Neutrons Flashcards

1
Q

Define Nuclear stability (include relationship to atomic number and line of stability)

A

Defined as the inherent ability of an atom to resist changing its atomic structure or energy. Line of stability is centered on the ratio of 1 neutron per proton. No known stable atom above an atomic number of 83.

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2
Q

Define Mass-energy equivalence

A

E = MC Squared.. Approximately 931.5 MeV/AMU

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3
Q

Define Mass Defect

A

When a nucleus is made up from its component parts (protons and neutrons), the mass of the nucleus is less than the mass of the individual protons and neutrons.

This mass difference is called the mass defect.

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4
Q

Define Binding Energy

A

It is the energy equivalent of the mass defect.

Binding energy represents that amount of energy that is released when an atom is formed from its component protons, neutrons, and electrons

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5
Q

Define Binding Energy per Nucleon

A

The average energy required to remove a nucleon from a given nucleus.

Found by dividing the total binding energy by the number of nucleons.

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6
Q

Given a plot of binding energy per nucleon, explain the changing slope of the curve.

A

The larger the mass number the greater number of protons and increasing electrostatic repulsion forces.

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7
Q

Define microscopic cross section.

A

σ is the probability that a given interaction will occur between a target nucleus and an incident neutron.

It is based on the makeup of the nucleus and, hence, on its “ability” to absorb the neutron, not on the “size” of the nucleus itself.

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8
Q

List each component of the total microscopic cross section.

A

For any given target (isotope):

  • Radiative capture
  • Fission
  • Elastic scattering
  • Inelastic scattering.

Classified as either absorption reactions (radiative capture/fission) or scattering reactions.

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9
Q

Describe radiative capture.

A

A neutron is absorbed by the target nucleus, resulting in an excited compound nucleus. The compound nucleus returns to ground state by emitting gamma rays

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10
Q

Describe fission.

A

Similar to radiative capture except that sufficient energy has been added to the target so that the target nucleus splits apart.

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11
Q

Describe elastic scattering.

A

Occurs when a nucleus deflects a neutron without absorbing the neutron.

Elastic scattering conserves kinetic energy and is often visualized as a “Billiard Ball” type of collision

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12
Q

Describe inelastic scattering.

A

Similar to elastic scattering, except that kinetic energy is not conserved.

The neutron is, for a short time, incorporated into the nucleus. The kinetic energy transferred from the neutron to the nucleus raises internal energy of target nucleus.

A neutron is subsequently released, with less kinetic energy than the incoming neutron had.

The target nucleus then gives up the energy it
received from the incident neutron and returns to ground state by emitting a gamma ray.

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13
Q

Define macroscopic cross section.

A

Σ is defined as the probability of an incident neutron interacting with a target nucleus per unit length of travel of the incident neutron. Σ=Nσ.

Where:

Σ = macroscopic cross section (cm-1)
N = atomic density (atoms/cm<sup><span style="font-size: 13.5px; line-height: 0px;">3</span></sup>)
σ = microscopic cross section (barns)
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14
Q

Explain how changing neutron energy will affect the magnitude of a cross section for a given isotope.

A

Increasing neutron energy results in a smaller magnitude of cross section.

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15
Q

List each component of the total macroscopic cross section.

A

It is the product of the atomic density and the microscopic cross section.

Σ=Nσ.

Where:
Σ = macroscopic cross section (cm-1)
N = atomic density (atoms/cm3)
σ = microscopic cross section (barns)

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16
Q

Define a fast neutron.

A

All fission neutrons are born as fast neutrons. They have a kinetic energy greater than 0.1 MeV (105 eV - 100,000 eV)

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17
Q

Define Critical Energy.

A

The minimum amount of energy required for fission to occur in a specific fuel type.

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18
Q

Define “Instantaneous” as it pertains to particle production from fission.

A

The immediate products of fission vice the delayed products.

83% of energy from kinetic energy of initial fission fragments and 6% from initial neutrons and gamma rays.

19
Q

Define “delayed” as pertaining to particle production from fission.

A

Accounts for 7% of energy from fission and results from the decay of the initial fission product fragments.

20
Q

Define fissile material.

A

A fuel type that will fission simply because of the binding energy of an incident neutron.

21
Q

Define fissionable material.

A

A fuel type that requires kinetic energy in addition to binding energy of an incident neutron for fission to occur.

22
Q

Given a graph of fission product yields, explain the shape of the distribution curve.

A

Maywest Curve

The curve shows that one of these fission fragments is lighter (A»95) while the other is heavier (A»139).

Splitting into two identical masses (A=118) is a low probability event.

23
Q

Give a typical value for the energy in MeV (mega electron volts) released from an average fission event.

A

Approximately 200 MeV.

24
Q

Given the origin, energy level, and production time of a neutron; classify the neutron.

A

Origin based:

  • Fission Neutrons or Source Neutrons

Energy Level:

  • Fast Neutrons are > 0.1 MeV,
  • Intermediate (epithermal) are 0.1 MeV - 1 eV
  • Slow Neutrons are <1 eV.

Production Time Classification

  • Prompt Neutrons are born less than 10-14 seconds after fission
  • Delayed Neutrons are born greater than 10-14 seconds after fission.
25
Q

Specify three energy ranges of neutron flux in a Pressurized Water Reactor core.

A

Slow/thermal ———- (< 1 eV)

Intermediate/epithermal ———– (1 eV to 105 eV)

Fast ————– (>105 eV)

1,000,000 (106) ev = 1 MeV

26
Q

Define a Prompt Neutron.

A

A fission neutron emitted within 10-14 seconds of a fission event

27
Q

Define a Delayed Neutron.

A

A neutron born more than 10-14 seconds after a fission event.

28
Q

Define Slowing Down Length?

A

The straight-line distance between the point of birth of the neutron and the point of thermalization is called the slowing down length.

29
Q

Define Thermal Diffusion Length.

A

It is the straight-line distance between the point of thermalization and the point of absorption.

30
Q

Define neutron generation time (prompt, delayed).

A

The total time, from fission to neutron absorption is referred to as the neutron generation time (the lifetime of one generation).

Prompt is 10-4 sec. Delayed is 12.5 sec.

31
Q

Explain the difference between prompt and delayed neutron generation times.

A

Prompt neutron generation time (l*) is the average length of time from fission to the absorption of resultant prompt neutrons. Approximately 10-4 seconds.

Delayed neutron generation time (ld ) is defined as the average length of time from fission and release of a neutron from a delayed neutron precursor to the absorption of resultant delayed neutrons.

Approximately 12.5 seconds.

32
Q

Explain why thermal neutrons are required for Pressurized Water Reactor operation.

A

The absorption cross section for thermal neutrons is much greater than that of intermediate or fast neutrons.

33
Q

Describe the mechanics of a neutron slowing down and the diffusion process.

A

A neutron will lose more energy in a collision with an atom of nearly its own mass than in a collision with an atom with a mass much greater than the neutron.

34
Q

Define the term Moderating Ratio.

A

It is determined by the microscopic cross section properties of the moderator and neutron energy levels. The density of a material does not alter the MR.

MR = (logarithmic enegy decrement (squigly) times the microscopic cross section for scattering) divided by the microscopic cross section for absorption.

MR=ςσsa

35
Q

Define the term Logarithmic energy decrement.

A

It is the average neutron energy loss per collision in a moderator.

36
Q

Describe the properties of an ideal moderator.

A

In addition to a high logarithmic energy decrement, other factors that affect the usefulness of a moderator are a high cross section for scattering, a low cross section for absorption, and a high atomic density.

37
Q

Explain how the density of a medium affects neutron slowing down length and thermal diffusion length.

A

As temperature of the water in the core goes up, the water molecules gain energy and spread out and the atomic density (N) of the water goes down. With fewer hydrogen molecules to interact with, neutrons will travel farther in the process of slowing down.

The microscopic cross section stays constant thus, the macroscopic cross section goes down.

38
Q

Define neutron flux (fast, thermal).

A

is the number of neutrons passing through a unit area per unit time. Neutron flux is designated by the Greek lower case phi.

39
Q

Define axial and radial flux distribution.

A

The magnitude of the axial flux is what would be seen clearly from a side view of the core

magnitude of the radial flux would be seen graphically from a top view.

40
Q

Define Reaction Rate.

A

It is how many reactions are occurring in a unit volume per unit time.

Dependent on the probability of the reaction happening, the amount of material to have the reaction with, and the number of neutrons available to cause the reaction.

The reaction rate is given the symbol R, and is expressed in units of reaction/cm3 second.

41
Q

Define power density.

A

It is the power generation per unit volume in the reactor core.

42
Q

Define Reactor Power.

A

It is the rate at which energy is emitted as a result of nuclear fissions.

It combines fission reaction rate and power density terms.

Note: Flux is the only thing the reactor operator has direct control over in the core.

43
Q
A