[en] The need in measuring and evaluating the subcriticality of multiplying systems without bringing them to critical state arises in reactor loading, operation of repositories, performing process operations on shut-down reactors. Two methods for reactivity measurements are proposed, which have not found a wide use. Nevertheless, they provide a high accuracy of measurements, simple measurement procedure and, hence, require minimum technical means and equipment. Therefore, they may prove to be more advantageous in comparing with the methods extensively used in practice, such as methods of source multiplication. (author)

[en] A number of nuclear physics design issues concerning Accelerator Driven Molten-Salt Reactor based on the so called ATW concept proposed by Los Alamos are discussed. General description of concept using internal moderation with graphite block is presented. Burn-up, salt processing and safety criteria (reactivity temperature coefficients and kinetics parameters) are presented for different spectra (graphite to salt ratio) and an optimal variant of the blanket with non-positive temperature reactivity coefficient is provided and results are discussed. (author)

[en] The classical notion of core reactivity is discussed, and various ways of this notion modifications are shown. Various methods for reactivity measurements are commented. The causes of significant disagreement between the calculation and experimental data on scram system efficiency are given, a suggestion on changing the interpretation of the results of VVER startup experiments with scram are made. (author)

[en] Results are presented of a pulsing α-method test for determination of the VVER system subcriticality. It permits to conduct measurements of systems subcriticality under the conditions of subcritical state and large neutron background. Therefore, this method can be used for the control of a subcriticality of store of burnup nuclear fuel and decommissioned reactor. (author)

[en] The assessment of coupled reactor physics and thermal-hydraulics computations with the coupled KIKO3D-ATHLET code system is provided, from two stand-alone codes. The details of data flow in the coupling are reviewed and some selected results of the validation are described. The validated coupled system code is used in the safety analysis for VVER reactors. (author)

[en] The 5.0 version of C-PORCA code includes not only integrated pinwise calculations but also developed data handling as well as renewed refuelling option. All reactor states are unambiguously identified, the data are grouped into relevant data sets. On the basis of identifiers all data can be stored in any data base form. The refuelling handling part of the new code makes possible numerous new operations: assembly rotations, single movement of assemblies, insertion of fresh absorbers and DPZs, fixing of factory numbers. Burned fuels from previous reactor states can be easily handled and used in any refuelling. (author)

[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] High Pu isotopes and minor actinides occur in contemporary reactors only in the very small amounts. It is shown how to predict operational safety characteristics and how to guarantee safe effective results using estimated errors of basic neutron data for the minor actinides and Pu. It is also shown how to use additional information (e.g. measurements with oscillating samples or irradiation of ampoules in the special reactors) and data from the beginning of operation to enlarge the precise of prediction. Because the problem is reduced to the calculation of derivatives and minimization of functionals, mathematical methods for the solutions are also discussed. (author)

[en] The sections of the meetings were: Spectral and Core Calculation Methods; Core Operation and Fuel Management; Core Monitoring, Surveillance and Testing; Safety Issues; Neutron Kinetics and Reactor Dynamics; Reactivity Evaluation, High Subcriticality; Criticality Safety and Spent Fuel; Spent Fuel Transmutations. All 54 papers of the meeting were indexed and abstracted separately for the INIS database. (R.P.)

[en] A brief review of the codes available for the neutron-physical calculations of the heterogeneous nuclear reactors and critical assemblies is presented. These programs can be used for simulation of the reactor part in the accelerator Driven Transmutations Technology (ADTT) for Molten Salt Reactor Project. The programs are described that do not use empirical coefficients typical for the existing engineering program complexes intended for the calculation of thermal reactors. This is particularly important for the calculation of the non-standard situations in reactor as well as for designing the new reactor concepts. (author)