Transport of neutrons with energies lower than a certain energy is performed
in FLUKA by a multigroup algorithm.
The energy boundary below which multigroup transport takes over depends in
principle on the cross section library used: in the library which is
presently distributed with the code this energy is 19.6 MeV (In FLUKA, there
are two neutron energy thresholds: one for high-energy neutrons (set by
option PART-THR) and one for low-energy neutrons (set by option LOW-BIAS).
The high-energy neutron threshold represents in fact the energy boundary
between continuous and discontinuous neutron transport.
The multi-group technique, widely used in low-energy neutron transport
programs, consists in dividing the energy range of interest in a
given number of intervals ("energy groups"). Elastic and inelastic reactions
are simulated not as exclusive processes, but by group-to-group transfer
probabilities forming a so-called "downscattering matrix".
The scattering transfer probability between different groups is
represented by a Legendre polynomial expansion truncated at the
(N+1)th term, as shown in the equation:
Sigma_s(g-->g',mu) = Sum_{i=0,N}(2i+1)/(4pi) P_i(mu) Sigma_i(g-->g')
where mu = Omega.Omega' is the scattering angle and N is the chosen
Legendre order of anisotropy.
The particular implementation used in FLUKA has been derived from that
of the MORSE program [Emm75] (although the relevant part of the
code has been completely rewritten). In the standard cross section library,
the energy range up to 19.6 MeV is divided into 72 energy groups of
approximately equal logarithmic width (one of which is thermal). The
angular probabilities for inelastic scattering are obtained by a
discretisation of a P5 Legendre polynomial expansion of the actual
scattering distribution which preserves its first 6 moments. The
generalised Gaussian quadrature scheme to generate the discrete
distribution is rather complicated: details can be found in the MORSE
manual [Emm75]. The result, in the case of a P5 expansion, is a
set of 6 equations giving 3 discrete polar angles (actually angle
cosines) and 3 corresponding cumulative probabilities.
In the library, the first cross-section table for an isotope (isotropic
term P_0) contains the transfer probabilities from each group g to
any group g': Sum_{g-->g'}/Sum_g, where Sum_g is the sum over all the
g' (including the "in-scattering" term g' = g). The next cross-section
table provides the P_1 term for the same isotope, the next the P_2
multigroup cross sections, etc.
Possible artifacts
The multigroup scheme adopted in FLUKA is reliable and much faster than
any possible approach using continuous cross sections. However, it is
important to remember that there are two rare situations where the
group approximation could give bad results.
One of such situations may occur when each neutron is likely to
scatter only once (e.g. in a very thin foil) before being scored: an
artifact then is possible, due to the discrete angular distribution.
In practice the problem vanishes entirely, however, as soon as there is
the possibility of two or more scatterings: it must be kept in mind, in
fact, that after a collision only the polar angle is sampled from a
discrete distribution, while the azimuthal angle is chosen randomly from
a uniform distribution. In addition, the 3 discrete angles are different
for each g --> g' combination and for each element or isotope. Thus, any
memory of the initial direction is very quickly lost after just a few
collisions.
The second possible artifact is not connected with the angular but
with the energy structure of the cross sections used. The group
structure is necessarily coarse with respect to the resonance
structure in many materials. A resonance in a material present in a
dilute mixture or as a small piece cannot affect much a smooth neutron
flux (case of so-called "infinite dilution") but if an isotope is
very pure and is present in large amounts, it can act as a "neutron
sink", causing sharp dips in the neutron spectrum corresponding to
each resonance. This effect, which results in a lower reaction rate
sigma*phi, is called "self-shielding" and is necessarily lost in
the process of cross section averaging over the width of each energy
group, unless a special correction is made. Such corrected cross
section sets with different degrees of self-shielding have been
included in the FLUKA library for a few important elements (Fe, Cu, Pb):
but it is the responsibility of the user to select the set with
the degree of self-shielding most suitable in each different case.
It is worth stressing that non-self-shielded materials are
perfectly adequate in most practical cases, because the presence of even
small amounts of impurities is generally sufficient to smooth out the
effect. On the other hand, in regions of non-resolved resonances
the multigroup approach is known to give very good results anyway.
Pointwise transport
For a few isotopes only, neutron transport can be done also using
continuous (pointwise) cross sections. For 1H, 6Li and 10B, it is
applied as a user option (above 10 keV in 1H, for all reactions in
6Li, and only for the reaction 10B(n,t gamma)4He in 10B). For the
reaction 14N(n,p)14C, pointwise neutron transport is always applied.
Secondary particle production
Gamma generation
In general, gamma generation by low-energy neutrons (but not gamma
transport) is treated in the frame of a multigroup scheme too. A
downscattering matrix provides the probability, for a neutron in a
given energy group, to generate a photon in each of 22 gamma energy
groups, covering the range 10 keV to 20 MeV. With the exception of a
few important gamma lines, such as the 2.2 MeV transition of Deuterium
and the 478 keV photon from 10B(n,gamma) reaction, the actual
energy of the generated photon is sampled randomly in the energy
interval corresponding to its gamma group. Note that the gamma
generation matrix does not include only capture gammas, but also
gammas produced in other inelastic reactions such as (n,n').
For a few elements (Cd, Xe, Ar), for which evaluated gamma production
cross sections could not be found, a different algorithm, based on
published energy level data, has been provided to generate explicitly
the full cascade of monoenergetic gammas [Fas01b].
In all cases, the generated gammas are transported in the same way
as all other photons in FLUKA, using continuous cross sections and
an explicit and detailed description of all their interactions with
matter, allowing for the generation of electrons, positrons, and even
secondary particles from photonuclear reactions.
Secondary neutrons
In the multigroup transport scheme, the production of secondary
neutrons via (n,xn) reactions is taken into account implicitly by the
so-called "non-absorption probability", a group-dependent factor
by which the weight of a neutron is multiplied after exiting a
collision. If the only possible reactions are capture and scattering,
the non-absorption probability is < 1, but at energies above the
threshold for (n,2n) reaction it can take values larger than 1.
Fission neutrons, however, are treated separately and created
explicitly using a group-dependent fission probability. They are
assumed to be emitted isotropically and their energy is sampled from
the fission spectrum appropriate for the relevant isotope and neutron
energy. The fission neutron multiplicity is obtained separately from
data extracted from European, American and Japanese databases.
Generation of charged particles
Recoil protons and protons from N(n,p) reaction are produced and
transported explicitly, taking into account the detailed kinematics
of elastic scattering, continuous energy loss with energy straggling,
delta ray production, multiple and single scattering.
The same applies to light fragments (alpha,3-H) from neutron
capture in 6-Li and 10-B, if pointwise transport has been
requested by the user. All other charged secondaries, including
fission fragments, are not transported but
their energy is deposited at the point of interaction (kerma
approximation).
Residual nuclei
For many materials, but not for all, group-dependent information on
the residual nuclei produced by low-energy neutron interactions
is available in the FLUKA library. This information can be used to
score residual nuclei, but it is important that the user check
its availability before requesting scoring.
Fission fragments are sampled separately, using evaluated data
extracted from European, American and Japanese databases.
The FLUKA neutron cross section library
As explained in 3}, an unformatted cross-section data set
must be available for low-energy neutron transport.
The Legendre expansion used in Fluka is P5, i.e. at each collision
the polar scattering angle is sampled from three discrete values (the
azimuthal angle is sampled instead from a uniform distribution between
0 and 2*pi). The energy group structure depends on the cross-section
set used. Here below the group structure of the currently available sets
is reported, and a list of the materials they contain.
One neutron cross-sections set is presently available. It was
originally prepared by G. Panini of ENEA [Cuc91], but it is being
continuously enriched and updated on the basis of the most recent
evaluations (ENDF/B, JEF, JENDL, etc.). It contains almost 150 different
materials (natural elements or single nuclides), selected for their interest
in physics, dosimetry and accelerator engineering. More materials can be
made available on request. Some cross sections are available at two or three
different temperatures. Doppler broadening is taken into account.
In some cases, data from different evaluated files are available for the
same material. For hydrogen, cross sections are provided for both gas
hydrogen and hydrogen bound in water. The maximum energy is 0.0196 GeV.
Note that the energy groups are numbered in order of DECREASING energy
(group 1 corresponds to the highest energy).
The standard cross section set has 72 neutron energy groups and 22 gamma groups,
a structure which has been chosen for practical considerations.
Gamma energy groups are used only for (n,gamma) production, since transport of
photons in FLUKA is continuous in energy and angle and is performed through the
EMF module).
Each material is identified by an alphanumeric name (a string not
longer than 8 characters, all in upper case), and by three integer
identifiers. Correspondence with FLUKA materials (standard or
user-defined) is based on any combination of name and zero or more
identifiers. In case of ambiguity, the first material in the list
fulfilling the combination is selected.
The convention generally used (but there may be exceptions) for the 3
identifiers is:
- Atomic number
- Mass number, or natural isotopic composition if negative
(exceptions are possible in order to distinguish between
data from different sources referring to the same nuclide)
- Neutron temperature in degrees Kelvin
Low-energy neutron transport is activated by option LOW-NEUT.
Energy group structure of the 72-neutron, 22-gamma groups ENEA
data set:
Neutron group n.: 1 upper limit 1.9600E-02 GeV
Neutron group n.: 2 upper limit 1.7500E-02 GeV
Neutron group n.: 3 upper limit 1.4918E-02 GeV
Neutron group n.: 4 upper limit 1.3499E-02 GeV
Neutron group n.: 5 upper limit 1.2214E-02 GeV
Neutron group n.: 6 upper limit 1.1052E-02 GeV
Neutron group n.: 7 upper limit 1.0000E-02 GeV
Neutron group n.: 8 upper limit 9.0484E-03 GeV
Neutron group n.: 9 upper limit 8.1873E-03 GeV
Neutron group n.: 10 upper limit 7.4082E-03 GeV
Neutron group n.: 11 upper limit 6.7032E-03 GeV
Neutron group n.: 12 upper limit 6.0653E-03 GeV
Neutron group n.: 13 upper limit 5.4881E-03 GeV
Neutron group n.: 14 upper limit 4.9659E-03 GeV
Neutron group n.: 15 upper limit 4.4933E-03 GeV
Neutron group n.: 16 upper limit 4.0657E-03 GeV
Neutron group n.: 17 upper limit 3.6788E-03 GeV
Neutron group n.: 18 upper limit 3.3287E-03 GeV
Neutron group n.: 19 upper limit 3.0119E-03 GeV
Neutron group n.: 20 upper limit 2.7253E-03 GeV
Neutron group n.: 21 upper limit 2.4660E-03 GeV
Neutron group n.: 22 upper limit 2.2313E-03 GeV
Neutron group n.: 23 upper limit 2.0190E-03 GeV
Neutron group n.: 24 upper limit 1.8268E-03 GeV
Neutron group n.: 25 upper limit 1.6530E-03 GeV
Neutron group n.: 26 upper limit 1.4957E-03 GeV
Neutron group n.: 27 upper limit 1.3534E-03 GeV
Neutron group n.: 28 upper limit 1.2246E-03 GeV
Neutron group n.: 29 upper limit 1.1080E-03 GeV
Neutron group n.: 30 upper limit 1.0026E-03 GeV
Neutron group n.: 31 upper limit 9.0718E-04 GeV
Neutron group n.: 32 upper limit 8.2085E-04 GeV
Neutron group n.: 33 upper limit 7.4274E-04 GeV
Neutron group n.: 34 upper limit 6.0810E-04 GeV
Neutron group n.: 35 upper limit 4.9787E-04 GeV
Neutron group n.: 36 upper limit 4.0762E-04 GeV
Neutron group n.: 37 upper limit 3.3373E-04 GeV
Neutron group n.: 38 upper limit 2.7324E-04 GeV
Neutron group n.: 39 upper limit 2.2371E-04 GeV
Neutron group n.: 40 upper limit 1.8316E-04 GeV
Neutron group n.: 41 upper limit 1.4996E-04 GeV
Neutron group n.: 42 upper limit 1.2277E-04 GeV
Neutron group n.: 43 upper limit 8.6517E-05 GeV
Neutron group n.: 44 upper limit 5.2475E-05 GeV
Neutron group n.: 45 upper limit 3.1828E-05 GeV
Neutron group n.: 46 upper limit 2.1852E-05 GeV
Neutron group n.: 47 upper limit 1.5034E-05 GeV
Neutron group n.: 48 upper limit 1.0332E-05 GeV
Neutron group n.: 49 upper limit 7.1018E-06 GeV
Neutron group n.: 50 upper limit 4.8809E-06 GeV
Neutron group n.: 51 upper limit 3.3546E-06 GeV
Neutron group n.: 52 upper limit 2.3054E-06 GeV
Neutron group n.: 53 upper limit 1.5846E-06 GeV
Neutron group n.: 54 upper limit 1.0446E-06 GeV
Neutron group n.: 55 upper limit 6.8871E-07 GeV
Neutron group n.: 56 upper limit 4.5400E-07 GeV
Neutron group n.: 57 upper limit 2.7537E-07 GeV
Neutron group n.: 58 upper limit 1.6702E-07 GeV
Neutron group n.: 59 upper limit 1.0130E-07 GeV
Neutron group n.: 60 upper limit 6.1442E-08 GeV
Neutron group n.: 61 upper limit 3.7267E-08 GeV
Neutron group n.: 62 upper limit 2.2603E-08 GeV
Neutron group n.: 63 upper limit 1.5535E-08 GeV
Neutron group n.: 64 upper limit 1.0677E-08 GeV
Neutron group n.: 65 upper limit 7.3375E-09 GeV
Neutron group n.: 66 upper limit 5.0435E-09 GeV
Neutron group n.: 67 upper limit 3.4662E-09 GeV
Neutron group n.: 68 upper limit 2.3824E-09 GeV
Neutron group n.: 69 upper limit 1.6374E-09 GeV
Neutron group n.: 70 upper limit 1.1254E-09 GeV
Neutron group n.: 71 upper limit 6.8257E-10 GeV
Neutron group n.: 72 upper limit 4.1400E-10 GeV
lower limit 1.0000E-14 GeV
Gamma group n.: 1 upper limit 2.0000E-02 GeV
Gamma group n.: 2 upper limit 1.4000E-02 GeV
Gamma group n.: 3 upper limit 1.2000E-02 GeV
Gamma group n.: 4 upper limit 1.0000E-02 GeV
Gamma group n.: 5 upper limit 8.0000E-03 GeV
Gamma group n.: 6 upper limit 7.5000E-03 GeV
Gamma group n.: 7 upper limit 7.0000E-03 GeV
Gamma group n.: 8 upper limit 6.5000E-03 GeV
Gamma group n.: 9 upper limit 6.0000E-03 GeV
Gamma group n.: 10 upper limit 5.5000E-03 GeV
Gamma group n.: 11 upper limit 5.0000E-03 GeV
Gamma group n.: 12 upper limit 4.5000E-03 GeV
Gamma group n.: 13 upper limit 4.0000E-03 GeV
Gamma group n.: 14 upper limit 3.5000E-03 GeV
Gamma group n.: 15 upper limit 3.0000E-03 GeV
Gamma group n.: 16 upper limit 2.5000E-03 GeV
Gamma group n.: 17 upper limit 2.0000E-03 GeV
Gamma group n.: 18 upper limit 1.5000E-03 GeV
Gamma group n.: 19 upper limit 1.0000E-03 GeV
Gamma group n.: 20 upper limit 4.0000E-04 GeV
Gamma group n.: 21 upper limit 2.0000E-04 GeV
Gamma group n.: 22 upper limit 1.0000E-04 GeV
lower limit 1.0000E-05 GeV
List of materials for which ENEA cross-sections are available:
"SS" means "self-shielded".
"Shield." means "partially self-shielded, appropriate for typical cast iron
used for shielding"
"SN" means "partially self-shielded, appropriate for a natural composition
without a dominant isotope".
A symbol Y or N in column "RN" refers to availability of information
about production of residual nuclei; in column "Gam" to information
about gamma production.
Material Temp. Origin RN Name Identifiers Gam
H H2O bound nat. Hydrogen (1) 293K JEF-2.2 N HYDROGEN 1 -2 293 Y
H CH2 bound nat. Hydrogen (1) 293K JEF-2.2 N HYDROGEN 1 -3 293 Y
H Bound nat. Hydrogen (1) 293K JEF-2.2 N HYDROGEN 1 -4 293 Y
H Free natural Hydrogen (1) 293K JEF-2.2 N HYDROGEN 1 -5 293 Y
1H H2O bound Hydrogen 1 293K JEF-2.2 N HYDROG-1 1 +1 293 Y
1H CH2 bound Hydrogen 1 293K JEF-2.2 N HYDROG-1 1 +11 293 Y
1H bound Hydrogen 1 293K JEF-2.2 N HYDROG-1 1 +21 293 Y
1H Free Hydrogen 1 293K JEF-2.2 N HYDROG-1 1 +31 293 Y
H H2O bound nat. Hydrogen (1) 87K JEF-2.2 N HYDROGEN 1 -2 87 Y
H Natural Hydrogen (1) 87K JEF-2.2 N HYDROGEN 1 -5 87 Y
2H Deuterium 293K JEF-1 N DEUTERIU 1 +2 293 Y
3He Helium 3 293K JEF-1 Y HELIUM-3 2 +3 293 N
He Natural Helium (1) 293K JEF-1 Y HELIUM 2 -2 293 N
Li Natural Lithium (1) 293K ENDF/B-VI Y LITHIUM 3 -2 293 Y
Li Natural Lithium (1) 87K ENDF/B-VI Y LITHIUM 3 -2 87 Y
6Li Lithium 6 293K ENDF/B-VI Y LITHIU-6 3 +6 293 Y
6Li Lithium 6 87K ENDF/B-VI Y LITHIU-6 3 +6 87 Y
7Li Lithium 7 293K ENDF/B-VI Y LITHIU-7 3 +7 293 Y
7Li Lithium 7 87K ENDF/B-VI Y LITHIU-7 3 +7 87 Y
9Be Beryllium 9 293K ENDF/B-VI Y BERYLLIU 4 +9 293 Y
B Natural Boron (1) 293K JEF-2.2 Y BORON 5 -2 293 Y
10B Boron 10 293K JEF-2.2 Y BORON-10 5 +10 293 Y
11B Boron 11 293K ENDF/B-VI Y BORON-11 5 +11 293 Y
C Natural Carbon 293K ENDF/B6R8 Y CARBON 6 -2 293 Y
C Natural Carbon 87K ENDF/B6R8 Y CARBON 6 -2 87 Y
N Natural Nitrogen (1,16) 293K JEF-2.2 Y NITROGEN 7 -2 293 Y
N Natural Nitrogen (1,16) 87K JEF-2.2 Y NITROGEN 7 -2 87 Y
16O Oxygen 16 293K ENDF/B6R8 Y OXYGEN 8 +16 293 Y
16O Oxygen 16 87K ENDF/B6R8 Y OXYGEN 8 +16 87 Y
19F Fluorine 19 293K ENDF/B-VI Y FLUORINE 9 -2 293 Y
23Na Sodium 23 293K JEF-2.2 Y SODIUM 11 -2 293 Y
Mg Nat. Magnesium (2) 293K JENDL3.2 Y MAGNESIU 12 -2 293 Y
Mg Nat. Magnesium (2) 87K JENDL3.2 Y available on request Y
27Al Aluminium 27 293K ENDF/B6R8 Y ALUMINUM 13 -2 293 Y
27Al Aluminium 27 87K ENDF/B6R8 Y ALUMINUM 13 -2 87 Y
27Al Aluminium 27 4K ENDF/B6R8 Y ALUMINUM 13 -2 4 Y
Si Natural Silicon (1) 293K JENDL3.3 Y SILICON 14 -2 293 Y
Si Natural Silicon (1) 87K JENDL3.3 Y available on request Y
31P Phosphorus 31 293K ENDF/B-VI Y PHOSPHO 15 -2 293 Y
S Natural Sulfur (3) 293K JENDL3.2 Y SULFUR 16 -2 293 Y
Cl Nat. Chlorine (4) 293K ENDF/B-VI N CHLORINE 17 -2 293 Y
Ar Natural Argon 293K ENDL Y ARGON 18 -2 293 Y
Ar Natural Argon 87K ENDL Y ARGON 18 -2 87 Y
K Nat. Potassium (5) 293K ENDF/B-VI Y POTASSIU 19 -2 293 Y
K Nat. Potassium (5) 87K ENDF/B-VI Y available on request Y
Ca Natural Calcium (6) 293K JENDL-3.3 Y CALCIUM 20 -2 293 Y
Ti Natural Titanium (7) 293K JEF-2.2 Y TITANIUM 22 -2 293 Y
V Vanadium 293K ENDF/B-VI Y VANADIUM 23 -2 293 Y
V Vanadium 87K ENDF/B-VI Y available on request Y
Cr Natural Chromium (1) 293K ENDF/B-VI Y CHROMIUM 24 -2 293 Y
Cr Natural Chromium (1) 87K ENDF/B-VI Y CHROMIUM 24 -2 87 Y
55Mn Manganese 55 293K ENDF/B-VI Y MANGANES 25 55 293 Y
55Mn Manganese 55 87K ENDF/B-VI Y available on request Y
Fe Natural Iron (1) 293K ENDF/B-VI Y IRON 26 -2 293 Y
Fe Nat. Iron SS (1,8) 293K ENDF/B-VI Y IRON 26 -4 293 Y
Fe Nat. Iron Shield. (1,9) 293K ENDF/B-VI Y IRON 26 -8 293 Y
Fe Natural Iron (1) 87K ENDF/B-VI Y IRON 26 -2 87 Y
Fe Natural Iron SS (1,8) 87K ENDF/B-VI Y IRON 26 -4 87 Y
Fe Natural Iron Shield. (1,9) 87K ENDF/B-VI Y IRON 26 -9 87 Y
59Co Cobalt 59 293K ENDF/B-VI Y COBALT 27 59 293 Y
59Co Cobalt 59 87K ENDF/B-VI Y available on request Y
Ni Natural Nickel (1) 293K ENDF/B-VI Y NICKEL 28 -2 293 Y
Ni Natural Nickel (1) 87K ENDF/B-VI Y NICKEL 28 -2 87 Y
Cu Natural Copper (1) 293K ENDF/B-VI Y COPPER 29 -2 293 Y
Cu Natural Copper SS (1,8) 293K ENDF/B-VI Y COPPER 29 -4 293 Y
Cu Natural Copper SN (1,10) 293K ENDF/B-VI Y COPPER 29 -5 293 Y
Cu Natural Copper (1) 87K ENDF/B-VI Y COPPER 29 -2 87 Y
Cu Natural Copper SS (1,8) 87K ENDF/B-VI Y COPPER 29 -4 87 Y
Cu Natural Copper SN (1,10) 87K ENDF/B-VI Y COPPER 29 -5 87 Y
Zn Natural Zinc 293K ENEA N ZINC 30 -2 293 Y
Zn Natural Zinc 87K ENEA N ZINC 30 -2 87 Y
Ga Natural Gallium 293K ENDF/B-VI N GALLIUM 31 -2 293 Y
Ga Natural Gallium 87K ENDF/B-VI N available on request Y
Ge Natural Germanium (1) 293K JEF-1 N GERMANIU 32 -2 293 N
75As Arsenic 75 293K JEF-1 Y ARSENIC 33 75 293 N
Br Natural Bromine (1) 293K JENDL-3.2 Y BROMINE 35 -2 293 N
Kr Natural Krypton (1) 293K JEF-2.2 Y KRYPTON 36 -2 293 N
Kr Natural Krypton (1) 120K JEF-2.2 Y KRYPTON 36 -2 120 N
Sr Natural Strontium (1) 293 JENDL3.2 Y STRONTIU 38 -2 293 N
90Sr Strontium 90 293K JEF-2.2 Y 90-SR 38 90 293 N
Zr Natural Zirconium (1) 293K BROND Y ZIRCONIU 40 -2 293 Y
93Nb Niobium 93 (11) 293K ENDF/B6R8 Y NIOBIUM 41 93 293 Y
93Nb Niobium 93 (11) 87K ENDF/B6R8 Y available on request Y
Mo Natural Molybdenum (1) 293K EFF2.4 Y MOLYBDEN 42 -2 293 Y
Mo Natural Molybdenum (1) 87K EFF2.4 Y available on request Y
99Tc Technetium 99 293K JEF-2.2 Y 99-TC 43 99 293 N
Ag Natural Silver (1,11) 293K JENDL-3.2 Y SILVER 47 -2 293 Y
Cd Natural Cadmium (17) 293K JENDL-3 Y CADMIUM 48 -2 293 (Y)(12)
In Natural Indium (1) 293K JEF-1 Y INDIUM 49 -2 293 N
Sn Natural Tin (1) 293K JENDL-3 Y TIN 50 -2 293 N
Sn Natural Tin (1) 87K JENDL-3 Y TIN 50 -2 87 N
Sb Natural Antimony (1) 293K ENDF/B-VI Y ANTIMONY 51 -2 293 N
Sb Natural Antimony (1) 87K ENDF/B-VI Y available on request N
127I Iodine 127 293K JEF-2.2 Y IODINE 53 127 293 N
129I Iodine 129 293K JEF-2.2 Y available on request N
Xe Natural Xenon (1,13) 293K JEF-2.2 Y XENON 54 -2 293 (N)
124Xe Xenon 124 93K JEF-2.2 Y 124-XE 54 124 293 (N)(13)
126Xe Xenon 126 293K JEF-2.2 Y 126-XE 54 126 293 (N)(13)
128Xe Xenon 128 293K JEF-2.2 Y 128-XE 54 128 293 (N)(13)
129Xe Xenon 129 293K JEF-2.2 Y 129-XE 54 129 293 (N)(13)
130Xe Xenon 130 293K JEF-2.2 Y 130-XE 54 130 293 (N)(13)
131Xe Xenon 131 293K JEF-2.2 Y 131-XE 54 131 293 (N)(13)
132Xe Xenon 132 293K JEF-2.2 Y 132-XE 54 132 293 (N)(13)
134Xe Xenon 134 293K JEF-2.2 Y 134-XE 54 134 293 (N)(13)
135Xe Xenon 135 293K JEF-2.2 Y 135-XE 54 135 293 (N)(13)
136Xe Xenon 136 293K JEF-2.2 Y 136-XE 54 136 293 (N)(13)
133Cs Cesium 133 293K JEF-2.2 Y CESIUM 55 133 293 N
135Cs Cesium 135 293K JEF-2.2 Y available on request N
137Cs Cesium 137 293K JEF-2.2 Y available on request N
Ba Natural Barium (1) 293K JEF-1 Y BARIUM 56 -2 293 N
Ce Natural Cerium (1) 293K JENDL-3.2 Y CERIUM 58 -2 293 N
Nd Natural Neodymium (1) 293K JENDL-3.2 Y NEODYMIU 60 -2 293 N
Sm Natural Samarium (1) 293K ENDF/B6R8 Y SAMARIUM 62 -2 293 N
Eu Natural Europium (1,11) 293K ENDF/B6R8 Y EUROPIUM 63 -2 293 Y
Gd Natural Gadolinium (1) 293K ENDF/B-VI Y GADOLINI 64 -2 293 N
Gd Natural Gadolinium (1) 87K ENDF/B-VI Y available on request N
181Ta Tantalum 181 (11,14) 293K ENDF/B6R8 Y TANTALUM 73 181 293 Y
181Ta Tantalum 181 (11,14) 87K ENDF/B6R8 Y available on request Y
W Nat. Tungsten (1,11) 293K ENDF/B6R8 Y TUNGSTEN 74 -2 293 Y
W Nat. Tungsten (1,11) 87K ENDF/B6R8 Y TUNGSTEN 74 -2 87 Y
186W Tungsten 186 293K ENDF/B6R8 Y 186-W 74 186 293 Y
Re Natural Rhenium (1) 293K ENDF/B-VI Y RHENIUM 75 -2 293 N
Re Natural Rhenium (1) 87K ENDF/B-VI Y available on request N
197Au Gold 197 (11) 293K ENDF/B-VI Y GOLD 79 197 293 Y
Hg Natural Mercury (1,15) 293K JENDL-3.3 Y MERCURY 80 -2 293 N
Pb Natural Lead (1) 293K ENDF/B6R6 Y LEAD 82 -2 293 Y
Pb Natural Lead SS (1,8) 293K ENDF/B6R6 Y LEAD 82 -4 293 Y
208Pb Lead 208 293K ENDF/B6R6 Y 208-PB 82 208 293 Y
208Pb Lead 208 SS (8) 293K ENDF/B6R6 Y 208-PB 82 1208 293 Y
Pb Natural Lead (1) 87K ENDF/B6R6 Y LEAD 82 -2 87 Y
Pb Natural Lead SS (1,8) 87K ENDF/B6R6 Y LEAD 82 -4 87 Y
209Bi Bismuth 209 (14) 293K JEF-2.2 Y BISMUTH 83 209 293 Y
209Bi Bismuth 209 (14) 87K JEF-2.2 Y available on request Y
230Th Thorium 230 293K ENDF/B-VI Y 230-TH 90 230 293 N
232Th Thorium 232 293K ENDF/B-VI Y 232-TH 90 232 293 Y
233U Uranium 233 293K ENDF/B-VI Y 233-U 92 233 293 Y
234U Uranium 234 293K ENDF/B-VI Y 234-U 92 234 293 N
234U Uranium 234 87K ENDF/B-VI Y available on request N
235U Uranium 235 293K ENDF/B-VI Y 235-U 92 235 293 Y
235U Uranium 235 87K ENDF/B-VI Y 235-U 92 235 87 Y
238U Uranium 238 293K ENDF/B-VI Y 238-U 92 238 293 Y
238U Uranium 238 87K ENDF/B-VI Y 238-U 92 238 87 Y
237Np Neptunium 237 293K ENDF/B-VI Y 237-NP 93 237 293 Y
239Pu Plutonium 239 293K ENDF/B-VI Y 239-PU 94 239 293 Y
239Pu Plutonium 239 87K ENDF/B-VI Y available on request Y
241Am Americium 241 293K ENDF/B6R8 Y 241-AM 95 241 293 Y
243Am Americium 243 293K ENDF/B6R8 Y 243-AM 95 243 293 Y
(1: Material of natural isotopic composition obtained by collapsing together single
isotope data obtained from the quoted "origin".
(2: The information on residual nuclei for natural Magnesium have been obtained
from the individual Mg isotope data from JENDL-3.2. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.2 individual isotopes have been used together with the JENDL-3.2
natural Magnesium cross sections.
(3: The information on residual nuclei for natural Sulfur have been obtained
from the individual S isotope data from JENDL-3.2. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.2 individual isotopes have been used together with the JENDL-3.2
natural Sulfur cross sections.
(4: The information on residual nuclei for natural Chlorine have been obtained
from the individual Cl isotope data from JENDL-3.3. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.3 individual isotopes have been used together with the ENDF/B-VI
natural Chlorine cross sections.
(5: The information on residual nuclei for natural Potassium have been obtained
from the individual K isotope data from JENDL-3.2. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.2 individual isotopes have been used together with the ENDF/B-VI
natural Potassium cross sections.
(6: The Calcium data have been processed for each isotope out of JENDL-3.3
since it is the only compilation containing data for individual calcium
isotopes with gamma production. They have problems in the kerma factors,
for 42-Ca, 46-Ca, 48-Ca, which could result from overall inconsistencies
in the data. These isotopes are however a tiny fraction of the total.
(7: The residual nuclei infos for natural Titanium have been obtained
from the individual Ti isotope data from JENDL-3.2. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.2 individual isotopes have been used together with the JEF-2.2
natural Titanium cross sections.
(8: Self-shielded.
(9: Partially self-shielded, appropriate for typical cast iron used for
shielding.
(10: Partially self-shielded, appropriate for a natural composition
without a dominant isotope.
(11: A couple of kerma values are incorrect.
(12: The cross sections for gamma generation in Cd in the JENDL-3.2 evaluated
file are considered unreliable. For this element, it is recommended to
activate the explicit gamma generation routine of FLUKA (LOW-NEUT,
with WHAT(6) = 1.0 or 11.0)
(13: Cross sections for gamma generation in Xe are missing from available
evaluated data files. However, for the single isotopes it is possible to
activate an explicit gamma generation routine of
FLUKA (capture gamma only!!). Use LOW-NEUT, with WHAT(6) = 1.0 or 11.0.
Gamma generation in Xe of natural composition can then be obtained by
defining a COMPOUND of the individual isotopes, each with its natural
abundance.
(14: The (capture) gamma multiplicity is strikingly different among different
evaluations
(15: The Mercury data have been processed for each isotope out of JENDL-3.3
since it is the only compilation containing data for individual Mercury
isotopes up to now. There are problems in the kerma factors, which could
result from overall inconsistencies in the data. Use with care!!!
(16: 14-N has been obtained from ENDF/B6R8, while 15-N from JENDL-3.2
(17: The information on residual nuclei for natural Cadmium have been obtained
from the individual Cd isotope data from JENDL-3.2. However these data
do not contain gamma production, so the residual nuclei data from
JENDL-3.2 individual isotopes have been used together with the JENDL-3
natural Cadmium cross sections.
