text: import openmc_tally_unit_converter as otuc
  import openmc
  from regular_mesh_plotter import plot_regular_mesh_tally_with_geometry
  
  # can be used to scale the color scale
  from matplotlib.colors import LogNorm
  
  statepoint_file = "statepoint.4.h5"
  
  # loads up the statepoint file with simulation results
  statepoint = openmc.StatePoint(filepath=statepoint_file)
  
  # gets the first tally using its name
  my_tally_xy = statepoint.get_tally(name="(n,Xa)_on_2D_mesh_xy")
  my_tally_xz = statepoint.get_tally(name="(n,Xa)_on_2D_mesh_xz")
  
  # gets number of neutron for a 1.3 mega joule shot
  source_strength = otuc.find_source_strength(fusion_energy_per_second_or_per_pulse=1.3e6)
  
  
  plot = plot_regular_mesh_tally_with_geometry(
      tally=my_tally_xy,
      dagmc_file_or_trimesh_object="dagmc.h5m",
      plane_origin=None,
      plane_normal=[0, 0, 1],
      scale=LogNorm(),
      rotate_mesh=180,
      vmin=None,
      label="(n,Xa) per pulse",
      filename="mesh_tally_xy.png",
  )
  
  # plot.show()
  
  plot = plot_regular_mesh_tally_with_geometry(
      tally=my_tally_xz,
      dagmc_file_or_trimesh_object="dagmc.h5m",
      plane_origin=[0,0.1,0],  # moves the slice 1mm into the 180 degree model to ensure geometry is sliced
      plane_normal=[0, 1, 0],
      rotate_geometry=90,
      scale=LogNorm(),
      vmin=None,
      label="(n,Xa)",
      filename="my_tally_xz.png",
  )
  
  # plot.show()
  