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297 lines
14 KiB
Python
297 lines
14 KiB
Python
import openmc
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import numpy as np
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names = ['H', 'O', 'Zr', 'U235', 'U238']
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def build_openmc_xs_lib(name, groups, temperatures, xsdict, micro=True):
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"""Build an Openm XSdata based on dictionary values"""
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xsdata = openmc.XSdata(name, groups, temperatures=temperatures)
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xsdata.order = 0
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for tt in temperatures:
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xsdata.set_absorption(xsdict[tt]['absorption'][name], temperature=tt)
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xsdata.set_scatter_matrix(xsdict[tt]['scatter'][name], temperature=tt)
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xsdata.set_total(xsdict[tt]['total'][name], temperature=tt)
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if (name in xsdict[tt]['nu-fission'].keys()):
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xsdata.set_nu_fission(xsdict[tt]['nu-fission'][name],
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temperature=tt)
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xsdata.set_chi(np.array([1., 0.]), temperature=tt)
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return xsdata
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def create_micro_xs_dict():
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"""Returns micro xs library"""
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xs_micro = {}
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reactions = ['absorption', 'total', 'scatter', 'nu-fission']
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# chi is unnecessary when energy bound is in thermal region
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# Temperature 300K
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# absorption
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xs_micro[300] = {r: {} for r in reactions}
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xs_micro[300]['absorption']['H'] = np.array([1.0285E-4, 0.0057])
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xs_micro[300]['absorption']['O'] = np.array([7.1654E-5, 3.0283E-6])
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xs_micro[300]['absorption']['Zr'] = np.array([4.5918E-5, 3.6303E-5])
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xs_micro[300]['absorption']['U235'] = np.array([0.0035, 0.1040])
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xs_micro[300]['absorption']['U238'] = np.array([0.0056, 0.0094])
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# nu-scatter matrix
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xs_micro[300]['scatter']['H'] = np.array([[[0.0910, 0.01469],
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[0.0, 0.3316]]])
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xs_micro[300]['scatter']['O'] = np.array([[[0.0814, 3.3235E-4],
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[0.0, 0.0960]]])
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xs_micro[300]['scatter']['Zr'] = np.array([[[0.0311, 2.6373E-5],
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[0.0, 0.0315]]])
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xs_micro[300]['scatter']['U235'] = np.array([[[0.0311, 2.6373E-5],
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[0.0, 0.0315]]])
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xs_micro[300]['scatter']['U238'] = np.array([[[0.0551, 2.2341E-5],
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[0.0, 0.0526]]])
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# nu-fission
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xs_micro[300]['nu-fission']['U235'] = np.array([0.0059, 0.2160])
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xs_micro[300]['nu-fission']['U238'] = np.array([0.0019, 1.4627E-7])
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# total
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xs_micro[300]['total']['H'] = xs_micro[300]['absorption']['H'] + \
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np.sum(xs_micro[300]['scatter']['H'][0], 1)
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xs_micro[300]['total']['O'] = xs_micro[300]['absorption']['O'] + \
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np.sum(xs_micro[300]['scatter']['O'][0], 1)
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xs_micro[300]['total']['Zr'] = xs_micro[300]['absorption']['Zr'] + \
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np.sum(xs_micro[300]['scatter']['Zr'][0], 1)
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xs_micro[300]['total']['U235'] = xs_micro[300]['absorption']['U235'] + \
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np.sum(xs_micro[300]['scatter']['U235'][0], 1)
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xs_micro[300]['total']['U238'] = xs_micro[300]['absorption']['U238'] + \
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np.sum(xs_micro[300]['scatter']['U238'][0], 1)
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# Temperature 600K
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xs_micro[600] = {r: {} for r in reactions}
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# absorption
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xs_micro[600]['absorption']['H'] = np.array([1.0356E-4, 0.0046])
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xs_micro[600]['absorption']['O'] = np.array([7.2678E-5, 2.4963E-6])
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xs_micro[600]['absorption']['Zr'] = np.array([4.7256E-5, 2.9757E-5])
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xs_micro[600]['absorption']['U235'] = np.array([0.0035, 0.0853])
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xs_micro[600]['absorption']['U238'] = np.array([0.0058, 0.0079])
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# nu-scatter matrix
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xs_micro[600]['scatter']['H'] = np.array([[[0.0910, 0.0138],
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[0.0, 0.3316]]])
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xs_micro[600]['scatter']['O'] = np.array([[[0.0814, 3.5367E-4],
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[0.0, 0.0959]]])
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xs_micro[600]['scatter']['Zr'] = np.array([[[0.0311, 3.2293E-5],
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[0.0, 0.0314]]])
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xs_micro[600]['scatter']['U235'] = np.array([[[0.0022, 1.9763E-6],
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[9.1634E-8, 0.0039]]])
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xs_micro[600]['scatter']['U238'] = np.array([[[0.0556, 2.8803E-5],
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[0.0, 0.0536]]])
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# nu-fission
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xs_micro[600]['nu-fission']['U235'] = np.array([0.0059, 0.1767])
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xs_micro[600]['nu-fission']['U238'] = np.array([0.0019, 1.2405E-7])
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# total
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xs_micro[600]['total']['H'] = xs_micro[600]['absorption']['H'] + \
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np.sum(xs_micro[600]['scatter']['H'][0], 1)
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xs_micro[600]['total']['O'] = xs_micro[600]['absorption']['O'] + \
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np.sum(xs_micro[600]['scatter']['O'][0], 1)
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xs_micro[600]['total']['Zr'] = xs_micro[600]['absorption']['Zr'] + \
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np.sum(xs_micro[600]['scatter']['Zr'][0], 1)
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xs_micro[600]['total']['U235'] = xs_micro[600]['absorption']['U235'] + \
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np.sum(xs_micro[600]['scatter']['U235'][0], 1)
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xs_micro[600]['total']['U238'] = xs_micro[600]['absorption']['U238'] + \
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np.sum(xs_micro[600]['scatter']['U238'][0], 1)
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# Temperature 900K
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xs_micro[900] = {r: {} for r in reactions}
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# absorption
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xs_micro[900]['absorption']['H'] = np.array([1.0529E-4, 0.0040])
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xs_micro[900]['absorption']['O'] = np.array([7.3055E-5, 2.1850E-6])
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xs_micro[900]['absorption']['Zr'] = np.array([4.7141E-5, 2.5941E-5])
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xs_micro[900]['absorption']['U235'] = np.array([0.0035, 0.0749])
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xs_micro[900]['absorption']['U238'] = np.array([0.0060, 0.0071])
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# total
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xs_micro[900]['total']['H'] = np.array([0.2982, 0.7332])
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xs_micro[900]['total']['O'] = np.array([0.0885, 0.1004])
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xs_micro[900]['total']['Zr'] = np.array([0.0370, 0.0317])
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xs_micro[900]['total']['U235'] = np.array([0.0061, 0.0789])
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xs_micro[900]['total']['U238'] = np.array([0.0707, 0.0613])
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# nu-scatter matrix
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xs_micro[900]['scatter']['H'] = np.array([[[0.0913, 0.0147],
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[0.0, 0.4020]]])
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xs_micro[900]['scatter']['O'] = np.array([[[0.0812, 4.0413E-4],
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[0.0, 0.0965]]])
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xs_micro[900]['scatter']['Zr'] = np.array([[[0.0311, 3.6735E-5],
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[0.0, 0.0314]]])
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xs_micro[900]['scatter']['U235'] = np.array([[[0.0022, 2.9034E-6],
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[1.3117E-8, 0.0039]]])
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xs_micro[900]['scatter']['U238'] = np.array([[[0.0560, 3.7619E-5],
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[0.0, 0.0538]]])
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# nu-fission
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xs_micro[900]['nu-fission']['U235'] = np.array([0.0059, 0.1545])
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xs_micro[900]['nu-fission']['U238'] = np.array([0.0019, 1.1017E-7])
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# total
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xs_micro[900]['total']['H'] = xs_micro[900]['absorption']['H'] + \
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np.sum(xs_micro[900]['scatter']['H'][0], 1)
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xs_micro[900]['total']['O'] = xs_micro[900]['absorption']['O'] + \
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np.sum(xs_micro[900]['scatter']['O'][0], 1)
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xs_micro[900]['total']['Zr'] = xs_micro[900]['absorption']['Zr'] + \
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np.sum(xs_micro[900]['scatter']['Zr'][0], 1)
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xs_micro[900]['total']['U235'] = xs_micro[900]['absorption']['U235'] + \
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np.sum(xs_micro[900]['scatter']['U235'][0], 1)
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xs_micro[900]['total']['U238'] = xs_micro[900]['absorption']['U238'] + \
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np.sum(xs_micro[900]['scatter']['U238'][0], 1)
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# roll axis for scatter matrix
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for t in xs_micro:
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for n in xs_micro[t]['scatter']:
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xs_micro[t]['scatter'][n] = np.rollaxis(xs_micro[t]['scatter'][n],
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0, 3)
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return xs_micro
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def create_macro_dict(xs_micro):
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"""Create a dictionary with two group cross-section"""
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xs_macro = {}
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for t, d1 in xs_micro.items():
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xs_macro[t] = {}
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for r, d2 in d1.items():
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temp = []
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xs_macro[t][r] = {}
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for n, v in d2.items():
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temp.append(d2[n])
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# The name 'macro' is needed to store data at the same level
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# of a xs_macro dictionary as for xs_micro and use it in
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# function build_openmc_xs_lib
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xs_macro[t][r]['macro'] = sum(temp)
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return xs_macro
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def create_openmc_2mg_libs(names):
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"""Built a micro/macro two group openmc MGXS libraries"""
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# Initialized library params
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group_edges = [0.0, 0.625, 20.0e6]
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groups = openmc.mgxs.EnergyGroups(group_edges=group_edges)
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mg_cross_sections_file_micro = openmc.MGXSLibrary(groups)
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mg_cross_sections_file_macro = openmc.MGXSLibrary(groups)
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# Building a micro mg library
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micro_cs = create_micro_xs_dict()
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for name in names:
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mg_cross_sections_file_micro.add_xsdata(build_openmc_xs_lib(name,
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groups,
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[t for t in
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micro_cs],
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micro_cs))
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# Building a macro mg library
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macro_xs = create_macro_dict(micro_cs)
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mg_cross_sections_file_macro.add_xsdata(build_openmc_xs_lib('macro',
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groups,
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[t for t in
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macro_xs],
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macro_xs))
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# Exporting library to hdf5 files
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mg_cross_sections_file_micro.export_to_hdf5('micro_2g.h5')
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mg_cross_sections_file_macro.export_to_hdf5('macro_2g.h5')
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# Returning the macro_xs dict is needed for analytical solution
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return macro_xs
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def analytical_solution_2g_therm(xsmin, xsmax=None, wgt=1.0):
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""" Calculate eigenvalue based on analytical solution for eq Lf = (1/k)Qf
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in two group for infinity dilution media in assumption of group
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boundary in thermal spectra < 1.e+3 Ev
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Parameters:
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----------
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xsmin : dict
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macro cross-sections dictionary with minimum range temperature
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xsmax : dict
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macro cross-sections dictionary with maximum range temperature
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by default: None not used for standalone temperature
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wgt : float
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weight for interpolation by default 1.0
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Returns:
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-------
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keff : np.float64
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analytical eigenvalue of critical eq matrix
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"""
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if xsmax is None:
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sa = xsmin['absorption']['macro']
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ss12 = xsmin['scatter']['macro'][0][1][0]
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nsf = xsmin['nu-fission']['macro']
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else:
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sa = xsmin['absorption']['macro'] * wgt + \
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xsmax['absorption']['macro'] * (1 - wgt)
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ss12 = xsmin['scatter']['macro'][0][1][0] * wgt + \
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xsmax['scatter']['macro'][0][1][0] * (1 - wgt)
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nsf = xsmin['nu-fission']['macro'] * wgt + \
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xsmax['nu-fission']['macro'] * (1 - wgt)
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L = np.array([sa[0] + ss12, 0.0, -ss12, sa[1]]).reshape(2, 2)
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Q = np.array([nsf[0], nsf[1], 0.0, 0.0]).reshape(2, 2)
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arr = np.linalg.inv(L).dot(Q)
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return np.amax(np.linalg.eigvals(arr).real)
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def build_inf_model(xsnames, xslibname, temperature, tempmethod='nearest'):
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""" Building an infinite medium for openmc multi-group testing
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Parameters:
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----------
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xsnames : list of str()
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list with xs names
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xslibname:
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name of hdf5 file with cross-section library
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temperature : float
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value of a current temperature in K
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tempmethod : {'nearest', 'interpolation'}
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by default 'nearest'
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"""
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model = openmc.Model()
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inf_medium = openmc.Material(name='test material', material_id=1)
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inf_medium.set_density("sum")
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for xs in xsnames:
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inf_medium.add_nuclide(xs, 1)
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INF = 11.1
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# Instantiate a Materials collection and export to XML
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materials_file = openmc.Materials([inf_medium])
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materials_file.cross_sections = xslibname
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model.materials = materials_file
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# Instantiate boundary Planes
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min_x = openmc.XPlane(boundary_type='reflective', x0=-INF)
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max_x = openmc.XPlane(boundary_type='reflective', x0=INF)
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min_y = openmc.YPlane(boundary_type='reflective', y0=-INF)
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max_y = openmc.YPlane(boundary_type='reflective', y0=INF)
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# Instantiate a Cell
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cell = openmc.Cell(cell_id=1, name='cell')
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cell.temperature = temperature
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# Register bounding Surfaces with the Cell
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cell.region = +min_x & -max_x & +min_y & -max_y
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# Fill the Cell with the Material
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cell.fill = inf_medium
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# Create root universe
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root_universe = openmc.Universe(name='root universe', cells=[cell])
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# Create Geometry and set root Universe
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model.geometry = openmc.Geometry(root_universe)
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# OpenMC simulation parameters
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batches = 200
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inactive = 5
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particles = 5000
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# Instantiate a Settings object
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settings_file = openmc.Settings()
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settings_file.batches = batches
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settings_file.inactive = inactive
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settings_file.particles = particles
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settings_file.energy_mode = 'multi-group'
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settings_file.output = {'summary': False}
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# Create an initial uniform spatial source distribution over fissionable zones
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bounds = [-INF, -INF, -INF, INF, INF, INF]
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uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
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settings_file.temperature = {'method': tempmethod}
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settings_file.source = openmc.IndependentSource(
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space=uniform_dist, constraints={'fissionable': True})
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model.settings = settings_file
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model.export_to_model_xml()
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