No abstract

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.

No abstract

Calculations of emergency situations occurring during the conversion of the VVR-SM reactor at the Institute of Nuclear Physics of the Republic of Uzbekistan are presented. Results are presented for the following situations: motion of a compensating rod (two cases), loss of electric power, and obstruction of a fuel assembly. It is shown that safety is maintained for all the emergency situations which were analyzed and the behavior of all zones which were analyzed is actually identical. The computational results will be used for the conversion of VVR-SM to low-enrichment fuel.This article presents the results of calculations and analysis of emergency situations for a core fully loaded with IRT-3M or IRT-4M fuel assemblies with high-and low-enrichment uranium, respectively, and 16 IRT-3M + 4 IRT-4M fuel assemblies. Four emergency situations are examined: extraction of an automatic rod, ejection of a compensating rod, loss of electric power, and obstruction of a fuel assembly.The VVR-SM core contains six tubular IRT-3M and IRT-4M fuel assemblies, containing tubular coaxial fuel elements with a square cross-section and 1.4 and 1.6 mm thick walls, respectively. The fuel element cladding, which is 0.45 mm thick, consists of the alloy SAV-1. A channel with a control rod or an experimental channel with the outer diameter 28 mm and inner diameter 24 mm is placed in a fuel assembly.The neutron-physical and thermohydraulic methods of calculation and calculation of fuel burnup and fuel assembly power were used, in turn, for thermohydraulic calculations of a stable state and emergency situations [1]. The emergency situations were calculated using the PARET program and a two-channel model [2]. The hottest and average fuel plates (first and second channel) and the associated water flows were chosen for the calculations. Since this program makes it possible to calculate only one type of fuel, which has definite geometric parameters and a composition, the data for a transitional core were obtained separately for IRT-3M and IRT-4M, after which the energy release was summed [3]. The following assumptions were made: accidents occur at the nominal maximum power (10 and 11 MW), the reactor stops at power 12 MW, the time delay between an excursion above maximum power and the lowering of a compensating rod into the core is 0.1 sec for both types of fuel assemblies. The reactor is stopped by lowering three emergency protection rods from the completely extracted position and six compensating rods from the critical position. The emergency protection rods are inserted to a distance of 0.6 m in 0.5 seconds, and a compensating rod is inserted completely at the constant velocity 35 mm/sec. For transient processes caused by an increase of reactivity, the three main pumps continue to operate after the reactor stops and stop feeding the water 4 sec after power loss; the auxiliary pump operates continually, powered by chargeable cells, with water flow rate 100 m 3 /h.