Nuclear reactor with forced convection liquid coolant and solid fuel assemblies, incorporating a liquid metal bath and material(s) (MCP) rated power evacuation system for residual power evacuation in case of accident.

Abstract

Forced convection liquid coolant nuclear reactor and solid fuel assemblies, incorporating a liquid metal bath and material(s) (MCP) rated power evacuation system for residual power evacuation in case of accident.The invention essentially consists of a solid-fuel nuclear reactor with a primary coolant of liquid metal or molten salt(s) that guarantees, in particular: - forced convection extraction of thermal power to the primary circuit in normal and accidental operating modes, even when the reactor is shut down, through the primary vessel, i.e., beyond the second containment barrier; - in the event of an incident or accident, the implementation of a compact passive residual heat removal system capable of ensuring the safety function for up to a predetermined period, typically 3 days, without any operator intervention, thanks to the presence of PCM material(s) that store the residual heat produced in the core and removed by the primary vessel. Figure 3.

Description

Title of the invention: Nuclear reactor with liquid coolant in forced convection and solid fuel assemblies, integrating a nominal power evacuation system with liquid metal bath and material(s) (MCP) for the evacuation of residual power in the event of an accident. Technical field

[0001] The present invention relates to the field of solid fuel nuclear reactors cooled with one or more heat transfer fluids of the molten salt or liquid metal type, in particular with liquid sodium, known as RNR (Fast Neutron Reactors) which are part of the family of so-called fourth generation reactors.

[0002] More particularly, the invention relates to a simplification of the architecture of these nuclear reactors while guaranteeing reliable evacuation of both nominal and residual power in the case of normal and accidental shutdown situations.

[0003] The invention applies to small power reactors or SMRs in English (acronym for “Small Modular Reactor”), and more specifically to MMRs (acronym for “Micro Modular Reactor”) typically with an operating power of less than 20MWth.

[0004] It is recalled here that the residual power of a nuclear reactor ("decay heat" in English) is the heat produced by the core after the nuclear chain reaction has stopped and consists mainly of the disintegration energy of the fission products. The residual power corresponds to a fraction of the nominal power and decreases over time. In all cases, it must be taken into account during normal shutdown phases of the reactor, during handling phases, but also during accident situations in order to size heat removal systems.

[0005] Although described with reference to a nuclear reactor cooled by one or more molten salts, the invention also applies to reactors cooled by liquid sodium or any other liquid metal, such as lead, as heat transfer fluid for a primary circuit of a nuclear reactor.

[0006] Also, although described with reference to a fast neutron reactor, the invention also applies to reactors operating in thermal and / or epithermal spectrum. Prior art

[0007] The fast neutron reactor sector was developed to enable a better management of nuclear fuel, in particular through sustainable management of the plutonium stock and the ability to recover the inventory of the uranium isotope 238, an isotope that cannot be recovered in thermal neutron reactors.

[0008] The uranium / plutonium regeneration cycle is only possible with high-energy neutrons. The resulting fast neutron spectrum is the key to sustainable management of the plutonium inventory and the possibility of recovering unused uranium 238 stocks through the more conventional process of light-water-cooled thermal neutron reactors.

[0009] Solid-fuel fast neutron reactors rely on a physical separation between the solid fuel and the coolant, by a casing forming a physical barrier (cladding). The fuel itself is made of solid materials at the target operating temperatures (in particular in the form of oxides, silicon carbides (SIC) or nitrides of fissile materials or directly in the form of a metal alloy).

[0010] Functionally, the physical barrier separating the fuel and the heat transfer fluid makes it possible to: - play the role of containment barrier in the sense of the safety function of controlling the containment of radioactive materials, in particular the containment of fission products generated in the gas phase, - implement a technical configuration and provisions for differentiated management of fuel and heat transfer fluid.

[0011] In this context, the main projects for solid-fuel fast neutron reactors have developed around concepts using liquid metals or gases to carry out the function of heat transfer fluid, and with a solid fuel physically separated from the heat transfer fluid by a barrier.

[0012] We can cite the RNR type systems cooled with sodium or liquid lead, or the high temperature reactor systems with gas heat transfer fluid.

[0013] The sodium (Na) cooled RNR sector is the most mature technology, particularly that of reactors in integrated configuration, i.e. with a sodium primary circuit located inside the reactor vessel. Feedback from the operation of several Na-RNRs in France and abroad [1] has highlighted the following advantages: - better use of nuclear fuel, i.e. a reduction in waste per unit of energy produced, which is valid for all fast neutron reactors; - a better thermodynamic efficiency, of around 42% compared to pressurized water reactors which is around 32-33%, thanks to a higher average coolant temperature since up to 550°C at the core outlet reactor; - very efficient heat transfer and fuel cooling thanks to the use of a liquid metal with very high thermal conductivity; - significant natural convection in the primary circuit, which contributes to the evacuation of residual power (EPuR) in accident situations; - the absence of pressurization of the heat transfer fluid in the primary circuit, given its fairly wide operating range in the liquid state at ambient pressure, from approximately 100°C to 880°C.

[0014] However, the integrated configuration of Na-NRs has the following disadvantages: - a risk of exothermic reaction in the event of interaction of sodium with water or air, which implies: • strict control of the quantity of oxygen in the liquid sodium loops, as well as strong sealing constraints; • having a secondary sodium loop in addition to pressurized water reactors between the primary circuit and the energy conversion system, in order to separate the radiological risk and the chemical risk; - risks of boiling of primary sodium in the event of a loss of flow or loss of cooling accident without activation of the reactor's emergency shutdown devices, which limits, with safety margins, the operating temperature to 550°C at the core outlet.

[0015] Lead-cooled reactors (RNR-Pb) make it possible to de facto eliminate the risks linked to the exothermic sodium / water interaction and to produce heat at high temperature by means of a solid fuel, while increasing the boiling margin of the heat transfer fluid, since the boiling temperature of lead is around 1750 °C.

[0016] Nevertheless, lead-cooled reactors present problems of erosion / corrosion of the components, increased at high temperatures and for very high heat transfer densities, which obliges them to operate at temperatures limited to 500-550 °C: [2].

[0017] Two other known reactor technologies are intrinsically capable of resolving or at least mitigating the risks associated with the exothermic sodium / water interaction and the boiling of the coolant in the event of an accident, and of producing high temperature heat (>550°C) by means of a solid fuel.

[0018] These are gas-cooled reactors (GFR for "Gas-cooled Fast Reactors") and very high temperature gas reactors (VHTGR for "Very High Temperature Gas Reactors") and combustion reactors. solid fuel and cooled by a liquid salt, called RNR-salt.

[0019] VHTGR reactors are characterized by a massive fuel, whose very high thermal inertia, low power density and very high melting temperature improve safety compared to sodium reactors. Reference may be made to patent FR2956773B1 or to publication [3]. To cool the fuel, a pressurized gas is used, which avoids the risks associated with the boiling of a liquid coolant in the event of an accident. In addition, the use of a gas makes it possible to reach very high temperatures as long as the structural materials are compatible, typically up to 1000 °C at the core outlet, and to achieve high efficiencies. Among the gases that have been considered as heat transfer fluids, helium has often been chosen as the main candidate due to its thermal properties (higher conductivity and specific heat compared to other gases) and its chemical inertia.

[0020] However, having a gas (helium) as a heat transfer fluid involves the following various disadvantages: - the low power density of the fuel requires increasing the volume of the primary tank which contains the nuclear fuel; - in order to increase the heat transfer of the gas and the efficiency of the installation, pressurization of the primary circuit is necessary. This implies a significant sizing of the primary tank, i.e. an increase in its wall thickness; - helium is a complicated fluid to use because it can penetrate solid barriers three times more easily than air: [4]. Also, a loss of helium inventory is to be considered during the operation of a VHTGR reactor; - to date, helium is a material that is not very abundant on earth and whose cost is high: [5]; - an operating risk linked to the depressurization of the primary circuit.

[0021] In order to solve the problems arising from the use of a liquid metal or a gas as a heat transfer fluid, a liquid salt can be used as a heat transfer fluid.

[0022] A nuclear reactor architecture using a liquid salt as a heat transfer fluid, of low power, of approximately 120 MWth, very compact is described in publication [6], and illustrated mainly in [Fig.l] of this publication.

[0023] This liquid salt cooled reactor architecture with solid fuel solves all the problems related to the use of helium mentioned above. Even if the operating temperature mentioned in the publication [6] is about 500-550°C, the boiling / dissociation temperature of the salt is high enough (> 1000°C) to allow operation at higher temperatures without risk boiling.

[0024] However, the reactor architecture according to publication [6] has the following drawbacks: - the presence of an intermediate circuit within the primary reactor vessel, which implies: • the need to have slab crossings above the primary tank and additional constraints linked to compliance with the boiler's watertightness; • a risk of gas bubbles being drawn into the core which, when it has a fast neutron spectrum, can cause a power peak; • potential activation of the intermediate fluid; • a larger tank size, • reduced overall reliability due to the addition of components and systems and a more complex configuration, • more complex construction, therefore longer and / or more expensive, - a stepped structure inside the primary tank which must allow the installation of an intermediate exchanger between the primary salt circuit and the secondary salt circuit, which probably complicates its manufacture; and more broadly, the internals of non-replaceable primary tanks subject to regulatory constraints on manufacturability and inspection and which may have an impact on the lifespan of the installation due to their non-replaceable nature, - the downward flow zone of the heat transfer salt in the primary tank, usually called "Downcomer" must be wide enough to allow the insertion of an intermediate exchanger. In fact, the latter is necessarily a large component because a liquid salt is a poor heat transfer fluid, with very low thermal conductivity, typically of the order of 0.5 to 1 W / mK, compared to liquid metals, such as that of sodium which is equal to approximately 60 W / mK. However, a wide primary tank is more difficult to manufacture in a factory and transport; - the absence of safety systems dedicated to the Evacuation of Residual Power (EPuR) in the event of an accident; - fuel handling is carried out by removing the reactor slab under an inert atmosphere, which requires significant downtime of the installation, as well as checks on the reactor's leaktightness before restarting.

[0025] There is therefore a need to improve solid fuel nuclear reactors cooled with liquid metal or liquid salt(s), in particular in order to overcome the in- aforementioned disadvantages.

[0026] Generally speaking, there is a need to improve the safety of solid fuel nuclear reactors cooled with liquid metal or liquid salt(s), by meeting all the points of a specification which can be defined as follows:

[0027] - operation with forced convection in the primary circuit over ranges of powers between 20 and 100 MWth depending on the operating modes,

[0028] - a guarantee of the EPuR function by a passive system, preferably compact,

[0029] - a simplification of the internal structures of the primary tank compared to those known,

[0030] - a limitation of the crossings of the primary tank,

[0031] - an overall simplification of the power evacuation function in function normal operation of the reactor,

[0032] - operation at atmospheric pressure without chemical risk of interaction exothermic with water and air,

[0033] - radiological protection as close as possible to nuclear materials, ensuring the 3rd containment barrier function.

[0034] The aim of the invention is to respond at least in part to this(these) need(s). Statement of the invention

[0035] To this end, the invention relates, in one of its aspects, to a nuclear reactor cooled by liquid metal or molten salt(s), comprising: - a vessel, called the primary vessel, axisymmetric around a central axis (X), filled with a first heat transfer fluid with at least one liquid metal or at least one inert liquid salt as heat transfer fluid for the primary circuit of the reactor, comprising: • a core made up of assemblies containing combustible nuclear material in the solid state, confined in at least one envelope; • at least one circulation pump for the first heat transfer fluid; • a structure forming a step, with a central axis coinciding with that of the primary vessel, the structure being arranged in the primary vessel to separate the interior of the latter into a central zone and a peripheral zone so that during operation of the reactor, the pump(s) circulate by forced convection the liquid metal or molten salt(s) coolant in a loop from the bottom of the central zone where the reactor core is arranged within which the fission reactions occur, from which it rises by heating to the top of the central zone where it is diverted towards the top of the peripheral zone to descend towards the bottom of the peripheral zone where it is diverted towards the core of the reactor;

[0036] - a tank called a secondary tank, arranged around the primary tank;

[0037]

[0038]

[0039]

[0040]

[0041]

[0042]

[0043] - a tank well, arranged around the secondary tank; - a closing slab, to enclose the first heat transfer fluid inside the primary tank; - a system for removing thermal power both during nominal operation and in nuclear reactor shutdown situations, the system comprising: • a shell arranged between the primary tank and the secondary tank, delimiting a volume with the primary tank filled with liquid metal; • a closed circuit, called a secondary circuit, filled with a second heat transfer fluid and adapted to evacuate the heat itself evacuated by conduction through the primary tank and transferred by the liquid metal, to an energy conversion system and / or a heat network; - a system for evacuating residual thermal power (EPuR) in accident situations of the nuclear reactor, the system comprising: • at least one phase change material (PCM) of solid / liquid type arranged inside the space delimited between the shell and the secondary vessel, the PCM material(s) being adapted to melt by storing by latent heat at least a part, preferably all, of the residual power emitted by the core in accident situations, for a predetermined duration. Advantageously, the inventory of the MCP material(s) allows, through the energy required for the phase change (latent heat), to absorb all the residual power of the core for a predetermined period of 3 days. There is therefore a design margin linked to the sensible heat of the molten salt in the liquid state which would allow, well before it boils, to absorb additional residual heat and therefore to leave a grace period before intervention greater than 3 days. By "sensible heat" we mean the heat that matter, in a given state, can absorb without changing phase. By "in shutdown situations" is understood here and within the scope of the invention a normal shutdown of the reactor and not in the event of an accident (accidental situations). The term "inert liquid salt" means a heat-transfer liquid salt containing neither fissile nor fertile elements and which does not react chemically with water and air. According to an advantageous embodiment, the circulation pump(s) is(are) a centrifugal pump(s) arranged vertically and mounted as a passage(s) through the cover plug of the primary tank with its(their) blades arranged above the step. Advantageously, the height of MCP material(s) between the shell and the secondary tank is greater than the height of inert liquid salt(s) between the primary tank and the shell. This ensures that the leakage of heat-transfer liquid salt is limited or prevented. by the primary tank in the event of an accident. Preferably, also, the height of the liquid metal between the primary tank and the shell is greater than the height of inert liquid salt(s) between the primary tank and the shell.

[0044] According to an advantageous construction variant, the primary and secondary tanks as well as the shell are straight cylinders arranged concentrically with each other.

[0045] Preferably, the inert liquid salt is chosen from chlorine-based salts, optionally enriched with Chlorine 37 to reduce the formation of radioactive C136, more preferably NaCl, KC1, MgCl2, CaCl2, ZnCl2 or a mixture thereof, in particular a mixture of molten salts NaCl-MgC12, NaCl-MgC12-KC1 ​​or even NaCl-MgC12-KCl-ZnCl2.

[0046] According to an advantageous embodiment variant, the closed circuit comprises a coil, the periphery of which is preferably provided with heat dissipation fins; the coil being arranged between the primary tank and the shell, in a helix around the latter. Preferably, the coil can be fixed, in particular by welding to the shell.

[0047] The liquid metal in the bath between the primary tank and the shell is made of pure aluminum.

[0048] Preferably, the MCP material(s) between the shell and the secondary tank is (are) in the form of a powder.

[0049] More preferably, the MCP material(s) between the shell and the secondary tank is / are made of pure aluminum.

[0050] More preferably, the primary tank is made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC).

[0051] More preferably, the secondary tank and the shell are each made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC), depending on the expected operating conditions.

[0052] The nuclear reactor according to the invention can be thermal, epithermal or fast neutron spectra.

[0053] Thus, according to a first alternative, the primary vessel is devoid of moderating material so that the operation of the reactor is with fast neutrons.

[0054] According to a second alternative, the reactor core houses at least one moderating material so that the operation of the reactor is with thermal neutrons or epithermal neutrons. The heat transfer fluid can advantageously play the role of moderator (chloride salt containing lithium or even fluoride salt).

[0055] In the context of the invention, the term "moderator material" means any material that can slow down neutrons. In the usual sense, the kinetic energy of a fast neutron is greater than leV, while that of a thermal neutron is less than leV, typically of the order of 0.025eV. Reference may be made to publication [8], and in particular in [Fig.4], which indicates, for several types of reactors, the thermal fraction and the fast fraction of the neutron flux.

[0056] Reference may be made to patent FR3025650B1 which describes the insertion of moderating materials into a fuel assembly of a fast neutron reactor.

[0057] According to an advantageous embodiment, the solid nuclear fuels are nuclear fuel assemblies and / or fuel particles, called TRISO, and / or fuel pellets housed individually in separate cells of a plate.

[0058] Solid nuclear fuels can be based on depleted, low-enriched uranium dioxide (UO2), preferably with an enrichment <5%, or reprocessing (URT), and / or plutonium dioxide (PuO2), or based on enriched uranium U235 (HALEU, an English acronym for “High-Assay Low-Enriched Uranium”), preferably with an enrichment between 5% and 20%.

[0059] Preferably, the nuclear reactor comprises a reactivity control system consisting either of control rods internal to the primary vessel, or of rotating drums external to the primary vessel.

[0060] The nuclear reactor which has just been described is particularly intended to have a power of between 20 and 100 MWth.

[0061] The invention therefore essentially consists of producing a nuclear reactor with solid fuel and primary heat transfer fluid with liquid metal or molten salt(s) which guarantees both: - extraction by forced convection in the primary circuit, of thermal power in normal and accidental operating modes and this, at shutdown through the primary vessel of the reactor, i.e. beyond the second containment barrier; - in the event of an incident or accident, the implementation of a passive residual heat removal system which is compact and capable of ensuring the safety function for up to a predetermined duration, typically 3 days, without any intervention by an operator, thanks to the presence of one or more MCP materials which store(s) the residual heat produced in the core and removed by the primary vessel; - a drastic reduction in the risk of gas entrainment in the core, due to the absence of an intermediate exchanger within the primary tank; - the simplification of the architecture of the primary circuit, the reactor pit and the reactor vessel without any crossing, with a fluid circuit for evacuating the thermal power, which: improves its reliability and inspectability, allows the replacement of primary components, which increases significantly the overall lifespan of the installation, • simplifies its constructability, • helps to simplify and therefore make the safety demonstration more robust, • simplifies all maintenance and handling operations, particularly fuel handling, - a second, extremely simplified barrier without any singular point, which passes exclusively through the limits of the primary tank, - due to the location outside the primary tank of the heat exchanger, constituted by the closed circuit, the possibility of carrying out maintenance and inspectability operations in a simplified manner while being beyond the second containment barrier.

[0062] The advantages of the invention are numerous, among which we can cite: - operation with active flow, i.e. by forced convection, of the heat transfer fluid of the primary circuit for reactors with a thermal power of 20 to a few tens of MWth, typically up to 100MWth, in normal / nominal operation and / or in residual power evacuation, by reducing pressure losses due to the absence of an intermediate exchanger within the primary vessel; - improved safety because the penetrations of the second containment barrier (primary vessel) are limited to the reactivity control devices, i.e. the penetrations of the control rods, and the primary circulation pumps which are penetrations in the upper part of the primary vessel, through the closure slab, and therefore above the liquid volume of the primary fluid, and to the usual instrumentation devices. In-vessel reactivity control devices, for example in the form of neutron reflectors rotating around the primary vessel, can also be considered outside the primary vessel to further reduce the slab penetrations; - a simplification of the design of the primary tank due to the absence of crossings for power evacuation, which limits singular points and increases the lifespan of the tank; - a simplification of the structure forming the redan, because, unlike state-of-the-art solutions, it no longer has to integrate intermediate exchangers, at the level of mechanical support and at the level of hydraulics; - the removal of an intermediate circuit; - a greater choice of fluid in the closed secondary circuit and therefore, of energy recovery methods, due to the evacuation of power beyond the second barrier (primary vessel). For example, it is possible to consider a pressurized fluid (gas or supercritical CO2) without the risk of having to consider, in the safety demonstration, the consequences of the entrainment of a gas bubble in the reactor core.

[0063] The preferred applications of the invention are small-sized reactors of the GenIV sector, in particular reactors cooled with sodium, lead and liquid salt.

[0064] Other advantages and characteristics of the invention will become more apparent upon reading the detailed description of examples of implementation of the invention given by way of illustration and not limitation with reference to the following figures. Brief description of the drawings

[0065] [Fig.l] [Fig.l] is a schematic cross-sectional view of a liquid salt-cooled nuclear reactor according to the invention, with a system for removing nominal and residual power in the event of a shutdown through the primary vessel and an EPuR system in an accident situation.

[0066] [Fig.2] [Fig.2] is a detail view of [Fig.l].

[0067] [Fig.3] [Fig.3] is a longitudinal sectional and perspective view of a reactor nuclear according to the invention.

[0068] [Fig.4] [Fig.4] is a detailed view of [Fig.3] according to an alternative embodiment of the secondary closed circuit of the reactor. Detailed description

[0069] Throughout the present application, the terms "vertical", "lower", "upper", "bottom", "top", "below" and "above" are to be understood by reference to a primary vessel filled with inert liquid salt of a fast neutron nuclear reactor according to the invention, as it is in the vertical operating configuration.

[0070] Figures 1 to 4 show a liquid salt-cooled fast neutron nuclear reactor 1 according to the invention.

[0071] Such a reactor 1 comprises a primary vessel 10 or reactor vessel filled with inert liquid salt, called primary salt S, and inside which is present the core 11 where a plurality of fuel assemblies, not shown, are installed, which generate thermal energy by the fissions of the fuel.

[0072] The inert liquid salt S is therefore the heat transfer fluid of the primary circuit: it stores and transports the heat of the core 11 and exchanges heat through the wall of the primary vessel 10. The primary salt has the physicochemical characteristics to ensure that it remains in the liquid state at atmospheric pressure over the temperature range in normal and accidental operation of the core 11. It is inert from a radioactive point of view, because it contains neither fertile nor fissile elements and it is also chemically inert with respect to the MCP materials and liquid metal bath detailed below, and with respect to all the reactor structures.

[0073] The primary salt S is chosen to have good thermal and physical properties to promote natural convection, thermal exchanges with the core 11 and through the wall of the primary tank 10.

[0074] The primary salt is thus advantageously chosen from NaCl, KC1, MgC12, CaC12, ZnC12 enriched with Chlorine 37 or a mixture of these, in particular a mixture of molten salts NaCl-MgC12, NaCl-MgC12-KC1 ​​or even NaCl-MgC12-KCl-ZnC12. For example, NaCl-KCl-MgC12, which has a melting point below 500°C, a good heat capacity (Cp) and a good coefficient of thermal expansion are advantageous for improving natural convection.

[0075] Solid nuclear fuels may be based on depleted uranium dioxide, low enrichment, preferably with an enrichment <5%, or reprocessing (URT) and / or plutonium dioxide (PuO2), or based on enriched uranium U235 (HALEU, an English acronym for “High-Assay Low-Enriched Uranium”, preferably enriched between 5% and 20%.

[0076] The fuel assemblies may comprise a SiC fuel cladding, in order to withstand very high temperatures. The claddings of the assemblies constitute the first containment barrier while the vessel 10 constitutes the second containment barrier for the radioactive materials contained in the core 11.

[0077] The primary tank 10 supports the weight of the liquid salt of the primary circuit as well as the internal components.

[0078] The support of the core 11 is provided by a mechanically welded structure called a base 12 in which the feet of the fuel assemblies 11 are positioned.

[0079] The core 11 is surrounded by a separation envelope 13 provided with a peripheral neutron reflector intended to ensure the maintenance of the neutron flux in the core. This separation envelope 13 of the core 11 makes it possible to separate the primary salt, in its so-called cold and hot temperatures. Thus, the primary salt at cold temperature surrounds the core 11 inside the primary vessel, while the primary salt at hot temperature, heated by circulating upwards in the core 11, is found in the upper central portion of the core.

[0080] Typically, the bed base 12 is made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC) depending on the operating conditions.

[0081] As shown, the primary tank 10 is a straight cylinder with a central axis X. Typically, the primary tank 10 is made of AISI 316L stainless steel, preferably with a very low boron content to prevent the risk of cracking at high temperatures. Its external surface is preferably made highly emissive by a pre-oxidation treatment, carried out to facilitate the radiation of heat. towards the outside during the residual power evacuation phase. The primary tank 10 can also be made of nickel-based alloy or silicon carbide (SiC) depending on the operating conditions.

[0082] The reactor vessel 10 comprises circulation pumps 100 of the liquid inert salt as primary heat transfer fluid.

[0083] The reactor vessel 10 is separated into two distinct zones by a separation structure consisting of at least one shell 14 arranged inside the reactor vessel 10. This separation device is also known as a redan.

[0084] As illustrated in [Fig.3], the step 14 of central axis X can comprise three ferrules 140, 141, 142 welded together, including: - the top ferrule 140, in the form of a straight cylinder; - the center ferrule 141, in the form of a truncated cone, - the lower ferrule 142, which surrounds the separation envelope 13, in the form of a right cylinder.

[0085] The step 14, more precisely its lower ferrule 142, can be welded to the base 12 as shown in [Fig.3] and is placed and welded against the bottom of the primary tank 10.

[0086] The step 14 further comprises through openings 143 made through the lower ferrule 142.

[0087] As symbolized by the arrows in [Fig.l], the step 14 is arranged in the primary vessel 10 by forming a central chimney, to separate the interior of the primary vessel 10 into a central zone and a peripheral zone so that during operation of the reactor, the pumps 100 circulate by forced convection the liquid metal or molten salt(s) heat transfer fluid, in a loop from the bottom of the central zone where the reactor core 11 is arranged, from which it rises by heating to the top of the central zone where it is diverted to the top of the peripheral zone to descend to the bottom of the peripheral zone where it is diverted towards the core 11.

[0088] Typically, the redan 14 is made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC), depending on the operating conditions.

[0089] Advantageously, as shown in [Fig. 3], the step 14 may comprise in its upper part a deflector 144 and a section reducer 145, each arranged around the upper shell 140 to reduce the passage section of the inert liquid salt S in the peripheral zone when it descends downwards. This makes it possible to increase the speed of the liquid salt S, in particular in this descending flow zone, which is usually referred to as "downcomer". This also makes it possible to advantageously improve the heat transfer by convection of the salt S towards the wall of the primary tank 10.

[0090] The central chimney shape of the redan 14 improves the circulation by natural convection of the liquid salt S.

[0091] A removable plug 15, called a core plug-cover, is arranged directly above the core 11 and closes the primary tank 10 to contain the liquid salt S, act as a barrier between the liquid salt S and the external environment, and also constitute with the primary tank the second containment barrier for the materials contained in the core 11.

[0092] Like the primary tank 10, the cap-cover 15 is chemically inert with respect to the MCP materials and liquid metal bath detailed below, and with respect to all of the structures of the reactor.

[0093] The cover plug 15 is provided with passages for the components of the control rods 16 and for the control and monitoring elements of the core 11 and the liquid salt inventory S, not shown.

[0094] The core cover plug 15 is therefore a plug, removable under an inert atmosphere, which carries all the handling systems as well as all the instrumentation necessary for monitoring the core including the control rods, the number of which depends on the type of core and its power, as well as the thermocouples and other monitoring devices. A system for maintaining the temperature of the plug will be provided in order to limit the risks of salt aerosol deposits.

[0095] Typically, the cover plug 15 is made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC) depending on the operating conditions. Typically also, the material used for the control rods is B4C.

[0096] In a variant, instead of the control rods, also made of B4C material, rotating drums can be arranged external to the primary tank 10, so as to advantageously eliminate the crossings of the cover 15 and therefore the risks of seizure associated with salt aerosols.

[0097] As shown in [Fig.3], the primary salt circulation pumps 100 are preferably centrifugal pumps arranged vertically and mounted as feedthroughs of the cover plug 15 of the primary tank, with their blades 101 arranged above the step 14.

[0098] The reactor vessel 10 comprises a sky, usually called a pile sky, filled with an inert gas, such as argon or helium, above the molten salt(s) fuel liquid. This sky makes it possible on the one hand to absorb the thermal expansion of the liquid within the reactor vessel, when it undergoes a variation in level and on the other hand to recover the gaseous fission products generated by the nuclear fissions within the fuels.

[0099] A support, confinement and thermal insulation assembly between the primary tank and the external environment E is arranged around the primary tank 10.

[0100] More precisely, as shown in figures 1 and 3, this assembly 2 comprises a well tank 20, inside which are inserted from the outside towards the inside a layer of thermal insulating material 21, a secondary tank 22 and the primary tank 10 of the reactor.

[0101] The tank well 30 is a block of generally cylindrical exterior shape which supports the weight of all the components within it. The functions of the tank well 30 are to provide biological protection and protection against external aggressions, and also to ensure cooling of the external environment to maintain low temperatures. Typically, the tank well 20 is a concrete block.

[0102] The layer of thermal insulating material 21 guarantees the thermal insulation of the tank well 20. Typically, the layer 21 is made of polyurethane foam or silicate-based.

[0103] The secondary tank 22 guarantees the retention of the liquid salt S in the event of a leak through the primary tank 10 and the protection of the tank well 30. Also, the secondary tank 22 contains a volume 23 of phase change material(s) (PCM) of the solid / liquid type.

[0104] The volume 23 of material(s) (MCP), preferably in the form of powder, is adapted to melt by storing by latent heat at least a part of the residual power emitted by the core in accident situations, for a predetermined duration. Thus, the physicochemical characteristics of the MCP material(s) make it possible to guarantee maintenance in the solid state at atmospheric pressure over the temperature range in normal operation of the core and a phase change when the core 11 enters an accident situation, and this, for a predetermined duration.

[0105] As shown in [Fig.2], the inventory of the total volume 23 of material(s) (MCP), makes it possible to guarantee a height H greater than that of the inert liquid salt S in the primary tank 10. Such a height contributes to maintaining all of the primary salt S inside the primary tank 10 in the event of a leak at the wall of the latter.

[0106] The MCP material(s) is(are) chemically inert with respect to the liquid metal bath and all of the reactor structures.

[0107] Typically, the MCP material is pure aluminum. Alternatively, the metal MCP may be replaced by a salt-type MCP, for example MgC12.

[0108] The secondary tank 22 rests against the tank well 20 and its upper part is welded to the closing slab 17.

[0109] Typically, the secondary tank 22 may be made of AISI 316L stainless steel or of nickel-based alloy or of silicon carbide (SiC) depending on the operating conditions.

[0110] A removable cover 24 of annular shape around the cover plug 15 closes the secondary tank 22 to contain the volume 23 of MCP material(s), ensuring the role of barrier between this volume 23 and the external environment E.

[0111] Like the secondary tank 20, the cover 24 is chemically inert with respect to MCP materials and liquid metal bath detailed below, and with respect to all reactor structures.

[0112] The cover 24 is provided with passages for the control and monitoring elements of the inventory made of MCP material(s), not shown, and meets the sealing requirements of the second containment barrier. During the handling phase, operations requiring the removal of this cover must take place under an inert atmosphere.

[0113] Typically, the cover 24 may be made of AISI 316L stainless steel or of a nickel-based alloy or of silicon carbide (SiC) depending on the operating conditions.

[0114] The nuclear reactor 1 further comprises a system 3 for removing thermal power, both in nominal operation and in situations of nuclear reactor shutdown.

[0115] This system 3 firstly comprises a cylindrical shell 30 arranged concentrically between the primary tank 10 and the secondary tank 22.

[0116] This shell 30 thus delimits a volume with the primary tank 10 filled with a bath of liquid metal 31.

[0117] Typically, the ferrule 30 may be made of AISI 316 stainless steel or nickel-based alloy or silicon carbide (SiC) depending on the operating conditions and is designed to chemically and mechanically resist the liquid metal bath 31 and the MCP volume 23.

[0118] A closed circuit 32, called the secondary circuit, is filled with a heat transfer fluid and is adapted to evacuate the heat evacuated by conduction through the primary tank 10 and transferred by the liquid metal bath 31, towards an energy conversion system and / or a heat network, not shown.

[0119] The liquid metal bath 31 thus improves the transfer of heat by conduction from the primary tank 10 to the secondary circuit 32.

[0120] Typically, the liquid metal bath is pure aluminum.

[0121] The liquid metal is chemically inert with respect to the liquid salt S, the MCP materials and with respect to all the structures of the reactor.

[0122] The closed circuit 32 is preferably constituted by a coil, arranged in a helix around the primary tank 10, preferably being welded to the internal wall of the shell 30.

[0123] The coil 32 has a diameter which is a function of the diameter of the primary tank 10 and a height sufficient to have the surface area necessary for the evacuation of the desired heat.

[0124] In other words, the total number of turns, the spacing and the diameter of these turns which make up the coil 32 depend on the diameter of the primary vessel 10 and the power of the core 11 of the nuclear reactor. For example, the pitch of the turns of the coil 32 can be equal to 10 cm, which is a good compromise for the manufacture and absorption of heat by conduction by the liquid metal bath.

[0125] For example also, the external diameter of the coil 32 is of the order of 5 to 10 cm with a pitch of turns of the order of 10 to 15 cm, in order to minimize the pressure losses, reduce the size of the pipes and maximize the surface exposed to the primary tank 10. The thickness of the coil 32 depends on the mechanical stresses exerted by the internal liquid metal and by its weight.

[0126] The material of the coil 32 must have good emissivity characteristics. Typically, the material of the coil is chosen from AISI 316L stainless steel, ferritic steels, nickel-based alloys. This material depends on the internal fluid used for the closed circuit 32.

[0127] This internal heat transfer fluid which circulates in the coil 32 is a liquid metal, chemically stable, of low viscosity, good conductor and heat transfer agent, chemically compatible with all the piping of the circuit 3 and capable of operating in natural or forced convection in a temperature range between 150-600 °C. Typically, the liquid metal of the circuit 3 can be chosen from an NaK, Pb-Bi alloy, sodium or one of the ternary alloys of liquid metals, ...

[0128] In order to improve the heat exchange between the liquid metal bath 31 and the tube forming the coil 32, the latter can be provided with heat dissipation fins 33, in particular of straight shape which extend radially to the tube, as shown in [Fig.4],

[0129] The operation of nuclear reactor 1 is now described in relation to the various normal, forced shutdown and accidental situations.

[0130] In normal operation of the reactor, all the heat generated by the fission reactions of the core 11 is evacuated by heat exchange during the passage of the salt S in the liquid state through the core, which is in forced convection thanks to the pumps 100. The liquid salt S exchanges this heat through the wall of the primary vessel 10 during its descent between the redan 14 and the primary vessel 10. This heat is then transferred by the liquid metal bath 31 to be evacuated by the secondary circuit 32 to a heat network and / or an energy conversion system. During this normal operation, the volume 23 of MCP remains in the solid state.

[0131] In the event of a shutdown caused by an operator, the same heat exchanges occur. Thus, the residual heat from the core 11 at the shutdown is also removed by heat exchange during the passage of the inert salt S in the liquid state through the core 11. Then, the salt S exchanges this heat through the wall of the primary vessel 10 during its descent between the step 14 and the primary vessel 10. This heat is then transferred by the liquid metal bath 31 to be removed by the secondary circuit 32 to a heat network and / or an energy conversion system. During this shutdown caused, volume 23 of MCP remains in solid state.

[0132] In accidental operation, in particular in the event of a Generalized Voltage Loss (GVL), which corresponds to a total loss of electrical power, such as the scenario of the Fukushima accident, the residual heat of the core 11 at shutdown is evacuated by heat exchange during the passage of the salt S in the liquid state through the core. Then, the salt S exchanges this heat through the wall of the primary vessel 10 during its descent between the step 14 and the primary vessel 10. This heat is absorbed by the initially solid MCP, which undergoes a transition from the solid state to the liquid state. The inventory of the volume 23 of MCP makes it possible, by the energy required for the phase change (latent heat), to absorb all the residual power of the core for a predetermined duration, typically 3 days.

[0133] The inventors carried out dimensioning studies to demonstrate the feasibility of a nuclear reactor 1 as just described and to propose orders of magnitude for its characteristic elements.

[0134] The studies carried out make it possible to cover the range from 20 to 100 MWth with normal and accidental operation, with circulation of the primary salt S exclusively by forced convection using the pumps 100.

[0135] The cold temperature of the inert liquid salt S, as primary fluid, is set at 600°C.

[0136] The hydraulic path of salt S is determined by the analytical study presented in the tables below.

[0137] We define the thermal power range in which the nuclear reactor 1 is intended to operate only in forced convection by the pumps 100 but without exchangers within the primary vessel.

[0138] The relevant characteristics are summarized in Table 1 below.

[0139] [Tables 1] Characteristics / input data Unit Value Nominal thermal power of the reactor 1 MWth between 20 and 100 Height of the primary vessel 10 m 20 Diameter of the primary vessel 10 m 1.6 Width of the annular space between redan 14 and primary vessel 10 m 0.063 Core inlet temperature 11 °C 600 Equivalent diameter of the core 11 at the baffle 144 m 1 m Type of primary salt S - NaCl-KCl-MgC12 in respective proportions 30-20-50

[0140] It is specified here that the characteristics of the salt considered were those indicated in the publication

[11] .

[0141] These input data, for a variable power between 20 and 100 MWth, were used by the inventors to carry out preliminary calculations using thermal calculation software, such as the COPERNIC software: [9],

[10] .

[0142] These sizing calculations were made in two sequential steps, as follows.

[0143] Step 1 / : a core configuration that is as less resistive as possible from a thermo-hydraulic point of view is chosen. An initial core dimensioning is necessary to calculate the total pressure losses of the primary hydraulic circuit.

[0144] Step 2 / : a flow of the primary salt S solely by forced convection within the primary tank 10 and a heat transfer which takes place solely by forced convection through the wall of the primary tank 10.

[0145] In step 1 / , a pre-design of a core with needle nuclear fuel assemblies was determined for a power range of 20 to 100 MWth. In this calculation, the hydraulic surface was maximized by imposing the physical quantities indicated in Table 2.

[0146] [Tables2] Physical quantities of the core 11 Unit Value Distance between fuel pins mm 3.08 Core diameter 11 m 1 Linear thermal power W / cm -100 Equivalent height of a pin m 2 External diameter of a pin mm 6.55 Thickness of a fuel cladding mm 0.45

[0147] Taking into account these physical quantities in Table 2, the hydraulic surface area for the passage of salt S is maximized by values ​​of the fissile height of the fuel and a number of fuel pins in the core. Having a maximized hydraulic surface area improves the flow of the fluid because the higher the surface area at the imposed flow rate, the more the pressure losses in the core are reduced.

[0148] The pre-dimensioning calculations for core 11 are summarized in Table 3, for respective thermal powers of 30, 40 and 100 MWth.

[0149] It is specified that the fuel assemblies considered are hexagonal tubes.

[0150] [Tables3] Core quantities 11 Unit Value Nominal thermal power MWth 30 40 100 Linear thermal power W / cm 102.02 100 163.38 Core height 11 mm 400 522.81 800 Equivalent height of a needle m 2 2 2 External diameter of a needle mm 6.55 6.55 6.55 Central hole diameter of a needle mm 0 0 0 Distance between fuel needles mm 3.08 3.08 3.08 Heating of the salt S within an assembly °C 54.49 63.95 114.8 Number of needles - 7651 7651 7651 Internal diameter of a cladding mm 5.65 5.65 5.65 External diameter of a fuel pellet mm 5.42 5.42 5.42 Temperature at the center of a fuel pellet °C 1814.05 1771.31 2293.05 Salt velocity S m / s 0.66 0.72 1.02 Pressure drop within a bundle bars 0.06 0.07 0.12 Surface area occupied by the core 11 m2 0.75 0.78 0.78 Equivalent diameter m 1 1 1 Hydraulic surface area of ​​the core 11 m2 0.489 0.509 0.509 Height / diameter ratio of the core 11 - 0.409 0.524 0.802 Fissile volume of the core 11 m3 0.068 0.092 0.141

[0151] The results in Table 3 provide the input data for the primary circuit sizing calculation. The latter determines the temperature difference at the internal wall of the primary tank 10 necessary to evacuate the heat transported by the primary salt S.

[0152] Table 4 below shows the results of reactor operation calculations based on the previous calculations. It is specified that the speed of the salt in the annular space was multiplied by a factor of 5 compared to that derived from circulation exclusively by natural convection.

[0153] [Tables4] Operating conditions Unit Value Nominal thermal power MWth 30 40 100 Flow section m2 0.51 0.51 0.51 AT between the outlet and the core inlet °C 54.50 63.95 114.80 Core inlet temperature 11 °C 600 600 600 Core outlet temperature 11 °c 654.50 663.95 714.80 Total core height 11 m 2 2 2 Pressure variation in the core 11 bar 0.022 0.025 0.045 Core diameter 11 with step 14 m 1.47 1.47 1.47 Boiler diameter m 1.6 1.6 1.6 Annular space thickness (between step 14 and primary tank 10) m 0.063 0.063 0.063 Hydraulic surface area of ​​the annular space m2 0.30 0.30 0.30 Hydraulic diameter of the annular space m 0.19 0.19 0.191 Height of the annular space m 20 20 20 Salt velocity S in the annular space m / s 4.95 5.63 7.89 AT salt S / internal wall of the primary tank 10 °C 56.73 67.62 126.30 Pressure drop in the annular space bar 0.30 0.38 0.68 Total pressure drop bar 0.32 0.40 0.72 Average temperature of the annular space °C 627.25 631.97 657.40

[0154] From the results in Table 4, it can be seen that an increase in the salt velocity contributes to a clear improvement in the heat transfer coefficient. between salt S and tank wall.

[0155] Compared to a natural convection architecture, a forced convection architecture according to the invention has the following advantages: - a reduction in the temperature difference between salt S and the wall, which implies an increase in the temperature of the fluid leaving the exchanger in the form of a coil 32, to approximately 600°C, and therefore in the electrical efficiency of the electrical conversion system connected to the exchanger; - a potential reduction in the height of the primary tank to 15m or 10m, from a height of 20m fixed in the calculations to promote exclusively natural circulation of salt S.

[0156] The invention is not limited to the examples which have just been described; it is possible in particular to combine characteristics of the illustrated examples within non-illustrated variants.

[0157] Other variants and embodiments may be envisaged without departing from the scope of the invention. List of cited references

[0158] [1]: H. OHSHIMA et al., “Handbook of Generation IVNuclear Reactors”, chapter 5, Elsevier, 2016.

[0159] [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.

[0160] [3]: MA 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. U. Sverdrup, «Assessing the Past and Future Sustainability of Global hélium,» Biophysical Economies and Sustainability (2020), 2020.

[0161] [6]: L. LIN et al., «Feasïbility of an innovative long-life molten chloride-cooled reactor», Nuclear Science and Techniques, vol. 33, pp. 1-15, 2020.

[0162] [7]: http: / / sme.vimaru.edu.vn / sites / sme.vimaru.edu.vn / files / volume_2_-_properties_and_se lection_nonf.pdf

[0163] [8]: Jiri Krepel et al. "Selj-Sustaining Breeding in Advanced Reactors: Characte- rization ofSelected Reactors”, Encyclopedia of Nuclear Energy 2021, Pages 801-819. https: / / www.sciencedirect.com / science / article / pii / B97801281972570012397via%3Dih ub

[0164] [9] : F. MORIN et al., "COPERNIC, A NEW TOOL BASED ON SIMPLIFIED CAL CULATION METHODS FOR INNOVATIVE LWRs CONCEPTUAL DESIGN STUDIES”, ICAPP 2017 Conférence, 2017.

[0165]

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

[0166]

[11] : Y.LI et al., "Survey and évaluation of équations for thermophysical properties of binary / temary 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.

Claims

1. Claims Nuclear reactor (1) cooled by liquid metal or inert liquid salt(s), comprising: - a tank (10), called primary tank, axisymmetric around a central axis (X), filled with a first heat transfer fluid with at least one liquid metal or at least one inert liquid salt as heat transfer fluid for the primary circuit of the reactor, comprising: • a core (11) consisting of assemblies containing combustible nuclear material in the solid state, confined, confined in at least one envelope; • at least one circulation pump (100) for the first heat transfer fluid; • a structure forming a step (14), with a central axis coinciding with that of the primary vessel, the structure being arranged in the primary vessel to separate the interior thereof into a central zone and a peripheral zone so that when the reactor is operating, the pump(s) circulate(s) the liquid metal or inert liquid salt(s) coolant by forced convection, in a loop from the bottom of the central zone where the reactor core (C) is arranged within which the fission reactions occur, from which it rises by heating, to the top of the central zone where it is diverted towards the top of the peripheral zone to descend towards the bottom of the peripheral zone where it is diverted towards the core of the reactor; - a tank (22) called the secondary tank, arranged around the primary tank; - a tank well (20), arranged around the secondary tank (22); - a closing cap-cover (15), for enclosing the first heat transfer fluid inside the primary tank; - a system (3) for evacuating thermal power both in nominal operation and in situations of nuclear reactor shutdown, the system comprising: a ferrule (30) arranged between the primary tank and the se- tank secondary, delimiting a volume with the primary vessel filled with liquid metal (31); • a closed circuit (32), called secondary circuit, filled with a second heat transfer fluid and adapted to evacuate the heat itself evacuated by conduction through the primary vessel and transferred by the liquid metal, to an energy conversion system and / or a heat network; - a system for evacuating residual thermal power (EPuR) in accident situations of the nuclear reactor, the system comprising: • at least one phase change material (PCM) of solid / liquid type (23) arranged inside the space delimited between the shell and the secondary vessel, the PCM material(s) being adapted to melt by storing by latent heat at least a part, preferably all, of the residual power emitted by the core in accident situations, for a predetermined duration.

2. Nuclear reactor (1) according to claim 1, the circulation pump(s) being centrifugal pumps (100) arranged vertically and mounted as feedthroughs of the cover plug (15) of the primary vessel with their blades (101) arranged above the step.

3. Nuclear reactor (1) according to claim 1 or 2, the height of MCP material(s) between the shell and the secondary vessel being greater than the height of inert liquid salt(s) between the primary vessel and the shell.

4. Nuclear reactor (1) according to one of the preceding claims, the primary and secondary vessels as well as the shell being straight cylinders arranged concentrically with each other.

5. Nuclear reactor (1) according to one of the preceding claims, the inert liquid salt being chosen from chlorine-based salts, optionally enriched with Chlorine 37, preferably NaCl, KC1, MgCl2, CaCl2, ZnCl2 or a mixture thereof, in particular a mixture of molten salts NaCl-MgCl2, NaCl-MgCl2-KC1 or NaCl-MgCl2-KCl-ZnCl2.

6. Nuclear reactor (1) according to one of the preceding claims, the closed circuit comprising a coil (32), the periphery of which is preferably provided with heat dissipation fins; the coil being arranged between the primary tank and the shell, in a helix around the latter.

7. Nuclear reactor (1) according to one of the preceding claims, the liquid metal of the bath between the primary vessel and the shell being made of pure aluminum.

8. Nuclear reactor (1) according to one of the preceding claims, the MCP material(s) between the shell and the secondary vessel being in the form of a powder.

9. Nuclear reactor (1) according to one of the preceding claims, the MCP material(s) between the shell and the secondary vessel being made of pure aluminum.

10. Nuclear reactor (1) according to one of the preceding claims, the primary vessel being made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC).

11. Nuclear reactor (1) according to one of the preceding claims, the secondary vessel and the shell each being made of AISI 316L stainless steel or nickel-based alloy or silicon carbide (SiC).

12. Nuclear reactor (1) according to one of the preceding claims, the primary vessel being devoid of moderating material so that the operation of the reactor is with fast neutrons.

13. Nuclear reactor (1) according to one of claims 1 to 10, the reactor core housing at least one moderating material so that the operation of the reactor is with thermal neutrons or with epithermal neutrons.

14. Nuclear reactor (1) according to one of the preceding claims, the solid nuclear fuels being nuclear fuel assemblies and / or fuel particles, called TRISO, and / or fuel pellets housed individually in separate cells of a plate.

15. Nuclear reactor (1) according to one of the preceding claims, the solid nuclear fuels being based on depleted, low-enriched uranium dioxide (UO2), preferably with an enrichment <5%, or reprocessing (URT) and / or plutonium dioxide (PuO2), or based on enriched uranium U235 (HALEU, an English acronym for “High-Assay Low-Enriched Uranium”, preferably with an enrichment between 5% and 20%.

16. Nuclear reactor (1) according to one of the preceding claims, comprising a reactivity control system constituted either by control rods internal to the primary vessel, or by rotating drums external to the primary vessel.

17. Nuclear reactor (1) according to one of the preceding claims, the power of which is between 20 and 100 MWth.

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