adding icf neutron spectra

[shimwell]

nesst has some nice neutron spectra, could include them as examples

[shimwell]

example making an openmc source term from NeSST

import matplotlib.pyplot as plt
import NeSST as nst
import numpy as np
import openmc
import openmc_source_plotter  # extends openmc.Source with plotting functions

Ion_temperature = 2.0  # keV

# Atomic fraction of D and T in scattering medium and source
nst.frac_D_default = 0.5
nst.frac_T_default = 0.5

DTmean, DTvar = nst.DTprimspecmoments(Ion_temperature)
DDmean, DDvar = nst.DDprimspecmoments(Ion_temperature)

Y_DT = 1.0  # DT neutron yield, all reactions scaled by this value
Y_DD = nst.yield_from_dt_yield_ratio("dd", Y_DT, Ion_temperature)
Y_TT = nst.yield_from_dt_yield_ratio("tt", Y_DT, Ion_temperature)

# single grid for DT, DD and TT grid
E_pspec = np.linspace(0, 20, 500)

dNdE_DT = Y_DT * nst.Qb(E_pspec, DTmean, DTvar)  # Brysk shape i.e. Gaussian
dNdE_DD = Y_DD * nst.Qb(E_pspec, DDmean, DDvar)  # Brysk shape i.e. Gaussian
dNdE_TT = Y_TT * nst.dNdE_TT(E_pspec, Ion_temperature)

dNdE_DT_DD_TT = dNdE_DT + dNdE_DD + dNdE_TT

my_source = openmc.IndependentSource()
my_source.space = openmc.stats.Point((0, 0, 0))
my_source.angle = openmc.stats.Isotropic()
my_source.energy = openmc.stats.Discrete(E_pspec * 1e6, dNdE_DT_DD_TT)
plot = my_source.plot_source_energy(n_samples=200000)

plot.update_yaxes(type="log")

plot.show()

[shimwell]

A material representing the fuel and pusher can then be made. Such as this revolver based one

def icf_neutron_source(
    materials: typing.Sequence[openmc.Material],
    radial_thicknesses: typing.Sequence[float],
):
    cells = []
    moving_radius = 0
    for i, (radius, material) in enumerate(zip(radial_thicknesses,materials)):
        moving_radius= moving_radius+radius
        surface = openmc.Sphere(r=moving_radius)
        if i == 0:
            region = -surface
        else:
            region = -surface & +surfaces[-1]
        cell = openmc.Cell(region=region, fill=material)
        cells.append(cell)

    my_materials = openmc.Materials(materials)
    geometry = openmc.Geometry(cells)

    source = openmc.IndependentSource()
    source.space = openmc.stats.Box(*cells[0].bounding_box) # neutrons created in the inner cell only
    source.domains=[cells[0]]  # reduces the box space to just the cell
    source.angle = openmc.stats.Isotropic()
    # source.energy should be set to the code in the first example

pusher_material = openmc.Material(name='Revolver')
pusher_material.add_element("Au", 1.0, percent_type="ao")
pusher_material.set_density("g/cm3", 1930)

fuel_material = openmc.Material(name='NIC Rev 5')
fuel_pusher_material.add_nuclide("H2", 0.5, percent_type="ao")
fuel_pusher_material.add_nuclide("H3", 0.5, percent_type="ao")
fuel_pusher_material.set_density("g/cm3", 300)

icf_neutron_source(
    materials=[pusher_material, fuel_material],
    radial_thicknesses = [0.0035, 0.0095-0.0035]
)