From 75b668f3bd797f49733d69274bfd0471a622c4bf Mon Sep 17 00:00:00 2001 From: shimwell Date: Wed, 15 Nov 2023 23:35:43 +0000 Subject: [PATCH 1/2] get_decay_photon_energy api update --- .../1_cell_based_shut_down_dose_rate_example.py | 7 +++---- 1 file changed, 3 insertions(+), 4 deletions(-) diff --git a/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py b/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py index fe766dfa..0abc19e2 100644 --- a/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py +++ b/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py @@ -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 @@ -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 @@ -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 @@ -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 From cc7424a556fbd79e0c3e7b82b48ed25f7e6f131b Mon Sep 17 00:00:00 2001 From: shimwell Date: Wed, 15 Nov 2023 23:37:23 +0000 Subject: [PATCH 2/2] typos --- .../1_cell_based_shut_down_dose_rate_example.py | 4 ++-- 1 file changed, 2 insertions(+), 2 deletions(-) diff --git a/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py b/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py index 0abc19e2..5f5b2a51 100644 --- a/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py +++ b/tasks/task_11_CSG_shut_down_dose_tallies/1_cell_based_shut_down_dose_rate_example.py @@ -198,7 +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]) @@ -207,7 +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