Method for measuring dose-equivalent in a neutron flux with an unknown energy spectra and means for carrying out that method

Disclosed is a method for measuring the dose-equivalent for exposure to an unknown and/or time varying neutron flux which comprises simultaneously exposing a plurality of neutron detecting elements of different types to a neutron flux and combining the measured responses of the various detecting elements by means of a function, whose value is an approximate measure of the dose-equivalent, which is substantially independent of the energy spectra of the flux. Also, a personnel neutron dosimeter, which is useful in carrying out the above method, comprising a plurality of various neutron detecting elements in a single housing suitable for personnel to wear while working in a radiation area