pyopensn.xs.MultiGroupXS

class pyopensn.xs.MultiGroupXS

Multi-group cross section.

Wrapper of opensn::MultiGroupXS.

CreateSimpleOneGroup(self: pyopensn.xs.MultiGroupXS, sigma_t: float, c: float) → None

Create a one-group cross section.

Parameters:

sigma_t (float) – Total cross section.
c (float) – Scattering ratio.

LoadFromOpenMC(self: pyopensn.xs.MultiGroupXS, file_name: str, dataset_name: str, temperature: float) → None: Load multi-group cross sections from an OpenMC cross-section file.

LoadFromOpenSn(self: pyopensn.xs.MultiGroupXS, file_name: str) → None

Load multi-group cross sections from an OpenSn cross section input file.

Format is as follows (for transfers, gprime denotes the departing group and g is the arrival group).

# Add comment lines, as needed
NUM_GROUPS ng
NUM_MOMENTS nmom

SIGMA_T_BEGIN
0 value
.
.
ng-1 value
SIGMA_T_END

SIGMA_A_BEGIN
0 value
.
.
ng-1 value
SIGMA_A_END

TRANSFER_MOMENTS_BEGIN
M_GPRIME_G_VAL 0 0 0 value
.
M_GPRIME_G_VAL moment gprime g value
.
M_GPRIME_G_VAL nmom-1 ng-1 ng-1 value
TRANSFER_MOMENTS_END

SetScalingFactor(self: pyopensn.xs.MultiGroupXS, factor: float) → None: Scale the cross sections by the specified factor.

__init__(self: pyopensn.xs.MultiGroupXS) → None: Create an empty multi-group cross section.

property chi: Get neutron fission spectrum.

property inv_velocity: Get inverse velocity.

property is_fissionable: Check if the material is fissile.

property nu_delayed_sigma_f: Get delayed neutron production due to fission.

property nu_prompt_sigma_f: Get prompt neutron production due to fission.

property nu_sigma_f: Get neutron production due to fission.

property num_groups: Get number of energy groups.

property num_precursors: Get number of precursors.

property scaling_factor: Get the arbitrary scaling factor.

property scattering_order: Get Legendre scattering order.

property sigma_a: Get absorption cross section.

property sigma_f: Get fission cross section.

property sigma_t: Get total cross section.