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Update api usage decay photon #254

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merged 2 commits into from
Nov 15, 2023
Merged

Update api usage decay photon #254

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75b668f
get_decay_photon_energy api update
shimwell Nov 15, 2023
cc7424a
typos
shimwell Nov 15, 2023
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11 changes: 5 additions & 6 deletions tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py
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Original file line number Diff line number Diff line change
Expand Up @@ -4,7 +4,6 @@
import numpy as np
import openmc
import openmc.deplete
import pint
from pathlib import Path
import math
from matplotlib.colors import LogNorm
Expand All @@ -13,7 +12,7 @@
# the chain file was downloaded with
# pip install openmc_data
# download_endf_chain -r b8.0
openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'
# openmc.config['chain_file'] = '/nuclear_data/chain-endf-b8.0.xml'
# openmc.config['cross_sections'] = 'cross_sections.xml'

# a few user settings
Expand Down Expand Up @@ -85,7 +84,7 @@

model_neutron = openmc.Model(my_geometry, my_materials, my_neutron_settings)

hour_in_seconds = pint.Quantity(1.0, "hour").to("s").magnitude
hour_in_seconds = 60*60

# This section defines the neutron pulse schedule.
# Warning, be sure to add sufficient timesteps and run the neutron simulation with enough
Expand Down Expand Up @@ -180,7 +179,7 @@
# gets the activated material using the material id
activated_mat = results[i_cool].get_material(str(material_id))
# gets the energy and probabilities for the
energy = activated_mat.decay_photon_energy
energy = activated_mat.get_decay_photon_energy()
strength = energy.integral()

if strength > 0.: # only makes sources for
Expand All @@ -199,7 +198,7 @@


# one should also fill the cells with the activated material
# the activated material contains ALL the iotopes produced during activation
# the activated material contains ALL the nuclides produced during activation
# sphere_cell_2.fill = results[i_cool].get_material("1")
# sphere_cell_3.fill = results[i_cool].get_material("2")
# my_geometry = openmc.Geometry([sphere_cell_1, sphere_cell_2, sphere_cell_3])
Expand All @@ -208,7 +207,7 @@
# so you could make use of openmc.deplete.Results.export_to_materials to export the modified activated material that
# just contains isotopes that appear in your cross_sections.xml

# however in this example we just use the original prisine material my_materials that were cloned earlier
# however in this example we just use the original pristine material my_materials that were cloned earlier
# my_geometry is also the same as the neutron simulation
pristine_mat_iron.id = 1
pristine_mat_aluminium.id =2
Expand Down