[en] The invention is aimed at improving the safety of a boiling water reactor with natural coolant circulation and increasing the reactor core power density by increasing the coolant flowrate and neutron flux stability as well as by reducing the medium compressibility effectiveness in pressure compensator in dynamic modes. The reactor vessel includes the core, draught section, heat exchangers and a pressure compensator. A part of the pressure compensator is separated by a barrier with calibrated openings possessing a limited capacity and hydrolocks. The calibrated openings in the barrier are located below the coolant level and a part of space separated by a barrier is filled with gas from external system. The part of the barrier projecting above the coolant level is adjacent to heat exchangers. In transitional regimes with the change of pressure in the circulation circuit a hydrolock facilitates to reactor vessel projection against repressing and keeps the barrier from excessive power load

[en] The fundamentals of reactor design for nuclear district-heating plants (NDHP) are examined. The main technical characteristics of the AST-500 plant, using a water-water vessel-type reactor, are given. The principal features of reactor layout and design ensuring the safety of the public when NDHPs are sited in the immediate vicinity of large towns are discussed. The main types of research and experimental work involved in planning NDHPs are enumerated. (author)

[en] Poster presentation

[en] A reactor designed for uses in nuclear heating stations is considered which is cooled by natural circulation of the water coolant at a pressure ranging from 1.2 to 1.5 MPa within the reactor. The reactor vessel is confined in a containment made of prestressed reinforced concrete and designed for total pressure within the reactor. Pool-type construction of the primary circuit equipment is provided for ensuring safe operation of the reactor and warranting the reactor to resist the maximum credible accident, i. e. rupture of the pressure vessel. Providing for an intermediate circuit reliably excludes penetration of radioactive coolant form the primary circuit into the heat consumer's circuit. (author)

[en] Full text: Russian technologies of nuclear shipbuilding are based on many outstanding achievements in science, engineering and industrial technologies, unique experience of ship nuclear reactors construction and operation. Modular ship reactors with water under pressure, developed in OKBM (more than 460 nuclear reactors), together with VVER-type power reactors are the most proven reactor technology, tested and confirmed by the successive experience of ship nuclear power plants operation. Total experience of different purpose ship reactors operation exceeds 6000 reactor years. The long-lived experience of ship plants development, construction and operation and the results of the performed R and D for the projects substantiation, process base and staff potential of Russian enterprises form the basis for the creation of highly reliable power sources for nuclear power. The report gives the description of average power VBER-300 reactor plant, developed on the basis of modular ship reactors and nuclear power stations with RP VBER-300 in groundbased and floating design. The main engineering solutions of VBER-300 reactor plant are the following: use of 300 MW(e) vessel-type PWR, the most proven in the world practice; block arrangement of the main equipment; leak-tight primary circuit; use of once-through steam generators of coiled type; cassette-type core with VVER-type fuel and reduced fuel rating corresponding to the proven technologies of VVER nuclear fuel cycle; use of passive safety systems. Safety of the nuclear power station with VBER-300 reactor plant together with realization of the reactor inherent self-protection is provided by defense in depth, it provides minimum influence upon the personnel, population and the environment. Indices of radiation safety allow the location of NPP in the immediate vicinity of the consumer. VBER-300 reactor plant design is evolutionary relative to modular ship reactor plants; increase of the reactor plant thermal power up to 850 MW is provided owing to the increase of overall dimensions at the maximum reservation of the reactor plant image and main design decisions on the reactor unit. VBER-300 reactor plant developed on the basis of nuclear ship building technology may be considered as basic for 'nuclear district heating' of Russian cities and towns. Calculations of main technical and economical indices of the plant and estimation of financial (commercial) efficiency of investment outlays show that NCP with two VBER-300 reactor plants is economically profitable and efficient. Application of VBER-300 RP for floating NPPs is extremely attractive. Economic efficiency of floating NPP is provided owing to: flexible scheme of construction at the ship building plant, low steel intensity, reduction of specific capital outlays, transportation to the consumer, simplicity of removal from operation, possibility of floating power unit utilization at the specialized enterprise. The design of floating NPP with two-loop VBER-300 reactor plant is a prospective option. At 100-150 MW(e) RP a long-term cycle of nuclear fuel operation without refueling (8 years) is provided, it allows to perform refueling, operations with spent fuel and maintenance at special enterprises - service centers. Spent fuel storage is not necessary at the floating power unit. Construction of such floating NPPs is possible at Russian ship building plants with the provision of floating power unit transportation by the Volga-Don channel to the potential consumers. So the performed design developments confirm the possibility of floating NPPs with 100-600 MW(e) power range creation, using unified equipment of VBER-300 reactor plant for electric energy, heat generation and sea water desalination. The supposed power sources have commercially attractive characteristics and high export potential. (author)

[en] Decription of basic design and schematic solutions of the VPBER-600 reactor are given. Technical basis and practical testing of the basis equipment are shown. Design principles of protection and localizing systems are presented. Safety concept based on the reactor inherant selfprotection and application of passive systems and device are formulated. The VPBER-600 reactor exceeds the best foreign facilities of the AP-600 type by the level of their safety

[en] Procedure of the research on behaviour of melt was tested by means of simple model experiments using paraffin with boric acid addition. Calculation of neutron-physical characteristics of assemblies was conducted by the PRIZMA D and MCNP programs. Analysis of the obtained experimental results allows number of detectors and their disposition to be optimized. Possibility for the generation of information on transfer of molten mass by means of small thermal neutron detectors is demonstrated