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NE 301 - Introduction to Nuclear Science Spring 2012. Classroom Session 8: Radiation Interaction with Matter Non-Charged Radiation Mass Attenuation Tables and Use Absorbed Dose (D), Kerma (K) Gray ( Gy ) = 100 rad Dose Calculations Analysis of Gamma Information (NAA)

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ne 301 introduction to nuclear science spring 2012

NE 301 - Introduction to Nuclear ScienceSpring 2012

Classroom Session 8:

Radiation Interaction with Matter

Non-Charged Radiation

Mass Attenuation Tables and Use

Absorbed Dose (D), Kerma (K)

Gray (Gy) = 100 rad

Dose Calculations

Analysis of Gamma Information (NAA)

Chemical Effects of Nuclear Reactions

reminder
Reminder
  • Load TurningPoint
    • Reset slides
    • Load List
    • Homework #2 due February 9
    • Next Tuesday February 14 – 1st Demo Session
      • MCA
      • Gamma Spectroscopy identification of isotopes
      • NAA of samples
ionizing radiation electromagnetic spectrum
Ionizing Radiation: Electromagnetic Spectrum

Ionizing Radiation

Each radiation have a characteristic , i.e.:

  • Infrared: Chemical bond vibrations (Raman, IR spectroscopy)
  • Visible: external electron orbitals, plasmas, surface interactions
  • UV: chemical bonds, fluorecense, organic compounds (conjugated bonds)
  • X-rays: internal electron transitions (K-shell)
  • Gamma-rays: nuclear transitions
  • Neutrons (@ mK, can be used to test metal lattices for example)

Ionizing

radiation interaction with matter
Radiation Interaction with Matter
  • Five Basic Ways:
    • Ionization
    • Kinetic energy transfer
    • Molecular and atomic excitation
    • Nuclear reactions
    • Radiative processes
radiation from decay processes
Radiation from Decay Processes
  • Charged
    • Directly ionizing (interaction with e-’s)
      • β’s, α’s, p+’s, fission fragments, etc.
    • Coulomb interaction – short range of travel
    • Fast moving charged particles
    • It can be completely stopped
  • Uncharged
    • Indirectly ionizing (low prob. of interaction – more penetrating)
    • , X-Rays, UV, neutrons
    • No coulomb interaction – long range of travel
    • Exponential shielding, it cannot be completely stopped

R

neutral interactions
Neutral Interactions
  • Stochastic (Probabilistic)
  • With an electron or a nucleus
  • Can be scattering – elastic or inelastic
  • Can be absorptive
  • It is still a collision:
    • Flux of particles is important
flux or intensity
Flux or Intensity
  • Flux is usually for neutrons (n)
  • Intensity is usually for photons (’s)

Target

Beam

Density of particles in the beam

Velocity of beam particles

attenuation of uncollided radiation
Attenuation of Uncollided Radiation
  • How do we calculate the change in the flux of (uncollided) particles as it moves through the slab?

Uncollided radiation is a simplification. In reality not every collided photon/neutron is lost and there are buildup factors (Bi)

attenuation of uncollided radiation1
Attenuation of Uncollided Radiation

Beam with intensity I, interacting with shield (1-D)

microscopic and macroscopic cross sections
Microscopic and Macroscopic Cross Sections
  • Sigma-N =
    • Linear Attenuation Coefficient or Macroscopic Cross Section ( or )
  • Notice Different Units:
  •  is measured in cm-1
  •  is measured in barns
    • 1 barn = 10-24 cm2

Constant of Proportionality or Microscopic Cross-Section

slide11

A beam of neutrons is normally incident on a slab 20 cm thick. The intensity of neutrons transmitted through the slab without interactions is found to be 13% of the incident intensity. What is the total interaction coefficient t for the slab material?

  • 0.01 cm-1
  • 0.1 cm-1
  • 1 cm-1
  • 10 cm-1
attenuation of uncollided radiation2
Attenuation of Uncollided Radiation

Beams of particles: with intensity I0, interacting with shield (1-D)

Point sources: Isotropic source emitting Sp particles per unit time

related concepts
Related Concepts
  • Mean Free Path (mfp or ):
    • Average distance a particle travels before an interaction
  • Half-thickness (x1/2) of the slab?
    • Thickness of slab that will decrease uncollided flux by half

Similar concepts to mean-life and half-life

slide15

It is found that 35% of a beam of neutrons undergo collisions as they travel across a 50 cm slab. What is the mfp and x1/2 for the slab?

  • 10 and 6.9 cm
  • 20 and 13.8 cm
  • 116 and 80 cm
  • 1000 and 693 cm
photon interactions tables
Photon Interactions -  tables
  • Photon energies:
    • 10 eV < E < 20 MeV
      • IMPORTANT radiation shielding design
  • For this energy range,

1. Photoelectric Effect

2. Pair Production

3. Compton Scattering

example photon interactions for pb
Example: Photon Interactions for Pb

High

Low

Intermediate

Energy

Compton Scattering

Pair Production

Photoelectric Effect

problem with photons
Problem with Photons
  • 100 mCi source of 38Cl is placed at the center of a tank of water 50 cm in diameter
    • What is the uncollided -flux at the surface of the tank?
problem with photons1
Problem with Photons
  • 100 mCi 38Cl, water tank 50 cm dia.
    • What is the uncollided -flux at the surface of the tank?
linear coefficients macroscopic cross sections
Linear Coefficients – Macroscopic Cross Sections
  • Linear Absorption Coefficient
    • μt
  • Linear Scattering Coefficient
    • μs
  • Macroscopic Fission Cross-section
    • Σf, μf for neutrons
for homogeneous mixes of any type
For homogeneous mixes of any type
  • Valid for any cross section type (fission, total, etc)
  • Valid for chemical compounds as well

DO NOT add microscopic cross-sections

slide29

In natural uranium (=19.21 g/cm3), 0.720% of the atoms are 235U, 0.0055% are 234U, and the remainder 238U. From the data in Table C.1.

What is the total linear interaction coefficient (macroscopic cross section) for a thermal neutron in natural uranium?

0.24 cm-1

0.0003 cm-1

238U: 0.59 cm-1

Who dominates?

absorbed dose d gray rad
Absorbed Dose, D (Gray, rad)

Energy absorbed per kilogram of matter (J/kg)

Gray: 1 Gy = 1 J/kg

The traditional unit:

Rad: 100 rad = 1 Gy

rad = Radiation Absorbed Man

Dose rate = dose/time

kerma approx dose for neutrons
Kerma (Approx. dose for neutrons)

Kerma

Kinetic Energy of Radiation absorbed per unit MAss

For uncharged radiation

Kerma is easier to calculate than dose for neutrons

Kerma and Dose: same for low energy

Kermaover-estimates dose at high energy

No account for “Bremsstrahlung” radiation loses.

calculating dose rate and kerma rate
Calculating Dose Rate and Kerma Rate

en(E)/ =mass interaction coefficient (table C3)

E = particle energy [MeV]

 = flux [particles/cm2 s]

Notice Difference

tr(E)/ =mass interaction coefficient (table C3)

E = particle energy [MeV]

 = flux [particles/cm2 s]

Engineering Equations – PLEASE Watch out for units!