Fusion Energy Neutronics Workshop

This work was funded by the RCUK Energy Programme
[Grant number EP/P012450/1]

Neutron creation

(n,n')

(n,f)

(n,n')

(γ,γ')

(n,nγ')

(n,pn')

(n,f)

(n,2n)

(n,α)

(n,γ)

Cerberus

Nuclear

How can neutronics help

Radioactivity - Neutrons activate material, making it radioactive leading to handling and waste storage requirements.

Hazardous - Neutrons are Hazardous to health and shielded will be needed to protect the workforce.

Produce fuel - Neutrons will be needed to convert lithium into tritium to fuel the reactor.

Electricity - 80% of the energy release by each DT reaction is transferred to the neutron.

Structural integrity - Neutrons cause damage to materials such as embrittlement, swelling, change conductivity …

Diagnose - Neutrons are an important method of measuring a variety of plasma parameters (e.g. Q value).

Topics Covered

Material

Geometry

Tally

Source

Cross sections

Materials

Li4SiO4, lithium lead, tungsten

Particle sources (point source, plasma, gamma)

Geometry (CSG and CAD)

Tallies (TBR, heat, spectra, dose, DPA)

Tally types (cell, mesh, surface)

Parameter studies

How to get setup

Install Docker

Download the docker image

Run the docker image

Navigate to the URL in the terminal

Install Paraview and Freecad

Instructions are on the Github repository along with a setup video

docker run -p 8888:8888 ghcr.io/fusion-energy/neutronics-workshop

docker pull ghcr.io/fusion-energy/neutronics-workshop

What are the tasks like

Collection of Jupyter notebooks for each task

Tasks have comments and instructions within the notebook

Outputs include
- numbers
- graphs
- images
- 3D files

Learning outcomes from each task

Why containerisation

OpenMC popularity

https://star-history.t9t.io/#openmc-dev/openmc

Remeber to shutdown the tasks

If your graphs are not showing it could be due to limited memory.

Switch to the Running tab

Introduction to the workshop

14:10

Introduction

Task 1 - Cross sections - parts 1, 2, 3, 4

14:00

Task 2 - Materials - parts 1, 2

Task 3 - CSG geometry parts 1, 3

Break

Task 4 - Sources - parts 1, 2, 3

Task 5 - TBR - part 1

Task 6 - DPA - part 1

Task 7 - Neutron and photon spectra - part 1 and part 2

14:40

15:15

15:45

16:20

16:30

17:15

17:30

17:55

Task 8 - Mesh tallies - parts 1 and 2

Task 10 - Making CAD geometry - part 1 and 2

15:00

17:45

Break

End of course survey www.surveymonkey.co.uk/r/W5CJ2PY

17:00

Task 1

Cross Sections

Press down to see the next slide

Cross section data is key to the neutronics workflow and provide us with the likelihood of a particular interaction.

Microscopic cross section

Cross sections can be measured experimentally with monoenergetic neutrons.

Probability of interaction is characterised by the microscopic cross-section (σ). It is the effective size of the nucleus.

Experimental data

Availability of experimental data varies for different reactions and different isotopes.

Typically the experimental data is then interpreted to create evaluation libraries, such as ENDF, JEFF, JENDL, CENDL.

Cross section evaluations exist for:

different isotopes
different interactions.

Interaction cross sections

www.oecd-nea.org/dbdata/data/manual-endf/endf102_MT.pdf

A list of interactions is available online

MT numbers using in this workshop include:

DPA 444
n,Xt 205

Cross section to reactions

Neutron flux (ɸ) in units of neutrons cm-2.s-1

\phi = n v

n is number of neutrons per cm3 (neutron density)

NA is Avogadro's number (6.022e23 atoms per mole)

\Sigma = N_D \sigma

N_D = \frac{\rho N_A} {M}

RR = \phi \Sigma

The atom number density (ND) is the number of atoms per cm3.

The reaction rate (RR) is the total number of neutron interaction per unit volume. Units of s-1

Macroscopic cross-section (Σ) corresponds to the total equivalent area of all the nuclei considered.

v is the neutron velocity

p is the density in gcm-3

M is the atomic weight in grams

Task 1 - Cross section plotting
Isotopes

Task 1 - Cross section plotting
Elements

Task 1 - Cross section plotting
Materials

Task 1 - Cross section plotting
Doppler broadening

Task 1 - Cross section plotting
Doppler Broadening

Task 2

Materials

Press down to see the next slide

Making materials

Protons

Neutron

Neutronics codes require the isotopes and the number density.

Making materials

import openmc

mat1 = openmc.Material()
mat1.add_element('Li', 2)
mat1.add_element('O', 1)
mat1.set_density('g/cm3', 2.01)


mat2 = openmc.Material()
mat2.add_nuclide('Li6', 0.0759*2)
mat2.add_nuclide('Li7', 0.9241*2)
mat2.add_nuclide('O16', 0.9976206)
mat2.add_nuclide('O17', 0.000379)
mat2.add_nuclide('O18', 0.0020004)
mat2.set_density('g/cm3', 2.01)

Simple material construction from elements or from isotopes.

Enriching materials

Simple material construction with lithium 6 enrichment

import openmc

mat1 = openmc.Material()
mat1.add_element('Li', 2,
                 enrichment_target='Li6',
                 enrichment=60)
mat1.add_element('O', 1)
mat1.set_density('g/cm3', 2.01)


mat2 = openmc.Material()
mat2.add_nuclide('Li6', 0.6*2)
mat2.add_nuclide('Li7', 0.4*2)
mat2.add_nuclide('O16', 0.9976206)
mat2.add_nuclide('O17', 0.000379)
mat2.add_nuclide('O18', 0.0020004)
mat2.set_density('g/cm3', 2.01)

Enriching materials

import openmc

mat1 = openmc.Material()
mat1.add_elements_from_formula('Li2O',
    enrichment_target='Li6',
    enrichment=60)
mat1.set_density('g/cm3', 2.01)

Simple material construction from formula with enrichment.

Making fusion relevant materials

import openmc

mat1 = openmc.Material()
mat1.add_nuclide('Fe56', 0.8149682034000001)
mat1.add_nuclide('Fe57', 0.0188211699)
mat1.add_nuclide('Fe58', 0.0025047522000000003)
mat1.add_nuclide('Fe54', 0.05191587450000001)
mat1.add_nuclide('B10', 1.982e-06)
mat1.add_nuclide('B11', 8.018e-06)
mat1.add_nuclide('C12', 0.0010383680999999998)
mat1.add_nuclide('C13', 1.1631899999999998e-05)
mat1.add_nuclide('N15', 1.4652000000000002e-06)
mat1.add_nuclide('N14', 0.00039853480000000003)
mat1.add_nuclide('O16', 9.976206000000001e-06)
mat1.add_nuclide('O18', 2.0003999999999998e-08)
mat1.add_nuclide('O17', 3.79e-09)
mat1.add_nuclide('Al27', 4e-05)
mat1.add_nuclide('Si29', 1.2176216e-05)
mat1.add_nuclide('Si28', 0.000239797168)
mat1.add_nuclide('Si30', 8.026615999999999e-06)
mat1.add_nuclide('P31', 2e-05)
mat1.add_nuclide('S33', 2.2460700000000002e-07)
mat1.add_nuclide('S32', 2.8512222e-05)
mat1.add_nuclide('S36', 4.374e-09)
mat1.add_nuclide('S34', 1.258797e-06)
mat1.add_nuclide('Ti46', 8.250000000000001e-07)
mat1.add_nuclide('Ti49', 5.410000000000001e-07)
mat1.add_nuclide('Ti50', 5.180000000000001e-07)
mat1.add_nuclide('Ti48', 7.372e-06)
mat1.add_nuclide('Ti47', 7.44e-07)
mat1.add_nuclide('V50', 5e-06)
mat1.add_nuclide('V51', 0.0019950000000000002)
mat1.add_nuclide('Cr53', 0.0085509)
mat1.add_nuclide('Cr50', 0.0039105)
mat1.add_nuclide('Cr52', 0.0754101)
mat1.add_nuclide('Cr54', 0.0021285)
mat1.add_nuclide('Mn55', 0.0055)
mat1.add_nuclide('Co59', 5e-05)
mat1.add_nuclide('Ni60', 2.62231e-05)
mat1.add_nuclide('Ni61', 1.1399e-06)
mat1.add_nuclide('Ni58', 6.80769e-05)
mat1.add_nuclide('Ni64', 9.256000000000001e-07)
mat1.add_nuclide('Ni62', 3.6345000000000005e-06)
mat1.add_nuclide('Cu63', 2.0745000000000002e-05)
mat1.add_nuclide('Cu65', 9.255e-06)
mat1.add_nuclide('Nb93', 5e-05)
mat1.add_nuclide('Mo95', 4.7619e-06)
mat1.add_nuclide('Mo96', 5.0018999999999995e-06)
mat1.add_nuclide('Mo97', 2.8746e-06)
mat1.add_nuclide('Mo94', 2.7560999999999998e-06)
mat1.add_nuclide('Mo92', 4.3947000000000005e-06)
mat1.add_nuclide('Mo98', 7.2876e-06)
mat1.add_nuclide('Mo100', 2.9232000000000002e-06)
mat1.add_nuclide('Ta180', 1.4412e-07)
mat1.add_nuclide('Ta181', 0.0011998558799999998)
mat1.add_nuclide('W183', 0.0015741)
mat1.add_nuclide('W184', 0.0033704)
mat1.add_nuclide('W186', 0.0031273)
mat1.add_nuclide('W180', 1.3199999999999997e-05)
mat1.add_nuclide('W182', 0.002915)
mat1.set_density('g/cm3', 2.01)

There are a large number of isotopes in a material

Making fusion relevant materials

import neutronics_material_maker as nmm

ref_mat = nmm.Material('eurofer')
mat1 = ref_mat.openmc_material

There is an app for that

Open-source

Pip install

Documented

Density with temperature

Materials used in fusion reactors such as water, FLiBe, lithium lead change density with temperature.

The NeutronicsMaterialMaker finds the density automatically

Density with pressure

Materials used in fusion reactors such as helium and CO2 change density with pressure.

The NeutronicsMaterialMaker finds the density automatically

Density with Li6 enrichment

Materials used in fusion reactors such as Li4SiO4 and Li2SiO3.

The NeutronicsMaterialMaker finds the density automatically

Task 3

CSG geometry

Press down to see the next slide

Making geometry

import openmc

surface_sphere = openmc.Sphere(r=10.0)
region_inside_sphere = -surface_sphere
cell_sphere = openmc.Cell(region=region_inside_sphere) 

cell_sphere.fill = steel

import openmc 

surf_sphere1 = openmc.Sphere(r=10.0)
surf_sphere2 = openmc.Sphere(r=20.0)
between_spheres = +surf_sphere1 & -surf_sphere2
cell_between = openmc.Cell(region= between_spheres) 

cell_between.fill = steel

End of the universe

import openmc

surface_sphere = openmc.Sphere(r=10.0)
region_inside_sphere = -surface_sphere
cell_sphere = openmc.Cell(region=region_inside_sphere)

import openmc 

surf_sphere = openmc.Sphere(r=10.0, boundary_type="vacuum")
between_spheres = -surf_sphere
cell_between = openmc.Cell(region= between_spheres)

The outer most surface must be identified using the boundary_type keyword

Particles crossing this surface are no longer tracked by the code and are terminated.

Making geometry

Making surfaces

surface_xp = openmc.XPlane(x0=0)
surface_yp = openmc.YPlane(y0=0)
surface_zp = openmc.ZPlane(z0=0)
surface_p = openmc.Plane(A=1.0, B=0.0, C=0.0, D=0.0)
surface_xcy = openmc.XCylinder(y0=0.0, z0=0.0, R=1.0)
surface_ycy = openmc.YCylinder(x0=0.0, z0=0.0, R=1.0)
surface_zcy = openmc.ZCylinder(x0=0.0, z0=0.0, R=1.0)
surface_s = openmc.Sphere(x0=0.0, y0=0.0, z0=0.0, R=1.0)
surface_xco = openmc.XCone(x0=0.0, y0=0.0, z0=0.0, R2=1.0)
surface_yco = openmc.YCone(x0=0.0, y0=0.0, z0=0.0, R2=1.0)
surface_zco = openmc.ZCone( x0=0.0, y0=0.0, z0=0.0, R2=1.0)
surface_quad = openmc.Quadric(a=0.0, b=0.0, c=0.0, d=0.0, e=0.0, f=0.0, g=0.0, h=0.0, j=0.0, k=0.0)

OpenMC makes use of constructive solid geometry (CSG)

CAD geometry is also supported via DAGMC

✅

Making geometry

Making volumes

| (union)

import openmc 

surf_sphere = openmc.Sphere(r=20.0)

surf_cy = openmc.ZCylinder(r=5)
surf_up_zp = openmc.ZPlane(z0=30)
surf_low_zp = openmc.ZPlane(z0=-30)

region_sph = -surf_sphere
region_cy = (-surf_cy & -surf_up_zp & +surf_low_zp)
            
cell = openmc.Cell(region = region_sph | region_cy) 

cell_sphere.fill = steel

|

=

Making geometry

Making volumes

~ (complement)

import openmc 

surf_sphere1 = openmc.Sphere(r=20.0)

surf_cy = openmc.ZCylinder(r=5)
surf_up_zp = openmc.ZPlane(z0=30)
surf_low_zp = openmc.ZPlane(z0=-30)

region_sph = -surf_sphere2
region_cy = (-surf_cy & -surf_up_zp & +surf_low_zp)

cell = openmc.Cell(region = region_sph & ~region_cy) 

cell_sphere.fill = steel

~

=

Making geometry

Making volumes

& (intersection)

import openmc 

surf_sphere1 = openmc.Sphere(r=20.0)

surf_cy = openmc.ZCylinder(r=5)
surf_up_zp = openmc.ZPlane(z0=30)
surf_low_zp = openmc.ZPlane(z0=-30)

region_sph = -surf_sphere2
region_cy = (-surf_cy & -surf_up_zp & +surf_low_zp)
            
cell = openmc.Cell(region = region_cy & region_sph) 

cell_sphere.fill = steel

&

=

Building and visualizing CSG geometry

Building and visualizing CSG geometry

Visualizing CSG geometry in 3D

hints

Open file and click 'Apply'

to initiate geometry viewing

hints

Use eyeball icon to toggle between displaying and hiding geometry

hints

Specify geometry to be coloured

with respect to cell ID

hints

Display geometry

surfaces

hints

Perform a threshold to isolate the geometry of interest (paraview automatically shows all geometry cells specified in OpenMC, including vacuum cells)

hints

Change threshold limits to isolate the geometry cells of interest (click 'Apply' to perform the threshold)

hints

Threshold function has removed the vacuum cells, allowing the geometry

to be viewed

hints

Shell can be clipped to view its structure more clearly

hints

Clipping plane can be

changed using these buttons

hints

Apply the clip

hints

View can be rotated

Task 4

Sources

Press down to see the next slide

Making neutron sources

Point source birth locations

import openmc

source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))

Making neutron sources

Point source birth directions

import openmc

source = openmc.Source()
source.angle = openmc.stats.Isotropic()

Making neutron sources
14 MeV monoenergetic

import openmc

source = openmc.Source()

# sets the location of the source to x=0 y=0 z=0
source.space = openmc.stats.Point((0, 0, 0))

# sets the direction to isotropic
source.angle = openmc.stats.Isotropic()

# sets the energy distribution to 100% 14MeV neutrons
source.energy = openmc.stats.Discrete([14e6], [1])

Making neutron sources
Watt distribution

import openmc

source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()

# Documentation on the Watt distribution is here
# https://docs.openmc.org/en/stable/pythonapi/generated/openmc.data.WattEnergy.html
source.energy = openmc.stats.Watt(a=988000.0, b=2.249e-06)

Energy distribution

Deuterium (D)

Tritium (T)

Helium 4

3.5MeV

1/5 of the energy

Neutron

14.1MeV

4/5 of the energy

Helium 5

17.6MeV

Making neutron sources
Muir distribution

import openmc

source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()

# Documentation on the Muir distribution is here
# https://docs.openmc.org/en/stable/pythonapi/generated/openmc.stats.Muir.html
source.energy = openmc.stats.Muir(
  e0=14080000.0,
  m_rat=5.0,
  kt=20000.0
)

MCF and ICF

Making neutron sources

Plasma source birth locations

import openmc

from parametric_plasma_source

import PlasmaSourcemy_plasma = PlasmaSource(
    elongation=1.557,
    ion_density_origin=1.09e20,
    ion_density_peaking_factor=1,
    ion_density_pedestal=1.09e20,
    ion_density_separatrix=3e19,
    ion_temperature_origin=45.9,
    ion_temperature_peaking_factor=8.06,
    ion_temperature_pedestal=6.09,
    ion_temperature_separatrix=0.1,
    major_radius=9.06,
    minor_radius=2.92,
    pedestal_radius=0.8 * 2.92,
    plasma_id=1,
    shafranov_shift=0.44789,
    triangularity=0.270,
    ion_temperature_beta=6
)

Making neutron sources

Plasma source birth directions

import openmc

from parametric_plasma_source

import PlasmaSourcemy_plasma = PlasmaSource(
    elongation=1.557,
    ion_density_origin=1.09e20,
    ion_density_peaking_factor=1,
    ion_density_pedestal=1.09e20,
    ion_density_separatrix=3e19,
    ion_temperature_origin=45.9,
    ion_temperature_peaking_factor=8.06,
    ion_temperature_pedestal=6.09,
    ion_temperature_separatrix=0.1,
    major_radius=9.06,
    minor_radius=2.92,
    pedestal_radius=0.8 * 2.92,
    plasma_id=1,
    shafranov_shift=0.44789,
    triangularity=0.270,
    ion_temperature_beta=6
)

Making neutron sources

Photon source energy spectrum

import openmc

# initialises a new source object
source = openmc.Source()

# sets the location of the source to x=0 y=0 z=0
source.space = openmc.stats.Point((0, 0, 0))

# sets the direction to isotropic
source.angle = openmc.stats.Isotropic()

# sets the energy distribution to 
# 50% 1.1MeV photons and 50% 1.3MeV photons
source.energy = openmc.stats.Discrete(
  [1.1732e6,1.3325e6],
  [0.5, 0.5]
)

# change particle type as the default is neutron
source.particle = 'photon'

hints

Open and threshold geometry

as shown in Task 3

hints

Reduce opacity of threshold

geometry to increase its transparency -

this allows neutron tracks to be seen

more easily

hints

Open .vtp track files

hints

Toggle visibility of each .vtp

file to view neutron tracks

Task 5

CSG cell tally - TBR

Press down to see the next slide

Task 5

CSG tally - TBR

TBR parameter study

Task 6

CSG cell tally - DPA

Press down to see the next slide

Task 6

Displacement

Finding the neutron damage

IAEA Issues Crowdsourcing Challenge for Materials for Fusion Technology

A visualization of the cascade of collisions leading to damage in a crystalline material. (Photo: A. Sand/University of Helsinki)

Stochastic volume calculation

\frac{particles\: in\: cube}{particles\: in\: sphere}=\frac{Volume\: of\: cube}{Volume\: of\: sphere}

\frac{s^3}{\frac{4}{3}\pi r^3}=\frac{8}{\frac{4}{3}\pi} = \frac{6}{\pi}

\frac{particles\: in\: cube}{particles\: in\: shape}=\frac{Volume\: of\: cube}{Volume\: of\: shape}

Task 7

CSG cell & surface tally - spectra

Press down to see the next slide

Neutron interactions

Material 1

Material 2

Neutron birth

(n,n')

(n,f)

(n,n')

(γ,γ')

(n,nγ')

(n,pn')

(n,f)

(n,2n)

(n,α)

(n,γ)

Neutron

Electron

Gamma

Alpha

Proton

Types of scattering

Neutron collides with the nucleus, the neutron scatters of the nucleus losing energy which is given to the recoil nucleus.

The angle and energy of both the neutron and he nucleus can be calculated by relativistic kinematics

Neutron capture by nucleus instantaneously re-emitted with less energy, leaving the nucleus in an excited state.

The nucleus relaxes to ground state by emitting gamma rays.

Angle

At low energies the angular distribution is often isotropic in the center of mass reference frame.
As the energy increases the scattering typically becomes more and more forward peaked.
If the energy goes through a resonance, dramatic swings from forwards scattering to backwards may be seen.

Image from tendl.web.psi.ch

Path length

Path length = 1 / Macroscopic cross section

Shorter path length for lower neutron energy

Path length

A 14MeV neutron will lose energy via scattering interactions.
The average path length between scattering interactions decreases.
Once thermalised the there is equal chance of energy lose or energy gain via collision.
Path length at thermal energy is more constant

Equations

\xi = 1+ \frac{(A-1)^2}{2A} ln \frac{(A-1)}{(A+1)}

The average logarithmic energy decrement (or loss) per collision (ξ) is related to the atomic mass (A) of the nucleus

	Hydrogen	Deuterium	Beryllium	Carbon	Uranium
Mass of nucleus	1	2	9	12	238
Energy decrement	1	0.7261	0.2078	0.1589	0.0084

Collisions

	Hydrogen	Deuterium	Beryllium	Carbon	Uranium
Number of collisions to thermalize	20	25	85	115	2172

n = \frac{1}{\xi} (ln E_0 - ln E)

The average number of collisions requried to reduce the energy of the neutron from Eo to E

*where a thermal neutron energy = 0.025eV

Moderating power

Moderating \; power = \xi \Sigma _s

We should account for the likelihood of scattering.

	Water	Heavy water	Beryllium	Carbon	Polyethylene
Moderating power	1.28	0.18	0.16	0.064	3.26

\Sigma _s = N_D \sigma_s

The number density of the nucleus (ND) and the microscopic cross section (σ) combine to produce the macroscopic scattering cross section (Σ)

Moderating ratio

	Water	Heavy water	Beryllium	Carbon	Polyethylene
Moderating ratio	58	2100	130	200	122

Moderating \; ratio = \xi \frac{\sigma_s}{\sigma_a} = \xi \frac{\Sigma_s}{\Sigma_a}

The probability of being absorbed is also important.

The Moderating ratio is a relative measure of a moderator to scatter neutrons without absorbing them.

Finding the neutron spectra

Neutron flux spectrum - Cell

Finding the neutron

Neutron current spectra - Surfaces

Finding the photon spectra

Photon flux spectrum - Cell

Task 8

CSG mesh tally

Press down to see the next slide

CSG regular mesh tally

# only 1 cell in the Y dimension
mesh.dimension = [4, 1, 4]

# physical limits (corners) of the mesh
mesh.lower_left = [-200, -1, -200]
mesh.upper_right = [200, 1, 200]

# only 1 cell in the Y dimension
mesh.dimension = [4, 4, 4]

# physical limits (corners) of the mesh
mesh.lower_left = [-200, -200, -200]
mesh.upper_right = [200, 200, 200]

2D mesh tally

Flux

Absorption

(n,2n)

Geometry

3D mesh tally

Click 'Apply' to initiate

tally viewing

hints

Specify colouring with

respect to tally value

hints

Display tally surfaces

hints

Rotate to view

tally more clearly

hints

Perform a threshold to

isolate the mesh tally

hints

Change threshold limits to

isolate tally range of interest

hints

Apply the threshold to view

tally results

hints

Task 9

CSG surface tally - dose

Press down to see the next slide

Dose

Effective dose coefficients

Dose

CSG cask geometry

Dose

CSG nested sphere cask geometry

Dose

CSG nested sphere cask dose

Task 10

CAD geometry

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Making CAD shapes

3D CAD geometry can be made by linking coordinates together with edges.

Edges can be straight, lines, splines, circles etc.

The resulting face can then be rotated, extrude of swept to make a volume.

RotateStraightShape

RotateSplineShape

ExtrudeSplineShape

ExtrudeStraightShape

The volume can then undergo Boolean opperations such as cut, intersected or unioned

Making CAD components

Finding the coordinates can be performed automatically by using a component.

Component construction is driven by parameters.

Boolean operations are also available.

Making CAD reactors

Components and Shapes can be assembled into reactors.

The construction is parameter driven.

Shapes / components

Plasma

TF coils

PF coils

Blanket

Reactor object

my_reactor = paramak.Reactor([pf_1, tf, blanket, plasma])

Ball reactor

Task 13

Parametric study sampling

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Techniques for sampling design space - Random

Techniques for sampling design space - Grid

Techniques for sampling design space - Halton

Techniques for sampling design space - Adaptive

Task 14

Parametric study optimisation

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Optimize a breeder

blanket for tritium production

Task currently under development

Task 15

Activation

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Task 15

Activation

Target

Nuclide

n,2n

n,g

n,p

n,pn

n,d

n,t

n,nd

n,a

n,He3

n,pd

n = neutron

g = gamma

t = tritium

p = proton

d = deuterium

He = helium

Common neutron induced reactions

Neutron number

Proton number

Task 15

Activation

Fe56

Neutron induced reactions in Fe56

Neutron number

Proton number

Fe57

Fe58

Fe59

Fe55

Fe54

Co59

Mn53

Mn54

Mn55

Mn56

Mn57

Mn58

Cr57

Cr56

Cr55

Cr54

Cr53

Cr52

Co58

Co57

Co56

Co60

Decay

Primary activation

Secondary activation

Co55

Task 15

Activation

Task 15

Irradiation Activation

Time

Number of atoms

Shut down

New isotopes build up during irradiation
Radioactive isotopes decay and will eventually reach a point where decay rate is equal to activation rate.
Decay is more noticeable once the plasma is shutdown.
The activity is related to the irradiation time and the nuclide half life.
The decay process emits gamma rays leading to dose.
The dose is dependent on the activity, the energy of the gamma and biological response.

Fusion Energy Neutronics Workshop

Cerberus

Nuclear

How can neutronics help

Topics Covered

How to get setup

What are the tasks like

OpenMC popularity

Introduction to the workshop

Task 1

Cross Sections

Microscopic cross section

Experimental data

Interaction cross sections

Cross section to reactions

Task 1 - Cross section plotting Isotopes

Task 1 - Cross section plotting Isotopes

Task 1 - Cross section plotting Isotopes

Task 1 - Cross section plotting Elements

Task 1 - Cross section plotting Elements

Task 1 - Cross section plotting Materials

Task 1 - Cross section plotting Materials

Task 1 - Cross section plotting Doppler broadening

Task 1 - Cross section plotting Doppler Broadening

Task 2

Materials

Making materials

Making materials

Enriching materials

Enriching materials

Making fusion relevant materials

Making fusion relevant materials

Density with temperature

Density with pressure

Density with Li6 enrichment

Task 3

CSG geometry

Making geometry

End of the universe

Making geometry

Making surfaces

Making geometry

Making volumes

| (union)

|

=

Making geometry

Making volumes

~ (complement)

~

=

Making geometry

Making volumes

& (intersection)

&

=

Building and visualizing CSG geometry

Building and visualizing CSG geometry

Visualizing CSG geometry in 3D

hints

hints

hints

hints

hints

hints

hints

hints

hints

hints

hints

Task 4

Sources

Making neutron sources

Point source birth locations

Making neutron sources

Point source birth directions

Making neutron sources 14 MeV monoenergetic

Making neutron sources Watt distribution

Energy distribution

Making neutron sources Muir distribution

Task 1 - Cross section plotting
Isotopes

Task 1 - Cross section plotting
Isotopes

Task 1 - Cross section plotting
Isotopes

Task 1 - Cross section plotting
Elements

Task 1 - Cross section plotting
Elements

Task 1 - Cross section plotting
Materials

Task 1 - Cross section plotting
Materials

Task 1 - Cross section plotting
Doppler broadening

Task 1 - Cross section plotting
Doppler Broadening

Making neutron sources
14 MeV monoenergetic

Making neutron sources
Watt distribution

Making neutron sources
Muir distribution