The article examines measures involved in severe accident management for NPP-2006 reactor plants based on the Leningrad NPP-2 prototype and their substantiation using deterministic analysis and special computer codes. The results obtained during the analysis should be considered in the development of severe accident management guidelines for these reactor plants in terms of the composition and priority of management actions.

The article considers the 3D modeling of a gaseous medium mixing during a NPP severe accident using the STAR-CCM+ commercial code. The mathematical model of a passive hydrogen recombiner is described. The results of calculating a severe accident scenario and its comparison with the results obtained using the KUPOL-M code of lumped parameters are given. An analysis of possible hydrogen combustion and detonation is carried out.

A jet-vortex condenser (JVC) used as part of the confinement system (CS) serves for limiting the pressure in sealed rooms to the maximum admissible value for which the civil structures of NPPs equipped with VVER-440 reactors made according to the V-230 and V-179 projects are designed under the loss of coolant accidents' conditions. Fitting the Kola NPP’s first stage equipped with the VVER-440 reactor (made according to the V-230 project) in combination with taking certain other measures opened the possibility to extend the NPP service life. To substantiate the possibility of extending the service life of NPP power units, their safety will be assessed using an analysis model. The article presents the results from such modeling aimed at analyzing the performance of the jet-vortex condenser that is used as part of the confinement system for limiting the pressure in the sealed rooms down to the maximum admissible value for which the VVER-440 reactor-based NPP civil structures are designed with due regard to loss of coolant accidents (LOCA). The jet-vortex condenser operation model incorporated into the KUPOL-M computer code was verified against the experimental data obtained on the test facility installed at the All-Russian Research Institute for Nuclear Power Plants Operation (VNIIAES). A comparative analysis has shown that the calculated and experimental data are in fairly good agreement with each other. The KUPOL-M code with the jet-vortex condenser model integrated into it was used for calculating the parameters of medium in the system of sealed rooms in modeling the large break LOCA and small break LOCA conditions for the Kola NPP’s first stage, the distinctive feature of which is that it contains a JVC in the confinement area. The calculation results have confirmed the JVC efficiency under the conditions of a wide range of loss of coolant accidents. At present, the JVC has been put into use in the Kola NPP units 1 and 2, in the Novovoronezh NPP units 3 and 4, and in the Kozloduy NPP units 3 and 4 in Bulgaria.

One of the problems of accidents in NPP with VVER is the evaporation of the primary coolant in the containment. Boric acid H3BO3 can ingress into the containment together with steam and crystallize, independently or in the form of salts, on heat-exchange and other surfaces of equipment placed in the containment, thereby lowering the efficiency of passive systems removing heat from the containment and the removal of hydrogen. The results of an experimental investigation of the deposition of insoluble aerosols and salts of boric acid on heat-exchange surfaces of the passive heat-removal system and the surfaces of autocatalytic hydrogen recombiners during serious accidents in NPP are reported. It is shown that the impact of crystallization of boric acid and its salts on the operability of these systems is very small.

Domestic and foreign data on free convection along vertical heated plates and tubes are analyzed. It is shown that the difference between the heat-emission coefficients during free convection can reach 70%. Heat emission along vertical tubes, especially in turbulent air-flow regimes, has been little studied. Meanwhile, a regime close to free convection complicated by heat exchange, for which a correction is made in the closure relations pertaining to heat emission for computational codes, is realized in the passive heatremoval system for the containment. Direct numerical modeling of free convection along vertical plates and tubes is performed.

As a result of catastrophic events on the nuclear power plant “Fukushima” the European organizations on regulation of nuclear power (ENSREG) initiated wide-scale measures for complex designs revision of already operating and under construction European and Russian NPPs. Inspection was made about resistance of power units to external influences of the natural character, being accompanied by multiple failures of safety systems. Within these works stress tests for constructed power units of LAES-2 and the Baltic NPP were executed. The structure of these checks included the settlement analysis of a condition of NPP at accident with loss of all AC power supply sources which results are presented in report materials. Accident calculations with a full blackout were executed on the best-estimated heat-hydraulics code KORSAR/GP for justification of power unit preservations in the intact condition within 72 hours from the accident beginning by means of SG PHRS. The system is developed for feed of the SG PHRS tanks and the fuel pool for working capacity extension the SG PHRS and power unit preservation in a stable condition more than 72 hours from the accident beginning. Use of system for feed of tanks the SG PHRS and the fuel pool allows to increase significantly resistance of the NPP to external influences of the natural character and to increase time of preservation of the blackout power unit in a stable condition more than 5 days.

Pressing questions concerning experimental modeling of thermohydraulic processes in NPP containment are examined. Domestic and foreign stands are briefly reviewed. It is shown that the KMS large-scale stand is necessary in order to obtain preliminary experimental data on the three-dimensional thermohydraulic processes occurring inside NPP containment and to verify modern computational codes.

The description of the experimental stand for studying of a local and average thermolysis is provided. The analysis of existing literary data on free convection at vertical warmed plates and pipes is carried out. It is shown that at free convection according to a number of works 70% can reach distinction of coefficients of a thermolysis. Rather low-studied question is the thermolysis at vertical pipes, especially at a turbulent mode of a current of air. Authors of article conducted research of free convection at vertical warmed and cooled single pipes under boundary conditions of the third sort for the subsequent validation of three-dimensional settlement codes.

The EUR-required passive system for protection against unanticipated accidents in nuclear reactors has been developed. This system and the computational results obtained using three-dimensional computer codes are described. The experimental data are compared with numerical experiments performed using three-dimensional hydrodynamic codes and the KUPOL-M lumped-parameter code. The experiments made it possible to refine the closure relations used in integral codes.

The main technical characteristics of the KMS stand as a large-scale model for NPP containment with VVER are presented. It is shown that this stand can be used to study three-dimensional heat and mass transfer processes in the containment atmosphere and to test systems and equipment used in NPP with VVER under close to anticipated and unanticipated accident conditions. In the future, with appropriate development work and additions to the stand it will be possible to study the rate of removal of aerosol particles having different chemical properties, perfect new designs of catalytic hydrogen recombiners and test equipment with the limiting parameters of the containment atmosphere.

Specialized three-dimensional codes and lumped-parameter codes are used to validate the effectiveness of the containment heat-removal system implemented in the AES-2006 design at Leningradskaya NPP-2. Great importance is attached to code verification in modeling the new system. To verify the software, experimental data on the heat emission and air speed under the heat exchanger and on the symmetry axis of the model of the hermetic volume were obtained under conditions as close as possible to natural conditions. The laser knife method was used to obtain a picture of the convective air flow established and the flow rate and temperature of the cooling water were measured. In spite of the high temperature (it can exceed 100°C) in the interior of the SMK stand, the air flow velocity was measured at 18 points under the heat exchanger.

The main elements of a mathematical model and test results for the specialized code PGS-TK to be used for calculating the flow of a multicomponent steam–gas mixture in the containment of a power-generating unit in NPP with passive heat-removal and volumetric and surface condensation of steam are presented. It is shown that owing to the modern mathematical models and numerical methods the code makes it possible to perform accurate calculations of the containment temperature and pressure for an unanticipated accident without using any adjustable parameters or specific experimental approximations.

The equipment of LNPP-2006 for severe accident’s managements includes the device of core melt localization–core catcher. By analogy with core catcher of the Tainvan NPP in China, the LNPP-2006 core catcher has been developed based on the crucible concept that combines corium retention inside the vessel (passive water cooling of metal surfaces that form the boundary of the corium localization zone) and the control of physical and chemical properties of the corium (by the use of sacrificial material). By present time design works in which the basic design parameters of the core catcher are defined have been performed. It was allowed to execute the verification calculations of core catcher processes taking into account behaviour of reactor’s cavity materials. For the analysis of core catcher severe accident’s behaviour code HEFEST-CC was used, for its adjustment three-dimensional hydraulic codes were applied. As a result of this analysis the quantity of the additional materials, which supplying from reactor’s cavity during concrete destructions under the influence of radiation was adjusted. The cooling ability of a passive water system of a melt surface has been confirmed. Introduced paper is devoted the analysis the formation dynamics and parameters of the molten core bath inside the core catcher vessel, to destruction the fracture dynamics of a thermal protection and response time the water passive feeding system on a melt surface, to definition the crisis store before heat exchange on the water-cooled vessel.

The course of an unanticipated accident with serious damage to the core is characterized by the emission of a large quantity of radioactive materials into the first loop and then into the space inside the protective shell. The majority of the fission products enter the protective shell in the form of radioactive aerosols, whose leakage largely determines the consequences of a serious accident. This makes it necessary to model the dynamics of multicomponent aerosols taking account of the coagulation, change of their composition and mass as a result of the condensation-evaporation and dissolution of the gaseous components as well as the precipitation of the particles on different surfaces. A program for calculating the aerosol kinetic during the course of a serious accident is examined.

During a loss-of-coolant and core meltdown accident at a nuclear power plant fission products enter the atmosphere inside the protective shell. One of the main factors determining the effect on the environment, the dose loads in personnel and the general public in the first days after the accident is radioactive iodine. To evaluate the mass of the radioactive iodine entering the environment, it is necessary to calculate its concentration in the protective shell. The model developed to describe the dynamics of iodine and the computational module in the integrated code make it possible to perform numerical modeling of the behavior of iodine in the protective shell in the case of accidents at nuclear power plants with VVER reactors. The model makes it possible to take account of the effect of physicochemical processes on the formation of volatile forms of iodine as well as the dynamics of thermohydraulic parameters on the iodine distribution. Verification showed an adequate description of the dynamics of the volatile forms of iodine in the protective shell.

The modeling of the operation of a passive condenser based on a numerical solution of three-dimensional equations of hydrodynamics is examined. Questions associated with correct modeling of turbulent transport under free-convection conditions are examined. A model taking account of the dynamics of a condensate film and the conditions of heat and mass transfer on its surface is proposed for surface condensation. The results are used to develop recommendations for closure relations used in point codes with whose help design validation of a passive system removing heat from beneath the protective shell is performed.

The project of nuclear station LNPP-2 with a reactor power plant VVER type by electrical power 1200 MVt involves a number of new design solutions to increase of parameters of safety. The passive containment heat removal system and heat removal system via steam generators is including of number of such solutions. Passive heat removal system via steam generators (PHRS/SG) is assigned for remove of residual heat of reactors core to final heat absorber (atmosphere) through a secondary circuit at DEC accident. The system PHRS/SG duplicates cooling-down system via SG to final heat absorber in case of impossibility of realization of its design functions. Containment heat removal system (PHRS/C) is assigned for remove of residual heat from containment in accidents with heat-transfer emissions from primary circuit. PHRS/C duplicates functions of a spray system to reduce of pressure under containment in case of spray system failure. In the substantiation of passive security systems the complex in SPbAEP of computational and experimental analysis was executed, the main results of which are shown in the present report.

Equipping new-generation nuclear power plants with passive means for controlling unanticipated accidents is one of the most promising directions for increasing safety, which is being implemented in the AES-2006 design for the site of the Leningradskaya nuclear power plant. An urgent problem is to obtain experimental validation of the passive system for removing heat from the protective envelope during unanticipated accidents with loss of coolant from the first loop in the case where the active systems fail. A particularity of the system is its state of constant readiness. The system functions with natural circulation of the coolant in both loops. Considering the importance of the passive heat removal system for ensuring the localizing properties of the protective envelope, OKBM Afrikantov has developed a large-scale stand and performed experimental investigations on validation of the effectiveness and serviceability of the cooling loop of a passive system for removing heat from the protective envelope.