A channel for neutron flux measurement

摘要

The design principles of a channel for measuring fluxes of thermal neutrons in a nuclear-reactor control and protection system are considered. The channel consists of a fission chamber, a set of boron-containing gas-filled ionization chambers, and an electronic unit that ensures a counting rate of fission pulses of up to ～1×10{sup}6 s{sup}(-1) It is shown that the widest linear range is achieved in channels equipped with a fission chamber and a gas-filled ionization chamber with electrodes coated with natural boron. The channel allows for measurements of the thermal-neutron flux density in a range of 0.5 to 5×10{sup}11 cm{sup}(-2) s{sup}(-1).