• The input data should NOT use Tabs. It is generally technically impossible to catch it in FORTRAN, so there is no appropriate diagnosis on the subject.
  • When the calculation fails at the multiprocessor system before blaming the administrator check all files *. LST and *. SCR for the presence of an error message.
  • If you can use the BOX or SBOX instead of RPP, it is best to use the RPP. This is a consequence of the universality of the body BOX, which uses mathematics difficult form the machine point of view. This may cause the loss of particles at the boundaries of the body.
  • Suppose that the network element is embedded in the lattice adjacent to the boundary of the container with the condition of reflection / transmission. When a particle exits the network through this boundary it may go directly outside the container and be lost (and you get a leak instead of the reflection / transmission). The solution in this case may be to set the container element of the lattice, so that it "cuts" the network a bit, i.e. outer edge of the network is outside the boundaries of the network carrier zone, and does not coincide with it or is inside it.
  • Submodule MOFITTG at fission with the release of several neutrons simply increases the weight of the simulated neutron, so non-analog methods for the thermal region is better to be used with "honest" FIMTOEN submodule which really generates fission neutrons.
  • Verification calculations show significant differences in the responses of some problems when using ready-made files for natural compounds from libraries and job-mix done by hand. (When used in calculations of critical experiments HMF85_10, 11 with copper reflector the constants of natural mixture of isotopes (cu.j32) give significant (1%) overestimation of Kef. Nuclide constants of copper provide the proper Kef within the experimental error.)
    Given the fact that modern libraries do not contain constants for natural compounds any more it is recommended to do the mixtures from nuclides by hand in the input data.
  • Inclusion into the a single tally both, for example, fuel and the central hole (which is e.g. O 1.E-7) damages accrued in the fuel the statistics due to rare tallying in a material with a low concentration (you get an arbitrary answer with a large error, i.e . everything is legally and honestly printed, but it is not funny).
  • Accumulating source is not currently running in multiprocessor mode.
  • If a program is used with Intel compiler, the parameter in the *. MEM
    + $ MRECL $ 4 $ general multiplier for RECL parameter of direct opened
    !: Files (may be 1 or 4 depending on compiler, four is safe)
    may be set as
    + $ MRECL $ 1 $ general multiplier for RECL parameter of direct opened
    !: Files (may be 1 or 4 depending on compiler, four is safe)
    At multiprocessor calculation it reduces the length of some files and thus the amount of disk operations, which will reduce the total computation time. At the same time the actual tallies are not affected. If for any compiler, this option is not possible, then the code should just stop running. However, one should treat this change with caution. Traditionally, for the Intel compiler that works.
  • In the calculations with the submodule MOFITTG of the physical module it can calculate the transition matrix in the energy bins (cards MSMT, ZSMT, OSMT) ONLY for the mixture, as in this case is the nuclide at which the interaction occurred with the medium is not chosen. Thus, if there is a need to use MOFITTG the NUCOFF card should be used in the appropriate place with the corresponding effects of this card. The second option is to use FIMTOEN submodule of the physical module.
  • In the calculation of macroscopic cross-sections using cards MADF, ZADF, OADF one need to understand that the results obtained in the calculation and derived in print under the heading 'F. Assm' for the tally suppose an integral over all the tally area defined in the cards, this may coincide with the concept of Full Assembly.
  • It is not recommended to use estimations based on absorption acquisitions in the calculation of flux, as you can get incorrect results. This is due to the fact that in some materials (e.g., water) for all nuclides of a material absorption cross sections are zero in some energy ranges, and the total cross section (sigma total over a given interval) is not zero. In connection with this the flux rate is underestimated, because the fact of tallying in the relevant energy range does not occur.
  • One should not select a zone consisting of a body-container system to be the net-carrier. There may be many Geometry avost-s.
  • It is recommended to check (NOT CHANGE) the default value of the model to calculate the scattering cross-sections of thermalization (MODS =), defined in DEFUALT.PHY from the data bank, as from code to code it may be different, unfortunately. To change the defaults use the card DEF.
  • In burnup calculations the following isotopes cannot be used:
    T, F-19, Na-22, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, S-32, Ar-40, Ca-43, Ca-44, Ca-46, Ca-48, Ti-47, Ti-48, Ti-49, Ti-50, Se-74, Sr-84, Ru-96, Ru-105, Te- 120, Ba-130, La-140, Ce-136, Ce-138, Ce-143, Pr-142, Pm-151, Dy-165, Au-197, Hg-196, Hg-198, Hg-199, Hg-200, Hg-201, Hg-202, Hg-204, Ac-226, Th-233, Pa-232, Pu-246, Am-242, Am-244, Cm-249, Bk-250, Fm- 255.
    Diagnosis of the use appears only when the burnup calculation starts.
  • If a code is compiled via Intel Compiler 12, the large number of remark #8291 is normal.


  • Total length of one sentence in input data files may be changed in *.mem file:
    +::$LBUFS $1500000$ - total length of a sentence
  • Output of numbers of tally zones that contain more than one material may be activated by the card (to be used in the part of the data for geometry module before the card FINISH):
    C=C RGMM