RU2242810C2 - Fuel assembly for water-moderated water-cooled reactor - Google Patents

Abstract

FIELD: fuel assemblies for water-cooled power reactors.

SUBSTANCE: proposed fuel assembly primarily designed for use in type VVER-440 nuclear reactors has uranium dioxide mass in fuel bundle, outer and inner diameters of fuel element can between 82.24 and 190.78 kg, 6.0·10 ^-3 and 8.00·10 ^-3 m, as well as between 5.09·10 ^-3 and 6.79·10 ^-3 m, respectively, for bundle of (174 - 216) fuel elements or uranium dioxide mass in bundle, outer and inner diameters of fuel element can are between 105.44 and 179.14 kg, 7.80·10 ^-3 and 8.79·10 ^-3 m, as well as between 6.6·10 ^-3 and 7.47·10 ^-3 m, respectively, for fuel bundle of (132 - 168) fuel elements, subchannel water-to-uranium ratio being chosen between 0.69 and 2.69.

EFFECT: reduced linear heat loads and fuel element depressurization probability, enlarged power control range, improved fuel utilization.

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Description

FIELD OF THE INVENTION

The invention relates to nuclear engineering and relates to the improvement of the design of fuel assemblies (FAs) from which the core of nuclear reactors is recruited, in which water (the so-called water-cooled nuclear reactors) is used as a heat carrier and used as a heat source for power plants in power plants etc., especially in reactors with thermal power of the order of (1150-1700) MW.

State of the art

The fuel loading of reactors consists of a large number of fuel elements (fuel elements), the number of which in the active zone of pressurized water reactors (VVER) is tens of thousands. To ensure the necessary rigidity of the rod fuel rods, as well as ease of installation, reloading, transportation and ensuring the required cooling conditions, they are combined in bundles. Each bundle represents a single fuel assembly design. The number of fuel rods in a fuel assembly can range from several to several tens or even hundreds. The fuel rods in the fuel assemblies are interconnected by means of two end and more than five to ten intermediate spacing grids installed at a certain distance from each other along the height of the assembly, which ensures rigid spacing of the fuel elements when they flow around the coolant and compliance with the clearances between the fuel rods for the passage of the coolant the necessary water-uranium ratio in the cross section of the assembly.

The active zones of water-cooled nuclear power reactors VVER-400 are recruited from fuel assemblies containing a bundle of fuel rods located in a hexagonal casing (see I.Ya. Emelyanov, V.I. Mikhan, M. Solonin, etc. Design of nuclear reactors , M., Energoizdat, 1982, pp. 76-78).

A fuel assembly, as a rule, consists of a bundle of fuel rods and a frame. The aircraft frame provides for the integration and fastening of fuel rods in the assembly and their spacing. The assembly frame consists of the following main parts: the supporting rod, end grids, spacer or guide grids, longitudinal connecting elements, various types of spacers and support runners, as well as crimp bushings. Moreover, fuel assemblies from fuel rods made with a length corresponding to the length of the active zone are supplemented by the following details: assembly head, to which the upper part of the assembly frame is attached; shank assembly, which is attached to the bottom of the frame; assembly suspension - a device with which fuel assemblies are moved, installed and held in a vertical channel; TVS shock absorber - a part of the assembly, which helps to reduce the shock load when the assembly falls on the support, as well as compensate for vibrations that occur during operation of the reactor; a calibration washer - a part of an assembly designed to determine the coolant flow rate through a fuel assembly (see G.N. Ushakov Technological channels and fuel elements of nuclear reactors, M., Energoizdat, 1981, pp. 84-86).

To reduce the share of structural material in the core, fuel assemblies may not have a casing, the so-called caseless fuel assemblies, in which the fuel rod bundle is connected by spacer grids, and the assembly support grids are connected by tubes and / or corners (see I.Ya. Emelyanov, V. I.Mikhan, Solonin V.I. et al. Design of nuclear reactors, M., Energoizdat, 1982, p.77, fig. 3.10 c).

The VVER-440 reactor cassette consists of a bunch of rod backs, a hexahedral housing, a cylindrical shank, a head and a cassette frame, with the help of which the fuel rods are mounted in the cassette. The cassette frame includes hexagonal spacer grids (lower support grid, upper and middle guide grids of zirconium alloy), which are mechanically connected to each other by a central tube of zirconium alloy. The lower ends of the fuel rods are fixed in the carrier grid, and the upper ends of the fuel rods have the possibility of longitudinal movement in the guide grid with thermal expansions. The lower carrier grill is attached to the cylindrical shank of the cartridge, and the upper guide grille, respectively, to the cartridge head. Using the shank and head, the cartridge is installed in the reactor vessel (see G.N. Ushakov Technological channels and fuel elements of nuclear reactors, M., Energoizdat, 1981, p. 89, Fig. 2.8 a).

The design of rod fuel elements and fuel assemblies for VVER-440 should provide mechanical stability and strength, including in emergency conditions at high temperatures, which is complicated by the presence of powerful long-term flows of neutrons and gamma radiation. Damage to a fuel rod entails radioactive contamination of the circuit with fission products. Violation of the initial geometric shape of a fuel element can worsen the conditions of heat transfer from the fuel element to the coolant. Therefore, when developing a fuel assembly design, it is necessary first of all to take into account the possibility of increasing the ratio of the heat transfer surface of a fuel element to the active volume occupied by nuclear fuel.

A known fuel assembly of a pressurized water reactor contains a frame and a bundle of rod fuel elements with nuclear fuel in the form of uranium dioxide (see “Future fuel: Vattenfall's new approach” Nuclear Engineering International, September 1997, p.25-31). The beam of the known fuel assembly of the VVER-440 reactor contains (120 ÷ 126) rod fuel elements made with an outer diameter of 8.80 · 10 ^-3 m to 9.14 · 10 ^-3 m and having an average linear thermal load on the fuel element of 12, 92 kW / m to 15.01 kW / m. Such a fuel rod provides a relatively high level of fuel burnup in a known fuel assembly and has proven itself during operation at domestic and foreign nuclear power plants with VVER-440 reactors. However, it should be noted that in the event of an overheat of the cladding of the fuel rods that occurs when the conditions for their cooling change, depressurization and even destruction of the fuel rods can occur. The fact is that the low thermal conductivity of the oxide fuel used by VVER-440 reactors causes its high temperature during normal operation, a relatively large amount of accumulated heat and, as a result, in an accident with de-energized nuclear power plants and in an accident with loss of coolant leads to a significant heating of the cladding of the fuel rods in the first few seconds. The temperatures achieved during accidents with loss of coolant when using standard fuel assemblies are largely dependent on the initial thermal linear loads on the fuel elements. So, with a large leak of the primary circuit of the VVER-440 reactor, fuel rods with a maximum heat load by the fifth second have an estimated sheath temperature of 850 ° C. At the same time, under the same conditions, fuel rods with a load close to average are heated to (550 ÷ 600) ° C.

Experimental and computational studies show that, from the point of view of preventing the possibility of depressurization of fuel rods in relation to accidents with loss of coolant, the limiting temperatures of the shells should not exceed ~ 700 ° C. Therefore, if the maximum thermal linear loads are reduced in the VVER-440 reactor core, then a possible heating of the shells would not exceed the aforementioned temperature limit. This fundamentally solves the problem of possible depressurization of fuel rods at the initial stage of an accident with loss of coolant. In particular, this problem is aggravated by increasing the fuel burnup depth, when the fuel rod performance, even under normal operating conditions, is close to the maximum permissible.

It follows from the foregoing that in order to increase the safety level of operating and newly designed NPPs with VVER-440, it is necessary to develop rod fuel elements of a container structure of reduced diameter with an increased number of fuel assemblies in the fuel assemblies (provided that the reactor power is maintained and the water-uranium ratio of the fuel grate is close to the standard fuel assembly) , which will fundamentally solve the problem of possible depressurization of fuel rods at the initial stage of an accident with loss of coolant. In addition, when developing the modernized core of the VVER-440 reactor, it is necessary to select the main parameters from the condition of maximum preservation of the design of the core and the nuclear power plant, as well as ensuring acceptable neutron-physical and thermal hydraulic characteristics close to the standard characteristics of the core of the VVER-440 reactor, since the objective of the present invention is not to develop a fundamentally new reactor.

This approach causes certain restrictions imposed on the selection of the main parameters of the modernized core, which boil down to the following:

- the step (147 _{+/- 0.3} mm) between the axes of the fuel assemblies and the height of the upgraded fuel assemblies in the core should be the same as in the standard design of the VVER-440 fuel assemblies;

- the “turnkey” size and height of the fuel cores of the upgraded fuel assemblies in comparison with the standard design of the VVER-440 fuel assemblies should not exceed 1.5% and 2.5%, respectively;

- the diameter of the fuel rods and their number in the upgraded fuel assemblies should ensure a decrease in linear thermal loads in the fuel rods of the upgraded core;

- the decrease in fuel loading in the upgraded fuel assemblies in comparison with the standard design of the fuel assemblies of the VVER-440 reactor should not exceed 10%;

- the increase in hydraulic friction losses in the upgraded fuel assemblies in comparison with the standard design of the fuel assemblies should not exceed the available reserves for the pressure of the main circulation pump (MCP) of the VVER-440 reactor;

- the number, diameter and placement of the CPS bodies should be the same as in the standard design of the VVER-440 reactor core.

When increasing the burnup depth of nuclear fuel or to increase operational safety at a given load due to restrictions associated with the permissible temperature of the fuel and heat sink, they tend to increase the ratio of the surface of a fuel rod to its weight, which ensures a decrease in heat flux due to an increase in surface. The reduction of specific thermal loads on the fuel rods can be achieved through the use of fuel rods with a reduced diameter, namely with the diameters of the fuel rods 6.0 · 10 ^-3 m and 6.80 · 10 ^-3 m (see Bek E.G., Gorokhov V. F., Dukhovensky A.S., Kolosovsky V.G., Lunin G.L., Panyushkin A.K. and Proshkin A.A. “Improving the fuel characteristics of VVER-440 and VVER-1000 reactors by reducing the diameter of the fuel elements” , report at the conference “Top Fuel-97”, Manchester, 1997). However, since the fuel loading in the upgraded fuel assemblies is 3.4% less than that of the standard fuel assemblies for the VVER-440 reactor, despite the fact that in the upgraded fuel assemblies with fuel rods with a diameter of 6.8 × 10 ^-3 m during initial enrichment, equal to the enrichment of fuel in a standard fuel assembly, we have a fuel burnup depth of 2.1% more than that of a standard fuel assembly, this does not completely compensate for the loss in the duration of the fuel load compared to the version with a standard fuel assembly. Therefore, the following should also be added to the above limitations:

- to ensure the projected duration of the fuel loading operation, the decrease in fuel loading in the upgraded fuel assemblies as compared to the standard design of the fuel assemblies should be compensated by an increase in the burnup depth in the upgraded fuel assemblies relative to the standard fuel assemblies.

The closest in technical essence to the described technical solution is a fuel assembly of a pressurized water power reactor containing a frame, hexagonal spacing grids, in the cells of which there is a bundle of rod fuel elements with a uranium dioxide fuel core enclosed in a shell (RU 2143144, G 21 C 3/32, 12/20/1999).

The use of such fuel assemblies in the modernized active zones of the VVER-440 reactor allows, by reducing the thermal loads of the fuel rods, to expand the maneuvering range of the reactor power, increase the fuel burnup depth and reduce the likelihood of depressurization of fuel rods.

However, a comparative assessment of the costs of the standard VVER-440 fuel assemblies (diameter of the fuel rods 9.1 · 10 ^-3 m) and the upgraded fuel assemblies (reduced-diameter fuel rods) showed that the factory cost of the upgraded fuel assemblies for VVER-440 reactors increased by 18%, which is one of the reasons why such fuel assemblies have not yet found practical application.

SUMMARY OF THE INVENTION

The objective of the present invention is the development and creation of new fuel assemblies of a pressurized-water power reactor with a thermal power of 1150 MW to 1700 MW, having improved characteristics, in particular, increased safety and reliability in the operation of newly designed and existing reactors, allowing to compensate for the increased cost of the upgraded fuel assemblies and to obtain overall increase in economic efficiency.

As a result of solving this problem during the implementation of the invention, technical results can be obtained consisting in reducing the thermal loads of the fuel elements, reducing the likelihood of depressurization of the claddings of the fuel rods, reducing the unevenness of energy release, expanding the range of maneuvering of the reactor power and improving fuel consumption by increasing the burnup depth of nuclear fuel.

These technical results are achieved in that in the fuel assembly of a pressurized water-water reactor containing a framework comprising hexagonal spacer grids, in the cells of which there is a beam of rod fuel elements with a uranium dioxide fuel core enclosed in a shell, the spacer grids contain 217 cells for the beam containing from 174 to 216 rod fuel elements with outer and inner shell diameters from 6.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.09 · 10 ^-3 m to 6.79 · 10 ^{- 3} m respectively naturally, and the mass of uranium dioxide in the beam is selected from 82.24 kg to 190.78 kg, or the spacer grids contain 169 cells for the beam containing from 132 to 168 rod fuel elements with outer and inner shell diameters from 7.80 · 10 ^{- 3} m to 8.79 · 10 ^-3 m and from 6.62 · 10 ^-3 m to 7.47 · 10 ^-3 m, respectively, and the mass of uranium dioxide in the beam is selected from 105.44 kg to 179.14 kg, moreover, the water-uranium ratio of the cell is selected from 0.69 to 2.69.

A distinctive feature of the present invention is that the spacer grids contain 217 cells for a beam containing from 174 to 216 rod fuel elements with outer and inner shell diameters from 6.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.09 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, and the mass of uranium dioxide in the beam is selected from 82.24 kg to 190.78 kg, or spacer grids contain 169 cells for a beam containing from 132 to 168 rod fuel elements with an outer diameter and inner diameter of the shell from 7.80 · 10 ^-3 m to 8.79 · 10 ^-3 m and from 6.62 · 10 ^-3 m to 7.47 · 10 ^-3 m, respectively, and the mass of uranium dioxide in the beam is selected from 105.44 kg to 179.14 kg, and the water-uranium ratio of the cell is selected from 0.69 to 2.69, which characterizes the new concept of the fuel assemblies of the VVER-440 reactor and, accordingly, the active zones of the VVER-440 reactor, which have increased efficiency both under normal operating conditions and in emergency conditions and is due to the following. Since the framework with which the beam of fuel rods is secured in the fuel assemblies must be similar to the framework of the standard fuel assemblies of the VVER-440 reactor, and the water-uranium ratio of the fuel grate should be close to the value of the standard fuel assemblies (water-uranium cell ratio of the standard fuel assembly-1 47), then the water-uranium ratio of the cell of the upgraded fuel assembly is selected from 0.69 to 2.69, and in the spacer grids 217 cells are made for a beam containing from 174 to 216 rod fuel elements with an outer and inner diameter of the shell of 6.00 10 ^-3 m to 8.00 × 10 ^-3 and from 5.09 · 10 ^-3 m to 6.79 × 10 ^-3 m, respectively, and the mass of uranium dioxide in the beam from 82.24 kg to about 190.78 kg or 169 cells for a bundle containing from 132 to 168 fuel rods with outer and inner shell diameters from 7.80 · 10 ^-3 m to 8.79 · 10 ^-3 m and from 6.62 · 10 ^-3 m to 7.47 · 10 ^-3 m, respectively, and the mass of uranium dioxide in beam from 105.44 kg to 179.14 kg, therefore, the average linear load on the fuel rods of the upgraded fuel assemblies decreases by 1.20 ÷ 1.76 times, provided that the rated power of the reactors is maintained and the neutron-physical and thermal-hydraulic characteristics are close regular characteristics VVER-440. Or, as calculations show, it is possible to increase the thermal power of the core, provided that the required safety of reactor operation is maintained by up to 3.6%, which is necessary to compensate for the increased cost of upgraded fuel assemblies.

It is advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, be from 131.23 kg to 163.02 kg, from 7.00 · 10 ^-3 m to 7.50 · 10 ^-3 m and from 5 , 94 · 10 ^-3 m to 6.36 · 10 ^-3 m, respectively, for a beam of (204 ÷ 210) rod fuel elements or that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, should be from 127.82 kg up to 157.76 kg, from 7.90 · 10 ^-3 m to 8.40 · 10 ^-3 m and from 6.70 · 10 ^-3 m to 7.13 · 10 ^-3 m, respectively, for a beam of (156 ÷ 162) rod fuel elements, and the water-uranium ratio of the cell is selected from 0.99 to 1.66.

It is also advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element be from 130.67 kg to 162.02 kg, from 7.20 · 10 ^-3 m to 7.70 · 10 ^-3 m and 6.11 · 10 ^-3 m to 6.53 · 10 ^-3 m, respectively, for a beam of (192 ÷ 198) rod fuel elements or for the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element to be from 124.04 kg to 153.11 kg, from 8.10 · 10 ^-3 m to 8.60 · 10 ^-3 m and from 6.87 · 10 ^-3 m to 7.30 · 10 ^-3 m, respectively, for a beam of (144 ÷ 150) rod fuel elements, moreover, the water-uranium ratio of the cell is selected from 0.86 to 1.51.

In addition, it is advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element ranged from 129.41 kg to 160.21 kg, from 7.40 · 10 ^-3 m to 7.90 · 10 ^-3 m and from 6.28 · 10 ^-3 m to 6.70 · 10 ^-3 m, respectively, for a beam of (180 ÷ 186) rod fuel elements or that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, are 124, 81 kg to 144.15 kg, from 8.30 · 10 ^-3 m to 8.70 · 10 ^-3 m and from 7.04 · 10 ^-3 m to 7.38 · 10 ^-3 m, respectively, for a beam of 138 rod heat dissipation the constituent elements, wherein the water-to-uranium ratio of the cell is selected from 0.74 to 1.37.

It is equally advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, be from 138.95 kg to 158.86 kg, from 7.00 · 10 ^-3 m to 7.30 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.19 · 10 ^-3 m, respectively, for a beam of 216 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element ranged from 134.19 kg to 152, 12 kg, of 7.80 x 10 ^-3 m to 8.10 × 10 ^-3 m and from 6.62 · 10 ^-3 m to 6.87 × 10 ^-3 m, respectively, for the beam of the 168 fuel rod elements s, wherein the water-to-uranium ratio of the cell is selected from 1.13 to 1.75.

It is most advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, be from 131.95 kg to 153.69 kg, from 7.60 · 10 ^-3 m to 8.00 · 10 ^-3 m and 6.45 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, for a beam of 174 rod fuel elements or that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, are from 122.28 kg to 140, 75 kg of 8,40 · 10 ^-3 m to 8.79 × 10 ^-3 m and from 7.13 · 10 ^-3 m to 7.46 × 10 ^-3 m, respectively, for a beam of 132 fuel rod lementov, wherein the water-to-uranium ratio of the cell is selected from 0.69 to 1.30.

It is also advisable that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element, be from 132.93 kg to 155.04 kg, from 7.50 · 10 ^-3 m to 7.90 · 10 ^-3 m and 6.36 · 10 ^-3 m to 6.70 · 10 ^-3 m respectively for a beam of 180 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element ranged from 124.81 kg to 144.15 kg, from 8.30 · 10 ^-3 m to 8.70 · 10 ^-3 m and from 7.04 · 10 ^-3 m to 7.38 · 10 ^-3 m for a bundle of 138 rod fuel elements, moreover, water Hurrah ovoe cell ratio is selected between 0.74 and 1.37.

It should be emphasized that only the entire set of essential features provides a solution to the problem of the invention and the receipt of the above new technical results. Indeed, fuel rods with an outer cladding diameter of 6.8 · 10 ^-3 m are known for fuel assemblies of the WWER-440 reactor. However, the choice is only a single value of the outer diameter of the cladding of a fuel rod without specifying the ranges of the necessary values of the inner and outer diameters of the cladding of the fuel rod, the corresponding range of the mass of fuel and their relationship, as well as without specifying the range of values of the water-uranium ratio of the fuel lattice (which involves combinations of the specific values ) does not allow to implement new technical results. In addition, the combination of values that make up the marked pairs of ranges of the inner and outer diameters of the fuel rods, without choosing the mass of the fuel (uranium dioxide), leads to the possibility of non-compliance with the permissible change in the water-uranium ratio of the fuel grate and / or the heating surface of the fuel rods, which allows you to fundamentally solve (subject to maintaining reactor power) the task.

List of drawings

Figure 1 shows a variant of a longitudinal section of a fuel assembly modernized in accordance with the present invention for a WWER-440 reactor; figure 2 shows a variant of the cross section of a spacer grid with a beam of fuel elements; figure 3 shows a variant of a longitudinal section of a fuel element for a modernized fuel assembly of a VVER-440 reactor; Fig. 4 shows a cross section of a modernized fuel assembly, and Fig. 5 shows curves characterizing the change in the maximum cladding temperature of the most energetically charged standard and upgraded fuel rod used in the described fuel assembly for a VVER-440 reactor in an accident with a pipeline break of DN 500.

Information confirming the possibility of carrying out the invention

The fuel assembly 1 of the VVER-440 reactor consists of a bundle of rod fuel elements 2, a hexagonal housing-cover 3, a cylindrical shank 4, a head 5 and a frame 6 (see Fig. 1). The frame 6 provides the fastening of the fuel rods 2 in the fuel assembly. The frame 6 of the assembly 1 includes hexagonal spacing grids (lower 7 supporting grid, upper 8 and middle 9 guiding grids of zirconium alloy), which are mechanically connected to each other by a central tube 10 of zirconium alloy. The distance gratings 7–9 for the described fuel assemblies have 169 or 217 cells 11 (see ft. 2). Depending on the selected number of fuel rods 2 in the beam, cylindrical displacers, burnable absorbers, process channels, etc. can be inserted into the free cells of the spacer grids 7-9. (not shown). The lower ends of the fuel rods 2 are rigidly fixed in the carrier 7 lattice, and the upper ends of the fuel rods 2 have the possibility of longitudinal movement in the guide 9 of the lattice during thermal expansions. The lower 7 supporting grid is attached to the cylindrical shank 4 of the assembly, and the upper 8 guide grid, respectively, to the head 5 of the fuel assembly. Using the shank 4 and head 5, the assembly is installed in the reactor core.

The fuel element 2 includes a fuel core, made with a diameter of from 5.00 · 10 ^-3 m to 7.33 · 10 ^-3 m and consisting of individual tablets 12 with a Central hole 13 with a diameter of from 0.79 · 10 ^-3 m to 1, 35 · 10 ^-3 m (or solid) or cylindrical rods with a length of 6.90 · 10 ^-3 m to 12.00 · 10 ^-3 m, placed in a shell 14 made with outer and inner diameters, respectively, from 6.00 · 10 ^-3 m to 8.79 · 10 ^-3 m and from 5.09 · 10 ^-3 m to 7.47 · 10 ^-3 m, which is a structural bearing element and to which end parts 14 are attached (see Fig. .3 and figure 4). The shell 14 during operation experiences stress due to expansion and swelling of the fuel, as well as due to gas evolution from the fuel, especially in places corresponding to the interface of tablets or rods. The elimination of these negative moments is carried out by profiling the shape of the tablets 12 (or rods), in particular by making their ends concave or with a conical shape of the side surface in the region of the ends (not shown).

As the material of tablets 12, it is most expedient to use compressed and sintered uranium dioxide with an average density (10.4 · 10 ³ ÷ 10.8 · 10 ³ ) kg / m ³ , but plutonium, thorium and uranium oxides or mixtures of these can also be used fissile materials. The mass of uranium dioxide in the fuel assembly is from 82.24 kg to 190.78 kg.

When choosing the thickness of the cladding 14 of the fuel rod of the upgraded core it is most advisable to keep the ratio of the cladding thickness to the outer diameter of the described fuel rod the same as in the standard fuel rods of the VVER-440 reactor, which, taking into account the preservation of the filling pressure with helium (0.2 ÷ 0.7) MPa allows us to guarantee the stability of the cladding of a fuel rod of a modernized core, not less than for regular fuel elements. In addition, it is also necessary to take into account the condition that the radial clearance between the tablets 12 of the fuel core and the cladding 14 in the described fuel rods was not less than 0.05 · 10 ^-3 m. This condition is due to technological difficulties in the assembly of fuel rods.

Due to the low thermal conductivity of the material of the tablets 12 of the fuel core, and also taking into account all the above conditions, the cladding 14 of the rod fuel rod of the fuel assembly described for the modernized core of the VVER-440 reactor must have outer and inner diameters (6.00 · 10 ^-3 ÷ 8.00 · 10 ^-3 ) m and (5.09 · 10 ^-3 ÷ 6.79 · 10 ^-3 ) m, respectively, for a bundle of (174 ÷ 216) fuel rods or (7.80 · 10 ^-3 ÷ 8.79 · 10 ^-3 ) m and (6.62 · 10 ^-3 ÷ 7.47 · 10 ^-3 ) m, respectively, for the beam (132 ÷ 168) of fuel elements. The fact is that from the first three of the above conditions it follows that the relative pitch h of the arrangement of the fuel rods (see figure 2) should provide a water-uranium ratio of cell 16 (see figure 3) for the upgraded core, close to water-uranium the ratio of the cell grids of the existing VVER-440. The water-uranium ratio of the cell for the lattices of the upgraded fuel assemblies is in the range from 0.69 to 2.69. Taking into account all the above conditions, as well as the results of neutron-physical, thermohydraulic, thermomechanical calculations and, first of all, the results of analyzes of VVER-440 accidents with coolant leaks from the primary circuit, the boundaries of the ranges of the main characteristics of the described fuel assemblies for the modernized VVER reactor core were determined -440. So, for a bundle containing from 174 to 216 fuel rods:

- the outer diameter of the cladding of the fuel rod is selected from 6.00 · 10 ^-3 m to 8.00 · 10 ^-3 m;

- the inner diameter of the cladding of the fuel rod is selected from 5.09 · 10 ^-3 m to 6.79 · 10 ^-3 ;

- the mass of uranium dioxide selected from 82.24 kg to 190.78 kg;

- 217 cells are made in the spacing grids, and for a beam containing from 132 to 168 fuel rods:

- the outer diameter of the cladding of the fuel rod is made from 7.80 · 10 ^-3 m to 8.79 · 10 ^-3 m;

- the inner diameter of the cladding of the fuel rod is made from 6.62 · 10 ^-3 m to 7.47 · 10 ^-3 m;

- the mass of uranium dioxide selected from 105.44 kg to 179.14 kg;

- 169 cells are made in the spacer grids, and the water-uranium ratio of the cell is selected from 0.69 to 2.69.

The implementation of the fuel rod of the described fuel assembly with a beam from 174 to 216 units, with an outer diameter of less than 6.00 · 10 ^-3 m, for example 5.90 · 10 ^-3 m, and, accordingly, the execution of a fuel element with an inner diameter of the shell of not more than 5.08 · 10 ^{- 3} m and fuel mass in fuel assemblies of not more than 82.23 kg, as well as non-observance of the above range of water-uranium ratio (0.69-2.69) leads to non-fulfillment of the condition for ensuring the design duration of the fuel load due to a decrease in fuel load in the upgraded fuel assemblies in comparison with the standard design of the WWER-440 fuel assemblies (which e must be compensated by increasing the burnup depth in the upgraded fuel assembly relative to the standard fuel assembly), and the performance of a fuel rod with an outer diameter of more than 8.00 · 10 ^-3 m (for example, 8.10 · 10 ^-3 m) and, accordingly, the execution of a fuel rod with an internal diameter a shell of at least 6.80 · 10 ^-3 m and a fuel mass of fuel assemblies of at least 190.79 kg leads to non-fulfillment of the condition regarding a possible increase in hydraulic friction losses in the upgraded fuel assemblies of the VVER-440 reactor compared to the standard design of the VVER-440 fuel assemblies (excess relative head pressure G H more than 19%). The execution of the fuel rod of the described fuel assembly with a beam of 132 to 168 units, with an outer diameter of less than 7.80 · 10 ^-3 m, for example 7.70 · 10 ^-3 m, and, accordingly, the execution of a fuel element with an inner diameter of the shell of not more than 6.61 · 10 ^-3 m and fuel mass in fuel assemblies of not more than 105.43 kg, as well as non-observance of the above range of water-uranium bearing also leads to non-fulfillment of the condition regarding ensuring the design duration of the fuel load due to a decrease in fuel loading in the upgraded fuel assemblies in comparison with the standard design of the VVER-440 fuel assembly (cat Roe must be compensated for increased burnup in the fuel assemblies upgraded with respect to the standard fuel assemblies) and fuel rod performance of external diameter greater than 8.79 × 10 ^-3 m (e.g., 8.90 · 10 ^-3 m) and accordingly performance of the fuel element with an inner diameter shells of at least 7.48 · 10 ^-3 m and fuel mass in fuel assemblies of at least 179.15 kg leads to failure to meet the conditions regarding a possible increase in hydraulic friction losses and upgraded fuel assemblies of the VVER-440 reactor compared to the standard design of the VVER-440 fuel assemblies (excess of relative pressure RCP greater than 10%).

It should be noted that the first four of the above conditions allow us to clarify the preferred boundaries of the ranges of the main characteristics of the described fuel element for the modernized core of the VVER-440 reactor, namely:

1) for fuel assemblies with spacer grids containing 217 cells:

- the bundle contains from 204 to 210 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.00 · 10 ^-3 m to 7.50 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 5.94 · 10 ^-3 m to 6.36 · 10 ^-3 m,

- the mass of uranium dioxide in the fuel assembly is selected from 131.23 kg to 163.02 kg,

- the water-uranium ratio of the cell is selected from 0.99 to 1.66 or the bundle contains from 192 to 198 fuel rods,

- the outer diameter of the cladding of the fuel rod is made from 7.20 · 10 ^-3 m to 7.70 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is made from 6.11 · 10 ^-3 m to 6.53 · 10 ^-3 m;

- the mass of uranium dioxide is selected from 130.67 kg to 162.02 kg,

- water-uranium ratio of the cell selected from 0.86 to 1.51 or

- the bundle contains from 180 to 186 fuel rods,

- the outer diameter of the cladding of the fuel rod is made from 7.40 · 10 ^-3 m to 7.90 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is made from 6.28 · 10 ^-3 m to 6.70 · 10 ^-3 m,

- the mass of uranium dioxide is selected from 129.41 kg to 160.21 kg,

- water-uranium ratio of the cell selected from 0.74 to 1.37.

2) for fuel assemblies with spacer grids containing 169 cells:

- the bundle contains from 156 to 162 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.90 · 10 ^-3 m to 8.40 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 6.70 · 10 ^-3 m to 7.13 · 10 ^-3 m,

- the mass of uranium dioxide in the fuel assembly is selected from 127.82 kg to 157.76 kg,

- the water-uranium ratio of the cell is selected from 0.99 to 1.66 or

- the bundle contains from 144 to 150 fuel rods,

- the outer diameter of the cladding of the fuel rod is made from 8.10 · 10 ^-3 m to 8.60 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is made from 6.87 · 10 ^-3 m to 7.30 · 10 ^-3 m,

- the mass of uranium dioxide selected from 124.04 kg to 153.11 kg,

- water-uranium ratio of the cell selected from 0.86 to 1.51 or

- the bundle contains 138 fuel rods,

- the outer diameter of the cladding of the fuel rod is made from 8.30 · 10 ^-3 m to 8.70 · 10 ^-3 m,

- the inner diameter of the cladding of a fuel rod is made from 7.04 · 10 ^-3 m to 7.38 · 10 ^-3 m,

- the mass of uranium dioxide selected from 124.81 kg to 144.15 kg,

- water-uranium ratio of the cell selected from 0.74 to 1.37.

In addition, from the first two and last two of the above conditions it follows that for the modernized core of the VVER-440 reactor, the most appropriate is the implementation of fuel assemblies with the following characteristics, namely:

1) for fuel assemblies with spacer grids containing 217 cells:

- the bundle contains 216 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.00 · 10 ^-3 m to 7.30 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 5.94 · 10 ^-3 m to 6.19 · 10 ^-3 ,

- the mass of uranium dioxide selected from 138.95 kg to 158.86 kg,

- water-uranium ratio of the cell selected from 1.13 to 1.75 or

- the bundle contains 174 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.60 · 10 ^-3 m to 8.00 · 10 ^-3 ,

- the inner diameter of the cladding of the fuel rod is selected from 6.45 · 10 ^-3 m to 6.79 · 10 ^-3 ,

- the mass of uranium dioxide selected from 131.95 kg to 153.69 kg,

- water-uranium ratio of the cell selected from 0.69 to 1.30 or

- the bundle contains 180 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.50 · 10 ^-3 m to 7.90 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 6.36 · 10 ^-3 m to 6.70 · 10 ^-3 m,

- the mass of uranium dioxide selected from 132.93 kg to 155.04 kg,

- water-uranium ratio of the cell selected from 0.74 to 1.37,

2) for fuel assemblies with spacer grids containing 169 cells:

- the bundle contains 168 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 7.80 · 10 ^-3 m to 8.10 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel element is selected from 6.62 · 10 ^-3 m to 6.87 · 10 ^-3 m,

- the mass of uranium dioxide selected from 134.19 kg to 152.12 kg,

- water-uranium ratio of the cell selected from 1.13 to 1.75 or

- the bundle contains 132 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 8.40 · 10 ^-3 m to 8.79 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 7.13 · 10 ^-3 m to 7.46 · 10 ^-3 m,

- the mass of uranium dioxide is selected from 122.28 kg to 140.75 kg,

- water-uranium ratio of the cell selected from 0.69 to 1.30 or

- the bundle contains 138 fuel rods,

- the outer diameter of the cladding of the fuel rod is selected from 8.30 · 10 ^-3 m to 8.70 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 7.04 · 10 ^-3 m to 7.38 · 10 ^-3 m,

- the mass of uranium dioxide selected from 124.81 kg to 144.15 kg,

- water-uranium ratio of the cell selected from 0.74 to 1.37.

An analysis of the operability and thermomechanical state of the fuel elements made it possible to clarify some basic structural parameters of the fuel elements of the fuel assembly described. As shown by computational studies, a significant reduction in the thermal load on the fuel rod allows us to abandon the design of a fuel pellet with a central hole that has become traditional for WWER reactors and has not found application in foreign PWR reactors. This solution is due, on the one hand, to a relatively small decrease in fuel temperature due to the central hole with reduced thermal loads on the fuel rods and an increased safety margin in relation to fuel melting, and, on the other hand, possible technological difficulties in the manufacture of tablets of reduced diameter with central holes less than 1.5 · 10 ^-3 m.

The case-cover 3, inside which the fuel rods 2 are placed, integrates all parts of the fuel assemblies and provides the necessary direction of flow of the coolant flow inside the fuel assembly between the individual fuel rods 2 in the assembly and between the fuel assemblies in the reactor core. The housing-cover 3 of the assembly is unloaded from the internal pressure of the coolant that occurs when the coolant flows through the fuel assembly. To obtain the same temperature of the coolant at the outlet of the fuel assembly, the flow rate of the coolant in the assemblies can be profiled in accordance with the distribution of heat generation along the radius of the reactor by installing throttle washers at the coolant inlet to the fuel assembly (not shown).

The manufacturing technology of the described designs of fuel elements and fuel assemblies is made using well-known standard equipment and has no differences in terms of production of similar devices.

Figure 5 presents, as an example, the curves characterizing the change at the maximum design basis accident (MPA) of the cladding temperatures of the fuel rods with the maximum load for the standard (the outer diameter of the shell of the standard fuel rod 9.10 · 10 ^-3 m) and modernized (the outer diameter of the fuel clad described Fuel assemblies 7.00 · 10 ^-3 m) of the active zones of the VVER-440 reactor. From the analysis of the state of the fuel rods in the indicated mode, it is seen (Fig. 5) that the fuel rod in the described fuel assembly has a significantly lower maximum clad temperature. For a “hot” fuel element (a fuel element with a maximum linear thermal load), the decrease in maximum temperature is 278 ° C, and for a fuel element with an average load of 150 ° C. Such values for lowering the cladding temperature of a fuel element fundamentally change the level of fuel element operability and the predicted degree of safety of VVER-440 . First of all, this is due to the strong dependence of the mechanical properties of the shell material on temperature in the region of T> 550 ° C, as well as the rapidly increasing contribution of the heat of the steam-zirconium reaction to the development of the emergency at temperatures T> 700 ° C. Therefore, the transition to the modernized zone and, accordingly a decrease in the maximum temperature at MPA from 900 ° C to a level below 600 ° C largely excludes the effect of the steam-zirconium reaction on the change in material properties and the geometric dimensions of the claddings of fuel elements.

It should also be noted that the fuel rods of the described fuel assemblies of the modernized VVER-440 reactor, due to the reduction of specific heat loads, have significantly lower fuel temperatures and have increased performance due to the reduced impact on the cladding of the fuel rod of the pressure of gaseous fission products. Their reduced output in the fuel rods of the upgraded core also leads to less corrosion on the fuel side of the cladding. This gives reason to believe (calculated justification) that in the fuel rods of the fuel assemblies of the modernized VVER-440 reactor, the average burnup of fuel (55 ÷ 60) MW · day / kg is real.

The operability of fuel rods in transient modes of operation associated with the required maneuvering power is due to many factors: the level of thermal loads, the history of operation, the speed and magnitude of the change in power, the corrosion effect on the cladding from the side of the fuel core, and others. To avoid the depressurization of fuel rods in maneuver modes in terms of speed and range of power rise of a standard reactor, which leads to economic losses. Values of the permissible "step" of power increase most sharply decrease with increasing both fuel burnup and the initial linear load. Therefore, reducing the linear thermal loads of fuel rods is one of the most effective ways to solve this problem. Reducing the maximum thermal linear loads from 40 kW / m to 23 kW / m gives virtually unlimited possibilities in changing power for modernized fuel assembly designs. The average linear load of the fuel rod of the described fuel assembly for the modernized VVER-440 reactor core with an outer diameter of 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m is (7.36 ÷ 7.54) kVg / m and ( 9.46 ÷ 9.69) kW / m for fuel elements with a cladding diameter from 7.8 · 10 ^-3 m to 8.79 · 10 ^-3 m (for a standard fuel element with a diameter of 9.1 · 10 ^-3 m average linear load equal to 12.82 kW / m). Therefore, the transition to reduced thermal loads in the fuel rods of the fuel assemblies of the upgraded VVER-440 reactor core fundamentally expands the range of maneuvering with reactor power at an increased speed.

It should also be noted that according to economic calculations, to compensate for the increased cost of a modernized fuel assembly, it is sufficient to extend the fuel cycle by a maximum of (25 ÷ 30) ef. days, or an increase in power unit by 3.6%. Assessments of the potential of the modernized core show that an increase in the duration of fuel cycles by 30 ef. days is achieved by implementing the overload scheme of upgraded fuel assemblies with a deeper decrease in neutron leakage, which is feasible on VVER-440 reactors, taking into account the growth of heat reserves when switching to a reduced fuel rod diameter. Thermohydraulic calculations of the modernized core of the VVER-440 reactor confirm the potential possibility of increasing the thermal power of the core when using fuel rods of reduced diameter by up to (15%) significantly more than the required (3.6%) to compensate for the increased cost of the upgraded fuel assemblies. Thus, the design of the upgraded fuel assembly for the VVER-440 reactor described above allows not only to compensate for the increased cost, but also to increase economic efficiency.

Based on the foregoing, it can be stated that the transition to a modernized core with the described fuel assemblies in VVER-440 reactors makes it possible to reduce the thermal load on the fuel elements by (1.33 ÷ 1.76) times. Such a significant reduction in linear thermal loads in the fuel rods of the fuel assemblies described in the modernized VVER-440 reactor core allows:

- increase the safety of power plants with a VVER-440 reactor;

- to provide an opportunity to solve the problem associated with maneuvering the power of the VVER-440 reactor;

- increase the operability of fuel rods under normal operating conditions, which gives reason to consider it realistic to achieve an average fuel burnup in fuel rods (55-60) MW · day / kg.

It should be noted that the described fuel assemblies can be used not only in VVER-440 reactors, but also in other pressurized water-water reactors (PWR), boiling water reactors (BWR), and in heavy water reactors.

Claims (7)

1. A fuel assembly of a pressurized water reactor containing a frame including hexagonal spacer grids, in the cells of which there is a bundle of rod fuel elements with a uranium dioxide fuel core enclosed in a shell, characterized in that the spacer grids contain 217 cells for a beam containing from 174 to 216 rod fuel elements with outer and inner shell diameters from 6.00 · 10 ^-3 to 8.00 · 10 ^-3 m and from 5.09 · 10 ^-3 to 6.79 · 10 ^-3 m, respectively and the mass of uranium dioxide in 82.24 to 190.78 kg is selected for the beam or the spacer grids contain 169 cells for the beam containing 132 to 168 rod fuel elements with outer and inner shell diameters of 7.80 · 10 ^-3 to 8.79 · 10 ^-3 m and from 6.62 · 10 ^-3 to 7.47 · 10 ^-3 m, respectively, and the mass of uranium dioxide in the beam is selected from 105.44 to 179.14 kg, moreover, the water-uranium ratio of the cell is selected from 0, 69 to 2.69. 2. The fuel assembly of the pressurized water reactor according to claim 1, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the core fuel element are from 131.23 to 163.02 kg, from 7.00 · 10 ^{- 3} to 7.50 · 10 ^-3 m and from 5.94 · 10 ^-3 to 6.36 · 10 ^-3 m, respectively, for a beam of 204 ÷ 210 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell rod fuel element is from 127.82 to 157.76 kg, from 7.90 · 10 ^-3 to 8.40 · 10 ^-3 m and from 6.70 · 10 ^-3 to 7.13 · 10 ^-3 m, respectively for a beam of 156 ÷ 162 rod fuel elements, and the water-uranium ratio of the cell is selected from 0.99 to 1.66. 3. The fuel assembly of the pressurized water reactor according to claim 1, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the core fuel element are from 130.67 to 162.02 kg, from 7.20 · 10 ^{- 3} to 7.70 · 10 ^-3 m and from 6.11 · 10 ^-3 to 6.53 · 10 ^-3 m, respectively, for a beam of 192 ÷ 198 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell rod fuel element is from 124.04 to 153.11 kg, from 8.10 · 10 ^-3 to 8.60 · 10 ^-3 m and from 6.87 · 10 ^-3 to 7.30 · 10 ^-3 m, respectively for a beam of 144 ÷ 150 rod fuel elements, and the water-uranium ratio of the cell is selected from 0.86 to 1.51. 4. The fuel assembly of the pressurized water reactor according to claim 1, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the core fuel element are from 129.41 to 160.21 kg, from 7.40 · 10 ^{- 3} to 7.90 · 10 ^-3 m and from 6.28 · 10 ^-3 to 6.70 · 10 ^-3 m, respectively, for a beam of 180 ÷ 186 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell rod fuel element is from 124.81 to 144.15 kg, from 8.30 · 10 ^-3 to 8.70 · 10 ^-3 m and from 7.04 · 10 ^-3 to 7.38 · 10 ^-3 m, respectively for a beam of 138 rod fuel elements, the water-uranium ratio of the cell being selected from 0.74 to 1.37. 5. The fuel assembly of the pressurized water reactor according to claim 1, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element are from 138.95 to 158.86 kg, from 7.00 · 10 ^{- 3} to 7.30 · 10 ^-3 m and from 5.94 · 10 ^-3 to 6.19 · 10 ^-3 m, respectively, for a beam of 216 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element ranged from 134.19 to 152.12 kg, 7.80 · 10 ^-3 to 8.10 · 10 ^-3 m and from 6.62 × 10 ^-3 to 6.87 · 10 ^-3 m, respectively, for beam from the rod 168 of the fuel elements, wherein the water-to-uranium ratio of the cell is selected from 1.13 to 1.75. 6. The fuel assembly of the pressurized water reactor according to claim 1, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the core fuel element are from 131.95 to 153.69 kg, from 7.60 · 10 ^{- 3} to 8.00 · 10 ^-3 m and from 6.45 · 10 ^-3 to 6.79 · 10 ^-3 m, respectively, for a beam of 174 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element ranged from 122.28 to 140.75 kg, 8.40 · 10 ^-3 to 8.79 · 10 ^-3 m and from 7.13 × 10 ^-3 to 7.46 · 10 ^-3 m, respectively, for beam from the rod 132 of the fuel elements, wherein the water-to-uranium ratio of the cell is selected from 0.69 to 1.30. 7. The fuel assembly of the pressurized water reactor according to claim 1 or 4, characterized in that the mass of uranium dioxide in the beam, the outer and inner diameters of the shell of the rod fuel element are from 132.93 to 155.04 kg, from 7.50 · 10 ^-3 to 7.90 · 10 ^-3 m and from 6.36 · 10 ^-3 to 6.70 · 10 ^-3 m, respectively, for a beam of 180 rod fuel elements or the mass of uranium dioxide in the beam, the outer and inner diameters of the shell rod fuel element is from 124.1 to 144.15 kg, from 8.30 · 10 ^-3 to 8.70 · 10 ^-3 m and from 7.04 · 10 ^-3 to 7.38 · 10 ^-3 m, respectively about for a beam of 138 rod fuel elements, and the water-uranium ratio of the cell is selected from 0.74 to 1.37.

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