Series: Nuclear and Reactor Constants

since 1971

Русский (РФ)

ISSN 2414-1038 (online)

DOI: 10.55176/2414-1038-2021-4-106-120

Authors & Affiliations

Kuzina Yu.A., Sorokin A.P., Denisova N.A.
A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia

Kuzina Yu.A. – Head of Nuclear Power Division, Cand. Sci. (Techn.). Contacts: 1, Bondarenko sq., Obninsk, Kaluga region, Russia, 249033. Tel.: +7 (484) 399-86-63; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Sorokin A.P. – Chief Researcher, Dr. Sci. (Techn.).
Denisova N.A. – Lead Engineer.


The results of experimental studies of heat transfer carried out at the State Research Center RF – IPPE on models of fuel assemblies in cores and screens of fast reactors BOR-60, BN-350, BN-600, BN-800 (the relative pitch of the fuel element simulators varied in the range 1.04≤s/d≤1.24) with natural and mixed convection of liquid metal coolants, which are the basis for the development of design codes are considered in the article.
Model fuel assemblies of fast reactors contain 37 fuel elements (fuel rod simulators) located in a triangular lattice and enclosed in a hexagonal cover. The elements were heated with nichrome heaters, which ensure a constant heat flux on the inner surface of the elements and along the length of the heat generation. The central, lateral and corner elements are rotary.
On their surface, 12 chromel-alumel thermocouples each in covers made of X18H9T steel (diameter of the cover 0.8–0.5 mm, diameter of thermoelectrodes 0.2 mm) are embedded, measuring the coolant temperature in the collectors of the models, as well as in each cell at the exit from the bundles. The spacing of the elements is carried out using wire coils; variants of smooth fuel rod simulators are also used.
Experimental studies on model fuel assemblies revealed regularities in the formation of temperature fields in fuel elements and coolant. Natural convection manifests itself in the region of low velocities, promotes fluid mixing between channels, leveling the uneven heating of the coolant in the cross section of the fuel assembly, and reduces the azimuthal non-uniformity of the temperature of the wall fuel elements in the fuel assembly.
The results of experimental studies show that the effect of natural convection manifests itself in the range of Pe<10 (Re<2000) numbers in gratings with relative steps (s/d<1.05) and in wide bundles (s/d<1.3 to a greater extent in bundles of smooth fuel rods in comparison with bundles of ribbed fuel rods. The introduction of displacers into the peripheral channels of a fuel assembly does not fundamentally change the nature of the temperature field in a fuel assembly as compared to the version of the geometry of a fuel assembly without displacers.
An attempt was made to generalize the experimental data using the Gr*Pr criterion, where Gr* is the modified Grashof number calculated from the local heat flux on the heat exchange surface and the axial coordinate measured from the beginning of heat generation.
In the dependences ΔTwmax=f(Gr*Pr2) at Ре<100 or ribbed side elements and Pe<10 for smooth side elements, two regions of ΔTwmax change in the Gr*Pr2 growth function are observed – first, ΔTwmax ncreases to a certain “limiting” value characteristic for fixed Pe, and then – a drop in ΔTwmax at large values of the parameter Gr*Pr2.
As the Peclet number increases, the “limiting” value of ΔTwmax hifts to the region of larger values of Gr*Pr2; and at Ре=370 or ribbed elements and Ре=26.5; 100 or smooth elements, the limiting value of ΔTwmax is not achieved in the investigated range of the Gr*Pr2 parameter variation. For smooth side elements, the limiting values of ΔTwmax are approximately the same for different Pe numbers and amount to ΔTwmax≈10.

fast reactor, core, fuel assembly, fuel element, liquid metal coolants, accident regines, natural convection, heat flux, heat exchange, temperature, heating, velocity, ribbing, peripheral, wall fuel element, displacer, Reynolds number, Peclet number, Grashof number

Article Text (PDF, in Russian)


UDC 536.24

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2021, issue 4, 4:10