From 2e635c74f4d2d87237e6acc6b4718f8783f7202a Mon Sep 17 00:00:00 2001
From: Jin Whan Bae <baej@ornl.gov>
Date: Thu, 2 Nov 2023 17:52:12 -0400
Subject: [PATCH 01/14] microxs from mg flux and chain file

---
 openmc/deplete/microxs.py | 114 +++++++++++++++++++++++++++++++++++---
 1 file changed, 105 insertions(+), 9 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index e145bce8763..ed112520945 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -22,6 +22,29 @@
 _valid_rxns.append('fission')
 
 
+def resolve_chain_file_path(chain_file:str):
+    # Determine what reactions and nuclides are available in chain
+    if chain_file is None:
+        if 'chain_file' in openmc.config:
+            raise DataError(
+                "No depletion chain specified and could not find depletion "
+                "chain in openmc.config['chain_file']"
+            )
+        else:
+            chain_file = openmc.config['chain_file']
+    return chain_file
+
+
+def resolve_openmc_data_path(openmc_data_path:str=None):
+    if not openmc_data_path:
+        if 'cross_sections' not in openmc.config:
+            raise ValueError("`openmc_data_path` is not defined nor `openmc.config['cross_sections']` is defined.")
+        else:
+            openmc_data_path = openmc.config['cross_sections']
+    return openmc_data_path
+
+
+
 def get_microxs_and_flux(
         model: openmc.Model,
         domains,
@@ -69,13 +92,7 @@ def get_microxs_and_flux(
     original_tallies = model.tallies
 
     # Determine what reactions and nuclides are available in chain
-    if chain_file is None:
-        chain_file = openmc.config.get('chain_file')
-        if chain_file is None:
-            raise DataError(
-                "No depletion chain specified and could not find depletion "
-                "chain in openmc.config['chain_file']"
-            )
+    chain_file = resolve_chain_file_path(chain_file)
     chain = Chain.from_xml(chain_file)
     if reactions is None:
         reactions = chain.reactions
@@ -151,6 +168,87 @@ def get_microxs_and_flux(
     return fluxes, micros
 
 
+
+
+def get_microxs_from_mg_flux(energies: list or str, mg_flux: list, 
+                             chain_file_path:str=None, openmc_data_path:str=None,
+                             temperature:int=294):
+    """Generated MicroXS object from a known flux and a chain file. The size of the MicroXs matrix depends
+       on the chain file.
+
+    Args:
+        energies: iterable of float or str
+            Energy group boundaries in [eV] or the name of the group structure
+        mg_flux: iterable of float
+            Energy-dependent multigroup flux values
+        chain_file_path: str, optional
+            Path to the depletion chain XML file that will be used in
+            depletion simulation.
+        openmc_dat_path: str, optional
+            Path to the cross section XML file that contains data library paths.
+        temperature: int, optional
+            Temperature for cross section evaluation in [K].
+    """
+    # check energies
+    if isinstance(energies, str):
+        energies = openmc.EnergyFilter.from_group_structure(energies).values
+    
+    # check ascending (low to high)
+    if not all(energies[i] <= energies[i+1] for i in range(len(energies) - 1)):
+         raise ValueError('Energy bin must be in ascending order')
+    
+    # check dimension consistency
+    if not len(mg_flux) == len(energies)-1:
+        raise ValueError('Length of flux array should be len(energies)-1')
+    
+    chain_file_path = resolve_chain_file_path(chain_file_path)
+    chain = openmc.deplete.Chain.from_xml(chain_file_path)
+    
+    # resolve data library
+    openmc_data_path = resolve_openmc_data_path(openmc_data_path)
+    data_lib = openmc.data.DataLibrary.from_xml(openmc_data_path)
+
+    # get reactions and nuclides from chain file
+    nuclides, reactions = chain.nuclides, chain.reactions
+    nuclides = [nuc.name for nuc in nuclides]
+    if 'fission' in reactions:
+        # send to back
+        reactions.append(reactions.pop(reactions.index('fission')))
+    # convert reactions to mt file
+    #! I feel like this zero indexing will backfire
+    #! There has to be a better way for this
+    mts = [list(REACTIONS[reaction].mts)[0] for reaction in reactions if reaction != 'fission']
+    if 'fission' in reactions:
+        mts.append(18) # fission is not in the REACTIONS
+
+    # normalize flux and get energy midpoint
+    norm_mg_flux = mg_flux / sum(mg_flux)
+    energies_midpoint = [(energies[i]+energies[i+1])/2 for i in range(len(energies)-1)]
+    temperature_key = str(temperature) + 'K'
+    microxs_arr = np.zeros((len(nuclides), len(mts)))
+
+    for nuc_indx, nuc in enumerate(nuclides):
+        mat_lib = data_lib.get_by_material(nuc, data_type='neutron')
+        if not mat_lib: # file does not exist
+            microxs_arr[nuc_indx, :] = 0
+            continue
+        else:
+            hdf5_path = mat_lib['path']
+        nuc_data = openmc.data.IncidentNeutron.from_hdf5(hdf5_path)
+        for mt_indx, mt in enumerate(mts):
+            try: # sometimes it fails cause there's no entry
+                total_xs = np.array(nuc_data[mt].xs[temperature_key](energies_midpoint))
+                # collapse
+                total_norm = sum(total_xs * norm_mg_flux)
+                microxs_arr[nuc_indx, mt_indx] = total_norm            
+            except KeyError:
+                microxs_arr[nuc_indx, mt_indx] = 0.0
+    
+    return openmc.deplete.MicroXS.from_array(nuclides=nuclides,
+                                             reactions=reactions,
+                                             data=microxs_arr)
+
+
 class MicroXS:
     """Microscopic cross section data for use in transport-independent depletion.
 
@@ -192,8 +290,6 @@ def __init__(self, data: np.ndarray, nuclides: List[str], reactions: List[str]):
         self._index_nuc = {nuc: i for i, nuc in enumerate(nuclides)}
         self._index_rx = {rx: i for i, rx in enumerate(reactions)}
 
-    # TODO: Add a classmethod for generating MicroXS directly from cross section
-    # data using a known flux spectrum
 
     @classmethod
     def from_csv(cls, csv_file, **kwargs):

From e70c89647acd3cfc3c6025e36f1810ded766da33 Mon Sep 17 00:00:00 2001
From: shimwell <mail@jshimwell.com>
Date: Fri, 3 Nov 2023 00:02:14 +0000
Subject: [PATCH 02/14] converted to cls method added test

---
 openmc/deplete/microxs.py                | 123 +++++++++++++++++++++--
 tests/unit_tests/test_deplete_microxs.py |  15 +++
 2 files changed, 128 insertions(+), 10 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index e145bce8763..30b7e49f88e 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -6,7 +6,7 @@
 
 from __future__ import annotations
 import tempfile
-from typing import List, Tuple, Iterable, Optional, Union
+from typing import List, Tuple, Iterable, Optional, Union, Sequence
 
 import pandas as pd
 import numpy as np
@@ -22,6 +22,29 @@
 _valid_rxns.append('fission')
 
 
+def _resolve_chain_file_path(chain_file:str):
+    # Determine what reactions and nuclides are available in chain
+    if chain_file is None:
+        if 'chain_file' in openmc.config:
+            raise DataError(
+                "No depletion chain specified and could not find depletion "
+                "chain in openmc.config['chain_file']"
+            )
+        else:
+            chain_file = openmc.config['chain_file']
+    return chain_file
+
+
+def _resolve_openmc_data_path(openmc_data_path:str=None):
+    if not openmc_data_path:
+        if 'cross_sections' not in openmc.config:
+            raise ValueError("`openmc_data_path` is not defined nor `openmc.config['cross_sections']` is defined.")
+        else:
+            openmc_data_path = openmc.config['cross_sections']
+    return openmc_data_path
+
+
+
 def get_microxs_and_flux(
         model: openmc.Model,
         domains,
@@ -69,13 +92,7 @@ def get_microxs_and_flux(
     original_tallies = model.tallies
 
     # Determine what reactions and nuclides are available in chain
-    if chain_file is None:
-        chain_file = openmc.config.get('chain_file')
-        if chain_file is None:
-            raise DataError(
-                "No depletion chain specified and could not find depletion "
-                "chain in openmc.config['chain_file']"
-            )
+    chain_file = _resolve_chain_file_path(chain_file)
     chain = Chain.from_xml(chain_file)
     if reactions is None:
         reactions = chain.reactions
@@ -192,8 +209,94 @@ def __init__(self, data: np.ndarray, nuclides: List[str], reactions: List[str]):
         self._index_nuc = {nuc: i for i, nuc in enumerate(nuclides)}
         self._index_rx = {rx: i for i, rx in enumerate(reactions)}
 
-    # TODO: Add a classmethod for generating MicroXS directly from cross section
-    # data using a known flux spectrum
+
+    @classmethod
+    def from_multi_group_flux(
+        cls,
+        energies: Union[Iterable[float], str],
+        multi_group_flux: Sequence[float],
+        chain_file: Optional[PathLike] = None,
+        openmc_data_path: Optional[PathLike] = None,
+        temperature:int=294
+    ):
+        """Generated MicroXS object from a known flux and a chain file. The size of the MicroXs matrix depends
+        on the chain file.
+
+        Parameters
+        ----------
+        energies: iterable of float or str
+            Energy group boundaries in [eV] or the name of the group structure
+        multi_group_flux: iterable of float
+            Energy-dependent multigroup flux values
+        chain_file: str, optional
+            Path to the depletion chain XML file that will be used in
+            depletion simulation.
+        openmc_dat_path: str, optional
+            Path to the cross section XML file that contains data library paths.
+        temperature: int, optional
+            Temperature for cross section evaluation in [K].
+
+        Returns
+        -------
+        MicroXS
+        """
+        # check energies
+        if isinstance(energies, str):
+            energies = openmc.EnergyFilter.from_group_structure(energies).values
+        else:
+            # if user inputs energies check they are ascending (low to high) as some
+            # depletion codes use high energy to low energy.
+            if not all(energies[i] <= energies[i+1] for i in range(len(energies) - 1)):
+                raise ValueError('Energy bin must be in ascending order')
+
+        # check dimension consistency
+        if not len(multi_group_flux) == len(energies)-1:
+            raise ValueError('Length of flux array should be len(energies)-1')
+
+        chain_file_path = _resolve_chain_file_path(chain_file)
+        chain = openmc.deplete.Chain.from_xml(chain_file_path)
+
+        openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
+        data_lib = openmc.data.DataLibrary.from_xml(openmc_data_path)
+
+        # get reactions and nuclides from chain file
+        nuclides, reactions = chain.nuclides, chain.reactions
+        nuclides = [nuc.name for nuc in nuclides]
+        if 'fission' in reactions:
+            # send to back
+            reactions.append(reactions.pop(reactions.index('fission')))
+        # convert reactions to mt file
+        #! I feel like this zero indexing will backfire
+        #! There has to be a better way for this
+        mts = [list(REACTIONS[reaction].mts)[0] for reaction in reactions if reaction != 'fission']
+        if 'fission' in reactions:
+            mts.append(18) # fission is not in the REACTIONS
+
+        # normalize flux and get energy midpoint
+        norm_multi_group_flux = np.array(multi_group_flux) / sum(multi_group_flux)
+        energies_midpoint = [(energies[i]+energies[i+1])/2 for i in range(len(energies)-1)]
+        temperature_key = f'{temperature}K'
+        microxs_arr = np.zeros((len(nuclides), len(mts)))
+
+        for nuc_indx, nuc in enumerate(nuclides):
+            mat_lib = data_lib.get_by_material(nuc, data_type='neutron')
+            if not mat_lib: # file does not exist
+                microxs_arr[nuc_indx, :] = 0
+                continue
+            else:
+                hdf5_path = mat_lib['path']
+            nuc_data = openmc.data.IncidentNeutron.from_hdf5(hdf5_path)
+            for mt_indx, mt in enumerate(mts):
+                try: # sometimes it fails cause there's no entry
+                    # perhaps a multigroup mgxs should be made as chosing the xs at the energy bin center is not always fair
+                    total_xs = np.array(nuc_data[mt].xs[temperature_key](energies_midpoint))
+                    # collapse
+                    total_norm = sum(total_xs * norm_multi_group_flux)
+                    microxs_arr[nuc_indx, mt_indx] = total_norm
+                except KeyError:
+                    microxs_arr[nuc_indx, mt_indx] = 0.0
+
+        return cls(nuclides=nuclides, reactions=reactions, data=microxs_arr)
 
     @classmethod
     def from_csv(cls, csv_file, **kwargs):
diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index 582c584654d..b92fd9be27f 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -58,3 +58,18 @@ def test_csv():
     assert np.all(ref_xs.data == temp_xs.data)
     remove('temp_xs.csv')
 
+def test_from_multi_group_flux():
+
+    MicroXS.from_multi_group_flux(
+        energies='CASMO-4',
+        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        temperature='293',
+        chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
+    )
+
+    MicroXS.from_multi_group_flux(
+        energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
+        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        temperature='293',
+        chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
+    )

From 626618e32de6793254173661d044c6e6814b6f4d Mon Sep 17 00:00:00 2001
From: shimwell <mail@jshimwell.com>
Date: Fri, 3 Nov 2023 00:07:00 +0000
Subject: [PATCH 03/14] asserting correct type is returned

---
 tests/unit_tests/test_deplete_microxs.py | 6 ++++--
 1 file changed, 4 insertions(+), 2 deletions(-)

diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index b92fd9be27f..fcda08ad688 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -60,16 +60,18 @@ def test_csv():
 
 def test_from_multi_group_flux():
 
-    MicroXS.from_multi_group_flux(
+    microxs = MicroXS.from_multi_group_flux(
         energies='CASMO-4',
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature='293',
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
+    assert isinstance(microxs, MicroXS)
 
-    MicroXS.from_multi_group_flux(
+    microxs = MicroXS.from_multi_group_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature='293',
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
+    assert isinstance(microxs, MicroXS)

From b81245b9802f370ad9ccb22ddca316855ba2ff91 Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <drshimwell@gmail.com>
Date: Mon, 6 Nov 2023 20:10:29 +0000
Subject: [PATCH 04/14] Code review improvements by @yardasol

Co-authored-by: Olek <45364492+yardasol@users.noreply.github.com>
---
 openmc/deplete/microxs.py | 16 ++++++++--------
 1 file changed, 8 insertions(+), 8 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 0be1084b2c1..69e1afacb5d 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -35,7 +35,7 @@ def _resolve_chain_file_path(chain_file:str):
     return chain_file
 
 
-def _resolve_openmc_data_path(openmc_data_path:str=None):
+def _resolve_openmc_data_path(openmc_data_path: str=None):
     if not openmc_data_path:
         if 'cross_sections' not in openmc.config:
             raise ValueError("`openmc_data_path` is not defined nor `openmc.config['cross_sections']` is defined.")
@@ -216,23 +216,23 @@ def from_multi_group_flux(
         multi_group_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
         openmc_data_path: Optional[PathLike] = None,
-        temperature:int=294
+        temperature: int=294
     ):
-        """Generated MicroXS object from a known flux and a chain file. The size of the MicroXs matrix depends
+        """Generated MicroXS object from a known flux and a chain file. The size of the MicroXS matrix depends
         on the chain file.
 
         Parameters
         ----------
-        energies: iterable of float or str
+        energies : iterable of float or str
             Energy group boundaries in [eV] or the name of the group structure
-        multi_group_flux: iterable of float
+        multi_group_flux : iterable of float
             Energy-dependent multigroup flux values
-        chain_file: str, optional
+        chain_file : str, optional
             Path to the depletion chain XML file that will be used in
             depletion simulation.
-        openmc_dat_path: str, optional
+        openmc_data_path : str, optional
             Path to the cross section XML file that contains data library paths.
-        temperature: int, optional
+        temperature : int, optional
             Temperature for cross section evaluation in [K].
 
         Returns

From 9197ee04ffe4e641aa65bbc36eacf4b1c920c944 Mon Sep 17 00:00:00 2001
From: shimwell <mail@jshimwell.com>
Date: Tue, 28 Nov 2023 23:57:04 +0000
Subject: [PATCH 05/14] faster collapse with collapse_rate

---
 openmc/deplete/microxs.py                | 65 +++++++++++++++---------
 tests/unit_tests/test_deplete_microxs.py |  4 +-
 2 files changed, 42 insertions(+), 27 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 69e1afacb5d..35b210a01f5 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -5,7 +5,7 @@
 """
 
 from __future__ import annotations
-import tempfile
+from tempfile import TemporaryDirectory
 from typing import List, Tuple, Iterable, Optional, Union, Sequence
 
 import pandas as pd
@@ -17,6 +17,7 @@
 import openmc
 from .chain import Chain, REACTIONS
 from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
+import openmc.lib
 
 _valid_rxns = list(REACTIONS)
 _valid_rxns.append('fission')
@@ -132,7 +133,7 @@ def get_microxs_and_flux(
     model.tallies = openmc.Tallies([rr_tally, flux_tally])
 
     # create temporary run
-    with tempfile.TemporaryDirectory() as temp_dir:
+    with TemporaryDirectory() as temp_dir:
         if run_kwargs is None:
             run_kwargs = {}
         else:
@@ -239,7 +240,9 @@ def from_multi_group_flux(
         -------
         MicroXS
         """
-        # check energies
+
+        check_type("temperature", temperature, int)
+        # if energy is string then use group structure of that name
         if isinstance(energies, str):
             energies = openmc.EnergyFilter.from_group_structure(energies).values
         else:
@@ -256,7 +259,6 @@ def from_multi_group_flux(
         chain = openmc.deplete.Chain.from_xml(chain_file_path)
 
         openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
-        data_lib = openmc.data.DataLibrary.from_xml(openmc_data_path)
 
         # get reactions and nuclides from chain file
         nuclides, reactions = chain.nuclides, chain.reactions
@@ -271,29 +273,42 @@ def from_multi_group_flux(
         if 'fission' in reactions:
             mts.append(18) # fission is not in the REACTIONS
 
-        # normalize flux and get energy midpoint
-        norm_multi_group_flux = np.array(multi_group_flux) / sum(multi_group_flux)
-        energies_midpoint = [(energies[i]+energies[i+1])/2 for i in range(len(energies)-1)]
-        temperature_key = f'{temperature}K'
         microxs_arr = np.zeros((len(nuclides), len(mts)))
 
-        for nuc_indx, nuc in enumerate(nuclides):
-            mat_lib = data_lib.get_by_material(nuc, data_type='neutron')
-            if not mat_lib: # file does not exist
-                microxs_arr[nuc_indx, :] = 0
-                continue
-            else:
-                hdf5_path = mat_lib['path']
-            nuc_data = openmc.data.IncidentNeutron.from_hdf5(hdf5_path)
-            for mt_indx, mt in enumerate(mts):
-                try: # sometimes it fails cause there's no entry
-                    # perhaps a multigroup mgxs should be made as chosing the xs at the energy bin center is not always fair
-                    total_xs = np.array(nuc_data[mt].xs[temperature_key](energies_midpoint))
-                    # collapse
-                    total_norm = sum(total_xs * norm_multi_group_flux)
-                    microxs_arr[nuc_indx, mt_indx] = total_norm
-                except KeyError:
-                    microxs_arr[nuc_indx, mt_indx] = 0.0
+        with TemporaryDirectory():
+            from ctypes import c_int, c_double
+
+            mat_all_nucs = openmc.Material()
+            for nuc in nuclides:
+                mat_all_nucs.add_nuclide(nuc, 1)
+            mat_all_nucs.set_density("atom/b-cm", 1)
+            materials = openmc.Materials([mat_all_nucs])
+
+            surf1 = openmc.Sphere(r=1, boundary_type="vacuum")
+            surf1_region = -surf1
+            surf1_cell = openmc.Cell(region=surf1_region)
+            geometry = openmc.Geometry([surf1_cell])
+
+            settings = openmc.Settings()
+            settings.particles = 1
+            settings.batches = 1
+
+            model = openmc.Model(geometry, materials, settings)
+            model.export_to_model_xml()
+
+            c_temperature = c_double(temperature)
+
+            openmc.lib.init()
+            for nuc_indx, nuc in enumerate(nuclides):
+                lib_nuc = openmc.lib.nuclides[nuc]
+                for mt_indx, mt in enumerate(mts):
+                    c_mt = c_int(mt)
+                    collapse = lib_nuc.collapse_rate(
+                        c_mt, c_temperature, energies, multi_group_flux
+                    )
+                    microxs_arr[nuc_indx, mt_indx] = collapse
+
+            openmc.lib.finalize()
 
         return cls(nuclides=nuclides, reactions=reactions, data=microxs_arr)
 
diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index fcda08ad688..11a6542f7a6 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -63,7 +63,7 @@ def test_from_multi_group_flux():
     microxs = MicroXS.from_multi_group_flux(
         energies='CASMO-4',
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
-        temperature='293',
+        temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)
@@ -71,7 +71,7 @@ def test_from_multi_group_flux():
     microxs = MicroXS.from_multi_group_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
-        temperature='293',
+        temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)

From 9f73cd3f7e0a7f06917c45d00df8055b531f58c6 Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <mail@jshimwell.com>
Date: Wed, 29 Nov 2023 15:31:35 +0000
Subject: [PATCH 06/14] checking xs avail before adding to mat

---
 openmc/deplete/microxs.py | 31 +++++++++++++++++++++----------
 1 file changed, 21 insertions(+), 10 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 35b210a01f5..63913ce42c5 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -219,8 +219,11 @@ def from_multi_group_flux(
         openmc_data_path: Optional[PathLike] = None,
         temperature: int=294
     ):
-        """Generated MicroXS object from a known flux and a chain file. The size of the MicroXS matrix depends
-        on the chain file.
+        """Generated MicroXS object from a known flux and a chain file.
+
+        The size of the MicroXS matrix depends on the chain file and cross
+        sections available. MicroXS entry will be 0 if the nuclide cross section
+        is not found.
 
         Parameters
         ----------
@@ -259,6 +262,7 @@ def from_multi_group_flux(
         chain = openmc.deplete.Chain.from_xml(chain_file_path)
 
         openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
+        data_lib = openmc.data.DataLibrary.from_xml(openmc_data_path)
 
         # get reactions and nuclides from chain file
         nuclides, reactions = chain.nuclides, chain.reactions
@@ -280,7 +284,10 @@ def from_multi_group_flux(
 
             mat_all_nucs = openmc.Material()
             for nuc in nuclides:
-                mat_all_nucs.add_nuclide(nuc, 1)
+                mat_lib = data_lib.get_by_material(nuc, data_type="neutron")
+                # limits the material to nuclides that have cross sections
+                if mat_lib:
+                    mat_all_nucs.add_nuclide(nuc, 1)
             mat_all_nucs.set_density("atom/b-cm", 1)
             materials = openmc.Materials([mat_all_nucs])
 
@@ -300,13 +307,17 @@ def from_multi_group_flux(
 
             openmc.lib.init()
             for nuc_indx, nuc in enumerate(nuclides):
-                lib_nuc = openmc.lib.nuclides[nuc]
-                for mt_indx, mt in enumerate(mts):
-                    c_mt = c_int(mt)
-                    collapse = lib_nuc.collapse_rate(
-                        c_mt, c_temperature, energies, multi_group_flux
-                    )
-                    microxs_arr[nuc_indx, mt_indx] = collapse
+                mat_lib = data_lib.get_by_material(nuc, data_type="neutron")
+                if mat_lib:
+                    lib_nuc = openmc.lib.nuclides[nuc]
+                    for mt_indx, mt in enumerate(mts):
+                        c_mt = c_int(mt)
+                        collapse = lib_nuc.collapse_rate(
+                            c_mt, c_temperature, energies, multi_group_flux
+                        )
+                        microxs_arr[nuc_indx, mt_indx] = collapse
+                else:
+                    microxs_arr[nuc_indx, :] = 0
 
             openmc.lib.finalize()
 

From 9d6577a099eb803bafaaef25fd7f5f9ebdf57eb2 Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <drshimwell@gmail.com>
Date: Wed, 29 Nov 2023 16:10:44 +0000
Subject: [PATCH 07/14] no longer writing summary.h5

---
 openmc/deplete/microxs.py | 2 +-
 1 file changed, 1 insertion(+), 1 deletion(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 63913ce42c5..a1a160a7ca7 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -299,7 +299,7 @@ def from_multi_group_flux(
             settings = openmc.Settings()
             settings.particles = 1
             settings.batches = 1
-
+            settings.output = {'summary': False}
             model = openmc.Model(geometry, materials, settings)
             model.export_to_model_xml()
 

From 8785cc5ef116c428af7f000a97c6299e7e78d1fe Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <drshimwell@gmail.com>
Date: Tue, 5 Dec 2023 11:21:22 +0000
Subject: [PATCH 08/14] [skip ci] allowing nuclides to be provided

---
 openmc/deplete/microxs.py | 19 +++++++++++++------
 1 file changed, 13 insertions(+), 6 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index a1a160a7ca7..4fa176c24b4 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -217,7 +217,8 @@ def from_multi_group_flux(
         multi_group_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
         openmc_data_path: Optional[PathLike] = None,
-        temperature: int=294
+        temperature: int=294,
+        nuclides: Optional[Iterable[str]] = None,
     ):
         """Generated MicroXS object from a known flux and a chain file.
 
@@ -233,11 +234,15 @@ def from_multi_group_flux(
             Energy-dependent multigroup flux values
         chain_file : str, optional
             Path to the depletion chain XML file that will be used in
-            depletion simulation.
+            depletion simulation.  Defaults to ``openmc.config['chain_file']``.
         openmc_data_path : str, optional
             Path to the cross section XML file that contains data library paths.
+            Defaults to ``openmc.config['cross_sections']``.
         temperature : int, optional
             Temperature for cross section evaluation in [K].
+        nuclides : list of str
+                Nuclides to get cross sections for. If not specified, all burnable
+                nuclides from the depletion chain file are used.
 
         Returns
         -------
@@ -262,10 +267,13 @@ def from_multi_group_flux(
         chain = openmc.deplete.Chain.from_xml(chain_file_path)
 
         openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
-        data_lib = openmc.data.DataLibrary.from_xml(openmc_data_path)
+        cross_sections = openmc.data.DataLibrary.from_xml(openmc_data_path)
+        nuclides_with_data = _get_nuclides_with_data(cross_sections)
 
         # get reactions and nuclides from chain file
-        nuclides, reactions = chain.nuclides, chain.reactions
+        if not nuclides:
+                nuclides = chain.nuclides
+        reactions = chain.reactions
         nuclides = [nuc.name for nuc in nuclides]
         if 'fission' in reactions:
             # send to back
@@ -307,8 +315,7 @@ def from_multi_group_flux(
 
             openmc.lib.init()
             for nuc_indx, nuc in enumerate(nuclides):
-                mat_lib = data_lib.get_by_material(nuc, data_type="neutron")
-                if mat_lib:
+                if nuc in nuclides_with_data:
                     lib_nuc = openmc.lib.nuclides[nuc]
                     for mt_indx, mt in enumerate(mts):
                         c_mt = c_int(mt)

From d7fefa7bacc9023d1135928f09cb22e4d65f6f47 Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <mail@jshimwell.com>
Date: Tue, 5 Dec 2023 11:52:32 +0000
Subject: [PATCH 09/14] added testing and working for nuclides

---
 openmc/deplete/coupled_operator.py       | 12 +++++++-----
 openmc/deplete/microxs.py                | 22 ++++------------------
 openmc/lib/nuclide.py                    |  2 +-
 tests/unit_tests/test_deplete_microxs.py | 13 +++++++++++++
 4 files changed, 25 insertions(+), 24 deletions(-)

diff --git a/openmc/deplete/coupled_operator.py b/openmc/deplete/coupled_operator.py
index 0d778c7450c..23d9bd8e0d1 100644
--- a/openmc/deplete/coupled_operator.py
+++ b/openmc/deplete/coupled_operator.py
@@ -10,6 +10,7 @@
 
 import copy
 from warnings import warn
+from typing import Optional
 
 import numpy as np
 from uncertainties import ufloat
@@ -33,18 +34,19 @@
 __all__ = ["CoupledOperator", "Operator", "OperatorResult"]
 
 
-def _find_cross_sections(model):
+def _find_cross_sections(model: Optional[str] = None):
     """Determine cross sections to use for depletion
 
     Parameters
     ----------
-    model : openmc.model.Model
+    model : openmc.model.Model, optional
         Reactor model
 
     """
-    if model.materials and model.materials.cross_sections is not None:
-        # Prefer info from Model class if available
-        return model.materials.cross_sections
+    if model:
+        if model.materials and model.materials.cross_sections is not None:
+            # Prefer info from Model class if available
+            return model.materials.cross_sections
 
     # otherwise fallback to environment variable
     cross_sections = openmc.config.get("cross_sections")
diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 4fa176c24b4..41750e943c6 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -36,16 +36,6 @@ def _resolve_chain_file_path(chain_file:str):
     return chain_file
 
 
-def _resolve_openmc_data_path(openmc_data_path: str=None):
-    if not openmc_data_path:
-        if 'cross_sections' not in openmc.config:
-            raise ValueError("`openmc_data_path` is not defined nor `openmc.config['cross_sections']` is defined.")
-        else:
-            openmc_data_path = openmc.config['cross_sections']
-    return openmc_data_path
-
-
-
 def get_microxs_and_flux(
         model: openmc.Model,
         domains,
@@ -216,7 +206,6 @@ def from_multi_group_flux(
         energies: Union[Iterable[float], str],
         multi_group_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
-        openmc_data_path: Optional[PathLike] = None,
         temperature: int=294,
         nuclides: Optional[Iterable[str]] = None,
     ):
@@ -235,9 +224,6 @@ def from_multi_group_flux(
         chain_file : str, optional
             Path to the depletion chain XML file that will be used in
             depletion simulation.  Defaults to ``openmc.config['chain_file']``.
-        openmc_data_path : str, optional
-            Path to the cross section XML file that contains data library paths.
-            Defaults to ``openmc.config['cross_sections']``.
         temperature : int, optional
             Temperature for cross section evaluation in [K].
         nuclides : list of str
@@ -266,15 +252,15 @@ def from_multi_group_flux(
         chain_file_path = _resolve_chain_file_path(chain_file)
         chain = openmc.deplete.Chain.from_xml(chain_file_path)
 
-        openmc_data_path = _resolve_openmc_data_path(openmc_data_path)
-        cross_sections = openmc.data.DataLibrary.from_xml(openmc_data_path)
+        cross_sections = _find_cross_sections(model=None)
+        data_lib = openmc.data.DataLibrary.from_xml(cross_sections)
         nuclides_with_data = _get_nuclides_with_data(cross_sections)
 
         # get reactions and nuclides from chain file
         if not nuclides:
-                nuclides = chain.nuclides
+            nuclides = chain.nuclides
+            nuclides = [nuc.name for nuc in nuclides]
         reactions = chain.reactions
-        nuclides = [nuc.name for nuc in nuclides]
         if 'fission' in reactions:
             # send to back
             reactions.append(reactions.pop(reactions.index('fission')))
diff --git a/openmc/lib/nuclide.py b/openmc/lib/nuclide.py
index 399bb346520..8078882cf35 100644
--- a/openmc/lib/nuclide.py
+++ b/openmc/lib/nuclide.py
@@ -91,7 +91,7 @@ def collapse_rate(self, MT, temperature, energy, flux):
         energy : iterable of float
             Energy group boundaries in [eV]
         flux : iterable of float
-            Flux in each energt group (not normalized per eV)
+            Flux in each energy group (not normalized per eV)
 
         Returns
         -------
diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index 11a6542f7a6..8e98fb48de5 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -60,6 +60,7 @@ def test_csv():
 
 def test_from_multi_group_flux():
 
+    # test with energy group structure from string
     microxs = MicroXS.from_multi_group_flux(
         energies='CASMO-4',
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
@@ -68,6 +69,7 @@ def test_from_multi_group_flux():
     )
     assert isinstance(microxs, MicroXS)
 
+    # test with energy group structure as floats
     microxs = MicroXS.from_multi_group_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
         multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
@@ -75,3 +77,14 @@ def test_from_multi_group_flux():
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)
+
+    # test with nuclides provided
+    microxs = MicroXS.from_multi_group_flux(
+        energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
+        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        temperature=293,
+        chain_file=Path(__file__).parents[1] / 'chain_simple.xml',
+        nuclides=['Gd157']
+    )
+    assert microxs.nuclides == ['Gd157']
+    assert isinstance(microxs, MicroXS)

From 296ac964b17fe4049cd007e8fa710f170e985d9a Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <mail@jshimwell.com>
Date: Tue, 5 Dec 2023 13:35:39 +0000
Subject: [PATCH 10/14] allowing floats or ints for temperature

---
 openmc/deplete/coupled_operator.py | 2 +-
 openmc/deplete/microxs.py          | 5 +++--
 2 files changed, 4 insertions(+), 3 deletions(-)

diff --git a/openmc/deplete/coupled_operator.py b/openmc/deplete/coupled_operator.py
index 23d9bd8e0d1..287a4e0d37b 100644
--- a/openmc/deplete/coupled_operator.py
+++ b/openmc/deplete/coupled_operator.py
@@ -69,7 +69,7 @@ def _get_nuclides_with_data(cross_sections):
     Returns
     -------
     nuclides : set of str
-        Set of nuclide names that have cross secton data
+        Set of nuclide names that have cross section data
 
     """
     nuclides = set()
diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 41750e943c6..1146e3dc80d 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -206,7 +206,7 @@ def from_multi_group_flux(
         energies: Union[Iterable[float], str],
         multi_group_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
-        temperature: int=294,
+        temperature: float=294.,
         nuclides: Optional[Iterable[str]] = None,
     ):
         """Generated MicroXS object from a known flux and a chain file.
@@ -235,7 +235,7 @@ def from_multi_group_flux(
         MicroXS
         """
 
-        check_type("temperature", temperature, int)
+        check_type("temperature", temperature, (int, float))
         # if energy is string then use group structure of that name
         if isinstance(energies, str):
             energies = openmc.EnergyFilter.from_group_structure(energies).values
@@ -294,6 +294,7 @@ def from_multi_group_flux(
             settings.particles = 1
             settings.batches = 1
             settings.output = {'summary': False}
+
             model = openmc.Model(geometry, materials, settings)
             model.export_to_model_xml()
 

From 88f039888b31d7eb2bd55d1c788050a37fe3541c Mon Sep 17 00:00:00 2001
From: Jonathan Shimwell <drshimwell@gmail.com>
Date: Wed, 13 Dec 2023 23:54:31 +0000
Subject: [PATCH 11/14] Norm the flux

---
 openmc/deplete/microxs.py | 4 ++--
 1 file changed, 2 insertions(+), 2 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 1146e3dc80d..82a66c4d996 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -299,7 +299,7 @@ def from_multi_group_flux(
             model.export_to_model_xml()
 
             c_temperature = c_double(temperature)
-
+            norm_multi_group_flux = [val*1/sum(multi_group_flux) for val in multi_group_flux]
             openmc.lib.init()
             for nuc_indx, nuc in enumerate(nuclides):
                 if nuc in nuclides_with_data:
@@ -307,7 +307,7 @@ def from_multi_group_flux(
                     for mt_indx, mt in enumerate(mts):
                         c_mt = c_int(mt)
                         collapse = lib_nuc.collapse_rate(
-                            c_mt, c_temperature, energies, multi_group_flux
+                            c_mt, c_temperature, energies, norm_multi_group_flux
                         )
                         microxs_arr[nuc_indx, mt_indx] = collapse
                 else:

From 917e55ac9db4e02b0c34ae94c9f69ca6633855a4 Mon Sep 17 00:00:00 2001
From: Paul Romano <paul.k.romano@gmail.com>
Date: Wed, 20 Dec 2023 10:07:31 -0600
Subject: [PATCH 12/14] Simplify logic in from_multigroup_flux

---
 openmc/deplete/abc.py                    |   3 +-
 openmc/deplete/microxs.py                | 116 +++++++++++------------
 tests/unit_tests/test_deplete_microxs.py |  14 +--
 3 files changed, 67 insertions(+), 66 deletions(-)

diff --git a/openmc/deplete/abc.py b/openmc/deplete/abc.py
index d135b2dae5a..40ef88c6783 100644
--- a/openmc/deplete/abc.py
+++ b/openmc/deplete/abc.py
@@ -68,8 +68,9 @@ def change_directory(output_dir):
         Directory to switch to.
     """
     orig_dir  = os.getcwd()
+    output_dir = Path(output_dir)
+    output_dir.mkdir(parents=True, exist_ok=True)
     try:
-        output_dir.mkdir(parents=True, exist_ok=True)
         os.chdir(output_dir)
         yield
     finally:
diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 82a66c4d996..8dda887cd8e 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -14,7 +14,9 @@
 from openmc.checkvalue import check_type, check_value, check_iterable_type, PathLike
 from openmc.exceptions import DataError
 from openmc import StatePoint
+from openmc.mgxs import GROUP_STRUCTURES
 import openmc
+from .abc import change_directory
 from .chain import Chain, REACTIONS
 from .coupled_operator import _find_cross_sections, _get_nuclides_with_data
 import openmc.lib
@@ -23,16 +25,15 @@
 _valid_rxns.append('fission')
 
 
-def _resolve_chain_file_path(chain_file:str):
+def _resolve_chain_file_path(chain_file: str):
     # Determine what reactions and nuclides are available in chain
     if chain_file is None:
+        chain_file = openmc.config.get('chain_file')
         if 'chain_file' in openmc.config:
             raise DataError(
                 "No depletion chain specified and could not find depletion "
                 "chain in openmc.config['chain_file']"
             )
-        else:
-            chain_file = openmc.config['chain_file']
     return chain_file
 
 
@@ -201,15 +202,16 @@ def __init__(self, data: np.ndarray, nuclides: List[str], reactions: List[str]):
         self._index_rx = {rx: i for i, rx in enumerate(reactions)}
 
     @classmethod
-    def from_multi_group_flux(
+    def from_multigroup_flux(
         cls,
-        energies: Union[Iterable[float], str],
-        multi_group_flux: Sequence[float],
+        energies: Union[Sequence[float], str],
+        multigroup_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
-        temperature: float=294.,
+        temperature: float = 293.6,
         nuclides: Optional[Iterable[str]] = None,
-    ):
-        """Generated MicroXS object from a known flux and a chain file.
+        **init_kwargs: dict,
+    ) -> MicroXS:
+        """Generated microscopic cross sections from a known flux.
 
         The size of the MicroXS matrix depends on the chain file and cross
         sections available. MicroXS entry will be 0 if the nuclide cross section
@@ -229,6 +231,8 @@ def from_multi_group_flux(
         nuclides : list of str
                 Nuclides to get cross sections for. If not specified, all burnable
                 nuclides from the depletion chain file are used.
+        **init_kwargs : dict
+            Keyword arguments passed to :func:`openmc.lib.init`
 
         Returns
         -------
@@ -238,28 +242,29 @@ def from_multi_group_flux(
         check_type("temperature", temperature, (int, float))
         # if energy is string then use group structure of that name
         if isinstance(energies, str):
-            energies = openmc.EnergyFilter.from_group_structure(energies).values
+            energies = GROUP_STRUCTURES[energies]
         else:
-            # if user inputs energies check they are ascending (low to high) as some
-            # depletion codes use high energy to low energy.
-            if not all(energies[i] <= energies[i+1] for i in range(len(energies) - 1)):
-                raise ValueError('Energy bin must be in ascending order')
+            # if user inputs energies check they are ascending (low to high) as
+            # some depletion codes use high energy to low energy.
+            if not np.all(np.diff(energies) > 0):
+                raise ValueError('Energy group boundaries must be in ascending order')
 
         # check dimension consistency
-        if not len(multi_group_flux) == len(energies)-1:
+        if len(multigroup_flux) != len(energies) - 1:
             raise ValueError('Length of flux array should be len(energies)-1')
 
         chain_file_path = _resolve_chain_file_path(chain_file)
-        chain = openmc.deplete.Chain.from_xml(chain_file_path)
+        chain = Chain.from_xml(chain_file_path)
 
         cross_sections = _find_cross_sections(model=None)
-        data_lib = openmc.data.DataLibrary.from_xml(cross_sections)
         nuclides_with_data = _get_nuclides_with_data(cross_sections)
 
-        # get reactions and nuclides from chain file
+        # If no nuclides were specified, default to all nuclides from the chain
         if not nuclides:
             nuclides = chain.nuclides
             nuclides = [nuc.name for nuc in nuclides]
+
+        # get reactions and nuclides from chain file
         reactions = chain.reactions
         if 'fission' in reactions:
             # send to back
@@ -273,49 +278,44 @@ def from_multi_group_flux(
 
         microxs_arr = np.zeros((len(nuclides), len(mts)))
 
-        with TemporaryDirectory():
-            from ctypes import c_int, c_double
-
+        with TemporaryDirectory() as tmpdir:
+            # Create a material with all nuclides
             mat_all_nucs = openmc.Material()
             for nuc in nuclides:
-                mat_lib = data_lib.get_by_material(nuc, data_type="neutron")
-                # limits the material to nuclides that have cross sections
-                if mat_lib:
-                    mat_all_nucs.add_nuclide(nuc, 1)
-            mat_all_nucs.set_density("atom/b-cm", 1)
-            materials = openmc.Materials([mat_all_nucs])
-
-            surf1 = openmc.Sphere(r=1, boundary_type="vacuum")
-            surf1_region = -surf1
-            surf1_cell = openmc.Cell(region=surf1_region)
-            geometry = openmc.Geometry([surf1_cell])
-
-            settings = openmc.Settings()
-            settings.particles = 1
-            settings.batches = 1
-            settings.output = {'summary': False}
-
-            model = openmc.Model(geometry, materials, settings)
-            model.export_to_model_xml()
-
-            c_temperature = c_double(temperature)
-            norm_multi_group_flux = [val*1/sum(multi_group_flux) for val in multi_group_flux]
-            openmc.lib.init()
-            for nuc_indx, nuc in enumerate(nuclides):
                 if nuc in nuclides_with_data:
-                    lib_nuc = openmc.lib.nuclides[nuc]
-                    for mt_indx, mt in enumerate(mts):
-                        c_mt = c_int(mt)
-                        collapse = lib_nuc.collapse_rate(
-                            c_mt, c_temperature, energies, norm_multi_group_flux
-                        )
-                        microxs_arr[nuc_indx, mt_indx] = collapse
-                else:
-                    microxs_arr[nuc_indx, :] = 0
-
-            openmc.lib.finalize()
-
-        return cls(nuclides=nuclides, reactions=reactions, data=microxs_arr)
+                    mat_all_nucs.add_nuclide(nuc, 1.0)
+            mat_all_nucs.set_density("atom/b-cm", 1.0)
+
+            # Create simple model containing the above material
+            surf1 = openmc.Sphere(boundary_type="vacuum")
+            surf1_cell = openmc.Cell(fill=mat_all_nucs, region=-surf1)
+            model = openmc.Model()
+            model.geometry = openmc.Geometry([surf1_cell])
+            model.settings = openmc.Settings(
+                particles=1, batches=1, output={'summary': False})
+
+            with change_directory(tmpdir):
+                # Export model
+                model.export_to_model_xml()
+
+                # Normalize multigroup flux
+                multigroup_flux = np.asarray(multigroup_flux)
+                norm_multigroup_flux = multigroup_flux / multigroup_flux.sum()
+
+                with openmc.lib.run_in_memory(**init_kwargs):
+                    # For each nuclide and reaction, compute the flux-averaged
+                    # cross section
+                    for nuc_index, nuc in enumerate(nuclides):
+                        if nuc not in nuclides_with_data:
+                            continue
+                        lib_nuc = openmc.lib.nuclides[nuc]
+                        for mt_index, mt in enumerate(mts):
+                            xs = lib_nuc.collapse_rate(
+                                mt, temperature, energies, norm_multigroup_flux
+                            )
+                            microxs_arr[nuc_index, mt_index] = xs
+
+        return cls(microxs_arr, nuclides, reactions)
 
     @classmethod
     def from_csv(cls, csv_file, **kwargs):
diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index 8e98fb48de5..1ea06b87303 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -58,30 +58,30 @@ def test_csv():
     assert np.all(ref_xs.data == temp_xs.data)
     remove('temp_xs.csv')
 
-def test_from_multi_group_flux():
+def test_from_multigroup_flux():
 
     # test with energy group structure from string
-    microxs = MicroXS.from_multi_group_flux(
+    microxs = MicroXS.from_multigroup_flux(
         energies='CASMO-4',
-        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)
 
     # test with energy group structure as floats
-    microxs = MicroXS.from_multi_group_flux(
+    microxs = MicroXS.from_multigroup_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
-        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
     )
     assert isinstance(microxs, MicroXS)
 
     # test with nuclides provided
-    microxs = MicroXS.from_multi_group_flux(
+    microxs = MicroXS.from_multigroup_flux(
         energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
-        multi_group_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
+        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
         temperature=293,
         chain_file=Path(__file__).parents[1] / 'chain_simple.xml',
         nuclides=['Gd157']

From 6f09829bdd555edf7e99d6878ab7694e016e7125 Mon Sep 17 00:00:00 2001
From: Paul Romano <paul.k.romano@gmail.com>
Date: Wed, 20 Dec 2023 12:38:57 -0600
Subject: [PATCH 13/14] Add reactions argument to from_multigroup_flux

---
 openmc/deplete/microxs.py                | 48 ++++++++++++------------
 tests/unit_tests/test_deplete_microxs.py | 31 +++++++--------
 2 files changed, 39 insertions(+), 40 deletions(-)

diff --git a/openmc/deplete/microxs.py b/openmc/deplete/microxs.py
index 8dda887cd8e..0721535ca4f 100644
--- a/openmc/deplete/microxs.py
+++ b/openmc/deplete/microxs.py
@@ -15,6 +15,7 @@
 from openmc.exceptions import DataError
 from openmc import StatePoint
 from openmc.mgxs import GROUP_STRUCTURES
+from openmc.data import REACTION_MT
 import openmc
 from .abc import change_directory
 from .chain import Chain, REACTIONS
@@ -208,7 +209,8 @@ def from_multigroup_flux(
         multigroup_flux: Sequence[float],
         chain_file: Optional[PathLike] = None,
         temperature: float = 293.6,
-        nuclides: Optional[Iterable[str]] = None,
+        nuclides: Optional[Sequence[str]] = None,
+        reactions: Optional[Sequence[str]] = None,
         **init_kwargs: dict,
     ) -> MicroXS:
         """Generated microscopic cross sections from a known flux.
@@ -217,6 +219,8 @@ def from_multigroup_flux(
         sections available. MicroXS entry will be 0 if the nuclide cross section
         is not found.
 
+        .. versionadded:: 0.14.1
+
         Parameters
         ----------
         energies : iterable of float or str
@@ -224,13 +228,16 @@ def from_multigroup_flux(
         multi_group_flux : iterable of float
             Energy-dependent multigroup flux values
         chain_file : str, optional
-            Path to the depletion chain XML file that will be used in
-            depletion simulation.  Defaults to ``openmc.config['chain_file']``.
+            Path to the depletion chain XML file that will be used in depletion
+            simulation.  Defaults to ``openmc.config['chain_file']``.
         temperature : int, optional
             Temperature for cross section evaluation in [K].
-        nuclides : list of str
-                Nuclides to get cross sections for. If not specified, all burnable
-                nuclides from the depletion chain file are used.
+        nuclides : list of str, optional
+            Nuclides to get cross sections for. If not specified, all burnable
+            nuclides from the depletion chain file are used.
+        reactions : list of str, optional
+            Reactions to get cross sections for. If not specified, all neutron
+            reactions listed in the depletion chain file are used.
         **init_kwargs : dict
             Keyword arguments passed to :func:`openmc.lib.init`
 
@@ -264,18 +271,17 @@ def from_multigroup_flux(
             nuclides = chain.nuclides
             nuclides = [nuc.name for nuc in nuclides]
 
-        # get reactions and nuclides from chain file
-        reactions = chain.reactions
-        if 'fission' in reactions:
-            # send to back
-            reactions.append(reactions.pop(reactions.index('fission')))
-        # convert reactions to mt file
-        #! I feel like this zero indexing will backfire
-        #! There has to be a better way for this
-        mts = [list(REACTIONS[reaction].mts)[0] for reaction in reactions if reaction != 'fission']
-        if 'fission' in reactions:
-            mts.append(18) # fission is not in the REACTIONS
+        # Get reaction MT values. If no reactions specified, default to the
+        # reactions available in the chain file
+        if reactions is None:
+            reactions = chain.reactions
+        mts = [REACTION_MT[name] for name in reactions]
 
+        # Normalize multigroup flux
+        multigroup_flux = np.asarray(multigroup_flux)
+        multigroup_flux /= multigroup_flux.sum()
+
+        # Create 2D array for microscopic cross sections
         microxs_arr = np.zeros((len(nuclides), len(mts)))
 
         with TemporaryDirectory() as tmpdir:
@@ -295,13 +301,9 @@ def from_multigroup_flux(
                 particles=1, batches=1, output={'summary': False})
 
             with change_directory(tmpdir):
-                # Export model
+                # Export model within temporary directory
                 model.export_to_model_xml()
 
-                # Normalize multigroup flux
-                multigroup_flux = np.asarray(multigroup_flux)
-                norm_multigroup_flux = multigroup_flux / multigroup_flux.sum()
-
                 with openmc.lib.run_in_memory(**init_kwargs):
                     # For each nuclide and reaction, compute the flux-averaged
                     # cross section
@@ -311,7 +313,7 @@ def from_multigroup_flux(
                         lib_nuc = openmc.lib.nuclides[nuc]
                         for mt_index, mt in enumerate(mts):
                             xs = lib_nuc.collapse_rate(
-                                mt, temperature, energies, norm_multigroup_flux
+                                mt, temperature, energies, multigroup_flux
                             )
                             microxs_arr[nuc_index, mt_index] = xs
 
diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index 1ea06b87303..c54f5cf1e94 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -59,32 +59,29 @@ def test_csv():
     remove('temp_xs.csv')
 
 def test_from_multigroup_flux():
+    energies = [0., 6.25e-1, 5.53e3, 8.21e5, 2.e7]
+    flux = [1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4]
+    chain_file = Path(__file__).parents[1] / 'chain_simple.xml'
+    kwargs = {'multigroup_flux': flux, 'chain_file': chain_file}
 
     # test with energy group structure from string
-    microxs = MicroXS.from_multigroup_flux(
-        energies='CASMO-4',
-        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
-        temperature=293,
-        chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
-    )
+    microxs = MicroXS.from_multigroup_flux(energies='CASMO-4', **kwargs)
     assert isinstance(microxs, MicroXS)
 
     # test with energy group structure as floats
+    microxs = MicroXS.from_multigroup_flux(energies=energies, **kwargs)
+    assert isinstance(microxs, MicroXS)
+
+    # test with nuclides provided
     microxs = MicroXS.from_multigroup_flux(
-        energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
-        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
-        temperature=293,
-        chain_file=Path(__file__).parents[1] / 'chain_simple.xml'
+        energies=energies, nuclides=['Gd157', 'H1'], **kwargs
     )
     assert isinstance(microxs, MicroXS)
+    assert microxs.nuclides == ['Gd157', 'H1']
 
-    # test with nuclides provided
+    # test with reactions provided
     microxs = MicroXS.from_multigroup_flux(
-        energies=[0., 6.25e-1, 5.53e3, 8.21e5, 2.e7],
-        multigroup_flux=[1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4],
-        temperature=293,
-        chain_file=Path(__file__).parents[1] / 'chain_simple.xml',
-        nuclides=['Gd157']
+        energies=energies, reactions=['fission', '(n,2n)'], **kwargs
     )
-    assert microxs.nuclides == ['Gd157']
     assert isinstance(microxs, MicroXS)
+    assert microxs.reactions == ['fission', '(n,2n)']

From ed4e29cf9c05d98f7fb1f74e28f0e1b9bcf2055e Mon Sep 17 00:00:00 2001
From: Paul Romano <paul.k.romano@gmail.com>
Date: Thu, 21 Dec 2023 14:43:48 -0600
Subject: [PATCH 14/14] Add one test to confirm that micros are the same with
 equivalent group structures

---
 tests/unit_tests/test_deplete_microxs.py | 24 ++++++++++++++++++++++++
 1 file changed, 24 insertions(+)

diff --git a/tests/unit_tests/test_deplete_microxs.py b/tests/unit_tests/test_deplete_microxs.py
index c54f5cf1e94..ad54026f014 100644
--- a/tests/unit_tests/test_deplete_microxs.py
+++ b/tests/unit_tests/test_deplete_microxs.py
@@ -51,6 +51,7 @@ def test_from_array():
                        r'match dimensions of data array of shape \(\d*\, \d*\)'):
         MicroXS(data[:, 0], nuclides, reactions)
 
+
 def test_csv():
     ref_xs = MicroXS.from_csv(ONE_GROUP_XS)
     ref_xs.to_csv('temp_xs.csv')
@@ -58,6 +59,7 @@ def test_csv():
     assert np.all(ref_xs.data == temp_xs.data)
     remove('temp_xs.csv')
 
+
 def test_from_multigroup_flux():
     energies = [0., 6.25e-1, 5.53e3, 8.21e5, 2.e7]
     flux = [1.1e-7, 1.2e-6, 1.3e-5, 1.4e-4]
@@ -85,3 +87,25 @@ def test_from_multigroup_flux():
     )
     assert isinstance(microxs, MicroXS)
     assert microxs.reactions == ['fission', '(n,2n)']
+
+
+def test_multigroup_flux_same():
+    chain_file = Path(__file__).parents[1] / 'chain_simple.xml'
+
+    # Generate micro XS based on 4-group flux
+    energies = [0., 6.25e-1, 5.53e3, 8.21e5, 2.e7]
+    flux_per_ev = [0.3, 0.3, 1.0, 1.0]
+    flux = flux_per_ev * np.diff(energies)
+    microxs_4g = MicroXS.from_multigroup_flux(
+        energies=energies, multigroup_flux=flux, chain_file=chain_file)
+
+    # Generate micro XS based on 2-group flux, where the boundaries line up with
+    # the 4 group flux and have the same flux per eV across the full energy
+    # range
+    energies = [0., 5.53e3, 2.0e7]
+    flux_per_ev = [0.3, 1.0]
+    flux = flux_per_ev * np.diff(energies)
+    microxs_2g = MicroXS.from_multigroup_flux(
+        energies=energies, multigroup_flux=flux, chain_file=chain_file)
+
+    assert microxs_4g.data == pytest.approx(microxs_2g.data)