A. P. Sorokin

Analysis of the fast reactors' fuel-rod bundle flow resistance

The authors analyze experimental data on flow resistance in wire-spaced fuel rod bundles. The formula is based on the data for the multirod bundles or for the rod bundles with peripheral pattern. Figures show the schematic cross section of the fast reactors fuel assemblies, and flow resistance factors in the cells of an infinite triangular lattice of wire-wrap spaced fuel elements. A more reliable generalizing formula than those currently available was obtained for the flow resistance in wire-wrap spaced fuel rod bundles. The formula is valid over a wide range of parameters being measured including close-packed lattices.

Specific features of the development of temperature fields in the system of interacting deformed fuel assemblies of the cores of fast reactors

In the fuel assemblies (FA), temperature fields form under the action of nonuniform energy release of the fuel elements, their radiation-induced deformation (shape change) during the operation period, the interchannel mass transfer and exchange of mass, momentum, and energy and interpacket heat exchange. The effect of interpacket heat exchange on the distribution of temperature in the fuel assemblies located in the peripheral region of the core has been demonstrated by several investigators. These studies have been conducted without considering the effect of the interpacket flow of the coolant and the changes occurring in the shape of the fuel assemblies on the temperature field of the sheaths in the system of the interacting FA which is important for carrying out a thermomechanical analysis. This paper deals with a study of the effect of the aforementioned factors on the temperature fields in the system of the interacting fuel assemblies.

Convective and diffusive interchannel transfer in fuel assemblies of a fast-reactor core

Diffusive interchannel heat transfer in liquid metals, compared with water or gaseous coolants, comprises a significant component of transfer due to heat conduction by the fuel elements. Other important questions are the specific details of momentum transfer compared with heat transfer, the rate of heat transfer between cells due to heat conduction by the fuel elements, and the effect of bundle deformation on the transfer rate. Systematic research on interchannel transfer that has been carried out at F~I has enabled us to answer these questions and obtain a complete system of coefficients for solving the equations of macrotransport in the channel approximation [8]. This paper presents the results of that research and gives a comparison of them with experimental data and the relations derived by other authors. Research has shown [7-9] that interchannel transfer of mass, momentum, and energy is an important factor in the formation of velocity and temperature fields in reactor fuel assemblies. CONVECTIVE INTERCHANNEL TRANSFER Central Region of the Fuel Assemblies As experimental research carried out on the basis of the electromagnetic method of measurement of local coolant flow rates in liquid-metal-cooled bundles has shown [5], transverse mass transfer with respect to the height of the gaps of the fuel elements obeys a periodic (sinusoidal) law (Fig. la). This accounts for the periodic (sinusoidal) behavior of local interchannel transfer coefficients with respect to the height of the gaps between fuel elements. The local interchannel transfer coefficient, which represents the mass flow from the i-th to the j-th channel through a unit length of gap, referred to the longitudinal flow rate of coolant in the channel, is described with sufficient accuracy by a single harmonic m

An analysis of hot spots in fast reactor fuel assembly cores

One of the most important safety requirements is avoiding the loss of fuel element intergity and fuel melt. The necessary condition for this is staying within the limit of the maximum clad and fuel temperature. Staying within the limit not only for the average values of parameters, but also during the greatest possible deviation from the nominal values, requires determining these deviations, investigating their influence on the temperature fields in the fuel assembly, figuring them into the thermal-hydraulic calculations for controlling the operating regime of the reactor core and into the calculation of the FBR fuel assembly operating capability. This paper analyzes such thermal-hydraulic deviations and provides computerized formulations of the reactor lattice parameters for the case of hot spots in the core-fuel configuration.

Temperature field in heat-liberating piles of fast reactors

The accuracy of thermophysical calculation of heat-liberatlng piles (HLP) of fast reactors is determined to a considerable extent by the accuracy of the initial calculational model of the fleld-element bundle and also by,taking correct account of all the factors influencing the formation of a temperature field in the bundle. The calculation is usually conducted for nominal parameters of the fuel elements and the HLP, under the assumption of a regular arrangement of fuel elements at the points of a regular triangular grid. For this model, the distribution of the heat-carrier flow rate over the individual channels, the heating of the heat carrier in the most stressed channels, and the corresponding temperature of the fuel-element casing are calculated for this model.