Update multipole docs and add a test

docs/source/usersguide/input.rst

@@ -987,6 +987,15 @@ Each ``<cell>`` element can have the following attributes or sub-elements:
 
     *Default*: A region filling all space.
 
+  :temperature:
+    The temperature of the cell in Kelvin. If windowed-multipole data is
+    avalable, this temperature will be used to Doppler broaden some cross
+    sections in the resolved resonance region. A list of temperatures can be
+    specified for the "distributed temperature" feature. This will give each
+    unique instance of the cell its own temperature.
+
+    *Default*: The temperature of the coldest nuclide in the cell's material(s)
+
   :rotation:
     If the cell is filled with a universe, this element specifies the angles in
     degrees about the x, y, and z axes that the filled universe should be

docs/source/usersguide/output/summary.rst

@@ -98,6 +98,10 @@ The current revision of the summary file format is 1.
     material.  The data is an array if the cell uses distributed materials,
     otherwise it is a scalar.
 
+**/geometry/cells/cell <uid>/temperature** (*double[]*)
+
+    Temperature of the cell in Kelvin.
+
 **/geometry/cells/cell <uid>/offset** (*int[]*)
 
     Offsets used for distribcell tally filter. This dataset is present only if

src/relaxng/geometry.rnc

@@ -9,6 +9,8 @@ element geometry {
       (element material { ( xsd:int | "void" )+ } |
         attribute material { ( xsd:int | "void" )+ })
     ) &
+    (element temperature { list { xsd:double+ } } |
+      attribute temperature { list { xsd:double+ } } )? &
     (element region { xsd:string } | attribute region { xsd:string })? &
     (element rotation { list { xsd:double+ } } | attribute rotation { list { xsd:double+ } })? &
     (element translation { list { xsd:double+ } } | attribute translation { list { xsd:double+ } })?

src/relaxng/geometry.rng

@@ -64,6 +64,24 @@
               </attribute>
             </choice>
           </choice>
+          <optional>
+            <choice>
+              <element name="temperature">
+                <list>
+                  <oneOrMore>
+                    <data type="double"/>
+                  </oneOrMore>
+                </list>
+              </element>
+              <attribute name="temperature">
+                <list>
+                  <oneOrMore>
+                    <data type="double"/>
+                  </oneOrMore>
+                </list>
+              </attribute>
+            </choice>
+          </optional>
           <optional>
             <choice>
               <element name="region">

tests/test_multipole/inputs_true.dat

@@ -0,0 +1 @@
+5c1cec635da5c4c869bdf58f62924a4cb1648e4ddecf0ce71b823cf45178767ba7f2e089b76d36e8616b1b21ffa43b32ab1b17d20bb74120f900b9e3e9ab9bcc

tests/test_multipole/results_true.dat

@@ -0,0 +1,12 @@
+k-combined:
+1.445285E+00 9.521660E-03
+Cell
+	ID             =	11
+	Name           =	
+	Material       =	2
+	Region         =	-10000
+	Temperature    =	[ 500.    0.  700.  800.]
+	Rotation       =	None
+	Translation    =	None
+	Offset         =	None
+	Distribcell index=	1

tests/test_multipole/test_multipole.py

@@ -0,0 +1,133 @@
+#!/usr/bin/env python
+
+import os
+import sys
+sys.path.insert(0, os.pardir)
+from testing_harness import TestHarness, PyAPITestHarness
+import openmc
+from openmc.stats import Box
+from openmc.source import Source
+
+
+class DistribmatTestHarness(PyAPITestHarness):
+    def _build_inputs(self):
+        ####################
+        # Materials
+        ####################
+
+        moderator = openmc.Material(material_id=1)
+        moderator.set_density('g/cc', 1.0)
+        moderator.add_nuclide('H-1', 2.0)
+        moderator.add_nuclide('O-16', 1.0)
+
+        dense_fuel = openmc.Material(material_id=2)
+        dense_fuel.set_density('g/cc', 4.5)
+        dense_fuel.add_nuclide('U-235', 1.0)
+
+        mats_file = openmc.MaterialsFile()
+        mats_file.default_xs = '71c'
+        mats_file.add_materials([moderator, dense_fuel])
+        mats_file.export_to_xml()
+
+
+        ####################
+        # Geometry
+        ####################
+
+        c1 = openmc.Cell(cell_id=1)
+        c1.fill = moderator
+        mod_univ = openmc.Universe(universe_id=1)
+        mod_univ.add_cell(c1)
+
+        r0 = openmc.ZCylinder(R=0.3)
+        c11 = openmc.Cell(cell_id=11)
+        c11.region = -r0
+        c11.fill = dense_fuel
+        c11.temperature = [500, 0, 700, 800]
+        c12 = openmc.Cell(cell_id=12)
+        c12.region = +r0
+        c12.fill = moderator
+        fuel_univ = openmc.Universe(universe_id=11)
+        fuel_univ.add_cells((c11, c12))
+
+        lat = openmc.RectLattice(lattice_id=101)
+        lat.dimension = [2, 2]
+        lat.lower_left = [-2.0, -2.0]
+        lat.pitch = [2.0, 2.0]
+        lat.universes = [[fuel_univ]*2]*2
+        lat.outer = mod_univ
+
+        x0 = openmc.XPlane(x0=-3.0)
+        x1 = openmc.XPlane(x0=3.0)
+        y0 = openmc.YPlane(y0=-3.0)
+        y1 = openmc.YPlane(y0=3.0)
+        for s in [x0, x1, y0, y1]:
+            s.boundary_type = 'reflective'
+        c101 = openmc.Cell(cell_id=101)
+        c101.region = +x0 & -x1 & +y0 & -y1
+        c101.fill = lat
+        root_univ = openmc.Universe(universe_id=0)
+        root_univ.add_cell(c101)
+
+        geometry = openmc.Geometry()
+        geometry.root_universe = root_univ
+        geo_file = openmc.GeometryFile()
+        geo_file.geometry = geometry
+        geo_file.export_to_xml()
+
+
+        ####################
+        # Settings
+        ####################
+
+        sets_file = openmc.SettingsFile()
+        sets_file.batches = 5
+        sets_file.inactive = 0
+        sets_file.particles = 1000
+        sets_file.source = Source(space=Box([-1, -1, -1], [1, 1, 1]))
+        sets_file.output = {'summary': True}
+        sets_file.export_to_xml()
+
+
+        ####################
+        # Plots
+        ####################
+
+        plots_file = openmc.PlotsFile()
+
+        plot = openmc.Plot(plot_id=1)
+        plot.basis = 'xy'
+        plot.color = 'cell'
+        plot.filename = 'cellplot'
+        plot.origin = (0, 0, 0)
+        plot.width = (7, 7)
+        plot.pixels = (400, 400)
+        plots_file.add_plot(plot)
+
+        plot = openmc.Plot(plot_id=2)
+        plot.basis = 'xy'
+        plot.color = 'mat'
+        plot.filename = 'matplot'
+        plot.origin = (0, 0, 0)
+        plot.width = (7, 7)
+        plot.pixels = (400, 400)
+        plots_file.add_plot(plot)
+
+        plots_file.export_to_xml()
+
+    def _get_results(self):
+        outstr = super(DistribmatTestHarness, self)._get_results()
+        su = openmc.Summary('summary.h5')
+        outstr += str(su.get_cell_by_id(11))
+        return outstr
+
+    def _cleanup(self):
+        f = os.path.join(os.getcwd(), 'plots.xml')
+        if os.path.exists(f):
+            os.remove(f)
+        super(DistribmatTestHarness, self)._cleanup()
+
+
+if __name__ == '__main__':
+    harness = DistribmatTestHarness('statepoint.5.*')
+    harness.main()