import openmc import openmc.deplete import numpy as np import matplotlib.pyplot as plt ############################################################################### # Simulation Input File Parameters ############################################################################### ## OpenMC simulation parameters batches = 300 inactive = 50 particles = 100000 ## Depletion simulation parameters ts = 40*24*60*60 # s td = 10*24*60*60 # s time_steps = np.array([ts,ts,ts,ts,ts,ts,ts,ts,ts,ts,td,ts,ts,ts,ts,ts,ts,ts,ts,ts,ts,td,ts,ts,ts,ts,ts,ts,ts,ts,ts,ts,td,ts,ts,ts,ts,ts,ts,ts,ts,ts,ts,td]) chain_file = '../chain_endfb71.xml' power = 17373 # W/cm, for 2D simulations only (use W for 3D) ############################################################################### # Define materials ############################################################################### ## Instantiate some Materials and register the appropriate Nuclides zircaloy = openmc.Material(material_id=3, name='Zircaloy 4') zircaloy.set_density('g/cm3', 6.49012) zircaloy.temperature = 600.0 zircaloy.add_nuclide('Sn112', 1.37041E-04, 'wo') zircaloy.add_nuclide('Sn114', 9.34709E-05, 'wo') zircaloy.add_nuclide('Sn115', 4.93221E-05, 'wo') zircaloy.add_nuclide('Sn116', 2.12757E-03, 'wo') zircaloy.add_nuclide('Sn117', 1.13348E-03, 'wo') zircaloy.add_nuclide('Sn118', 3.60520E-03, 'wo') zircaloy.add_nuclide('Sn119', 1.28950E-03, 'wo') zircaloy.add_nuclide('Sn120', 4.93338E-03, 'wo') zircaloy.add_nuclide('Sn122', 7.12568E-04, 'wo') zircaloy.add_nuclide('Sn124', 9.05713E-04, 'wo') zircaloy.add_nuclide('Fe54', 7.34249E-05, 'wo') zircaloy.add_nuclide('Fe56', 1.19419E-03, 'wo') zircaloy.add_nuclide('Fe57', 2.80871E-05, 'wo') zircaloy.add_nuclide('Fe58', 3.77460E-06, 'wo') zircaloy.add_nuclide('Cr50', 4.17870E-05, 'wo') zircaloy.add_nuclide('Cr52', 8.37035E-04, 'wo') zircaloy.add_nuclide('Cr53', 9.67293E-05, 'wo') zircaloy.add_nuclide('Cr54', 2.44824E-05, 'wo') zircaloy.add_nuclide('Ni58', 4.03133E-04, 'wo') zircaloy.add_nuclide('Ni60', 1.60602E-04, 'wo') zircaloy.add_nuclide('Ni61', 7.09947E-06, 'wo') zircaloy.add_nuclide('Ni62', 2.29762E-05, 'wo') zircaloy.add_nuclide('Ni64', 6.07653E-06, 'wo') zircaloy.add_nuclide('Zr90', 4.97990E-01, 'wo') zircaloy.add_nuclide('Zr91', 1.09810E-01, 'wo') zircaloy.add_nuclide('Zr92', 1.69691E-01, 'wo') zircaloy.add_nuclide('Zr94', 1.75712E-01, 'wo') zircaloy.add_nuclide('Zr96', 2.89111E-02, 'wo') uo2 = openmc.Material(name='Fuel Batch 1') uo2.set_density('g/cc' ,10.07) uo2.temperature = 900.0 uo2.add_nuclide('O16' ,1.18530E-01,'wo') uo2.add_nuclide('U234' ,3.92254E-04,'wo') uo2.add_nuclide('U235' ,4.36326E-02,'wo') uo2.add_nuclide('U238' ,8.37445E-01,'wo') uo2.depletable = True borated_water = openmc.Material(material_id=4, name='Borated water') borated_water.set_density('g/cm3', 0.7245) borated_water.temperature = 600.0 borated_water.add_nuclide('B10',8.40178E-05,'wo') borated_water.add_nuclide('B11',3.71839E-04,'wo') borated_water.add_nuclide('H1', 1.11832E-01,'wo') borated_water.add_nuclide('H2', 3.35290E-05,'wo') borated_water.add_nuclide('O16',8.85414E-01,'wo' ) borated_water.add_nuclide('O17',2.26386E-03,'wo' ) borated_water.add_s_alpha_beta('c_H_in_H2O') ############################################################################### # Create geometry ############################################################################### ## Instantiate ZCylinder surfaces fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.41266, name='Fuel OR') clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.474364, name='Clad IR') left = openmc.XPlane(surface_id=4, x0=-0.632458, name='left') right = openmc.XPlane(surface_id=5, x0=0.632458, name='right') bottom = openmc.YPlane(surface_id=6, y0=-0.632458, name='bottom') top = openmc.YPlane(surface_id=7, y0=0.632458, name='top') up = openmc.ZPlane(surface_id=8, z0=50.0, name='up') down = openmc.ZPlane(surface_id=9, z0=-50.0, name='down') left.boundary_type = 'reflective' right.boundary_type = 'reflective' top.boundary_type = 'reflective' bottom.boundary_type = 'reflective' up.boundary_type = 'reflective' down.boundary_type = 'reflective' ## Instantiate Cells fuel = openmc.Cell(cell_id=1, name='cell 1') clad = openmc.Cell(cell_id=3, name='cell 3') water = openmc.Cell(cell_id=4, name='cell 4') ## Use surface half-spaces to define regions fuel.region = -fuel_or & -up & +down clad.region = +fuel_or & -clad_ir & -up & +down water.region = +clad_ir & +left & -right & +bottom & -top & -up & +down ## Register Materials with Cells fuel.fill = uo2 clad.fill = zircaloy water.fill = borated_water ## Instantiate Universe root = openmc.Universe(universe_id=0, name='root universe') ## Register Cells with Universe root.add_cells([fuel, clad, water]) ## Instantiate a Geometry, register the root Universe geometry = openmc.Geometry(root) ############################################################################### # Set volumes of depletable materials ############################################################################### ## Compute cell areas ## area = {} ## area[fuel] = np.pi * fuel_or.coefficients['r'] ** 2 ## Set materials volume for depletion. Set to an area for 2D simulations uo2.volume = 53.49763956 ############################################################################### # Transport calculation settings ############################################################################### ## Instantiate a Settings object, set all runtime parameters, and export to XML settings_file = openmc.Settings() settings_file.batches = batches settings_file.inactive = inactive settings_file.particles = particles settings_file.photon_transport = True ## Create an initial uniform spatial source distribution over fissionable zones bounds = [-0.632458, -0.632458, -50, 0.632458, 0.632458, 50] # uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:]) uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True) settings_file.source = openmc.source.IndependentSource(space=uniform_dist) ## settings_file.electron_treatment = 'ttb' ## settings_file.cutoff = {'energy_photon' : 1000.0} #entropy_mesh = openmc.RegularMesh() #entropy_mesh.lower_left = [-0.39218, -0.39218, -50] #entropy_mesh.upper_right = [0.39218, 0.39218, 50] #entropy_mesh.dimension = [10, 10, 1] #settings_file.entropy_mesh = entropy_mesh ############################################################################### # Initialize and run depletion calculation ############################################################################### tallies_file = openmc.Tallies() t1 = openmc.Tally(name='edep') t1.filters = [openmc.MaterialFilter([uo2,zircaloy,borated_water])] #t1.filters.append(openmc.ParticleFilter(['neutron'])) #,'photon','electron','positron'])) t1.scores = ['flux','fission', 'nu-fission','heating-local','heating'] tallies_file.append(t1) tallies_file.export_to_xml() fission_q={"U232": 193.0442774*1E6, "U233": 199.7961836*1E6, "U234": 200.632851*1E6, "U235": 202.27*1E6, "U236": 203.4043119*1E6, "U237": 196.429643*1E6, "U238": 206.8513817*1E6, "Np237": 205.3704803*1E6, "Np238": 208.6993709*1E6, "Pu238": 209.5400125*1E6, "Pu239": 208.0185328*1E6, "Pu240": 208.6125667*1E6, "Pu241": 211.2377153*1E6, "Pu242": 212.0721865*1E6, "Am241": 211.2167986*1E6, "Am242_m1": 215.1453706*1E6, "Am243": 212.952779*1E6, "Cm242": 212.7864913*1E6, "Cm243": 213.375296*1E6, "Cm244": 217.9267669*1E6, "Cm245": 214.6240222*1E6, "Cm246": 220.179494*1E6, "Th232": 197.1083894*1E6, "Pa231": 194.0999425*1E6} ## op = openmc.deplete.Operator(geometry, settings_file, chain_file,fission_q=fission_q) op = openmc.deplete.Operator(geometry, settings_file, chain_file, energy_mode="energy-deposition") ## Perform simulation using the predictor algorithm integrator = openmc.deplete.PredictorIntegrator(op, time_steps, power) integrator.integrate() ## openmc.run(mpi_args=['mpiexec', '-n', '32', '--bind-to', 'core']) ############################################################################### # Read depletion calculation results ############################################################################### ## Open results file #results = openmc.deplete.ResultsList.from_hdf5("depletion_results.h5") ## Obtain K_eff as a function of time #time, keff = results.get_eigenvalue() ## Obtain U235 concentration as a function of time #time, n_U235 = results.get_atoms('1', 'U235') ## Obtain Xe135 absorption as a function of time #time, Xe_gam = results.get_reaction_rate('1', 'Xe135', '(n,gamma)') # t_days, n_cm3 = res.get_atoms("1", "Xe135", nuc_units="atom/cm3", time_units="d") ############################################################################### # Generate plots ############################################################################### #plt.figure() #plt.plot(time/(24*60*60), keff, label="K-effective") #plt.xlabel("Time (days)") #plt.ylabel("Keff") #plt.show() #plt.figure() #plt.plot(time/(24*60*60), n_U235, label="U 235") #plt.xlabel("Time (days)") #plt.ylabel("n U5 (-)") #plt.show() #plt.figure() #plt.plot(time/(24*60*60), Xe_gam, label="Xe135 absorption") #plt.xlabel("Time (days)") #plt.ylabel("RR (-)") #plt.show() #plt.close('all') #tallies_file = openmc.Tallies() #energy_filter = openmc.EnergyFilter([0., 20.0e6]) ## Instantiate flux Tally in moderator and fuel #tally = openmc.Tally(name='flux') #tally.filters = [openmc.CellFilter(fuel)] #tally.filters.append(energy_filter) #tally.scores = ['flux'] #tallies_file.append(tally) ## Instantiate reaction rate Tally in moderator #tally = openmc.Tally(name='fuel rxn rates') #tally.filters = [openmc.CellFilter(fuel)] #tally.filters.append(energy_filter) #tally.scores = ['fission'] #tally.nuclides = ['O16', 'H1'] #tallies_file.append(tally) #tallies_file.export_to_xml()