The neutron flux density from 0.025 eV to 12 MeV has been measured experimentally in all channels of the VVR-SM core by the activation method using threshold monitors (Au, Ni, Fe. Ti, Mg, Y). Comparing with a calculation of the neutron flux density at different energy using the IRT-2D computer code showed agreement to within 5%. The distribution of the neutron fluxes and spectra in the core, which is of practical utility for radiation technologies, was obtained. A series of irradiations has been conducted and experimental dependences of the irradiation time on the channel position in the core as well as on the size of the stones for obtaining a standard light blue and dark blue color have been obtained. The irradiation conditions making it possible to lower the induced radioactivity of the minerals three-fold as a result of increasing the ratio of the fast to thermal neutron fluxes are found.The main parameters characterizing research reactors are the neutron flux density in the core and the neutron energy spectrum. The flux density of the main neutron energy groups in the experimental channels of water moderated and cooled reactors is of great interest for practical applications of activation analysis, radiation technology, as well as testing of different materials and semiconductor devices for radiation resistance of their functional characteristics. In this connection, the prime problem is finding the distribution of the neutron flux and spectrum over the channels inside the core of the reactor; the difficulty is that it is impossible to detect neutrons directly and determine their energy.It is known that the thermal neutron flux density can be determined from the (n, γ) reaction products, which are formed when a monitor is irradiated, as the difference of the indications of a counter without and with a filter. Threshold reactions are used to recover the fast-neutron spectrum. Since the differential measurements of a neutron spectrum are difficult to perform, and there is no need for them in analytical measurements, ordinarily the integral fluxes of neutrons above some threshold value are evaluated.The structural features of some irradiation channels in research reactors make it impossible to determine the neutron flux experimentally and to reconstruct the neutron energy spectrum. In such cases, these parameters are determined by, for example, solving a system of differential equations governing diffusion [1]. Several types of applied programs are available for calculating the neutron flux density. A program for calculating the neutron flux density in three-dimensional space by the Monte Carlo method has been developed at the Los Alamos National Laboratory (USA) [2]. This program is considered to be universal and is the only program that takes account of published data on neutron-physical interaction constants for neutrons interacting with different atoms at different energies as well as the energy spectrum of neutrons formed as a result of the fission of 235 U nuclei in fuel. In addition, it permits setting t...

The neutron fluxes and the intensity of γ radiation are measured in 26 channels of a VVR-SM reactor and its thermal column. The fast neutron fluxes in the channels are determined using Ni, Fe, Co, Au, and Mn element monitors with different threshold energies, together with a theoretical calculation using the MCNP-4C program. The energy distribution of the neutron flux inside the fuel assembly is obtained for selected channels around the core. The flux of neutrons with energies >1 MeV is in the range (0.5-43)·10 12 cm -2 sec -1 , depending on the location of the channel. A linear correlation is discovered between the induced optical absorption at the 215 nm line (E′ center) of SiO 2 -BaO glass and the fast neutron flux in the channels. The γ-ray intensity in the thermal channel is estimated for the reactor during operation (~38.4 Gy/sec) and 24 hours after it is shut down (~24.7 Gy/sec) using the E′ centers induced in pure quartz glasses. The observed difference in the efficiency with which oxygen defects are formed during dry and wet irradiation of glass owing to the radiolysis of water must be taken into account when developing radiation technology and during the burial of radioactive waste.Discrete determinations of the neutron flux and the accompanying γ radiation in nuclear reactors are needed in studies of the formation of radiation defects and in the development of radiation technologies. The dose rate for γ radiation is estimated customarily at 15 Gy/sec for a reactor power of 1 MW. The γ radiation is detected using ionization chambers, silicon detectors, and solid state dosimeters based on phosphate or quartz glass with germanium doping [1,2].A linear correlation between the increase in the optical density of glass in the 215 nm band (E′ center) and the fast neutron flux has been established by irradiating samples of specially pure quartz glass and nickel monitors in 10 vertical channels of a reactor operating at 10 MW [3]. Irradiation of lithium fluoride (LiF) crystals in a 60 Co device at 7.65 Gy/sec and parallel irradiation in a shut-down reactor showed that the γ dose rate was 26 Gy/sec [4].This paper reports an experimental determination, by means of quartz glasses and element monitors, of the fluxes of neutrons at various energies and γ-rays in all the test channels in the core and in the center of the fuel assemblies of the IRT-3M with 36% enriched 235 U for the operating and shut-down VVR-SM reactor.Experimental Data. The neutron flux and the γ-ray intensity were measured in six channels in the center of the fuel assembly and in 20 vertical channels in the first and second series surrounding the core (Fig. 1a) at various times after reloading the fuel assembly and with differing degrees of burnup.

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