Nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor vessel shaft delimiting a water basin in which the SMR reactor block is immersed.

Abstract

A nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor vessel well delimiting a water basin in which the SMR reactor unit is immersed. The invention relates to a nuclear installation comprising: at least one SMR reactor unit (4) for heat generation, delimited by a reactor vessel (40, 41, 42); a reactor vessel well (100) delimiting a water-filled basin (B) forming part of a secondary circuit, in which the reactor vessel is immersed; at least one heat exchanger (49) between the primary circuit of the SMR reactor and the secondary circuit, arranged outside the reactor vessel and in the water basin. Figure for the abstract: Fig. 6

Description

Title of the invention: Nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor vessel shaft delimiting a water basin in which the SMR reactor block is immersed. technical field

[0001] The present invention relates to the field of nuclear power plants, in particular those comprising pressurized water reactors (PWRs). More specifically, it relates to the field of so-called small or medium power reactors or SMRs (acronym for "Small Modular Reactor"), designed for heat generation.

[0002] The invention thus has as its main objective the significant simplification of the operation and the improvement of the safety of a PWR reactor for heat generation which operates at low pressure, typically less than 6 bar and which is intended to provide a relatively low thermal power, on the order of a few tens of MWth, typically between 5 and 200MWth.

[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] One of the current development topics for nuclear reactors concerns so-called heat-generating reactors designed to provide a level of thermal power of a few tens of MWth, mainly for the purpose of supplying so-called urban heat, i.e. in urban networks, for cities / agglomerations of several hundred thousand inhabitants.

[0008] Among the various technological solutions that provide heat from nuclear fission, it is generally accepted that to date, pressurized water reactors (PWRs) are the most suitable for providing heat at relatively low temperatures.

[0009] Indeed, boiling water reactors (BWRs) are designed to produce steam in a primary circuit which is directly used in a turbo-alternator group to produce electricity.

[0010] Fourth generation fast neutron reactors (FNRs) provide heat at temperature levels beyond the required levels, and have as their main disadvantage a construction and operating cost which makes them incompatible with the exclusive supply of urban heat.

[0011] Similarly, graphite moderator and gas coolant reactors are intended to provide heat at relatively high temperatures.

[0012] Finally, emerging concepts such as molten salt reactors do not have sufficient technological maturity for deployment in the relatively short term.

[0013] It is recalled that a pressurized water reactor (PWR) comprises three cycles (fluidic circuits) whose general principle of normal operation is as follows.

[0014] 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, in the reactor core.

[0015] Next, this water under high pressure and high temperature, typically 155 bar and 300 °C, enters a steam generator (SG) and transfers its energy to 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.

[0016] The water from the secondary circuit is then condensed via a condenser using a third cycle, the tertiary or cooling cycle, as a cold source.

[0017] 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.

[0018] The main elements of a PWR primary circuit are shown in [Fig. 1]:

[0019] - a reactor building 1 performing various functions, including in particular a contribution to the containment safety function,

[0020] - a reactor vessel 20, located in the center of building 1, housing the core C of the reactor,

[0021] - a primary circuit 2 in pressurized water comprising the tank 20.

[0022] These main elements are therefore common, their constitution and the number of components varying according to the power of the reactor.

[0023] 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.

[0024] As illustrated in [Fig.2], from publication [1], the primary circuit 2 consists of the following main components:

[0025] - a reactor vessel 20,

[0026] - primary loops 21 each comprising a primary pump 22 and a steam generator 23,

[0027] - a single pressurizer 24.

[0028] In addition, the reactor core control rod mechanisms and control rod clusters 25 are distinguished on this [Fig.2].

[0029] 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.

[0030] The reactor building 1 is therefore sized, among other things, to house all the components of the primary circuit 2.

[0031] 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 reactor building 1, which provides the function of a third containment barrier, can be seen in particular.

[0032] 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.

[0033] More specifically, for a given steam generator 23, the reactor building 1 is crossed by a line called hot line 30 which evacuates the steam from the steam generator 23 to evacuate the power and bring it to the turbine 32, and by a line called cold line 31 which supplies liquid water to the steam generator 23.

[0034] Among the already studied solutions for heat-producing PWR reactors, three main categories can be distinguished, corresponding to the main design architectures of the primary circuit: namely, respectively, the so-called "pool" type reactors, those with a primary loop(s), and those of the integrated type, generally illustrated in small-power reactors, designated by the Anglo-Saxon acronym SMR. (Small Modular Reactor), intended for power generation. See publications [2] and [3].

[0035] Pool reactors implemented to date have been used experimentally with low power outputs, typically 10 MWth: the pressure in the primary circuit can be close to atmospheric pressure, resulting in both moderate neutron fluences within the core, and the primary circuit temperature is limited to a maximum of approximately 100°C. In a pool reactor, the height of liquid water above the core allows for a slight increase in the primary pressure within the fissile core, while remaining on the order of a few bars. The advantage of such a reactor lies in its simple design, as the reactor vessel is not considered a pressure vessel; the concrete reactor vessel well and its sealed internal lining form the reactor vessel surrounding the primary circuit.Therefore, it is primarily the maintenance of radiological containment, as in the case of a spent fuel pool, that governs the design quality of this "sandwich" component, consisting of the concrete reactor vessel shaft and its internal leak-tight lining. The reactor vessel itself is assembled on-site and must, of course, withstand all extreme external stresses, including major earthquakes and aircraft crashes. Furthermore, the concrete must be subjected to temperatures compatible with maintaining its mechanical properties over time. This type of design is no longer favored due to the stringent requirements for demonstrating quality of construction and on-site inspection for the second containment barrier of a nuclear reactor, rules established by the RCC-M, the French code defining the design and construction rules for the mechanical components of the nuclear islands of PWR reactors.A self-supporting metal tank, manufactured in a factory, is preferred, and also allows for higher primary pressures.

[0036] An example of a pool reactor is the SLOWPOKE project conducted in the 1970s and 1980s by Canada (AECL) concerning NHP reactors (Nuclear Heating Plants). The figure on page 13 of publication [3] illustrates this demonstration reactor, with a thermal output of 10 MWth, whose water-filled reactor vessel contains the entire primary circuit and the heat exchangers between the primary and secondary circuits, all enclosed by a slab at ground level. In this pool reactor, the primary fluid circulates by natural convection, thus simplifying electrical loss transients and general system maintenance. The reactor also has significant thermal inertia, due to the large water inventory of the primary circuit relative to the reactor power, which also improves overall safety with regard to incidental transients, and the... Finally, the thickness of the reactor vessel well and its lack of lateral or bottom penetrations ensure by design that the primary circuit cannot be drained or depressurized. Such a pooled reactor therefore meets many criteria for design simplicity, design safety, and ease of operation. However, it presents a major drawback: the manufacturing quality of the casing and the maintenance of the performance of the concrete forming the reactor vessel well are difficult to demonstrate, respectively, initially and over time.

[0037] A more recent example, the Chinese DHR 400 project by CNNC, appears to demonstrate its reproducibility, since the reactor vessel consists of a thick-walled, prestressed concrete sandwich component, typically on the order of one meter, partially lined internally with a 5 mm thick stainless steel liner, all enclosed within a 10 mm thick carbon steel outer cylinder. However, this reactor can only be reasonably considered if the predominant criteria of modularity and maximizing component manufacturing in the factory, with transport to the site, are met.

[0038] An example of a pool reactor, with a traditional reactor vessel well and associated casing, is presented in the Russian RUTA-70 reactor project of the NIKIET company.

[0039] An example of a heat-generating loop reactor is the Chinese HAPPY 200 reactor from SPIC, dedicated to district heating in Beijing. With a unit capacity of 200 MWth, it is designed in sets of two units, thus matching the performance of the aforementioned CNNC DHR400 project. Unlike the latter, the reactor vessel is a self-supporting steel structure, manufactured in a factory and assembled on site, although the primary loop is fully welded, requiring extensive construction work. The thermal power is removed from the core via two primary loops supplying plate heat exchangers by forced convection using two pumps. A water pool surrounds the entire reactor vessel, but without direct contact due to a double jacket. In the event of an accidental transient, this double jacket is flooded by the cold water in the pool.This looped reactor configuration thus offers the advantage of being able to provide heat at a temperature relatively close to that exiting the core, thanks to forced convection in the primary circuit, which facilitates the extraction of thermal power. Furthermore, the presence of plate heat exchangers between the primary and secondary circuits allows for a low temperature gradient between the primary and secondary circuits. The major drawbacks of this looped configuration lie in the forced convection in the primary circuit, due to the complexity of studying transient power loss, and the maintenance costs associated with its operation. Pumps. Another significant drawback is the factory-built nature of the loops and the reduction in on-site assembly phases (construction time). Furthermore, in terms of compactness, simplicity of civil engineering, and reduction of volumes in controlled and monitored areas, a loop configuration is not optimal due to the extent of the primary circuit within the reactor building. Finally, a loop configuration carries a risk of Loss of Coolant Accident (LOA).

[0040] The last category is that of so-called integrated reactors, which comprise a block delimited by a reactor vessel entirely manufactured in a factory and transported to the site, and which houses the entire primary circuit, including the heat exchangers between the primary and secondary circuits. This type of integrated reactor has the same configuration as the main concepts of currently existing SMR reactors with a power-generating function, namely a configuration based on the integration of the steam generator, or even all the components of the primary circuit, including the pressurizer and the primary pumps, inside the reactor vessel. These SMRs are called integrated SMRs.

[0041] SMR reactors have the primary advantages over existing PWRs of allowing simplification of systems, mainly for safety purposes, and increased modularity capacity through significant manufacturing of components in the factory for transport to the construction site.

[0042] In addition to the gain in compactness, integrated SMRs have the advantage of eliminating the need for overhead pressurized water lines, except for the control circuit lines, which have very small diameters, typically a few centimeters. This considerably reduces the risk of accidents and associated consequences related to primary circuit line ruptures (APRP-type accidents). Thus, on-site installation is greatly simplified by being limited to secondary piping connections, apart from the branch connections for the volumetric and chemical systems of the primary circuit, which are of small diameters.

[0043] By way of example, the nuclear power plant project with the acronym NUWARD™, is a power plant intended for power generation consisting of two integrated SMRs, with a unit power of 170MWe, each of which includes a block housing all the components of the primary circuit inside the reactor vessel.

[0044] Other integrated SMR projects with a power generation purpose 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 of the Chinese manufacturer CNNC with a power of 100 MWe.

[0045] Figure 4 shows an example of an integrated SMR currently under development. Such an integrated SMR reactor, the block of which is generally designated by the reference digital 4 includes a fixed compartment 40 and a removable compartment 41 in the form of a cover, for fuel handling or reactor internals maintenance phases.

[0046] A pioneering concept for an integrated heat-generating reactor is the THERMOS project, jointly conducted by the Applicant and Technicatome: [4]. The reactor under this project had a thermal output of 100 MWth and was intended to supply district heating to the city of Grenoble. According to this project, the reactor vessel integrated the entire primary circuit, thus enabling operation at a pressure higher than that of a tank reactor, necessary to provide district heating of approximately 120°C. With a diameter of 5 m and a height of 9 m, the proposed reactor vessel was fully factory-assembled and housed, in particular, the heat exchangers between the primary and secondary circuits in its upper section. The low temperature gradient in the core necessitated the presence of primary fluid pumping units, arranged in the hot section, which is not ideal in terms of safety and ease of maintenance.Furthermore, the thermal inertia of the primary and secondary circuits is relatively limited, as the basin in which the reactor is immersed is thermally disconnected from the normal operation of the reactor, which impairs operational safety and the smoothing of power demand transients in the customer's heating network.

[0047] Recent studies have been dedicated to heated reactors in Finland, conducted primarily by VTT, for the district heating of the city of Helsinki: [5]. In particular, a 50 MWTh integrated vessel reactor was studied, containing the heat exchangers between the primary and secondary circuits in its upper part, the reactor vessel itself being contained within a nearby containment structure. The primary fluid circulates by natural convection during nominal operation, and the secondary circuit is a closed-loop liquid circuit with a heat exchanger connected to the tertiary circuit, which operates by forced convection using a pump. The entire vessel and containment structure are immersed in a pool, which is the fuel handling pool (IRWST, an acronym for "In-containment Refueling Water Storage Tank").More specifically, the IRWST pool comprises a pit forming the reactor vessel well, within which the reactor vessel and containment structure are partially submerged. This IRWST pool also serves as a cold source for design accidents. This type of integrated reactor offers the same advantages of factory buildability and modularity as integrated SMRs with power generation capabilities, notably those from NuScale Power. However, it has a major drawback: a lack of thermal inertia in the secondary circuit, inertia which is necessary for stable operation. in response to a load demand. Furthermore, the permanent thermal leaks linked to the IRWST pool configuration, which promotes thermal conduction through the reactor vessel, lead to the need to constantly cool said pool, resulting in unexploited waste heat.

[0048] In summary, integrated SMR type reactors appear to be the best candidates as heat-producing reactors, insofar as they allow both optimal safety (due to the simplification of systems and the elimination of overhead fluid lines) and increased modularity capacity allowing the manufacture of components in the factory and their assembly on site.

[0049] However, the inventors have analyzed that the integrated SMR type reactors that have been envisaged or as currently envisaged have several disadvantages.

[0050] First, the architecture of the safety systems used to maintain the primary circuit water inventory to prevent the reactor core from uncovering during an accident presents some weaknesses. These systems utilize dedicated safety injection water reservoirs that are directly connected to the reactor vessel via fluid lines. These lines are susceptible to rupture, which would effectively lead to an APRP (Accident-Resistant Precipitation Risk).

[0051] A simplification of the safety architecture to make a heat-producing SMR very robust to all accidental events would be desirable insofar as such a reactor will have to be located in areas close to the place to be supplied with heat.

[0052] Furthermore, the SMR reactors intended for heat generation planned to date in the form of integrated SMRs, with secondary and tertiary circulation loops to send the heat produced to a network, have insufficient flexibility insofar as the operating conditions of the primary circuit vary greatly during variations in power demands, which makes the operation of such reactors complex.

[0053] Moreover, heat exchangers between primary and secondary circuits as designed to date present integration difficulties and manufacturing complexity which do not sufficiently facilitate the construction and maintenance of the reactor.

[0054] Finally, the limitation of heat losses is not optimal and relies on thermal insulation arranged around the reactor vessel. It would be necessary to manage heat leaks through the vessel, and to provide temperature protection for the reactor vessel shaft and, more generally, for the concrete surrounding the Pilot nuclear reactor.

[0055] There is therefore a need to improve integrated SMR type reactors, particularly when considered as heat-producing reactors, in order to overcome the disadvantages mentioned above.

[0056] There is therefore a need for a nuclear installation with SMR-type reactor(s) for a heat-generating purpose which offers both: - an optimized level of safety, - a level of flexibility and adaptability to the needs of the network(s), - simplicity of construction, - ease of installation and removal (minimizing the number of connecting fluid lines), - ease of operation.

[0057] The aim of the invention is therefore to meet at least part of this need(s). Description of the invention

[0058] To this end, the invention relates, in one of its aspects, to a nuclear installation comprising:

[0059] - at least one SMR reactor block for heat generation delimited by a tank of reactor;

[0060] - a tank well delimiting a water-filled pool forming part of a secondary circuit, in which the reactor vessel is immersed;

[0061] - at least one heat exchanger between the primary circuit of the SMR reactor and the secondary circuit, arranged outside the reactor vessel and in the secondary water pool.

[0062] Advantageously, the water pressure of the secondary circuit in the basin is greater than or equal to that prevailing in the reactor vessel.

[0063] Advantageously, the water pressure in the basin is less than or equal to 20 bar.

[0064] According to an advantageous operating mode, the SMR reactor is configured to, when in operation, circulate the primary circuit water by natural convection between the exchanger(s) and the inside of the reactor vessel by passing through the fuel assemblies defining the reactor core.

[0065] According to an advantageous embodiment, the water basin is configured so that, when the SMR reactor is in 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 and the exchanger(s) between primary and secondary circuits are immersed and the top of the basin at a so-called hot temperature, the water in the basin circulating by natural convection between the exchanger(s) and the thermocline.

[0066] Preferably the cold water comes from at least one heat exchanger between the secondary circuit and the tertiary circuit.

[0067] According to an advantageous construction variant, the installation further comprises an enclosure arranged around the reactor vessel and adapted to guide a flow of water from the basin around the reactor vessel, by natural convection.

[0068] Preferably, the level of the thermocline is fixed so as to be above the reactor vessel.

[0069] According to an advantageous configuration, the heat exchanger is fixed outside the reactor vessel, the reactor vessel having inlet and outlet openings leading respectively into the inlet and outlet manifold of the part of the primary circuit in the exchanger.

[0070] According to an advantageous arrangement, the heat exchanger includes an inlet manifold and an outlet manifold for the water from the basin, the level of the thermocline being fixed so as to be above the outlet of the outlet manifold.

[0071] According to an advantageous embodiment, the installation includes a fuel assembly reloading pool, intended to be inserted into the core of the SMR reactor, the reactor vessel well being arranged below the bottom of the pool and being closed by a removable metal cover forming the separation wall with the bottom of the pool.

[0072] According to this method, and an advantageous embodiment, the installation includes at least one heat exchanger suspended by the removable cover of the tank well, adapted to exchange heat between a closed circuit of water from the bottom of the pool and the water from the secondary circuit of the pool, the water from the bottom of the pool flowing by gravity into the suspended exchanger and rising by natural convection to join the pool.

[0073] According to an advantageous embodiment, the reactor block comprises a plate with holes, a part of which individually houses a control rod for reactivity control bars, and at least one hole of which houses a valve for regulating the flow of primary water circulating in the reactor.

[0074] Advantageously, the structure delimiting the water basin includes a bottom configured to support the reactor vessel.

[0075] The SMR reactor designed for heat generation is preferably intended to be connected to a heat network

[0076] Thus, the invention essentially consists of immersing an SMR type reactor block and the heat exchangers between the primary and secondary circuits in a liquid water basin which is part of the secondary circuit.

[0077] To date, an integrated SMR reactor involves a block with exchangers between primary and secondary circuits inside the reactor vessel, which makes the assembly and mounting phases more complex, and also complicates the overall sizing of the vessel and the exchanger components themselves.

[0078] This classic design is explained by the interest in eliminating any possibility of a major primary breach compared to a conventional primary loop PWR reactor, essentially for applications where the reactor is intended for power generation and for which the primary circuit is subjected to significant pressures, on the order of 150 bars or even higher.

[0079] However, to address the issues of construction, maintenance, and safety, the inventors analyzed that by considering a heat-generating reactor with a secondary water circuit comprising a secondary water basin contained within the reactor vessel shaft, which forms the third containment barrier, and with a primary circuit subjected to a pressure of a few bars, typically 3 to 6 bar, the risks of primary circuit breach were significantly reduced, and any risk of core uncovering was eliminated in the event of a breach. Indeed, in such a case, the secondary circuit can be close to the primary pressure, thus eliminating any significant loss of primary inventory in the event of a breach.

[0080] The inventors also judiciously considered arranging the heat exchangers between the primary and secondary circuits outside the reactor vessel. This makes it possible to overcome the aforementioned integration constraint of known integrated SMR reactors, a constraint which complicates the assembly and mounting phases, and also complicates the overall sizing of the reactor vessel and the heat exchanger components.

[0081] This secondary water basin surrounding the reactor vessel is advantageously at a slightly higher pressure than the vessel itself to prevent the primary circuit flow from escaping in the event of a breach. The secondary water basin thus constitutes a passive barrier for radiological containment.

[0082] Responding in fact to the need for flexibility, one advantage is that the secondary water basin is used as a buffer storage basin, offering high thermal inertia and allowing to absorb a large part of the power variations on the heat network in which the SMR reactor for heat generation is connected.

[0083] Also addressing the need for simplicity of intervention and operation, a subsequent advantage of a secondary water basin, which does not require connecting / disconnecting secondary water circulation piping, lies in the possibility of easily replacing or moving the entire reactor block, particularly for refueling phases. During these phases, the reactor block can be completely moved, greatly facilitating the handling of fuel assemblies, notably by avoiding the need for long handling poles, as is the current state of the art, and the associated risks of jamming or falling. In the event of a major failure of reactor block components, it is also possible to easily replace the reactor vessel and / or the heat exchangers. Primary and secondary circuits are used to minimize downtime. Also addressing safety requirements in the event of a severe external event such as an earthquake, the absence of primary or secondary piping connections between the reactor vessel and the installation greatly simplifies accident studies associated with this type of accident. Indeed, in this configuration, the reactor vessel is fixed to the base of the reactor vessel well, freely immersed in the secondary water basin, without fixed fluid transport connections to dissipate heat.

[0084] Advantageously, from a safety and thermal efficiency standpoint, a vertical thermal stratification is established in the secondary basin to form a thermocline, with the hot water at the top of the basin and the cold water at the bottom. This cold water advantageously comes from at least one heat exchanger between the secondary and tertiary circuits.

[0085] Secondary water circulation occurs by natural convection between the heat exchangers immersed in the basin and the thermocline.

[0086] This natural convection can be enhanced by the presence of a casing arranged around the reactor vessel, which guides the flow of water from the basin by natural convection around it. The casing can be positioned a few centimeters around the reactor vessel. This flow allows for the exchange of thermal power between the primary and secondary circuits through the reactor vessel. This additional exchange surface area makes it possible to reduce the size of the immersed heat exchangers and to utilize the heat losses through the reactor vessel. The casing guiding the flow ensures that the thermal power thus exchanged through the reactor vessel does not destabilize the thermocline of the basin.

[0087] In an accident situation, the residual heat can be dissipated by natural convection through the heat exchangers between the primary and secondary circuits immersed in the basin and through the reactor vessel. The design of the safety systems dedicated to core cooling is thus greatly simplified compared to the state of the art, due to the sole requirement of cooling the volume of water in the secondary circuit. Because of the potential for significant thermal inertia from the volume of secondary water, which can be substantial—typically several days before any external intervention—dissipating the residual heat from the reactor core is facilitated and simplified both through the heat exchangers used for normal operation and directly by conduction through the secondary water via the reactor vessel walls.In addition to the evacuation of residual power through the primary exchangers, the jacket guiding the flow of water around the reactor vessel also allows, in accidental conditions. In extreme cases, the core is cooled by direct thermal conduction with the vessel using the secondary volume of cold water surrounding the vessel. This ultimate power dissipation method compensates for any potential common-mode failure related to heat dissipation through the heat exchangers, by judiciously utilizing the reactor vessel's immersion in the secondary cold source.

[0088] In order to properly balance the flow rates by natural convection in primary and secondary water, shunting valves are provided and judiciously arranged.

[0089] The primary water throttling valve is advantageously integrated within the reactor vessel, and its control mechanism is linked to the reactivity control rod piloting system. The primary water flow rate is thus dependent on the neutron power level and therefore the thermal level of the core. More specifically, adjusting the height of the throttling valve's position allows for regulating the primary water flow rate and thereby the heat exchange between the primary and secondary circuits, using rod followers and mechanical rod positioning systems identical to those used for reactivity control. Preferably, the passive safety system allows for the simultaneous release of all the reactivity control rods to ensure their gravity descent into the core and the throttling valve.In its lowest position, the primary water flow opening is at its maximum, thus facilitating the evacuation of residual power to the secondary water volume.

[0090] Furthermore, the reactor's operation by natural convection in primary and secondary water meets the need for adaptability to the customer's district heating network, as it makes it inherently stable with respect to the thermal power demands of the district heating network for which the installation is intended. This contributes to operational reliability and, in fact, simplifies control and management constraints. For example, in the event of a shutdown or a sudden drop in thermal power draw from the secondary water volume, the supply of secondary water at its hot temperature through the heat exchangers between the primary and secondary circuits quickly leads to a thickening of the layer of secondary water at its hot temperature relative to the layer of water at its cold temperature surrounding the reactor block. This results in a lower thermocline altitude, consequently causing a decrease in flow rate by natural convection in the heat exchangers on the secondary side.At constant nuclear power output, the heat exchange between the primary and secondary circuits decreases, which in turn leads to a rise in the average temperature of the primary fluid. From a safety perspective, the neutron feedback induced by the expansion of the primary fluid in the core, on the one hand, and in the nuclear fuel, on the other, must be considered. The Doppler effect, on the other hand, generates an immediate decrease in the heart's reactivity, thus counteracting the rise in the heart's average temperature.

[0091] There is therefore a stable natural or passive reaction of reduction of the neutron power of the core, without modification of the position of the reactivity control bars, during a reduction of the supply of thermal energy to the customer network, guaranteeing safety and security of operation.

[0092] In addition to the core power adjustment by the control rods on the one hand, the adjustments by primary and secondary throttling valves as explained previously, and by the adjustment of the primary water pressure using the pressurizer on the other hand, the safety advantage is to be able to operate this heat-generating reactor in a passive and simplified manner, eliminating all of the following accidental transients, which are usually encountered in pressurized water reactors intended for power generation according to the state of the art, namely:

[0093] - a Primary Refrigerant Loss Accident due to the impossibility of significant loss of the primary water inventory and therefore the risk of core dewatering

[0094] - a Water Pipe Rupture and a Steam Pipe Rupture, due to the absence of steam generation and a secondary water supply by direct immersion,

[0095] - a Reactivity Accident due to clear water dilution, since the reactor is operating without primary soluble boron,

[0096] - a primary flow loss, due to a primary natural circulation,

[0097] - a secondary flow loss, due to a secondary natural circulation,

[0098] - an external electrical network loss, due to the passive nature of operation nominal.

[0099] Generally, supplying heat at around 90 to 110°C for a district heating network leads to sizing the primary circuit's cold and hot temperatures between 110 and 140°C, and slightly lower for the secondary circuit's hot temperature. The return to cold standstill of the primary circuit and to atmospheric pressure is therefore relatively rapid. Similarly, the pressure of the secondary water volume is also limited to a few bars, and a sudden return to atmospheric pressure, for example due to a pipe rupture or the wellhead cap failure, cannot lead to massive boiling of the secondary water and a significant loss of water inventory.

[0100] As can be seen from the above, the average temperature of the secondary water, which is the equilibrium temperature of the volume at its cold temperature and of that at its hot temperature, is less than 100°C, and therefore cannot spontaneously boil in the event of sudden depressurization, thus preventing the risk of uncovering the reactor core.

[0101] The invention can be implemented with secondary water pressures in the basin up to about 12 bars, provided that the walls of this basin are appropriately sized and in certain specific secondary volume confinement configurations explained later.

[0102] Addressing the need for modularity and factory construction, compared to a conventional installation with a state-of-the-art PWR reactor and a pool, the solution according to the invention has the advantage of a primary circuit design manufactured in the factory (SMR reactor) and delivered to the site, for improved mechanical implementation in accordance with nuclear equipment construction rules (RCCM). The structure delimiting the secondary pool according to the invention can be built either on-site or also manufactured in the factory and transported to the site, with fewer manufacturing and control constraints compared to those of the primary circuit.

[0103] With the aim of further enhancing safety, an advantageous configuration involves locating the reactor vessel well below the bottom of a fuel assembly refueling pool, where the fuel assemblies are to be inserted into the core of the SMR reactor. The vessel well is closed by a removable metal cover forming the separation wall with the bottom of the pool. This configuration is designed to allow the handling of fuel assemblies at the bottom of the pool with the reactor vessel open. The water depth of the IRWST pool, approximately 8 to 10 meters, ensures biological protection during the handling of the fuel assemblies, which requires dismantling the vessel to access the reactor core and storing the removable portion of the vessel 41, 45 underwater in the pool.

[0104] This configuration is also provided for in cases where resistance to secondary water pressure at its maximum temperature can no longer be ensured solely by the prestressed concrete wall covered with a metallic liner forming the tank well.

[0105] This configuration allows a natural flow by gravity of part of the water present in this fuel refueling pool, to achieve the cooling of the secondary metallic tank detached from the tank well and itself ensuring resistance to secondary pressure.

[0106] In normal reactor operation, a small flow of water from the pool naturally descends along a self-supporting metal tank inside the secondary tank well, and returns to the basin once heated, by natural convection.

[0107] In the event of a reactor incident transient, such as an electrical failure preventing, for example, the normal discharge of the nuclear power produced, a greater flow rate of water from the pool can be released, which maximizes the cooling of the thick-walled vessel, and therefore the volume of secondary water contained within it. The primary fluid, the reactor core, is then also cooled by the volume of secondary water surrounding it, both by direct conduction through the reactor vessel, but also by the exchangers between the primary and secondary circuits which continue to operate in their heat exchange function.

[0108] The installation as planned with its heat-generating reactor is intended to supply urban heating networks, industrial processes such as desalination, drying, and food processing.

[0109] Advantageously, a supplementary power generation unit can also be added, using an organic Rankine cycle, to obtain local emergency power generation in case of an ultimate need to supplement conventional battery bank solutions generally used. For reasons of simplicity of design and maintenance, the implementation of classified active backup means, such as diesel combustion systems, is not planned.

[0110] 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

[0111] [Fig. 1] [Fig. 1] is a schematic perspective and partial cross-sectional view of an existing PWR type nuclear reactor.

[0112] [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.

[0113] [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.

[0114] [Fig.4] [Fig.4] is a schematic perspective view of an SMR of type integrated as it is currently envisaged.

[0115] [Fig. 5] [Fig. 5] is a schematic perspective view of an SMR reactor calogen according to the invention.

[0116] [Fig.5A] [Fig.5A] is an exploded perspective view of the parts of the SMR calogen reactor vessel according to the invention.

[0117] [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.

[0118] [Fig.6A] [Fig.6A] is a schematic detail view of a control cluster assembly, control rod, and rod control mechanism of a calogen SMR reactor according to the invention.

[0119] [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.

[0120] [Fig.9] [Fig.9] is a perspective view of part of the reactor according to [Fig.5], and which shows in detail the installation of a water flow control valve for the secondary circuit.

[0121] [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.

[0122] [Fig. 11] [Fig. 11] is a partial longitudinal cross-sectional view of the upper part of the reactor according to figures 5 and 6 showing a pressurizer according to the invention.

[0123] [Fig. 12] [Fig. 12] is a schematic longitudinal and transparent cross-sectional view of a nuclear installation according to the invention with a heat-generating reactor according to figures 5 and 6 in a sealed vessel shaft.

[0124] [Fig. 13] [Fig. 13] is a schematic representation of the primary circuits in natural convection, secondary in natural and forced convection and tertiary in forced convection of a nuclear installation with a reactor according to figures 5 and 6.

[0125] [Fig. 14] [Fig. 14] is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to figures 5 and 6 in a sealed vessel well and the various components (piping, heat exchangers, pumps, valves, hydraulic distribution block) of the primary, secondary and tertiary circuits.

[0126] [Fig. 15] [Fig. 15] includes the hydraulic distribution block with the two exchangers between the secondary and tertiary circuits which can be connected in series or in parallel fluidic.

[0127] [Fig. 16] [Fig. 16] is a perspective view showing a variant of the hydraulic components of [Fig. 14].

[0128] [Fig. 17] [Fig. 17] is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to figures 5 and 6 in a sealed vessel shaft, arranged on the bottom of a handling pool (IRWST) whose water constitutes the safety cold source.

[0129] [Fig.17A] [Fig.17A] is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to Figures 5 and 6, during the handling phase, with the reactor vessel well open, and with the top of the reactor block dismantled and moved into an IRWST pool storage station.

[0130] [Fig. 18] [Fig. 18] is a schematic and transparent view of a nuclear installation according to the invention with a heat-generating reactor according to figures 5 and 6, comprising two redundant heat exchangers dedicated to the evacuation of residual power, which are fixed to the upper wall of the reactor vessel shaft.

[0131] [Fig.19] [Fig.19] is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to figures 5 and 6, with a metallic vessel well cover integrating two heat exchangers also dedicated to the evacuation of residual power, in the form of cooling panels immersed in the secondary water volume at its hot temperature.

[0132] [Fig.20] [Fig.20] is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to Figures 5 and 6, comprising a thick metal vessel filled with secondary water, which is decoupled from the vessel well, and is dedicated to applications at higher secondary water pressure and temperature.

[0133] [Fig.20A] [Fig.20A] is a schematic view of the thick metal tank filled with secondary water according to [Fig.20]. Detailed description

[0134] 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 whose reactor vessel shaft is arranged at least in part within a pool, in particular excavated below, and which can serve as an ultimate cold source and biological protection during the handling phases according to the invention.

[0135] By "primary water", "secondary water", "tertiary water", we mean the water which constitutes the fluid respectively of the primary, secondary and tertiary circuit.

[0136] Figures 1 to 4 have already been detailed in the preamble, so they will not be commented on below.

[0137] It should be noted that the various temperatures, power outputs, volumes, flow rates, etc., indicated are for guidance purposes only. For example, other temperatures may be considered depending on the configuration, in particular the power output of the SMR reactor, the volume of water in the secondary basin, and the power requirements of the heating network.

[0138] With reference to Figures 5 and 6, a pressurized water type nuclear reactor 4 is described, according to an integrated SMR type primary circuit configuration according to an embodiment of the invention.

[0139] This reactor 4 has a unit power of 20 MW thermal, with a heat-generating purpose, i.e., dedicated to the supply of hot water, typically at 90°C. Its power unit capacity can, however, vary upwards or downwards, in a range of approximately 10 MW to 100 MW, and the hot water supply temperature can also vary depending on the customer's heat network.

[0140] 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 bodies 41 and 45 is removable to allow handling of the assemblies contained in the reactor vessel 40. The removable compartment 41 is a dome-shaped cover 6, the central chimney of which incorporates a valve 64, adapted for cooling the reactor pressurizer as detailed below.

[0141] 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.

[0142] The vessel body 40 houses a cylinder 43, supporting an assembly basket usually referred to as the "core support basket", dedicated to holding the fuel assemblies 42, and a separation envelope 40 with its peripheral neutron reflector 440 intended to ensure the maintenance of the neutron flux in the core.

[0143] A set of flanges is bolted between the fixed compartments 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 set 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 is intended to provide access to the upper internals of the core, the control rod mechanisms, and to the flow rolling ring. The pressurizer associated with compartment 41 and its internal elements can also thus be separated from the rest of the tank block for intervention and maintenance.

[0144] The studies carried out by the Applicant provide for scheduled fuel refueling stops during ten-year outages, without intervention on the core between these periods.

[0145] 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.

[0146] The free volume above the reactor core C allows the control rods 42 to be fully extended, as well as the standby position of so-called emergency absorber 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.

[0147] 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 rolling valve 481.

[0148] A flow control valve 481 for the primary circuit water is arranged around the periphery of the plate 48. This valve 481 is in the form of a flow control ring 481 which follows the inner periphery of the compartment 45 of the tank and extends over a height sufficient to cover the primary water outlet openings, thus allowing the flow rate of the primary water to be regulated.

[0149] 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.

[0150] 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.

[0151] In the event of a power failure or emergency stop activation, the gravitational fall of the control rods 46 also triggers the gravitational fall 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, in order to remove residual heat and cool the primary circuit. This gravity-fed operation of the control valve ensures reliability and safety in the event of a power outage or reactor failure.

[0152] 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 section. There, it flows through the various outlet openings 400 corresponding to the inlet manifolds of the heat exchangers 49 between the primary and secondary circuits and into a central upper portion of the core, in the form of a riser. This central riser, not detailed, contains, in addition to the control rod control mechanisms, the core parameter instrumentation sensors.

[0153] 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 on the outside of 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.

[0154] Above the outlet openings 400, within the reactor 4, a separation plate 7 separates the riser from the interior of the dome 41 of the vessel containing a pressurizer. 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.

[0155] The upper part incorporating the reactor pressurizer will be detailed later with reference to [Fig.11].

[0156] 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 median elevation of the heat exchangers 49 and the average height of the fissile zone of the core, defined by the core C.

[0157] As already mentioned, the adjustment of primary pressure losses and therefore the flow rate is achieved by the throttling valve 481 whose control rods are in position 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 system, it is the primary water flow control 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. Therefore, the operation of the heat reactor can be easily controlled by adjusting the core power using all the control rods 42, in addition to this primary water pressure setting.

[0158] 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.

[0159] Within an exchanger 49, the flow is downward for primary water, and upward for secondary water.

[0160] 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. 12]. 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.

[0161] This secondary circuit with liquid water basin B is an open medium delimited by the tank well, without a circulation pump.

[0162] 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 resulting in the loss of primary water, 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.

[0163] The internal circuit within a heat exchanger 49, which is part of the secondary circuit of reactor 4, therefore carries a flow of liquid water as a secondary fluid. This water is heated by 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 thus creates a volume greater than that at a so-called hot temperature. The separation layer between the 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 Figures 6, 11, and 12.

[0164] 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 cooling flow rate of the secondary circuit through the heat exchangers 49. The closed-loop circulation S, solely by natural convection of the secondary water, is symbolized by the gray arrows in [Fig. 6].

[0165] 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.

[0166] 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.

[0167] 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.

[0168] 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 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 power demand The heat flow to the primary circuit is then at its maximum, and the average temperature of the primary water decreases. This decrease in primary water temperature leads to average cooling of the moderator in the core, consequently increasing core reactivity and therefore its thermal power. Maximum heating conditions of the secondary water volume are accompanied by a natural increase in core power, so reactor 4 is naturally stable. As previously mentioned, the position of the primary water throttling valve 480, combined with the reactivity control rod positions 42, allows the increase in core reactivity to be limited, keeping it within the temperature rise range of the entire reactor block 4 and its reactor vessel well.

[0169] 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. Consequently, the circulation of secondary water by natural convection decreases, thereby reducing the 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. There is thus a de facto 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 thermal 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 establishment 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.

[0170] 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, for a total volume for basin B of between 300 and 450 m³. Typically, an altitude difference of approximately 4 meters between the median plane of the heat exchangers 49 and the position of the secondary thermocline separating a layer of cold secondary water at 65°C from the layer of hot water at 105°C allows for the establishment of natural circulation of a cold secondary fluid at 65°C at the inlet of heat exchanger 490, and a hot secondary fluid at 105°C at the outlet. of exchanger 491, generating approximately 1000 Pascals of driving pressure required to overcome the pressure losses due to the passage through the exchangers 49, from the inlet 490 to the outlet 491, including the pressure losses adjustable using the secondary flow control valves 5.

[0171] The thermocline position can only be maintained at a fixed level 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 transfer, or an unplanned 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 continuous removal of the thermal power produced by the reactor, without an external escape route, necessitates shutting down the reactor to remove only the residual power through specifically dedicated residual power removal systems. Typically, a continuous thermal output of 20 MW, with a tertiary water supply at 90°C and a return at 45°C via the customer's heating network, requires pumping 123 liters per second, or 442 m³ per hour, from the hot water layer at 105°C above the secondary thermocline and returning the same quantity to the bottom of the reactor vessel at 65°C.Preferably, this discharge and return can be implemented using piping from the top of the reservoir well, in order to avoid lateral connections that could cause leaks or lateral mechanical strength issues that would restrict expansion or seismic resistance. The heat transfer of 20 MW from the pumped secondary water layer to the customer's tertiary water circuit is carried out using one or more heat exchangers sized to transfer 20 MW with a hot pinch of 15°C (from 105°C to 90°C) and a cold pinch 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 case of failure or the need for intervention.

[0172] 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.

[0173] As previously stated, the reactor's primary circuit operates solely by natural convection, i.e. without a pumping unit.

[0174] 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.

[0175] 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.

[0176] 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.

[0177] In normal operation, the thermocline level is fixed sufficiently above the pressurizer, in particular so as to be located above the central exhaust stack 62, as illustrated in Figures 11 and 12.

[0178] Thus, 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 steam of the primary circuit inside the tank and thus reduce the pressure within the tank. Typically, the cold temperature of the liquid water entering 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 steam underlying it, to a primary pressure saturation temperature of 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 the condensation of 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.

[0179] The central chimney 62 incorporates a control valve 64 which regulates the flow rate of secondary liquid water circulating in space E and thus regulates the liquid cooling itself. 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 and will join 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.

[0180] Valve 64 can be a butterfly valve like secondary flow valve 5 illustrated in Figures 10A, 10B, 10C.

[0181] The two walls 60, 61 of the dome 6 are metallic, preferably stainless steel.

[0182] 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.

[0183] 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 its surface, preferably by welding or brazing. These fins 65 thus increase the contact area with the steam of the primary circuit and therefore improve heat exchange by conduction between said steam and the dome 6. 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 typically have a thickness of a few centimeters and a length of a few tens of centimeters along the inside of the wall 60.

[0184] 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.

[0185] The electrical resistances 8 may be of the type described in US patent 4135552.

[0186] As shown in [Fig. 12], the reactor vessel 40, 41, 45 of reactor block 4 with the heat exchangers 49 is supported by a metal base 102, preferably made of stainless steel or black steel coated with an anti-corrosion coating. Preferably, this base is held in the bottom of the reactor vessel well 100 by a mechanical locking system, not shown, adapted to prevent its displacement and uplift in the event of an earthquake. However, during fuel handling or component replacement, the base 102 and the reactor block 4 it supports can be lifted and brought to the top of the reactor vessel well. To do this, the mechanical locking system of the base must be able to be easily unlocked by tools accessible from the top of the reactor vessel well 100.

[0187] The base 102, which supports reactor block 4, is rigidly connected to a foundation slab 103, and the metallic cladding, with which the prestressed concrete wall 101 is coated, is rigidly and hermetically fixed to the slab 103. As already mentioned, reactor 4 has a completely physical impossibility of a serious accident occurring, with significant core meltdown and primary circuit vessel breach. Therefore, there is no specific corium recovery device at the bottom of the reactor vessel shaft, as the corium is confined in all conceivable physical configurations inside the reactor vessel, permanently immersed in a liquid.

[0188] The tank well 100 is closed by a removable and watertight cover plug 104, adapted to withstand the pressure of the secondary circuit. Preferably, the cover plug 104 is a welded metal slab, preferably honeycomb or formed of cavities.

[0189] A gas space 105, preferably nitrogen, is defined by the free level of the secondary water volume in the tank well 100. This gas space 105 allows control of the secondary water pressure and adaptation to variations in free level and expansion of the secondary water. The mechanical connection between the metallic lining of the tank well and the cover plug is formed by a metallic gasket or a high-temperature elastic gasket, in order to maintain a complete seal.

[0190] If necessary, as a cooling system for the tank well 100, a pipe not shown with water circulating within it can be embedded in the concrete wall 101, at a distance close to the inside and the metal lining forming the third barrier.

[0191] Reactor block 4, 102 has only instrumentation and core reactivity control connections, in the form of easily removable electrical connectors. An additional flexible or semi-rigid connection, also removable, allows for the continuous sampling and return of a small fraction of the primary circuit water for chemical treatment and adjustment of the total primary water volume. Thus, reactor block 4, 102 is easy to disconnect and relocate for potential complete replacement.

[0192] For the implementation of the tertiary circuit of the installation, forced convection can be implemented.

[0193] An example of such a configuration by forced convection is shown schematically in [Fig. 13].

[0194] The tertiary water circuit intended to supply heat to a heat network includes at least one exchanger 200 with the secondary circuit, of which an inlet 201 is fluidly connected to an inlet pipe 210 and an outlet 202 is fluidly connected to an outlet pipe 220.

[0195] The portion of the secondary circuit that is connected to the exchanger 200 comprises:

[0196] - a 230 piping that opens directly into the secondary water of the tank well 100 at the hot temperature T2 above the thermocline and is connected to an inlet 203 of the exchanger 200;

[0197] - a pipe 240 connected to an outlet 204 of the exchanger 200 which opens directly into the secondary water of tank well 100 at the cold temperature Tl below the thermocline.

[0198] A pump 250 ensures forced circulation in a closed loop of secondary water from the hot temperature T2 through respectively the piping 230, the exchanger 200, the piping 240 to the cold temperature Tl.

[0199] Figures 14 and 15 show an advantageous embodiment of the configuration according to [Fig. 13].

[0200] In this mode, two exchangers 200.1, 200.2 are implemented between the secondary and tertiary circuits with a hydraulic distribution block 260 which allows these two exchangers 200.1, 200.2 to be fluidly connected either in series or in parallel.

[0201] As illustrated in figures 14 and 15, the exchangers 200.1, 200.2 are of the plate type and counter-current flow.

[0202] As shown in [Fig. 15], different pipes connect each of the heat exchangers 200.1, 200.2 to the hydraulic distribution block 260 to achieve this series or parallel connection as follows: - the piping 261.1 filled with hot secondary water connects the block 260 to the hot inlet of the exchanger 200.1; - the piping 262.1 filled with secondary water at cold temperature connects the cold outlet of the exchanger 200.1 to the block 260; - the piping 263.1 filled with cold temperature tertiary water connects the block 260 to the cold inlet of the exchanger 200.1; - the piping 264.1 filled with hot temperature tertiary water connects the outlet of the exchanger 200.1 to the block 260; - the piping 261.2 filled with hot secondary water connects the block 260 to the hot inlet of the exchanger 200.2; - the piping 262.2 filled with secondary water at cold temperature connects the cold outlet of the exchanger 200.2 to the block 260; - the piping 263.2 filled with cold temperature tertiary water connects the block 260 to the cold inlet of the exchanger 200.2; - the piping 264.2 filled with hot temperature tertiary water connects the outlet of the exchanger 200.2 to the block 260.

[0203] Operating the heat exchangers 200.1 and 200.2 in series allows the heat exchange power to be divided in half in order to reduce the power supply to a heat network while maintaining the same operating parameters in temperature, or for a maintenance intervention on one of these two exchangers, operating at half power.

[0204] Parallel operation of heat exchangers 200.1 and 200.2 allows for the maintenance of nominal power dissipation, with one of the two heat exchangers 200.1 or 200.2 fluidically isolated and accessible for maintenance. In this scenario, heat exchangers 200.1 or 200.2 are each sized at 100% of the thermal power, whereas heat exchangers 200.1 and 200.2 in series are sized at 50%.

[0205] For series operation, the block 260 includes a series of isolation valves and connections between the secondary supply and supply pipes 210 and 220 on the one hand, and the pipes 261.1, 262.1, 261.2, 262.2 on the other hand. Similarly, a series of isolation valves and connections exist between the tertiary pipes 230 and 240 on the one hand, and the pipes 263.1, 264.1, 263.2, 264.2 on the other hand.

[0206] For a series connection, the block 260 connects the pipe 220 to the pipe 261.1, the pipe 262.1 to the pipe 261.2, and the pipe 262.2 to the pipe 210. The block 260 also connects the pipe 240 to the pipe 263.1, the pipe 264.1 to the pipe 263.2, and the pipe 264.2 to the pipe 230.

[0207] For parallel mounting, block 260 connects one or the other of the heat exchangers, while the other is isolated. For example, for operation of heat exchanger 200.1, block 260 connects piping 220 to the hot inlet piping of heat exchanger 261.1. Block 260 also connects outlet 262.1 to piping 210. On the tertiary side, block 260 connects piping 240 to the cold inlet piping 263.1, and block 260 connects the hot outlet 264.1 to piping 230.

[0208] Advantageously, as shown in [Fig. 14], for forced convection circulation of tertiary water, a pump 270 can be installed, for example on the cold supply pipe 210 of the tertiary circuit, and an isolation valve 212, 222, 232, 242 can be placed on each individual pipe 210, 220, 230, 240.

[0209] Fig. 16 illustrates a variant with a duplication of the pumps of the secondary and tertiary circuits which are put in fluidic parallel for intervention and maintenance of operation during this intervention.

[0210] Thus, two pumps 250.1, 250.2, each with an associated isolation valve 242.1, 242.2, are connected in parallel on the piping 240.

[0211] And two pumps 270.1, 270.2 each with an associated isolation valve 212.1, 212.2 are put in parallel on the piping 210.

[0212] [Fig. 17] illustrates an advantageous configuration in which the tank well 100 is arranged on the bottom 90 of a pool 9 called IRWST intended for the fuel handling phases as shown in [Fig. 17A].

[0213] The tank well 100 containing the secondary water basin is sealed by a removable metal cover 104 forming the separating wall with the bottom 90 of the pool 9.

[0214] The heat exchangers shown in [Fig. 14], 15, and 16 between the secondary and tertiary circuits are housed in the nuclear facility near the IRWST pool. The secondary suction piping 230 and secondary discharge piping 240 are also shown in [Fig. 17] and 17A.

[0215] Pool 9 contains a volume of water necessary for the complete covering of reactor block 4 in the open handling position, as illustrated [Fig.17A], for the cooling of the fuel assemblies on the one hand, and for biological protection and radiological containment on the other.

[0216] In addition, the quantity of cold water contained in pool 9, which is protected from external aggressions, serves as a safety cold source for accidental situations where the normal cooling systems, being shut down, cannot operate.

[0217] As mentioned previously, a piping not shown with water circulating within it from a water intake from the pool 9, embedded in the concrete wall 101, at a distance close to the interior and the metal lining forming the third barrier, can be provided as a cooling system for the tank well 100.

[0218] Another system is illustrated in [Fig. 18], with two redundant safety heat exchangers 500, of the drum type, fixed to the upper part of the tank well 100 at the level of the secondary hot water layer. As symbolized by the arrows in [Fig. 18], the cold water from the pool 9 flows by gravity down to the inlet manifold 501 of each heat exchanger, is heated through the heat exchanger tubes in contact with the secondary water volume B at its hot temperature, and exits into the pool 9 through the outlet manifold 502, forming a discharge stack. Between each of the inlet manifolds 501 and outlet manifolds 502 and the heat exchanger 500 to which they are connected, an isolation valve 503, 504 is arranged to isolate the heat exchanger systems 600 during normal operation and thus dedicate their operation solely to accidental phases.The positioning of these 500 exchangers in the 100 vessel well does not prevent handling, nor the complete removal of reactor block 4.

[0219] Another system that can be considered consists of two redundant 600 exchangers in the form of panels, each suspended directly from the cover plug 104 of the tank well, with a closed circuit of water from the bottom 90 of the pool 9, as shown in [Fig. 19]. As symbolized by As indicated by the arrows in [Fig. 18], the water from pool 9 flows by gravity into each suspended heat exchanger 600, from an inlet manifold 601 configured for lateral intake at the bottom 90 of pool 9, then through a central coaxial pipe 605, and rises by natural convection to an outlet manifold 602 forming a central discharge stack, after being heated before returning to pool 9. Between each of the inlet manifolds 601 and outlet manifolds 602 and the heat exchanger 600 to which they are connected, an isolation valve 603, 604 is arranged to isolate the heat exchangers 600 during normal operation and thus dedicate their operation solely to emergency situations. Thermal insulation, not shown, inside the metal dome 104 isolates the bottom of pool 9 from the secondary water temperature conditions.During the handling phases, all components 600 to 605 and the dome 104 from which they are suspended are removed, which does not impair the cooling of the secondary water volume in direct contact with the water in the IRWST 9 pool.

[0220] Another ultimate supplementary system consists of a direct connection by opening a valve to directly discharge the water from pool 9 into the secondary water volume of basin B, after the latter has been depressurized and brought into equilibrium with the pressure of pool 9. This ultimate supplementary system is intended to be implemented only for ultimate emergency situations where all the other preceding systems are not operational, and pool 9 must then be kept watertight by a watertight closure cover in order to maintain the integrity of the third containment barrier. The building housing pool 9 may be sealed in place of, or in addition to, the pool cover.

[0221] In the illustrated example in [Fig. 12], as shown, the tank well 100 is delimited by a prestressed concrete wall 101 lined with a liner-type metal lining. This configuration is suitable when the maximum temperature of the secondary water is below 100-110 °C and the pressure is below 5 bar.

[0222] When secondary fluid pressure and / or temperature conditions are higher, a thick-walled metal tank 300, independent of the concrete structure of the reactor vessel well 100, can be considered, as shown in [Fig. 20]. This thick-walled metal tank 300, which then forms the third containment barrier of the reactor, is separated from the inner concrete wall 101 of the reactor vessel well 100 but rests on its bottom, and is fixed by welding or bolting to a metal flange 301 embedded directly in the reactor vessel well's base slab. At its upper end, the tank 300 is also welded to the upper flange 302, and this flange 302 rests on the flange 303 embedded directly on the upper face of the reactor vessel well 101. Sealing is also achieved by bolting with compression of a metal gasket, not shown. An inlet chamber 106 is located at the bottom of the well. During installation and during the ten-year inspections, the tank ensures that the connection between the bottom of the tank and 301 is properly sealed. In normal operation, the airlock 106 must be completely sealed to prevent loss of inventory from pool 109. An intermediate space between the inner wall of the tank well 100 and the outer diameter of the tank 301 allows some of the water from pool 9 to flow by gravity, as indicated in [Fig. 20] by the arrows of the circulation loop R. Thus, the water from pool 9 can flow around the secondary water volume to maintain an acceptable temperature for the concrete of the tank well wall 101, but with a flow rate that must be limited to avoid excessive heat loss. This flow rate can be increased in an emergency to serve as a means of dissipating residual heat.

[0223] Water circulation flow control flaps for the pool 9 can be arranged at the inlet or outlet of the space between the thick metal tank and the inner wall of the tank well, at the penetrations 105 of the tank well, in order to regulate the cooling flow of the thick tank, protecting the concrete wall from excessive temperatures.

[0224] A coating on the inner wall 101 of the tank well prevents direct contact between the pool water 9 and the concrete of the tank well. Such a coating can be a paint or an epoxy resin.

[0225] In order to limit the heat losses of the secondary water volume, thermal insulation may be added to the internal wall of the thick tank 300, provided that it does not prevent the control and inspection of the tightness of the thick tank.

[0226] Preferably, the thick-walled tank should not exceed 7 meters in width to allow for its transport by road or sea, with road transport by special convoy. This thick-walled tank is manufactured in a factory and assembled either as a single module or as several modules to be assembled vertically on site, simply, for example, by automated welding around the perimeter. The vertical and horizontal expansion of the thick-walled tank is ensured by the flexibility of its walls, which are 10 to 20 mm thick and whose geometric shape is optimized to avoid exceeding the maximum thermomechanical stresses at the rigid connection welds.

[0227] The thermal protection of the tank well may only concern its part facing the secondary water at its hot temperature, its cold temperature being compatible with the strength of the concrete of the wall 101 of the tank well, and in particular the bottom of the well.

[0228] Preferably, the circulation of the cooling flow from the thick tank by a portion of the water taken from the bottom of the pool can be regulated according to two modes: - In the event of normal operation of reactor 4, a small flow of water from pool 9 flows by gravity along the thick tank, and returns by natural convection, once heated, to the pool. Typically, a flow rate of around 2 m3 / h is sufficient to ensure thermal protection of the tank well; - In the event of a transient incidental operation of reactor 4, such as a power outage preventing, for example, the normal release of the generated nuclear power, a full opening of the control dampers can maximize the cooling of the thick-walled reactor vessel, and therefore the volume of secondary water contained within it. The normal pool cooling system is designed to dissipate the heat generated by the minimum cooling flow rate during normal reactor operation. In an incidental or accidental event, the overall thermal inertia of pool 9 is sufficient to ensure reactor cooling for at least 7 days, and indefinitely under extreme conditions where boiling of pool 9 is permitted.The primary water within the reactor vessel is then also cooled by the surrounding secondary water, both by direct conduction through the reactor vessel and by the heat exchangers 49, which continue their heat exchange function. The primary water flow control valve 481 and each secondary flow control valve 5 are then automatically, by gravity drop of the ring 481 and the weight 53 respectively, brought to their fully open position to maximize this heat exchange. In the fully open position, the aforementioned control flaps can allow a flow rate of water from the pool 9 by natural convection exceeding 0.5 l / s, up to approximately 100 l / s.

[0229] The thermal power dissipation is then approximately 3 to 5 MW at the level of the metal wall of the thick tank 300, a value much higher than the residual power released by reactor 4. Consequently, a rapid decrease in average temperature is obtained in the total volume of secondary water, i.e. of basin B.

[0230] This type of direct thermal conduction cooling of the metal tank 300 is similar to the system shown in [Fig. 18], with cooling panels installed on the upper wall of the tank well 100, except that the wall of the tank 300 acts as a heat exchanger instead of the heat exchangers 500 shown in [Fig. 18]. The performance, in terms of exchange surface area and heat extraction, is essentially equivalent.

[0231] The inventors carried out preliminary thermo-hydraulic sizing calculations on all the primary, secondary and tertiary circuits of the installation according to [Fig. 12] using thermomechanical calculation software. This can be a conventional software, such as the pre-sizing code known as CATHARE: [6].

[0232] Considering a pinch of 15°C for the exchangers 49 between primary and secondary circuits, and a hot water supply to a heat network at 90°C, the calculations give an average temperature at the outlet of the reactor core C of 120°C.

[0233] It is recalled here that a pinch of exchanger 49 is the minimum temperature difference between the primary water and the secondary water at a given point of the exchanger.

[0234] With a temperature gradient of 40°C between the inlet and outlet of the reactor core, we obtain the thermo-hydraulic operating parameters allowing us to pre-size the exchangers 49, and an estimate of the positioning of the latter with respect to the core.

[0235] The following table 1 summarizes these different parameters.

[0236] [Tables 1] Parameters Values ​​Primary water pressure 4 bar Core inlet temperature 80°C Core outlet temperature 120°C Primary water flow rate 118.6 kg / s Secondary water pressure 5 bar Heat exchanger inlet temperature T1 65°C Heat exchanger outlet temperature T2 105°C Secondary water flow rate 119.1 kg / s Equilibrium water head between core and heat exchangers 3.16 m

[0237] Based on these data, an estimated volume for the heat exchangers 49 can be established, with a minimum height of 3.16 m relative to the average altitude of the fissile zone. A sizing margin of 4 m is recommended to allow for the addition of supplementary pressure losses via the throttling valve 481. The primary water pressure of 4 bar is determined such that the boiling margin relative to the average core outlet temperature is 20°C.

[0238] The reactor vessel 40, 41 of reactor block 4 has an overall height of about 9 m, with an overall diameter of 3 m incorporating the main shell 42 with a diameter of the lower compartment 40 of the vessel of 2.74 m and a diameter of the upper compartment 41 of 2.15 m.

[0239] Three identical heat exchangers 49 are selected, each with a useful heat exchange volume of approximately 1.2 m3 and a heat exchange height between the primary and secondary circuits of 2m. The three heat exchangers 49 are fixed with their manifolds 400, 401, 490, 491 at 120° to each other to compartment 45 of the reactor vessel as illustrated [Fig.5A].

[0240] In fact, the thickness of the reactor vessel is not dictated by considerations of pressure resistance, since it is immersed in the secondary water of basin B, but rather by constraints of mechanical rigidity, resistance to buckling, and core support. It should be noted here that the primary pressure is lower than the secondary pressure.

[0241] An average thickness of approximately 20 mm is therefore selected. The material envisaged for the reactor vessel 4 and the heat exchangers 49 is stainless steel.

[0242] Computer-aided design (CAD) gives a secondary water volume of approximately 200 m3 for its hot temperature T2 and approximately 100 m3 for its cold temperature TL. The secondary thermocline height is fixed at 4 meters, depending on the operating conditions in natural secondary circulation allowing the evacuation of 20MW thermal at nominal power, and an overall pressure loss on the secondary side of approximately 1000 Pa.

[0243] The average temperature of the secondary water is around 80 °C, which is a temperature lower than the boiling point of water in the atmosphere.

[0244] The primary water volume is on the order of 30 m3, which is approximately 10 times smaller than the secondary water volume. This results in a very high thermal inertia, which greatly contributes to the operational safety of reactor block 4, both in terms of absorbing any heating of the primary water and also with regard to the pressure rise of the primary water in the event of a residual power evacuation failure.

[0245] Furthermore, the consequences of a Primary Cooling Loss Accident (PCLA) are greatly reduced by the configuration of the secondary water which completely surrounds the primary circuit.

[0246] The following table 2 illustrates the characteristics of the thermohydraulic sizing of the secondary and tertiary and quaternary circuits (heat network).

[0247] [Tables2] Parameters Values ​​Temperature T1 at heat exchanger inlet 49-65°C Temperature T2 at heat exchanger outlet 49-105°C Secondary water flow rate 119.1 kg / s Secondary water pressure 5 bar Temperature T3 at heat exchanger inlet 230 65°C Temperature T4 at heat exchanger outlet 240 105°C Tertiary water flow rate 119.1 kg / s Electrical pumping power of tertiary water 2 kilowatts Natural convection equilibrium head S of secondary water 4m Temperature T5 at heat exchanger inlet 210 45°C Temperature T6 at heat exchanger outlet 220 90°C Tertiary circuit water flow rate 106.2 kg / s Electrical pumping power of secondary water final dependent on the client

[0248] The operating conditions are therefore substantially identical to those governing the primary water circulation.

[0249] A number of two identical exchangers 200.1 and 200.2 is retained with each having a useful heat exchange volume of approximately 2.4 m3 and an exchange height between secondary and tertiary circuits of 2m, for the parallel configuration, i.e. each discharging 100% of the nominal power.

[0250] In the series configuration, each heat exchanger 200.1 and 200.2 is sized at 50% of its nominal power, and the heat exchange length is then halved, to approximately 1 m. The secondary pumping hydraulic power, corresponding to the pumping units 250.1 and 250.2 in [Fig. 16], is estimated at approximately 1 kW, which leads to an electrical power of approximately 1.7 to 2 kW per pump. This power is highly dependent on the diameter and length of the pipes 230 and 240 considered. The power of the tertiary pumping unit is highly dependent on the quaternary circuit, i.e., the customer heating network, and cannot be estimated.

[0251] The driving height of natural convection of secondary water is at least 4 m, which fixes the maximum altitude position of the thermocline.

[0252] The level of the reactor vessel shaft foundation slab 103 is located approximately 25 m above ground level. In the CAD design, the bottom of pool 9 is located 10 m below ground level. The reactor building is therefore entirely underground, with the top of the IRWST pool corresponding to ground level.

[0253] The invention is not limited to the examples just described; in particular, features of the illustrated examples can be combined in unillustrated variants.

[0254] Other variants and embodiments may be envisaged without departing from the scope of the invention. List of cited references

[0255] [1]: The World Nuclear Industry Status Report 2017. https: / / www.worldnuclearreport.org / IMG / pdf / 20170912wnisr2017-en-lr.pdf.

[0256] [2]: IAEA-TECDOC-397 “Potential of Low-temperature Nuclear Heat Applications.

[0257] [3]: IAEA-TECDOC-463 «S mall R eactors for Low-température Nuclear Heat Applications».

[0258] [4] : https : / / inis.iaea.org / collection / NCLCollectionStore / _Public / 10 / 494 / 10494922.pdf

[0259] [5] :https: / / www.ecosmr.fi / wp-content / uploads / 2021 / 06 / Leppanen_EcoSMR_15062021 .pdf

[0260] [6]: G. Geffraye et al. “ CATHARE 2 V2.5 2: A single version for varions applications” Nuclear Engineering and Design 241 (2011) 4456-4463.

Claims

Demands

1. Nuclear installation comprising: - at least one SMR reactor block (4) for heat generation delimited by a reactor vessel (40, 41, 42), the reactor block comprising all the components and part of the fluidic circuit, including the reactor core which generates heat by nuclear fission reactions, which is housed inside the reactor vessel; - a vessel well (100) delimiting a water-filled basin (B) forming part of a secondary circuit, in which the reactor vessel is immersed; - at least one heat exchanger (49) between the primary circuit of the SMR reactor and the secondary circuit, arranged outside the reactor vessel and in the water basin.

2. Nuclear installation according to claim 1, the secondary circuit water pressure in the basin being greater than or equal to that prevailing in the reactor vessel.

3. Nuclear installation according to claim 1 or 2, the water pressure in the basin being less than or equal to 20 bar.

4. Nuclear installation according to any one of the preceding claims, the SMR reactor being configured to, when in operation, circulate the primary circuit water by natural convection between the heat exchanger(s) and the interior of the reactor vessel by passing through the fuel assemblies defining the reactor core.

5. Nuclear installation according to any one of the preceding claims, the water basin being configured to, when the SMR reactor is in operation, achieve 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 and the exchanger(s) between primary and secondary circuits are immersed and the top of the basin at a so-called hot temperature, the water in the basin circulating by natural convection between the exchanger(s) and the thermocline.

6. Nuclear installation according to claim 5, water at cold temperature coming from at least one heat exchanger (200) between the secondary circuit and a tertiary circuit.

7. Nuclear installation according to claim 5 or 6, further comprising an enclosure arranged around the reactor vessel and adapted to guide a flow of water from the basin around the reactor vessel, by natural convection.

8. Nuclear installation according to claim 5, 6 or 7, the level of the thermocline being fixed so as to be above the reactor vessel.

9. Nuclear installation according to any one of the preceding claims, the heat exchanger (49) being fixed outside the reactor vessel, the reactor vessel (40, 42) having inlet (400) and outlet (401) openings respectively into the inlet and outlet manifold of the part of the primary circuit in the exchanger.

10. Nuclear installation according to claim 9 in combination with claim 8, the heat exchanger (49) comprising an inlet manifold (490) and an outlet manifold (491) of the basin water, the thermocline level being fixed so as to be above the outlet of the outlet manifold.

11. Nuclear installation according to any one of the preceding claims, comprising a fuel assembly reloading pool (9) for fuel assemblies intended to be inserted into the core of the SMR reactor, the reactor vessel well (100) being arranged below the bottom (90) of the pool (9) and being closed by a removable metal cover (104) forming the separating wall with the bottom of the pool (9).

12. Nuclear installation according to claim 11, comprising at least one heat exchanger suspended by the removable cover (104) of the tank well, adapted to exchange heat between a closed circuit of water from the bottom (90) of the pool (9) and the water of the secondary circuit of the pool, the water from the bottom of the pool flowing by gravity into the suspended exchanger and rising by natural convection to rejoin the pool.

13. Nuclear installation according to any one of the preceding claims, the reactor block comprising a perforated plate (48) a portion of which individually houses a control rod for control rods 41 of reactivity (47) and of which at least one hole (480) houses a valve for regulating the flow of primary water circulating in the reactor.

14. Nuclear installation according to any one of the preceding claims, the structure delimiting the water basin comprising a bottom configured to support the reactor vessel.

15. Nuclear installation according to one of the preceding claims, in which the SMR reactor for heat generation is intended to be connected to a heat network.

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