Integrated fast neutron reactor including safety devices to mitigate core meltdown accidents

Abstract

This invention relates to an integral fast neutron reactor including safety devices to mitigate core meltdown accidents. The main objective of this invention is to provide an integral fast neutron nuclear reactor comprising: a main tank including a core; an integrated main loop including a main pump, an intermediate heat exchanger, and a heat exchanger for discharging excess power; and a convex bastion dividing the volume of heat transfer fluid into a heat collector and a cold collector. The core includes safety devices consisting of at least one basic safety device specifically designed to mitigate core meltdown accidents. The at least one basic safety device includes a delivery pipe and a hollow through-tube positioned to pass through the core grid and a strong backplate to the bottom of the tank. The delivery pipe includes one or more injection nozzles and a reduced-thickness region to increase the puncture velocity in the event of a core meltdown accident. Equipment for recovering molten fuel is located at the bottom of the tank.

Description

Technical Field

[0001] This invention relates to the field of fast neutron nuclear reactors (FNR), and more particularly to monolithic fast neutron nuclear reactors. Such reactors can be cooled by a heat transfer fluid, typically in the form of a liquid metal, and more specifically, by sodium. Therefore, such reactors are called sodium-cooled fast neutron nuclear reactors (FNR-Na) or SFR-type reactors representing sodium-cooled fast reactors. This type of sodium-cooled fast neutron nuclear reactor is part of the "generation four" nuclear reactor family.

[0002] Therefore, this invention proposes an integrated fast neutron nuclear reactor that includes safety devices specifically designed to mitigate core meltdown accidents. Background Technology

[0003] The operating principle of fast neutron nuclear reactors has been known for several years. Therefore, a fast neutron nuclear reactor is a nuclear reactor that uses fast neutrons (with a kinetic energy greater than 0.907 MeV) instead of thermal neutrons (with a kinetic energy less than 0.025 eV). Furthermore, unlike conventional nuclear reactors, the core of a fast neutron nuclear reactor is not moderated (neutrons are neither slowed down nor heated).

[0004] Furthermore, 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.

[0005] Sodium-cooled nuclear reactors typically consist of a vessel in which the core is positioned, with a core control plug above it. The core includes numerous fuel assemblies, usually surrounded by breeder assemblies, an internal storage area containing used fuel assemblies, and reflector and absorber assemblies that act as neutron shielding. Heat extraction is achieved by circulating sodium through the core via a pumping system. This heat is transferred to the intermediate loop via one or more intermediate exchangers (IEs) before being used to generate steam in the steam generator (VG). The steam then passes through a turbine to convert steam into mechanical energy, which is in turn converted into electrical energy.

[0006] The intermediate loop includes sodium. Because a violent reaction could occur between sodium and steam in the event of a potential pipe rupture in the steam generator, the purpose of this loop is to isolate the primary sodium (in the tank) from the steam contained within the steam generator. This architecture reveals two sodium loops: one, called the primary loop, is responsible for transferring heat between the reactor core and the intermediate heat exchanger, and the other, called the secondary loop, is responsible for transferring heat from the intermediate heat exchanger to the steam generator.

[0007] All sodium reactors share common technical features. The tank is sealed by a top plate to prevent the main sodium from contacting the outside air. All components (heat exchangers, pumps, piping, etc.) pass vertically through this plate, allowing for disassembly by vertically lifting these components using a hoisting device. The size of the passageways in this plate depends on the size and number of components. The larger the number and size of the passageways, the larger the diameter of the tank.

[0008] The main loop can be arranged according to two main families. Therefore, in fast neutron nuclear reactors, there exist "monolithic" reactors and "loop" reactors. It should be noted that, preferably, the present invention relates to a monolithic fast neutron nuclear reactor.

[0009] A loop reactor is characterized by the fact that the intermediate exchanger and the equipment for pumping the main sodium are located outside the tank. Figure 1 The design principle of a "loop-type" sodium-cooled fast neutron nuclear reactor is shown in axial cross-section.

[0010] exist Figure 1 In the loop reactor R, sodium passes through core 1 to carry away the generated heat. At the output of core 1, the sodium enters region 2 of tank 3 in reactor R. This region 2 is commonly referred to as the "heat collector". Through the loop, pipe 4 descends into heat collector 2 to draw up primary sodium and guide it to an intermediate exchanger (not shown), where the sodium transfers heat to secondary sodium. At the output of the intermediate exchanger, the primary sodium is pumped and delivered directly to the core input via pipe 5.

[0011] For a given power output, the main advantage of the loop design is that it results in a smaller tank diameter than that of an integral reactor, as the tank contains fewer components. Therefore, the tank is easier to manufacture and thus less expensive. However, the disadvantage of the loop design is that it involves the main sodium exiting the tank, which is more complex from the perspective of the main loop architecture and poses significant safety concerns. Therefore, the additional costs incurred through the addition of equipment associated with the loop design and special devices for managing potential leaks of the main sodium outweigh the benefits associated with the smaller tank size and easier tank manufacturing.

[0012] For reactors with an integrated design, the reactor is characterized by the fact that the intermediate exchanger and the device for pumping the main sodium are located in the tank, which makes it possible to avoid the main loop leaving the tank, and thus constitutes a significant safety advantage of this family of designs over those with loops. Figure 2 The design principle of the "monolithic" sodium-cooled fast neutron nuclear reactor is shown in axial cross-section.

[0013] exist Figure 2In the monolithic reactor, sodium passes through the core 11 to carry away the generated heat. At the outlet of the core 11, sodium enters region 12 of the reactor vessel 13, which is sealed by a top plate 24. This region 12 is commonly referred to as the "heat collector." The heat collector 12 is separated by a wall 15 from another region 14, referred to as the "cold collector," which has an overall cylindrical-conical shape called a "fortress." An intermediate exchanger 16, consisting of a bundle of tubes (not shown), passes through the fortress 15. Primary sodium enters the intermediate exchanger 16 through an inlet window 17 located in the heat collector 12. As the primary sodium moves along the tubes, it releases its heat to secondary sodium and exits the intermediate exchanger 16 through a window 18 located in the lower portion of the intermediate exchanger 16 within the cold exchanger 14. Secondary sodium enters the intermediate exchanger 16 through a conduit 28 and exits through a conduit 29. In the cold collector 14, sodium is absorbed by the pumping device 19 and fed directly back to the input end of the core 11 via the core grid 30, which is used to supply components. The core grid 30 is a pressurized chamber into which components, fuel, breeder, internal storage, or devices used as neutron shields are assembled. The core grid 30 is supported by a mechanical support structure called a strong backplate 31.

[0014] Sodium flows through the intermediate exchanger 16 by gravity between the heat collector 12 and the cold collector 14. Due to the size and geometry of the intermediate exchanger 16, the sodium drive head between the two collectors is set to approximately 2m, corresponding to the difference between the liquid level 20 in the heat collector 12 and the liquid level 21 in the cold collector 14. For maximum efficiency, components passing through the convex fortress 15, the intermediate exchanger 16, and the pumping device must have as many seals as possible at these channels 22 and 23 to prevent the main sodium from bypassing the intermediate exchanger 16.

[0015] The bulging bunker 15 is an essential component of this type of reactor. Large components (intermediate exchanger and pumps) pass through a conical section located in the lower part of the bulging bunker 15. A cylindrical section forms the vertical shroud located in the upper part of the bulging bunker 15. The bulging bunker 15 is a steel component typically manufactured by mechanical welding, which is difficult to design due to: the shape and size of the bulging bunker, approximately fifteen meters in length; the pressure difference between the two collectors (approximately two meters for a sodium column); the thermomechanical stress caused by the temperature difference between the hot and cold collectors (approximately 150°C for current reactors); and the sealing stress at the passages through the conical section of the bulging bunker 15 via the intermediate exchanger and pumping system. Because bypassing the intermediate exchanger 16 must be avoided, the bulging bunker 15 must be sealed, and the sealing system must allow for component disassembly for maintenance. However, the slight bypass at the component, combined with the heat leakage through the convex fortress 15, results in a cooler fluid at the bottom of the heat collector 12, which can be carried along multiple structures and cause thermomechanical stress on the structure due to the instability of the flow path.

[0016] In fact, once a convex fort design is chosen, it cannot be easily modified retrospectively. Furthermore, in addition to normal operation, nuclear reactor designers must consider reactor shutdown scenarios: therefore, all reactors must have a usable system (referred to in French as EPUR) responsible for venting the excess power from the core. This excess power comes from the radioactive decay of fission products generated during the nuclear reaction when the reactor starts up. For safety reasons and to ensure maximum possible redundancy, these loops must be as different as possible from the normal loops to vent thermal power during reactor startup; that is, these loops must not use steam generators. The overall architecture of the systems used for venting excess power must further be compatible with normal reactor operation. Typically, these devices for venting excess power are only activated when the reactor is shut down.

[0017] What is common to most embodiments or schemes is that the means for discharging excess power includes several specific switches, which are dedicated to the function of discharging excess power. These switches 25 (see...) Figure 2 The exchangers are vertical and pass through the top plate 24. Due to their function, the exchangers 25 are smaller than the intermediate exchangers 16. In order to function, especially in the event of a failure of the pumping unit 19, the main sodium must be able to flow through natural convection between the core 11 and the exchangers 25 to discharge excess power.

[0018] Generally, the reliability and effectiveness of natural convection involves defining the simplest feasible hydraulic path while adhering to the following recommendations: the heat source (in this case, the reactor core) must be located in the lower section; the cold source (in this case, an exchanger dedicated to venting excess power) must be located in the upper section; the hydraulic path forming the hot column must be located between the output of the heat source and the input of the cold source and must be as monotonic as possible (no non-monotonic height changes: the hot sodium must always rise); the hydraulic path forming the cold column must be located between the output of the cold source and the input of the heat source and must be as monotonic as possible (no non-monotonic height changes: the cold sodium must always fall); the hot and cold columns must be separated to avoid mixing of the heat transfer fluids between the two columns.

[0019] In sodium-cooled reactors with an integral design, exchangers dedicated to discharging excess power are located within either the heat collector or the cold collector. Regardless of the exchanger's location, the main sodium hydraulic path passes through intermediate exchangers, exhibiting significant height variations within the hot and / or cold columns, thus reducing the hydraulic performance of natural convection. Therefore, in Figure 2 In the heat collector 12, an exchanger 26 for discharging excess power is positioned. The hydraulic path includes a hot column 26 and a cold column 27. The hot column 26 rises regularly with a monotonous height variation. However, the cold column 27 involves a non-monotonic height variation. In practice, the sodium at the output of the exchanger 25 must return towards the input window 17, located in the upper part of the heat collector, before entering the intermediate exchanger 16, to reach the core 11 after passing through the pumping system 19. In the heat collector 12, the hot and cold columns are not physically separated, which does not correspond to the optimal conception regarding natural convection, as the cold sodium leaving the exchanger 25 can mix with the hot sodium entering the same exchanger in the heat collector.

[0020] For those skilled in the art, one possible improvement would be to place a dedicated excess power recirculation (EPuR) exchanger between the heat collector and the cold collector, passing through the bulge bastion, as in the case of the intermediate exchanger. This is not done because during normal operation, this would mean creating a bypass to the intermediate exchanger via the EPUR exchanger, thus degrading the reactor's normal operating performance. Therefore, there is an inherent technical contradiction in the loop used for heat recirculation: a technical solution optimized for normal operation reduces operation under excess power recirculation conditions, and vice versa.

[0021] The final drawback of reactors with an integrated design involves the constraints of arranging all the components necessary for proper reactor operation within the reactor: therefore, the tank size is larger than that of a loop reactor, and the passageways through pumps and exchangers significantly increase the burden on the bulge. This results in limitations on the size of passive equipment, which can be added at the bulge to facilitate natural convection operation of the reactor while allowing excess power to be discharged through the exchangers.

[0022] As mentioned above, fast neutron nuclear reactors can withstand accidents resulting in complete core meltdown. Therefore, mitigation safety procedures exist for these reactors to mitigate the effects of such core meltdown accidents.

[0023] Therefore, considering that a core meltdown accident is a holistic event, the reactor design limits the consequences of such an accident through the presence of mitigation equipment. In the case of core physical degradation, mitigation of reactivity transients is achieved by adding extraction equipment for degrading fuel, particularly delivery pipes, associated with core melt traps.

[0024] These two devices serve multiple purposes. At the onset of an accident, this involves using delivery pipes to briefly (approximately several seconds) reposition the fuel and molten structural material (i.e., the core melt) outside the fission core region. This repositioning of the core melt is thus directed downwards out of the region of neutron flux. The kinetics of removing the core melt are crucial, as this involves extracting the core melt before it forms a large molten pool, which has significant recriticality potential due to the massive compression of the core.

[0025] The second objective involves radially limiting the degradation region of the reactor core to prevent melt propagation to other components and internal storage, specifically the area surrounding the core that includes used fuel assemblies.

[0026] Furthermore, the third objective involves repositioning the molten core material over a prolonged period or in a sequence exceeding several minutes to a trap plate located at the bottom of the main tank, beneath the core. The trap reduces the risk of criticality through the diffusion of the molten core material and also allows for its cooling and containment.

[0027] In practice, to achieve the above objectives, the commonly chosen structural arrangement is to implement specific equipment within the reactor core for extracting and subsequently recovering the molten core material. This equipment consists of pipes 40 for conveying the molten core material C from the fission zone ZF of core 11, and is associated with a molten core trap 41, such as... Figure 3As shown in the figure, this illustrates the principle of discharging molten core material through delivery pipes. The core 11 includes several delivery pipes 40, the layout of which is optimized for discharging molten core material.

[0028] In the event of a complete meltdown of core 11, these tubes 40 must rapidly allow the molten core to be unloaded into a core melt trap 41 located at the bottom of tank 13. These safety devices serve only to mitigate core meltdown accidents and must be as transparent as possible in other reactor operating conditions.

[0029] From an architectural perspective, the delivery tube 40 is a hollow structure that provides a channel for the molten core stream to pass through the core grid 30 and the strong backplate 31.

[0030] During a core meltdown accident, several operational issues arise. In fact, regardless of the initiator that triggers the meltdown, the fuel assembly suffers degradation of at least one rod after the fuel melts (a fuel assembly typically includes a base for supplying sodium, a bundle of rods within a hexagonal tube (HT), components for securing the rods, an upper neutron shield (UNS), and a sustaining head). Consequently, the steel structure of the rods (cladding, wires, plugs) melts rapidly, thus losing the first limiting barrier, the cladding seal. The defect rapidly spreads across the entire fuel rod bundle on an assembly scale. The core melt, including fuel and molten structural material, comes into contact with the inner wall of a 4-5 mm thick steel hexagonal tube at approximately 2700°C; the inner wall melts and is punctured within seconds. Therefore, the core melt pool propagates throughout the core to other fuel assemblies, leading to a complete meltdown accident.

[0031] The delivery pipes constrain and limit the radial propagation of the molten core within the core. Each delivery pipe is used to melt and open a path to the molten core, which flows internally by gravity out of the neutron flux region, helping to reduce reactivity. Thus, the trajectory of the molten core continues through the core grid plates, then through the strong backplate, to reach vertically the molten core trap plate positioned at the bottom of the tank.

[0032] From the perspective of a series of accidents, in order to limit the consequences of an accident, it is important to remove the molten core material outside the flux zone quickly enough. To ensure the correct trajectory of the molten core material to the trap, it is also necessary to limit any obstructions within the delivery pipe.

[0033] Several issues also arose related to the normal operation of the reactor. In fact, since the tubes used to transport the molten core can only operate during a meltdown accident, they must be as transparent as possible during normal reactor operation. Specifically, the tubes must have no impact on operation in terms of mechanical, neutron, and thermal-hydraulic behavior.

[0034] From a mechanical perspective, the tubes used to transport the molten core are part of the core structure, and like other components, this structure is subjected to mechanical stresses during various stages of its lifespan (i.e., manufacturing, transportation, maintenance, and radiation). Furthermore, the radiation and damage levels of the mechanical structure of the transport tubes are as high as those of the fuel assembly mechanical structures (up to 100 dPa). The primary mechanical loads that the transport tube structure must withstand are the contact stresses between components within the core. These contact stresses at the hexagonal tubes manifest as compressive, bending, and torsional forces encountered during normal operation and during earthquakes.

[0035] From a neutron perspective, the delivery tubes are hollow structures positioned within the fission reactor core. The presence of these potentially vacant elements must have a tolerable impact on the core's neutron performance. This particularly implies the criteria for the core's criticality, as the presence of "pores" in the core reduces the fission volume and flattens the power distribution.

[0036] From a thermohydraulic perspective, the proposed delivery pipe design must not result in a direct connection between the heat collector and the cold collector. Therefore, a pressure difference of approximately 150 mbar between the two collectors would generate a hot sodium flow from the heat collector, which, under overpressure, descends into the cold collector within the pipe. However, injecting hot sodium (550°C) into structures positioned within cold sodium (400°C) (core grids, strong backplates, etc.) is undesirable for the mechanical strength and aging (thermomechanical stress caused by the temperature gradient) of such structures. Specifically, it is necessary to demonstrate that the service life of the structure equals the service life of the equipment. Furthermore, the free surface of sodium in the heat collector is known to be one of the main sources of gas release from the heat collector. These bubbles dissolve in the sodium of the heat collector under high temperatures, and if the gas enters the cold collector, there is a risk of witnessing the gas re-nucleating (nucleation phenomenon) and reforming into bubbles. However, the presence of gas in the cold collector must be limited to avoid any risk of gas entering the assembly, a phenomenon that could lead to fuel rod cooling failure and power surges. In summary, the delivery pipe should not allow for the injection of sodium between collectors to limit mechanical damage to the structure or to promote the release of gas from the cold collector.

[0037] Therefore, the following requirements and constraints arise. In the event of a series of core meltdown accidents, it is required that the molten core material be rapidly discharged outside the neutron flux region and then into the core melt trap. During normal reactor operation, it is required to ensure a tight seal between the hot and cold collectors. These two constraints are designed to resist mechanical loads throughout the reactor's lifespan without degrading the core's neutron performance.

[0038] In the prior art, it can be mentioned that the first FNR-Na, known in French as Rapsodie and Phénix, did not include equipment in the core specifically designed to mitigate core meltdown accidents. Safety assessments of the FNR-Na reactor were conducted through dedicated experiments using qualified simulation tools.

[0039] The FNR-Na SuperPhénix includes a core melt trap at the bottom of the tank, but no mitigation device in the core.

[0040] Following the Chernobyl and Fukushima nuclear accidents, safety objectives were reinforced to establish a set of requirements that must be met for the next generation of reactors, namely the fourth generation reactors.

[0041] These new safety objectives could lead to the addition of specific equipment in the reactor core dedicated to mitigating core meltdown accidents and enabling the transport of molten core material to the traps. The first proposed idea, adding a transport pipe to the core, was not entirely satisfactory because the transport pipe created thermal-hydraulic problems by bypassing the hot and cold collectors.

[0042] The same applies to recommendations for fully or partially sealed delivery pipes, because while delivery pipes do meet the sealing requirements between collectors, they may facilitate the release of gas from the core (sealing the upper part could lead to the risk of gas accumulation within the delivery pipe) or impede the flow of molten core material (in the case of a “fusible” sealed lower part). In reality, devices with fusible films carry a reliability risk of melting during a series of core meltdown accidents, which accumulates and is difficult to simulate and characterize.

[0043] In addition, the patent documents also propose equipment for mitigating core meltdown accidents, particularly equipment specifically designed for transporting molten core material.

[0044] For example, Japanese patent application JP H10-227884A relates to a concept for an absorber assembly (control rod) that functions in mitigating core meltdown accidents. An opening is provided in a hexagonal tube, leading to an interassembly space defined by adjacent fuel assemblies. The base of the assembly also leads to this interassembly space, thus allowing cooling of the absorber rods during normal operation. The base is closed at its ends. During a fuel rod meltdown accident, the core melt melts the hexagonal tube of the fuel assembly and enters the interassembly space. The core melt then flows by gravity to the base of the absorber assembly, where it is stored. This application highlights the repositioning of the core melt outside the region of neutron flux and the reduction in reactivity. However, aside from the fact that the progress of the core melt can be significantly slowed in the small cross-section of the interassembly space, this concept is not intended for a device to transport the core melt to a trap positioned beneath the core.

[0045] Furthermore, US patent application US 2012 / 0201342 A1 relates to a concept for a fuel assembly and an FNR reactor that function in venting and recovering molten core material during a core meltdown accident. The assembly includes a solid body in the portion below its fuel rods, pierced by a series of holes to ensure external cooling via hexagonal tubes during a series of core meltdown accidents containing molten core material within the assembly. The fuel rods are modified to improve behavior during core meltdown accidents. The end of the assembly base is closed. A sealed container for absorbing material is positioned below the core grid perpendicular to the assembly; this sealed container serves as a molten core trap. This application aims to reposition the molten core material, then vent it, and prevent its radial propagation within the core. However, several disadvantages arise: the ends of the base and the top of the container are fusible walls that must disappear upon contact with the molten core, and the reliability of equipment with fusible films is deemed unsatisfactory; integrating equipment for venting the molten core into fuel assemblies supplied with pressurized sodium flow if the main pumps do not stop risks obstructing proper flow of the molten core; the thickness of the assembly body in its lower portion significantly increases the amount of irradiated waste and complicates the manufacture of the assembly body. Furthermore, it must be verified that this series of orifices will not interfere with thermal hydraulics during operation.

[0046] Furthermore, the principle of the FAIDUS assembly (for fuel assemblies with internal ducting structures) from the Japan Atomic Energy Agency can be mentioned. Therefore, in order to limit the consequences of core meltdown accidents using the CDA (for core failure accidents) approach, the JSFR reactor (for the Japanese sodium-cooled fast reactor) integrates a mitigation device called FAIDUS into its fuel assembly design. Due to the reactivity characteristics of the JSFR reactor, heating from the CDA and sodium-induced boiling produce localized power spikes in the JSFR assembly (a positive reactivity effect in the case of sodium flaring). Therefore, the FAIDUS design integrated into the JSFR's fuel assembly aims to rapidly extract liquid fuel to avoid excessive compression and a return to an excessively high critical state during a CDA event. FAIDUS is not a patented concept, but rather the subject of several international exchanges.

[0047] Two alternative FAIDUS assemblies exist. The reference solution, intended for advanced research (computational and experimental) purposes, proposes small channels for core melt flow, arranged at the angles of hexagonal tubes. Molten fuel is unloaded towards the top of the assembly by a motor that initiates an initial early power surge (internal pressure) within the assembly. In the upper portion of the JSFR core, a trap plate is present to receive the ejected fissile material. In the second alternative, the channel for discharging the core melt is located at the center of the fuel rod bundle. The molten fuel must be discharged by gravity through the bottom of the assembly. This option has not been widely adopted, partly due to manufacturing difficulties and partly because the technological constraints for conveying the core melt to the substructure have been considered very strong.

[0048] It should be noted that the amount of fuel injected via FAIDUS is insufficient to prevent other fuel in the remainder of the core from later returning to criticality. A second option for unloading fuel requires an empty structure of the absorber control rod (CRGT, for the control rod guide tube) held in the upper position. Furthermore, questions arise regarding whether the proper passage of molten fuel through the vacuum generation system would obstruct the channels in the CRGT's base. Finally, even though the FAIDUS design has the advantage of slightly interfering with core neutronics, integrating the channels for venting molten core material into fuel assemblies supplied with pressurized sodium flow poses a risk of hindering proper flow of molten core material if the main pumps are not stopped.

[0049] In summary, the strategy of venting core melt through the top of the FAIDUS fuel assembly depends on the neutron and thermal-hydraulic characteristics of the JSFR core. This strategy cannot be immediately applied to another FNR-Na core. Summary of the Invention

[0050] One objective of the present invention is to overcome, at least in part, the aforementioned needs and disadvantages associated with embodiments of the prior art.

[0051] More specifically, the object of this invention is to improve the behavior of the FNR-Na reactor core in the event of a complete core meltdown accident. In particular, the invention seeks to reduce the consequences of a core meltdown accident while maintaining the integrity of the second barrier (the reactor vessel) and the sealing of the third safety barrier (the reactor building and base plate) to prevent damage from mechanical effects (e.g., energy release due to the steam expansion of molten and then evaporated material) and thermal effects (e.g., corrosion of the structure by the melted core material) generated during the accident. The invention generally seeks to comply with restrictions on the release of radioactive materials into the population and to restore the reactor to a safe state after the accident.

[0052] Therefore, according to one aspect of the invention, the object of the invention is an integral fast neutron nuclear reactor cooled by a liquid metal heat transfer fluid, the reactor comprising:

[0053] - The main tank, suspended from the protective plate, includes the volume of heat transfer fluids and the reactor core, which comprises multiple fuel assemblies assembled into core grids supported by a strong backplate.

[0054] - The main loop, integrated into the main tank, includes at least one main pump, at least one intermediate heat exchanger, and at least one heat exchanger. The at least one intermediate heat exchanger is used to discharge the power generated by the reactor core during normal operation, and the at least one heat exchanger is used to discharge excess power.

[0055] - The internal structure, referred to as the convex fortress, divides the volume of the heat transfer fluid into at least two regions to form a heat collector at the output end of the core and a cold collector at the output end of the at least one intermediate exchanger.

[0056] The reactor core is characterized by further including safety equipment specifically designed to mitigate core meltdown accidents. This safety equipment comprises at least one basic safety device positioned near at least one fuel assembly in the core for conveying molten fuel to a cold collector at the bottom of the main tank.

[0057] The at least one basic safety device includes a hollow and removable delivery pipe and a hollow through-tube, the delivery pipe being fitted into the core grid plate through a first opening in the core grid plate, and the through-tube being in fluid communication with the delivery pipe.

[0058] The through-pipe is positioned to pass through the core grid from the first opening to a second opening located between the core grid and the strong back plate, and then through the second opening to a third opening in the strong back plate, leading to a cold collector at the bottom of the main tank.

[0059] The delivery pipe includes one or more injection nozzles located inside the pipe and positioned at least partially within a portion of the pipe, which is fitted into the core grid and oriented toward the heat collector. The injection nozzles are designed to supply heat transfer fluid to the core grid.

[0060] The delivery pipe includes at least one region with a reduced thickness on its outer wall, the at least one region being located at the fission region of at least one fuel assembly to increase the puncture velocity of the outer wall in the event of a core meltdown accident and to allow molten fuel to flow from the at least one fuel assembly into the delivery pipe.

[0061] The reactor also includes equipment for recovering molten fuel, which is located in a cold collector at the bottom of the main tank and aligned with the third opening.

[0062] The fast neutron nuclear reactor according to the invention may further include one or more of the following features, which may be employed individually or in any technically possible combination.

[0063] Delivery pipes can have the external features of commonly used fuel assemblies.

[0064] Advantageously, the supply of heat transfer fluid to the core grids through one or more injection nozzles is carried out in the same manner as with conventional fuel assemblies. Furthermore, advantageously, the delivery pipe is fitted into the core grids at the location of a conventional fuel assembly. As with conventional fuel assemblies, the delivery pipe is easily removed for venting and replacement.

[0065] Advantageously, the delivery tube may include: a head that enables the delivery tube to be held; a base that is fitted into the core grid; and a hollow body having a hexagonal cross-section between the head and the base.

[0066] The cross-section of the base can be smaller than that of the hollow body. The upper part of the base can have a funnel shape to facilitate the passage of molten fuel. The lower part of the base can be opened to allow molten fuel to flow into the through pipe.

[0067] In addition, the head may include a central channel and a vent, the central channel having a cross-section smaller than that of the hollow body to reduce the risk of molten fuel being discharged toward the heat collector, and the vent being located at the interface with the hollow body to facilitate the discharge of gas during immersion in the heat transfer fluid.

[0068] Furthermore, the at least one region with reduced thickness can be obtained by milling the outer wall of the delivery pipe.

[0069] The at least one region with reduced thickness can be positioned at the hollow body of the conveying pipe and not at the angle of the hexagonal cross-section of the hollow body in order to maintain the mechanical rigidity of the conveying pipe.

[0070] In addition, one or more injection nozzles can be positioned at the base of the delivery tube.

[0071] The delivery tube may include at least one injection nozzle having an annular shape, particularly a single injection nozzle having an annular shape.

[0072] Alternatively, the delivery tube may include multiple injection nozzles regularly distributed in the delivery tube, particularly three injection nozzles forming the vertices of an equilateral triangle.

[0073] In addition, the reactor core may include safety devices consisting of multiple basic safety devices specifically designed to mitigate core meltdown accidents. A portion of the multiple basic safety devices is located on the periphery of a region of the reactor core that includes multiple fuel assemblies, while another portion of the multiple basic safety devices is located in the central portion of a region of the reactor core that includes multiple fuel assemblies.

[0074] Each basic safety device can be attached to the fuel assembly in two directions. Attached Figure Description

[0075] The invention can be better understood by reading the following detailed description of non-limiting embodiments of the invention, and by examining the schematic and partial views in the accompanying drawings, in which:

[0076] - Figure 1 The design principle of a "loop-type" sodium-cooled fast neutron nuclear reactor is shown in axial cross-section.

[0077] - Figure 2 The design principle of the "monolithic" sodium-cooled fast neutron nuclear reactor is shown in axial cross-section.

[0078] - Figure 3 The axial cross-section illustrates the principle of venting fuel, and particularly molten core material, through delivery pipes in the event of a complete core meltdown accident in a fast neutron nuclear reactor.

[0079] - Figure 4 The axial cross-section illustrates the principle of evacuating fuel, and particularly evacuating molten core material, using the basic safety devices of the reactor according to the invention in the event of a complete core meltdown accident.

[0080] - Figure 5A A partial perspective view shows the implementation details of the hollow body of the delivery pipe of the basic safety device of the reactor according to the present invention.

[0081] - Figure 5B It is along Figure 5A A cross-sectional view of AA.

[0082] - Figure 5C It is along Figure 5A Cross-sectional view of BB

[0083] - Figure 6A An embodiment of the basic safety device for a reactor according to the present invention is shown in axial cross-section.

[0084] - Figure 6B It is along Figure 6A A cross-sectional view of AA.

[0085] - Figure 6C It is along Figure 6A Cross-sectional view of BB

[0086] - Figure 7A Another embodiment of the basic safety device for the reactor according to the invention is also shown in axial cross-section.

[0087] - Figure 7B It is along Figure 7A A cross-sectional view of AA.

[0088] - Figure 7C It is along Figure 7A A cross-sectional view of BB, and

[0089] - Figure 8 An example of the construction of the reactor core according to the present invention is shown in top view.

[0090] In all these figures, the same reference numerals may denote the same or similar elements.

[0091] Furthermore, to make the accompanying drawings easier to read, the different parts shown in the drawings are not necessarily drawn to the same scale. Detailed Implementation

[0092] Figures 1 to 3 The invention has been described above with reference to the prior art and the general context of the present invention.

[0093] Reference Figures 4 to 8 The features and structural arrangement of the present invention will now be described, which enable the transport of the core melt to a trap 41 at the bottom 13 of the tank via a set of transport pipes 40 to mitigate the consequences of a hypothetical accident involving the complete meltdown of the core of a fast neutron reactor, particularly a sodium-cooled fast neutron reactor (FNR-Na).

[0094] It should be noted that the above has already been referenced. Figures 1 to 3 Elements specific to the fast neutron nuclear reactor R described herein need not be described again. Therefore, the same reference numerals denote the same elements. Furthermore, in the example described below, the liquid-metal heat transfer fluid of reactor R is sodium, but this choice is by no means limiting.

[0095] Furthermore, it should be noted that the inventors have investigated a solution proposed in French patent application FR 3053827A1, which cools the reactor core through natural convection having a direct hydraulic path descending from the heat collector toward the cold collector. However, the present invention relates to a completely different technical problem, which is logically related to mitigating core meltdown accidents by hydraulically locking empty delivery pipes.

[0096] The international community, particularly Japan, employs safety procedures involving the rapid removal of molten fuel from the reactor core to prevent a violent power surge during a core meltdown accident. However, the technical arrangements chosen to achieve this objective depend on the neutron and thermal-hydraulic characteristics of the reactor core and its behavior during an accident. Therefore, the principles of this invention, developed specifically for the core of the ASTRID reactor, differ from existing solutions, particularly from the solutions adopted by Japan in the aforementioned FAIDUS fuel assembly concept.

[0097] One of the unique features of the ASTRID reactor is its innovative CFV heterogeneous core design (a core with low reactivity in the event of sodium emissions). Therefore, in the event of an accident, a power surge caused by sodium emissions from components is impossible. In this case, the goal of mitigation methods in the ASTRID reactor is to avoid significant compressive diffusion of fuel onto the degraded core, and thus prevent excessively high transient criticality. To this end, complementary safety devices must be implemented in the CFV core to allow molten fuel, i.e., the core melt, to be discharged sufficiently quickly outside the core via flow channels. In practice, the flow channels can be integrated into existing components (fuel, control rods, etc.) or formed into completely separate components adjacent to the fuel assemblies. This invention is based on the second possibility, namely, using a dedicated delivery pipe 40 to maximize the cross-sectional area of ​​the channel used to discharge the core melt without affecting the performance of other components.

[0098] Furthermore, these objects are characterized by their ability to facilitate the axial transport of liquid fuel by gravity through the lower structure of the reactor (core grid and reinforcing backplate). Thus, some or all of the molten fuel in the core can be collected solely through these devices at the bottom of the tank in a core melt trap 41 specifically designed for this purpose.

[0099] Therefore, the present invention is based on the use of a hydraulic locking system in the delivery pipe, which does not have the disadvantages of existing solutions (fusible walls, bypass between the hot and cold collectors, risk of gas release, delivery of core melt below the main flow, etc.). On the other hand, the present invention is based on the optimization of the pipe design and the installation of the pipe in the core to achieve the purpose of mitigation.

[0100] Figure 4 The axial cross-section illustrates the principle of fuel discharge according to the invention, in the event of a complete meltdown of the reactor core 11, by means of a safety device consisting of at least one basic safety device 45 of the reactor R, where the fuel is in the form of the core melt.

[0101] According to the present invention, the core 11 of reactor R includes a safety device consisting of at least one basic safety device 45 specifically designed to mitigate a possible core meltdown accident.

[0102] The basic safety device 45 is adjacent to the fuel assembly 43 of the core 11 and enables the molten core material to be transported to the cold collector 14 at the bottom of the main tank 13.

[0103] The basic safety device 45 includes a delivery tube 40, which is similar to an assembly-type removable hollow structure and is positioned or mounted on the core grid 30 through a first opening 50 in the core grid 30.

[0104] The basic safety device 45 also includes a hollow through-tube 42 in fluid communication with the delivery tube 40 in an extension of the delivery tube 40. The through-tube 42 is positioned to pass through the core grid plate 30 from a first opening 50 to a second opening 54 located between the core grid plate 30 and the strong back plate 31, and from the second opening 54 through the strong back plate 31 to a third opening 56 of the strong back plate 31, so as to lead perpendicularly to the cold collector 14 at a low pressure of approximately 0 bar. The device 41 is in the form of a trap plate for recovering molten core material located in the cold collector 14 at the bottom of the main tank 13.

[0105] The core grid 30 serves to supply cooled sodium to the assembly under high pressure, or approximately 3 bar. For this purpose, the assembly's base and shield tubes have radial openings facing each other. Furthermore, the reinforcing backplate 31 serves to support the core 11 and the core grid 30.

[0106] The delivery pipe 40 is a component specifically designed for discharging molten core material. Preferably, the external architecture of the delivery pipe is identical to that of other components in the core 11, and it has the same interfaces.

[0107] The delivery tube 40 includes: a head 60 located in the upper portion to enable the tube to be held; a base 62 to enable the tube to be held on the core grid plate 30; and a hollow body 61 having a hexagonal cross-section between the head 60 and the base 62.

[0108] The base 62 differs slightly from the bases of other components of the core 11. The cross-section of the base 62 is smaller than that of the hollow body 61. Furthermore, the upper portion 62a of the base 62 may have a funnel shape to allow the core molten material to pass through the base 62, which has a smaller cross-section. Additionally, the lower portion 62b of the base 62 is open to allow the core molten material to flow to the through-tube 42 and thus through the core grid 30 and the reinforcement backplate 31 to the trap plate 41. The base 62 also includes an injection nozzle 53, as described below.

[0109] Furthermore, the head 60 includes a central channel 60a, the inner cross-section of which is smaller than the cross-section of the hollow body 61, to reduce the risk of venting molten core material through the top at the onset of a series of core meltdown accidents. Additionally, the lower portion of the head 60 includes a vent 60b to facilitate the venting of gases during assembly immersion in sodium. The solid nature of the steel head 60 also serves as a neutron shield by limiting axial neutron leakage.

[0110] The hollow body 61 does not include an internal structure and is filled with sodium during normal operation. During a core meltdown accident, the hollow body 61 is punctured by the melt, initially forming a channel for the core melt to flow by gravity to the base 62.

[0111] Regarding the aforementioned requirement to rapidly expel the molten core outside the neutron flux region, and considering the dynamics that cannot be influenced by the flow of the molten core within the delivery pipe 40, a time gain can be achieved at the onset of an accident by accelerating the penetration of the hollow body 61 by the molten core from adjacent fuel assemblies 43. This is achieved by reducing the thickness of the hollow body to decrease its thermal inertia, thereby allowing the temperature of the hollow body 61 to rise more rapidly.

[0112] Therefore, as referenced Figure 5A , Figure 5B and Figure 5C As can be seen, the delivery pipe 40 includes at least one region 65 with reduced thickness on its outer wall 66, which is positioned to face the fission region ZF of at least one fuel assembly 43, in order to increase the puncture velocity of the outer wall 66 in the event of a core meltdown accident and to allow core melt to flow from the at least one fuel assembly 43 to the delivery pipe 40.

[0113] In practice, this structural arrangement is only applicable to the region of the delivery pipe 40 most likely to come into contact with the core melt, namely the hollow body 61 facing the fission zone ZF. Of course, the axial positioning of one or more regions 65 with reduced thickness on the hollow body 61 can be optimized through appropriate calculations.

[0114] From a manufacturing perspective, the reduced thickness in one or more regions 65 can be achieved by milling the outer wall 66 of the delivery tube 40, specifically by milling the outer surface of the delivery tube. The advantage of this solution is that it uses a completely standard tube initially, and the milling operation presents no technical difficulties. The milling is performed only on the surface of the hollow body 61, rather than at the angles of the hexagonal cross-section of the hollow body, to maintain good mechanical rigidity and thus meet the aforementioned requirements for resisting mechanical loads during the lifespan of the tube.

[0115] Milling is a more advantageous solution, especially compared to solutions that involve, for example, manufacturing pipe sections with reduced thickness and then welding these reduced-thickness sections to other pipe sections with standard thickness. In reality, there are many difficulties in manufacturing such pipes and in obtaining the necessary welds.

[0116] Furthermore, still according to the present invention and referring to Figures 6A to 7C The delivery pipe 40 includes one or more injection nozzles 53 located inside the base 62 and positioned at least partially within the hollow body 62 of the pipe 40, which is fitted into the core grid plate 30 and oriented toward the heat collector 12. The one or more injection nozzles 53 are intended to be supplied with heat transfer fluid from the core grid plate 30.

[0117] These nozzles 53 are supplied with "high-pressure" sodium from the core grid 30 in a manner similar to that of the fuel assembly 43. The dimensions of these nozzles 53 in terms of diameter and number are defined such that the back pressure generated by this injection compensates as precisely as possible for the head loss of the intermediate exchanger during operation, which represents the pressure difference between collectors 12 and 14. Therefore, the flow velocity exiting through the top of the delivery pipe 40 into the hot collector 12 is almost zero and is offset when the main pump stops, resulting in a zero reduction in the pressure difference between collectors 12 and 14. However, the flow velocity exiting through the bottom of the delivery pipe 40 corresponds to the flow velocity of the injection nozzles 53. This outflow velocity is injected into the cold collector 14 through the through-pipe 42, core grid 30, and reinforcing backplate 31.

[0118] The back pressure generated by the injection of sodium supplied by the main pump locks the descending sodium flow in the hexagonal delivery pipe 40 at all pumping speeds. Therefore, during normal operation of reactor R, the requirement to ensure a seal between the hot collector 12 and the cold collector 14 is met as specified above, since there is no hydraulic bypass between collectors 12 and 14 and no upper closure of the pipe (no gas release).

[0119] Similarly, in the event of a series of core meltdown accidents, if the main pumps are still operational, the resulting hydraulic lock-in will not produce an increased flow velocity of the heat transfer fluid, thus preventing the transport of molten core material toward the trap plate 41. This helps to comply with the requirements listed above for allowing molten core material to be rapidly discharged outside the neutron flux region. The fact that the delivery pipe 40 remains open and is not blocked at its base by a fusible wall also facilitates this requirement by not obstructing or slowing the flow of molten core material toward the trap plate 41. (Refer to the following...) Figures 7A to 7C The alternative described toroidal injector 53 is also in this case, which minimizes obstacles to the flow of core melt within the base 62.

[0120] therefore, Figure 6A , Figure 6B and Figure 6C A specific embodiment of the delivery tube 40 is shown, which has three injection nozzles 53 and includes a circular inner cross-section for the base 62 and a hexagonal outer shape for the hollow body 61, similar to that used for a standard fuel assembly. The three injection nozzles 53 form the vertices of an equilateral triangle.

[0121] also, Figure 7A , Figure 7B and Figure 7C Another embodiment is shown, which is similar to Figure 6A , Figure 6B and Figure 6C The difference in this embodiment lies in the fact that three injection nozzles 53, arranged in a transverse cross-section according to the vertices of an equilateral triangle, are replaced by a single injection nozzle 53 with an annular shape, which directly contacts the inner wall of the base 62. This annular shape of the injection nozzle 53... Figure 7B It is particularly visible in the middle.

[0122] also, Figure 8 An example of the construction of the core 11 of the reactor R according to the present invention is shown in top view.

[0123] Therefore, the core 11 includes safety equipment consisting of a plurality of basic safety devices 45 specifically designed to mitigate core meltdown accidents. A portion of the basic safety devices 45 is located at the periphery of the region of the core 11 including the fuel assemblies 43, while another portion of the basic safety devices 45 is located in the central portion of the region of the core 11 including the fuel assemblies 43.

[0124] Therefore, most of the delivery pipes 40 are positioned on the periphery of the fuel core to limit the radial propagation of the molten pool. Some delivery pipes 40 are also positioned in the central portion of the core to discharge a portion of the core molten material as early as possible. The installation of the central pipe 40 must comply with neutron studies designed to define the number and spacing of the absorber rods and to be used for the control and shutdown of reactor R.

[0125] The number and spacing of the delivery pipes 40 around the fuel core were determined by calculating the scenario of a core meltdown accident using code simulating the propagation of molten core material within the core 11. An optimal trade-off was sought among a limited number of pipes 40 to avoid degrading neutron performance while enabling the removal of a satisfactory volume of molten core material within a short timeframe.

[0126] The same applies to the optimal determination of the number of central tubes 40 (as early as possible in comparison to neutron performance) and the optimal determination of the number of tubes 40 on the periphery of core 11 (the limited number in comparison to the spatial layout of core melt discharge).

[0127] It should also be noted that the installation of pipes 40 for conveying the core melt and at the location of fuel assemblies 43 in core 11 translates into reduced core power. To avoid reducing the neutron performance of core 11, the power loss could be compensated for by increasing the number of fuel assemblies 43 and thus increasing the diameter of core 11. This option is undesirable because it results in a direct increase in costs associated with the increased size of the tank 13 and reactor buildings, which in turn increases the costs of manufacturing, installing, and removing the fuel assemblies 43.

[0128] Alternatively, the power could be increased not by increasing the number of fuel assemblies 43, but by increasing the concentration of plutonium (Pu) or another fissile material in the (U,Pu)O2 oxide fuel used. However, this must be done carefully to adhere to the maximum Pu concentration limits set by radiation feedback and manufacturing capabilities. Ultimately, neutron studies are needed to verify that the criterion for maximum linear power in the fuel rods is being followed.

[0129] Of course, the present invention is not limited to the embodiments just described. Those skilled in the art can make various modifications to the present invention.

[0130] In particular, the heat-transferring liquid metal can be a metal other than sodium, such as lead or lead-bismuth.

Claims

1. Monolithic fast neutron nuclear reactor (R) cooled by a liquid metal heat transfer fluid, said reactor comprising: - a main tank (13) suspended on a protection plate (24), said main tank comprising a volume of heat transfer fluid and a core (11) of said reactor (R), said core (11) comprising a plurality of fuel assemblies (43) assembled into a core grid (30) supported by a strong backplate (31), - a primary circuit integrated into said main tank (13), said primary circuit comprising at least one primary pump (19), at least one intermediate heat exchanger (16) for evacuating the power produced by said core (11) during normal operation, and at least one heat exchanger (25) for evacuating the residual power, - an internal structure (15), called a crenel, dividing said volume of heat transfer fluid into at least two zones to form a hot collector (12) at the output of said core (11) and a cold collector (14) at the output of said at least one intermediate heat exchanger (16), characterized in that said core (11) of said reactor (R) also comprises safety equipment dedicated to slowing down a core meltdown accident, said safety equipment consisting of at least one basic safety equipment (45) positioned in the vicinity of at least one fuel assembly (43) of said core (11) for transporting molten fuel to said cold collector (14) at the bottom of said main tank (13), said at least one basic safety equipment (45) comprising a hollow and removable transport pipe (40) assembled into said core grid (30) through a first opening (50) in said core grid (30) and a hollow through pipe (42) in fluid communication with said transport pipe (40), said through pipe (42) being positioned to pass through said core grid (30) from said first opening (50) to a second opening (54) positioned between said core grid (30) and said strong backplate (31) and to pass through said strong backplate (31) from said second opening (54) to a third opening (56) of said strong backplate (31) to open onto said cold collector (14) at the bottom of said main tank (13), said transport pipe (40) comprising one or more injection nozzles (53) located inside said transport pipe (40) and positioned at least partially in a portion (40a) of said transport pipe (40) assembled into said core grid (30) while being oriented towards said hot collector (12), said one or more injection nozzles (53) being intended to be fed by the heat transfer fluid of said core grid (30). The transfer pipe (40) comprises, on its outer wall (66), at least one zone (65) of reduced thickness, positioned at the fissile zone (ZF) of at least one fuel assembly (43), to increase the piercing speed of the outer wall (66) in the event of a core meltdown accident and to enable the flow of molten fuel from the at least one fuel assembly (43) to the transfer pipe (40), and the reactor (R) further comprises a device (41) for recovering molten fuel, positioned in the cold collector (14) at the bottom of the main tank (13) and positioned in alignment with the third opening (56).

2. The reactor of claim 1, wherein, The transfer pipe (40) comprises a head (60) enabling the holding of the transfer pipe, a base (62) fitted into the core grid (30), and a hollow body (61) having a hexagonal cross-section between the head (60) and the base (62).

3. The reactor of claim 2, wherein, The cross-section of the base (62) is smaller than the cross-section of the hollow body (61), an upper part (62a) of the base (62) has a funnel shape to facilitate the passage of molten fuel, and a lower part (62b) of the base (62) is open to enable the flow of molten fuel towards the through pipe (42).

4. The reactor according to claim 2 or 3, characterized in that The head (60) comprises a central passage (60a) having a cross-section smaller than the cross-section of the hollow body (61) to reduce the risk of expulsion of molten fuel towards the hot collector (12), and a vent (60b) at the interface with the hollow body (61) to facilitate the expulsion of gases during immersion in the heat transfer fluid.

5. The reactor of any one of claims 1 to 3, wherein, The at least one zone (65) of reduced thickness is obtained by milling the outer wall (66) of the transfer pipe (40).

6. The reactor of claim 2 or 3, wherein The at least one zone (65) of reduced thickness is positioned at the hollow body (61) of the transfer pipe (40) and does not exist at the angles of the hexagonal cross-section of the hollow body (61) to preserve the mechanical rigidity of the transfer pipe (40).

7. The reactor of claim 2 or 3, wherein The one or more injection nozzles (53) are positioned at the base (62) of the transfer pipe (40).

8. The reactor of any one of claims 1 to 3, wherein, The transfer pipe (40) comprises at least one injection nozzle (53) having an annular shape.

9. The reactor of any one of claims 1 to 3, wherein, The transfer pipe (40) comprises a plurality of injection nozzles (53) regularly distributed in the transfer pipe (40).

10. The reactor of any one of claims 1 to 3, wherein, The core (11) comprises safety devices consisting of a plurality of basic safety devices (45) dedicated to slowing down a core meltdown accident, a part of the plurality of basic safety devices (45) being positioned at the periphery of the zone of the core (11) comprising the plurality of fuel assemblies (43), and another part of the plurality of basic safety devices (45) being positioned in the central part of the zone of the core (11) comprising the plurality of fuel assemblies (43).

11. The reactor of claim 8, wherein, The transfer pipe (40) has a single injection nozzle (53) having an annular shape.

12. The reactor of claim 9, wherein, The delivery tube (40) comprises three injection nozzles (53) forming the vertices of an equilateral triangle.

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