Civil Engineering Reference
In-Depth Information
Fig. 2.7
Microscopic capture cross sections of Cadmium and Gadolinium isotopes and B-10 (n,
α
)
cross sections [
13
]
For many practical applications it is sufficient to solve the neutron diffusion
equation which is an approximation to the Boltzmann neutron transport equation.
The microscopic cross sections shown in Figs.
2.3
,
2.4
and
2.7
are collected in a
special format in cross section libraries, e.g. JEFF [
15
], ENDF/B [
16
], JENDL
[
17
]. Their continuous energy range can be approximated and divided into a
number of energy groups with specifically defined microscopic group cross sections
applying codes, e.g. NJOY [
19
]orMC
2
-3 [
20
,
21
]. The heterogeneous cell geom-
etry of the reactor core (Fig.
2.5
) can be accounted for by codes, e.g. WIMS [
22
]or
MC
2
-3 [
20
,
21
]. These computational methods are summarized in [
23
,
24
].
Whole-core calculations can be done in diffusion theory by codes, e.g. DIF3D
[
25
] or SIMULATE-4 [
26
]. Whole-core (Fig.
2.6
) codes applying Boltzmann
neutron transport theory were developed for two- and three-dimensional geome-
tries. Examples for such computer codes are, e.g. DANTSYS [
27
] or PARTISN
[
28
]. Monte Carlo Codes are available for both lattice (Fig.
2.5
) and whole-core
(Fig.
2.6
) geometries. Such codes include, e.g. MCNP5 [
29
] or VIM [
30
].
The number of neutrons in the reactor core can be controlled by moving or
adding, e.g. absorber materials (neutron poisons). This is done in a k
eff
-range, where
the delayed neutrons are dominating the transient behavior of the neutron flux. The
delayed neutrons come into being in a time range of seconds. Therefore, the number
of neutrons or the power in reactor cores can also be controlled safely by moving
absorber materials in the time range of seconds [
1
,
2
,
6
,
11
].
