A nuclear reactor having a forced convection liquid coolant and a solid fuel assembly incorporating a heat removal system having a liquid metal bath for removing nominal heat and a phase-change material (PCM) for removing decay heat during an accident

Abstract

To provide a nuclear reactor with forced-convection liquid coolant and solid fuel assemblies, incorporating a heat removal system using a liquid metal bath for removing nominal heat and a phase-change material(s) (PCM) for removing decay heat in the event of an accident.SOLUTION: The present invention essentially relates to a solid-fuel nuclear reactor with a liquid metal or molten salt primary coolant, which simultaneously ensures, in particular: heat removal by forced convection in a primary circuit, in normal and accident operating modes, during shutdown through the primary vessel of the reactor, that is, beyond a second containment barrier; and implementation of a compact passive decay heat removal system capable of, in the event of an incident or accident, performing a safety function for a predetermined period, typically three days, by means of one or more PCMs that store decay heat produced in the core and removed by the primary vessel without any intervention by an operator.SELECTED DRAWING: Figure 3

Description

Technical Field

[0001] The present invention relates to the field of solid fuel nuclear reactors cooled by one or more molten salts or liquid metals, particularly liquid sodium, which are part of the fourth generation of nuclear reactors, called fast neutron reactors (FNR).

[0002] More specifically, the present invention relates to simplifying the architecture of these nuclear reactors while ensuring reliable removal of both nominal heat and decay heat in normal and accident shutdown situations.

[0003] The present invention is applicable to small modular reactors (SMR), and more specifically to micro modular reactors (MMR) typically having an operating power of less than 20 MWth.

[0004] Here, the decay heat of a nuclear reactor is the heat generated by the reactor core after the nuclear chain reaction has stopped, and it will be recalled that it mainly consists of the decay energy of fission products.

[0005] Decay heat corresponds to a part of the nominal heat and decreases over time. In any case, it must be considered during normal reactor shutdown stages, handling stages, and accident situations in order to determine the size of the heat removal system.

[0006] The present invention is described with reference to a nuclear reactor cooled by one or more molten salts, but it is also applicable to nuclear reactors cooled by any other liquid metal such as liquid sodium or lead used as a coolant in the primary circuit of the nuclear reactor.

[0007] Similarly, the present invention is described with reference to fast neutron reactors, but it is also applicable to thermal spectrum nuclear reactors or epithermal spectrum nuclear reactors.

Background Art

[0008] Fast neutron reactor systems have been developed to enable improved management of nuclear fuel, particularly through sustainable management of plutonium stocks and the ability to recover stocks of uranium isotope 238 that cannot be recovered in thermal neutron reactors.

[0009] The uranium-plutonium breeding cycle is only possible with high-energy neutrons. The resulting fast neutron spectrum is key to the sustainable management of plutonium stocks and to the potential to recover stocks of uranium 238 not used by conventional systems in light water-cooled thermal neutron reactors.

[0010] Solid fuel fast neutron reactors are based on the physical separation between the solid fuel and the coolant by a barrel that forms a physical barrier (cladding). The fuel itself is composed of materials that are solid at the target operating temperature, particularly in the form of oxides of fissile materials, silicon carbide (SiC), or nitrides, or in the form of direct metal alloys.

[0011] Optionally, the physical barrier separating the fuel and the coolant - functions as a containment barrier in the sense of a safety function that controls the containment of radioactive substances, particularly gaseous fission products that occur - enables implementing different technical configurations and management regulations for the fuel and the coolant to be possible.

[0012] In this context, major solid fuel fast neutron reactor projects have been developed around the concept of implementing a liquid metal or a gas to perform the coolant function and physically separating the liquid fuel from the coolant by a barrier.

[0013] Examples can include FNRs cooled by liquid sodium or liquid lead, or high-temperature gas-cooled reactors.

[0014] The system of FNR cooled by sodium (Na), in particular, an integral reactor, i.e., one in which the primary sodium circuit is arranged inside the reactor vessel, is the most mature technology. The feedback [1] obtained from the operation of several FNR-Na reactors both within and outside France highlights the following advantages: - Improved use of nuclear fuel, i.e., reduction of waste per unit of generated energy, which is effective for all fast neutron reactors, and - Due to a higher average coolant temperature of up to 550 degrees Celsius at the core outlet, an improved thermodynamic efficiency of approximately 42% compared to pressurized water reactors with a thermodynamic efficiency of about 32 - 33%, and - Very efficient heat conduction and cooling of the fuel through the use of a liquid metal with very high thermal conductivity, and - Significant natural convection in the primary circuit that contributes to decay heat removal (DHR) in accident situations, and - No pressurization of the coolant in the primary circuit due to a fairly wide operating range in the liquid state at ambient pressure from about 100°C to 880°C were emphasized.

[0015] However, integral FNR-Na reactors have the following drawbacks: - The risk of an exothermic reaction when an interaction occurs between sodium and water or air, · Along with strict sealing requirements, strict management of the amount of oxygen in the liquid sodium loop, and · In order to separate the radiation risk and the chemical risk, having an additional secondary sodium loop between the primary circuit and the energy conversion system compared to pressurized water reactors, and the risks that require, - The risk of boiling of the primary sodium in the event of an accident of loss of flow or loss of cooling without the actuation of the reactor emergency stop device that limits the operating temperature to 550°C at the core outlet due to the safety margin, and have.

[0016] The lead-cooled reactor (FNR-Pb) enables the generation of high-temperature heat with solid fuel while increasing the boiling margin of the coolant and eliminating risks associated with exothermic interactions between sodium and water, since the boiling temperature of lead is approximately 1750°C.

[0017] Nevertheless, lead-cooled reactors have problems of erosion / corrosion of components that deteriorate at high temperatures and very high coolant densities, which requires operation at temperatures limited to 500 - 550°C [2].

[0018] Two other well-known reactor technologies can essentially overcome or at least mitigate the risks associated with exothermic reactions between sodium and water and coolant boiling in case of accidents, and can generate high-temperature heat (>550°C) with solid fuel.

[0019] These are the gas-cooled fast reactor (GFR), the very high temperature gas reactor (VHTGR), and the molten salt-cooled solid fuel reactor called MSFR.

[0020] The VHTGR is characterized by solid fuel with very high thermal inertia, low power density, and very high melting temperature, which improves safety compared to sodium reactors. Reference can be made to FR2956773B1 or publication [3]. Pressurized gas is used to cool the fuel, which makes it possible to avoid risks associated with liquid coolant boiling in case of accidents. In addition, using gas allows for raising the temperature up to very high levels, typically up to 1000°C at the core outlet if the structural materials are suitable, achieving significant efficiency levels. Among the gases considered as coolants, helium has often been selected as the main candidate due to its thermal properties (higher conductivity and specific heat than other gases) and its chemical inertia.

[0021] However, having gas (helium) as the coolant has various drawbacks as follows, - The low power density of the fuel requires an increase in the volume of the primary vessel that houses the nuclear fuel, - Pressurization of the primary circuit is necessary to enhance the heat transfer of the gas and the efficiency of the facility which causes. This requires larger dimensions of the primary vessel, i.e., an increase in its wall pressure, - Helium is a complex fluid to use as it can penetrate solid barriers three times more easily than air [4]. Therefore, during the operation of the VHTR, helium stock losses must be taken into account, - To date, helium is not very abundant on Earth and is a costly material [5], - Operating risks associated with depressurization of the primary circuit exist.

[0022] To overcome the problems resulting from using liquid metal or gas as the coolant, liquid salts can be used as the coolant.

[0023] A very compact reactor architecture with a low power output of about 120 MWth using liquid salt as the coolant is described in publication [6], mainly shown in Figure 1 of said publication.

[0024] This architecture of the liquid salt cooled solid fuel reactor solves all of the problems associated with the use of helium described above. The operating temperature mentioned in publication [6] is about 500 - 550 °C, but the boiling / dissociation temperature of the salt is high enough (>1000 °C) to allow operation at higher temperatures without any risk of boiling.

[0025] However, the reactor architecture according to publication [6] has the following drawbacks, - The presence of an intermediate circuit within the primary vessel of the reactor, · The need to have feedthroughs in the closure on the primary vessel, and additional requirements related to the sealing of the boiler, · In the case of a high-speed neutron spectrum core, the risk of entrainment of gas bubbles into the core, which may cause an output peak, and · The potential activation of the intermediate fluid, and · A larger vessel size, and · The addition of components and systems, and the resulting reduced overall reliability due to a more complex configuration, and · A more complex and thus longer and / or more costly constructability and The presence of an intermediate circuit, including - A redundant structure within the primary vessel that must allow for the installation of an intermediate heat exchanger between the primary salt circuit and the secondary salt circuit, which may complicate its manufacture, and more generally, alternative in-vessel structures that are subject to regulatory constraints regarding manufacturability and inspection and may affect the facility's service life due to their irreplaceable nature, - The downward flow region of the cooling salt within the primary vessel, commonly referred to as the "downcomer," must be large enough to allow for the insertion of the intermediate heat exchanger. Liquid salts are poor coolants with very low thermal conductivities typically on the order of 0.5 to 1 W / mK compared to liquid metals such as sodium with a conductivity of about 60 W / mK, so the heat exchanger is necessarily a large component. However, a large primary vessel is more difficult to manufacture and transport in a factory, - The absence of a dedicated decay heat removal (DHR) safety system in case of an accident, - Fuel handling is done by removing the reactor closure in an inert atmosphere, which requires a significant facility downtime and a check of the reactor's seal before restart, having.

[0026] Therefore, there is a need to improve solid fuel reactors cooled using liquid metal or liquid salt, especially to overcome the aforementioned drawbacks.

[0027] Generally, the following, - Depending on the operating mode, operation with forced convection in the primary circuit within the output range between 20 MWth and 100 MWth, - A DHR function guaranteed by a preferably compact passive system, - Simplification of the internal structure of the primary vessel compared to well-known vessels, - Limitation of the feedthrough of the primary vessel, - Overall simplification of the heat removal function during normal reactor operation, - Operation at atmospheric pressure without chemical risk of exothermic interaction with water and air, - Radioactive substance protection fulfilling a third containment barrier function as close as possible to the nuclear material It is necessary to improve the safety of a solid fuel reactor cooled using a liquid metal or a liquid salt by complying with all points of the specification that can be defined as follows.

Prior Art Documents

Patent Documents

[0028]

Patent Document 1

Summary of the Invention

Problems to be Solved by the Invention

[0029] An object of the present invention is to at least partially meet this need / these needs.

Means for Solving the Problems

[0030] To achieve this, the present invention, in one of its aspects, relates to a reactor cooled using a liquid metal or one or more molten salts, the reactor comprising - A vessel called a primary vessel that is axisymmetric with respect to a central axis (X) and filled with a first coolant that uses at least one liquid metal or at least one inert liquid salt as the coolant for the primary circuit of the reactor, · A reactor core consisting of an assembly containing nuclear fuel material in solid state, housed in at least one barrel, · At least one pump for circulating a first coolant, · A structure forming a redan having a central axis coinciding with the axis of the primary vessel, during operation of the nuclear reactor, the pump circulates a liquid metal or molten salt coolant in a loop by forced convection from the bottom of the central zone where the nuclear reactor core where nuclear fission reaction occurs is disposed, the liquid metal or molten salt coolant rises by heating therefrom to the upper surface of the central zone, where it is diverted towards the upper surface of the peripheral zone so as to descend towards the bottom of the peripheral zone, where it is diverted towards the core of the nuclear reactor, the structure being disposed in the primary vessel so as to separate the interior of the primary vessel into a central zone and a peripheral zone, the structure and Comprising a vessel, - A vessel called a secondary vessel, disposed around the primary vessel, - A nuclear reactor pit disposed around the secondary vessel, - A nuclear reactor closure for enclosing the first coolant within the primary vessel, - A system for removing heat in both normal operation and in the situation where the nuclear reactor is shut down, the system being · A shell disposed between the primary vessel and the secondary vessel, defining the volume of the primary vessel filled with liquid metal, · A closed circuit called a secondary vessel, filled with a second coolant, capable of removing the heat itself removed by conduction through the primary vessel and transferred by the liquid metal to an energy conversion system and / or a heat network, Comprising a system, - A decay heat removal (DHR) system for removing decay heat in the accident situation of the nuclear reactor, the system being ·At least one solid-liquid phase change material (PCM) disposed inside the space delimited between the shell and the secondary container, wherein the PCM can melt while accumulating at least a part, preferably all, of the decay heat released from the core in the event of an accident by latent heat for a predetermined period of time. comprising, a system and comprising.

[0031] Advantageously, the stock of PCM makes it possible to absorb all of the decay heat from the core during a predetermined three-day duration by the energy (latent heat) required for the phase change. Thus, there is a sizing margin associated with the sensible heat of the molten salt in the liquid state, which makes it possible to absorb additional decay heat well before boiling and thus to leave a grace period of more than three days before intervention.

[0032] "Sensible heat" means the heat that a material can absorb without changing phase in a given state.

[0033] Here, in the context of the present invention, "shutdown situation" is given to mean the shutdown of a normal nuclear reactor and not the shutdown of the nuclear reactor in the event of an accident (accident situation).

[0034] "Inert liquid salt" is given to mean a liquid salt coolant that does not contain any fissionable element or breeding element and does not chemically react with water or air.

[0035] According to one advantageous embodiment, the circulation pump is a centrifugal pump that is vertically arranged and attached as a feed-through of the core head plug of the primary container, with its blades arranged above the redan.

[0036] Advantageously, the height of the PCM between the shell and the secondary container is greater than the height of the inert liquid salt between the primary container and the shell. This ensures that in the event of an accident, leakage of the liquid salt coolant through the primary container is limited or prevented. Similarly, preferably, the height of the liquid metal between the primary container and the shell is greater than the height of the inert liquid salt between the primary container and the shell.

[0037] According to one advantageous structural modification, the primary and secondary containers and the shell are straight circular cylinders arranged concentrically.

[0038] Preferably, the inert liquid salt is selected from chlorine-based salts optionally enriched with chlorine 37 to reduce the formation of radioactive Cl36, and more preferably is selected from NaCl, KCl, MgCl2, CaCl2, ZnCl2, or mixtures thereof, in particular the molten salt mixtures NaCl-MgCl2, NaCl-MgCl2-KCl, or NaCl-MgCl2-KCl-ZnCl2.

[0039] According to one advantageous variant embodiment, the closed circuit comprises a serpentine coil, preferably provided with heat dissipation fins around the serpentine coil, and the serpentine coil is arranged spirally around the shell between the primary container and the shell. Preferably, the serpentine coil can be fixed, in particular by welding to the shell.

[0040] The liquid metal of the bath between the primary container and the shell consists of pure aluminum.

[0041] Preferably, the PCM between the shell and the secondary container is in powder form.

[0042] More preferably, the PCM between the shell and the secondary container is made of pure aluminum.

[0043] More preferably, the primary container is made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC).

[0044] More preferably, the secondary container and the shell are made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC) according to the planned operating conditions.

[0045] The nuclear reactor according to the present invention can be a thermal spectrum nuclear reactor, a fast spectrum nuclear reactor, or a fast neutron reactor.

[0046] Thus, according to the first alternative, there is no moderator in the primary container so that the nuclear reactor operates using fast neutrons.

[0047] According to the second alternative, the reactor core contains at least one moderator so that the nuclear reactor operates using thermal neutrons or epithermal neutrons. The coolant can advantageously act as a moderator (a chloride salt or a fluoride salt containing lithium).

[0048] In the context of the present invention, "moderator" is given to mean any material capable of slowing down neutrons. In the normal sense, the kinetic energy of fast neutrons is greater than 1 eV, while the kinetic energy of thermal neutrons is less than 1 eV, typically about 0.025 eV. Reference can be made to publication [8], particularly Figure 4, which shows the thermal neutron fraction and the fast neutron fraction of the neutron flux in some types of nuclear reactors.

[0049] Reference can be made to FR3025650B1, which describes the insertion of a moderator into the fuel assembly of a fast neutron reactor.

[0050] According to one advantageous embodiment, the solid nuclear fuel is fuel pellets individually housed in separate cells of a nuclear fuel assembly and / or fuel particles called TRISO, and / or plates.

[0051] The solid nuclear fuel is degraded, preferably low-enriched with a concentration of <5%, or reprocessed (URT) uranium dioxide (UO2), and / or plutonium dioxide (PuO2), or preferably enriched uranium U with a concentration between 5% and 20%235 can be based on (HALEU, or High Assay Low Enriched Uranium).

[0052] Preferably, the reactor comprises a reactivity control system constituted by either control rods within the primary vessel or a rotating drum outside the primary vessel.

[0053] The reactor described above is in particular intended to have an output between 20 MWth and 100 MWth.

[0054] Therefore, the present invention essentially consists of - Heat removal by forced convection within the primary circuit, through the primary vessel of the reactor, i.e., during shutdown beyond the second containment barrier, in both normal and accident operating modes. - The implementation of a passive decay heat removal system that can perform safety functions for a predetermined period, typically three days, in a compact manner and without operator intervention, due to the presence of one or more PCMs that store the decay heat generated in the core and removed by the primary vessel. - A significant reduction in the risk of gas entrainment into the core due to the absence of an intermediate heat exchanger within the primary vessel. ○ Improving its reliability and inspectability, ○ Enabling the replacement of primary components that significantly extend the overall service life of the facility, ○ Simplifying its constructability, ○ Simplifying the demonstration of safety and thus helping to make it more robust, ○ Simplifying all maintenance and handling operations, especially fuel handling. - Simplifying the architecture of the primary circuit, the reactor pit, and the reactor vessel by using a heat removal fluid circuit without any feedthrough. - A very simplified second barrier with no singularities derived only from the limitations of the primary vessel. - The possibility of performing maintenance and inspection operations in a simplified manner beyond the second containment barrier, due to the position of the heat exchanger constituted by a closed circuit outside the primary circuit. It consists of manufacturing a solid fuel nuclear reactor using a primary vessel coolant of liquid metal or molten salt that guarantees both simultaneously.

[0055] Among the many advantages of the present invention are - Due to the reduction of pressure drop caused by the absence of an intermediate heat exchanger in the primary vessel, an active flow of coolant in the primary circuit, i.e., flow by forced convection, for reactors with a thermal output of 20 to several tens of MWth, typically up to 100 MWth at most, in normal / nominal operation and / or for decay heat removal. - Improved safety because the feedthroughs of the second containment barrier (primary vessel) are limited to reactivity control devices, i.e., control rod feedthroughs that pass through closures and are thus above the primary fluid level in the upper part of the primary vessel, and the circulation pumps of the primary vessel, and normal instrumentation devices. To further reduce the closure feedthroughs, it is also possible to assume a vessel reactivity control device in the form of a neutron reflector rotating around the primary vessel, for example, outside the primary vessel. - Simplification of the design of the primary vessel due to the absence of feedthroughs for heat removal, which limits singularities and enables a longer service life of the vessel. - Different from prior art solutions, there is no longer a need to incorporate an intermediate heat exchanger, so from a mechanical support and hydraulic point of view, simplification of the structures forming redundancy. - Elimination of the intermediate circuit. - A wider choice of fluids in the closed secondary circuit due to heat removal across the second barrier (primary vessel), and thus a wider choice of energy recovery methods. It is also possible to assume, for example, a pressurized fluid (gas or supercritical CO2) without the risk of having to consider the influence of gas bubble entrainment into the reactor core in the demonstration of safety.

[0056] A preferred application of the present invention is the fourth generation of small nuclear reactors, in particular, nuclear reactors cooled using sodium, lead, or liquid salts.

[0057] Further advantages and features of the present invention will become more clearly apparent by reading the following detailed description of exemplary embodiments of the present invention given as non - limiting examples with reference to the following figures.

Brief Description of the Drawings

[0058]

Figure 1

Figure 2

Figure 3

Figure 4

Modes for Carrying Out the Invention

[0059] Throughout this application, the terms "vertical", "lower", "upper", "bottom", "top surface", "downward", and "upward" should be understood with reference to a primary vessel filled with an inert liquid salt of a fast - neutron nuclear reactor according to the present invention in its vertical operating configuration.

[0060] FIGS. 1 to 4 show a liquid - salt - cooled fast - neutron nuclear reactor 1 according to the present invention.

[0061] Such a nuclear reactor 1 comprises a primary vessel 10 or a reactor vessel filled with an inert liquid salt called a primary salt S and containing a core 11 in which a plurality of fuel assemblies (not shown) that generate thermal energy by nuclear fission of fuel are immersed.

[0062] Therefore, the inert liquid salt S is the coolant of the primary circuit, stores and transports the heat from the core 11, and exchanges heat through the wall of the primary vessel 10. The primary salt has physical and chemical properties that enable it to remain in a liquid state at atmospheric pressure within the normal and accident operating temperature ranges of the core 11. Since it contains no breeding elements or fissile elements, it is inert from a radioactive perspective and is also chemically inert with respect to the PCM and liquid metal bath described in detail below, as well as all structures of the nuclear reactor.

[0063] The primary salt S is selected to have good thermal and physical properties to facilitate natural convection and heat exchange with the core 11 and through the wall of the primary vessel 10.

[0064] Therefore, the primary salt is preferably selected from NaCl, KCl, MgCl2, CaCl2, or ZnCl2 in which chlorine 37 is enriched, or a mixture thereof, particularly the molten salt mixtures NaCl-MgCl2, NaCl-MgCl2-KCl, or NaCl-MgCl2-KCl-ZnCl2. For example, NaCl-KCl-MgCl2, which has a melting point below 500 °C, good heat capacity (Cp), and good coefficient of thermal expansion, is advantageous for improving natural convection.

[0065] The solid nuclear fuel can be based on deteriorated, preferably low-enriched, <5% enrichment, or reprocessed (URT) uranium dioxide, and / or plutonium dioxide (PuO2), or preferably enriched uranium U 235 (HALEU, or high assay low-enriched uranium).

[0066] The fuel assembly can be provided with a SiC fuel cladding to withstand very high temperatures. The cladding of the fuel assembly forms a first containment barrier, and the vessel 10 forms a second containment barrier for the radioactive substances contained within the core 11.

[0067] The primary vessel 10 supports the weight of the liquid salt and internal structures of the primary circuit.

[0068] The core 11 is supported by a welded structure called a diagrid 12 where the legs of the fuel assembly 11 are located.

[0069] The core 11 is surrounded by a separation barrel 13 provided with a peripheral neutron reflector intended to ensure that the neutron flux is retained within the core. This separation barrel 13 of the core 11 enables the separation of the low-temperature primary salt and the high-temperature primary salt. Thus, the low-temperature primary salt surrounds the core 11 within the primary vessel, and the high-temperature primary salt heated by circulating upward within the core 11 is at the upper central part of the core.

[0070] Typically, the diagrid 12 is made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC) depending on the operating conditions.

[0071] As shown in the figure, the primary vessel 10 is a straight cylinder having a central axis X. Typically, the primary vessel 10 is made of AISI 316L stainless steel having a very low boron content, preferably, to prevent the risk of cracking at high temperatures. Its outer surface is made highly radioactive by a pre-oxidation treatment preferably carried out to facilitate the radiation of heat to the outside during the decay heat removal stage. The primary vessel 10 can also be made of a nickel-based alloy, or silicon carbide (SiC) depending on the operating conditions.

[0072] The reactor vessel 10 is equipped with a pump 100 for circulating an inert liquid salt as the primary coolant.

[0073] The reactor vessel 10 is divided into two different zones by a separation structure consisting of at least one shell 14 arranged inside the reactor 10. This separation structure is also known as a redan.

[0074] As shown in FIG. 3, the redan 14 having a central axis X can include three shells 140, 141, 142 welded to each other, namely, - an upper shell 140 in the shape of a straight cylinder, - a central shell 141 in the shape of a frustum of a cone, - a bottom shell 142 in the shape of a straight cylinder that surrounds the separation barrel 13. It can be provided with these.

[0075] The redan 14, and more specifically, its bottom shell 142 can be welded to the diagrid 12 as shown in FIG. 3 and is disposed and welded at the bottom of the primary container 10.

[0076] The redan 14 further includes a through opening 143 made through the bottom shell 142.

[0077] As indicated by the arrows in FIG. 1, during the operation of the nuclear reactor, the inert liquid salt circulates by natural convection in the loop from the bottom of the central zone where the nuclear reactor core 11 is disposed, the pump 100 circulates the liquid metal or molten salt coolant in the loop by forced convection from the bottom of the central zone where the nuclear reactor core 11 is disposed, the liquid metal or molten salt coolant rises from there to the upper surface of the central zone by heating, is diverted toward the upper surface of the peripheral zone so as to descend toward the bottom of the peripheral zone there, and is diverted toward the core 11 there. Thus, the redan 14 is disposed in the primary container 10 so as to form a central chimney that separates the inside of the primary container 10 into a central zone and a peripheral zone.

[0078] Typically, the redan 14 is made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC) according to the operating conditions.

[0079] Advantageously, as shown in FIG. 3, in order to reduce the flow area of the inert liquid salt S in the peripheral zone when the inert liquid salt S descends toward the bottom, the reducer 14 can include, at its upper surface portion, a deflector 144 and a section reducer 145, each disposed around the upper shell 140. This makes it possible, in particular, to increase the speed of the liquid salt S in this downflow zone, which is usually referred to as a "downcomer". It also makes it possible to advantageously improve the heat transfer by convection from the salt S to the wall of the primary vessel 10.

[0080] The central chimney shape of the reducer 14 improves the circulation by natural convection of the liquid salt S.

[0081] A removable plug 15, called a core head plug, is disposed directly above the core 11, contains the liquid salt S, functions as a barrier between the liquid salt S and the external environment, and closes the primary vessel 10 to form a second containment barrier for the substances contained in the core 11 together with the primary vessel.

[0082] Similar to the primary vessel 10, the core head plug 15 is chemically inert with respect to the PCM and the liquid metal bath, which will be described in detail below, and with respect to all the structures of the nuclear reactor.

[0083] The core head plug 15 is provided with feedthroughs for the components of the control rods 16, as well as feedthroughs for the elements for controlling and monitoring the core 11 and the liquid salt S, not shown. Thus, the core head plug 15 is a plug that can be removed in an inert atmosphere and is equipped with all the handling systems and all the instrumentation necessary for monitoring the core, including control rods, the number of which depends on the type of core and its output, as well as thermocouples and other monitoring devices. A system for maintaining the temperature of the plug is provided to limit the risk of deposition of salt aerosols.

[0084] Typically, the core head plug 15 is made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions. Also typically, the material used for the control rods is B4C.

[0085] In one variant, in order to advantageously eliminate the feedthrough of the plug 15 and thus advantageously eliminate the risk of fouling associated with the salt aerosol, instead of the control rods, a rotating drum, also made of B4C, can be arranged outside the primary vessel 10.

[0086] As shown in FIG. 3, the pump 100 for circulating the primary salt is preferably a centrifugal pump arranged vertically and attached as a feedthrough of the core head plug 15 of the primary vessel, with its blades 101 arranged above the redan 14.

[0087] The reactor vessel 10 comprises, above the molten salt liquid fuel, a plenum, usually called a cover gas plenum, filled with an inert gas such as argon or helium. This plenum absorbs the thermal expansion of the liquid when the liquid level in the reactor vessel changes and enables the recovery of gaseous fission products generated by nuclear fission in the fuel.

[0088] A support, containment, and insulation assembly between the primary vessel and the external environment E is arranged around the primary vessel 10.

[0089] More specifically, as shown in FIGS. 1 and 3, this assembly 2 comprises a reactor pit 20, and within the reactor pit 20, a layer of insulating material 21, a secondary vessel 22, and the primary vessel 10 of the reactor are inserted from the outside towards the inside.

[0090] The reactor pit 20 is a block having a generally cylindrical outer shape that supports the weight of all components inside it. The reactor pit 20 has the function of providing biological protection and protection against external attacks, and also has the function of providing cooling of the external environment to maintain a low temperature. Typically, the reactor pit 20 is a concrete block.

[0091] The layer 21 of heat-insulating material ensures heat insulation of the reactor pit 20. Typically, the layer 21 is made from polyurethane or silicate-based foam.

[0092] The secondary container 22 guarantees that the liquid salt S is retained in case of leakage from the primary container 10 and protects the reactor pit 20. The secondary container 22 also includes a volume 23 of a solid-liquid phase change material (PCM).

[0093] Preferably, the volume 23 of the PCM in powder form can melt while storing at least a part of the decay heat released by the core in latent heat for a predetermined duration in an accident situation. Thus, the physical and chemical properties of the PCM enable the PCM to remain in a solid state at atmospheric pressure within the normal operating temperature range of the core and to guarantee a phase change for a predetermined duration when the core 11 enters an accident situation.

[0094] As shown in FIG. 2, the stock of the total volume 23 of the PCM makes it possible to ensure a height H that is higher than the height of the inert liquid salt S in the primary container 10. Such a height serves to hold all of the primary salt S within the primary container 10 in case of leakage at the wall of the primary container 10.

[0095] The PCM is chemically inert with respect to the liquid metal bath and all structures of the reactor.

[0096] Typically, the PCM is made from pure aluminum. As a variant, the metallic PCM can be replaced by a salt PCM, such as MgCl2.

[0097] The secondary container 22 is located opposite to the reactor pit 20, and its upper surface is welded to the reactor closure 17.

[0098] Typically, the secondary container 22 can be made of AISI 316L stainless steel, or nickel-based alloy, or silicon carbide (SiC) according to the operating conditions.

[0099] The annular removable closure 24 around the core head plug 15 encloses the PCM volume 23 and closes the secondary container 22 to function as a barrier between this volume 23 and the external environment E.

[0100] Similar to the secondary container 22, the closure 24 is chemically inert with respect to the PCM and the liquid metal bath, which will be described in detail below, and with respect to all structures of the reactor.

[0101] The closure 24 is provided with feedthroughs for elements for controlling and monitoring the PCM stock (not shown), which meet the sealing requirements of the second containment barrier. In the handling stage, operations that require the removal of this closure must be carried out in an inert atmosphere.

[0102] Typically, the closure 24 can be made of AISI 316L stainless steel, or nickel-based alloy, or silicon carbide (SiC) according to the operating conditions.

[0103] The reactor 1 further comprises a system 3 for removing heat both during normal operation and in the situation where the reactor is shut down.

[0104] This system 3 first comprises a cylindrical shell 30 arranged concentrically between the primary container 10 and the secondary container 22.

[0105] Therefore, this shell 30 defines a volume with the primary container 10 filled with the liquid metal bath 31.

[0106] Typically, the shell 30 can be made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions, and is designed to be chemically and mechanically resistant to the liquid metal bath 31 and the volume 23 of the PCM.

[0107] The closed circuit 32, called the secondary circuit, is filled with a coolant and can remove the heat removed by conduction through the primary vessel 10 and transferred by the liquid metal bath 31 to an energy conversion system and / or a heat network (not shown).

[0108] Thus, the liquid metal bath 31 improves heat transfer by conduction from the primary vessel 10 to the secondary circuit 32.

[0109] Typically, the liquid metal bath is made of pure aluminum.

[0110] The liquid metal is chemically inert with respect to the liquid salt S, the PCM, and all the structures of the nuclear reactor.

[0111] The closed circuit 32 preferably consists of a serpentine coil arranged spirally around the primary vessel 10 and preferably welded to the inner wall of the shell 30.

[0112] The serpentine coil 32 has a diameter that depends on the diameter of the primary vessel 10 and a height sufficient to have the surface area required for the desired heat removal.

[0113] In other words, the total number, spacing, and diameter of these turns that make up the serpentine coil 32 depend on the diameter of the primary vessel 10 and the output of the nuclear reactor core 11. For example, the pitch of the turns of the serpentine coil 32 can be equal to 10 cm, which is a good compromise between manufacturing and heat absorption by conduction by the liquid metal bath.

[0114] Again, for example, in order to minimize the pressure drop, reduce the ground contact area of the pipe, and maximize the surface area exposed to the primary vessel 10, the outer diameter of the serpentine coil 32 is set to about 5 to 10 cm, and the turn pitch is set to about 10 to 15 cm. The thickness of the serpentine coil 32 depends on the mechanical stress applied by the internal liquid metal and its weight.

[0115] The material of the serpentine coil 32 must have good emissivity characteristics. Typically, the material of the serpentine coil is selected from AISI 316L stainless steel, ferritic steel, and nickel-based alloys. This material depends on the internal fluid used in the closed circuit 32.

[0116] This internal coolant circulating inside the serpentine coil 32 is a chemically stable low-viscosity liquid metal that is a good heat conductor and heat medium, is chemically compatible with all the pipes of the circuit 3, and can operate in natural convection or forced convection in the temperature range of 150 to 600 °C. Typically, the liquid metal of the circuit 3 can be selected from among a NaK alloy, a Pb-Bi alloy, sodium, or a ternary alloy of liquid metals.

[0117] To improve the heat exchange between the liquid metal bath 31 and the pipe forming the serpentine coil 32, the serpentine coil can be provided with heat dissipation fins 33 that extend radially from the pipe, particularly in a linear shape, as shown in FIG. 4.

[0118] Here, the operation of the nuclear reactor 1 will be described with respect to different normal situations, planned shutdown situations, and accident situations.

[0119] During normal operation of the nuclear reactor, all of the heat generated by the nuclear fission reaction in the reactor core 11 is removed by heat exchange when the liquid salt S in forced convection by the pump 100 passes through the reactor core. When the liquid salt S falls between the riser 14 and the primary vessel 10, it exchanges this heat through the wall of the primary vessel 10. Then, this heat is transferred by the liquid metal bath 31 so as to be removed by the secondary circuit 32 to the heat network and / or the energy conversion system. During normal operation, the volume 23 of the PCM remains in the solid state.

[0120] In the case of a planned shutdown caused by an operator, the same heat exchange takes place. The decay heat from the shutdown reactor core 11 is also removed by heat exchange when the liquid inert salt S passes through the reactor core 11. Then, when the salt S falls between the riser 14 and the primary vessel 10, it exchanges this heat through the wall of the primary vessel 10. Then, this heat is transferred by the liquid metal bath 31 so as to be removed by the secondary circuit 32 to the heat network and / or the energy conversion system. During this planned shutdown, the volume 23 of the PCM remains in the solid state.

[0121] During operation in an accident situation, particularly in the case of a station blackout (SBO) corresponding to a complete loss of power supply as in the Fukushima accident, the decay heat from the shutdown reactor core 11 is removed by heat exchange when the liquid salt S passes through the reactor core. Then, when the salt S falls between the riser 14 and the primary vessel 10, it exchanges this heat through the wall of the primary vessel 10. This heat is absorbed by the PCM which was initially solid, and the PDM changes from the solid state to the liquid state. The stock of the volume 23 of the PCM enables it to absorb all of the decay heat from the reactor core for a predetermined duration, typically three days, by the energy (latent heat) required for the phase change.

[0122] The inventors conducted a sizing study to demonstrate the feasibility of the nuclear reactor 1 as described above and to propose the order of magnitude for its characteristic elements.

[0123] The conducted research covers the range of 20 - 100 MWth in normal operation and accident operation, involving the circulation of primary salt S by forced convection only by pump 100.

[0124] The low temperature of the inert liquid salt S as the primary fluid is set at 600 °C.

[0125] The hydraulic path of salt S is determined by the analytical study shown in the following table.

[0126] The thermal output range is defined within the range where reactor 1 is intended to operate without a heat exchanger in the primary vessel by forced convection only by pump 100.

[0127] The related characteristics are summarized in Table 1 below.

[0128]

Table 1

[0129] Note that the characteristics of the salt in question are as described in publication

[11] .

[0130] These input data for variable power between 20 MWth and 100 MWth were used by the inventors in preliminary calculations using thermal calculation software such as COPERNIC: [9],

[10] .

[0131] These sizing calculations were performed according to the following two consecutive steps.

[0132] Step 1 / : A core configuration with the lowest possible resistance to heat flow is selected. The first core sizing is required to calculate the total pressure drop of the primary fluid pressure circuit.

[0133] Step 2 / : The flow of primary salt S by forced convection only within primary vessel 10, and heat transfer by forced convection only through the wall of primary vessel 10.

[0134] In Step 1 / , a preliminary reactor core design with a nuclear fuel pin assembly was determined for a power interval of 20 to 100 MWth. In this calculation, the hydraulic area was maximized by applying the physical quantities given in Table 2.

[0135]

Table 2

[0136] Considering the physical quantities in Table 2, the fluid pressure flow area of salt S is maximized by the nuclear fission height value of the fuel and the number of fuel pins in the reactor core. The larger the area for a given flow rate, the lower the pressure drop in the reactor core. Thus, having a maximized fluid pressure flow area improves the fluid flow.

[0137] The preliminary sizing calculations for reactor core 11 are summarized in Table 3 for thermal outputs of 30 MWth, 40 MWth, and 100 MWth, respectively.

[0138] Note that the fuel assembly in question has a hexagonal sheath.

[0139]

Table 3

[0140] The results in Table 3 provide input data for the sizing calculation of the primary circuit. This determines the temperature difference on the inner wall of the primary vessel 10 required to remove the heat carried by the primary salt S.

[0141] The following Table 4 shows the results of the reactor operation calculation based on the above calculations. It is specified that the velocity of the salt in the annular space is five times that obtained from circulation by natural convection only.

[0142]

Table 4

[0143] From the results of Table 4, it can be seen that the increase in the salt velocity contributes to a significant improvement in the heat transfer coefficient between salt S and the vessel wall.

[0144] Compared with the architecture by natural convection, the forced convection architecture according to the present invention has the following advantages: - A decrease in the temperature difference between salt S and the wall, accompanied by an increase in the temperature of the fluid at the outlet of the heat exchanger in the form of the serpentine coil 32 to about 600 °C, and thus an increase in the electrical efficiency of the electrical conversion system connected to the heat exchanger. - A potential reduction in the height of the primary vessel from a fixed height of 20 m to 15 m or 10 m in the calculation to promote only the natural circulation of salt S. has.

[0145] The present invention is not limited to the examples described above, and the features of the illustrated examples can be combined, in particular, within modifications not shown.

[0146] Further modifications and embodiments can be envisaged without departing from the scope of the present invention.

[0147] (References) [1]: H. OHSHIMA et al., “Handbook of Generation IV Nuclear Reactors”, chapter 5, Elsevier, 2016. [2]: M. TARANTINO et al., “Overview on Lead-Cooled Fast Reactor Design and Related Technologies Development in ENEA”, MDPI Energies, vol. 14, p. 5157, 2021. [3]: M.A. FUTTERER et al., “Status of the very high temperature reactor system”, Progress in Nuclear Energy, vol. 77, pp. 266 - 281, 2014. [4]: GLENN D. CONSIDINE ed., “Van Nostrand’s Encyclopedia of Chemistry”, Wiley - Interscience, 5th edition, 2005. [5]: A.H.O. H. USverdrup, “Assessing the Past and Future Sustainability of Global Helium”, Biophysical Economics and Sustainability(2020), 2020. [6]: L. LIN et al., “Feasibility of an innovative long - life molten chloride - cooled reactor”, Nuclear Science and Techniques, vol. 33, pp. 1 - 15, 2020. [7]: http: / / sme.vimaru.edu.vn / sites / sme.vimaru.edu.vn / files / volume_2_-_properties_and_selection_nonf.pdf [8]: Jiri Krepel et al. “Self - Sustaining Breeding in Advanced Reactors: Characterization of Selected Reactors”, Encyclopedia of Nuclear Energy 2021, Pages 801 - 819. https: / / www.sciencedirect.com / science / article / pii / B9780128197257001239?via%3Dihub [9]: F.MORIN et al., “COPERNIC, A NEW TOOL BASED ON SIMPLIFIED CALCULATION METHODS FOR INNOVATIVE LWRs CONCEPTUAL DESIGN STUDIES”, ICAPP 2017 Conference, 2017.

[10] : P.GAUTHE et al., “Innovative and inherently safe small SFR as a response to the dilemma ‘safety vs cost’”, ICAPP 2019 Conference, 2019.

[11] : Y.LI et al., “Survey and evaluation of equations for thermophysical properties of binary / ternary eutectic salts from NaCl, KCl, MgCl2, CaCl2, ZnCl2 for heat transfer and thermal storage fluids in CSP”, Solar Energy, vol.152, pages 57 - 79, 2017.

Explanation of symbols

[0148] 1 Liquid - salt - cooled fast neutron reactor, reactor 2 Assembly 3 System, circuit 10 Primary vessel, vessel, reactor vessel 11 Core, fuel assembly, reactor core 12 Diagrid 13 Separation barrel 14 Shell, redan 15 Removable plug, core head plug 17 Reactor closure 20 Reactor pit 21 Layer of insulation material, layer 22 Secondary vessel 23 Volume, volume of PCM 24 Annular removable closure, closure 30 Cylindrical shell, shell 31 Liquid metal bath 32 Closed circuit, secondary circuit, serpentine coil 33 Heat dissipation fin 100 Pump 101 Blade 140 Shell, top shell 141 Shell, central shell 142 Shell, bottom shell 144 Deflector 145 Section reducer

Claims

1. A nuclear reactor (1) cooled by a liquid metal or one or more molten salts, said nuclear reactor comprising: - A vessel (10), called a primary vessel, which is axisymmetric with respect to a central axis (X) and is filled with a first coolant using at least one liquid metal or at least one inert liquid salt as the coolant for the primary circuit of said nuclear reactor, - A reactor core (11) consisting of an assembly containing a nuclear fuel material in a solid state, housed in at least one barrel, - At least one pump (100) for circulating said first coolant, - A structure forming a redan (14) having a central axis coinciding with the axis of said primary vessel, during operation of said nuclear reactor, said pump causing the liquid metal or molten salt coolant to circulate in a loop by forced convection from the bottom of a central zone where a nuclear reactor core (C) where nuclear fission reactions occur is located, and said liquid metal or molten salt coolant rising therefrom by heating to the upper surface of said central zone, where it is diverted towards the upper surface of said peripheral zone towards the bottom of said peripheral zone, where it is diverted towards said reactor core of said nuclear reactor, said structure being arranged in said primary vessel so as to separate the interior of said primary vessel into said central zone and said peripheral zone, a structure comprising a vessel, - A vessel (22), called a secondary vessel, arranged around said primary vessel, - A nuclear reactor pit (20) arranged around said secondary vessel (22), - A core head plug (15) for enclosing said first coolant within said primary vessel, - A system for removing heat in both normal operation and in a situation where said nuclear reactor is shut down, said system comprising: - A shell (30) arranged between said primary vessel and said secondary vessel and defining the volume of said primary vessel filled with said liquid metal (31), - A closed circuit (32), called a secondary circuit, filled with a second coolant and capable of removing the heat itself removed by conduction through said primary vessel and transferred by said liquid metal to an energy conversion system and / or a heat network, comprising a system, - A system for removing decay heat (DHR) in an accident situation of said nuclear reactor, said system comprising - At least one solid-liquid phase change material (PCM) disposed inside the space delimited between the shell and the secondary container, the PCM being capable of melting while accumulating at least a part of the decay heat released from the core in the accident situation by latent heat for a predetermined period, at least one PCM A system including A nuclear reactor (1) comprising

2. The circulation pump is a centrifugal pump (100) vertically arranged and attached as a feed-through of the core head plug (15) of the primary container, and its blade (101) is arranged above the redan. The nuclear reactor (1) according to claim 1

3. The height of the PCM between the shell and the secondary container is greater than the height of the inert liquid salt between the primary container and the shell. The nuclear reactor (1) according to claim 1

4. The primary and secondary containers and the shell are straight cylinders arranged concentrically. The nuclear reactor (1) according to claim 1

5. The inert liquid salt is a chlorine-based salt. The nuclear reactor (1) according to any one of claims 1 to 4

6. The closed circuit includes a serpentine coil (32), and the serpentine coil is arranged spirally around the shell between the primary container and the shell. The nuclear reactor (1) according to claim 1

7. The liquid metal of the bath between the primary container and the shell consists of pure aluminum. The nuclear reactor (1) according to claim 1

8. The PCM between the shell and the secondary container is in powder form. The nuclear reactor (1) according to claim 1

9. The PCM between the shell and the secondary container is made of pure aluminum. The nuclear reactor (1) according to claim 1

10. The primary container is made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC). The nuclear reactor (1) according to claim 1

11. The secondary container and the shell are each made of AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC). The nuclear reactor (1) according to claim 1

12. The primary container has no moderator so that the nuclear reactor operates using fast neutrons. The nuclear reactor (1) according to claim 1

13. The nuclear reactor (1) according to claim 1, wherein the reactor core contains at least one moderator so that the nuclear reactor operates using thermal neutrons or epithermal neutrons.

14. The nuclear reactor (1) according to claim 1, wherein the solid nuclear fuel is a nuclear fuel assembly and / or fuel particles called TRISO, and / or fuel pellets individually housed in separate cells of plates.

15. wherein the solid nuclear fuel is based on deteriorated, low-enriched, or reprocessed (URT) uranium dioxide (UO ２ ), and / or plutonium dioxide (PuO ２ ), or enriched uranium U ２３５ (HALEU, or high assay low enriched uranium), the reactor (1) according to claim 1.

16. The nuclear reactor (1) according to claim 1, comprising a reactivity control system constituted by either a control rod in the primary vessel or a rotating drum outside the primary vessel.

17. The nuclear reactor (1) according to claim 1, having an output between 20 MWe and 100 MWe.

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