RU2833667C2 - Integrated fast neutron nuclear reactor which includes protective device designed to minimize consequences of accidents with core melting - Google Patents

Abstract

FIELD: nuclear reactors.

SUBSTANCE: invention relates to an integral fast neutron nuclear reactor. Reactor comprises a main tank including an active zone and an integral primary circuit comprising a primary pump, an intermediate heat exchanger and a heat exchanger for removing residual power, as well as a step dividing the heat carrier volume into hot collector (12) and a cold collector. Core comprises a protective device consisting of at least one elementary protective device (45) designed to minimize the consequences of accidents with core melting, comprising transfer pipe (40) and through pipe (42) passing through lower support plate (30) of the core and support belt (31), opening into the lower part of the reservoir. Transfer pipe (40) comprises one or more injection nozzles and a zone of reduced thickness to increase the perforation speed in case of an accident with core melting. Device (41) for extracting molten fuel is located in the lower part of the reservoir.

EFFECT: minimization of consequences of an accident with core melting of an integral fast neutron reactor.

10 cl, 8 dwg

Description

Field of technology

The present invention relates to the field of fast neutron nuclear reactors (FNR) and, in particular, to integral reactors. Such reactors can be cooled by a coolant, usually in the form of a liquid metal, in particular sodium. Thus, a reactor of this type is called a sodium-cooled fast neutron nuclear reactor (FNR-Na) or a reactor of the SFR type, which stands for sodium-cooled fast neutron reactor. This type of sodium-cooled fast neutron nuclear reactor is part of the family of generation IV nuclear reactors.

Thus, the present invention provides an integrated fast neutron nuclear reactor that includes a safety device designed to minimize the consequences of core meltdown accidents.

Background of the invention

The operating principle of fast neutron reactors has been known for several years. Thus, a fast neutron reactor is a nuclear reactor that uses fast neutrons (with a kinetic energy greater than 0.907 MeV) as opposed to thermal neutrons (with a kinetic energy less than 0.025 eV). In addition, unlike traditional nuclear reactors, the core of a fast neutron reactor is not slowed down (i.e., there is no slowing down or thermalization of the neutrons).

Moreover, although other technologies have been studied, the vast majority of fast neutron nuclear reactors use liquid sodium as a coolant because of its high boiling point, among other things.

Sodium-cooled nuclear reactors typically include a vessel containing the core, with a core control plug located above the core. The core consists of a large number of fuel assemblies, typically surrounded by fertile assemblies, with an inner storage area containing spent fuel assemblies and reflector and absorber assemblies that act as a neutron shield. Heat is removed by circulating sodium in the core using pumping systems. This heat is transferred to the intermediate circuit through one or more intermediate heat exchangers (IE) before being used to produce steam in a steam generator (VG). This steam is then passed through a turbine to convert the heat of the steam into mechanical energy, which in turn is converted into electrical energy.

The intermediate loop contains sodium. Because of the violent reactions between sodium and steam that can occur in the event of a possible rupture of a steam generator tube, the purpose of this loop is to isolate the primary sodium (in the tank) from the steam contained in the steam generator. This architecture allows for two sodium loops: one loop, called the primary, is designed to transfer heat between the core and the intermediate heat exchanger, and the other loop, called the secondary, is designed to transfer heat from the intermediate heat exchanger to the steam generator.

All sodium reactors have common technical characteristics. The tank is sealed with a cover plate so that the primary sodium does not come into contact with the outside air. All components (heat exchangers, pumps, pipes, etc.) pass through this plate vertically so that they can be disassembled by lifting them vertically with a lifting device. The dimensions of the passage holes in this plate depend on the size and number of components. The larger the holes (in size and number), the larger the diameter of the tank.

The primary circuit can be divided into two main families. Thus, among fast neutron nuclear reactors there are "integral" reactors and "loop" reactors. It should be noted that the present invention preferably relates to integral type fast neutron nuclear reactors.

Loop reactors are characterized by the fact that the intermediate heat exchanger and the device for pumping primary sodium are located outside the reservoir. Fig. 1 shows the axial section of the design principle of a "loop" nuclear reactor on fast neutrons with sodium coolant.

In the loop reactor R shown in Fig. 1, sodium passes through the core 1, carrying away the calories produced. At the exit from the core 1, it enters zone 2 of the reservoir 3 of the reactor R. This zone 2 is usually called the "hot collector". Through the loop, pipe 4 descends into the hot collector 2 to suck up the primary sodium and send this sodium to an intermediate heat exchanger, not shown in the drawing, where it transfers heat to the secondary sodium. At the exit from the intermediate heat exchanger, the primary sodium is taken up by a pump and sent directly to the entrance of the core through pipe 5.

The main advantage of the loop design is to obtain, for a given power, a tank with a smaller diameter than the integral reactor, since it contains fewer components. Thus, the tank is easier to manufacture and, therefore, cheaper. However, the disadvantage of the loop concept is that the primary sodium leaks out of the tank, which is more complex in terms of the primary loop architecture and creates significant safety issues. Thus, the benefits associated with the reduction in size and simplification of the tank are offset by the additional costs caused by the addition of devices associated with the loop design and special means for managing possible leaks of primary sodium.

As for the integrated concept reactors, they are characterized by the fact that the intermediate heat exchangers and the means for pumping the primary sodium are located in the tank, which avoids the exit of the primary circuit of the tank and thus represents a significant advantage of this family of solutions compared to the family of solutions with loops from the safety point of view. Fig. 2 shows the design principle of an "integral" fast neutron nuclear reactor with sodium coolant in axial section.

In the integral reactor shown in Fig. 2, sodium passes through the active zone 11, carrying away the produced calories. At the exit from the active zone 11, it enters the zone 12 of the reactor tank 13, sealed by the cover plate 24. This zone 12 is usually called the "hot collector". This hot collector 12 is separated from another zone 14, called the "cold collector", by a wall 15, which has a generally cylindrical-conical shape, called a "redan". The intermediate heat exchanger 16, consisting of a bundle of tubes, not shown in the drawing, passes through the step 15. The primary sodium enters the intermediate heat exchanger 16 through the inlet windows 17, located in the hot collector 12. As it moves through the tubes, it transfers its heat to the secondary sodium and leaves the intermediate heat exchanger 16 through the windows 18 in the lower part of the intermediate heat exchanger 16, located in the cold collector 14. The secondary sodium enters the intermediate heat exchanger 16 through the pipe 28 and leaves it through the pipe 29. In the cold collector 14, the sodium is taken by the pump device 19 and directed back directly to the inlet of the core 11 through the lower support plate 30 of the core, which is used to feed the assemblies. The lower support plate 30 of the core is a pressure vessel that contains assemblies, fuel, breeder, internal storage, or acts as a neutron shield. The lower support plate 30 of the core is supported by a mechanical support structure called a support belt 31.

The circulation of sodium in the intermediate heat exchanger 16 is carried out under the action of gravity between the hot collector 12 and the cold collector 14. For reasons related to the dimensions of the intermediate heat exchanger 16 and the geometric mass, this driving pressure of sodium between the two collectors is set at approximately 2 m, which corresponds to the difference in level 20 of the hot collector 12 and at level 21 of the cold collector 14. For reasons of maximum efficiency, the components that pass through the step 15, the intermediate heat exchanger 16 and the pumping means must have the greatest possible sealing in these passages 22 and 23 in order to avoid bypassing the intermediate heat exchanger 16 by the primary sodium.

Redan 15 is an essential component of this type of reactor. Large components (intermediate heat exchangers and pumps) pass through the conical section located at the bottom of redan 15. The cylindrical section is a vertical containment shell located at the top of redan 15. Redan 15 is a component made of steel, usually mechanically welded, which is difficult to design for the following reasons: its shape and size, approximately fifteen meters; the pressure difference (approximately two meters of sodium column) to which it is subjected between the two collectors; thermomechanical stresses caused by the temperature difference between the hot and cold collectors (approximately 150°C for operating reactors); sealing stresses in the passages through redan 15 in its conical section from the intermediate heat exchangers and pumping systems. The 15 step must be sealed, since bypass of the intermediate heat exchanger 16 must be avoided, and the sealing system must allow disassembly of the components for their maintenance. However, a small bypass in the assemblies, coupled with thermal leaks through the 15 step, leads to the presence of a colder fluid medium in the lower part of the hot collector 12, which can be transferred along the structures and can cause thermomechanical stresses on the latter due to destabilization of the flow threads.

In reality, once a redan design has been selected, it cannot be easily changed a posteriori. Moreover, in addition to normal operation, designers of nuclear power reactors must take into account the reactor shutdown situation: all reactors must therefore have systems available to remove residual power from the core (EPuR in French). This residual power results from the radioactive decay of the fission products that were formed during the nuclear reactions when the reactor was started. For safety reasons and to ensure the greatest possible redundancy, these schemes must be as different as possible from the normal scheme for removing thermal energy when the reactor is started, i.e. they must not use a steam generator. The overall architecture of the residual power removal systems must be compatible with normal reactor operation. As a rule, these means for removing residual power are only activated when the reactor is shut down.

The means for removing residual power, common to most embodiments or projects, consists of several specific heat exchangers intended for the function of removing residual power. These heat exchangers 25 (see Fig. 2) are arranged vertically and pass through the cover plate 24. Due to their purpose, these heat exchangers 25 have a smaller size than the intermediate heat exchangers 16. In order to be effective, in particular in the event of a failure of the pump means 19, the primary sodium must be able to circulate by natural convection between the active zone 11 and the heat exchangers 25 for removing residual power.

In general, the reliability and efficiency of natural convection involves the definition of the simplest possible hydraulic path, which can be obtained by observing the following recommendations: the heat source (in this document, the nuclear reactor core) should be located at the bottom; the cold source (in this document, the heat exchanger intended for residual power removal) should be located at the top; the hydraulic path forming the hot column located between the outlet of the heat source and the inlet of the cold source should be as monotonic as possible (without non-monotonic altimetric variation: hot sodium should always rise upwards); the hydraulic path forming the cold column located between the outlet of the cold source and the inlet of the heat source should be as monotonic as possible (without non-monotonic altimetric variation: cold sodium should always fall downwards); the hot and cold columns should be separated to avoid mixing of the coolant between the two columns.

In the sodium-cooled reactor with an integral design, the heat exchanger intended for the removal of residual power is located either in the hot header or in the cold header. Regardless of its location, the hydraulic path of the primary sodium passes through an intermediate heat exchanger with altimetric variations in the hot and/or cold column, which worsens the hydraulic characteristics of natural convection. Thus, the heat exchanger 26 (Fig. 2) for the removal of residual power is located in the hot header 12. The hydraulic path consists of a hot column 26 and a cold column 27. The hot column 26 rises uniformly, while the altimetric variation remains monotonic. However, the cold column 27 includes a non-monotonic altimetric variation. Indeed, the sodium leaving the heat exchanger 25 must return upward to the inlet windows 17 located at the top of the hot collector before entering the intermediate heat exchanger 16 in order to enter the active zone 11 after passing through the pumping system 19. In the hot collector 12, the hot column and the cold column are not physically separated, which does not correspond to the optimal concept with respect to natural convection, since the cold sodium leaving the heat exchanger 25 can mix in the hot collector with the hot sodium entering this same heat exchanger.

For a person skilled in the art, one possible improvement would be to place the EPUR heat exchangers between the hot collector and the cold collector when passing through the redan, as in the case of the intermediate heat exchangers. This is not done because, in normal operation, this would mean creating a bypass of the EPuR intermediate heat exchangers and deteriorating the normal operating characteristics of the reactor. Thus, there is an internal technical contradiction in the heat removal circuit with technical solutions that optimize operation in a normal situation, which deteriorates operation in a residual power removal situation, and vice versa.

The last drawback of the reactors with an integral concept is related to the limitation of placing inside the latter all the components necessary for its correct operation: thus, the tank is larger than the loop reactors, and the step is heavily loaded by the passage of pumps and heat exchangers. This leads to a limitation of the size of the passive devices that could be added to the step to improve the operation due to natural convection in the situation of residual power removal by means of heat exchangers.

In fast reactors, as described above, accidents with complete core meltdown can occur. Therefore, in order to reduce the consequences of such core meltdown accidents, these reactors have a safety procedure to minimize the consequences of accidents.

Thus, the consideration of situations related to accidents with core meltdown is generalized, starting from the reactor design, due to the presence of devices to minimize the consequences of such accidents. Minimization of reactivity transients in the situation of physical destruction of the core is carried out by adding devices, in particular transfer tubes, for the extraction of burned nuclear fuel, connected to the corium catcher.

There are several tasks that need to be solved for these two devices. At the onset of an accident, there is a short-term (approximately a few seconds) movement of fuel and molten structural materials, namely corium, outside the core area using transfer tubes. This movement of corium is thus directed downwards from the neutron flux zone. The kinetics of corium removal is an important element, as it involves the extraction of corium before a large melt pool is formed, which will have a significant potential for restoring criticality due to large-scale compaction of the core.

The second task involves radially confining the destroyed core so as not to cause the meltdown to spread to other assemblies or to the internal storage, which is the area around the periphery of the core containing the spent fuel assemblies.

Furthermore, the third task involves long-term movement, or movement over a period exceeding several minutes, of the corium to a catcher plate located at the bottom of the main tank beneath the core. The catcher reduces the risk of criticality by moving the corium, and also ensures cooling and containment of the corium.

In practice, the design arrangement that is generally selected to solve the above-mentioned problems is the implementation in the active zone of specific devices for extracting and then catching the corium. These devices are pipes 40 for transferring the corium C coming from the fission zone ZF, the active zone 11, connected to the corium catcher 41, as shown in Fig. 3, which illustrates the principle of extracting the corium through transfer pipes. The active zone 11 includes several transfer pipes 40, the arrangement of which is optimized for unloading the corium.

In the event of an accident with a complete meltdown of the core 11, these pipes 40 must quickly ensure the discharge of the melted core into the corium trap 41 located in the lower part of the tank 13. These protective devices have only one function, associated with minimizing the consequences of an accident with a meltdown of the core, and must be as invisible as possible in other operating situations of the reactor.

From an architectural point of view, the transfer tubes 40 are hollow structures that provide channels for the flow of corium through the lower support plate 30 of the active zone and the support belt 31.

During a core meltdown accident, a number of operational problems arise. Indeed, regardless of the initiator that triggers the core meltdown accident situation, the fuel assembly is subject to at least one element failure after the fuel meltdown (the fuel assembly typically contains the sodium feed bed, the fuel elements inside the hexagonal tube (HT), the fuel element support piece, the upper neutron shield (UNS) and the supported head). The steel structural elements of the fuel element (cladding, wires, plugs) quickly melt, and thus the first protective barrier, namely the sealing of the cladding, is lost. The defect quickly spreads to the entire fuel element bundle at the assembly scale. The corium, at a temperature of approximately 2700°C, consisting of fuel and molten structural materials, contacts the inner walls of the hexagonal tube, made of 4 to 5 mm thick steel, which melt and become perforated in a few seconds. In this way, the corium bath moves in the active zone towards the other fuel assemblies, which leads to complete meltdown.

The transfer tubes allow to define and limit the radial advance of the corium in the core. Each transfer tube is designed to melt to open a path for the corium to flow inwards under the action of gravity from the neutron flux zone, which helps to reduce reactivity. Thus, the corium trajectory continues through the lower support plate, and then through the support belt to arrive vertically in line with the corium catcher plate located at the bottom of the vessel.

From the point of view of the accident development sequence, it is important that the removal of corium from the flow zone occurs quickly enough to limit the consequences of the accident. In order to ensure the correct trajectory of the corium flow into the trap, it is also necessary to limit the obstacles present inside the transfer pipe.

There are also several problems associated with the normal operation of the reactor. Indeed, since the corium transfer tubes are only supposed to operate during a core meltdown accident, they must be as unnoticeable as possible during normal reactor operation. In particular, they must not affect the operation in terms of mechanical, neutronic and thermal-hydraulic characteristics.

From a mechanical point of view, the corium transfer tubes are a core structure that is subjected, like other components, to mechanical loads at various stages of the life cycle, namely during production, transportation, maintenance, irradiation. Moreover, the mechanical structure of the transfer tubes has an irradiation degree and, therefore, damage as strong as that of the fuel assemblies (up to 100 dPa). The predominant mechanical loads that the design of the transfer tubes must withstand are the contact stresses between the components in the core. These manifest themselves in the hexagonal tube as compaction, bending and torsion forces that occur during normal operation and during an earthquake.

From a neutron point of view, the transfer tubes are hollow structures located in the core. The impact of the presence of these potentially empty elements on the neutron characteristics of the core must be acceptable. In particular, this means that there are criteria for the criticality of the core due to the reduction in the volume of fissile material and the smoothing of the power distribution due to the presence of "holes" in the core.

From a thermal-hydraulic point of view, the presented concept of the transfer tube should not lead to a direct connection of the hot collector with the cold collector. Thus, a pressure difference of approximately 150 mbar between these two collectors will lead to the formation of a hot sodium flow, leaving the hot collector under excess pressure and descending inside the tubes into the cold collector. However, the injection of hot sodium (550°C) into structures (in the lower support plate, support belt, etc.) located in cold sodium (400°C) is undesirable from the point of view of their mechanical strength and their aging (due to thermomechanical stresses caused by temperature gradients). In particular, it is necessary to be able to justify a service life of the mentioned structures equal to the service life of the nuclear power plant. Moreover, it is known that the free surface of sodium in the hot collector is one of the main sources of gas evolution in the hot collector. These bubbles dissolve in the sodium of the hot collector under the effect of high temperature, with the risk of re-bubbling of this gas (bubbling phenomenon) and transformation of the bubbles if it passes into the cold collector. However, the presence of gas in the cold collector must be limited to avoid any risk of gas passing into the assemblies, i.e. a phenomenon that could lead to failure of cooling of the fuel elements and to the risk of a sharp increase in power. Therefore, the transfer tube should not lead to sodium injection between the collectors in order to limit mechanical damage to the structures or to promote gas evolution in the cold collector.

Thus, the following requirements and constraint functions emerge. In the sequence of the development of an accident with a core meltdown, it is necessary to ensure rapid removal of corium from the neutron flux zone and then into the corium trap. In normal reactor operation, it is necessary to ensure tightness between the hot and cold collectors. Two constraint functions are to withstand mechanical loads at different stages of the life cycle and not to degrade the neutron characteristics of the core.

It should be noted with reference to the state of the art that the first FNR-Na in France, called Rapsodie and Phénix, did not include devices in the core designed to minimize the consequences of accidents with a meltdown of the core. The safety demonstration of FNR-Na reactors was studied using qualified modeling tools in specific experiments.

The FNR-Na SuperPhénix includes a corium catcher at the bottom of the vessel, but does not include devices to minimize the consequences of accidents in the core.

Following the nuclear accidents at Chernobyl and then at Fukushima, safety measures were tightened in order to create a set of requirements that the new generation of reactors, namely Generation IV, had to meet.

These new safety measures could lead to the addition of special devices in the core, designed to minimize the consequences of accidents with a meltdown of the core and to ensure the transfer of corium to the trap. The first proposed idea, namely the addition of transfer pipes in the core, was not entirely satisfactory, as it created a thermal-hydraulic bypass problem between the hot collector and the cold collector.

The same is true for proposals for fully or partially closed transfer tubes, since although they do meet the sealing requirements between the headers, there is a risk that they will promote gas release in the core (closure at the top may lead to the risk of gas accumulation inside the transfer tube) or block the flow of corium (case of "melting" closure at the bottom). Indeed, fusible membrane devices pose a risk to the reliability of the meltdown during the accident sequence with a core meltdown, which is associated with difficulties in modeling and qualification.

In addition, the patent literature also proposes devices for minimizing the consequences of accidents with core meltdown, in particular, those designed for corium transfer.

For example, Japanese patent application JP H10-227884 A relates to the concept of an absorber assembly (control rod) that plays a role in minimizing the consequences of accidents with a meltdown of the core. A hexagonal tube has openings that open into an inter-assembly space limited by adjacent fuel assemblies. The base of the assembly also opens into this inter-assembly space and thus allows cooling of the absorber rod during normal operation. The base is closed at its end. During an accident with a meltdown of fuel elements, corium melts the hexagonal tube of the fuel assembly and falls into the inter-assembly space. It then flows by gravity to the base of the absorber assembly, where it is stored. The present application describes in detail the aspects of moving corium from the neutron flux zone and reducing reactivity. However, in addition to the fact that the advancement of corium may be significantly slowed down in the small cross-section of the inter-assembly space, this concept does not provide a device for transporting the corium to a trap located below the core.

Moreover, the US patent application US 2012/0201342 A1 relates to a concept of a fuel assembly and a FNR reactor with a function of removal and extraction of corium during a core meltdown accident. This assembly includes a solid body in its portion below the fuel elements, wherein the body is perforated to form a series of holes allowing for cooling through the outside of a hexagonal tube during a core meltdown accident sequence with the presence of corium inside the assembly. The fuel elements are modified to improve the behavior during a core meltdown accident. The end of the base of the assembly is closed. A sealed container, which may contain an absorbent material, is located below the lower support plate vertically in alignment with the assembly, acting as a corium trap. This application aims to move and then remove the corium and prevent it from moving radially in the core. However, several disadvantages arise: the end of the base and the top of the container are fusible walls that must disappear upon contact with the corium, and fusible membrane devices have poor reliability; the integration of a corium removal device into a fuel assembly supplied with a compressed sodium flow poses a risk of obstructing the normal flow of corium if the primary pumps are not stopped; the thickening of the assembly body at its bottom significantly increases the amount of irradiated waste and complicates its manufacture. In addition, it is necessary to ensure that a number of holes do not interfere with the thermal hydraulics during operation.

In addition, the FAIDUS (for Fuel Assembly with Internal Channel Structure) assembly principle of the Japan Atomic Energy Agency can be mentioned. Therefore, in order to limit the consequences of a core meltdown accident situation with a CDA (for Core Disintegration Accident) scenario, the JSFR (Japanese Sodium-Cooled Fast Reactor) reactor integrates an accident mitigation device called FAIDUS into the design of its fuel assemblies. Due to its reactivity characteristics, the heating from the CDA and the induced boiling of sodium generate a local power peak (a positive reactivity effect in the case of sodium drain) in the JSFR assembly. Therefore, the FAIDUS design integrated into the JSFR fuel assemblies aims to quickly extract the liquid fuel to avoid too much compaction and return to criticality, which is too high during the occurrence of CDA events. FAIDUS is not a patented concept, but has been the subject of several international reports.

There are two alternative designs for FAIDUS assemblies. The standard solution, which has been the subject of extensive studies (calculations and experiments), proposes a small corium flow channel located in the corner of a hexagonal tube. The molten fuel is discharged through it towards the upper part of the assembly due to the driving force of the first early power step initiated in the assemblies (internal pressure). The upper parts of the JSFR core contain catcher trays to receive the ejected fissile material. The second design has a corium discharge channel located in the center of the fuel element bundle. The molten fuel is to be discharged through it by gravity through the lower part of the assembly. This design is not given much importance, on the one hand, due to manufacturing difficulties, and on the other hand, due to the fact that the technical limitations for creating a corium transport to the lower structures have been determined to be strong.

It should be noted that the amount of fuel ejected via FAIDUS is not sufficient to avoid other subsequent criticality returns in the rest of the core. And a second fuel unloading scenario is required via the empty structures of the absorbent control rods (for the control rod guide tube (CRGT)) remaining in the upper position. Moreover, questions arise about the proper passage of the molten fuel through the vacuum generation system, which prevents it from passing through the base of the CRGT. Finally, even if the FAIDUS concept has the advantage of slightly disturbing the neutronic characteristics of the core, the integration of the corium bleed channel into the fuel assembly supplied by the pressurized sodium flow poses the risk of impeding the proper corium flow if the primary pumps are not stopped.

In conclusion, the corium removal strategy using the upper part of the FAIDUS fuel assemblies depends on the neutronic and thermal-hydraulic characteristics of the JSFR core. This strategy cannot be directly applied to another FNR-Na core.

The essence of the invention

One object of the invention is to at least partially satisfy the above-mentioned needs and to eliminate the disadvantages associated with the embodiments of the prior art.

More specifically, the invention is aimed at improving the behavior of the FNR-Na reactor core in the event of an accident with a complete meltdown of the core. In particular, it is aimed at reducing the consequences of an accident with a meltdown of the core while maintaining the integrity of the second barrier (the reactor vessel) and the sealing of the third safety barrier (the reactor building and the support plate) due to mechanical effects caused during the accident (e.g., energy release caused by the expansion of vapors of melted and then evaporated materials), as well as due to thermal effects caused during the accident (e.g., erosion of structures by corium). In general, the invention is aimed at maintaining the limits of radioactive emissions dangerous to the population and returning to a safe state after an accident.

Thus, the objective of the present invention, according to one of its aspects, is to implement an integral fast neutron nuclear reactor cooled by a liquid metal coolant, comprising:

- a main reservoir suspended from a protective plate and including a volume of coolant and an active zone of the reactor, wherein the active zone includes a plurality of fuel assemblies installed in a lower support plate of the active zone of a nuclear reactor supported by a support belt,

- a primary circuit integrated into the main tank and comprising at least one primary pump, at least one intermediate heat exchanger for removing the power produced by the core during normal operation, and at least one heat exchanger for removing residual power,

- an internal structure called a redan, dividing the volume of the coolant into at least two zones, forming a hot collector at the outlet of the active zone and a cold collector at the outlet of the mentioned at least one intermediate heat exchanger,

characterized in that the reactor core additionally includes a protective device designed to minimize the consequences of accidents with core meltdown, wherein said protective device consists of at least one elementary protective device located near at least one fuel assembly of the core, for transferring melted fuel to a cold collector at the bottom of the main reservoir,

wherein said at least one elementary protective device comprises a transfer tube, hollow and removable, installed in the lower support plate of the active zone through a first opening in the lower support plate of the active zone, and a hollow through tube in fluid communication with the transfer tube,

the through pipe passes through the lower support plate of the core from the first opening to the second opening located between the lower support plate of the core and the support belt, and through the support belt from the second opening to the third opening of the support belt opening into the cold collector at the bottom of the main tank,

the transfer tube comprises one or more injection nozzles located inside the tube and located at least partially on a section of the tube installed in the lower support plate of the active zone, while oriented towards the hot collector, wherein the injection nozzles are intended for feeding coolant into them from the lower support plate of the active zone,

the transfer tube includes at least one zone of reduced thickness on its outer wall, located in the fission zone of at least one fuel assembly, to increase the rate of perforation of the outer wall in the event of an accident with melting of the core and to ensure the flow of molten fuel from the said at least one fuel assembly into the transfer tube,

and the reactor further includes a device for extracting molten fuel, located in the cold collector at the bottom of the main tank and located in alignment with the third opening.

The fast neutron nuclear reactor according to the present invention may further include one or more of the following features, taken individually or in any technically possible combinations.

The transfer tube may have the external features of traditional fuel assemblies.

The coolant is preferably supplied through the injection nozzles to the lower support plate in the same manner as for conventional fuel assemblies. Moreover, the installation of the transfer tube in the lower support plate of the core is preferably carried out instead of and at the location of the conventional fuel assembly. Like the latter, it is easily removed for removal and replacement.

The transfer tube may advantageously include a head providing the ability to support it, a base mounted on the lower support plate of the active zone, and a hollow body with a hexagonal cross-section between the head and the base.

The base may have a cross-section smaller than the cross-section of the hollow body. The upper part of the base may be funnel-shaped to facilitate the passage of molten fuel. The lower part of the base may be open to allow the flow of molten fuel to the through pipe.

Moreover, the head may include a central channel having a cross-section smaller than the cross-section of the hollow body in order to reduce the risks of ejection of molten fuel towards the hot collector, and a reduction in the cross-section at the boundary with the hollow body to facilitate the removal of gases during immersion in the coolant.

In addition, said at least one zone having a reduced thickness can be obtained by milling the outer wall of the transmission pipe.

The mentioned at least one zone having a reduced thickness may be located on the hollow body of the transmission pipe, and may not be located at the corners of the hexagonal cross-section of the hollow body in order to maintain the mechanical rigidity of the transmission pipe.

In addition, injection nozzles can be located at the base of the transfer pipe.

The transfer pipe may include at least one injection nozzle having an annular shape, in particular a single injection nozzle having an annular shape.

Alternatively, the transfer tube may include a plurality of injection nozzles uniformly distributed within the transfer tube, in particular three injection nozzles forming the vertices of an equilateral triangle.

In addition, the active zone may include a protective device consisting of a plurality of elementary protective devices intended to minimize the consequences of accidents with meltdown of the active zone, wherein some of the elementary protective devices are located along the periphery of the active zone containing fuel assemblies, and another part of the elementary protective devices are located in the central part of the active zone containing fuel assemblies.

Each elementary safety device can be attached to a fuel assembly in a one-to-one manner.

Brief description of the drawings

The invention may be better understood by reading the following detailed description of non-limiting embodiments thereof and by studying the schematic and partial figures of the accompanying drawings, in which:

- Fig. 1 illustrates in axial section the design principle of a fast-neutron nuclear reactor with sodium coolant;

- Fig. 2 illustrates in axial section the design principle of an “integral” fast-neutron nuclear reactor with sodium coolant;

- Fig. 3 illustrates in axial section the principle of fuel removal, in particular corium, using transfer pipes in the event of an accident with a complete meltdown of the active zone of a fast neutron nuclear reactor;

- Fig. 4 illustrates in axial section the principle of pumping out fuel, in particular corium, through an elementary reactor protection device according to the invention in the event of an accident with a complete meltdown of the reactor core;

- Fig. 5 illustrates in partial perspective a detail of the implementation of the hollow body of the transfer tube of the elementary protective device of the reactor according to the invention;

- Fig. 5A - section along line AA (Fig. 5);

- Fig. 5B – section along line BB (Fig. 5);

- Fig. 6 illustrates in axial section an embodiment of an elementary reactor protection device according to the invention;

- Fig. 6A – section along line AA (Fig. 6);

- Fig. 6B – section along line BB (Fig. 6);

Fig. 7 also illustrates in axial section another embodiment of the elementary reactor protection device according to the invention,

Fig. 7A is a section along line AA in Fig. 7;

Fig. 7B – section along line BB in Fig. 7; And

Fig. 8 illustrates, in a top view, an example of the design of the reactor core according to the invention.

In all these drawings, identical references may designate the same or similar element.

Moreover, the various parts shown in the drawings are not necessarily to the same scale in order to make the drawings more readable.

Detailed description of the invention

Figs. 1-3 have already been described above with reference to the prior art and in the general context of the present invention.

Next, with reference to Figs. 4-8, the features and design solutions of the invention will be described, which make it possible to minimize the consequences of a hypothetical accident with a complete meltdown of the active zone of a fast neutron nuclear reactor, in particular, with sodium cooling (FNR-Na), by transferring corium into the trap 41 of the lower part 13 of the tank through a set of transfer pipes 40.

It should be noted that the elements specific to the fast neutron nuclear reactor R, already described above with reference to Figs. 1-3, will not necessarily be described again. Thus, the same references designate the same elements. Moreover, in the examples described below, the liquid metal coolant of the reactor R is sodium, although this choice is in no way limiting.

Moreover, it should be noted that the inventors have worked on a solution for cooling the core by natural convection with a direct hydraulic path descending from the hot collector to the cold collector, presented in the French patent application FR 3 053 827 A1. However, the present invention relates to a completely different technical problem related to the logic of minimizing the consequences of accidents with core meltdown by hydraulically locking the empty transfer tube.

The safety procedure, which includes the rapid removal of melted fuel from the core to avoid sudden power surges during a core meltdown accident, is shared by the international community, in particular by Japan. However, the technical means chosen for its implementation depend on the neutron-physical and thermal-hydraulic characteristics of the core and its behavior during accident scenarios. Thus, the principle of the present invention, developed in particular with respect to the ASTRID reactor core, differs from the solutions of the state of the art and, in particular, from the principles adopted by Japan in the FAIDUS fuel assembly concept presented above.

Indeed, one of the special features of the ASTRID reactor is the innovative design of its heterogeneous CFV core (low reactivity core in case of sodium leakage). Thus, a power surge caused by a sodium leakage from an assembly in an emergency is unlikely. In such a scenario, the challenge for the ASTRID mitigation approach is to avoid a significant compaction of the fuel distribution in the decaying core and thus prevent an immediate criticality that is too high. To this end, additional protective devices must be provided in the CFV core to ensure that the molten fuel, namely the corium, is removed from the core through flow channels sufficiently quickly. In practice, the flow channels can either be integrated into existing assemblies (fuel elements, control rods, etc.) or formed in completely separate assemblies adjacent to the fuel assemblies. The present invention is based on the second possibility, namely the use of special transfer pipes 40 so as not to affect the operation of other units and to maximize the cross-section of the channels for removing corium.

Moreover, the peculiarity of these objects is to facilitate the axial transfer under the action of gravity of the liquid fuel through the lower structures of the reactor (the lower support plate and the support belt). Thus, part or all of the molten fuel in the core can be collected only by means of these devices in the lower part of the tank in the corium catcher 41 intended for this purpose.

Thus, the invention is based on the use of a hydraulic blocking system in the transfer pipes, which does not have the disadvantages compared to previous solutions (fuse wall, bypass between hot and cold collectors, risk of gas release, transfer of corium under the primary flow, etc.) and, on the other hand, on the optimization of the design of the pipes and their installation in the active zone, taking into account the task of minimizing the consequences of accidents.

Fig. 4 illustrates in axial section the principle of removing fuel, in this case in the form of corium, through a protective device consisting of at least one elementary protective device 45 of the reactor R according to the invention in the event of an accident with a complete meltdown of the active zone 11 of the reactor.

According to the invention, the active zone 11 of the reactor R includes a protective device consisting of at least one elementary protective device 45, designed to minimize the consequences of a possible accident with meltdown of the active zone.

This elementary protective device 45 is adjacent to the fuel assembly 43 of the active zone 11 and allows the transfer of molten corium to the cold collector 14 in the lower part of the main reservoir 13.

The elementary protective device 45 includes a transfer tube 40, resembling a removable hollow structure of an assembly type, which is placed or installed in the lower support plate 30 of the active zone through the first opening 50 on the lower support plate 30 of the active zone.

It also includes, in continuation of the transfer pipe 40, a hollow through pipe 42, in fluid communication with the transfer pipe 40. This through pipe 42 passes through the lower support plate 30 of the active zone from the first opening 50 to the second opening 54, located between the lower support plate 30 of the active zone and the support belt 31, and through the support belt 31 from the second opening 54 to the third opening 56 of the support belt 31, in order to open into the cold collector 14 at a low pressure, approximately 0 bar, vertically in accordance with the device 41 in the form of a catching plate for extracting molten corium, located in the cold collector 14 in the lower part of the main tank 13.

The function of the lower support plate 30 of the active zone is to supply the assemblies with cold sodium under high pressure or approximately 3 bar. For this purpose, the tubes of the protective shell and the base of the assemblies have radial openings facing each other. In addition, the function of the support belt 31 is to support the active zone 11 and the lower support plate 30 of the active zone.

The transfer tube 40 is a unit designed for the removal of corium. Its external architecture is preferably identical to the architecture of the other units in the active zone 11 and has the same coupling means.

The transfer tube 40 includes a head 60 in the upper part, designed with the possibility of supporting it, a base 62, allowing the tube to be supported in the lower support plate 30 of the active zone, and a hollow body 61 with a hexagonal cross-section between the head 60 and the base 62.

The base 62 differs slightly from that of the other units of the active zone 11. It has a cross-section smaller than the cross-section of the hollow body 61. In addition, the upper part 62a of the base 62 can have the shape of a funnel to facilitate the passage of corium into the base 62, which has a smaller cross-section. Moreover, the lower part 62b of the base 62 is open to allow the flow of corium to flow to the through pipe 42 and, thus, to the catcher plate 41 through the lower support plate 30 of the active zone and the support belt 31. This base 62 additionally contains injection nozzles 53, which are described below.

Moreover, the head 60 includes a central channel 60a having an internal cross-section smaller than the cross-section of the hollow body 61 in order to reduce the risks of corium removal through the upper part at the beginning of the sequence of the development of an accident with a core meltdown. In addition, the lower part of the head 60 includes cross-sectional narrowings 60b to facilitate gas removal during the immersion of the assembly in sodium. The robust design of the steel head 60 also acts as a neutron shield, limiting axial neutron leaks.

The hollow body 61 does not contain any internal structure and is filled with sodium during normal operation. During a core meltdown accident, this hollow body 61 is perforated by meltdown and forms, first of all, a channel for the flow of corium under the action of gravity to the base 62.

With regard to the above requirement regarding the possibility of rapid unloading of corium from the neutron flux zone and taking into account that it is impossible to influence the kinetics of the corium flow inside the transfer tube 40, at the beginning of the accident scenario a gain in time can be obtained by accelerating the perforation of the hollow body 61 by corium coming out of the adjacent fuel assemblies 43. A faster increase in the temperature of the hollow body 61 is achieved by reducing its thermal inertia by reducing its thickness.

Thus, as shown in Fig. 5, 5A and 5B, the transfer tube 40 includes at least one zone 65 of reduced thickness on its outer wall 66, located opposite the fission zone ZF of at least one fuel assembly 43, to increase the rate of perforation of the outer wall 66 in the event of an accident with a meltdown of the core and to ensure the flow of corium from at least one mentioned fuel assembly 43 into the transfer tube 40.

Indeed, this design solution is applicable only to the area of the transfer tube 40 that is most exposed to contact with the corium, i.e. to the hollow body 61 facing the division zone ZF. Of course, the axial arrangement on the hollow body 61 of the zone(s) of reduced thickness 65 can be optimized by means of suitable calculations.

From a production point of view, one or more zones of reduced thickness 65 can be obtained by milling the outer wall 66 of the transfer tube 40, in particular by milling its outer surfaces. The advantage of this solution is that a completely standard tube is used at the beginning, whereby the milling operation does not present any technical difficulties. The milling is carried out only on the surfaces, and not at the angles of the hexagonal cross-section of the hollow body 61, so as to maintain good mechanical rigidity and thus fulfill the above-mentioned function of counteracting mechanical loads during the life cycle stages.

The milling solution is preferable to the one that involves, for example, manufacturing and then joining by welding a section of pipe of reduced thickness to other sections of pipe of standard thickness. Indeed, manufacturing such pipes and performing their welds is associated with numerous difficulties.

Moreover, according to the present invention and with reference to Figs. 6-7B, the transfer tube 40 comprises one or more injection nozzles 53 located inside the base 62 and located at least partially in the hollow body 62 of the tube 40 installed in the lower support plate 30 of the core, while oriented towards the hot collector 12, wherein the injection nozzles 53 are intended to supply coolant to them from the lower support plate 30 of the core.

In this document, sodium under "high pressure" is supplied to these nozzles 53 from the lower support plate 30 of the core, similar to the fuel assemblies 43. The size of these nozzles 53 in diameter and number is determined in such a way that the back pressure created by this injection compensates as accurately as possible for the loss of pressure of the intermediate heat exchangers during operation, which reflects the pressure difference between the collectors 12, 14. Thus, the flow rate entering the hot collector 12 in the upper part of the transfer pipe 40 is almost zero and is neutralized in the event of a stop of the primary pumps, while the reduced pressure difference between the collectors 12, 14 is zero. However, the velocity of the flow exiting the lower part of this transfer pipe 40 corresponds to the velocity of the flow in the injection nozzles 53. This exiting flow is deflected into the cold collector 14 through the through pipes 42, the lower support plate 30 of the active zone and the support belt 31.

The back pressure created by the injection of sodium supplied by the primary pumps blocks the flow of sodium descending in the hexagonal transfer pipe 40 at all pumping speeds. Thus, during normal operation of the reactor R, i.e., when there is neither a hydraulic bypass between the collectors 12, 14, nor an upper closure of the pipe (no gas release), the above-formulated requirement for ensuring tightness between the hot collector 12 and the cold collector 14 is fulfilled.

Similarly, in the case of a core meltdown accident sequence, and if the primary pumps are still running, the resulting hydraulic blockage does not create an increasing coolant flow rate capable of counteracting the transfer of corium to the catcher plate 41. This helps to meet the above-mentioned requirement for the ability to quickly unload corium from the neutron flux zone. The fact that the transfer tube 40 remains open and is not closed at its base by a fusible wall also helps to meet this requirement due to the absence of blocking or slowing down the corium flow to the catcher plate 41. It is also possible to have an annular injector 53, described below with reference to Figs. 7-7B, which minimizes the obstruction of the corium flow inside the base 62.

Thus, in Fig. 6, 6A and 6B, a specific embodiment of the transfer tube 40 is shown, having three injection nozzles 53 and comprising a circular internal cross-section for the base 62 and a hexagonal external shape for the hollow body 61, as for standard fuel assemblies. These three nozzles 53 form the vertices of an equilateral triangle.

Moreover, Fig. 7, 7A and 7B show another embodiment which differs from the embodiments shown in Fig. 6, 6A and 6B in that the three nozzles 53 located in cross-section at the vertices of an equilateral triangle are replaced by a single injection nozzle 53 having an annular shape and being in direct contact with the inner wall of the base 62. This annular shape of the injection nozzle 53 is shown, in particular, in Fig. 7A.

In addition, Fig. 8 shows, in a top view, an example of the design of the active zone 11 of the reactor R according to the invention.

Thus, the active zone 11 includes a protective device consisting of a plurality of elementary protective devices 45 intended to minimize the consequences of accidents with melting of the active zone. Part of the elementary protective devices 45 is located along the periphery of the active zone 11 containing fuel assemblies 43, and another part of the elementary protective devices 45 is located in the central part of the active zone 11 containing fuel assemblies 43.

Thus, most of the transfer tubes 40 are located on the periphery of the core in order to limit the radial advance of the molten bath. Some transfer tubes 40 are also located in the central part of the core in order to remove part of the corium as early as possible. The installation of the central tubes 40 must correspond to neutron studies aimed at determining the number and distance between the absorber rods, as well as for the control and shutdown of the reactor R.

The choice of the number and spacing of the transfer tubes 40 inside and around the core is based on the results of calculations of core meltdown accident scenarios using a code simulating the movement of corium inside the core 11. The best compromise is sought between the limited number of tubes 40 so as not to degrade the neutron performance while allowing a satisfactory volume of corium to be removed in a short time.

The same is true for the optimal determination of the number of central tubes 40 (earliest possible unloading depending on neutron characteristics) and the number of tubes 40 around the periphery of the core 11 (limited number depending on the spatial arrangement of the corium to be unloaded).

It should also be noted that the installation of pipes 40 for transferring corium instead of fuel assemblies 43 in the active zone 11 leads to a decrease in the power of the latter. In order not to worsen the neutron characteristics of the active zone 11, it is possible to compensate for the loss of power by increasing the number of fuel assemblies 43 and, consequently, the diameter of the active zone 11. This option is undesirable, since it will lead to a direct increase in costs associated with an increase in the dimensions of the reservoir 13 and the reactor building, which are associated with an increase in the costs of manufacturing, installing and dismantling fuel assemblies 43.

Thus, an alternative way to increase the power is not to increase the number of fuel assemblies 43, but rather to increase the concentration of plutonium (Pu) or other fissile material in the oxide fuel (U, Pu)O ₂ used. However, this must be done carefully to meet the maximum Pu concentration limit set by radiation feedback and production capacity. Neutron studies are required to ultimately ensure that the maximum linear power criterion in the fuel elements is met.

Of course, the invention is not limited to the above-described embodiments. Specialists in the field of technology can make various modifications to it.

In particular, the liquid metal coolant may be other than sodium, such as lead or lead-bismuth, among others.

Claims (19)

1. An integrated fast-neutron nuclear reactor (R), cooled by a liquid metal coolant, comprising: - a main reservoir (13) suspended from a protective plate (24) and including a volume of coolant and an active zone (11) of the reactor (R), wherein the active zone (11) includes a plurality of fuel assemblies (43) installed in a lower support plate (30) of the active zone, supported by a support belt (31), - a primary circuit integrated into the main tank (13) and comprising at least one primary pump (19), at least one intermediate heat exchanger (16) for removing the power produced by the active zone (11) during normal operation, and at least one heat exchanger for removing residual power (25), - an internal structure (15), called a redan, dividing the volume of the coolant into at least two zones, forming a hot collector (12) at the outlet of the active zone (11) and a cold collector (14) at the outlet of the mentioned at least one intermediate heat exchanger (16), characterized in that the active zone (11) of the reactor (R) additionally includes a protective device designed to minimize the consequences of accidents with a meltdown of the active zone, wherein said protective device consists of at least one elementary protective device (45) located near at least one fuel assembly (43) of the active zone (11), for transferring molten fuel to a cold collector (14) in the lower part of the main reservoir (13), wherein said at least one elementary protective device (45) comprises a transfer tube (40), hollow and removable, installed in the lower support plate (30) of the active zone through the first opening (50) in the lower support plate (30) of the active zone, and a hollow through tube (42) in fluid communication with the transfer tube (40), a through pipe (42) is located passing through the lower support plate (30) of the active zone from the first opening (50) to the second opening (54) located between the lower support plate (30) of the active zone and the support belt (31), and through the support belt (31) from the second opening (54) to the third opening (56) of the support belt (31), opening into the cold collector (14) in the lower part of the main reservoir (13), the transfer tube (40) comprises one or more injection nozzles (53) located inside the tube (40) and located at least partially on a section (40a) of the tube (40) installed in the lower support plate (30) of the active zone, while oriented towards the hot collector (12), wherein the injection nozzles (53) are intended to supply coolant to them from the lower support plate (30) of the active zone, the transfer tube (40) includes at least one zone of reduced thickness (65) on its outer wall (66), located in the fission zone (ZF) of at least one fuel assembly (43), to increase the rate of perforation of the outer wall (66) in the event of an accident with a meltdown of the core and to ensure the flow of molten fuel from the said at least one fuel assembly (43) into the transfer tube (40), and the reactor (R) further includes a device (41) for extracting molten fuel, located in the cold collector (14) at the bottom of the main tank (13) and located in alignment with the third opening (56). 2. The reactor according to claim 1, characterized in that the transfer tube (40) includes a head (60) that provides the ability to support it, a base (62) installed in the lower support plate (30) of the active zone, and a hollow body (61) with a hexagonal cross-section between the head (60) and the base (62). 3. The reactor according to claim 2, characterized in that the base (62) has a smaller cross-section than the cross-section of the hollow body (61), in that the upper part (62a) of the base (62) has the shape of a funnel to facilitate the passage of molten fuel, and in that the lower part (62b) of the base (62) is open to ensure the flow of molten fuel to the through pipe (42). 4. A reactor according to claim 2 or 3, characterized in that the head (60) includes a central channel (60a) having a smaller cross-section than the cross-section of the hollow body (61) in order to reduce the risk of ejection of molten fuel in the direction of the hot collector (12), and a narrowing (60b) of the cross-section at the boundary with the hollow body (61) to facilitate the removal of gas during immersion in the coolant. 5. A reactor according to any of the preceding claims, characterized in that said at least one zone of reduced thickness (65) is obtained by milling the outer wall (66) of the transfer tube (40). 6. A reactor according to any one of claims 2-5, characterized in that said at least one zone of reduced thickness (65) is located on the hollow body (61) of the transfer tube (40) and is not located in the corners of the hexagonal cross-section of the hollow body (61) in order to maintain the mechanical rigidity of the transfer tube (40). 7. A reactor according to any one of paragraphs 2-6, characterized in that the injection nozzles (53) are located in the base (62) of the transfer pipe (40). 8. A reactor according to any of the preceding claims, characterized in that the transfer tube (40) includes at least one injection nozzle (53) having an annular shape, in particular a single injection nozzle (53) having an annular shape. 9. A reactor according to any one of claims 1 to 7, characterized in that the transfer tube (40) includes a plurality of injection nozzles (53) uniformly distributed in the transfer tube (40), in particular three injection nozzles (53) forming the vertices of an equilateral triangle. 10. A reactor according to any of the preceding paragraphs, characterized in that the active zone (11) includes a protective device consisting of a plurality of elementary protective devices (45) designed to minimize the consequences of accidents with meltdown of the active zone, wherein part of the elementary protective devices (45) is located along the periphery of the active zone (11) containing fuel assemblies (43), and another part of the elementary protective devices (45) is located in the central part of the active zone (11) containing fuel assemblies (43).