parameters\nbatches = 100\ninactive = 10\nparticles = 1000\n\n\n###############################################################################\n#                 Exporting to OpenMC materials.xml file\n###############################################################################\n\n\n# Instantiate some Materials and register the appropriate Nuclides\nuo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')\nuo2.set_density('g/cm3', 10.29769)\nuo2.add_element('U', 1., enrichment=2.4)\nuo2.add_element('O', 2.)\n\nhelium = openmc.Material(material_id=2, name='Helium for gap')\nhelium.set_density('g/cm3', 0.001598)\nhelium.add_element('He', 2.4044e-4)\n\nzircaloy = openmc.Material(material_id=3, name='Zircaloy 4')\nzircaloy.set_density('g/cm3', 6.55)\nzircaloy.add_element('Sn', 0.014  , 'wo')\nzircaloy.add_element('Fe', 0.00165, 'wo')\nzircaloy.add_element('Cr', 0.001  , 'wo')\nzircaloy.add_element('Zr', 0.98335, 'wo')\n\nborated_water = openmc.Material(material_id=4, name='Borated water')\nborated_water.set_density('g/cm3', 0.740582)\nborated_water.add_element('B', 4.0e-5)\nborated_water.add_element('H', 5.0e-2)\nborated_water.add_element('O', 2.4e-2)\nborated_water.add_s_alpha_beta('c_H_in_H2O')\n\n# Instantiate a Materials collection and export to XML\nmaterials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])\nmaterials_file.export_to_xml()\n\n\n###############################################################################\n#                 Exporting to OpenMC geometry.xml file\n###############################################################################\n\n# Instantiate ZCylinder surfaces\nfuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')\nclad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')\nclad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')\nleft = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')\nright = openmc.XPlane(surface_id=5, x0=0.62992, name='right')\nbottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')\ntop = openmc.YPlane(surface_id=7, y0=0.62992, name='top')\n\nleft.boundary_type = 'reflective'\nright.boundary_type = 'reflective'\ntop.boundary_type = 'reflective'\nbottom.boundary_type = 'reflective'\n\n# Instantiate Cells\nfuel = openmc.Cell(cell_id=1, name='cell 1')\ngap = openmc.Cell(cell_id=2, name='cell 2')\nclad = openmc.Cell(cell_id=3, name='cell 3')\nwater = openmc.Cell(cell_id=4, name='cell 4')\n\n# Use surface half-spaces to define regions\nfuel.region = -fuel_or\ngap.region = +fuel_or & -clad_ir\nclad.region = +clad_ir & -clad_or\nwater.region = +clad_or & +left & -right & +bottom & -top\n\n# Register Materials with Cells\nfuel.fill = uo2\ngap.fill = helium\nclad.fill = zircaloy\nwater.fill = borated_water\n\n# Instantiate Universe\nroot = openmc.Universe(universe_id=0, name='root universe')\n\n# Register Cells with Universe\nroot.add_cells([fuel, gap, clad, water])\n\n# Instantiate a Geometry, register the root Universe, and export to XML\ngeometry = openmc.Geometry(root)\ngeometry.export_to_xml()\n\n\n###############################################################################\n#                   Exporting to OpenMC settings.xml file\n###############################################################################\n\n# Instantiate a Settings object, set all runtime parameters, and export to XML\nsettings_file = openmc.Settings()\nsettings_file.batches = batches\nsettings_file.inactive = inactive\nsettings_file.particles = particles\n\n# Create an initial uniform spatial source distribution over fissionable zones\nbounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]\nuniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\nsettings_file.source = openmc.source.Source(space=uniform_dist)\n\nentropy_mesh = openmc.Mesh()\nentropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]\nentropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]\nentropy_mesh.dimension = [10, 10, 1]\nsettings_file.entropy_mesh = entropy_mesh\nsettings_file.export_to_xml()\n\n\n###############################################################################\n#                   Exporting to OpenMC tallies.xml file\n###############################################################################\n\n# Instantiate a tally mesh\nmesh = openmc.Mesh()\nmesh.type = 'regular'\nmesh.dimension = [100, 100, 1]\nmesh.lower_left = [-0.62992, -0.62992, -1.e50]\nmesh.upper_right = [0.62992, 0.62992, 1.e50]\n\n# Instantiate some tally Filters\nenergy_filter = openmc.EnergyFilter([0., 4., 20.e6])\nmesh_filter = openmc.MeshFilter(mesh)\n\n# Instantiate the Tally\ntally = openmc.Tally(tally_id=1, name='tally 1')\ntally.filters = [energy_filter, mesh_filter]\ntally.scores = ['flux', 'fission', 'nu-fission']\n\n# Instantiate a Tallies collection and export to XML\ntallies_file = openmc.Tallies([tally])\ntallies_file.export_to_xml()\n", "path": "examples/python/pincell/build-xml.py"}]}