Merge pull request #9 from cnerg/anu1217-py

Added Spherical Shell Python script and associated files

SphericalShell/SS_h5_Reader.py

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+import matplotlib.pyplot as plt
+import openmc
+
+
+# Get results from statepoint
+with openmc.StatePoint('statepoint.10.h5') as sp:
+    t = sp.get_tally(name="Flux spectrum")
+    k = sp.get_tally(name="Neutron tally")
+
+    # Get the energies from the energy filter
+    energy_filter = t.filters[0]
+    energies = energy_filter.bins[:, 0]
+
+    # Get the flux values
+    mean = t.get_values(value='mean').ravel()
+
+    #Flux/elastic/absorption tallies:
+    tal = k.get_values(value='mean').ravel()
+    print(tal)
+
+# Plot flux spectrum
+fix, ax = plt.subplots()
+ax.loglog(energies, mean, drawstyle='steps-post')
+ax.set_xlabel('Energy [eV]')
+ax.set_ylabel('Flux [neutron-cm/source]')
+ax.grid(True, which='both')
+plt.savefig('Neutron_flux_vs_energy.png')
+plt.show()

SphericalShell/SphericalShell.py

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+# -*- coding: utf-8 -*-
+"""
+Created on Thu Feb  8 08:16:12 2024
+
+@author: Anupama Rajendra
+"""
+import openmc
+import numpy as np
+#import tkinter as tk
+
+# Create materials & export to XML:
+#Simulating tungsten shell:
+W = openmc.Material(name='W_Shell')
+W.set_density('g/cm3', 19.28)
+W.add_element('W', 1.0)
+materials = openmc.Materials([W])
+materials.export_to_xml()
+
+# Create geometry
+#Spherical shell:
+R_1= openmc.Sphere(r=1000) #sphere of radius 1000cm
+inside_sphere_1 = -R_1
+outside_sphere_1 = +R_1
+R_2 = openmc.Sphere(r=1005, boundary_type='vacuum')
+inside_sphere_2 = -R_2
+outside_sphere_2 = +R_2
+R_3 = outside_sphere_1 & inside_sphere_2 #filled with tungsten
+
+# Mapping materials to geometry:
+Void = openmc.Cell(fill=None, region = inside_sphere_1)
+Shell = openmc.Cell(fill=W, region=R_3)
+geometry = openmc.Geometry([Void, Shell])
+geometry.export_to_xml()
+
+
+# # Source distribution:
+PointSource = openmc.stats.Point(xyz=(0.0, 0.0, 0.0))
+Prob = openmc.stats.Discrete(14E+06, 1.0)
+
+# Assign simulation settings
+settings = openmc.Settings()
+settings.batches = 10
+settings.inactive = 1
+settings.particles = 100000
+settings.source = openmc.Source(space=PointSource, energy=Prob, strength = 1.0, particle = 'neutron')
+settings.run_mode = 'fixed source'
+settings.export_to_xml()
+
+# Define tallies
+neutron_tally = openmc.Tally(name="Neutron tally")
+neutron_tally.scores = ['flux', 'elastic', 'absorption']
+# Implementing filter for neutron tally through W shell
+cell_filter = openmc.CellFilter([Shell])
+neutron_tally.filters = [cell_filter]
+
+# Creating a tally to get the flux energy spectrum.
+# An energy filter is created to assign to the flux tally.
+e_min, e_max = 5e2, 14.001e6
+groups = 500
+energies = np.logspace(np.log10(e_min), np.log10(e_max), groups + 1)
+energy_filter = openmc.EnergyFilter(energies)
+
+spectrum_tally = openmc.Tally(name="Flux spectrum")
+# Implementing energy and cell filters for flux spectrum tally
+spectrum_tally.filters = [energy_filter, cell_filter]
+spectrum_tally.scores = ['flux']
+
+# Collecting and exporting tallies to .xml
+tallies = openmc.Tallies([neutron_tally, spectrum_tally])
+tallies.export_to_xml()