10 239 94 5 4 -2 Pu 1.2 x 10 years 6.6 x 10
years 1.0 x 10 240 94 11 3 3 Cf 66.0 years
2.65 years 2.3 x 10 252 98 12 Another
intrinsic neutron source is a reaction
involving natural boron and fuel. In some
reactors, natural boron is loaded into the
reacto
may also be regarded as the effective area
the nucleus presents to the neutron for the
particular reaction. The larger the effective
area, the greater the probability for
reaction. Because the microscopic cross
section is an area, it is expressed in units
actually occur, it is necessary to know how
many neutrons are traveling through the
material and how many centimeters they
travel each second. It is convenient to
consider the number of neutrons existing
in one cubic centimeter at any one instant
and the
average decrease per collision in the
logarithm of the neutron energy. This
quantity is represented by the symbol
(Greek letter xi). (2-8) where: = average
logarithmic energy decrement Ei = average
initial neutron energy Ef = average final
neutron energy
and free neutrons. Beryllium-9 undergoes
an alpha-neutron reaction (alpha from the
decay of plutonium, polonium, or radium)
and becomes carbon-12. Beryllium-9
undergoes a gamma-neutron reaction (high
energy gamma from decay of antimony124) and becomes ber
(installed) neutron sources that are
incorporated into the design of the reactor.
The neutrons produced by sources other
than neutron-induced fission are often
grouped together and classified as source
neutrons. EO 1.1 DEFINE the following
terms: a. Intri
determined largely by the delayed neutron
generation time. The following equation
shows this mathematically. (2-12) Example:
Assume a prompt neutron generation time
for a particular reactor of 5 x 10 seconds
and -5 a delayed neutron generation time
of 12.
The neutron energy spectrum varies widely
for different types of reactors. EO 4.1 STATE
the average energy at which prompt
neutrons are produced. EO 4.2 DESCRIBE
the neutron energy spectrum in the
following reactors: a. Fast reactor b.
Thermal reactor EO
stable isotope beryllium-9 has a weakly
attached last neutron with a binding
energy of only 1.66 MeV. Thus, a gamma
ray with greater energy than 1.66 MeV can
cause neutrons to be ejected by the (
,n) reaction as shown below. NEUTRON
SOURCES DOE-HDBK-1019/
three distinct regions. At low neutron
energies (< 1 eV) are designated thermal
neutrons. The process of reducing the
energy of a neutron to the thermal region
by elastic scattering is referred to as
thermalization, slowing down, or
moderation. The materi
cm 1 a 1 a 1 0.01345 cm 1 74.3 cm s 1 s 1
0.0903 cm 1 11.1 cm DOE-HDBK-1019/1-93
Reactor Theory (Neutron Characteristics)
NUCLEAR CROSS SECTIONS AND NEUTRON
FLUX Rev. 0 Page 13 NP-02 Step 2: The
macroscopic cross sections for absorption
and scattering are
are released after the decay of fission
products and are called delayed neutrons.
Although delayed neutrons are a very
small fraction of the total number of
neutrons, they play an extremely
important role in the control of the reactor.
EO 3.1 STATE the or
immediately following the first beta decay
of a fission fragment known as a delayed
neutron precursor. An example of a
delayed neutron precursor is bromine-87,
shown below. PROMPT AND DELAYED
NEUTRONS DOE-HDBK-1019/1-93 Reactor
Theory (Neutron Characteris
generation time is about 12.5 seconds. & A
prompt neutron generation time is the sum
of the amount of time it takes a fast
neutron to thermalize, the amount of time
the neutron exists as a thermal neutron
before it is absorbed, and the amount of
time betw
OBJECTIVES TERMINAL OBJECTIVE 2.0
Given the necessary information for
calculations, EXPLAIN basic concepts in
reactor physics and perform calculations.
ENABLING OBJECTIVES 2.1 DEFINE the
following terms: a. Atom density d. Barn b.
Neutron flux e. Macrosco
The most probable velocity (vp ) of a
thermal neutron is determined by the
temperature of the medium and can be
determined by Equation (2-13) . (2-13)
where: v = most probable velocity of
neutron (cm/sec) p k = Boltzman's constant
(1.38 x 10 erg/ K) -16 T
neutron generation times. 3.5 Given
prompt and delayed neutron generation
times and delayed neutron fraction,
CALCULATE the average generation time.
3.6 EXPLAIN the effect of delayed neutrons
on reactor control. Rev. 0 Page ix NP-02
OBJECTIVES DOE-HDBK-10
which involve absorption (for
example, ) , ) , ) ) must be corrected for the
existing temperature. acf The following
formula is used to correct microscopic
cross sections for temperature. Although
the example illustrates absorption cross
section, the same
especially important at resonance
energies, where the absorption cross
sections are large. Summary The important
information in this chapter is summarized
below. DOE-HDBK-1019/1-93 Reactor
Theory (Neutron Characteristics) NUCLEAR
CROSS SECTIONS AND NEUTRO
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globe 348 Natasha Tsoy Senior Ta
constant fraction of the neutron energy per
collision (on the average), independent of
energy; thus, the neutron loses larger
amounts of energy per collision at higher
energies than at lower energies. The fact
that the neutrons lose a constant fraction
of
NEUTRON FLUX Reactor Theory (Neutron
Characteristics) NP-02 Page 6 Rev. 0
Introduction Fission neutrons are born
with an average energy of about 2 MeV.
These fast neutrons interact with the
reactor core materials in various
absorption and scattering react
a high logarithmic energy decrement and a
good slowing down power, but it is a poor
moderator because of its high probability
of absorbing neutrons. The most complete
measure of the effectiveness of a
moderator is the moderating ratio. The
moderating rati
microscopic cross section, . T = + T a s Both
the absorption and the scattering
microscopic cross sections can be further
divided. For instance, the scattering cross
section is the sum of the elastic scattering
cross section ( se) and the inelastic
scatte
(2-1). (2-1) where: N = atom density
(atoms/cm )3 ' = density (g/cm )3 N =
Avogadro's number (6.022 x 10
atoms/mole) A 23 M = gram atomic weight
N NA M 2.699 g cm3 6.022 x 1023 atoms
mole 26.9815 g mole 6.024 x 1022 atoms
cm3 DOE-HDBK-1019/1-93 Reactor Th
are usually less than 10 barns. With the
exception of hydrogen, for which the value
is fairly large, the elastic scattering cross
sections are generally small, for example, 5
barns to 10 barns. This is close to the
magnitude of the actual geometric cross
iron is 2.56 barns 3 and the gram atomic
weight is 55.847 g. Solution: Step 1: Using
Equation (2-1), calculate the atom density
of iron. Step 2: Use this atom density in
Equation (2-2) to calculate the macroscopic
cross section. ' ' DOE-HDBK-1019/1-93
NUC
neutrons per unit area and time
(neutrons/cm -sec) falling on 2 a surface
perpendicular to the direction of the beam.
One can think of the neutron flux in a
reactor as being comprised of many
neutron beams traveling in various
directions. Then, the neutro
NP-02 Page 32 Rev. 0 Prompt and Delayed
Neutrons Summary & Prompt neutrons are
released directly from fission within 10
seconds of the -13 fission event. & Delayed
neutrons are released from the decay of
fission products that are called delayed
neutron pr