1. Chapter 3. Basic Instrumentation for Nuclear Technology
Outline of experiment:
􀂄 get particles (e.g. protons, …)
􀂄 accelerate them
􀂄 throw them against each other
􀂄 observe and record what happens
􀂄 analyse and interpret the data
Accelerators
Detectors
Reactors

2. 1.Accelerators
History-Why
Particle Sources
Acceleration stage
Space charge
Diagnostics
Application

3. 2. Detectors Gas-Filled Radiation Detectors Scintillation Detectors
ionization chambers proportional counters Geiger-Muller counters
Gas-Filled Radiation Detectors
Scintillation Detectors
Semiconductor Detectors
Personal Dosimeters
Others
Particle identification
Measurement theory
Detection Equipment
Photomultiplier tube
photographic films photographic emulsion plates
Cloud and Bubble Chambers
E-ΔE, TOF

4. 3. Reactors Reactions Involving Neutrons
Thermal-Neutron Properties of Fuels
General features
The Neutron Life Cycle in a Thermal Reactor
Homogeneous and Heterogeneous Cores
Reflectors
Reactor Kinetics
Reactivity Effects

5. n + 235U  X + Y+ E MeV
“The energy produced by the breaking down of the atom is a very poor kind of thing. Anyone who expects a source of power from the transformations of these atoms is talking moonshine.” Lord Ernest Rutherford, 1933.

6. Self-sustaining Chain reaction
Dec. 2, 1942, Fermi achieved sustained chain reaction, and the first fission reactor provided data for future design of nuclear reactors.

7. The Vision
“It is not too much to expect that our children will enjoy in their homes [nuclear generated] electrical energy too cheap to meter.”
– Lewis Strauss, Chairman of the U.S.
Atomic Energy Commission (1954)

8. The total and fission cross section for 235U based on NJOY-processed ENDF/B (version V) data.

9. The fast fission cross section for three fissionable uranium isotopes based on NJOY processed ENDF/B (version V) data

10. Reactions Involving Neutrons

12. Neutron Scattering
The elastic scattering is the main mechanism in moderating neutrons in thermal nuclear reactors.

13. The corresponding neutron energy loss
Average Logarithmic Energy Loss
on a logarithmic energy scale a neutron loses the same amount of logarithmic energy per
elastic scatter, regardless of its initial energy
average number of scatters required to bring a neutron of initial energy E1to a lower energy E2

14. Slowing of neutron (moderation) by various materials.
Here n is the number of elastic scatters to slow, on the average, a neutron from 2 MeV to eV
Is H2O a good moderator (慢化剂)

15. Thermal Neutrons Cross Sections
Thermal neutron capture cross sections (c) Thermal neutron cross section for fission (f)
1H 2H 12C N O 113Cd
 c /b ,820
Moderators: H2O vs. D2O vs. C
Fermi’s used Cd for emergency
Nuclear Fission

16. Why neutron moderation is needed?
Fission neutron energy spectrum
The average energy of prompt fission neutrons is about 2 MeV

17. Neutron Capture Reactions
Neutron leakage
Safety consideration

19. Spontaneous fission:
The fission produced in these cases is insignificant for energy production However the phenomenon is important since represents an uncontrollable source of neutrons in a reactor and it is, furthermore, possible to make use of it in the start-up stage. An example of the use of this fission is the neutron source of 252 californium.

20. Induced fission:
Certain heavy nuclei can be induced to fission, as result of one neutron capture. Consequently, several high-energy neutrons are produced, which permit to maintain the chain reaction process. The nuclei 235U,233U, 239Pu and 241Pu experience fission with low-energy thermal neutrons and they are called fissile materials.
The nuclei 238U and 232Th fission with fast neutrons. The radiative capture of neutrons by 238U and 232Th leads to the formation of the fissionable materials 239Pu and 233U, so they are called fertile materials.

22. Thermal-Neutron Properties of Fuels
σf and σγ are the cross-sections for fission and capture
v is the average number of emitted neutrons per nuclear fission
The number of neutrons emitted when one neutron is absorbed in the nucleus expressed as η

23. 1. 233U has the largest value of η, the number of fission neutrons produced per thermal neutron absorbed, and hence is the best prospect for a thermal breeder reactor (增值反应堆）. A breeder reactor needs an η of at least two since one neutron is needed to sustain the chain reaction and one neutron must be absorbed in the fertile material （增值材料） to breed a new fissile fuel atom. Fertile materials are those such as 232Th and 238U that, upon thermal neutron absorption, may yield fissile materials

24. 2. Although the plutonium isotopes produce almost 3 fission neutrons per thermal fission, their relatively high radiative capture (n,γ) cross sections result in low values of η.
E> ~100 keV, 239Pu and 241Pu, η >3. Thus fast reactors using plutonium as fuel are attractive as breeder reactors.

25. 3. The fertile isotopes 232Th and 238U have absorption cross sections of about 1% or less than those of their conversion fissile isotopes
4. The fertile isotope 240Pu has a large capture cross section for the production of the fissile isotope 241Pu.
η-values for important fissile nuclides

26. general features
Active core: (1) fissile fuel which through its fissioning is the main source of neutrons, (2) moderator material if the fission neutrons are to slow down, (3) coolant if the heat generated by the fissions is to be removed from the core, and (4) structural material which maintains the physical integrity of the core.
Reflector: scatter neutrons back towards the core
Blanket region: captures neutrons leaking from the core to produce useful isotopes such as 60Co
Shield
Control: allow the chain reaction to be started up, maintained at some desired level, and safely shutdown

27. Reactors are broadly classified according to the energy of the neutrons :
fast reactor, the fast fission neutrons do not slow
down very much before they are absorbed by the fuel and cause the production of a new generation of fission neutrons.
thermal reactor, almost all fissions are caused by neutrons that have slowed down and are moving with speeds comparable to those of the atoms of the core material, i.e., the neutrons are in thermal equilibrium with the surrounding material.

28. 3. Reactors Reactions Involving Neutrons
Thermal-Neutron Properties of Fuels
General features
The Neutron Life Cycle in a Thermal Reactor
Homogeneous and Heterogeneous Cores
Reflectors
Reactor Kinetics
Reactivity Feedback

29. The Neutron Life Cycle in a Thermal Reactor
The neutron life cycle in a thermal reactor showing the major mechanisms for the loss and gain of neutrons. The n fast neutrons beginning the cycle produce n‘ second-generation fast neutrons which, in turn, begin their life cycle

30. critical state supercritical state subcritical state
neutrons generated by nuclear fission +
neutrons increased by (n,2n), etc.
= absorbed neutrons + leaked neutrons
supercritical state
right side > left side
subcritical state

31. Quantification of the Neutron Cycle
1. fast fission factor ε （快中子增殖因数）: the ratio of the total number of fast neutrons produced by both thermal and fast fission to the number produced by thermal fission alone. ε -> 1 238U
2. resonance escape probability p （逃脱共振俘获概率）: the probability that a fast fission neutron slows to thermal energies without being absorbed p -> 1 235U
This is the measure of how many neutrons can go through resonances without being absorbed:

32. 3. thermal utilization f （热中子利用系数）: probability that, when a thermal neutron is absorbed, it is absorbed by the "fuel" (F) and not by the "nonfuel" (NF)
Σ is called the macroscopic cross section
for a homogeneous core
=0->1

34. 4. thermal fission factor η （热裂变中子数）: number of fast fission neutrons produced per thermal neutron absorbed by the "fuel."
Equivalently, η is the average number of neutrons per thermal fission (v) times the probability a fission occurs when a thermal neutron is absorbed by the fuel,
η > 1, sustaining chain reaction
is a property of the fuel material alone and is unaffected by the type and amount of nonfuel material in the core.

36. 5. thermal non-leakage probability （热中子在扩散过程中不泄漏概率） :
probability a thermal neutron does not leak from the core before it is absorbed.
R: spherical core of radius
there is no leakage
临界屈曲
for a homogeneous mixture of fuel (F) and moderator (M)
D is the thermal diffusion coefficient

37. （快中子在慢化过程中不泄漏概率）
the probability a fast neutron does not leak from the core as it slows to thermal energies.
Moderator properties for thermal ( eV) neutrons. L is the thermal diffusion length and Г is the Fermi age from fission to thermal energies.
Г ：one-sixth the mean squared distance between the point at which a fast fission neutron is born and begins to slow down and the point at which it reaches thermal energies.

38. thermal utilization

39. fastabsorp.

40. Effective Multiplication Factor （有效增殖因数）
For an infinite medium, there is no neutron leakage.
"four-factor formula"

41. subcritical
supercritical
critical
self-sustaining

42. What is of a homogeneous mixture of 235U and graphite
with an atomic uranium to carbon ratio of 1 to 40,000?
For such a dilute mixture of fully enriched uranium and carbon,
so that

43. What is the radius R of a critical bare sphere composed of a homogeneous mixture of 235U and graphite with a uranium to carbon atom ratio of 1 to 40,000?
For criticality,
R = 125 cm

44. Variation of Keff and its factors with the fuel-to-moderator
ratio. This example is for a homogeneous mixture of water and 2%-enriched uranium. Here NF /NM = atom density of uranium to molecular density of water.

45. 3. Reactors General features Reactions Involving Neutrons
Thermal-Neutron Properties of Fuels
The Neutron Life Cycle in a Thermal Reactor
Homogeneous and Heterogeneous Cores
Reflectors
Reactor Kinetics
Reactivity Effects
Fission Product Poisons

46. Homogeneous and Heterogeneous Cores
The least expensive fuel to use in a reactor assembly is natural uranium (0.72 atom-% 235U). But
a fast fission neutron would lose so little energy in each scatter from a uranium nucleus that over 2000 scatters would be required to slow the neutron to thermal energies
238U has large absorption cross sections,
for a pure natural uranium core, the resonance escape probability p would be very small so that <<1

47. The simplest assembly is a homogeneous mixture of natural uranium and a moderator material
Small, too little moderation, p is very small
Large, the thermal neutrons are not absorbed
easily by the fuel, f is small
Optimum moderator-fuel ratios for a homogeneous mixture of natural uranium and moderator

48. Increase fast fission factor ε
heterogeneous core
the fast neutrons are thermalized in the moderator away from the 238U and hence they can slow through the energy
ranges of the 238U resonances with little likelihood of being captured
increase p
Fast neutrons born in the fuel lumps have a greater probability of causing fast fissions in 238U if they are surrounded by only uranium atoms. In a homogeneous system, a fast fission neutron may first encounter a moderator atom, scatter, and
lose so much energy that it is no longer capable of causing fast fission.
Increase fast fission factor ε

49. Cross-section of a heterogeneous core
Cross-section of a heterogeneous core. Each unit cell of pitch a contains a 1.25-cm radius fuel rod (black circles) of natural uranium metal.
The remainder of each lattice cell is graphite.

50. Variation of core parameters with cell size for the natural uranium and graphite core

51. Reflectors
Most reactor cores are surrounded by some material that has a high scattering to-absorption cross section ratio (typical of moderators). This material, called a reflector
it reflects some of the neutrons which would escape or leak from a bare core back into the core, thereby increasing the nonleakage probabilities.
raise the thermal flux density near the core edges. For heat-transfer purposes it is desirable to maintain as constant a thermal flux profile

52. Reactor Kinetics A Simple Reactor Kinetics Model
Consider a core in which the neutron cycle takes l' seconds to complete
The change Δn in the total number of thermal neutrons in one cycle at time t or

53. =1.001
Uncontrollable !

54. 瞬发中子

55. A small fraction β (0.65% for 235U) of fission neutrons are emitted, not during the fission event, but by the radioactive decay of daughters of certain fission products at times up to minutes after the fission event that created the fission products.
The fission products, whose daughters decay by neutron emission, are called delayed neutron precursors and the emitted neutrons are called delayed neutrons.
An example of a fission product whose decay leads to a delayed neutron.

56. Delayed-neutrons are grouped by the apparent half-lives of the observed emission rates
delayed neutron precursors with similar half-lives are placed in the same delayed-neutron group.
Half-lives and yield-fractions ,βi of a six delayed-neutron group

57. proportion β of delayed neutrons is insignificant. For 235U, it is 0
proportion β of delayed neutrons is insignificant. For 235U, it is 0.65%, while the proportion of the prompt neutrons is 99.35%. Though the proportion of delayed neutrons is small but they have a very important effect in the control of the reactor.
Delayed neutrons, the average energy being
about one-half of that for prompt fission neutrons.
lengthening of the neutron cycle time , causes the neutron
population and reactor power to vary sufficiently slowly that control of the chain reaction is possible.

58. Reactivity and Delta-k
the degree of departure from criticality
reactivity
The factor that determines how subcritical or supercritical a reactor may be is
They are all positive for a supercritical system, negative for a subcritical reactor, and zero at criticality

59. Revised Simplified Reactor Kinetics Models
Consider a thermal reactor fueled with 235U
Delayed neutron average lifetime is
A fraction β of the fission neutrons requires a cycle time of
while a fraction (1 -β) is the prompt-neutron fraction and
requires a cycle time of only
The average or effective generation time required for all the neutrons produced in a single neutron cycle is thus
=0.083 s

60. -> s
controllable !

61. 3. Reactors General features Reactions Involving Neutrons
Thermal-Neutron Properties of Fuels
The Neutron Life Cycle in a Thermal Reactor
Homogeneous and Heterogeneous Cores
Reflectors
Reactor Kinetics
Reactivity Effects
Fission Product Poisons

62. REACTIVITY EFFECTS

63. Temperature effects on reactivity
The extent to which the reactivity is affected by changes in temperature is described in terms of temperature coefficient of reactivity, denoted as αT. This is defined by the relation:

64. Poisoning effect
Fission product mass yield per fission induced by thermal neutrons for important
fissile nuclides[2]. Total yield is 200%.

65. 135Xe has a large thermal neutron cross-section, and it greatly affects the reactivity control of the reactor
Fission products decay chain including 135Xe

66. Fission product yield and decay constant for 135I and 135Xe in the fission of 235U
will generally cause a divergent oscillation in the output power
135Xe works as positive feedback for the output power. Thus, it makes the output power unstable. However, since the half life of 135Xe is sufficiently long, a small fluctuation can be easily controlled with control rods.

67. Fuel burn-up

68. The reactivity when all control material is removed from the core is called the excess reactivity
Excess reactivity change during burnup

70. The following conditions have to be satisfied for nuclear fission reactions
① Exoergic reaction
② Sustainable as a chain reaction
③ Controllable
n + 235U  X + Y+ E MeV
Neutron evolution in a nuclear reactor