#!/usr/bin/env python
# coding: utf-8

# In[10]:


from collections import defaultdict
import matplotlib.pyplot as plt
import openmc.deplete
import openmc.data


# In[3]:


chain = openmc.deplete.Chain.from_xml('/home/romano/openmc-data/depletion/chain_casl_pwr.xml')


# In[4]:


u235 = chain['U235']


# In[5]:


u235.yield_data


# The yield data is organized by incident neutron energy

# In[6]:


u235.yield_data.energies


# You can index the yield data by energy:

# In[7]:


thermal_yield = u235.yield_data[0.0253]
thermal_yield


# And then by a product:

# In[8]:


thermal_yield['I135']


# `thermal_yield` will act as a dictionary, so you can use methods like `items()` on it:

# In[9]:


list(thermal_yield.items())


# In[11]:


chain_yields = defaultdict(float)
for nucname, y in thermal_yield.items():
    z, a, m = openmc.data.zam(nucname)
    chain_yields[a] += y
    
mass, yields = zip(*sorted(chain_yields.items()))


# In[18]:


plt.plot(mass, yields)
plt.xlabel('Mass number')
plt.ylabel('Chain yield')


# In[ ]: