diff --git a/.gitignore b/.gitignore
index 8a8e189..a5a08f8 100644
--- a/.gitignore
+++ b/.gitignore
@@ -98,4 +98,9 @@ ENV/
 *.png
 
 # Temp files
-*.xml~
\ No newline at end of file
+*.xml~
+
+# Cobalt
+*.output
+*.error
+*.cobaltlog
\ No newline at end of file
diff --git a/2x2-periodic/build-xml.py b/2x2-periodic/build-xml.py
index 8ef379f..6d73dc5 100644
--- a/2x2-periodic/build-xml.py
+++ b/2x2-periodic/build-xml.py
@@ -170,6 +170,10 @@ openmc_geometry.export_to_xml()
 
 #### Create OpenMC "settings.xml" file
 
+# Query the user on whether to use multipole cross sections
+multipole = input('Use multipole cross sections? (y/n): ').lower()
+multipole = True if multipole == 'y' else False
+
 # Construct uniform initial source distribution over fissionable zones
 lower_left = two_by_two.bounds[:3]
 upper_right = two_by_two.bounds[3:]
@@ -184,6 +188,10 @@ settings_file.ptables = True
 settings_file.output = {'tallies': False}
 settings_file.source = source
 settings_file.sourcepoint_write = False
+
+if multipole:
+    settings_file.temperature = {'multipole': True, 'tolerance': 1000}
+
 settings_file.export_to_xml()
 
 
@@ -276,4 +284,4 @@ capture_rates.scores = ['absorption', 'fission']
 tallies_file = openmc.Tallies([fission_rates, capture_rates])
 cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
 mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
-tallies_file.export_to_xml()
\ No newline at end of file
+tallies_file.export_to_xml()
diff --git a/2x2-reflector/build-xml.py b/2x2-reflector/build-xml.py
index c2f2cd2..04424e8 100644
--- a/2x2-reflector/build-xml.py
+++ b/2x2-reflector/build-xml.py
@@ -178,6 +178,10 @@ openmc_geometry.export_to_xml()
 
 #### Create OpenMC "settings.xml" file
 
+# Query the user on whether to use multipole cross sections
+multipole = input('Use multipole cross sections? (y/n): ').lower()
+multipole = True if multipole == 'y' else False
+
 # Construct uniform initial source distribution over fissionable zones
 lower_left = reflector.bounds[:3]
 upper_right = reflector.bounds[3:]
@@ -196,6 +200,10 @@ settings_file.ptables = True
 settings_file.output = {'tallies': False}
 settings_file.source = source
 settings_file.sourcepoint_write = False
+
+if multipole:
+    settings_file.temperature = {'multipole': True, 'tolerance': 1000}
+
 settings_file.export_to_xml()
 
 
@@ -288,4 +296,4 @@ capture_rates.scores = ['absorption', 'fission']
 tallies_file = openmc.Tallies([fission_rates, capture_rates])
 cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
 mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
-tallies_file.export_to_xml()
\ No newline at end of file
+tallies_file.export_to_xml()
diff --git a/assembly/build-xml.py b/assembly/build-xml.py
index c70f5fb..3a4f5cb 100644
--- a/assembly/build-xml.py
+++ b/assembly/build-xml.py
@@ -90,6 +90,10 @@ openmc_geometry.export_to_xml()
 
 #### Create OpenMC "settings.xml" file
 
+# Query the user on whether to use multipole cross sections
+multipole = input('Use multipole cross sections? (y/n): ').lower()
+multipole = True if multipole == 'y' else False
+
 # Construct uniform initial source distribution over fissionable zones
 lower_left = fuel_assembly.bounds[:3]
 upper_right = fuel_assembly.bounds[3:]
@@ -104,6 +108,10 @@ settings_file.ptables = True
 settings_file.output = {'tallies': False}
 settings_file.source = source
 settings_file.sourcepoint_write = False
+
+if multipole:
+    settings_file.temperature = {'multipole': True, 'tolerance': 1000}
+
 settings_file.export_to_xml()
 
 
@@ -196,4 +204,4 @@ capture_rates.scores = ['absorption', 'fission']
 tallies_file = openmc.Tallies([fission_rates, capture_rates])
 cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
 mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
-tallies_file.export_to_xml()
\ No newline at end of file
+tallies_file.export_to_xml()
diff --git a/pin-cell/build-xml.py b/pin-cell/build-xml.py
index 383fac4..411e1ef 100644
--- a/pin-cell/build-xml.py
+++ b/pin-cell/build-xml.py
@@ -90,6 +90,10 @@ openmc_geometry.export_to_xml()
 
 #### Create OpenMC "settings.xml" file
 
+# Query the user on whether to use multipole cross sections
+multipole = input('Use multipole cross sections? (y/n): ').lower()
+multipole = True if multipole == 'y' else False
+
 # Construct uniform initial source distribution over fissionable zones
 lower_left = pin_geometry.bounds[:2] + [-10.]
 upper_right = pin_geometry.bounds[3:5] + [10.]
@@ -104,6 +108,10 @@ settings_file.ptables = True
 settings_file.output = {'tallies': False}
 settings_file.source = source
 settings_file.sourcepoint_write = False
+
+if multipole:
+    settings_file.temperature = {'multipole': True, 'tolerance': 1000}
+
 settings_file.export_to_xml()
 
 
@@ -153,4 +161,4 @@ mgxs_lib.build_library()
 # Create a "tallies.xml" file for the MGXS Library
 tallies_file = openmc.Tallies()
 mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
-tallies_file.export_to_xml()
\ No newline at end of file
+tallies_file.export_to_xml()