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-MATWHAT(1)= number of neutron groups in the cross section set used. The FLUKA standard neutron library has 260 groups (see 10}).Default= 260WHAT(2)= number of gamma groups No default ifWHAT(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 printedDefault: 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 = 31.0Default= 31.0WHAT(6)= i0 + 10 * i1: i0 > 0: available pointwise cross sections (1-H, 2-H, 3-He, 4-He, 6-Li, 12-C/Cnat) are used (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 set at default values: 10^-5 eV for free gas materials 3.059023 eV (260 groups) for materials for which S(a,b) treatment is available (default for some DEFAULTS, PRECISIOn ...) = 2: lower thresholds for pointwise treatment set at: 10^-5 eV for free gas materials 3.059023 eV (260 groups) for materials for which S(a,b) treatment is available = 3: lower thresholds for pointwise treatment set at: 3.059023 eV (260 groups) for all materials = 4: lower thresholds for pointwise treatment set at: 10^-5 eV for all materials = 0: ignored =<-1: resets to the default (pointwise cross sections are not used) 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 withSDUM= 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 usedDefault(option LOW-NEUT not given): if option DEFAULTS is used withSDUM= 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 cutoffWHAT(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 6-Li, 10-B, 12-C, 40-Ar, x-Xe and 113-Cd, by settingWHAT(6)= 1.0-4.0 or 11.0-14.0 (see Note 5 for the additional requirements applying to 6-Li, 12-C, and 40-Ar). 5) Pointwise neutron transport, fully alternative to the groupwise one, is available only for 1-H, 2-H, 3-He, 4-He, 6-Li, and 12-C/Cnat, by settingWHAT(6)= 1.0-4.0 or 11.0-14.0. In the case of 2-H, and 6-Li, in order to get the pointwise treatment it is mandatory to define the respective monoisotopic material through a MATERIAL card and name them DEUT...,LITHIU-6 respectively (or with a character string containing LI-6 or 6-LI for 6-Li). 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. Pointwise treatment has been developed also for 40-Ar but with some limitations making it not suitable for all applications. It requires an additional cross section file to be requested to the authors. In addition, as for 6-Li, one has to define the respective monoisotopic material and call it ARGON-40 (or with a character string containing AR-40 or 40-AR). The physical gammas from the 40-Ar(n,gamma)41-Ar capture reaction can be obtained only in the context of the pointwise treatment. 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 DEFAULTSSDUMoptions: 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 otherSDUMvalue 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:*...+....1....+....2....+....3....+....4....+....5....+....6....+....7....+...LOW-NEUT 260.0 42.0 0.020 2.0 31.0 11.0* The low-energy neutron library used is the (260n, 42gamma) standard* multigroup library. The user requests a printout of cross sections, kerma* factors, probabilities, downscattering matrices and n-->gamma transfer* probabilities. Pointwise treatment and generation of secondary charged* particles and correlated photon cascades will be performed where available* (seeWHAT(6)and Note 5 above), and only one neutron per low-energy fission* will be emitted, with an adjusted weight.