A monte carlo pin cell spectral code for nuclear engineering applications.
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- _ -
_A :
Isotope
_A_plus_one_squared :
Isotope
_A_squared :
Isotope
_absorb_xs :
Isotope
_absorb_xs_energies :
Isotope
_alive :
neutron
_all_tallies :
TallyBank
_alpha :
Isotope
_alpha1 :
Geometry
_alpha2 :
Geometry
_AO :
Isotope
_batch_mu :
Tally
_batch_num :
neutron
_batch_rel_err :
Tally
_batch_std_dev :
Tally
_batch_variance :
Tally
_beta :
Geometry
_bin_delta :
Tally
_bin_spacing :
Tally
_boundary_type :
Surface
_buckling_squared :
Geometry
,
Material
,
Region
_capture_rescaled :
Isotope
_capture_xs :
Isotope
_capture_xs_energies :
Isotope
_cdf :
Fissioner
_cdf_energies :
Fissioner
_centers :
Tally
_collided :
neutron
_computed_statistics :
Tally
_dancoff :
Geometry
_delta_lethargy :
EquivalenceRegion
,
Isotope
_density_unit :
Material
_E_max :
Fissioner
_E_to_kT :
Isotope
_edges :
Tally
_elapsed_time :
Timer
_elastic_rescaled :
Isotope
_elastic_xs :
Isotope
_elastic_xs_energies :
Isotope
_end_lethargy :
EquivalenceRegion
,
Isotope
_end_time :
Timer
_energy :
neutron
_energy_bands :
pinspec.process.RITrue
_Eprime_to_E :
Isotope
_eta :
Isotope
_fission_rescaled :
Isotope
_fission_xs :
Isotope
_fission_xs_energies :
Isotope
_fissionable :
Isotope
_fissioner :
Geometry
_flux :
pinspec.process.GroupXS
,
pinspec.process.RIEff
_fuel :
Geometry
_fuel_radius :
EquivalenceRegion
,
Geometry
_geometry :
GeometryTally
_geometry_name :
Geometry
_geometry_tallies :
TallyBank
_group_expand_bins :
Tally
_half_width :
EquivalenceRegion
_infinite_medium :
Geometry
_isotope :
IsotopeTally
,
neutron
,
pinspec.process.RITrue
_isotope_name :
Isotope
_isotope_tallies :
TallyBank
_isotopes :
Material
_isotopes_AO :
Material
_kB :
Isotope
_material :
MaterialTally
,
neutron
,
Region
_material_atomic_mass :
Material
_material_density :
Material
_material_name :
Material
_material_number_density :
Material
_material_tallies :
TallyBank
_moderator :
Geometry
_mu_avg :
Isotope
_n :
Isotope
,
Material
,
Region
,
Surface
_name :
pinspec.process.GroupXS
,
pinspec.process.RIEff
,
pinspec.process.RITrue
_num_absorb_xs :
Isotope
_num_batches :
Geometry
,
Tally
_num_bins :
Fissioner
,
Tally
_num_capture_xs :
Isotope
_num_edges :
Tally
_num_elastic_xs :
Isotope
_num_energies :
Isotope
_num_fission_xs :
Isotope
_num_neutrons_per_batch :
Geometry
_num_prob :
EquivalenceRegion
,
Geometry
_num_RIs :
pinspec.process.RIEff
,
pinspec.process.RITrue
_num_thermal_cdf_bins :
Isotope
_num_thermal_cdfs :
Isotope
_num_threads :
Geometry
_num_total_xs :
Isotope
_num_xs :
pinspec.process.GroupXS
_old_energy :
neutron
_other_region :
EquivalenceRegion
_path_length :
neutron
_pitch :
EquivalenceRegion
,
Geometry
_prob_energies :
EquivalenceRegion
,
Geometry
_prob_ff :
EquivalenceRegion
,
Geometry
_prob_mf :
EquivalenceRegion
,
Geometry
_r :
ZCylinder
_r_squared :
ZCylinder
_rate :
pinspec.process.GroupXS
,
pinspec.process.RIEff
_reaction :
pinspec.process.RITrue
_region :
neutron
,
RegionTally
_region_name :
Region
_region_tallies :
TallyBank
_region_type :
Region
_regions :
Geometry
_rescaled :
Isotope
_rho :
Isotope
_RI :
pinspec.process.RIEff
_RIs :
pinspec.process.RITrue
_running :
Timer
_sigma_e :
Geometry
_source_sampling_radius :
Geometry
_spatial_type :
Geometry
_start_lethargy :
EquivalenceRegion
,
Isotope
_start_time :
Timer
_surface :
neutron
_surface_name :
Surface
_surface_type :
Surface
_surfaces :
BoundedRegion
_T :
Isotope
_tallies :
Tally
_tally_domain :
Tally
_tally_name :
Tally
_tally_type :
Tally
_thermal_cdfs :
Isotope
_thermal_cutoff :
Isotope
_thermal_dist :
Isotope
_timer_splits :
Timer
_total_xs :
Isotope
_total_xs_energies :
Isotope
_trigger_precision :
Tally
_trigger_type :
Tally
_u :
neutron
_uid :
Isotope
,
Material
,
Region
,
Surface
_use_thermal_scattering :
Isotope
_v :
neutron
_volume :
Material
,
Region
_w :
neutron
_x :
neutron
,
XPlane
_x0 :
ZCylinder
_xs :
pinspec.process.GroupXS
_y :
neutron
,
YPlane
_y0 :
ZCylinder
_z :
neutron
Generated on Wed Oct 1 2014 09:28:14 by
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