A. Ya. Kramerov

RBMK safety investigation for accidents initiated by partial breaks in the circulation loop

The present status of RBMK safety investigations for accidents initiated by partial breaks in the head part of the circulation loop is analyzed. The analysis shows that the RELAP code, which is the main tool for simulating such accidents, is inapplicable. An estimate of the maximum temperature of the fuel-element cladding in an accident situation with critical partial rupture is estimated on the basis of an analytical analysis of the thermal conductivity, RELAP5/MOD3.2 code calculations of a hypothetical RBMK-1000 accident, and taking account of the uncertainty of the experimental data accumulated in the investigation of similar PWR accidents. It is found that this temperature can exceed the applicability criteria.

Status and prospects for pressure-tube water-cooled graphite-moderated reactors

An overview is given on the 50 year experience in the development of pressure-tube water-cooled graphite-moderated reactors (WCGMR) in Russia and operation of NPPs with RBMK-type reactors. Advantages and shortcomings of WCGMR are pointed out. Also described is the systematic upgrading effort for some RBMK systems and components meant to enhance their reliability, safety and lifetime. A possibility of further improvement of the economic performance of NPPs is discussed. The paper gives technical features of the MKER-800 reactor plant designed to replace power units with RBMK-type reactors at the end of their lifetime. Pressure-tube and pressure-vessel reactor concepts adopted in the Russian nuclear power system are mutually complementary, which is an important merit and advantage of such a model of power development.

Analysis of the accident in the second power-generating unit of the chernobyl nuclear power plant caused by inadequate makeup of the reactor cooling loop

The accident in the second power-generating unit of the Chernobyl nuclear power plant on October 11, 1991 was the result of unauthorized connection of the TG-4 turbogenerator, which was shut down for repairs, into the grid (in the off-design asynchronous engine mode), and this resulted in a serious fire in the machine room and subsequent failure of systems which are important for safety and which ensure the design mode of reactor cooling: These were primarily failures of the feed and emergency feed pumps and failure of the BRU-B control valve, which regulates steam release during cooling.

On the development of fuel assemblies

Abstract Investigations of RBMK fuel assembly models of various geometries were carried out to develop fuel assemblies with reduced uranium dioxide temperatures and decreased void reactivity effect. The investigations include thermophysical experiments and neutron and physical calculations. The influence of the following factors on bundle thermophysical characteristics was studied: number and diameter of fuel elements located within a channel 80 mm in diameter, relative arrangement of fuel rods within the channel tube, gap between the fuel rods and between the peripheral row of fuel rods and the channel tube, peaking factor of power distribution over the fuel assembly radius, and the use of heat exchange intensifiers. The RBMK fuel assembly of ultimate geometry (18 fuel rods 13.6 mm in diameter within a channel 80 mm in diameter) was compared in neutron and physical calculations with three variants: 36 fuel rods 10.2 mm in diameter within a channel 80 mm in diameter and 18 fuel rods 14.9 mm in diameter without and with holes in the pellets within a channel 86 mm in diameter. It was determined that the fuel assemblies with 36 fuel rods within the ultimate channel 80 mm in diameter give no essential reduction of the void reactivity effect and have little effect of the critical power increase when heat exchange intensifiers are used. Therefore efforts are being made to increase the channel diameter and use the 18-rod fuel assembly with axial holes in the fuel pellets.

Thermal stresses in reactor structures

Abstract The conditions in which thermal stresses develop in reactors are considered, their magnitude is estimated, and the extent of the danger they present. The influence of the shape of the fuel elements on the temperature gradient and on the magnitude of the thermal stresses is analysed; recommendations are made for decreasing the harmful influence of temperature stresses. The methods of elasticity theory used in the calculation of the thermal stresses have important limitations. In many eases, when one is estimating the size and extent of the danger arising from the thermal stresses superimposed on mechanical stresses and also in the search for measures to decrease them, one must take into account the additional influences of yield, creep, initial breakdown and of microscopic processes.

The selection of the optimum parameters for an atomic electric generating station

A system of equations has been derived, giving a set of optimum parameters for an atomic electric generating station, which insure minimum cost of the electrical energy produced. It is assumed that the layout, materials, and type and elements of equipment have been previously selected, and that the problem is one of finding the optimum numerical values of the constructional and operational parameters of the elements of the station. Quite general approximate relationships are used to express the costs of the elements of the installation as a function of the parameters being sought. Attention is paid to the mutual relationships which exist among the parameters, through the equations which describe the processes going on in the station, as well as to the limitations which are imposed by the need for reliable operation.The system of equations derived may be used to check the optimum parameters of various projected two-loop installations with nonbolling reactors, employing the maximum allowable fuel element temperatures. Examples are given to show the importance of the independent problem of expressing some of the optimum parameters in terms of other parameters obtained from individual equations of the system.

Thermal stresses in reactor constructions

Conditions for the appearance of thermal stresses in reactors are investigated; also their magnitude and the danger they create are estimated. The influence of the form of the heat-generating elements (HGE) on the temperature drop and magnitudes of thermal stresses is analyzed; recommendations are given with the aim of decreasing the harmful effect of thermal stresses.The methods from the theory of elasticity employed in the calculation of thermal stresses have significant limitations. In many cases when estimating the magnitude and degree of danger created by the thermal stresses, when combining such stresses with mechanical stresses, and also when seeking a way to decrease them, other effects such as fluidity, creep, initial breakdown, and microscopic processes must be taken into consideration.