LIGHT WATER REACTOR NUCLEAR POWER PLANT AND HIGH-TEMPERATURE WATER ELECTROLYSIS SYSTEM(S) FOR THE PRODUCTION OF HYDROGEN USING HEAT FROM THE BOILING WATER REACTOR

Description

technical field

[0001] The present invention relates to the field of light water reactors (LWRs), particularly pressurized water reactors (PWRs) or boiling water reactors (BWRs). More specifically, the invention relates to cogeneration plants comprising such nuclear reactors. Unless otherwise specified, "cogeneration" herein and within the scope of the invention means the simultaneous or separate production of electricity and, where applicable, heat, and the production of hydrogen by electrolysis of water from the heat and electricity supplied by the nuclear reactors.

[0002] The main objective of the invention is to optimize the efficiency of the primarily thermal coupling in order to optimize the production of hydrogen from the heat supplied by nuclear reactors.

[0003] The production of hydrogen within the framework of the invention is carried out by high-temperature water electrolysis (HTE, or EVHT for high-temperature steam electrolysis, or HTE, an English acronym for " High Temperature Electrolysis ", or HTSE, an English acronym for " High Temperature Steam Electrolysis " also to solid oxides (SOEC, English acronym for " Solid Oxide Electrolysis Cell ".

[0004] Although described with reference to a pressurized water reactor, the invention applies to any nuclear reactor with an indirect thermodynamic cycle belonging to the second, third, and fourth generation (GEN IV) reactor family. It applies in particular to boiling water reactors (BWRs), fast neutron reactors cooled with liquid metal, notably liquid sodium, also known as SFRs (Sodium Fast Reactors), which are part of the GEN IV reactor family. Previous technique

[0005] In the context of climate and energy transition, the nuclear industry must meet several challenges for the future. Indeed, to address tomorrow's energy and societal challenges, it will be essential to design nuclear reactors that enable: to limit the need for so-called "environmental" liquid cold sources (rivers, streams, sea) and associated discharges into the environment; to be more flexible and therefore more complementary to other so-called renewable energies (RE), to meet fluctuating electricity demand and the intermittency of RE; to decarbonize processes by supplying heat to consuming industries (desalination, heat networks, hydrogen, etc.) while increasing energy efficiency; to capture atmospheric CO2 to limit the effects of global warming and contribute to closing the carbon cycle as a carbon source for industrial processes; and this without degrading the profitability of the installation, either by economically benefiting from the new services provided, or by significantly increasing the amount of electricity produced during the day.

[0006] A pressurized water reactor (PWR) typically comprises three cycles (fluidic circuits), the general principle of whose normal operation is explained below with reference to the figure 1 The temperatures and yields are given for illustrative purposes only.

[0007] The primary circuit 1 is a closed-loop fluid circuit comprising mainly the reactor core 2, at least one steam generator (SG) acting as a heat exchanger (referred to as the primary heat exchanger 3), and a hydraulic pump 4 to circulate the heat transfer fluid, which is water maintained in a liquid state within the reactor's operating temperature range, typically around 320°C-330°C during normal operation. Other equipment, such as a pressurizer and all the devices ensuring safe operation, is not described here.

[0008] Thus, the high-pressure water of the primary circuit extracts the energy supplied, in the form of heat, by the fission of uranium nuclei in the core of reactor 1.

[0009] Next, this water under high pressure and high temperature, typically 155 bar and 320°C-330°C, enters the intermediate exchanger 3 and transmits its energy to a secondary circuit 5, which also uses pressurized water as a heat transfer fluid in a closed loop.

[0010] This secondary circuit 5 includes the intermediate exchanger 3, a turbine 6 comprising a high-pressure body 60 and a low-pressure body 61, a condenser 7 and a hydraulic pump 8 to circulate water in the form of steam as a heat transfer fluid.

[0011] Thus, in this secondary circuit 5, water in the form of steam, at high pressure, typically about 70 bar, is expanded in the high-pressure body of the turbine, then superheated before continuing its expansion in the low-pressure bodies 61. The turbine drives an alternator 9 which produces electricity.

[0012] The water from the secondary circuit is then condensed via the condenser 7 in a third cycle, the cooling cycle 10, as a so-called "cold" source. This cycle 10 mainly comprises humid air cooling towers 11, which are hollow towers in their center, creating a natural airflow that enters at the bottom and exits at the top. As this airflow passes through, it extracts heat from the water in the cooling circuit and disperses it into the atmosphere as a cloud of water vapor. This process is repeated continuously, with the water distributed as fine droplets. This allows for efficient heat exchange between the water and the air, thus bringing the water temperature close to that of the ambient air, and also saturates the airflow circulating from bottom to top within the tower with water vapor.Part of the water flow evaporates in tower 11, while the rest falls as rain into the basin below the tower, where it is pumped and returns to cool condenser 7. The evaporated water is replaced by tertiary water, also known as "environmental" water, pumped upstream from a river, stream, or sea. This significantly increases the temperature of these waterways, which, during periods of high water and / or low flow, may lead a nuclear facility operator to reduce their power output or even shut them down.

[0013] As shown on the figure 1 For example, the thermodynamic efficiency of a PWR is around 33 to 34%, the water temperature at the inlet of condenser 7 is around 20°C and 35°C at its outlet.

[0014] In the conventional PWR sector, reactors are classified by major categories of use: reactors known as power-generating reactors, which are dedicated solely to the production of electricity; reactors known as heat-generating reactors, which are dedicated solely to the production of heat; and cogeneration reactors, dedicated to both the production of electricity and heat, or to the production of electricity and, where applicable, heat and hydrogen, simultaneously or not.

[0015] As detailed in [1], one principle of combined heat and power (CHP) from a nuclear reactor involves modifying the design of the energy conversion cycle so that the heat is released at the cold source at a temperature that allows for its use. This principle is typically applied to installations such as seawater desalination plants or district heating systems, where the required heat is at a relatively low temperature, typically around 150°C.

[0016] Limiting global warming requires minimizing heat losses at all levels, and particularly at the cold source of a thermodynamic installation. This cogeneration objective becomes even more relevant for a nuclear reactor, as industrial or domestic heat is often traditionally obtained by burning fossil fuels, which are responsible for greenhouse gas emissions.

[0017] To achieve this objective, a first configuration consists of modifying the components of the electrical production system of a PWR installation in order to adjust the water temperature at the cold source level.

[0018] In a classic configuration, illustrated on the figure 1 However, this modification remains limited. It does not affect the high-pressure turbine 60 but only the low-pressure turbine 61, which drives the Rankine cycle. This modification is illustrated in the figure 2This involves reducing the operating point P of the low-pressure turbine 61 to a pressure on the order of one bar, instead of approximately 50 mbar, so that the water exiting the condenser has a sufficiently high temperature, typically 70°C, to be used, for example, in a district heating network 12. This modification is initially accompanied by a decrease in the electrical power produced, since the thermodynamic efficiency drops to 27%. There is also an increase in the pressure within the condenser 7.

[0019] Recently, numerous studies have highlighted the possibility of combining nuclear power reactors with hydrogen production units from high-temperature water electrolysis, in order to create cogeneration plants that efficiently produce electricity and hydrogen: [2], [3].

[0020] Water electrolysis is an electrochemical reaction that decomposes water into dioxygen and dihydrogen gas with the help of an electric current according to the reaction: H2O → H2 + ½ O2.

[0021] To carry out the electrolysis of water, it is advantageous to carry it out at high temperature typically between 600 and 950°C, because part of the energy required for the reaction can be supplied by heat, which is cheaper than electricity, and carrying out the reaction is more efficient at high temperature and finally it may not require a catalyst.

[0022] To implement high-temperature electrolysis, it is known to use a reactor, also called a SOEC type electrolyzer (English acronym for " Solid Oxide Electrolysis CellThis system consists of a stack of elementary units, each containing a solid oxide electrolysis cell made up of three stacked anode / electrolyte / cathode layers, and interconnecting plates, for example, made of metallic alloys, also called bipolar plates or interconnectors. The interconnectors ensure both the passage of electric current and the circulation of gases in the vicinity of each cell (injected steam, extracted hydrogen and oxygen in a high-temperature (HTE) electrolyzer) and separate the anodic and cathodic compartments, which are the gas circulation compartments on the anode and cathode sides of the cells, respectively. To perform high-temperature steam electrolysis (HTE), steam (H₂O) is injected into the cathodic compartment.Under the influence of the current applied to the cell, the dissociation of water molecules into vapor occurs at the interface between the hydrogen electrode (cathode) and the electrolyte: this dissociation produces dihydrogen gas (H₂) and hydroxide ions (OH⁻). The dihydrogen is collected and discharged from the hydrogen compartment. The hydroxide ions (OH⁻) migrate through the electrolyte and recombine into dioxygen at the interface between the electrolyte and the oxygen electrode (anode).

[0023] To increase the flow rates of hydrogen and oxygen produced, it is known to stack several elementary electrolysis cells on top of each other, separated by interconnecting devices, usually called interconnectors or bipolar interconnecting plates. The assembly is positioned between two end interconnecting plates that support the electrical and gas supplies to the electrolyzer (electrolysis reactor).

[0024] A high-temperature water electrolyzer (HTE) thus comprises at least one, usually a plurality of electrolysis cells stacked on top of each other, each elementary cell being formed of an electrolyte, a cathode and an anode, the electrolyte being intercalated between the anode and the cathode.

[0025] In steady-state operation, a high-temperature electrolyzer (HTE) requires upstream combined thermal and electrical power to bring the reactants to the temperature required for electrolysis, and within itself an electrical power input to carry out the electrolysis.

[0026] To ensure the supply of heat and electricity to an EHT electrolyzer, patent CN207603212 proposes a configuration for extracting high-temperature, high-pressure steam from the outlet of a nuclear reactor's steam generator (SG) and injecting it directly into the EHT electrolyzer. This configuration has several major drawbacks: Regarding reactor safety: drawing steam from the steam generator outlet to feed the high-temperature electrolyzer (HTE) impacts the water inventory of the secondary circuit. This configuration necessitates the implementation of a water supply and treatment unit with the physicochemical properties required by the secondary circuit to maintain its water inventory continuously. Furthermore, any failure in this unit could lead to a partial or total loss of the steam generator's power removal function on the primary circuit, thereby impacting the overall safety of the installation. Regarding the operation of the overall installation: having water supplying both the secondary circuit and the HTE requires compatibility between the physicochemical requirements of the secondary circuit water and the electrolyzer feedwater.As a preliminary example, the requirements for non-aggressive secondary water on the secondary loop materials necessitate chemical treatments, which appear difficult to reconcile with water intended to feed an electrolyzer. Indeed, the water introduced into an EHT electrolyzer must be as pure as possible because impurities remain within the electrolyzer and accumulate over time.These impurities can in fact end up disrupting electrolytic reactions, firstly by the formation of sludge and secondly by the action of chlorides (halides in general) which destroy the nickel plating or the solid nickel generally required for the anodes of an EHT electrolyzer; at the level of the overall installation safety, due to the transport of pressurized steam from the conventional island of the nuclear installation (machine room) to the EHT electrolysis unit, there is a constraint of compatibility between the ability to separate the installation from the unit and the risk analysis imposing a minimum distance between them: reference can be made to publication [4] which indicates an estimated safety distance of 1 km which must be able to be reduced in view of the risk analysis carried out.

[0027] Publication [3] proposes a coupling configuration between a Rankine cycle PWR nuclear power plant and an EHT electrolyzer unit. In the proposed coupling configuration, heat extraction to power an EHT electrolyzer is achieved indirectly, i.e., without direct steam extraction. This avoids impacting the water inventory of the secondary circuit. This coupling configuration thus proposes the implementation of a fluid loop that extracts steam directly from the steam generator outlet, i.e., at approximately 280°C, and reinjects the steam, after passing through a heat exchanger, directly into the Rankine cycle condenser. In this configuration, the extracted heat is not fully utilized by the proposed EHT unit, which is not thermally efficient.

[0028] In summary, the energy efficiency of nuclear cogeneration facilities as mentioned above is not optimal, because the coupling between the nuclear reactor(s) for power generation and / or heat generation and the EHT electrolyzer unit for hydrogen production is not optimal.

[0029] However, this energy efficiency is of primary importance to the economic viability of producing decarbonized hydrogen. Designing efficient and safe nuclear cogeneration plants will help ensure the feasibility of deploying such a sector.

[0030] Therefore, there is a need to improve nuclear cogeneration facilities, i.e., those producing electricity and hydrogen simultaneously or not from the heat supplied by the nuclear reactor(s) of the installation, by optimizing the thermal coupling between the nuclear reactor(s) and the high-temperature electrolysis (HTE) unit dedicated to hydrogen production, in particular: with a performance at least equal to the solutions proposed so far, which does not compromise the safety of the nuclear reactor(s), and which does not present a risk to the overall safety of the installation, and this without complicating the operating conditions of both the nuclear reactor(s) and of the EHT unit considered separately.

[0031] The aim of the invention is to at least partially meet this need. Description of the invention

[0032] To this end, the invention relates, in one of its aspects, to a nuclear power cogeneration plant, intended to produce electricity and, where applicable, heat and hydrogen, comprising: at least one nuclear reactor, in particular a pressurized water reactor (PWR) or a boiling water reactor (BWR), comprising: a first fluidic circuit, called the primary circuit, comprising at least one first steam generator as the first intermediate heat exchanger; a second fluidic circuit, called the secondary circuit, comprising: at least one turbine comprising a high-pressure casing connected to the first steam generator and a low-pressure casing connected to the high-pressure casing by at least one fluidic branch, a high-pressure heater connected in a closed loop to the high-pressure casing, to a water reservoir, called the feedwater tank, and to the first steam generator, a low-pressure heater connected on one side by at least one fluidic branch to the low-pressure casing and on the other side by at least one fluidic branch to the feedwater tank, a condenser connected on one side to the low-pressure casing and on the other side in a closed loop to the low-pressure heater; an alternator mechanically coupled to the turbine, intended to be connected to an electrical network.

[0033] According to the invention, the installation comprises at least one high-temperature electrolysis unit (HTE) thermally coupled to the nuclear reactor by a third fluidic circuit, called the injection coupling circuit, comprising: a second intermediate heat exchanger connected in a closed loop to a draw-off tap, made in the fluidic branch between the high pressure body and the low pressure body of the turbine, and to the feed tank, a second steam generator, called the coupling steam generator, connected on one side in a closed loop to the second heat exchanger, and on the other side to a fluidic branch whose inlet is connected to a liquid water supply, separate from the primary and secondary circuits of the nuclear reactor and the outlet is connected to the inlet of the high temperature electrolysis unit so as to inject the steam produced by the coupling steam generator.

[0034] Advantageously, the liquid water supplying the coupling steam generator is demineralized water.

[0035] For the purposes of this invention, a "high-temperature water electrolysis unit" is defined as a hydrogen production unit comprising one or more high-temperature water (HTW) electrolyzers in series or parallel flow. Each electrolyzer comprises at least one, and generally a plurality, of stacked electrolysis cells. Each cell consists of an electrolyte, a cathode, and an anode, with the electrolyte interposed between the anode and the cathode. The electrolysis unit further includes all components upstream and downstream of the electrolyzer(s), such as heat exchangers, hydrogen and oxygen compressors, etc. Reference may be made to the figures 2 And 3 from publication [5].

[0036] According to an advantageous configuration, the nuclear reactor is a PWR reactor comprising a steam dryer, arranged on the fluidic branch connecting the high-pressure body and the pressure body of the turbine, the withdrawal tap being made at a point in the branch between the high-pressure body and the steam dryer.

[0037] Advantageously, the injection coupling circuit is adapted to draw steam from the draw-off tap at a temperature of no more than 200°C, advantageously still between 150 and 180°C, so that the coupling steam generator transforms the liquid water that feeds it into steam at a temperature of no more than 40°C, advantageously between 10 and 35°C.

[0038] Preferably, the heat transfer fluid in the closed loop connecting the coupling steam generator to the second heat exchanger is pressurized water.

[0039] According to an advantageous embodiment, the high-temperature electrolysis production unit (EHT) is further thermally coupled to the nuclear reactor by a fourth fluidic circuit, called the reinjection coupling circuit, comprising: a fourth intermediate heat exchanger connected in a closed loop to a first reinjection tap made between the low pressure heater and the feed tank, and to a second reinjection tap made between the condenser and the low pressure heater, a fifth intermediate heat exchanger, called the coupling exchanger, connected on one side to a fluid branch whose inlet is connected to a liquid water supply from the high temperature electrolysis unit and whose outlet is connected to at least one cooling circuit of the high temperature electrolysis unit and on the other side in a closed loop to the fourth heat exchanger, so as to remove at least part of the excess heat from the high temperature electrolysis unit and reinject it into the secondary circuit of the nuclear reactor.

[0040] According to this method, the reinjection coupling circuit is preferably adapted to take liquid water from the high-temperature electrolysis unit at a temperature of at least 80°C, advantageously between 85 and 95°C.

[0041] Preferably, the heat transfer fluid in the closed loop connecting the coupling exchanger to the fourth heat exchanger is pressurized water.

[0042] Advantageously, the alternator is suitable to supply at least some of its electricity to the high-temperature electrolysis unit.

[0043] According to an advantageous embodiment, the draw-off connection is equipped with a valve for regulating the inlet pressure in the third intermediate heat exchanger. Preferably, the regulating valve is also a valve for regulating the inlet pressure in the high-pressure heater.

[0044] Advantageously, the second intermediate heat exchanger has the same structure as the high-pressure heater.

[0045] The invention essentially consists of achieving optimal indirect thermal coupling between a nuclear reactor and a high-temperature electrolysis hydrogen production unit via a strategically chosen extraction port downstream of the high-pressure turbine core of the nuclear reactor's Rankine cycle. This port transfers the extracted heat, via a closed loop between an intermediate heat exchanger and a coupling steam generator, to liquid water supplying the production unit. This liquid water then injects steam at the high temperature required for electrolysis. This indirect heat input can be continuous or gradually reduced until the heat released by the electrolysis reaction is sufficient to power the unit itself.

[0046] By "indirect" we mean here and within the framework of the invention the fact that there is no mass transfer between the cycles of the nuclear reactor and the hydrogen production unit by electrolysis EHT, and therefore that the heat transfer takes place through heat exchangers.

[0047] In an advantageous embodiment, another closed-loop heat exchanger is also implemented to achieve indirect coupling by returning waste heat from the electrolysis hydrogen production unit (EHT) to the nuclear reactor. In other words, according to this embodiment, the thermal discharges from the electrolysis hydrogen production unit (EHT) are at least partially reinjected into the nuclear reactor.

[0048] In conclusion, the proposed indirect thermal coupling for transferring heat from a reactor to a high-temperature electrolysis unit, advantageously with indirect coupling through reinjection of waste heat from the electrolysis unit, offers numerous advantages, including: an overall efficiency of the nuclear cogeneration plant at least equal to existing solutions, unmodified safety of the nuclear reactor, no risk to the overall plant, operating conditions of the nuclear reactor and the EHT unit which remain the same and independent of each other.

[0049] Other advantages and features of the invention will become clearer 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

[0050] [ Fig 1 ] there figure 1schematically illustrates a configuration of a pressurized water reactor (PWR) operating solely as a power-generating reactor according to the state of the art. Fig 2 ] there figure 2 is a schematic view of a pressurized water reactor (PWR) configuration modified to operate as a state-of-the-art cogeneration reactor. Fig 3 ] there figure 3 is a schematic view showing in more detail the components of the secondary and tertiary circuits of a pressurized water reactor (PWR) nuclear power plant according to the state of the art. Fig 4 ] there figure 4 is a schematic view of a cogeneration installation of a pressurized water reactor (PWR) thermally coupled with a high-temperature electrolysis unit according to the invention. Detailed description

[0051] Throughout this application, the terms "upstream" and "downstream" are to be understood by reference to the direction of flow of a heat transfer fluid within one of the fluid circuits of a nuclear cogeneration plant according to the invention.

[0052] Points A to H of the various relevant fluidic lines of the figure 4 are symbolized by black dots.

[0053] THE figures 1 and 2 The aspects relating to the state of the art have already been detailed in the preamble, so they will not be commented on below.

[0054] 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 the figures 1 to 4 .

[0055] A nuclear installation with a PWR reactor, as it currently exists, has been represented.

[0056] The primary circuit 1 is a closed-loop fluidic circuit comprising mainly the reactor core 2, a heat exchanger 3 as a steam generator (SG), and a hydraulic pump 4 to circulate the heat transfer fluid which is water which changes from the liquid state to the vapor state, typically around 300°C at high pressure, typically around 70 bar, in normal operation.

[0057] Other equipment, such as a pressurizer and all the devices that ensure operation under the required safety conditions, is not described here.

[0058] The tertiary circuit 10 mainly comprises a turbine 6 consisting of a high-pressure body 60 and a low-pressure body 61 connected to the high-pressure body 60 by a fluidic branch 62, a condenser 7 connected to the low-pressure body 61 by a fluidic branch 63 and a hydraulic pump 80 to circulate water in the form of steam as a heat transfer fluid.

[0059] Thus, in this circuit 10, the water in the form of steam is expanded in the high-pressure body of the turbine, then superheated before continuing its expansion in the low-pressure body 61. The turbine drives an alternator 9 which produces electricity.

[0060] The water from the secondary circuit is then condensed via condenser 7, as a so-called "cold" source.

[0061] The steam generator 3 produces steam for the high and low pressure bodies 60, 61 of the turbine 6, which is characteristic of a Rankine cycle with the operating modalities of a generator cycle of the installation and must be able to operate according to the needs of the electrical network.

[0062] Steam generator 3 is typically sized to remove 1.5 times the power of the nuclear reactor. It should be noted that the turbine housings 60 and 61 are sized based on the peak steam flow rate produced by steam generator 30.

[0063] The hydraulic pump 4 is designed to operate at least at the availability coefficient Kd of the nuclear reactor and must be able to operate according to the fluctuations in the electricity requirements of the electrical network to which the alternator 9 of the nuclear reactor is electrically connected.

[0064] The flow rate of pump 4 must allow, taking into account the heat capacity of the heat transfer fluid and the sizing of the steam generator 3, to supply the latter with heat transfer fluid at a flow rate that allows it to meet the power demands of the electrical network.

[0065] Pump 4 has metal walls resistant to the chemical attack of the heat transfer fluid at high temperatures, typically above 300°C. Several pumps can be positioned in parallel to distribute the pumping flow rate, and a redundant pump can be provided for safety reasons.

[0066] Circuit 10 also includes a circuit for heating the feedwater of the steam generator 3 before it is injected / expanded as steam into the pressure body 60 of the turbine 6.

[0067] This heating circuit includes firstly a high-pressure heater 64 connected in a closed loop to the high-pressure body 60, to a water tank, called the feed tank 66, and to the first steam generator 30. It also includes a low-pressure heater 65 connected on one side by at least one fluid branch to the low-pressure body and on the other side by at least one fluid branch to the feed tank 66.

[0068] In this circuit, a condensate extraction pump 81 from the condenser 7 conveys it to the feed tank 66 via the low pressure heater 65. The condensate is typically at a temperature of 40°C at a pressure of around 0.075 bar at the outlet of the condenser 7.

[0069] As shown in the figure 3The low-pressure heater 65 uses as a heat source the steam drawn from the low-pressure body 61 of the turbine 6. The low-pressure heater 65 thus heats the feedwater of the steam generator 30. When the feedwater leaves the low-pressure heater 65, it typically has a temperature between 80 and 100 °C.

[0070] Pump 80 brings this feed water from the feed tank 66 to the high-pressure heater 64 to bring it to a temperature of around 150°C at a pressure of 40 bars in the steam generator 30.

[0071] As shown in the figure 3 , the high pressure heater 64 uses as a heat source the steam drawn from the high pressure body 60 of the turbine 6.

[0072] We will not detail again all the different relationships and functions of the common elements between a cogeneration plant with hydrogen production according to the invention and a plant with a PWR nuclear reactor according to the state of the art, as illustrated in the figure 3 Only the elements of the indirect thermal couplings according to the invention are described.

[0073] The nuclear cogeneration plant according to the invention illustrated in the figure 4 includes, in addition to the usual components of a typical PWR reactor installation, which has just been described in relation to the figure 3 , at least one EHT 100 high temperature electrolysis unit thermally coupled to the nuclear reactor by an injection coupling circuit 200.

[0074] This injection coupling circuit 200 includes first of all an intermediate heat exchanger 202 connected in a closed loop to a withdrawal tap A, made in the fluidic branch 62 between the high pressure body 60 and the low pressure body 61 of the turbine, and to the feed tank 66. More precisely, taking into account the temperature and pressure of the hot steam that we wish to withdraw, the withdrawal tap A is at a point in the branch 62 which connects the high pressure body 60 of the turbine and the dryer, not shown of the PWR reactor.

[0075] As detailed later, this tapping at point A is chosen at the appropriate temperature level to provide the latent heat necessary to vaporize the feed water in unit 100.

[0076] The tapping of the A supply can be carried out in a strictly identical manner (type of weld, characteristic diameter of pipe, etc.) to that of the tapping supplying the high pressure heater 64.

[0077] Branch A is advantageously equipped with a regulating valve for the inlet pressure in the intermediate heat exchanger 202. This regulating valve can also be used as an inlet pressure regulating valve in the high-pressure heater 64. This allows the valve to be operated in two different modes. This regulating valve optimizes the operation and control of the coupling cycle 200, offering a high degree of flexibility.

[0078] In the closed loop from the draw-off point A, the fluid branch 201 allows the steam to be channeled and transferred to the exchanger 202. After heat exchange, the fluid branch 203 returns the liquid water to a point B of the feed tank.

[0079] This circuit also includes a second steam generator, called the coupling steam generator 205, adapted to produce feed steam at an inlet point F of the electrolysis unit 100 from demineralized liquid water from a point E. This coupling generator 205 is also connected in an intermediate closed loop to the heat exchanger 202.

[0080] In this intermediate closed loop, a circulation pump 206 brings pressurized liquid water via the fluid branch 204 from the exchanger 202 to the coupling generator 205 and then after heat exchange, returns the water via the fluid branch 207 to the exchanger 202.

[0081] To further improve the thermal coupling between the PWR reactor and the EHT electrolysis unit, at least some of the waste heat can be recovered, i.e. the heat emitted by the latter that was not used for the electrolysis reaction.

[0082] To achieve this, the installation includes a 300 coupling circuit by reinjection of heat into the tertiary circuit 10 of the reactor.

[0083] This circuit 300 includes first of all an intermediate heat exchanger 302 connected in a closed loop to a first reinjection branch made at point C between the low pressure heater 65 and the feed tank 66, and to a second reinjection branch made at point D between the condenser 7 and the low pressure heater 65.

[0084] In this closed loop from the draw-off point D, the fluid branch 303 brings pressurized liquid water to the exchanger 302 and after exchange, the fluid branch 301 allows the pressurized liquid water to be channeled and transferred to point C, which allows at least part of the excess calories from the EHT 100 unit to be reinjected into the Rankine conversion cycle 10 of the reactor.

[0085] This circuit also includes an intermediate heat exchanger 305, called a coupling exchanger, connected on one side to a fluid branch 308 whose inlet at the draw-off point G is connected to a liquid water supply from the EHT 100 unit and the outlet is connected to at least one cooling circuit of the EHT 100 unit.

[0086] This coupling exchanger 305 is connected in an intermediate closed loop to the heat exchanger 302, so as to remove at least part of the excess heat from the high-temperature electrolysis unit and reinject it into the tertiary circuit of the nuclear reactor.

[0087] The liquid water from the EHT 100 unit is at a temperature of at least 80°C, advantageously between 85 and 95°C.

[0088] In this intermediate closed loop, a circulation pump 306 brings pressurized liquid water via the fluid branch 304 from the coupling exchanger 305 to the exchanger 302 and then after heat exchange, returns the water via the fluid branch 307 to the exchanger 202.

[0089] In the illustrated example, all the fluid branches 201, 203, 204, 207, and 208 of circuit 200 and the fluid branches 301, 303, 304, 307, and 308 of circuit 300, each consisting of a cylindrical pipe, preferably with metallic walls, are externally insulated with high-temperature insulation. The diameter of each pipe is calculated to allow the dissipation of all the thermal power with a maximum permissible flow velocity of the heat transfer fluid, typically on the order of 1 to 10 m / s.

[0090] The electrical network connected to alternator 9 is designed to transport and distribute electricity to end users according to their needs. It is a high-voltage electrical network operating according to power demands related to electricity use, and must be able to accept the peak electrical power produced by the cogeneration plant.

[0091] In addition, alternator 9 can supply the electricity requirements of the EHT 100 unit, including electrolyzers which require direct current, to implement high-temperature electrolysis.

[0092] The inventors performed a sizing calculation for all the components of the cogeneration plant, characterizing each point of the 200 and 300 circuits in terms of temperature and pressure. This sizing was established using software, called CYCLOP, which the applicant qualified for the steady-state sizing of thermodynamic conversion cycles.

[0093] The CYCLOP software essentially allows the modeling of an energy conversion cycle, consisting of different loops connected by thermal, mechanical, or electrical exchanges. Each loop is made up of components (heat exchangers, pumps, turbines, etc.) linked to each other by fluid circulation.

[0094] This software calculates each thermodynamic point of the complete cycle and deduces the useful energy production and therefore its efficiency. Each component is characterized by selected macroscopic quantities, but this tool can be linked to more detailed pre-sizing modules to obtain more precise characteristics of a given cycle.

[0095] The software also allows the optimization of the efficiency of a cycle according to its free parameters (turbine pressure ratio, withdrawal pressures and flow rates, etc.), using deterministic or genetic optimization algorithms.

[0096] The use of this software and its relevance are described for example in [6] or [7]. The dimensioning can also be carried out using other commercial software, notably that under the name THERMOFLEX ®< .

[0097] For the heat injection coupling circuit 200, a flow rate of approximately 300 kg / sa was considered with thermal coupling of a PWR reactor producing approximately 540 MWth and 180 MWe and considering hydrogen production by unit 100 of equivalent electrical power.

[0098] From a practical standpoint, assuming a flow velocity of 1.5 m / s in the fluid branches 204, 207, and 208, their pipe diameter can be approximately 500 mm, resulting in a pressure drop of around 45 Pa / m, considering only regular pressure losses. For a diameter of 300 mm, a velocity of 4.25 m / s would be achieved with a pressure drop of 850 Pa / m, also considering only regular pressure losses. Alternatively, three 300 mm diameter fluid branches 204, 207, and 208 can be considered in parallel, leading to a pressure drop of 45 Pa / m.

[0099] For the heat reinjection coupling circuit 300, a flow rate of approximately 100 kg / sa was considered with a thermal coupling of a PWR reactor producing approximately 540 MWth and 180 MWe and considering a hydrogen production by unit 100 of equivalent electrical power.

[0100] From a practical point of view, assuming a flow velocity of 1.5 m / s in the fluidic branches, 304, 307, 308, their pipe diameter can be about 300 mm with a resulting pressure loss of the order of 45 Pa / m considering only regular pressure losses.

[0101] Table 1 below gives different values ​​of water temperature and pressure at the various points A to E mentioned above, as well as in the various relevant fluid branches of circuits 200 and 300, which were obtained using the CYCLOP software. [Table 1] Fluidic point / branch Temperature (°C) Pressure (bar) Steam withdrawal point A 167 7,4 Liquid water injection point B 165 7,0 Liquid water injection point C 85 7,0 Liquid water draw-off point D 33 7,4 Demineralized liquid water dispensing point E 25 2,1 Vapor injection point F 115 1,7 Liquid water sampling point G 90 40 Liquid water reinjection point H 40 39,6 Pressurized Liquid Water Fluid Branch 204 150°C 50 Fluidic branch 207 of pressurized liquid water 122°C 49,6 Pressurized liquid water fluid branch 304 87°C 1,4 Pressurized Liquid Water Fluid Branch 307 36°C 1,8

[0102] The invention is not limited to the examples just described; in particular, features of the illustrated examples can be combined in unillustrated variants.

[0103] Other variants and embodiments may be considered without departing from the scope of the invention.

[0104] In the illustrated example, the nuclear reactor is 100% dedicated to hydrogen production. The installation can be operated in other configurations with partial production under nominal conditions, for example, 50% of reactor power dedicated to hydrogen production and 50% to selling electricity to the grid, or with variable operating conditions, for example, reactor power that allows the reactor to sell electricity when market prices are high and to produce hydrogen when prices are lower. The nuclear cogeneration plant just described in relation to a pressurized water reactor can be implemented with all indirect-cycle nuclear reactors, in which the heat production cycle is physically separated from the energy conversion cycle, such as a boiling water reactor or a Generation IV nuclear reactor.

[0105] Thus, if the detailed example concerns a conventional pressurized water reactor, one can consider a cogeneration installation with a large or smaller reactor (SMR).

[0106] Although the detailed example relates to a pressurized water reactor, the invention can equally well be applied to a boiling water reactor, the primary circuit 1 comprising the steam generator 3 as the first intermediate heat exchanger then being constituted by the boiling water reactor vessel from which the generated steam is sent directly into the high pressure turbine body 60, the water from the exchanger 64 then being injected as feedwater to the reactor vessel. List of cited references

[0107] [1]: "Improving energy efficiency by using cogeneration in electricity production" Jean-Marie Loiseaux, Henri Safa, Bernard Tamain, Réseau Sauvons le Climat. [2]: Fujiwara et al. "Hydrogen production by high temperature electrolysis with nuclear reactor", Progress in Nuclear Energy 50 (2008). [3]: Richard D. Boardman et al. "Developing a low-cost renewable supply of hydrogen with high-temperature electrochemistry", MRS Bulletin, volume 47, March 2022, mrs.org / bulletin. [4]: ​​Austin Glover et al. "Assessment of hydrogen plant risks for sitting near nuclear power plants", PSAM 16, June 26-July 1, 2022. [5]: "Dynamic performance analysis of a high-temperature steam electrolysis plant integrated within nuclear-renewable hybrid energy systems". Applied Energy Volume 228, 15 October 2018, Pages 2090-2110. https: / / doi.org / 10.1016 / j.apenergy.2018.07.060. [6]: H.D. Nguyen, N. Alpy, D. Haubensack. "Insight on electrical and thermal powers mix with a Gen2 PWR: Rankine cycle performances under low to high temperature grade cogeneration." Energy, Elsevier, 2020, 202, pp.117518. ff10.1016 / j.energy.2020.117518ff. ffcea-02569231f. [7]: D. Haubensack et al., "The COPERNIC / CYCLOP computer tool: pre-conceptual design of generation 4 nuclear systems, HTR-2004", 2nd International Topic Conference for the HTGR, September 22-24, 2004, Beijing, China, 2004.

Claims

1. Electronuclear cogeneration plant, suitable for generating electricity, heat if required, and hydrogen, comprising: - at least one nuclear reactor, in particular pressurized water reactor (PWR) or boiling water reactor (BWR), comprising: • a first fluidic circuit, referred to as primary circuit (1), comprising at least a first steam generator (3) as first intermediate heat exchanger; • a second fluidic circuit, referred to as secondary circuit (10), comprising: at least one turbine (6, 60), characterized in that the at least one turbine (6, 60) comprises a high-pressure body (60) connected to the first steam generator and a low-pressure body (61) connected to the high-pressure body by at least one fluidic arm (62), a high-pressure heater (64) connected in a closed loop to the high-pressure body (60), to a water container, referred to as feed-water tank (66), and to the first steam generator (30), a low-pressure heater (65) connected to the low-pressure body (61) by at least one fluidic arm and also to the feed-water tank by at least one fluidic arm, a condenser (7) connected to the low-pressure body and also, in a closed loop, to the low-pressure heater; an alternator (9) mechanically coupled to the turbine and intended for connection to an electrical network; - at least one high-temperature electrolysis HTE unit (100) thermally coupled to the nuclear reactor by a third fluidic circuit, referred to as injection coupling circuit, comprising: a second intermediate heat exchanger connected in a closed loop to a bleed-off branch, provided in the fluidic arm between the high-pressure body and the low-pressure body of the turbine, and to the feed-water tank, a second steam generator, referred to as coupling steam generator, as third heat exchanger, connected in a closed loop to the second heat exchanger and also to a fluidic arm (208) of which the inlet is connected to a liquid water feed, separate from the primary and secondary circuits of the nuclear reactor and the outlet is connected to the inlet of the high-temperature electrolysis unit so as to inject there the steam generated by the coupling steam generator.

2. Cogeneration plant according to Claim 1, the nuclear reactor being a PWR reactor comprising a steam dryer, arranged on the fluidic arm (62) connecting the high-pressure body and the pressure body of the turbine, the bleed-off branch (A) being provided at a point of the arm (62) between the high-pressure body (60) and the steam dryer.

3. Cogeneration plant according to either of Claims 1 and 2, the injection coupling circuit being designed to withdraw steam from the bleed-off branch at a temperature of at most 200°C, advantageously comprised between 150°C and 180°C so that the coupling steam generator converts into steam the liquid water fed to it at a temperature of at most 40°C, advantageously comprised between 10°C and 35°C.

4. Cogeneration plant according to one of the preceding claims, the heat-transfer fluid of the closed loop that connects the coupling steam generator to the second heat exchanger being pressurized water.

5. Cogeneration plant according to one of the preceding claims, the high-temperature electrolysis HTE generation unit (100) being additionally thermally coupled to the nuclear reactor by a fourth fluidic circuit, referred to as re-injection coupling circuit, comprising: a fourth intermediate heat exchanger connected in a closed loop to a first re-injection branch provided between the low-pressure heater and the feed-water tank, and to a second re-injection branch provided between the condenser and the low-pressure heater, a fifth intermediate heat exchanger, referred to as coupling exchanger, connected both to a fluidic arm (308) of which the inlet is connected to a feed of liquid water coming from the high-temperature electrolysis unit and the outlet is connected to at least one cooling circuit for the high-temperature electrolysis unit, and in a closed loop to the fourth heat exchanger, so as to discharge at least some of the excess heat from the high-temperature electrolysis unit and to re-inject it into the secondary circuit of the nuclear reactor.

6. Cogeneration plant according to Claim 5, the re-injection coupling circuit being designed to withdraw the liquid water coming from the high-temperature electrolysis unit at a temperature of at least 80°C, advantageously comprised between 85°C and 95°C.

7. Cogeneration plant according to either of Claims 5 and 6, the heat-transfer fluid of the closed loop that connects the coupling exchanger to the fourth heat exchanger being pressurized water.

8. Cogeneration plant according to one of the preceding claims, the alternator being designed to supply at least some of its electricity to the high-temperature electrolysis unit (100).

9. Cogeneration plant according to one of the preceding claims, the bleed-off branch being provided with a valve for controlling the intake pressure in the third intermediate exchanger.

10. Cogeneration plant according to Claim 9, the control valve additionally being a valve for controlling the intake pressure in the high-pressure heater.

11. Cogeneration plant according to one of the preceding claims, the second intermediate heat exchanger having an identical structure to that of the high-pressure heater.