From c1082420e488b8c28c207d31e8b72f3b5221a1f3 Mon Sep 17 00:00:00 2001
From: Paul Romano <paul.k.romano@gmail.com>
Date: Wed, 17 Jan 2018 06:47:03 -0600
Subject: [PATCH] Use argparse in build-fresh.py

---
 smr/build-fresh.py | 28 ++++++++++++----------------
 1 file changed, 12 insertions(+), 16 deletions(-)

diff --git a/smr/build-fresh.py b/smr/build-fresh.py
index 026bec5..5819516 100644
--- a/smr/build-fresh.py
+++ b/smr/build-fresh.py
@@ -3,6 +3,7 @@
 import os
 import shutil
 import copy
+import argparse
 
 import numpy as np
 
@@ -13,24 +14,19 @@ from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
 from smr.core import core_geometry
 
 
-#### Query the user for options
+# Define command-line options
+parser = argparse.ArgumentParser()
+parser.add_argument('-m', '--multipole', action='store_true',
+                    help='Whether to use multipole cross sections')
+parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='cell',
+                    help='Whether to use distribmats or distribcells for tallies')
+args = parser.parse_args()
 
-# Query the user on whether to use multipole cross sections
-multipole = input('Use multipole cross sections? (y/n): ').lower()
-multipole = (multipole == 'y')
-
-# Query the user on whether to use distribmats or distribcells
-# If using distribmats, the geometry must be "differentiated" with unique
-# material instances for each instance of a fuel cell
-distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
-
-if distrib not in ['cell', 'mat']:
-    raise InputError('Distrib type "{}" is unsupported'.format(distrib))
 
 geometry = core_geometry()
 
 #### "Differentiate" the geometry if using distribmats
-if distrib == 'mat':
+if args.tallies == 'mat':
     # Count the number of instances for each cell and material
     geometry.determine_paths(instances_only=True)
 
@@ -68,7 +64,7 @@ settings.output = {'tallies': False}
 settings.source = source
 settings.sourcepoint_write = False
 
-if multipole:
+if args.multipole:
     settings.temperature = {'multipole': True, 'tolerance': 1000}
 
 settings.export_to_xml()
@@ -85,7 +81,7 @@ tallies = openmc.Tallies()
 materials = geometry.get_materials_by_name(name='Fuel', matching=False)
 
 # If using distribcells, create distribcell tally needed for depletion
-if distrib == 'cell':
+if args.tallies == 'cell':
     # Extract all cells filled by a fuel material
     fuel_cells = []
     for cell in geometry.get_all_cells().values():
@@ -98,7 +94,7 @@ if distrib == 'cell':
             tallies.append(tally)
 
 # If using distribmats, create material tally needed for depletion
-elif distrib == 'mat':
+elif args.tallies == 'mat':
     tally = openmc.Tally(name='depletion tally')
     tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
                     'fission', '(n,2n)', '(n,3n)', '(n,4n)']