FLUKA: LOW-NEUT
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LOW-NEUT
Activates low-energy neutron transport (which in fact is already activated
by several DEFAULTS choices), dumps on request information concerning
groupwise treatment ingredients, and allows to activate generation of
secondary charged particles and correlated photon cascades, as well as
pointwise cross section treatment, both available only for special cases
See also LOW-BIAS, LOW-MAT
WHAT(1)
= number of neutron groups in the cross section set used.
The FLUKA standard neutron library has 260 groups (see 10}).
Default
= 260
WHAT(2)
= number of gamma groups
No default if WHAT(1)
is given, 42 otherwise. (The standard
FLUKA neutron library has 42 gamma groups).
WHAT(3)
= maximum energy of the low-energy cross section neutron library.
For the standard FLUKA neutron library, the maximum energy is
0.020 GeV.
Default
= 0.020 GeV.
WHAT(4)
= printing flag: from 0.0 to 4.0 increases the amount of
output about cross sections, kerma factors, etc.
1.0 : Standard output includes integral cross sections,
kerma factors and probabilities
2.0 : In addition, downscattering matrices and group
neutron-to-gamma transfer probabilities are printed
3.0 : In addition, scattering probabilities and angles are
printed
4.0 : In addition, information on residual nuclei is printed
Default
: 0.0 (minimum output)
WHAT(5)
= number of neutron groups to be considered thermal ones. (The
standard FLUKA neutron library has 31 thermal groups).
= 0, ignored
< 0: resets to the default = 31.0
Default
= 31.0
WHAT(6)
= i0 + 10 * i1:
i0 > 0: if no full pointwise treatment is activated
(see LOW-PWXS) a few pointwise cross sections
(1-H, 2-H, 3-He, 4-He, 12-C/Cnat) are used in a
mixed approach with the groupwise treatment (see
important details in Note 5 below), explicit and
correlated secondary generation for 10-B(n,alpha)7-Li
is activated, as well as correlated photon cascade for
x-Xe(n,gamma)x+1-Xe and 113-Cd(n,gamma)114-Cd
= 1: lower thresholds for pointwise treatment, including
when full pointwise treatment is invoked with LOW-PWXS,
set at default values:
10^-5 eV for free gas materials
3.059023 eV (260 group library) for materials for
which S(a,b) treatment is available
(default for some DEFAULTS, PRECISIOn ...)
= 2: same as for i0=1
= 3: lower thresholds for pointwise treatment set at:
3.059023 eV (260 group library) for all materials,
including when full pointwise treatment is invoked
with LOW-PWXS
= 4: lower thresholds for pointwise treatment set at:
10^-5 eV for all materials, including when full
pointwise treatment is invoked with LOW-PWXS
= 0: ignored
=<-1: resets to the default (pointwise cross sections
are not used, do not use if LOW-PWXS is invoked)
i1 = 1, fission neutron multiplicity forced to 1, with
proper weight
= 0, ignored
=<-1: resets to the default (normal fission multiplicity)
Default
= -11., unless option DEFAULTS has been chosen with
SDUM
= CALORIMEtry, DAMAGE, HADROTHErapy, ICARUS,
NEUTRONS or PRECISIOn, in which case the default is
1.0 (pointwise treatment - see Note 5 - and generation
of secondary charged particles and correlated photon
cascades are performed when available, and fission
multiplicity is not forced)
SDUM
: Not used
Default
(option LOW-NEUT not given): if option DEFAULTS is used with
SDUM
= CALORIMEtry, DAMAGE, EET/TRANsmut, HADROTHErapy, ICARUS,
NEUTRONS, NEW-DEFAults, PRECISIOn or SHIELDINg, low-energy
neutrons are transported and a suitable cross section library must
be available.
In all other cases, low-energy neutrons are not transported, and
their energy is deposited as explained in Note 2).
Notes:
1) In FLUKA, transport of neutrons with energies lower than a certain
threshold is performed by a multigroup algorithm. For the neutron
cross section library currently used by FLUKA, this threshold is
0.020 GeV. The multigroup transport algorithm is described in Chap.
10}.
2) If low-energy neutrons are not transported (because of the chosen
DEFAULTS, or because so requested by the user, see Note 3), the
energy of neutrons below threshold (default or set by PART-THR) is
deposited on the spot.
This is true also for evaporation neutrons.
3) If there is no interest in low-energy neutron transport, but that
feature is implicit in the DEFAULTS option chosen, it is suggested
to request LOW-NEUT, and to use PART-THRes with an energy cutoff
WHAT(1)
= 0.020. However, even in this case the availability of the
low-energy neutron cross sections for the materials defined in input
is checked. To avoid the run being stopped with an error message,
the user should issue a LOW-MAT command for each material for which
cross sections are missing, pointing them to any available material.
4) Gamma data are used only for gamma generation and not for transport
(transport is done via the FLUKA ElectroMagnetic module EMF using
continuous cross sections). The actual precise energy of a photon
generated by (n,gamma) or by inelastic reactions such as (n,n') is
sampled randomly within the gamma energy group concerned, except for
a few important reactions where a single monoenergetic photon is
emitted, as the 1-H(n,gamma)2-H reaction where the actual photon
energy of 2.226 MeV is used. It is possible to get (single or
correlated) physical gammas also for the capture in 10-B, 12-C,
x-Xe and 113-Cd, by setting WHAT(6)
= 1.0-4.0 or 11.0-14.0
(see Note 5 for the additional requirements applying to 12-C).
5) Full, correlated, pointwise treatment is available for all
isotopes when invoking the LOW-PWXS option. Otherwise, in the
framework of groupwise cross sections, special pointwise
neutron transport is available only for a few isotopes, 1-H,
2-H, 3-He, 4-He, and 12-C/Cnat, by setting WHAT(6)
= 1.0-4.0
or 11.0-14.0. In the case of 2-H in order to get the pointwise
treatment it is mandatory to define the respective monoisotopic
material through a MATERIAL card and name it DEUT.... (it must
start with DEUT).
In the case of 3-He, and 4-He, in order to get the pointwise
treatment it suffices to define the respective monoisotopic
material through a MATERIAL card and associate it to the low
energy material HELIUM-3, HELIUM-4, respectively.
For pointwise treatment for 1-H, if activated, it is sufficient to
name the material through a MATERIAL} card ...HYDR.... If
the material has natural composition, the (small) amount of 2-H
will be neglected. Of course one can define two monoisotopic
materials for 1-H and 2-H and mix them in the proper proportions
with a COMPOUND card.
Pointwise cross sections for 12-C are activated, if requested,
when the low energy neutron data set associated with the material
has name CARBO-12}, or CARBON..., for both 12-C and natural
Carbon. In the latter case the 13-C small abundance is neglected.
6) Recoil protons are always transported explicitly, and so is the
proton from the 14-N(n,p) reaction.
7) The groups are numbered in DECREASING energy order (see 10}
for a detailed description).
The energy limits of the thermal neutron groups in the standard
FLUKA neutron library neutron library are reported in 10.4.1.1}
8) Here are the settings for transport of low-energy neutrons
corresponding to available DEFAULTS SDUM
options:
CALORIMEtry, DAMAGE, HADROTHErapy, ICARUS, NEUTRONS, PRECISIOn:
low-energy neutrons are transported, with generation of
charged secondaries, correlated photon cascades and use of
pointwise cross sections when available
EET/TRANsmut, NEW-DEFAults (or DEFAULTS missing), SHIELDINg:
low-energy neutrons are transported using always
multigroup cross sections
Any other SDUM
value of DEFAULTS: no low-energy neutron transport
9) If treatment of low energy neutrons is requested, one must make sure
that the transport threshold for neutrons (set with PART-THR) be
equal to the minimum energy needed for neutron transport (typically
1E-5 eV). Please note that the behaviour of the PART-THR option for
neutrons has changed with respect to past releases.
10) For a given neutron group and material (see example below), the residual
nuclei production probabilities printed in the output are normalized to
the non-elastic cross section (e.g. a 0.5 probability means that that
nucleus is produced on average every other non-elastic interaction).
Example:
LOW-NEUT 260.0 42.0 0.020 2.0 31.0 11.0
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