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[en] Equations of neutron dynamics and reactivity are analyzed within the framework of an adiabatic model for prolonged transient processes in the core. Calculations are presented for the adiabatic model, point approximation of the adiabatic model, point model of the reactor core, quasistationary approximation, delayed neutrons in the form function, eigenvalue with the absorption operator, transient processes, reactivity, and reactivity and the state space of the core. Based on experimental data, the transient process in the core can be divided into stages: perturbation, stabilization with redistribution of the delayed-neutron sources, an established transient process with established form functions, and distribution of delayed-neutron sources and constant reactivity. The indications of reactimeters, based on formulas of the inverse solution of kinetic or in-hour equations and calculated with respect to local neutron flux density, are valid only at the stages of an established transient process. For other stages, a coordinate- and time-dependent quantity can be introduced as a parameter of the exponential of the amplitude factor. The reactivity can be determined for all stages of the transient processes by using the equations presented, and an estimate of the local values can be introduced. 5 refs

[en] Programs are described which are intended for processing data obtained in physical experiments at critical assemblies and laboratory stands. The calculation algorithms and the texts of: KRIT-1 program, used for the statistical processing of measured data for calculation of moderator critical heights and changes in the reactivity rate as a function of the core flooding with moderator; KRIT-2 program, intended for processing the conversion register readings and determining activity drop parameters of the monitoring detectors and relative activation of the detectors; programs for drawing core cartograms; and also NORMIR program used for processing measured data on the neutron flux distributions in radial and azimuthal directions, are presented. The programs are written in BASIC and FORTRAN and are intended for BESM-6 computations. The programs have been used for processing experimental data obtained at a full-scale assembly simulating the WWER-1000 reactor core

[en] The description of mathematical statement of a task for definition the spectral characteristics of neutron fields with use developed in RRC KI unified activation detectors (UKD) is resulted. The method of processing of results offered by authors activation measurements and calculation of the parameters used for an estimation of the neutron spectra characteristics is discussed. Features of processing of the experimental data received at measurements of activation with using UKD are considered. Activation detectors UKD contain a little bit specially the picked up isotopes giving at irradiation peaks scale of activity in the common spectrum scale of activity. Computing processing of results of the measurements is applied on definition of spectrum parameters for nuclear reactor installations with thermal and close to such power spectrum of neutrons. The example of the data processing, the measurements received at carrying out at RRC KI research reactor F-1 is resulted

[en] Basic principle and requirements, stipulated by the nuclear safety rules(NSR) for NPP reactor facilities, introduced from September 1, 1990 are considered. The new rules are characterized by safety concept fundamental change, conditioned by the Chernobyl accident. The rules contain a safety target orientator, expressed in probabalistic terms, as well as new requirements, forming deterministic criterium of acceptable safety level, the main feature of which is communication of the inner safety properties to a reactor

[en] Analysis of experimentally recorded a behavior distributed of neutron flux density (DNF) in RRC KI required more detail consideration of methodical base and approximations, which are used for the description of the behavior. Mathematical models and main features of transients are based on adiabatic approximation. This approximations and term 'reactivity' used for description of DNF transients in reactor cores and volumes containing fissile materials are discussed. However the value of reactivity have large errors for reasons of approximations: methodical, measuring and total effect models. (author)

[en] The paper presents main topics and problems which arise during creating of regulatory requirements for post accident situations at NPP

[en] The method proposed for estimation of parameters thermal and epithermal parts of energy distribution of neutrons is described. The method based on application of activation measuring with use of unified composition detectors (UCD) and samples of fuel. The method is applicable for definition of neutron spectrum parameters and velocities of division in fuel of nuclear installations. Theoretical bases and the description of a method, expedients of manufacturing and calibration for the detectors, the experimental data, carried out in RRC KI are given and processing of experimental data, and also. The parametric model of a spectrum constructed on the basis of Westcott's formalism is described. The parameter of stiffness is entered and its role for temperature of neutron gas, spectral coefficients of isotopes of detectors, the transition area thermal and epithermal parts of neutron spectra is observationally appreciated. It is offered to confirm the found results by calculations with use of MCU Monte Carlo code

[en] The report is devoted to the 55-year experience in operation of the first Soviet nuclear reactor F-1, started-up on December 25, 1946. The material science, physical, reactor and medicinal studies, accomplished at this reactor, are enumerated. Various studies are continued presently at the F-1 reactor. It has the license of the Gosatomnadzor for operation and is certified as the calibration standard for the neutron flux density

[en] Analysis of experimentally recorded a behavior distributed of neutron flux density [1,2] an RRC 'Kurchatov Institute' required more detail consideration of methodical base and approximations, which used for description of the behavior. mathematical models and main features of transients are based on adiabatic approximation [3]. This approximation and term 'reactivity' used for descriptions of DNF transients in reactor cores and volumes containing fissile materials are discussed. To know the reactivity values is necessary but not sufficient to the description this process: sometimes the values is uninformative, incorrect calculations and using of it could result to unreal presentations of localization, duration and velocity of distributed DNF transients. Value of reactivity have large errors for reasons of approximations: methodical, measuring and total effect models. (Author)