The results of computational and experimental validation of the design of the emergency heat-removal system for BN-1200 are presented. The results of experimental studies performed on a water model of the V-200 integrated stand are presented, and the planning of experiments on the TISEI stand for validation of the efficiency and operating regimes using the BURNA, GRIF, and Flow Vision computational design codes is discussed.The removal of the residual heat released in the core remains a pressing problem, increasing as reactor power increases. In accordance with current trends, the problem here is to meet during emergency cool-down safety requirements that are more stringent than those determined in OPB 88/97, viz., for advanced designs the probability of serious damage to the reactor core must not exceed 10 -6 per reactor per year.At the same time, the current trends in nuclear power are toward validated reduction of safety costs with a positive economic outcome. For the sodium-cooled fast reactor BN-1200, the technical task is to lower the cost of a power-generating unit to the cost of VVER at the same power. The cost reduction of the emergency heat-removal system can make an appreciable contribution to the reduction of the cost of a power-generating unit.The heat-removal system based on heat-exchangers immersed in the coolant in the first loop and final heat removal accomplished through an intermediate loop into the environment was chosen at the design variant for BN-1200.Accurate modeling of heat and mass transfer on small-scale models with a natural coolant is impossible because the similarity criteria are not satisfied: Peclet number (Pe = wl/a), Reynolds number (Re = w/ν), and Froude number (Fr = w 2 /gβΔTl) [1]. The experimental facilities and research are expensive because large-scale models with natural coolant are used. The main objective of the computational and experimental validation is to confirm the serviceability of the system passively removing the residual heat release. For routine functioning, it must keep core operation within safe operating limits according to the maximum admissible temperature of the fuel-element cladding and the reactor vessel, which are set during the validation of the corresponding materials in the design [2].Two stands were developed to study the thermohydraulic processes in the new-generation emergency heat-removal system: V-200 and TISEI. The integrated V-200 stand takes account of the characteristics of the in-take arrangement of BN-1200 as a whole on a 1:10 scale [3]. The TISEI stand with a 1:5 scale ensures more accurate geometric similarity of the

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Предложена математическая модель процесса разрушения тканевых композиционных материалов, основанная на использовании метода асимптотического осреднения и конечно-элементного решения локальных задач на ячейках периодичности. В качестве критерия прочности матрицы используется модифицированная модель Писаренко-Лебедева, а критерия прочности армирующих нитейдвухуровневая модель повреждаемости пучка моноволокон. Модель позволяет прогнозировать процесс распространения микроразрушения в ячейке периодичности композита при изменении действующей на композит системы нагрузок. Приведены численные примеры, демонстрирующие возможности разработанной модели при применении ее к численному исследованию процессов микроразрушения тканевых композитов.

An approach to validating the safety of sodium-cooled fast reactors which was devised in Russia as these reactors were developed is presented. The collection of computational codes used for validating safety is described. Attention is focused on a group of codes which are intended for systematic analysis of all phases of a serious beyond design-basis accident resulting in the destruction of the core. The domestic approach to the analysis of the consequences of serious accidents is compared to the methodology developed abroad. Up to now, the code system under development has made it possible to perform the main tasks of design validation, but the chosen path of development, where a separate computational code was developed for each individual problem, is not exhausted. On the one hand, the complexity of the mathematical models drawn into the analysis has increased significantly, while on the other hand significant advances have been made in the development of computational technologies. The answer to the present situation is the tendency observed in the world nuclear community toward developing integrated codes and multipurpose computational systems. In Russia, this tendency is realized in the program for the development of a new generation of code systems within the framework of a special federal program.The approach to the safety analysis of sodium-cooled fast reactors was formed on the basis of domestic and foreign experience accumulated in the process of designing and operating these reactors. One of the main tools for solving this general problem is a set of computational codes used for validating the characteristics of fast reactors and analyzing their changes under different initial events, including accidents. As numerical modeling methods develop and the computational hardware improves, computer codes play an increasingly important role in the validation process. In practice, both the code developers and code users rejected, first and foremost, the list of applied problems whose solution required complete and comprehensive validation of the reactor parameters and safety. Such lists are, on the one hand, universal and applicable for definite types of reactors, for example, sodium-cooled fast reactors, while on the other hand the characteristics of each particular design must be taken into account in them. The lists are compiled by the joint efforts of design organizations and oversight agencies in application to each design being developed and include a list of disruptions of normal operation, design-basis and beyond design-basis accidents.The task of safety analysis is to show that the operational limits and safety criteria formulated in the regulatory documents are not violated in all initial failures and accidents included on the list.Since the lists of initial disruptions and accidents cover a large number of phenomena on different levels, a set of different codes differing by the spectrum of the phenomena being modeled, the degree of detail of the modeling and, in consequence, the computational accura...

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