Maximal intensity of the neutron flux allowed to the professional workers working with the radioactive material depends from the neutron energy. The definition "fast neutron" means various neutron energies in the different field of science and technology. For instance, for the 20 MeV neutrons the maximal intensity equals to ~9.5 n*cm-2*c-1 for the isotropic irradiation. This value may be refined by the local low. Having this value and  the minimal distance to the source that is necessary for work one can estimate the maximum source intensity. 

First Thank you Andrey V Daniel. I know the Categories of the neutrons which depend on the energies and the penetrating power,  but my question about the dose rate of neutrons which say that the neutron source is very dangerous and cannot be handled. in other words the limit of neutron counts emitted from point sources like Cf-252 and Am-Be sources. 

Your question is unfortunately not fully complete. If the emission rate or fluence of a neutron source is known, we can calculate the same value at any distance using inverse square law. For the neutron sources mentioned by you, the fluence-to-dose equivalent conversion at a distance, d , can be computed using conversion factors for the particular neutron energy: Please refer to: NRC:10CFR 20.1004. Let me give you some values: For Am-Be, average energy of neutrons is 4.46 MeV and for Cf-252, average energy is 2.14 MeV and the corresponding fluence-to-dose equivalent factors are: Am-Be (QF=8.5): 1 rem = 24 x 106 n/cm2 and for Cf-252(QF=10); 1 rem = 28 x 106 n/cm2. The maximum permissible dose equivalent for a radiation worker is 20mSv/year or about 1mrem/hour (1 year = 52 weeks, 1 week = 40 working hours). From these values, you can work out whether, at the distance  away from the source you normally work, the dose is within the permissible limit. Otherwise, working backwards, calculate the distance at which you are safe. It may be of interest for you to know that you can design a shield for your neutron source by surrounding it by a paraffin cylinder on all sides to fully thermalize the fast neutrons which, in turn, is surrounded by a cadmium cylinder to absorb the thermal neutrons. The capture gamma rays produced by cadmium can be suppressed by surrounding this assembly by a further layer of lead on all sides. The high quality factor (QF) for neutrons implies neutrons are much more hazardous than gamma rays (QF=1). Personnel monitoring neutron badges are available from regulatory authorities. 

Hope the above analysis helps. For better accuracy, the energy spectrum of neutrons should be taken into account.

The problem is more complicated than the determination of the "allowed" fast neutron fluence rate, mentioned in the three responses above.

The question contains the quantity  "fast neutron counts/m" . The documentation of the used dosimeter shall contain curves (or at least detailed tables) of the count rate to fluence rate conversion factors as function of neutron energy. Depending on the size of the moderator mounted around the detector, the user has to have a set of those. In case of monoenergetic neutron source the calculation counts rate --> fluence rate at the detector --> dose rate at the detector is a simple task. On the other side, in case of source with many energies, the count rate --> dose rate conversion is approximation only. (And what about the detector -- person distance?)

This has reference to the comments of Egon Zondi above. My comments above were in response to the query of Muhammad Mansy who wished to know about safety from a source with known fluence rate. I agree with the comments of Egon Zondi. If I were to suggest the best procedure, it would be the actual measurement of ambient dose equivalent rate at the point where the operator is working near the source. There are many types of such neutron monitors commercially available. Many years ago, we in BARC, were using neutron "rem counters". The present day dose equivalent monitors are only adaptations of the rem counter. These rem counters can be used from thermal to 15 MeV neutron energies with an accuracy of about plus or minus 20%. I have stated above my calculation above is only approximate since I have assumed average energy. A more accurate computation requires a knowledge of the neutron spectrum and QF weightage has to be done for the entire spectrum. Re: the point about detector-person distance, this can be taken into account by placing the monitor at the position of the person. A good exercise for Muhammad Mansy will be experimental verification of the calculation above with a rem counter/ambient dose equivalent neutron survey meter. Hope this helps. Mr Muhammad Mansy may please give his feedback.

Thank you Sankaran and Egon for your detailed answers and efforts in this discussion.