RU2032946C1 - Water-cooled nuclear reactor - Google Patents

Abstract

FIELD: nuclear technology. SUBSTANCE: water- cooled nuclear reactor has active zone with spherical nuclear fuel with protective coat and light-water moderator-heat transfer agent surrounder with reflector and placed into pressure vessel. Enrichment of nuclear fuel in each elementary cell of active zone amounts to 5-25% and its volumetric share is from 0.015 to 0.76. Volumetric share of water in cell amounts to 0.1-0.25. Remaining volume is filled with expelling agent produced from material having smaller microsection of absorption of thermal neutrons than that of light-water moderator. EFFECT: enhanced operational efficiency. 9 cl, 9 dwg, 3 tbl

Description

 The invention relates to nuclear engineering and can be used to improve water-cooled nuclear reactors.

 Currently, the efforts of the developers are mainly focused on the creation of technical solutions that exclude the possibility of an accident, as a result of which fuel elements (fuel elements) can be destroyed (melted), but unfortunately this cannot be guaranteed yet. It is unlikely, but nonetheless possible events leading to a complete loss of primary coolant for existing water-cooled reactors such as VVER and RBMK containing an active zone with rod fuel elements and a light-water coolant moderator surrounded by a reflector and placed in the reactor vessel, and for reactors based on new technologies, for example, HTGR, containing an active zone with spherical fuel rods, in the graphite matrices which are microfuel dispersed, and a helium coolant surrounded by graphites m reflector and placed in the high pressure housing, as well as VTRS containing an active zone with spherical fuel rods and a molten-salt coolant-moderator surrounded by a reflector and placed in the reactor body, will inevitably cause destruction of fuel rods (microfuel) as a result of heating of the active zone due to the residual energy release. In such an accident, depressurization of the primary circuit is possible, and the retention of radioactive fission products can only be ensured by complex and expensive engineering means of localization, the operability of which is very difficult to reliably substantiate in the situations for which they are designed, since experimental investigation of severe accidents is difficult, and computational modeling , due to the abundance and complexity of the processes, very approximately.

A special place in a number of promising projects is occupied by high-temperature gas-cooled modular reactors containing an active zone with spherical fuel rods, in the graphite matrix of which microfuel nuclear fuel is dispersed, made in the form of spherical particles 0.3-0.8 mm in size and coated with several layers of protective shells to prevent the exit of radioactive fission products from nuclear fuel, and a gas coolant surrounded by a reflector and placed in a high-pressure housing, in which even with complete loss of heat the carrier, the residual energy is removed from the core due to the thermal conductivity of the spherical structure of the core, the reflector and the reactor vessel, and fission fragments do not come out of the fuel rods. The decisive role in such a high resistance to severe accidents is played by microtel reactors of this type. Moreover, the protective shells of the microfuel can not only achieve high operating temperatures, but also do not worsen the neutron balance and do not complicate the process of fuel regeneration. While rod fuel rods, which are widely used in WWER and RBMK reactors and containing fuel cores consisting of individual tablets or cylindrical rods enclosed in a pipe shell made of zirconium alloy, can be operated at lower temperatures, since if the fuel cladding temperature exceeds 650 ^{° C,} they begin to swell due to reduction of its strength and increase the pressure differential between the interior fuel rods and the coolant, and at a temperature of about lochek above 1000 ^C. begins the chemical reaction of zirconium with steam to produce hydrogen and zirconium dioxide.

 However, now it is hardly possible to expect that radical developments: a new fuel rod, new coolant and new equipment of nuclear power plants can find practical implementation in the next 15-20 years.

 In view of the foregoing, it seems promising to consider the possibility of using microfuel in reactors with a light-water coolant, with the maximum use of the experience in the development, manufacture and operation of such reactors. They are basic in the nuclear energy industry of our country and the entire industry is mainly focused on their production.

 Known water-cooled nuclear reactor containing an active zone formed from technological channels for spherical fuel elements, and a light-water coolant surrounded by a reflector and placed in the housing, and in the upper and lower parts of each technological channel, respectively, loading and unloading devices for spherical fuel elements are made. During operation of the reactor inside the technological channels, a pseudo-boiling layer of spherical fuel elements is arranged from the bottom up through the channel. The energy release is equalized in the reactor core by changing the porosity in the process channel by changing the flow rate of the light-water coolant and, therefore, the channel power.

The disadvantages of this technical solution include:
possible disturbances in the homogeneity of the fluidized bed of spherical fuel particles and related fluctuations in reactivity;
positive steam reactivity coefficient;
the lack of measures in the reactor design to ensure guaranteed heat removal from spherical fuel rods during dehydration of the core.

Also known is a water-cooled nuclear reactor containing an active zone with spherical fuel rods and a liquid coolant, which are proposed as H ₂ O, D ₂ O and an organic compound pumped from the bottom up with a speed ensuring the formation of a fluidized bed in the entire volume of the active zone surrounded by a reflector and enclosed in the case. In this case, continuous mixing of ball fuel rods occurs. A characteristic feature of the core is an automatic decrease in reactivity in the event of a pump failure or when the speed of the pump motor is exceeded. Spherical fuel elements in this case are concentrated, respectively, either in the lower part of the active zone or in the upper, where they are carried away by the coolant flow.

 The claimed technical solution is aimed at solving the problem of improving the nuclear and radiation safety of water-cooled nuclear reactors during design and beyond design basis accidents.

 To solve this problem, in a water-cooled nuclear reactor containing an active zone with spherical nuclear fuel with a protective coating and a light-water moderator-coolant, surrounded by a reflector and placed in a pressure casing, nuclear fuel with enrichment of 5-25% and s is placed in each unit cell of the active zone the volume fraction in the cell is 0.015-0.76, and the volume fraction of water in the cell is 0.1-0.25, and the remaining volume of the cell is filled with a displacer made of material with a micro neutron absorption cross section c, less than that of a light-water moderator.

 Light water can be used as a reflector, and microfuel and / or modules, in the graphite matrix of which microfuel, which are connected by a displacer in at least one block, can be used as nuclear fuel, and the structure of microfuel and / or modules in the block and / or their shape is made with the possibility of the formation of a bulk structure of nuclear fuel, limited by the inner surface of at least the lower part of the pressure housing during melting of the displacer. In addition, a ball fuel rod can be used as a module, and aluminum and / or its alloys can be used as a propellant.

 Another difference is that graphite elements are dispersed in the matrix of the block, moreover, the graphite elements, as well as microfuel elements and modules are provided with antioxidant protective coatings.

 A further difference is that the inner surface of at least the lower part of the pressure housing is provided with a protective coating of a material that does not interact with the displacer melt. In addition, the difference lies in the fact that the block has at least one vertical channel for passage through the block of coolant and / or absorbing element.

 It should be noted that the implementation of the active zone of a water-cooled vessel-mounted nuclear reactor in which nuclear fuel with enrichment of 5-25% and with its volume fraction in the unit cell of 0.015-0.76 is placed in each unit cell, the volume fraction of water in the cell being 0, 1-0.25, and the remaining volume of the cell is filled with a displacer made of a material with a micro neutron absorption cross section less than that of a light-water moderator (for example, Al, its alloys, graphite) allows the negative reactivity coefficients (my nostnogo, hollow and steam) and thus eliminate the destruction of the core, i.e. improve nuclear safety of the reactor.

 The above requirements are explained as follows. With an enrichment of less than 5%, the criticality condition of the reactor will not be achieved. With enrichment of more than 25%, positive reactivity coefficients are possible, which does not ensure the stability of the power plant. With a volume fraction of fuel of less than 0.015, it is difficult to achieve the criticality condition of the reactor and a positive void reactivity coefficient appears. The volume fraction of spherical fuel above 0.76 cannot be achieved due to the impossibility of a denser packing of spherical bodies in a given volume. With a volume fraction of water of less than 0.1, it is difficult to ensure heat removal from the fuel elements and, all other things being equal, it is difficult to ensure criticality conditions. With a volume fraction of water of more than 0.25, the reactor will have a positive void reactivity coefficient. A displacer is necessary to provide a volume fraction of water in the range of 0.1-0.25. The above boundaries were obtained after a set of computational studies conducted using the WIMS-D4, CRISTALL, DRAKON, SFERA, TEPLO, CTAPT-3, FLY programs.

 It should be especially noted that the use of microfuel fuel rods and / or modules made, for example, in the form of spherical fuel rods, with an antioxidant protective coating, which are dispersed in at least one matrix of a block made of a material selected as a displacer, moreover, with a nuclear fuel structure in the block, which provides the possibility, in an accident with a complete loss of coolant, of the formation of a bulk structure of spherical fuel inside the lower part of the reactor vessel or melting Therians propellant would ensure the guaranteed heat from the nuclear fuel through the reactor vessel and thereby prevent overheating microfuel above their maximum allowable operating temperature and thereby retain fission products within a fuel even if complete dewatering of the core, i.e. to increase the radiation safety of a water-cooled reactor.

 Thus, in the proposed technical solution, the task is to increase the safety of water-cooled reactors by making the fullest possible use of the intrinsic safety properties of the reactor (such as negative reactivity coefficients, high heat-storage ability of fuel rods and unit, the inability to form secondary critical masses, chemical passivity of core materials to a friend and with respect to the environment and the passive cooling mechanism of fuel rods and / or microfuel during design design basis accidents) can only be ensured by the totality of essential features set forth in the claims.

 Figure 1 shows a schematic diagram of the loop layout of a water-cooled nuclear reactor with forced circulation of the coolant; figure 2 is a schematic diagram of the integrated layout of a water-cooled nuclear reactor with natural circulation of light-water coolant; figure 3 schematic diagrams of the integrated layout of a water-cooled nuclear reactor with forced circulation of the coolant; figure 4 is a structural diagram of a block fuel assembly of a water-cooled reactor; figure 5 is a longitudinal section of a block fuel assembly; Fig.6 is a cross section of a cylindrical block fuel assembly; 7 is a cross section of a prismatic block fuel assembly; Fig.8 is a variant of the scheme for damping the chaotic filling of ball fuel rods through the reactor vessel after melting the material of the displacer blocks; Fig. 9 is a variant of the scheme for damping the chaotic backfill of microfuel and spherical fuel elements through the reactor vessel after melting of the material of the displacer blocks.

 A water-cooled nuclear reactor contains an active zone 1, surrounded by a light-water reflector 2 and placed in a pressure casing 3 (see Fig. 1-3). As a coolant-moderator and reflector use light water. In FIG. 1 is a schematic diagram of the loop layout of a water-cooled nuclear reactor with forced circulation of the coolant by pumps 4 through the core 1 and steam generators (heat exchangers) 5. The reactor, pumps, steam generators and other equipment are located in pressurized boxes 6 of the reactor building of the station. Figure 2 shows a schematic diagram of the integrated layout of a water-cooled nuclear reactor with natural circulation of light-water coolant. In this case, the steam generators 5 are placed inside the pressure vessel 3, and the reactor itself is placed in the safety case 7. Figure 3 shows the schematic diagrams of the integrated layout of a water-cooled nuclear reactor with forced circulation of the coolant by pumps 4 through the active zone 1 and steam generators 5. Moreover, in the left half the drawing shows the placement of the pump 4 under the active zone 1, and in the right half above the active zone 1.

 The active zone of the inventive water-cooled nuclear reactor can be made in the form of a single monoblock or recruited from block fuel assemblies (see figure 4). Moreover, in each unit cell of the active zone, the enrichment of nuclear fuel is 5–25% and its volume fraction is 0.015–0.76. In addition, the volume fraction of water in the cell is 0.1-0.25, and the remaining volume is filled with a displacer made of a material with a micro neutron absorption cross section less than that of a light-water moderator (graphite, aluminum and / or its alloys).

 The indicated limits are explained by the fact that when going beyond them it is difficult to ensure the criticality condition of the reactor and it is possible to obtain a positive void reactivity coefficient in some operating modes. So, for example, when enriching 5% of the volume fraction of nuclear fuel 0.3 and the water fraction in the unit cell of the active zone 0.15, the criticality of the reactor is achieved, but there is no reactivity margin for burnup (i.e., the reactor will not work in the energy mode). With a fuel fraction of less than 0.015, a positive void reactivity coefficient is obtained, which does not satisfy the requirements for a safe reactor. It is generally impossible to obtain a volume fraction of microtuels greater than 0.76 from geometrical restrictions, since this value is the limit for spherical bodies placed in a given volume. A displacer is necessary in order to provide a volume fraction of water in the range of 0.1-0.25, as well as in order to bind the fuel assembly fragments to each other. The best material for this is aluminum. It has a high thermal conductivity of ≈200-250 W / m˙ K, a small neutron absorption cross section of ≈0.15 bar and a melting point of ≈900 K. This, on the one hand, ensures its operability in nominal conditions at T ≈600K, and with on the other hand, it makes it possible to fragment the zone with passive means at a temperature level of ≈900-1000 K, which is much lower than the maximum operating temperature of microfuel 1900 K.

 In FIG. 4 shows a structural diagram of a block fuel assembly of the inventive water-cooled reactor. In this assembly, microfuel 8 (designed for high-temperature gas-cooled reactors) and / or modules 9, in the graphite matrix 10 of which microfuel 8 are dispersed, are most used. Nuclear fuel rods designed for high-temperature gas-cooled reactors are used as modules 9. Micro-fuel rods and / or modules (ball fuel rods), in turn, are dispersed into a monoblock matrix 11 made of a material with a micro-section of thermal neutron absorption less than that of a light-water coolant. As such a material, aluminum and / or its alloys can be used. Moreover, in order to reduce the absorption of thermal neutrons into the monoblock matrix material, graphite elements 12 can be dispersed between microfuel and / or spherical fuel rods (see Figs. 4 and 5). Graphite elements, microfuel and spherical fuel rods are provided with antioxidant protective coatings (not shown in the drawing). In addition, the structure of microfuel and modules dispersed in the monoblock matrix is made with the possibility of its scattering and the formation of chaotic filling during melting of the aluminum matrix of the block, due to the residual energy release in nuclear fuel. The spherical shape of microfuel and spherical fuel rods most fully provides this condition, and graphite elements 12, made, for example, also spherical, do not interfere with the dispersion of fuel during the melting of aluminum or its alloys.

 Figure 5 shows a longitudinal section of a block fuel assembly, in the center of which there is a vertical channel 13 for passage through a block of coolant and / or absorbing element 14. Figure 6 shows a cross section of a cylindrical block fuel assembly, and Fig. 7 is a cross section of a prismatic assemblies equipped with spacer ribs 15. In the core, the block fuel assemblies can be combined into a single unit, for example, bondage (not shown), which also provides for their suspension in the reactor vessel, in the inner surface of which, at least in its lower part, is provided with a protective coating 16 made of a material that does not interact with the molten aluminum and / or its alloys.

 On Fig and 9 shows options for chaotic backfilling of nuclear fuel at the bottom of the reactor vessel 3 after melting of aluminum due to residual energy in the fuel during dehydration of the active zone. Moreover, Fig. 8 shows that the aluminum melt 17 is located between the spherical fuel rods 9 lying randomly on the bottom of the reactor vessel, and Fig. 9 shows that the spherical fuel rods 9 are partially immersed in the aluminum melt 17, and the microfuel 8 is completely immersed in the melt bottom of the reactor vessel.

 As a specific example of the inventive water-cooled nuclear reactor, two options are considered in which cylindrical blocks with spherical fuel rods in an aluminum matrix are used as fuel assemblies.

 The first version is a medium-power reactor with a thermal capacity of 1000 MW and an electric capacity of 300 MW. The reactor can be placed in the reactor vessel VVER-1000. The second option is an AST-type reactor with a thermal capacity of 500 MW, designed for heat supply.

 Consideration of these two options of the inventive water-cooled reactor is of interest for the following reasons. So, the first option is a power plant for the fuel compositions under consideration, placed in the VVER casing, which ensures the retention of fission products in microfuel in the case of a hypothetical accident associated with a complete loss of coolant. The second variant is a reactor for heat supply stations, which, due to less stressful working conditions, acquires the properties of maximum safety. Moreover, it should be noted that the lower heat intensity of the core allows you to expect in the nominal mode the possibility of heat removal due to the natural circulation of the coolant (see figure 2), thereby increasing its reliability in normal and emergency conditions.

 In the table. 1 shows the main characteristics of a medium-cooled water-cooled reactor. The active zone of this reactor consists of 500 cylindrical block fuel assemblies (see Fig. 4), made up closely to each other in the form of a hexagonal lattice. The height of the active zone is 4 m. The dimensions of the zone are determined from the permissible level of energy intensity of the fuel assemblies (FA), which guarantees the non-melting of aluminum in all operating modes. The fuel assembly in height is composed of four blocks (see Fig. 5 and 6). In the lower block there are 120 ball fuel rods and small graphite balls, the space between which is filled with aluminum. The calculations of the distribution of temperature fields in the fuel assembly show that the permissible energy level is about 6 kW / ball, taking into account the uncertainty factors, it is accepted: the average energy voltage is 4 kW / ball.

For a given energy intensity level of 2 MW / fuel assembly, which corresponds to an average core energy intensity of 17 MW / m ³ , for a reactor with a capacity of 1000 MW (t), we obtain an equivalent core radius of 2.2, m.

It should be noted that the choice of the optimal fuel loading is associated with the determination of its characteristics, ensuring the non-positivity of the reactivity coefficients and good characteristics of the fuel cycle. For fuel loads of more than 40 g UO ₂ / ball with an enrichment of U ²³⁵ above 8% and a water fraction of about 15%, all reactivity coefficients, temperature, power and steam become negative, which suggests the possibility of creating a fairly effective reactor control system.

It is customary to evaluate fuel efficiency by the criterion of specific consumption of natural uranium. In the table. Figure 2 shows the calculation results for a stationary mode of operation (steady-state overload mode) of the specific consumption of natural uranium for various charges and enrichment of fuel. From the table. Figure 2 shows that the minimum values of the consumption of natural uranium ≈0.220-230 kg U _eats / MW day (the highest fuel cycle efficiency) are achieved when loading fuel of about 60 g / ball and enriching 8-10% in U ²³⁵ (the same value for the specific consumption of natural uranium for VVER-1000 reactors, R = 0.240 kg U _{eats /} VMt.day, RBMK R 0.240 kg U _{eats /} MW.day (see table 1).).

For the considered option of a medium-cooled water-cooled reactor, a load of 60 g UO ₂ / ball was selected, with an initial fuel enrichment of 8% in U ²³⁵ . The fuel campaign for these parameters is 1170 days; the average burnup of the unloaded fuel is 81,000 MW. day / t. The total fuel load in the reactor is 14.4 tons From the above data it can be seen that the proposed version of the reactor from the point of view of the fuel cycle is comparable in characteristics with modern reactors with a water coolant.

Consider the second option, a water-cooled reactor for AST. This reactor uses the same fuel assemblies as for the medium power reactor discussed above. Reducing the power of the reactor can reduce the specific energy density to 13.7 MW / m ³ . While maintaining the same core height of 4 m, the equivalent radius of the core is 1.7 m. In this case, 310 fuel assembly units with a unit power of 1.6 MW are placed in the reactor.

Due to the lower power density and a lower operating temperature coolant supply to the melting temperature of aluminum displacer compared to the previous embodiment chitsya increased by approximately 150 ° ^C. The characteristics of the fuel cycle do not differ from the data reported for the previous embodiment.

 The main characteristics of a water-cooled reactor for AST are given in Table 3.

 The inventive water-cooled nuclear reactor operates as follows. The light-water coolant pump 4 is pumped from the bottom up through the active zone 1 both between the blocks and through channels 13 made in the blocks where it is heated, and in the steam generator 5 it transfers heat to the secondary coolant and enters the pump inlet 4. Now we consider the course of a severe accident associated with with a complete loss of coolant in the inventive water-cooled nuclear reactor.

 As noted above, the main advantage of the proposed reactor, compared with other water-cooled nuclear reactors, is its ability to retain fission products in microfuel in severe accident conditions associated with loss of heat removal from the core. This is achieved due to the proposed composition of the core and the design of the blocks, which make it possible to organize the development of a severe accident in such a way that under no circumstances the maximum temperature of the microfuel does not exceed the permissible value, which ensures the integrity of the microfuel and the retention of almost all fission products in them.

The pattern of events in a severe accident is as follows:
in case of loss of coolant (light water), a large negative reactivity is introduced (about 30%), the reactor becomes subcritical and the only source of energy release will be the residual heat from fission products;
the core is heated, aluminum and / or its alloys in the blocks are melted and flows to the bottom of the reactor vessel. Nothing holds ball spherical fuel rods (microfuel) in the blocks, and they also spill out to the bottom of the reactor vessel;
molten aluminum and spherical fuel rods (microfuel) are collected at the bottom of the reactor vessel and the residual energy is discharged through the wall of the reactor vessel through heat conduction into the external environment by any known method, for example, by pumping air, water, etc. along the reactor vessel ( see Figs. 8 and 9) or the gulf of the space between the reactor vessel and the water safety vessel, and also due to the placement of any known heat exchange devices (not shown in the drawing) in this space.

 Let us consider this scheme in more detail and evaluate the working conditions in which microfuel can be at different stages of the accident.

 Loss of coolant.

 We assume that the loss of water in the core occurs instantly. From the point of view of estimating temperatures in microfuel, this is the worst option. The loss of a water moderator leads to a decrease in reactivity of about 30%; the reactor becomes subcritical. Such a change in reactivity is provided by the choice of parameters of the fuel composition, namely: in each unit cell of the active zone, the enrichment of nuclear fuel is 5–25%; its volume fraction is in the range 0.015–0.75; the volume fraction of water in the cell under normal reactor operation is 0 , 1-0.25, the remaining volume is filled with a displacer made of a material with a micro-section of thermal neutron absorption less than that of a light-water moderator, for example, graphite, aluminum or its alloys. Moreover, the reactor becomes subcritical with the loss of any amount of water from the amount in the normal operating mode of the reactor 10-25%. Thus, as a result of the loss of coolant (light water), the only heat source is residual energy.

 Core heating.

Due to the residual energy release, heating of the core begins. The aluminum-graphite block has a high heat capacity and for its heating from nominal temperatures to the melting temperature of aluminum and for the melting of aluminum it is necessary to spend a sufficiently large amount of energy. According to estimates, this amount of energy is equal to
W heating + melting ≈137 MJ / kg. This amount of energy, due to the residual energy of fission products, will be allocated for
τ heating + melting ≈0.3 hours

 As the aluminum melts, it will flow out of the fuel assembly, releasing ball fuel rods and accumulate with them at the bottom of the reactor vessel. The experience of studying the motion of ball fuel rods in VTGR reactors shows that there is the possibility of organizing this process of core fragmentation in an ordered manner so that the non-uniformity of the thickness of the ball bed layer on the bottom of the shell is minimal.

Ultimately, when all the matrix material of the blocks is melted, a layer of random spherical filling of fuel rods partially immersed in aluminum forms on the bottom of the reactor vessel (see Fig. 8). Based on the ratio of the densities of graphite and aluminum 2/3 (γ _gr 1.0 g / cm ³ , γ _AI = 2.7 g / cm ³ ), we find that in 2/3 layers of ball filling between the fuel rods can be filled with aluminum. This will require a small amount of aluminum (for example, 1/2 of the amount of aluminum in the core) to be placed, for example, on the bottom of the reactor vessel, so that its quantity and the aluminum that drains out after melting from the core is sufficient to fill 2/3 of the free volume of ball filling . If this is not foreseen, then only half of the ball fill will be immersed in molten aluminum. The indicated circumstance, the presence of aluminum between the balls in the backfill is significant, since in the presence of aluminum the thermal conductivity will be significantly higher compared to the case when gas is located between the spherical fuel rods and heat is exchanged by radiation and heat conduction through the contact points of the spherical fuel rods at 150-200 ^о С .

 Removal of residual heat through the wall of the reactor vessel.

To calculate the temperature of the fuel at this stage of the development of a severe accident, the temperature fields were calculated according to the height of the ball filling of fuel rods. The level of residual energy release was determined taking into account the delay for the core melting time. The decay of the residual energy release over time was set using tabular values
Q _ost Q _nom 0.0665 F (τ + τ heating +
+ melting)
where Q _{nom is the} core energy density in normal mode,
F (τ) is the decay function of the residual energy release, which was calculated taking into account the recommendations and was given in the table; linear interpolation was carried out in the intervals between the table values.

 When calculating the reactor for AST, it was believed that only half of the ball filling is immersed in aluminum, i.e. only aluminum from the core. For a reactor of gray power, calculations were performed when 1/2 and 2/3 of the ball-bed layer were immersed in aluminum.

For ball filling immersed in aluminum, the effective coefficient of thermal conductivity was determined by the following formula:
λ _eff. λ _AI AIε _AI + λ _g (1 ε _g ), where λ _g thermal conductivity of small graphite balls;
λ _AI thermal conductivity of aluminum;
ε _AI , ε _gr volume fraction of aluminum and graphite.

 For ball filling without aluminum, we used data on the thermal conductivity coefficient in the ball filling of VTGR reactors.

The calculation was carried out for the planar one-dimensional case. As boundary conditions on the outer wall of the reactor vessel, the condition was accepted
T _article 100 ^about C (temperature of boiling water).

 At the upper (free) boundary of the ball filling, it was believed that heat was removed only through radiation, i.e.

λ ▽ Т С σ _о (Т ⁴ Θ ⁴ ), where С is the reduced emissivity characterizing the intensity of heat transfer by radiation at the free boundary;
σ _about 5.66 ˙10 ^-8 W / m ² to ⁴ Stefan-Boltzmann constant;
T is the temperature at the border;
Θ temperature of the external system.

The value of the coefficient C in the calculations ranged from 0 to 0.5. The thickness of the entire layer of chaotic ball filling was determined based on the total number of balls in the core and the area of the bottom of the body. For the considered embodiment of reactors it is obtained that the layer thickness of the ball filling ≈1,5 m. The results of these calculations are presented in Table 4, where different heat transfer conditions on the free surface backfill maximum temperature determined microfuel share microfuel a temperature above 1600 ^{° C} and time after which the maximum temperature is reached.

As seen from the presented results of calculations for both reactors embodiment modes when microfuel maximum temperature does not exceed 1600 ^{° C,} it can be adequately implemented. In addition, it should be borne in mind that in the review, there are rather conservative approximations from the point of view of the formation of heat removal conditions, heat removal through the side walls of the housing was not taken into account; in the lower part, near the case, aluminum will be in a solid state, and its thermal conductivity is 3-4 times greater than that of the molten one, as a result of heat removal will be more intense than in the calculations, it was assumed that the active zone is drained instantly and after stopping all the energy of the reactor is spent only on the melting of the blocks, however, there is reason to believe that in reality this process will be longer and naturally the subsequent heating of the core will be less intense. The inventive water-cooled nuclear reactor with nuclear fuel in the form of microfuel, designed for high-temperature gas-cooled nuclear reactors, is based on current technologies that solve the problem of nuclear and radiation safety of power plants in the case of the most severe, hypothetical accident associated with the complete loss of the primary coolant .

It should also be noted that the characteristics of the fuel cycle of the proposed reactor are not inferior to modern reactors with a water coolant and provide a flow rate of natural uranium at the level of 0.230 kg U _eats / MW / day, while the specific fuel loading is 14.4 kg / MW.

In addition, the achievable core energy level for the proposed block design is 17–20 MW / m ³ (equivalent to 4–6 kW / ball) and the reactor power corresponding to this energy level (1000 MW t) allows placing its core in the reactor vessel VVER-1000.

Claims (9)

 1. A WATER-COOLED NUCLEAR REACTOR containing an active zone surrounded by a reflector and placed in a pressure housing with spherical nuclear fuel with a protective coating and a light-water moderator-coolant, characterized in that in each unit cell of the active zone the volume fraction of nuclear fuel enriched from 5 to 25% is 0.015 0.76, the volume fraction of water is from 0.1 0.25, and the remaining volume is filled with a displacer made of a material with a micro neutron absorption cross section smaller than that of a light-water moderator.  2. The reactor according to claim 1, characterized in that light water is used as a reflector, and microfuel and / or modules connected by a propellant into at least one block are used as nuclear fuel, and microfuel is dispersed into the graphite matrix of which, and the microfuel structure and / or modules in the block and / or their execution form are selected from the condition of ensuring free fall on the bottom of the housing during melting of the displacer.  3. The reactor according to claim 2, characterized in that the module is made in the form of a ball fuel rod.  4. The reactor according to paragraphs. 1 and 2, characterized in that as a displacer used aluminum and / or its alloys.  5. The reactor according to claims 1 to 4, characterized in that the protective coating of the nuclear fuel is antioxidant.  6. The reactor according to claims 1-5, characterized in that the graphite elements are dispersed in the block matrix.  7. The reactor according to claims 1 to 6, characterized in that an antioxidant protective coating is applied to the surface of the graphite elements.  8. The reactor according to claim 2, characterized in that at least one vertical channel is made in the block for passage through the block of coolant and / or absorbing element.  9. The reactor according to claims 2 and 4, characterized in that the inner surface of at least the lower part of the pressure housing is provided with a protective coating of a material that does not interact with aluminum and / or its alloys.

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