# Plot yz slices
plotter.plotSlice(geometry, plane='YZ', lim1=[-1., 1.], lim2=[-1., 1.], \
filename='fuel', gridsize=50)
plotter.plotSlice(geometry, plane='YZ', lim1=[-1., 1.], lim2=[-1., 1.], \
filename='moderator', gridsize=50)
py_printf('INFO', 'Plotting the fission site source distribution...')
plotter.plotFissionSourceDist(geometry)
py_printf('INFO', 'Plotting neutron tracks...')
plotter.trackANeutron(geometry, num_moves=100, plane='xy', \
lim1=[-0.75, 0.75], lim2=[-0.75,0.75])
plotter.trackANeutron(geometry, num_moves=100, plane='yz', \
lim1=[-0.75, 0.75], lim2=[-3., 3.])
plotter.trackANeutron(geometry, num_moves=100, plane='xz', \
lim1=[-0.75, 0.75], lim2=[-3.,3.])
# Plot the fluxes
plotter.plotFlux([total_flux, moderator_flux, fuel_flux], uselegend=True)
# Compute k-infinity using tally arithmetic operators and print to screen
k_inf = tot_fiss_rate * nu / tot_abs_rate
k_inf.setTallyName('k-infinity')
k_inf.printTallies(uncertainties=True)
py_printf('TITLE', 'Finished')
elastic_xs = process.GroupXS(total_flux_xs, elastic_rate)
capture_xs = process.GroupXS(total_flux_xs, capture_rate)
fission_xs = process.GroupXS(total_flux_xs, fission_rate)
absorb_xs = process.GroupXS(total_flux_xs, absorb_rate)
transport_xs = process.GroupXS(total_flux_xs, transport_rate)
diffusion_coeff = process.GroupXS(total_flux_xs, diffusion_rate)
total_xs = process.GroupXS(total_flux_xs, total_rate)
elastic_xs.printXS(uncertainties=True)
capture_xs.printXS(uncertainties=True)
fission_xs.printXS(uncertainties=True)
absorb_xs.printXS(uncertainties=True)
transport_xs.printXS(uncertainties=True)
diffusion_coeff.printXS(uncertainties=True)
total_xs.printXS(uncertainties=True)
# Compute k-infinity using tally arithmetic operators and print to screen
k_inf = tot_fiss_rate * nu / tot_abs_rate
k_inf.setTallyName('k-infinity')
k_inf.printTallies(uncertainties=True)
# Compute moderator-to-fuel flux ratios and print to screen
flux_ratio = moderator_flux_ratio / fuel_flux_ratio
flux_ratio.setTallyName('Moderator-to-Fuel Flux Ratios')
flux_ratio.printTallies(uncertainties=True)
# Plot the fluxes
plotter.plotFlux([total_flux, moderator_flux, fuel_flux])
py_printf('TITLE', 'Finished')
################################ Create Tallies ##############################
###############################################################################
py_printf('NORMAL', 'Initializing flux tally...')
flux = TallyFactory.createTally(region_mix, FLUX)
flux.generateBinEdges(1E-2, 1E7, 10000, LOGARITHMIC)
TallyBank.registerTally(flux)
###############################################################################
######################### Run Monte Carlo Simulation #########################
###############################################################################
# Run Monte Carlo simulation
geometry.runMonteCarloSimulation()
###############################################################################
############################ Process Output Data #############################
###############################################################################
py_printf('INFO', 'Plotting flux...')
plotter.plotFlux(flux, uselegend=False, filename='flux', \
title='Infinite medium flux')
py_printf('INFO', 'Writing tally batch statistics to output file...')
TallyBank.outputBatchStatistics()
py_printf('TITLE', 'Finished')
RI = process.RIEff(abs_rate_flux, u238_abs_rate)
RI.setName('Effective RI (U/H=%1.0E)' % u_h_ratios[u_h_ratio])
Eff_RIs[temp].append(RI)
# Compute absorption rate ratios
abs_rate_ratio = u238_abs_rate / tot_abs_rate
abs_rate_ratio.setTallyName('U238 / Tot for U/H=%1.0E' \
% u_h_ratios[u_h_ratio])
abs_rate_ratios[temp].append(abs_rate_ratio)
# Append flux for this simulation to an array and deregister it
# for next simulation
fluxes[u_h_ratio].normalizeBatchMu()
TallyBank.deregisterTally(fluxes[u_h_ratio])
# Plot fluxes
plotter.plotFlux(fluxes,title='Flux for Temp = ' + str(temps[temp]) \
+ 'K', filename='flux-temp-'+str(temps[temp])+'K', \
uselegend=True)
# print the reaction rate ratios and resonance integrals to the shell
process.printTallies(abs_rate_ratios[temp],
header='U238 Abs. / Tot. Abs.')
process.printRIs(Eff_RIs[temp], header='Temp=%dK'%temps[temp])
process.printRIs(True_RIs, header='T=300-1200K')
py_printf('TITLE', 'Finished')
h1.collideNeutron(neutron)
h1_fluxes[j].tally(neutron)
py_printf('INFO', 'Simulating generational flux from C-12...')
for i in range(num_neutrons):
neutron._energy = max_energy # initialize energy to 2 MeV
for j in range(num_generations):
c12.collideNeutron(neutron)
c12_fluxes[j].tally(neutron)
py_printf('INFO', 'Plotting the generational fluxes...')
plotter.plotFlux(h1_fluxes[1:], loglog=False, uselegend=False, \
filename='h1-generational-flux', title='H1 Generational Flux')
plotter.plotFlux(c12_fluxes[1:], loglog=False, uselegend=False,
filename='c12-generational-flux', title='C12 Generational Flux')
###############################################################################
########################### Problems 3 and 4 ##############################
###############################################################################
num_generations = 50
py_printf('HEADER', 'Problems 3 and 4')
py_printf('INFO', 'Initializing lethargy and energy flux tallies...')
lethargy_flux = TallyFactory.createTally(c12, COLLISION_RATE)
elastic_xs = process.GroupXS(total_flux_xs, elastic_rate)
capture_xs = process.GroupXS(total_flux_xs, capture_rate)
fission_xs = process.GroupXS(total_flux_xs, fission_rate)
absorb_xs = process.GroupXS(total_flux_xs, absorb_rate)
transport_xs = process.GroupXS(total_flux_xs, transport_rate)
diffusion_coeff = process.GroupXS(total_flux_xs, diffusion_rate)
total_xs = process.GroupXS(total_flux_xs, total_rate)
elastic_xs.printXS(uncertainties=True)
capture_xs.printXS(uncertainties=True)
fission_xs.printXS(uncertainties=True)
absorb_xs.printXS(uncertainties=True)
transport_xs.printXS(uncertainties=True)
diffusion_coeff.printXS(uncertainties=True)
total_xs.printXS(uncertainties=True)
# Compute k-infinity using tally arithmetic operators and print to screen
k_inf = tot_fiss_rate * nu / tot_abs_rate
k_inf.setTallyName('k-infinity')
k_inf.printTallies(uncertainties=True)
# Compute moderator-to-fuel flux ratios and print to screen
flux_ratio = moderator_flux_ratio / fuel_flux_ratio
flux_ratio.setTallyName('Moderator-to-Fuel Flux Ratios')
flux_ratio.printTallies(uncertainties=True)
# Plot the fluxes
plotter.plotFlux([total_flux, moderator_flux, fuel_flux])
py_printf('TITLE', 'Finished')
reached_one_ev = False
# Simulate neutron until it is absorbed in H-1
while(neutron._alive):
h1_material.collideNeutron(neutron)
flux.tally(neutron)
times.tally(neutron)
coll_rate.tally(neutron)
if neutron._energy < 1.0:
reached_one_ev = True
if not reached_one_ev:
coll_rate_1eV.tally(neutron)
py_printf('INFO', 'Plotting the flux...')
plotter.plotFlux(flux, title='H-1 Flux', filename='h-1-flux')
num_collisions = process.computeMeanNumCollisions(coll_rate_1eV, num_neutrons)
py_printf('RESULT', 'Mean # of collisions to 1 eV: %f', num_collisions)
num_collisions = process.computeMeanNumCollisions(coll_rate, num_neutrons)
py_printf('RESULT', 'Mean # of collisions to death: %f', num_collisions)
mean_lifetime = process.computeMeanNeutronLifetime(times, num_neutrons)
py_printf('RESULT', 'Avg neutron lifetime: %1.2E seconds', mean_lifetime)
py_printf('TITLE', 'Finished')
neutron._energy = fissioner.emitNeutroneV()
neutron._alive = True
reached_one_ev = False
# Simulate neutron until it is absorbed in H-1
while (neutron._alive):
h1_material.collideNeutron(neutron)
flux.tally(neutron)
times.tally(neutron)
coll_rate.tally(neutron)
if neutron._energy < 1.0:
reached_one_ev = True
if not reached_one_ev:
coll_rate_1eV.tally(neutron)
py_printf('INFO', 'Plotting the flux...')
plotter.plotFlux(flux, title='H-1 Flux', filename='h-1-flux')
num_collisions = process.computeMeanNumCollisions(coll_rate_1eV, num_neutrons)
py_printf('RESULT', 'Mean # of collisions to 1 eV: %f', num_collisions)
num_collisions = process.computeMeanNumCollisions(coll_rate, num_neutrons)
py_printf('RESULT', 'Mean # of collisions to death: %f', num_collisions)
mean_lifetime = process.computeMeanNeutronLifetime(times, num_neutrons)
py_printf('RESULT', 'Avg neutron lifetime: %1.2E seconds', mean_lifetime)
py_printf('TITLE', 'Finished')
geometry.addRegion(region_mix)
###############################################################################
################################ Create Tallies ##############################
###############################################################################
py_printf('NORMAL', 'Initializing flux tally...')
flux = TallyFactory.createTally(region_mix, FLUX)
flux.generateBinEdges(1E-2, 1E7, 10000, LOGARITHMIC)
TallyBank.registerTally(flux)
###############################################################################
######################### Run Monte Carlo Simulation #########################
###############################################################################
# Run Monte Carlo simulation
geometry.runMonteCarloSimulation()
###############################################################################
############################ Process Output Data #############################
###############################################################################
py_printf('INFO', 'Plotting flux...')
plotter.plotFlux(flux, uselegend=False, filename='flux', \
title='Infinite medium flux')
py_printf('INFO', 'Writing tally batch statistics to output file...')
TallyBank.outputBatchStatistics()
py_printf('TITLE', 'Finished')
