Modular pressurized water reactor (PWR) type (SMR) with pressurizer without water spray.

Abstract

Modular pressurized water reactor (PWR) of the SMR type with a pressurizer without water spray. The invention relates to a pressurized water reactor (4) of the modular reactor (SMR) type comprising: - a reactor vessel with a central axis (X) including a cover (41) in the form of a dome (6) and housing at least part of the primary circuit; - a pressurizer of the primary circuit whose steam cooling and condensation part comprises: • two walls (60, 61) of the dome, separated from each other to form a space (E) within which water in liquid form can circulate from the bottom to the top of the dome forming a central discharge stack (52), so as to condense the steam in the saturation state of the primary circuit inside the vessel and thus reduce the pressure within the vessel; • a regulating valve (6) for the flow of liquid water circulating in the space, arranged in the central exhaust stack.Figure for the abridged version: fig. 11.

Description

Title of the invention: Modular type pressurized water reactor (PWR) with pressurizer without water spray. technical field

[0001] The present invention relates to pressurized water nuclear reactors, in particular those of the integrated modular reactor type, known as small or medium power or SMR in English (acronym for "Small Modular Reactor").

[0002] The invention aims to overcome a major drawback of manufacturing complexity, associated with a significant cost, and complexity of piloting and operating such an SMR reactor, in particular its pressurizer.

[0003] By "SMR reactor", we mean here and within the framework of the invention, the usual technological meaning, namely a nuclear fission reactor, of smaller size and power than conventional REL reactors, of which a block is manufactured in a factory and transported to a nuclear site for installation.

[0004] By “reactor block”, we mean here and within the framework of the invention, the vessel, called reactor vessel as well as all the components and part of the fluidic circuit, in particular the reactor core which creates heat by nuclear fission reactions, which is housed inside the reactor vessel.

[0005] For the purposes of this invention, "heat-producing reactor" refers to a nuclear installation, nuclear power plant, or nuclear reactor whose power output is primarily dedicated to heat production. A heat-producing reactor may be 100% dedicated to heat production. However, a small portion of its power output may also be used to generate electricity.

[0006] For the purposes of this invention, "power-generating" refers to a nuclear installation, a nuclear power plant, or a nuclear reactor whose power output is primarily dedicated to electricity generation. The power output of a power-generating reactor may be 100% dedicated to electricity generation. However, a small portion of its power output may also be used to generate heat. Previous technique

[0007] A significant portion of the current fleet of pressurized water reactor (PWR) nuclear power plants is nearing the end of its operating period for which the reactors were designed and licensed, in a context where the energy transition with the decarbonization of uses will increase the need for electricity (non-intermittent, high availability and competitive electricity).

[0008] A pressurized water reactor (PWR) comprises three cycles (fluidic circuits) whose general principle of normal operation is as follows.

[0009] The high-pressure water of a primary circuit extracts the energy supplied, in the form of heat, by the fission of uranium nuclei, and where applicable plutonium nuclei, in the reactor core.

[0010] Next, this water under high pressure and high temperature, typically 155 bar and 300 °C, circulates through a steam generator (SG) where it exchanges its heat with a secondary circuit, which also uses pressurized water as a heat transfer fluid. This water, in the form of steam at high pressure, typically around 70 bar, is then expanded via an expansion device that transforms the change in the fluid's enthalpy into mechanical work and then electrical work in the presence of an electric generator.

[0011] The water from the secondary circuit is then condensed via a condenser using a third cycle, the cooling cycle, as the cold source.

[0012] The design principles of PWR reactors according to these three cycles have been substantially the same since the beginning of the commissioning of the first ones operated.

[0013] The main elements of a PWR primary circuit are shown in [Fig. 1]:

[0014] - a reactor building 1 performing various functions, including in particular a contribution to the containment safety function,

[0015] - a reactor vessel 20, located in the center of building 1, housing the core C of the reactor,

[0016] - a primary circuit 2 in pressurized water comprising the tank 20.

[0017] These main elements are therefore common, their constitution and the number of components varying according to the power of the reactor.

[0018] Typically, the building envelope of reactor 1 can consist of several thicknesses. For example, as illustrated in [Fig. 1], a building of reactor 1 can consist of an outer wall of reinforced concrete 12, an inner wall of prestressed concrete 10 separated from the outer wall 12 by an annular space 13 devoid of material, and a metallic skin 11 on the inside of the prestressed concrete wall 10, for a 1650 MWe reactor.

[0019] As illustrated in [Fig.2], from publication [1], the primary circuit 2 consists of the following main components:

[0020] - a reactor vessel 20,

[0021] - primary loops 21 each comprising a primary pump 22 and a steam generator 23,

[0022] - a single pressurizer 24.

[0023] In addition, the reactor core control rod mechanisms and control rod clusters 25 are distinguished on this [Fig.2].

[0024] Depending on the power of the reactor, the number of loops can be three for a 900MWe reactor or 4 for a 1300MWe reactor and above.

[0025] The reactor building 1 is therefore sized, among other things, to house all the components of the primary circuit 2.

[0026] Figure 3 illustrates the energy transfer cycle (heat then electricity) of a PWR reactor. In Figure 3, the distribution of the positioning of the components relative to the reactor building 1, which provides the function of a third containment barrier, can be seen in particular.

[0027] The fluid connections between the inside and outside of the reactor building 1 are provided by lines 30, 31 of the external circuit of the steam generators 23 to the secondary circuit 3 comprising a turbine 32 connected to the electric generator 33, a condenser 34, a feed pump 35 and a heater not shown.

[0028] More specifically, for a given steam generator 23, the reactor building 1 is crossed by a line called the hot line 30 which evacuates the steam from the steam generator 23 to extract the power and brings it to the turbine 32, and by a line called the cold line 31 which supplies liquid water to the steam generator 23.

[0029] A currently emerging technology is that of small modular reactors (SMRs). The main advantages of these SMRs compared to existing PWRs are the simplification of systems, primarily for safety purposes, and increased modularity through the extensive manufacturing of components in factories for transport to the construction site.

[0030] In addition, SMRs are flexible due to their low power level and their ability to be integrated into the territory.

[0031] They thus appear as a competitive solution for the future. To date, approximately 70 SMR projects have been identified worldwide at various stages of development, a quarter of which use mature, generation 3 (Gen-III) technologies, such as those used in the French fleet.

[0032] Among the SMRs currently under development, some offer a configuration based on the integration of the steam generator, or even all the components of the primary circuit, including the pressurizer and primary pumps, inside the reactor vessel. These SMRs are called integrated SMRs.

[0033] In addition to the gain in compactness, the integrated SMRs have the advantage of no longer requiring overhead pressurized water fluidic lines, with the exception of the control circuit lines of very limited diameter, typically a few cm, which considerably reduces the risks of accident and associated consequences related to the rupture of the primary circuit lines.

[0034] By way of example, the nuclear power plant project with the acronym NUWARD™, is a power plant consisting of two integrated SMRs, with a unit power of 170MWe, with all the primary circuit components inside the reactor vessel.

[0035] Other integrated SMR projects are under development or have been studied, among which we can mention the SCOR project with a power of 150 to 200 MWe on behalf of the Applicant or the ACP100 project with a power of 100 MWe.

[0036] The gain in compactness of integrated type SMRs complicates operations in operation, compared to those carried out in a conventional PWR.

[0037] Indeed, the main operability and maintainability structural operations on the architecture for a reactor primary circuit are as follows:

[0038] - fuel loading / unloading operations which require, in appropriate radiation protection conditions, and access to the inside of the reactor vessel,

[0039] - maintenance operations on equipment that require accessibility to the equipment.

[0040] If we refer to [Fig.2], we see that the loops of a primary circuit 2 of a conventional PWR are designed to allow maintenance on each component without impacting, or in a very limited way, the other components and that the fuel handling operations are carried out by opening the lid of the tank 20 without impacting the primary loops 21.

[0041] Conversely, due to the integration of components in an integrated SMR, access to the fuel area for loading / unloading operations may require removing functional parts of the primary circuit, which is more significant than handling the tank cover.

[0042] Figure 4 shows an example of an integrated SMR currently under development. Such an integrated SMR reactor, generally designated by reference numeral 4, comprises a fixed compartment 40 and a removable compartment 41 in the form of a lid, for fuel handling or maintenance of the reactor internals.

[0043] The inventors have analyzed that integrated SMR reactors as currently envisaged have several disadvantages, particularly related to the complexity of their pressurizer.

[0044] Indeed, just as with a conventional PWR reactor, the pressurizer of an integrated SMR reactor includes an electric heating part which produces steam and thereby increases the pressure in the pressurizer, and a steam cooling and condensation part which causes a drop in pressure.

[0045] To date, this part of cooling and condensing the steam consists of a liquid water spraying / injection device by pumping, from a sample of the primary circuit.

[0046] In addition to the manufacturing, control and operating complexity inherent in such a spraying device within an SMR reactor, the inventors have chosen not to arrange a primary pumping group in the reactor vessel, as proposed in the patent application filed today and entitled "Nuclear installation comprising at least one modular nuclear reactor (SMR) and a vessel well delimiting a water basin in which the SMR reactor and the exchangers between primary and secondary circuits are immersed".

[0047] There is therefore a need to find a solution which makes it possible to overcome the aforementioned disadvantages related to a liquid water spraying / injection device as part of steam cooling and condensation in pressurized water reactors (PWRs) of the SMR type.

[0048] The aim of the invention is to meet at least part of this need. Description of the invention

[0049] To this end, the invention relates, in one of its aspects, to a pressurized water reactor (PWR), of the modular reactor (SMR) type, comprising:

[0050] - a central axis (X) reactor vessel comprising a lid in the form of a dome and housing at least part of the primary circuit;

[0051] - a primary circuit pressurizer, the cooling and Vapor condensation includes:

[0052] • two walls of the dome, separated from each other, forming a space (E) at inside which water can circulate in liquid form from the bottom to the top of the dome forming a central exhaust chimney, so as to condense the steam in the saturation state of the primary circuit inside the tank and thus reduce the pressure within the tank;

[0053] • a valve for regulating the flow of liquid water circulating in space, arranged in the central exhaust chimney.

[0054] The term "primary circuit" is understood in its usual sense, namely the fluid circuit that removes the heat generated in the reactor core by means of a pressurized water circulation, known as primary coolant. The reactor vessel, which houses part of the primary circuit, is the second of the three containment / safety barriers that prevent the dispersion of radioactive products.

[0055] By "secondary circuit" is meant either the usual meaning of a closed loop circuit, or an open medium comprising a secondary water basin, contained in the space of the reactor vessel well forming the third containment barrier.

[0056] Preferably, the two walls of the dome are metallic, preferably stainless steel.

[0057] According to an advantageous embodiment, the inner wall forms the pressure-resistant enclosure of the primary circuit. Preferably, the thickness of the inner wall is between 10 and 20 mm.

[0058] According to an advantageous embodiment, the reactor includes a passive heat sink arranged inside the dome to cool the steam from the primary circuit.

[0059] According to this mode, and an advantageous embodiment, the heat sink comprises a plurality of cooling fins arranged inside the inner wall, preferably distributed uniformly over the surface of the latter.

[0060] Advantageously, the fins are welded or brazed to the inner wall.

[0061] Preferably, the space between the inner and outer wall is between 0.5 and 5 cm.

[0062] According to another advantageous embodiment, the external wall of the dome is covered with a cap housing within it a thermal insulator.

[0063] Preferably, the pressurizer liquid water intended to circulate between the two walls of the dome is at least 10°C lower than the primary circuit liquid water temperature.

[0064] According to another advantageous embodiment, the heating part of the pressurizer comprises a plurality of electrical resistors wrapped in electrical insulation, arranged inside the dome.

[0065] According to an advantageous configuration, the liquid water from the pressurizer intended to circulate between the two walls of the dome is the water from the secondary circuit of the reactor.

[0066] According to an advantageous embodiment, the secondary circuit includes a water-filled basin contained in the reactor vessel well space and supplying the bottom of the space between the dome walls.

[0067] According to this mode, and an advantageous embodiment variant, the water basin is configured so that, when the SMR reactor is in normal operation, it achieves a vertical thermal stratification resulting in the formation of a thermocline delimited between the bottom of the basin at a so-called cold temperature in which the SMR reactor vessel is immersed and the top of the basin at a so-called hot temperature, the thermocline being positioned above the dome so that the circulation of liquid water from the basin into the space between the walls of the pressurizer is achieved by natural convection.

[0068] The invention further relates to a nuclear installation comprising at least one SMR nuclear reactor as described above.

[0069] Thus, the invention essentially consists of a pressurizer whose vapor cooling and condensation part is integrated directly into the lid in the form of a double-walled dome inside which liquid water can preferably circulate by natural convection from the bottom to the top of the dome through a central chimney housing a valve regulating the flow of circulating water and thus the cooling of the primary water vapor inside the dome.

[0070] In conclusion, the nuclear reactor with a pressurizer integrated within the reactor vessel head according to the invention offers numerous advantages, including:

[0071] the removal of water spraying / injection devices as part of steam cooling and condensation for an integrated SMR reactor pressurizer;

[0072] improving safety through efficient cooling solely by natural convection;

[0073] the strong simplification of the design of the SMR reactor with a calogen function;

[0074] the significant reduction in associated costs, whether for capital expenditure (CAPEX) or for operating expenditure (OPEX).

[0075] Other advantages and features of the invention will become clearer from 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

[0076] [Fig. 1] [Fig. 1] is a schematic perspective and partial cross-sectional view of an existing PWR type nuclear reactor.

[0077] [Fig.2] [Fig.2] is a schematic view of a nuclear reactor primary circuit of the REP type according to the state of the art in a configuration with three primary loops.

[0078] [Fig.3] [Fig.3] is a schematic view of the three cycles of a nuclear reactor REP type according to the state of the art.

[0079] [Fig.4] [Fig.4] is a schematic perspective view of an SMR type integrated as it is currently envisaged.

[0080] [Fig. 5] [Fig. 5] is a schematic perspective view of an SMR reactor calogen which incorporates a pressurizer according to the invention.

[0081] [Fig.5A] [Fig.5A] is an exploded perspective view of the parts of the SMR calogen reactor vessel according to the invention.

[0082] [Fig.6] [Fig.6] is a longitudinal cross-sectional view of the reactor according to [Fig.5], and which illustrates the natural convection circulation of water in the primary and secondary circuits.

[0083] [Fig.ôA] [Fig.ôA] is a schematic detail view of a control cluster assembly, the control rod, and the rod control mechanism of a calogenous SMR reactor according to the invention.

[0084] [Fig.7] [Fig.8] Figures 7 and 8 are detailed perspective views showing the primary circuit flow control valve respectively in an intermediate position and the fully open position.

[0085] [Fig.9] [Fig.9] is a perspective view of part of the reactor according to [Fig.5], and which shows in detail the placement of a water flow control valve for the secondary circuit.

[0086] [Fig.8] [Fig.1OA] [Fig.1OB] [Fig.1OC] Figures 10A, 10B and 10C are perspective views showing an example of a secondary circuit water flow control valve and its integration into an exchanger outlet manifold, the valve being respectively in the fully open position, an intermediate position and the fully closed position.

[0087] [Fig. 11] [Fig. 11] is a partial longitudinal cross-sectional view of the upper part of an SMR reactor according to Figures 5 and 6 showing a pressurizer according to the invention. Detailed description

[0088] Throughout this application, the terms "vertical", "lower", "upper", "bottom", "top", "below" and "above" are to be understood by reference to an SMR nuclear reactor, as provided for in a vertical operating configuration and arranged in a water basin according to the invention.

[0089] Figures 1 to 4 have already been detailed in the preamble, so they will not be commented on below.

[0090] For the sake of clarity, the same element according to the invention and according to the prior art is designated by the same numerical reference in all of Figures 1 to 11.

[0091] With reference to [Fig.5], a pressurized water type nuclear reactor 4 is described, according to an integrated SMR type primary circuit configuration, according to the invention.

[0092] This reactor 4 has a unit power of 20 MW thermal, with a heat-generating purpose, i.e., dedicated to supplying hot water at 90°C. Its unit power can, however, vary up or down, in a range of approximately 10 MWth to 100 MWth, and the hot water supply temperature can also vary up to approximately 150°C.

[0093] The reactor 4 with central axis X [Fig. 5] comprises a block delimited by a reactor vessel 40, an intermediate body 45, and a metallic dome 41, preferably made of stainless steel, with a thickness of approximately 10 to 20 mm, and formed of a hemispherical bottom and a vertical cylinder. This reactor vessel consists of a fixed compartment 40 and a removable compartment 45 and 41 as shown [Fig. 5A], located above the reactor core for fuel handling or maintenance of the reactor internals. The compartment formed by the The body 41 and 45 is removable to allow handling of the assemblies contained in the vessel body 40. The removable compartment 41 is a cover in the form of a dome 6 whose central chimney integrates a valve 64, adapted for cooling the reactor pressurizer as detailed later.

[0094] The reactor core C comprises a set of fuel assemblies such as those conventionally used in PWR-type reactors but with a fissile height adapted to obtain the desired total thermal power. Each fuel assembly has several missing fuel rod locations, replaced by absorber rods that can move up or down in the assembly to control the reaction and forming the control rods 42. Data from preliminary studies carried out by the Applicant consider a number of 52 assemblies and a cycle time of 10 years, with a fissile height of 1.5 m.

[0095] The vessel body 40 houses in its lower part a cylinder 43, supporting an assembly basket usually designated as the "core support basket", dedicated to holding the fuel assemblies, and a separation envelope 40 with its peripheral neutron reflector 440 intended to ensure the maintenance of the neutron flux in the core.

[0096] A set of flanges is bolted between the fixed compartment 40 and the removable compartment 45, and the dome 4L. The seal between the flanges of compartments 40 and 45 on the one hand, and 45 and 41 on the other, is advantageously ensured by a metal gasket. Removing the bolted flange located between compartments 40 and 45 allows for the complete handling of the fuel assemblies during core refueling phases. The block formed by the entire assembly of compartments 45 and 41, with the heat exchangers 49 attached, is completely removed to provide direct access to the reactor core during the handling phase. Removing the bolted flange connecting compartments 45 and 41 allows access to the upper internals of the core, control rod mechanisms, and flow control ring. The pressurizer associated with compartment 41 and its internal components can also thus be separated from the rest of the reactor vessel block for intervention and maintenance.

[0097] The studies carried out by the Applicant provide for scheduled fuel refueling stops during ten-year visits, without intervention on the core between these periods.

[0098] Above the core C, control rods 46 of the control rod assembly allow the insertion of nuclear reactivity control rods 42, in a manner similar to that commonly found in conventional PWR reactors. The control rods 42 are rods made of neutron-absorbing material.

[0099] The free volume above the reactor core C allows the control rods 42 to be fully extended, as well as the standby position of the so-called emergency fuel rods, dedicated to the safe shutdown of the nuclear reaction. The control rods 46 are individually steered vertically by means of the rod control mechanisms 47. Above the control mechanisms 47, a plate 48 with holes 480 is fixed, allowing the passage of the hot primary fluid exiting the core into the central part called the "riser". Peripheral holes also allow the passage of the control rods of the throttling valve 481.

[0100] A flow control valve 481 for the primary circuit water is arranged around the periphery of the plate 48. This valve is in the form of a flow control ring 481 that follows the inner periphery of the compartment 42 of the tank and extends over a height sufficient to cover the primary water outlet openings, thus allowing the flow rate of this primary water to be regulated.

[0101] This throttling valve 481 is designed to regulate the natural flow rate of water in the primary circuit passing through the openings 400 that constitute the inlets of the primary water manifolds of the heat exchangers 49 between the primary and secondary circuits. The positioning of this regulating valve is controlled by a motor or rod control mechanism, with the control rod 482 linked to the ring 481, which is advantageously similar to those used for the control rods of the reactivity control rods 42.

[0102] In an intermediate position, as illustrated in [Fig.7], the rolling ring 481 leaves the openings 400 partially unobstructed, which determines the flow rate of primary water that passes through them towards the exchangers 49.

[0103] In the event of a power failure or emergency shutdown, the gravitational drop of the control rods 46 also triggers the gravitational drop of the primary fluid control valve. In the lowest position, as illustrated in [Fig. 8], this control valve 481 allows all water from the primary circuit to pass through the openings 400, thus maximizing the flow rate in the heat exchangers 49 to remove residual heat and cool the primary circuit. This gravitational drop operation of the control valve ensures reliability and safety in the event of a power loss or reactor failure.

[0104] In reactor 4 of Figures 5 and 6, during normal operation, the thermal power generated by the nuclear chain reaction within the reactor core is dissipated by the primary circuit fluid, which rises by natural convection to the upper part, where it can then flow through the various outlet openings 400 corresponding to the inlet manifolds of the heat exchangers 49 between the primary and secondary circuits, and into an upper central portion of the core, in the form of a chimney. This central chimney is not detailed, called riser, contains in addition to the control rod piloting mechanisms, the sensors for instrumenting core parameters.

[0105] Thus, the separation jacket 44 of the core C allows the water, the fluid of the primary circuit, to be separated into its so-called cold and hot temperatures. Thus, the primary water at a cold temperature surrounds the core C inside the jacket 44, while the primary water at a hot temperature, heated by circulating upwards in the core C, is found in the upper central portion of the core.

[0106] Above the outlet openings 400, within the reactor 4, a separation plate 7 separates the interior of the dome 41 from the vessel containing a pressurizer, and from the riser. This separation plate 7 is a plate with through holes, providing thermal insulation and pressure differentials for the integrated pressurizer. This separation plate may be of the type already described in patent application WO2012 / 158929 A3.

[0107] The upper part incorporating the reactor pressurizer will be detailed later with reference to [Fig.11].

[0108] After cooling through the heat exchangers 49 in a downward direction, the primary circuit water passes through the openings 401, which constitute the primary water outlet manifolds of the heat exchangers 49, and then returns in a closed loop to the lower part of the reactor core for a further heating phase. The closed-loop circulation P, driven solely by natural convection of the primary water, is symbolized by the white arrows in [Fig. 6]. The driving force of the primary circuit by natural convection is controlled by the difference in height between the position of the heat exchangers 49 and the average height of the fissile zone of the core, defined by the assemblies 42.

[0109] As already mentioned, the primary pressure drop, and therefore the flow rate, is regulated by the throttling valve 481, whose control mechanisms are housed in one of the holes 480 in the plate 48. The inlet and outlet temperatures of the primary water are regulated by the neutron flux conditions, i.e., the thermal power of the core, by the positions of the reactivity control rods 42 in the core, and by the saturation temperature and pressure conditions in the pressurizer. Here, due to circulation being solely by natural convection, i.e., in the absence of any active primary water pumping means, it is the primary water flow throttling valve and the thermo-hydraulic parameters (hot and cold temperatures) that determine the circulation and heat exchange conditions between the primary and secondary circuits in relation to the power produced in the core.In fact, the control of the operation of the heat generator reactor can be carried out simply with, in addition to this primary water pressure setting, the adjustment of the core power by all the control rods 42. .

[0110] The heat exchangers 49 between the primary and secondary circuits are preferably plate heat exchangers, advantageously made of stainless steel, and designed to withstand the water pressure of the primary circuit. Advantageously, these heat exchangers 49 are manufactured by stacking grooved metal plates assembled together either by hot isostatic compression (HIC) or by hot uniaxial compression (HUC) so as to obtain diffusion welding between the metal plates, or by brazing.

[0111] Within an exchanger 49, the flow is downward for the primary water, and upward for the secondary water.

[0112] As shown in [Fig.6], the secondary circuit of this reactor 4 is not a closed-loop circuit as in conventional PWR reactors, but includes a water basin B, as schematically shown in [Fig.5]. This basin B is contained within the space of the reactor vessel well forming the third containment barrier, and the reactor vessel 4 is immersed in it.

[0113] This secondary circuit with liquid water basin B is an open medium delimited by the tank well, without a circulation pump.

[0114] With such a liquid water basin B for the secondary circuit, the heat exchangers 49 are not integrated into the reactor vessel 40, 41 but are arranged and fixed outside of it. Such an arrangement is possible because the unlikely event of a rupture of the primary water inlet or outlet pipes, causing a large-diameter breach, does not have significant accidental consequences for the reactor, thanks in particular to the primary and secondary pressures. Indeed, the liquid water basin B completely surrounds the reactor vessel 4, and an accident of this type cannot lead to a risk of core uncovering, endangering the physical integrity of the reactor core.

[0115] The internal circuit within a heat exchanger 49, which is part of the secondary circuit of reactor 4, therefore sees a flow of liquid water as a secondary fluid that heats up in contact with the primary water within the heat exchanger 49, by natural suction from its inlet manifold 490 at the bottom to its outlet manifold 491 at the top. The secondary water then creates a volume greater than a so-called hot temperature. The separation layer between a so-called cold temperature and the so-called hot temperature of the secondary water is designated as a thermocline, as symbolized by the term thermocline in [Fig. 6].

[0116] In other words, when the SMR reactor is in normal operation, the water basin B is configured to achieve vertical thermal stratification, resulting in the formation of a thermocline delimited between the bottom of the basin at a cold temperature, in which the reactor vessel 4 is immersed, and the top of the basin at a hot temperature. The height of the thermocline layer determines the flow rate. cooling of the secondary circuit through the exchangers 49. The closed circuit circulation S solely by natural convection of the secondary water is symbolized by the grey arrows in [Fig.6].

[0117] The natural convection flow rate of the secondary water is regulated by control valves 5 integrated into each of the outlet manifolds 491 of the heat exchangers 49, as illustrated in [Fig. 9]. The cold secondary water temperature is governed by the temperature conditions of the secondary water basin B. The hot water temperature is set by the control valves 5 in the outlet manifolds 491, and by the heat exchange within the heat exchangers 49 themselves.

[0118] An example of integrating a control valve 5 in the form of a butterfly valve 50 into an outlet manifold 491 is shown in Figures 10A, 10B, and 10C, which show the valve in a fully open position, allowing the maximum flow of secondary water from the basin, an intermediate position, and a fully closed position, allowing no flow. The butterfly 5 is rotated by the output shaft 51 of an electric motor 52.

[0119] Advantageously, the end of the shaft 51 opposite that connected to the butterfly 50 is connected to a remote weight 53. As shown in [Fig.1OA], in the event of an electrical failure or an emergency stop being triggered, the gravitational fall of the weight 53 puts the valve 5 in its fully open position so as to circulate the maximum flow of secondary water from basin B.

[0120] During the start-up phase of the nuclear reactor, the thermocline is completely aligned with the upper free level of the secondary water basin B. The driving head of the secondary water circulation is then at its maximum due to the maximum weight of the cold water column supplying the inlets of the heat exchangers 49. The thermal power demand on the primary circuit is then at its maximum, and the average temperature of the primary water decreases. The lowering of the primary water temperature leads to average cooling of the moderator in the core, thereby inducing an increase in core reactivity, and thus an increase in its thermal power. Maximum thermal heating conditions of the secondary water volume are accompanied by a natural increase in core power; the reactor 4 is therefore naturally stable.As already mentioned, the position of the primary water shunt valve, combined with the positions of the 42 control rods for reactivity, makes it possible to limit the increase in core reactivity, in order to remain within the temperature rise range of the entire reactor block 4 and its reactor vessel well.

[0121] Conversely, when the thermocline level drops, this implies a rise in the secondary hot water layer, and therefore a decrease in the driving head of secondary water through the heat exchangers 49 since the height of the cold water column decreases. Thus, the circulation of secondary water by natural convection decreases. This effectively reduces heat exchange between the primary and secondary circuits. In core C, the decrease in power output leads to an increase in the average temperature of the primary water, and therefore to an increase in the average temperature of the moderator in the core. Consequently, there is a decrease in reactivity due to moderator expansion, and the neutron and thermal power produced decreases. The reactor is therefore naturally stable for heat output and storage to the secondary water volume defined by basin B.Typically, a difference in altitude between the median plane of core C and the median plane of the exchangers 49 of about 4 meters allows the development of the natural circulation of a cold primary fluid at 80°C at the outlet of exchanger 49, and hot at 120°C at the outlet of core C, generating about 1100 Pascals of driving pressure necessary to overcome the pressure losses of core C, the exchangers 49, and the rest of the primary circuit including the additional adjustable pressure loss formed by the rolling ring 480.

[0122] The secondary water volume is determined by the dimensions of the tank well on the one hand, and by the height dedicated to the cold and hot zones of the secondary water on the other. Typically, the secondary water volumes are on the order of 200 to 300 m3 for the cold zone, and 100 to 150 m3 for the hot zone, i.e. a total volume for basin B of between 300 and 450 m3.Typically, an altitude difference of about 4 meters between the median plane of the exchangers 49 and the position of the secondary thermocline separating a secondary cold water layer at 65°C from the hot water layer at 105°C allows the development of the natural circulation of a secondary cold fluid at 65°C at the inlet of exchanger 490, and hot at 105°C at the outlet of exchanger 491, generating about 1000 Pascals of driving pressure necessary to overcome the pressure losses due to the passage through the exchangers 49, from the suction 490 to the outlet 491, including the pressure losses adjustable by means of the secondary flow control valves 5.

[0123] The thermocline position can only be maintained at a fixed position if a quantity of secondary water at its hot temperature is continuously drawn off and replaced by the same quantity of secondary water at its cold temperature. Therefore, an adjustable pumping system exists to transport the power corresponding to customer demand—that is, the power required by the heating network—to a district heating network. In the event of an unexpected interruption of this heat removal, or an unforeseen shutdown of the pumping system, the stability conditions described above allow the power produced by the reactor core to be temporarily stored by modifying the ratio between the secondary water at its cold temperature and its hot temperature, and by lowering the thermocline level. After several minutes of operation, the removal continues. The lack of an external vent for the thermal power produced by the reactor necessitates shutting down the reactor to remove only the residual power through dedicated residual power removal systems. Typically, a continuous thermal power output of 20 MW, with secondary water supplied at 90°C and returning at 45°C, requires pumping 123 liters per second, or 442 m³ per hour, from the hot water layer above the thermocline and returning the same quantity to the bottom of the reactor vessel. Preferably, this pumping and return can be implemented using piping from the top of the reactor vessel to avoid lateral connections that could cause leaks or lateral structural integrity issues, thus limiting expansion and seismic resistance.

[0124] The heat transfer of 20 MW thermal energy from the pumped secondary water layer to the tertiary water circuit of the customer network is carried out by means of one or more heat exchangers sized to transfer 20 MW thermal energy with a hot pinch-off of 15°C (from 105°C to 90°C) and a cold pinch-off of 20°C (from 65°C to 45°C). This secondary water is pumped using pumping units, preferably installed in parallel, to provide operational redundancy in the event of a failure or the need for intervention.

[0125] The presence of this piping must not interfere with the transport of the entire reactor block, as detailed later for its removal from the reactor vessel well using heavy handling equipment.

[0126] As previously stated, the reactor's primary circuit operates solely by natural convection, i.e. without a pumping unit.

[0127] Consequently, the inventors were faced with a problem of realizing a pressurizer whose primary fluid vapor cooling and condensation part cannot be designed with a liquid water spraying / injection device, from a sample of the primary circuit as according to the state of the art.

[0128] The inventors then thought of modulating the heat losses by conduction through the dome 6, to control the depressurization of the primary steam of the pressurizer, taking advantage of the fact that the metal casing 60 of the cover 41 is of low thickness, typically between 10 and 20 mm.

[0129] Thus, as illustrated in [Fig. 1 1], the steam cooling and condensation part comprises a double-walled dome 6 60, 61 separated from each other forming a space E within which liquid water from basin B can flow from the bottom to the top of the dome forming a central exhaust chimney 62. Typically, the space E has a constant height of the order of 0.5 to 2 cm.

[0130] In normal operation, the thermocline level is fixed sufficiently above the pressurizer, in particular so as to be above the central exhaust stack 62, as illustrated in [Fig.11].

[0131] Thus, as illustrated in [Fig. 11], the liquid water circulating solely by natural convection in the space E delimited by the two walls 60, 61, from a cold temperature below the thermocline, will condense the saturated vapor of the primary circuit inside the tank and thus reduce the pressure within the tank. Typically, the cold temperature of the liquid water entering the space E at the bottom of the dome 6 is around 65°C, which allows for efficient and rapid cooling of the dome 6, and in particular of the inner wall 60 forming the enclosure of mechanical resistance to the pressure of the primary circuit, and consequently of the primary water vapor underlying it, to a saturation temperature of the primary pressure around 3.5 bar, or approximately 140°C. Typically, such a device allows the extraction by natural circulation of the order of 0.3 MW thermal, and thus condenses about 0.15 kg / s of steam at saturation.The saturated steam inventory in the pressurizer, under normal operating conditions, is on the order of a few kilograms, depending on the required pressurizer volume. The primary depressurization capacity is therefore fully compatible with the primary pressure control requirements.

[0132] The central chimney 62 incorporates a regulating valve 64, or in other words, a laminating valve, which allows the flow rate of secondary liquid water circulating in space E to be adjusted, thus regulating the liquid cooling process. Indeed, in a fully closed position of the valve 64, the water layer is trapped and stratified in space E. Conversely, in an open position, particularly a fully open position, the hot water rises naturally in space E and then through the central chimney 62, joining the upper hot water layer of basin B, while the cold water from basin B is drawn in through the lower inlet of the double wall 60, 61.

[0133] Valve 64 can be a butterfly valve like secondary flow valve 5 illustrated in Figures 10A, 10B, 10C.

[0134] The two walls 60, 61 of the dome 6 are metallic, preferably stainless steel.

[0135] The outer wall 61 of the dome 6 is advantageously covered with a cap 63 housing within it a thermal insulator to prevent the cooling of the dome when the valve 64 is in the closed position.

[0136] According to an advantageous embodiment, the reactor 4 comprises, as a passive heat sink, a plurality of cooling fins 65 arranged inside the inner wall, preferably distributed uniformly over the surface of the latter, preferably by being welded or brazed. These cooling fins This increases the vertical condensation surfaces inside the pressurizer, thus allowing for better gravity flow of the steam condensate film. In this way, a high primary steam condensation rate is maintained by removing and replacing the liquid film created on the cold surface.

[0137] These fins 65 also increase the total contact area with the steam in the primary circuit and thus improve heat exchange by conduction between said steam and the dome 5. In the illustrated example, these fins 65 are straight and extend over a major part of the height of the dome. These fins 65 are preferably made of the same material as the walls 60, 61 of the dome 6, and are typically a few centimeters thick and a few tens of centimeters long along the inside of the wall 60.

[0138] Furthermore, the heating portion of the pressurizer comprises a plurality of electrical resistors 8 wrapped in electrical insulation and supplied by electrical cables, arranged inside the dome, preferably on the separating plate 7, which in its center comprises a perforated portion 70 enabling the thermal insulation and pressure differential functions of the integrated pressurizer. Such a perforated portion 70 is, for example, as described in the device in patent application WO 2012 / 158929A3.

[0139] The electrical resistances 8 may be of the type described in US patent 4135552.

[0140] The invention is not limited to the examples just described; in particular, features of the illustrated examples can be combined in unillustrated variants.

[0141] Other variants and embodiments may be envisaged without departing from the scope of the invention. List of cited references

[0142] [1]: The World Nuclear Industry Status Report 2017. https: / / www.worldnuclearreport.org / IMG / pdf / 20170912wnisr2017-en-lr.pdf

Claims

Demands

1. Pressurized water reactor (4) (PWR), of the modular reactor (SMR) type, comprising: - a central axis reactor vessel (X) including a cover (41) in the form of a dome (6) and housing at least part of the primary circuit; - a primary circuit pressurizer whose steam cooling and condensation part comprises: • two walls (60, 61) of the dome, separated from each other forming a space (E) within which water in liquid form can circulate from the bottom to the top of the dome forming a central exhaust stack (52), so as to condense the steam in the saturation state of the primary circuit inside the vessel and thus reduce the pressure within the vessel;• a regulating valve (6) for the flow of liquid water circulating in the space, arranged in the central exhaust stack - a secondary water circuit, the liquid water intended to circulate between the two walls of the dome being the water from the reactor's secondary circuit.

2. Reactor according to claim 1, the two walls of the dome being metallic, preferably stainless steel.

3. Reactor according to claim 1 or 2, the inner wall (50) of the dome forming the enclosure for mechanical resistance to pressure of the primary circuit.

4. Reactor according to claim 3, the thickness of the internal wall being between 10 and 20 mm.

5. Reactor according to any one of the preceding claims, comprising a passive heat sink arranged inside the dome to cool the steam of the primary circuit.

6. Reactor according to claim 5, the heat sink comprising a plurality of cooling fins (65) arranged inside the inner wall, preferably being distributed uniformly over the surface of the latter.

7. Reactor according to claim 6, the fins being welded or brazed to the inner wall of the dome.

8. Reactor according to any one of the preceding claims, the space between the inner and outer walls of the dome being between 0.5 and 5 cm.

9. Reactor according to any one of the preceding claims, the outer wall of the dome being covered with a cap (63) housing within it a thermal insulator.

10. Reactor according to any one of the preceding claims, the pressurizer liquid water intended to circulate between the two walls of the dome being at least 10°C lower than the primary circuit liquid water temperature.

11. Reactor according to any one of the preceding claims, the heating part of the pressurizer comprising a plurality of electrical resistors (8) wrapped in electrical insulation, arranged inside the dome.

12. Reactor according to any one of the preceding claims, comprising a secondary water circuit including a water-filled basin contained in the reactor vessel well space and supplying the bottom of the space between the dome walls.

13. Reactor according to claim 12, the water basin being configured so that, when the SMR reactor is in normal operation, it achieves vertical thermal stratification resulting in the formation of a thermocline delimited between the bottom of the basin at a so-called cold temperature in which the SMR reactor vessel is immersed and the top of the basin at a so-called hot temperature, the thermocline being positioned above the dome so that the circulation of liquid water from the basin into the space between the walls of the pressurizer is achieved by natural convection.

14. Nuclear installation comprising at least one SMR nuclear reactor according to one of the preceding claims.

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