onix
– Utilities and functions¶
Reaction classes¶
These classes enable users to build their own custom nuclear libraries. Instantiations of these classes can be used as nuclear libraries for simulations.
Decay_lib objects allow users to build and store custom, small size decay libraries. |
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xs_lib objects allow users to build and store custom, small size one-group cross section libraries. |
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fy_lib objects allow users to build and store custom, small size fission yield libraries. |
Functions¶
Convenient functions to be used directly by users and used in the source code.
Converts a decay constant into a half life in specified units. |
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Converts a half life into a decay constant. |
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Converts a half life from specified units into seconds. |
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Check whether an object is an integer. |
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Check whether a string is a nuclide’s z-a-m id. |
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Check whether a string is a nuclide’s name. |
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Check whether a list is redundant. |
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Returns the redundant elements in a list. |
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Converts a list of nuclides’ z-a-m ids into a list of names. |
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Converts a list of nuclides’ names into a list of z-a-m ids. |
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Gets the atomic number (z) from a nuclide’s z-a-m id. |
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Gets the atomic number (z) from a nuclide’s name. |
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Gets the mass number (a) from a nuclide’s z-a-m id. |
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Gets the number of neutrons (n) from a nuclide’s z-a-m id. |
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Gets the state of a nuclide from a its z-a-m id. |
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Converts a nuclide’s z-a-m id into the nuclide’s name. |
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Converts a nuclide’s name into the nuclide’s z-a-m id. |
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Gets the mass of heavy metal in a Passlist object. |
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Gets the atomic mass of a nuclide (in grams). |
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Converts the mass quantity (in grams) of a given nuclide type into number of atoms. |
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Converts the quantity of a given nuclide from number of atoms into mass (in grams). |
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Computes the factor that converts seconds into burnup units (MWd/kg) from a given volume and a Initial Heavy Metal mass (IHM). |
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Gets the list of nuclides from a decay dictionnary. |
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Gets the list of nuclides from a cross section dictionnary. |
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Gets the list of fission products from a fission yield dictionnary. |
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Check whether elements from a list (lista) are all contained in another list (listb). |
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Gets the list of fission parents from a fission yield dictionnary. |
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Check whether s is a float. |
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Converts a nuclide’s name written with OpenMC format (‘U235_m1’) into ONIX format (‘U-235*’). |
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Converts a nuclide’s name written with ONIX format (‘U-235*’) into OpenMC format (‘U235_m1’). |
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Converts a list of nuclides’ names written with ONIX format (‘U-235*’) into OpenMC format (‘U235_m1’). |
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Converts a list of nuclides’ names written with OpenMC format (‘U235_m1’) into ONIX format (‘U-235*’). |
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Orders a list of nuclides’ z-a-m ids according to their atomic number (z). |
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Orders a list of nuclides’ names (in OpenMC format) according to their atomic number (z). |
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Orders a list of nuclides’ z-a-m ids according to their mass number (a). |
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Returns the list of nuclides for which there are cross section data in a specified HDF5 cross section library directory (produced with OpenMC). |
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Converts a neutron spectrum output file computed in an ONIX simulations to a format suitable to be used to fold cross sections in JANIS4.1. |
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Gets the natural abundance of a nuclide (values from 0 to 1). |
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Gets the natural abundance of a nuclide (values from 0 to 1). |
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Finds the precursor of a nuclide via a specified reaction (except fission reactions). |
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Linearly interpolates between two points to find the ordinate to a given abscissa value (x). |
Data processors¶
Various functions to extract, read and visualize output results.
Plots a network diagram of the destruction and production reaction rates of a specified nuclide at a given macrostep for a given BUCell. |
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Plots the density evolution of a given nuclide (in \(atm barn^{-2}cm{-1}\)) in a given BUCell against time (in days). |
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Plots the total density evolution of a group of nuclides (in \(atm barn^{-2}cm{-1}\)) in a given BUCell against time (in days). |
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Plots the one-group cross section evolution (in barn) of a given reaction for a given nuclide in a given BUCell against time (in days). |
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Plots the one-group cross section evolution (in barn) of a given reaction for a given nuclide in a given BUCell against burnup (MWd/kg). |
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Plots multiple cross sections from multiple simulation directories in a series of subplots with the same burnup y-axis. |
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Plots the multiplication factor against time (in days). |
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Plots the neutron flux against time (in days). |
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Plots the neutron spectrum for different macrosteps for multiple BUCells. |
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Plots the neutron spectrum in lethargy units for different macrosteps for multiple BUCells. |
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Plots the evolution of a cross section, a nuclide density and the neutron flux in a specified BUCell in three subplots witht the same y-axis in days. |
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Reads the time sequence from a simulation’s directory and returns a list of the time sequence. |
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Creates a list with time length between each macrosteps. |
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Reads the burnup sequence from a simulation’s directory and returns a list of the burnup sequence. |
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Reads the multiplication factor evolution from a simulation’s directory and returns a list of the multiplication factor evolution. |
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Reads the macrostep-wise neutron flux evolution from a simulation’s directory and returns a list with the neutron flux evolution. |
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Reads the microstep-wise neutron flux evolution from a simulation’s directory and returns a list with the neutron flux evolution. |
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Returns a list with the macrostep-wise fluence evolution in a specificed BUCell |
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Returns a list with the microstep-wise fluence evolution in a specificed BUCell |
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Reads the neutron flux spectra for multiple macrosteps from a flux spectrum output file and returns it in a list. |
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Reads the mid-points of the energy groups against which the multi-group neutron flux is stored and returns it in a list. |
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Reads the energy interval of each energy bin group used to store the multi-group neutron flux and returns it in a list. |
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Reads the density (in \(atm barn^{-2}cm^{-1}\)) of a specified nuclide from the provided density file. |
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Returns the total density of the material (in \(atm barn^{-2}cm^{-1}\)) in a BUCell at a given macrostep. |
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Returns the total mass density of the material (in \(g cm^{-3}\)) in a BUCell at a given macrostep. |
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Reads the cross section evolution (in barn) for a specified reactions for a given nuclide in a given BUCell and returns it in a list. |
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Computes time-averaged cross section (in barn) for a specified reactions for a given nuclide in a given BUCell. |
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Computes the time-averaged neutron flux (in \(cm^{-2}s^{-1}\)) in a given BUCell. |
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Computes the total one-group cross section evolution (in barn) for a specified nuclide in a given BUCell. |
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Returns the list of nuclides for which one-group cross sections have been calculated during a simulation in a given BUCell. |
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Returns the list of nuclides for which densities have been calculated during a simulation in a given BUCell. |
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Print a ranking of nuclides in a BUCell according to their densities for multiple macrosteps in a text file. |
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Plots the transmuation matrix for a given BUCell at a given macrostep. |
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Plots a chart of the nuclide where nuclides are colored only if present in the decay, cross section or fission yield library. |
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Plots a chart of the nuclide where nuclides are colored only if present in the decay, cross section or fission yield library. |