RU2241262C2 - Water-moderated power reactor core - Google Patents

Abstract

FIELD: nuclear power engineering; cores for water-moderated reactors.

SUBSTANCE: proposed core used in particular for type VVER-440 reactors is characterized in that water-to-uranium ratio of its fuel assembly spacer grid, outer and inner diameters of fuel element can are 0.98 to 2.52 and 7.00 m · 10 ^-3 to 8.00· 10 ^-3 m, and 5.94 · 10 ^-3 to 6.79 · 10 ^-3 m, respectively, for fuel assemblies incorporating 174 to 216 fuel elements, or water-to-uranium ratio of fuel assembly spacer grid, outer and inner diameters of fuel element can are 1.14 to 2.76, 7.80 · 10 ^-3 to 8.79 · 10 ^-3 m, and 6.62 · 10 ^-3 to 7.47 · 10 ^-3 m, respectively, for fuel assemblies incorporating 132 to 168 fuel elements; ratio of core height to fuel assembly length is 0.7488 to 0.7897.

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

6 cl, 6 dwg

Description

FIELD OF THE INVENTION

The invention relates to nuclear engineering and relates to the improvement of the core areas of nuclear reactors in which water (so-called water-cooled nuclear reactors) is used as a coolant and a moderator, used as a heat source for power plants, in power plants, etc., especially in thermal reactors power of the order (1150-1700) MW.

State of the art

The prospects for the development of nuclear energy are largely determined by solving the issue of ensuring the safety of nuclear power plants (NPPs). When creating active zones that provide a qualitatively new level of NPP safety, it is necessary to be based on proven technical solutions, positive experience in the design and operation of existing NPPs. The most significant consequences for nuclear power plants, in particular with pressurized water reactors (VVER), are accidents with loss of the primary coolant, the development of which, when passive and active safety systems are repeatedly redundant, which introduce cooling water with a neutron absorber into the primary circuit, can lead to serious consequences.

The problem of improving the safety level of operating NPPs with VVER reactors has various solutions. However, at present, it is being solved, as a rule, by increasing the reliability of protective systems, improving individual components and equipment, optimizing modes and operating procedures.

At the same time, the issues of reducing in the normal mode the thermal stress of fuel rods, the shells of which are one of the main barriers to the spread of radioactive substances and which can be depressurized in emergency situations, primarily because of their overheating, are not addressed. This trend is mainly due to many years of successful experience in operating nuclear fuel of an existing design and its well-established production.

VVER-type reactors in the process of introduction into nuclear power have not undergone changes in the basic technical solutions. Such decisions incorporated in the designs of domestic VVER-type reactors include:

- all devices inside the reactor vessel must be removable for possible repair, replacement and for monitoring the internal surface of the reactor vessel;

- installation in the upper part of the reactor of the organs of the control and protection system (CPS) and equipment for monitoring the operation of the reactor for convenient operational maintenance;

- fuel assemblies (FAs), allowing to create an active zone configuration close to cylindrical, are placed in a removable basket, the bottom of which is the supporting structure of the active zone;

- the coolant in the core moves from the bottom up, which makes it possible to cool fuel assemblies in natural circulation mode.

The core of the VVER-440 reactor, with a nominal electric power of the power unit with which is 440 MW (with a correspondingly thermal power of the reactor of 1175 MW), is recruited from hexagonal fuel assemblies that are installed almost close to each other in the core basket. In fuel assemblies, rod fuel rods are installed in a triangular step. Pressed and sintered uranium dioxide tablets are used as nuclear fuel. A hollow tube is located in one cell of the (central) fuel assembly. Sensors for measuring water temperature and energy release detectors are placed inside this tube (see I.Ya. Emelyanov, V.I. Mikhan, M.I. Solonin, etc. Design of nuclear reactors, M., Energoizdat, 1982, p. 76) .

A fuel assembly of a VVER-440 reactor consists of a bunch of rod fuel elements, a hexagonal housing-cover, a cylindrical shank, a head and an assembly frame, with the help of which the fuel elements are mounted in the assembly. The fuel assembly frame includes hexagonal spacer grids (lower bearing 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 rigidly fixed in the carrier grid, and the upper ends of the fuel rods have the possibility of longitudinal movement in the guide grid with temperature expansions. The lower support grid is attached to the cylindrical shank of the assembly, and the upper guide grill, respectively, to the assembly head. Using the shank and the head, the fuel assembly is installed in the reactor basket (see G.N. Ushakov Technological channels and fuel elements of nuclear reactors, M., Energoizdat, 1981, p. 89, fig. 2.8 a).

The design of the rod fuel rods and the core for VVER reactors should provide mechanical stability and durability of the fuel rods, including in design emergency conditions at high temperatures and in the presence of long-term powerful fluxes 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 the design of the active zone, it is necessary to take into account the positive effect of increasing the ratio of the heat transfer surface of the fuel element to the active volume occupied by nuclear fuel.

The well-known active zone of the VVER-440 pressurized water reactor is composed of 349 (or 313) hexagonal fuel assemblies assembled from rod fuel elements (see I.Ya. Emelyanov, V.I. Mikhan, M. Solonin and others. Design of nuclear Reactors, M., Energoizdat, 1982, p. 61, 76, Fig. 3.10a). The VVER-440 reactor core has a shape close to a cylinder with a height of 2.42 m and an equivalent diameter of 2.88 m. The total height of the fuel assembly is 3.217 m; there is a slight water gap between the fuel assemblies (3 · 10 ^-3 m) Each fuel assemblies of the VVER-440 reactor contain 126 rod fuel elements made with an outer diameter of 9.1 · 10 ^-3 m and have an average linear thermal load on the fuel element of 12.82 kW / m.

Such a fuel rod provides a relatively high level of fuel burnup in the aforementioned fuel assemblies and has proved itself well 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 in the 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 elements 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 the level of (700-750) ° С. Therefore, if in the core of the VVER-440 reactor the maximum thermal loads were reduced to the average level, then the 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 addition, this problem of relatively high fuel temperature in the nominal mode is compounded by an increase in the fuel burnup depth, when the fuel rod performance is even close to the maximum permissible even under normal operating conditions.

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 an active zone with rod fuel rods of a container structure of reduced diameter (provided that the reactor power is kept close to the regular active zone of the water-uranium ratio of the fuel grate), which will allow solve the problem of possible depressurization of fuel rods at the initial stage of an accident with loss of coolant. In addition, when developing a modernized core of the VVER-440 reactor, it is necessary to select the main parameters from the condition of maximally preserving the design of the core and the nuclear power plant, as well as providing neutron-physical and thermal hydraulic characteristics close to the standard characteristics of the core of the VVER-440 reactor, as an object of the present invention, it is not the development of a 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 assembly and the height of the upgraded core should be the same as in the standard design of the core of the VVER-440 reactor;

- the turnkey size (145 · 10 ^-3 m) and the height of the fuel assemblies of the upgraded core should be the same as in the standard design of the VVER-440 fuel assemblies;

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

- the decrease in fuel loading in the fuel assemblies of the modernized core compared to the standard design of the fuel assemblies of the VVER-440 reactor should not exceed 10%;

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

- the number, design and placement of the CPS in the modernized core should be the same as in the regular core of the VVER-440 reactor.

To increase the burnup depth of nuclear fuel or to increase operational safety at a given load due to limitations associated with the permissible temperature of the fuel and heat sink, they seek to increase the ratio of the surface of a fuel rod to its volume, which ensures a decrease in heat flux due to an increase in the 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 fuel elements”, conference report “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 the upgraded fuel assemblies with fuel rods with a diameter of 6.8 × 10 ^-3 m with the initial enrichment selected equal to the enrichment of the standard fuel assemblies, it has a fuel burn-up depth of 2.1% more than that of a standard fuel assembly, this does not fully compensate for the loss in the duration of the fuel load compared to a standard fuel assembly. Therefore, the following should also be added to the above limitations:

- to ensure the projected duration of the fuel loading, the decrease in fuel loading in the upgraded fuel assemblies in comparison with the standard design of the fuel assemblies should be compensated by an increase in the burnup depth of nuclear fuel.

The closest in technical essence to the described technical solution is the active zone of a pressurized water reactor containing fuel assemblies assembled from rod fuel elements (RU 2126999, G 21 C 1/04, 02.27.99)

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 permissible fuel burnup depth and reduce the chance of fuel depressurization.

However, a comparative assessment of the cost of the standard VVER-440 fuel assemblies (diameter of the fuel rods is 9.1 · 10 ^-3 m) and the upgraded fuel assemblies (reduced fuel rods of 6.8 · 10 ^-3 m) showed that the factory cost of the upgraded fuel assemblies for VVER-440 reactors increased by 18%, which is one of the reasons why the active zones with such fuel elements 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 reactor with a thermal power of 1150 to 1700 MW, which have 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, when implementing the invention, new technical results can be obtained, consisting in reducing the thermal loads of the fuel elements, reducing the likelihood of depressurization of the cladding 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 allowable burnup depth of nuclear fuel .

These technical results are achieved in that in the active zone of a pressurized water reactor containing fuel assemblies with a hexagonal fuel grid, assembled from rod fuel elements, at least one fuel assembly contains from 174 to 216 rod fuel elements having external and internal the diameters of the shell of the fuel element from 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 0.98 to 2.52 or the fuel assembly contains from 132 to 168 rod fuel elements having the outer and inner diameters of the shell of the fuel element 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 water – uranium ratio of the fuel grate is selected from 1.14 to 2.76, and the ratio of the height of the core to the length of the fuel assembly is from 0.7488 to 0.7897.

A distinctive feature of the present invention is that at least one fuel assembly contains from 174 to 216 rod fuel elements having outer and inner diameters of the shell of the fuel element from 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 0.98 to 2.52 or the fuel assembly contains from 132 to 168 rod fuel elements having an outer and inner diameter of the shell of the fuel element 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 water-uranium ratio of the fuel lattice is selected from 1.14 to 2.76, and the ratio of the height of the core to the length of the fuel assembly is from 0.7488 to 0.7897, which characterizes the new concept of the VVER-440 reactor core, which has increased working capacity both under normal operating conditions and in emergency conditions and is due to the following. Since the described active zone, as well as the regular core of the VVER-440 reactor, is composed of 349 or 313 hexagonal fuel assemblies, which have a “turnkey” size, the height and structure of the frame, with which the beam bundle is secured to the fuel assemblies, are similar to the standard fuel assemblies the VVER-440 reactor, and the bundle contains from 174 to 216 rod fuel elements having an outer and inner diameter of the cladding of a fuel element from 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 0.98 to 2.52, or from 132 to 168 st rods having outer and inner diameters of the cladding of a fuel rod from 7.8 · 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 water-uranium ratio of the fuel lattice is selected from 1.14 to 2.76, and the ratio of the height of the active zone to the length of the fuel assembly is from 0.7488 to 0.7897, the average linear thermal load on the fuel rods of the modernized core decreases by 1.33-1.76 times, while maintaining the rated power of the reactor and providing neutron-physical and thermohydraulic characteristics close to regular ha VVER-440 reactor specifications. Or, as calculations show, it is possible to increase the thermal power of the core, provided that the required safety of the reactor is maintained, by up to 3.6%, which is necessary to compensate for the increased cost of upgraded fuel assemblies.

It should be noted that it is advisable that at least one fuel assembly contains from 180 to 210 rod fuel elements having the outer and inner diameters of the shell of the fuel element from 7.00 · 10 ^-3 m to 7.90 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.70 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was selected from 1.07 to 2.40 or so that the fuel assembly contained 138 to 162 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.9 · 10 ^-3 m to 8.7 · 10 ^-3 m and from 6.70 · 10 ^-3 m to 7.38 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was chosen from 1.24 to 2.49.

It is also advisable that at least one fuel assembly contains from 186 to 204 rod fuel elements having outer and inner diameters of the shell of the fuel element from 7.10 · 10 ^-3 m to 7.80 · 10 ^-3 m and from 6.02 · 10 ^-3 m up to 6.62 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was selected from 1.17 to 2.18 or the fuel assembly contained from 144 to 156 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 8.00 · 10 ^-3 m to 8.6 · 10 ^-3 m and from 6.79 · 10 ^-3 m to 7.30 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was selected from 1.35 to 2.24.

In addition, it is advisable that, but at least one fuel assembly contains from 192 to 198 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.20 · 10 ^-3 m to 7.70 · 10 ^-3 m and from 6.11 · 10 ^{- 3} m to 6.53 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was selected from 1.28 to 1.98 or the fuel assembly contained 144 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 8.10 · 10 ^-3 m to 8.5 · 10 ^-3 m and from 6.87 · 10 ^-3 m to 7.21 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was chosen from 1.53 to 2.15.

It is most advisable that at least one fuel assembly contains 174 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.6 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 6.45 · 10 ^-3 m to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate was selected from 1.20 to 1.89 or the fuel assembly contained 132 rod fuel elements having outer and inner diameters of the fuel element shell from 8.40 · 10 ^-3 m to 8.79 · 10 ^-3 m and from 7.13 · 10 ^-3 m to 7.46 · 10 ^-3 m, corresponding of course, and the water – uranium ratio of the fuel grate was chosen from 1.43 to 2.16.

It is equally advisable that at least one fuel assembly contains 180 rod fuel elements having outer and inner diameters of the shell of the fuel element from 7.5 · 10 ^-3 m to 7.90 · 10 ^-3 m and from 6.36 · 10 ^-3 m to 6.70 · 10 ^-3 m, respectively, and the water-uranium ratio was selected from 1.24 to 1.89 or the fuel assembly contained 138 rod fuel elements having outer and inner diameters of the shell of the fuel element from 8.30 · 10 ^-3 m to 8.7 · 10 ^- 3 m and from 7.04 · 10 ^-3 m to 7.38 · 10 ^-3 m, respectively, and water The ratio was chosen from 1.44 to 2.10.

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 new technical results. Indeed, as noted earlier, fuel rods with an outer cladding diameter of 6.8 · 10 ^-3 m are known, but, however, this feature is not enough to solve the problem. Failure to fulfill at least one of the essential features included in the independent claim will not allow us to solve the problem and provide new technical results. So, for example, the absence of a sign regarding the water-uranium ratio leads to a violation of the first three of the above conditions, i.e. the principle of choosing the basic geometric parameters of the fuel grid of the modernized core is violated, which should be carried out from the condition of maintaining the design of the core and ensuring the basic neutron-physical and thermal hydraulic characteristics of the VVER-440 reactor close to the design values of the standard core.

It should also be noted that for the manufacture of an active zone with the above-mentioned essential features when designing a modernized active zone, it is necessary to specify the outer and inner diameters of the fuel cladding from the above ranges, and then, using simple calculations, determine the water-uranium ratio of the fuel grate taking into account the above requirements. And, if the obtained value of the water-uranium ratio of the fuel grate does not correspond to the claimed range of values, then it is necessary to make changes to the specified initial data and recalculate.

List of drawings

Figure 1 shows a longitudinal section of a fuel assembly of the described core for a VVER-440 reactor, Figure 2 shows a cross section of a fuel assembly of the described core for a VVER-440 reactor, Figure 3 shows a longitudinal section of a fuel cell for the described active zones of the VVER-440 reactor, Fig. 4 presents curves characterizing the change in the maximum temperature of the cladding of the most energetically charged fuel rod of the standard and described core of the VVER-440 reactor in the event of an accident with by calling the Du-500 pipeline, Fig. 5 shows the change in the tensile strength and stresses in the shell of the fuel element of the VVER-440 reactor core during an accident with a rupture of the Du-500 pipeline, Fig. 6 shows the change in the tensile strength and stresses in the fuel shell element of the described VVER-440 reactor core during an accident with a rupture of the Du-500 pipeline.

Information confirming the possibility of carrying out the invention

The modernized core, according to the new VVER-440 reactor concept, is composed of 349 or 313 hexagonal fuel assemblies 1 having the same overall dimensions (turnkey size and height). Moreover, at least one of the 349 or 313 fuel assemblies of the upgraded core has the following structure (see figure 1 and figure 2). The fuel assembly 1 of the claimed core consists of a bundle of rod fuel rods 2, a hexagonal housing-cover 3, a cylindrical shank 4, a head 5 and a frame 6 of assembly 1. Using the latter, the fuel rods 2 are secured to the fuel assembly. The frame 6 of the fuel 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 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 assembly. Using the shank 4 and head 5, the assembly is installed in the core basket (see Fig. 1). The fuel assemblies of the claimed core of the modernized VVER-440 reactor contain from 174 to 216 rod fuel elements with the outer and inner diameters of the cladding of a fuel rod, respectively, from 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m and from 5.94 · 10 ^-3 m to 6.79 · 10 ^-3 m, and the water-uranium ratio of the fuel grate in the upgraded core is selected from 0.98 to 2.52 and can vary in the indicated range during fuel overloads. Or they contain from 132 to 168 rod fuel elements with outer and inner shell diameters, respectively, from 7.8 · 10 ^-3 m to 8.79 · 10 ^-3 m and from 6.62 · 10 ^-3 m to 7.47 · 10 ^-3 m, and the water-uranium ratio the fuel grate is selected from 1.14 to 2.76 and can vary in the indicated range during fuel overloads. Moreover, in the modernized core, the ratio of the height of the core to the length of the fuel assembly is from 0.7488 to 0.7897.

The fuel element 2 includes a fuel core, consisting of solid tablets 11 or having a central hole 12 (or cylindrical rods) placed in a shell 13, which is a structural bearing element and to which end parts 14 are tilted (see Fig. 3). The shell 13 during operation is subjected to stresses 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 11 (or rods). The elimination of these negative moments is carried out by profiling the shape of the tablets 11 (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 in the drawing).

As the material of tablets 11, 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 or thorium oxide, uranium nitrides and carbides can also be used, as well as mixtures of said fissile materials.

When choosing the thickness of the cladding of a fuel rod of a modernized core, it is most advisable to keep the ratio of cladding thickness to the outer diameter of the described fuel rod the same as in standard VVER-440 fuel rods, which, taking into account the preservation of the filling pressure with helium (0.2-0.7) MPa, can guarantee increased stability cladding of a fuel rod of a modernized core compared to a standard one. In addition, it is also necessary to take into account the condition that the radial clearance between the tablets 11 of the fuel core and the cladding 13 in the fuel rods described in the active zone 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 11 of the fuel core, as well as taking into account all the above conditions. the cladding 14 of the core rod of the described core for the modernized VVER-440 reactor should have outer and inner diameters (7.00 · 10 ^-3 -8.00 · 10 ^-3 ) m and (5.94 · 10 ^-3 -6.79 · 10 ^-3 ) m, respectively , for a bundle of (174-216) fuel rods or from (7.8 · 10 ^-3 -8.79 · 10 ^-3 ) m and (6.62 · 10 ^-3 -7.47 · 10 ^-3 ) m, respectively, for a bundle of (132- 168). The fact is that from the first three of the above conditions it follows that the relative step between the fuel rods should provide the water-uranium ratio of the fuel grate for the modernized core, close to the water-uranium ratio of the fuel gratings operating VVER-440. 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 active zone for the modernized VVER-reactor were determined 440. So, for fuel assemblies containing from 174 to 216 rod fuel elements:

- the outer diameter of the cladding of the fuel rod is selected from 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m,

- the inner diameter of the cladding of the fuel rod is selected from 5.94 · 10 ^-3 m to 6.79 · 10 ^-3 m,

- the water-uranium ratio of the fuel grate is selected from 0.98 to 2.52, and for fuel assemblies containing from 132 to 168 rod fuel rods:

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

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

- water-fuel ratio of the fuel grid is selected from 1.14 to 2.76,

moreover, the ratio of the height of the active zone to the length of the fuel assembly is from 0.7488 to 0.7897.

It should be noted that, as calculations have shown, for the claimed core in which fuel assemblies are installed with fuel rods having an outer cladding diameter of 7.00 · 10 ^-3 m, the average linear power of such a fuel element will be (7.36-7.54) kW / m, and the average the specific power per heating surface unit will be (334.6-343.0) kW / m ² , and the margin before the boiling crisis (DNBR) in a fuel assembly with a maximum power will be ~ 4.5.

The implementation of the fuel rod of the described active zone (fuel assemblies with a bundle of fuel rods from 174 to 216 units) with an outer diameter of less than 7.00 · 10 ^-3 m, for example 6.90 · 10 ^-3 m, and, accordingly, the execution of a fuel rod with an internal diameter of no more than 5.93 · 10 ^-3 m and non-compliance with the above ranges of the water-fuel ratio of the fuel grate (0.98-2.52) and the ratio of the height of the active zone to the length of the fuel assembly (0.7488-0.7897) leads to failure to meet the conditions for ensuring the design duration of the fuel load due to a decrease in fuel load in the modernized TVS, Wed Avoiding the standard design of the VVER-440 fuel assemblies (which should be compensated by increasing the burnup depth in the upgraded fuel assemblies with respect to the standard fuel assemblies), and making fuel rods with an outer diameter of more than 8.00 · 10 ^-3 m (for example, 8.10 · 10 ^-3 m) and, accordingly, the performance of a fuel rod with an inner diameter of at least 6.80 · 10 ^-3 m leads to non-fulfillment of the condition regarding a possible increase in hydraulic friction losses in the modernized fuel assemblies of the VVER-440 reactor compared to the standard design of the VVER-440 fuel assemblies (exceeding the value relative to more than 19% of the pressure of the MCP). The implementation of the fuel rod of the described active zone (fuel assemblies with a bunch of fuel rods from 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 rod with an internal diameter of not more than 6.61 10 ^-3 m and non-compliance with the above ranges of the water-oranium ratio of the fuel grate (1.14-2.76) and the ratio of the height of the active zone to the length of the fuel assembly (0.7488-0.7897) also does not fulfill the condition for ensuring the design duration of the fuel load due to a decrease in the specific fuel load in modernized T C, in comparison with the possibility of changing the relative fuel loading in the upgraded fuel assemblies of the VVER-440 reactor compared to the standard design of the WWER-440 fuel assemblies (which should be offset by an increase in the burnup depth in the upgraded fuel assemblies in relation to the standard fuel assemblies), and the execution of a fuel rod with an outer diameter of more than 8.79 · 10 ^-3 m (e.g., 8.90 · 10 ^-3 m) and coootvetstvenno, performance of the fuel element casing with an inner diameter of at least 7.48 · 10 ^-3 m leads to default conditions relating to the possible increase of hydraulic friction losses in vogue saltatory VVER-440 in comparison with the standard design VVER-440 (exceeding the value of the relative pressure of more than 10% MCP).

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. So for a fuel assembly containing from 180 to 210 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 7.00 · 10 ^-3 m to 7.90 · 10 ^-3 m,

- the inner diameter of the shell of the core fuel rod is selected from 5.94 · 10 ^-3 m to 6.70 · 10 ^-3 m,

- water-fuel ratio of the fuel grid is selected from 1.07 to 2.40,

and for a fuel assembly containing from 138 to 162 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 7.90 · 10 ^-3 m to 8.70 · 10 ^-3 m,

- the inner diameter of the shell of the core fuel rod is selected from 6.70 · 10 ^-3 m to 7.38 · 10 ^-3 m,

- water-fuel ratio of the fuel grid is selected from 1.24 to 2.49.

For a fuel assembly containing from 186 to 204 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 7.10 · 10 ^-3 m to 7.80 · 10 ^-3 m,

- the inner diameter of the shell of the core fuel rod is selected from 6.02 · 10 ^-3 m to 6.62 · 10 ^-3 ,

- water-fuel ratio of the fuel grid is selected from 1.17 to 2.18,

and for a fuel assembly containing from 144 to 156 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 8.00 · 10 ^-3 m to 8.60 · 10 ^-3 m,

- the inner diameter of the shell of the rod fuel rod is selected from 6.79 · 10 ^-3 m to 7.30 · 10 ^-3 m,

- water-fuel ratio of the fuel grid is selected from 1.35 to 2.24.

For a fuel assembly containing from 192 to 198 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 7.20 · 10 ^-3 m to 7.70 · 10 ^-3 m,

- the inner diameter of the shell of the core fuel rod is selected from 6.11 · 10 ^-3 m to 6.53 · 10 ^-3 m,

- the water-uranium ratio of the fuel grate is selected from 1.28 to 1.98, and for a fuel assembly containing 144 rod fuel elements:

- the outer diameter of the shell of the core fuel rod is selected from 8.10 · 10 ^-3 m to 8.50 · 10 ^-3 m,

- the inner diameter of the shell of the core fuel rod is selected from 6.87 · 10 ^-3 m to 7.21 · 10 ^-3 m,

- water-fuel ratio of the fuel grid is selected from 1.53 to 2.15.

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.

So for a fuel assembly containing 174 rod fuel elements:

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

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

- water-fuel ratio of the fuel grid is selected from 1.20 to 1.89,

- and for a fuel assembly containing 132 rod fuel elements:

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

- the inner diameter of the shell of the core fuel rod is selected from 7.13 · 10 ^-3 m to7.46 · 10 ^-3 ,

- water-fuel ratio of the fuel grid is selected from 1.43 to 2.16.

For a fuel assembly containing 180 rod fuel elements:

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

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

- water-fuel ratio of the fuel grid is selected from 1.24 to 1.89,

and for a fuel assembly containing 138 rod fuel elements:

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

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

- water-fuel ratio of the fuel grid is selected from 1.44 to 2.10.

The analysis of the operability and thermomechanical state of the fuel rods made it possible to clarify some basic structural parameters of the fuel rods of the described active zone. As shown by computational studies, a significant reduction in the thermal load on the fuel rods allows us to abandon the design of a fuel pellet with a central hole that has become traditional for VVER 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 elements and an increased safety margin in relation to fuel melting, and, on the other hand, to possible technological difficulties in the manufacture of tablets with central holes less than 1.5 10 ^-3 m.

Heat carrier - water in the active zone moves from bottom to top, which, in particular, provides cooling of fuel assemblies in the natural circulation mode. The case-cover 3, inside which the fuel rods 2 are placed, integrates all parts of the fuel assemblies into a single unit and provides the necessary direction of flow of the coolant flow inside the TBC 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 that occurs when the coolant flows through the active zone. 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 inlet of the fuel assembly (not shown). Water heated in the core is sent to the steam generators, where it transfers its heat to the water of the second circuit, and then returns to the core.

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

Figure 4, as an example, presents curves characterizing the change at the maximum design basis accident (MPA) of the cladding temperature of the fuel rods with a maximum load for the standard (outer cladding diameter of the standard clutch 9.10 · 10 ^-3 m) and upgraded (outer cladding diameter described Fuel assembly 7.00 · 10 ^-3 m) of the active zones of the VVER-440 reactor. An analysis of the state of the fuel elements in this mode shows that the fuel element in the described fuel assembly has a significantly lower maximum clad temperature. So, for a "hot" fuel element (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 of decreasing the temperature of the cladding of fuel rods fundamentally change the level of operability of the fuel rods 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 and the development of an emergency at temperatures T> 700 ° C. The transition to the modernized zone and, consequently, a decrease in the maximum temperature at MPA from 900 ° C to a level below 600 ° C largely excludes the influence of the heat of the steam-zirconium reaction on the change in the material properties and the geometric dimensions of the cladding of fuel rods.

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 power maneuvering is determined by 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 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 core of the VVER-440 reactor with an outer diameter of 7.00 · 10 ^-3 m to 8.00 · 10 ^-3 m is (7.36-7.54) kW / m and (9.46-9.69) kW / m for fuel elements with a cladding diameter of 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, the average linear load is 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.

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 either extend the fuel cycle by a maximum of (25-30) day / night, or increase the power unit by 3.6%. Estimates of the potential of the upgraded core show that an increase in the duration of fuel cycles by 30 days per day is achieved by overloading the upgraded fuel assemblies with a deeper decrease in neutron leakage, which will become feasible in VVER-440 reactors, taking into account the increase in heat reserves when switching to a reduced diameter fuel rods. 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 obtain an increase in economic efficiency.

Figure 5 and figure 6 shows the change in the health parameters of a standard fuel rod (with a diameter of 9.1 · 10 ^-3 m) and a correspondingly modernized (diameter of 7.00 · 10 ^-3 m) fuel rod during an accident with a rupture of the pipeline DN 500 at the inlet to the reactor. The calculated value of the probability of destruction of the shell of a standard fuel element is 0.4, and the dynamics of changes in the probability of destruction of a standard fuel element is shown in Fig.5. It should be noted that the nature of the mechanical loading of the fuel cladding is determined by the pressure drop of the coolant and the internal pressure in the fuel, as well as the temperature regime of the fuel cladding. During the above accident, the pressure of the coolant exceeds the pressure in the fuel rod at a high level of cladding temperatures (over 800 ° C), which leads to negative plastic deformations of the cladding and, as a result, subsequent contact of the fuel with the cladding. Due to the mechanical interaction of the fuel and the cladding of the fuel element (in the first 4 seconds), the circumferential stresses take positive values, which leads to an increase in the probability of destruction of the cladding of a regular fuel element. As calculations showed, for the fuel elements of the described VVER-440 reactor core (see FIG. 6), their cladding does not lose stability during the accident, since they have a lower cladding temperature level (~ 280 ° С).

As can be seen from Fig.6, for the fuel rods of the described active zone there is a significant margin until the shell is destroyed by exceeding the strength criterion (the probability of depressurization of the fuel cladding of the described active zone is zero).

Neutron-physical studies of the described core for the VVER-440 reactor showed:

- the upgraded core allows to ensure the basic design, and in a number of parameters improved neutron-physical characteristics with a significant increase in the safety of the VVER-440 reactor and the operational reliability of the fuel rods;

- the maximum value of the linear thermal load on a fuel rod of the described active zone in a stationary fuel load does not exceed q ₁ <12.5 kW / m (without taking into account uncertainties), while for standard fuel elements this value is q ₁ <28.5 kBt / m;

- fuel loading (according to U ²³⁵ ) is the same as in the VVER-440 serial reactor;

- the unevenness of the energy release field in the stationary fuel loading has a maximum power unevenness coefficient of fuel assembly K $\genfrac{}{}{0ex}{}{\text{max}}{\text{g}}$ = 1.24, and the maximum coefficient of uneven power of the fuel rods does not exceed K _g K $\genfrac{}{}{0ex}{}{\text{max}}{\text{k}}$ = 1.37, which is below the design values;

- for the described active zone, it is possible to organize an in-in-out overload circuit with a low neutron leakage over a wider range, since there is a large margin in the magnitude of the non-uniformity of the energy release field. At the same time, fuel consumption improves (by ~ 15%) and the neutron fluence to the reactor vessel decreases;

- estimates of the fuel use of the described active youth in the four-year fuel cycle showed that the duration of the stationary fuel load is at least 300 eff. days, and the average fuel burnup is 40.6 MW · day / kg with an average enrichment of fuel recharge 3.74%.

Thermophysical studies of the described core for the VVER-440 reactor showed:

- the increase in hydraulic friction losses in the described core in comparison with the standard design of the core of the VVER-440 reactor does not exceed the available reserves for the pressure of the main circulation pumps. For example, the specific hydraulic resistance of the entire core when replacing standard fuel assemblies with upgraded fuel assemblies with fuel rods with a diameter of 7.0 · 10 ^-3 m and 7.8 · 10 ^-3 m increases respectively by ~ (0.02-0.025) MPa;

- the power reserve to boiling in the described core (with 10% by volume vapor content) is 1.22-1.25, and the reserves before the heat transfer crisis (DNBR) in the nominal mode do not fall below ~ 9 (without taking into account uncertainties).

A comparative analysis of the operability of the fuel rods of the standard and described active zones of the VVER-440 reactor in accidents with rupture of pipelines of the primary circuit showed:

- the temperature of the claddings of the most heat-stressed fuel rods of the described active zone in the first ~ 10 sec of an accident with rupture of the Du 500 pipeline is 300 ° C lower than that of standard fuel rods and does not exceed 600 ° C, which virtually eliminates the development of a steam-zirconium reaction;

- the operational reserves of the fuel rods according to the criterion of the tensile strength of the cladding material at emergency temperatures for the standard core are practically absent (i.e., the claddings are depressurized), and for the fuel rods of the described core, even with the maximum initial load, the margin by this criterion exceeds K> 5;

- for medium-loaded fuel rods of the described active zone, an increase in the temperature of the shells in the first 10 seconds of the process does not occur at all.

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

- increase the safety of the power plant from the 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 active zone can be used not only in VVER-440 reactors, but also in other pressurized water-cooled reactors.

Claims (6)

1. The active zone of a water-to-water power reactor containing fuel assemblies with a hexagonal fuel grid assembled from rod fuel elements, characterized in that at least one fuel assembly contains from 174 to 216 rod fuel elements having an outer and inner diameter of the shell fuel element from 7.00 · 10 ^-3 to 8.00 · 10 ^-3 m and from 5.94 · 10 ^-3 to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel lattice is selected from 0, 98 to 2.52 or fuel assembly contains from 132 to 168 rod t plovydelyayuschih elements having outer and inner diameter of the cladding 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 water the uranium ratio of the fuel lattice is selected from 1.14 to 2.76, and the ratio of the height of the core to the length of the fuel assembly is from 0.7488 to 0.7897. 2. The active zone of the water-water power reactor according to claim 1, characterized in that at least one fuel assembly contains from 180 to 210 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.00 · 10 ^{- 3} to 7.90 · 10 ^-3 m and from 5.94 · 10 ^-3 to 6.70 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.07 to 2.40 or the fuel assembly contains from 138 to 162 rod fuel elements having an outer and inner diameter of the cladding of a fuel rod from 7.90 · 10 ^-3 to 8.7 · 10 ^-3 m and from 6.70 · 10 ^-3 to 7.38 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.24 to 2.49. 3. The active zone of the water-water power reactor according to claim 1 or 2, characterized in that at least one fuel assembly contains from 186 to 204 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.10 10 ^-3 to 7.80 · 10 ^-3 m and from 6.02 · 10 ^-3 to 6.62 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate is selected from 1.17 to 2.18 or fuel the assembly contains from 144 to 156 rod fuel elements having an outer and inner diameter of a fuel rod cladding from 8.00 · 10 ^-3 to 8.6 · 10 ^-3 m from 6.79 · 10 ^-3 to 7.30 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.35 to 2.24. 4. The active zone of the water-water power reactor according to claim 1, or 2 or 3, characterized in that at least one fuel assembly contains from 192 to 198 rod fuel elements having an outer and inner diameter of the shell of the fuel element , 20 · 10 ^-3 to 7.70 · 10 ^-3 m and from 6.11 · 10 ^-3 to 6.53 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.28 to 1, 98 or the fuel assembly contains 144 rod fuel elements having an outer and inner diameter of the cladding of a fuel rod from 8.10 · 10 ^-3 to 8.5 · 10 ^{- 3} m and from 6.87 · 10 ^-3 to 7.21 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.53 to 2.15. 5. The active zone of the water-water power reactor according to claim 1, characterized in that at least one fuel assembly contains 174 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.6 · 10 ^-3 to 8 , 00 · 10 ^-3 m and from 6.45 · 10 ^-3 to 6.79 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grate is selected from 1.20 to 1.89 or the fuel assembly contains 132 core fuel elements having an outer and inner diameter of the cladding of a fuel rod from 8.40 · 10 ^-3 to 8.79 · 10 ^-3 m and from 7.13 · 10 ^-3 to 7, 47 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.43 to 2.16. 6. The active zone of the water-water power reactor according to claim 1 or 2, characterized in that at least one fuel assembly contains 180 rod fuel elements having an outer and inner diameter of the shell of the fuel element from 7.5 · 10 ^-3 up to 7.90 · 10 ^-3 m and from 6.36 · 10 ^-3 to 6.70 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.24 to 1.89 or the fuel assembly contains 138 rod fuel elements having an outer and inner diameter of a fuel cladding from 8.30 · 10 ^-3 to 8.7 · 10 ^-3 m and from 7.04 · 1 0 ^-3 to 7.38 · 10 ^-3 m, respectively, and the water-uranium ratio of the fuel grid is selected from 1.44 to 2.10.

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