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Foundation Class > Fission > Flashcards

Fission Flashcards

1
Q

Describe Elastic Scattering

A

Interaction in which momentum and kinetic energy are conserved.
ie: total kinetic energy (and momentum) is transferred from neutron before the collision to the nucleus/neutron after the collision.
Nucleus is NOT excited.

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2
Q

Describe Inelastic Collision.

A

Collision where some or all Kinetic energy is transferred to some amount of excitation energy of the nucleus it collided with.
Kinetic energy is NOT conserved.
Momentum IS conserved.

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3
Q

Describe Radioactive Capture.

A

A neutron is absorbed (captured) by a nucleus which then becomes excited.
Excited nucleus then emits a gamma to decay back to ground state.

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4
Q

Describe Particle Ejection.

A

A neutron is absorbed by a nucleus which then becomes excited.
The excited nucleus then emits a particle (alpha, proton etc) to decay.

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5
Q

Describe Fission.

A

Neutron is absorbed by a nucleus which then becomes excited.
The excited nucleus then splits (fissions) into 2 smaller nuclides and also typically releases 2-3 neutrons.

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6
Q

Define Excitation Energy

A

Measure of how far the energy level of a nucleus is above ground state.

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7
Q

Define Critical Energy.

A

Amount of Excitation Energy required for a particular nuclide to Fission.

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8
Q

Explain Fission Process

A

Nuclide nucleus is like a drop of water. As a neutron is absorbed by the drop of water, the bead (drop) distorts and is unstable. 2 halves of the drop start to form and if the excitation is high enough, the drop will separate (split in half). During this process 2 smaller drops are left with more neutrons emitted as well.

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9
Q

Define Fissile Material

A

Material that will Fission when it absorbs any neutron (regardless of energy of the neutron)

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10
Q

Define Fissionable Material

A

Material that can Fission if it absorbs a neutron with enough energy added to it above its normal binding energy.
May not fission if the KE of the neutron is not high enough.

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11
Q

Define Fertile Material

A

Fissionable Material than can undergo transmutation to become Fissile Material.
Fissionable Material that thru absorption of a neutron and a series of decay chain can result in a Fissile material.

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12
Q

Define Thermal Neutrons

A

Very low Kinetic Energy. Said to be of same KE of it’s surroundings (thermal energy).
KE of neutron also changes as thermal energy of surroundings (ie. Temperature) changes.

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13
Q

What is Binding Energy per Nucleon?

A

BE/A
Average energy required to remove a single nucleon from a specific nucleus.

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14
Q

Describe the BE/A Curve

A

Binding Energy per Nucleon curve.
BE/A increases (initially rapidly) in nuclides from mass number 1 to 60 (which is the peak).
After that the BE/A decreases from mass number 60 to >240.
Nuclides w/higher mass number have more neutrons/proton take less energy to remove nucleons.

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15
Q

What is the average amount of energy released from fission (Instantaneous and Delayed)?

A

200 MeV

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16
Q

What is the average energy released Instantaneously (immediately) from fission.

A

187 MeV total

165 MeV KE of Fission Fragments
5 MeV KE of Fission Neutrons
7 MeV Instantaneous Gamma Rays
10 MeV Capture Gamma Rays

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17
Q

What is the average amount of energy released as Delayed Energy from Fission?

A

13 MeV total

7 MeV KE of Beta Particles
6 MeV Decay Gamma Rays

(Neutrinos produce 10 MeV, however it rant quantified because neutrinos don’t interact with anything).

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18
Q

Which fission product nuclides are most likely to result from fission?

A

Typically mass numbers around 95 and 140.

Rb-93 and Cs-140 are very likely.
Also Sr-94 and Xe-140.

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19
Q

Describe fission energy released from Change in Binding Energy.

A

Binding energy per nucleon raises as the mass number lowers (reference BE/A curve). BE/A is higher in the fission products than the original nuclide.
Ex: U-235 BE is 1,786 MeV and the FPs Cs-140 and Rb-93 are 1,176 MeV/809 MeV respectively for a total of 1,985 MeV. Difference of 199 MeV.

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20
Q

Describe the fission energy released using conservation of mass energy.

A

Calculate mass difference x Energy per amu. Gives energy released.

Energy released=m(reactants)-m(products)•931.5 MeV/amu

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21
Q

How to estimate Decay Energy from fission

A

Given: Rb-93->(B-) Sr-93->(B-) Zr-93->(B-) Nb-93

E(decay)=[m(Rb-93)-(m(Nb-93)+4m(electron))]•931.5 MeV/amu

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22
Q

How is heat produced from fission?

A

Ionization and scattering transforms KE and Electrostatic Force into Thermal (Heat) energy.

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23
Q

Purpose and importance of Source Neutrons

A

Ensure neutron population during shutdown is high enough to allow visible indication of SR NI’s.

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24
Q

What are Intrinsic Neutron Sources?

A

Nuclei that yield neutron producing reactions and that occur in reactor core fuel related materials.

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25
Q

What are the Intrinsic Neutron Sources

A

Spontaneous Fission
Photo-neutron reactions
Alpha-neutron reactions

26
Q

How are Spontaneous Fission Neutrons produced?

A

Some heavy nuclei May fission spontaneously and emit neutrons.
U-235, U-238, Cm-242 and Cm-244 contribute most to this.
Cm=Curium.
Significant contributor prior to startup (BOL).

27
Q

Describe how Photo-neutrons are produced.

A

Significant after more power history (EOL)
Gamma (photon) interacts with deuterium and yields hydrogen and a neutron.
The gammas are high energy and come from fission -> more at EOL.
Decreases quickly over time.

28
Q

Describe how Alpha-Neutrons are produced.

A

Alphas come from decay of heavy elements in fuel.
These interact with O-18 and B-11 and produce Neutrons.
Alphas also produced from Transuranic elements over time during power operations.
Transuranic alphas become the most significant by EOL.

29
Q

What are Transuranic Elements?

A

Elements beyond Uranium. (Atomic number>92).

30
Q

Order of Intrinsic Neutron sources by strength at BOL immediately after Rx S/D

A

1: Photo-neutrons
2: Spontaneous Fission
3: Alpha-neutrons (Transuranic)

31
Q

Order of Intrinsic Neutron sources by strength at BOL several weeks after Rx S/D

A

1: Spontaneous Fission
2: Photo-neutrons
3: Alpha-neutrons (Transuranic)

32
Q

Order of Intrinsic Neutron sources at EOL immediately after Rx S/D

A

1: Photo-neutrons
2: Alpha-neutrons (Transuranic)
3: spontaneous neutrons

33
Q

Order of Intrinsic Neutron sources at EOL several weeks after Rx S/D

A

1: Alpha-neutrons (transuranic)
2: Spontaneous neutrons
3: Photo-neutrons

34
Q

Purpose of Installed Neutron Sources.

A

Ensure reliable and sufficient number of source neutrons because intrinsic neutrons may not be enough.
Ensures detectable neutron levels at all times.
Ensure Rx can be started up.

35
Q

4 types of Installed Neutron Sources.

A

1: Californium-252
Beryllium Sources
2: Alpha-neutron Beryllium Sources
3: Photo-neutron Beryllium Sources
4: Antimony-Beryllium Sources

36
Q

Define Atomic Density

A

Number of atoms per volume
N=(density•Avocados #)/M

N=atomic density
M=gram atomic weight

37
Q

Define Neutron Flux

A

How many neutrons travel thru a material and the distance travelled per second.
Also number of neutrons passing thru the unit area per unit time.
Also total path length covered by all neutrons in 1 cm^3 per second.
Neutron flux=nv
n=neutron density
v=neutron velocity

38
Q

Define Thermal/Fast Neutron Flux

A

Same as Neutron Flux but specific to Thermal/Fast Neutrons.

39
Q

Define Cross Section

A

Probability of a neutron interacting with a nucleus for a particular reaction.
Depends on energy of the neutron and the particular nucleus involved.
Also varies with the type of reaction.

40
Q

Define Microscopic Cross Section

A

Probability a neutron will react with a nucleus.
Absorption (fission or capture) or scattering (elastic or Inelastic).

41
Q

Define Barns

A

Buildings for farm animals.

42
Q

Define Barns with respect to Cross Sections you dummy!

A

Unit of microscopic cross section.
1 barn=1x10^-24 cm^2

(Probability, not area)

Different nuclei have different microscopic cross sections-doesn’t depend on size of nucleus.

43
Q

Define Macroscopic Cross Section (Sigma)

A

Probability that neutrons will interact with a certain material per unit volume.
Takes into account atomic density, and microscopic cross section
Sigma=N•microscopic cross section
N=atomic density

44
Q

Difference between microscopic and macroscopic cross sections.

A

Microscopic is effective area a single nucleus presents to a bombarding particle.
Macroscopic is total effective target area of the nuclei in a cubic cm of the material.

45
Q

Define Mean Free Path.

A

Path taken by the righteous.
The path of kindness.

46
Q

Define Mean Free path with regards to nuclear fission.

A

Lambda
Probability of neutron interaction in 1 cm of travel in a material.
Inverse of Macroscopic Cross Section
Lambda=1/Sigma.

47
Q

Define Fast Neutrons

A

Fission neutrons are born fast.
Greater than 0.1 MeV
(Chart of Nuclides indicates >1MeV, contradicts lesson plan)

48
Q

Define Intermediate Neutron

A

Neutrons with energy levels between 1eV and 0.1MeV (or 1MeV, depending on the source).

49
Q

Define Slow Neutrons

A

Neutrons with energy levels <1eV

50
Q

Define Thermal Neutrons

A

Energy levels of 0.025eV at 68C.
Note: These are also Slow Neutrons.

51
Q

Describe neutron absorption cross section at low (slow, 1/v region) energies

A

Cross section is inversely proportional to neutron velocity. ie: as velocity lowers, absorption probability rises.
Absorption probability in this region is relatively high.

52
Q

Describe absorption cross section in the Intermediate (epithermal) region.

A

Resonance peaks cause high probability of absorption because at these peaks the sum of KE and BE of the neutron is equal to the amount of energy required to raise a nucleus to a higher quantum energy state.

53
Q

Describe absorption cross section for the Fast region.

A

Absorption probability is lower at the start of this region and steadily decreases as neutron energy increases.

54
Q

How does temperature affect Macroscopic Cross section and Mean Free Path?

A

As temperature rises->density of material lowers causing atoms to spread further apart. Therefore less interactions per unit volume. As a result of this:
1: Macroscopic cross sections LOWER
2: Mean Free Path RISES

55
Q

Units of Neutron Flux

A

n/(sq cm•sec)

56
Q

What is Axial Flux?

A

Flux distribution from top to bottom of the core.

57
Q

What is Radial Flux?

A

Flux distribution from side to side of the core (or from the centerline to the outside edge, radius).

58
Q

What is Self Shielding?

A

Local neutron flux is depressed (shielded) within a material due to neutron absorption near the surface of the material. Ex: inside of a fuel pellet is shielded by the outside of the fuel pellet because neutrons don’t make it all the way into the middle due to being absorbed by the outer layer.

59
Q

What is Reaction Rate?

A

Number of interactions taking place in a cubic centimeter in one second.
Found by multiplying total path length of all neutrons in a cm^3 (n flux) by the probability of interaction in that volume (Sigma, Macroscopic).
R=sigma•neutron flux

60
Q

Describe Reaction Rate over core life.

A

As fuel burns up, reaction rate lowers due to macroscopic cross section lowering. Therefore to maintain the same reaction rate, neutron flux must be raised to compensate (this is why we continuously dilute over core life).

61
Q

How is Reaction Rate related to Reactor Power?

A

Power equals reaction rate x core volume.
P=[n(flux)•sigma(fission)•V]/3.12x10^10 (fissions/watt•sec)
V=core volume

Foundation Class (8 decks)

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