Nuclear facility comprising at least one modular nuclear reactor (SMR) and a vessel pit delimiting a water basin in which the smr reactor block is submerged

Abstract

The invention relates to a nuclear facility comprising: - at least one SMR reactor block (4) for the purpose of heat generation, delimited by a reactor vessel (40, 41, 42); - a vessel pit (100) delimiting a water-filled basin (B) forming part of a secondary circuit, in which the reactor vessel is submerged; - 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:

Description

[0001] Description

[0002] Title: Nuclear installation comprising at least one modular nuclear reactor (SMR) and a reactor pit delimiting a water basin in which the SMR reactor block is immersed.

[0003] Technical field

[0004] The present invention relates to the field of nuclear power plants, in particular those comprising pressurized water nuclear reactors (PWRs). More particularly, it relates to the field of so-called small or medium power reactors or SMRs in English (acronym for "Small Modular Reactor"), for calogenic purposes.

[0005] The main objective of the invention is therefore to greatly simplify the operation and improve the safety of a PWR reactor for calogenic purposes which operates at low pressure, typically less than 6 bar and which is intended to provide relatively low thermal power, of the order of a few tens of MWth, typically between 5 and 200 MWth.

[0006] By "SMR reactor" is meant here and within the framework of the invention, the usual technological meaning, namely a nuclear fission reactor, of smaller size and power than those of conventional REL reactors, a block of which is manufactured in a factory and transported to a nuclear installation site for installation there.

[0007] By "reactor block" is meant here and within the scope of the invention, the vessel, called the reactor vessel as well as all the components and part of the fluid circuit, in particular the core of the reactor creating heat by nuclear fission reactions, which is housed inside the reactor vessel.

[0008] By "heat-generating" is meant here and within the scope of the invention, a nuclear installation, a nuclear power plant or a nuclear reactor whose power is mainly dedicated to the supply of heat. The power of a heat-generating reactor can be 100% to supply heat. A small part of its power can still be used to supply electricity.

[0009] By "electrical purpose" is meant here and within the scope of the invention, a nuclear installation, a nuclear power plant or a nuclear reactor whose power is mainly dedicated to the supply of electricity. The power of a reactor with an electric purpose can be 100% to supply electricity. A small part of its power can still be used to supply heat.

[0010] Prior art

[0011] One of the current developments for nuclear reactors concerns so-called calogenic reactors intended to provide a thermal power level of a few tens of MWth, mainly for the purpose of providing so-called urban heat, i.e. in urban networks, for towns / agglomerations of several hundred thousand inhabitants.

[0012] Among the various technological solutions that provide heat from nuclear fission, it is commonly accepted that to date, pressurized water reactors (PWRs) are the most suitable for providing heat at relatively low temperatures.

[0013] In fact, boiling water reactors (BWR) are designed to produce steam in a primary circuit directly used in a turbo-alternator group in order to produce electricity.

[0014] Fourth-generation fast neutron reactors (FNRs) provide heat at temperatures beyond those required, and their main disadvantage is the cost of construction and operation, which makes them incompatible with the exclusive supply of district heat.

[0015] Similarly, graphite-moderated and gas-cooled reactors are designed to provide heat at relatively high temperatures.

[0016] Finally, emerging concepts such as molten salt reactors do not have sufficient technological maturity for deployment in the relatively short term.

[0017] It should be remembered that a pressurized water nuclear reactor (PWR) comprises three cycles (fluid circuits) whose general principle of normal operation is as follows.

[0018] The high-pressure water in a primary circuit extracts the energy supplied, in the form of heat, by the fission of uranium nuclei, and where appropriate plutonium, in the reactor core.

[0019] Then, this water under high pressure and high temperature, typically 155 bars and 300 °C, enters a steam generator (SG) and transmits its energy to a secondary circuit, also using pressurized water as a heat transfer fluid. This water in the form of steam, at high pressure, typically around 70 bars, is then expanded via an expansion device transforming the variation in enthalpy of the fluid into mechanical and then electrical work in the presence of an electric generator.

[0020] 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.

[0021] The design principles of PWR reactors according to these three cycles have been essentially the same since the first ones were put into service.

[0022] The main elements of a PWR primary circuit are shown in Figure 1:

[0023] - a reactor building 1 providing various functions including in particular a contribution to the containment safety function,

[0024] - a reactor vessel 20, located in the center of building 1, housing the reactor core C,

[0025] - a primary circuit 2 in pressurized water comprising tank 20.

[0026] These main elements are therefore common, their composition and the number of components varying according to the power of the reactor.

[0027] Typically, the envelope of the reactor building 1 may be made of several thicknesses. For example, as illustrated in Figure 1, a reactor building 1 may be made of a reinforced concrete outer wall 12, a prestressed concrete inner wall 10 separated from the outer wall 12 by an annular space 13 devoid of material, and a metal skin 11 on the inside of the prestressed concrete wall 10, for a 1650 MWe reactor.

[0028] As illustrated in Figure 2, from publication [1], the primary circuit 2 consists of the following main components:

[0029] - a reactor vessel 20,

[0030] - primary loops 21 each comprising a primary pump 22 and a steam generator 23,

[0031] - a single 24 pressurizer.

[0032] In addition, this figure 2 shows the control rod mechanisms of the reactor core and control clusters 25. Depending on the power of the reactor, the number of loops can be three for a 900 MWe reactor or 4 for a 1300 MWe reactor and above.

[0033] The reactor 1 building is therefore sized, among other things, to house all the components of primary circuit 2.

[0034] Figure 3 illustrates the energy transfer cycle (heat then electricity) of a PWR reactor. In this figure 3, we can see in particular the distribution of the positioning of the components in relation to reactor building 1, which acts as the third containment barrier.

[0035] The fluid connections between the inside and the outside of the reactor building 1 are provided by the 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.

[0036] More precisely, for a given steam generator 23, the reactor building 1 is crossed by a line called the hot line 30 which evacuates the steam from the steam generator 23 to evacuate the power and bring it to the turbine 32, and by a line called the cold line 31 which supplies liquid water to the steam generator 23.

[0037] Among the solutions already studied for PWR reactors for calogenic purposes, 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 primary loop(s), and those of the integrated type generally illustrated in small power reactors, designated by the English acronym SMR ("Small Modular Reactor"), for electrogenic purposes. Reference may be made to publications [2] and [3].

[0038] The pool reactors implemented to date have been implemented experimentally with low power, typically 10 MWth: the pressure in the primary circuit can be close to atmospheric pressure, involving both moderate neutron fluences within the core, and the temperature of the primary circuit is limited and close to 100°C maximum. In a pool reactor, the height of liquid water above the core, however, allows the primary pressure within the fissile core to be slightly increased, while remaining within the order of magnitude of a few bars. The advantage of such a reactor lies in the simplicity of design, the reactor vessel not being considered as a pressure vessel, the concrete reactor pit and its internal sealed casing forming the reactor vessel surrounding the primary circuit.It is therefore mainly the maintenance of radiological containment, as in the case of a fuel storage pool, which governs the design quality of this "sandwich" component consisting of the concrete reactor pit and its internal watertight casing. The reactor vessel itself is assembled on site, and must of course withstand all extreme external attacks, including major earthquakes and aircraft crashes. In addition, the concrete must be subjected to temperatures compatible with maintaining its mechanical characteristics over time. This type of design is no longer used due to the strong constraints of demonstrating the quality of construction and on-site inspection for the construction of the second containment barrier of a nuclear boiler, rules set by the RCC-M, which is the French code defining the rules for the design and construction of mechanical equipment for the nuclear islands of PWR reactors.A self-supporting metal tank, manufactured in the factory, is preferred, and also allows for higher primary pressures.

[0039] An example of a pool reactor is the SLOWPOKE project carried out in the 1970s and 1980s by Canada (AECL) concerning NHP reactors (an English acronym for "Nuclear Heating Plants"). The figure on page 13 of the publication [3] illustrates this demonstration reactor, with a thermal output of 10 MWth, whose water-filled reactor pit contains the entire primary circuit and the exchangers between the primary and secondary circuits, the whole being closed by a slab at ground level. In this pool reactor, the circulation of the primary fluid is achieved by natural convection, which simplifies electrical loss transients and general system maintenance.The reactor also has significant thermal inertia, due to the importance of the water inventory in the primary circuit in relation to the reactor power, which also provides a gain in general safety in relation to incidental transients, and operation. Finally, the thickness of the reactor pit and its absence of lateral or lower crossings ensure by design the impossibility of draining the primary circuit, nor its depressurization. Such a pool reactor therefore meets many criteria of simplicity of design, safety in design, and ease of control. However, it has a major drawback: the manufacturing quality of the casing and the maintenance of the performance of the concrete forming the reactor pit are difficult to demonstrate respectively initially and over time.

[0040] A more recent example of implementation with the Chinese DHR 400 project of the CNNC company seems to show its reproducibility, since the reactor vessel is made up of a sandwich component with a thick prestressed concrete shell, typically of the order of one meter, partly coated internally with a 5mm thick stainless steel liner, the whole being enclosed in an external cylinder of 10mm thick carbon steel. However, this reactor can only be reasonably retained on the condition that the criteria of modularity and maximization of manufacturing of components in the factory, with transport to the site, which are preponderant, are respected.

[0041] An example of a pool reactor, with a traditional reactor pit and associated casing, is presented in the Russian RUTA-70 reactor project of the NIKIET company.

[0042] An example of a heat-generating loop reactor is the Chinese HAPPY 200 reactor from SPIC, dedicated to district heating in the city of Beijing. With a unit capacity of 200 MWth, it is designed in batches of two units, thus equaling the performance of the aforementioned DHR400 project from CNNC. Unlike the latter, the reactor vessel is a self-supporting steel structure, designed in a factory and assembled on site, although the primary loop requires complete welding, thus requiring heavy construction work. The thermal power is removed from the core via two primary loops feeding 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 an accidental transient situation, this double jacket is flooded by the cold water present in the pool.This loop reactor configuration thus has the advantage of being able to provide heat at a temperature relatively close to that leaving the core, thanks to forced convection in the primary circuit, which facilitates the extraction of thermal power, and due to the presence of plate exchangers between the primary and secondary circuits, which also allows a low thermal gradient between the temperatures of the primary circuit and the secondary circuit. The major disadvantages of this loop configuration lie in the forced convection in the primary circuit for reasons of simplicity in studying the transient loss of electricity, and the maintenance cost associated with the operation of these pumps. Another major disadvantage is the factory constructability of the loops and the reduction of the phases of their assembly on site (construction site delay).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 in the reactor building. Finally, with a loop configuration, there is a risk of a Primary Coolant Loss Accident (LOCA).

[0043] The last category is that of so-called integrated reactors, which comprise a block delimited by a reactor vessel entirely manufactured in the factory and transported to the site, and which houses the primary circuit in its entirety, and in particular the exchangers between the primary and secondary circuits. This type of integrated reactor has the same configuration as the main concepts of so-called SMR reactors currently existing, with an electrogenic function, namely a configuration based on the integration of the steam generator, or even all the components of the primary circuit, in particular the pressurizer and the primary pumps, inside the reactor vessel. These SMRs are called integrated SMRs.

[0044] The main advantages of SMR reactors over existing PWRs are that they allow for system simplification, mainly for safety reasons, and increased modularity through significant factory manufacturing of components for transport to the construction site.

[0045] In addition to the gain in compactness, integrated SMRs have the advantage of no longer requiring overhead fluid lines with pressurized water, with the exception of control circuit lines with a very limited diameter, typically a few cm, which considerably reduces the risk of accidents and associated consequences linked to the rupture of the primary circuit lines (APRP type accidents). Thus, on-site installation is greatly facilitated by limiting itself to secondary piping connections, except for the connections of the volumetric and chemistry system of the primary circuit, which are of small diameters.

[0046] For example, the nuclear power plant project, acronym NUWARD™, is a power plant 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.

[0047] Other integrated SMR projects for electricity generation are under development or have been studied, including the SCOR project with a power of 150 to 200 MWe on behalf of the Applicant or the ACP100 project from the Chinese manufacturer CNNC with a power of 100 MWe.

[0048] Figure 4 shows an example of an integrated SMR currently in the planning stage. Such an integrated SMR reactor, the block of which is generally designated by the reference numeral 4, comprises a fixed compartment 40 and a removable compartment 41 in the form of a cover, for the fuel handling or maintenance phases of the reactor internals.

[0049] A pioneering concept for an integrated calogenic reactor is the THERMOS project jointly conducted by the Applicant and Technicatome: [4]. The reactor according to this project had a thermal power of 100 MWth and was intended to supply district heat to the city of Grenoble. According to this project, the reactor vessel integrates the entire primary circuit, thus allowing operation under a pressure higher than that of a basin reactor, necessary to provide district heat of around 120°C. With a diameter of 5 m and a height of 9 m, the reactor vessel that was proposed was thus entirely factory-assembled, and housed in particular the exchangers between the primary and secondary circuits in its upper part. The low temperature gradient in the core required the presence of primary fluid pumping units, arranged in the hot part, which is not ideal in terms of safety and ease of maintenance.In addition, the thermal inertia of the primary and secondary circuits is relatively limited, the basin in which the reactor is immersed being thermally disconnected from the normal operation of the reactor, which is detrimental to operational safety and the smoothing of power call transients from the customer's heating network.

[0050] Recent studies are dedicated to calogenic reactors in Finland, carried out mainly by the VTT organization, for the district heating of the city of Helsinki: [5]. In particular, a reactor with a thermal power of 50 MWTh of the integrated tank type was studied, containing the exchangers between primary and secondary circuits in its upper part, the reactor tank itself being contained in a nearby containment enclosure. The circulation of the primary fluid is by natural convection in nominal operation, and the secondary circuit is a liquid loop circuit including an exchanger with the tertiary circuit which operates by forced convection by means of a pump. The tank and containment enclosure assembly is immersed in a pool which is the fuel handling pool (IRWST anglo-Saxon acronym for "In-containment Refueling Water Storage Tank").More specifically, the IRWST pool includes a pit forming the reactor pit, inside which the assembly consisting of the reactor vessel and the containment building is partially submerged. This IRWST pool also acts as a heat sink for design-basis accidents. This type of integrated reactor has the same advantages of factory constructability and modularity as those of integrated SMRs with a generator function, notably from the NuScale Power company. However, it has a major drawback of a lack of thermal inertia in the secondary circuit, an inertia which is necessary to have operating stability with respect to a load demand. In addition, the permanent thermal leaks linked to the IRWST pool configuration favoring thermal conduction through the reactor vessel mean that the pool must be constantly cooled, which leads to unexploited waste heat.

[0051] In summary, integrated SMR type reactors appear to be the best candidates as calogenic reactors, as they allow both optimal safety to be guaranteed (due to the simplification of the systems and the elimination of overhead fluid lines) and increased modularity capacity allowing the factory manufacture of components and their assembly on site.

[0052] However, the inventors analyzed that the integrated SMR type reactors that have been considered or as currently considered have several drawbacks.

[0053] First, the architecture of the safety systems used to maintain the primary circuit water inventory to prevent the reactor core from flooding in an accident situation presents weak points. Indeed, these systems implement dedicated safety injection water reserves that are directly connected by fluid lines to the reactor vessel. These lines are likely to be broken, which would in fact lead to an APRP.

[0054] A simplification of the safety architecture to make a heat-generating SMR very robust to all accidental events would be desirable since such a reactor will have to be located in areas close to the place to be supplied with heat.

[0055] Then, the SMR reactors for calogenic purposes planned to date in the form of integrated SMRs, with secondary and tertiary circulation loops to send the heat produced to a network, present 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.

[0056] Furthermore, the heat exchangers between the primary and secondary circuits as currently designed present integration difficulties and manufacturing complexity that do not sufficiently facilitate the construction and maintenance of the reactor. Finally, the limitation of thermal losses is not optimal and relies on heat insulation arranged around the vessel. It would be necessary to manage thermal leaks through the vessel, and the temperature protection of the reactor pit and more generally of the concrete surrounding the nuclear island.

[0057] There is therefore a need to improve integrated SMR type reactors, particularly when they are considered as calogenic reactors, in order to overcome the drawbacks mentioned above.

[0058] There is therefore a need for a nuclear installation with SMR type reactor(s) for a heat generating purpose which presents both:

[0059] - an optimized level of security,

[0060] - a level of flexibility and adaptability to the needs of the network(s),

[0061] - simplicity of construction,

[0062] - ease of installation and removal (minimization of fluid connection lines),

[0063] - ease of operation.

[0064] The aim of the invention is therefore to respond at least in part to this(these) need(s).

[0065] Statement of the invention

[0066] To this end, the invention relates, in one of its aspects, to a nuclear installation comprising:

[0067] - at least one SMR reactor block for calogenic purposes delimited by a reactor vessel, the reactor block comprising all the components and part of the fluidic circuit, including the reactor core creating heat by nuclear fission reactions, which is housed inside the reactor vessel;

[0068] - a tank well delimiting a pool filled with water forming part of a secondary circuit, in which the reactor vessel is immersed;

[0069] - 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.

[0070] Advantageously, the water pressure of the secondary circuit in the basin is greater than or equal to that prevailing in the reactor vessel. Advantageously again, the water pressure in the basin is less than or equal to 20 bar.

[0071] According to an advantageous operating mode, the SMR reactor is configured to, when in operation, circulate the water of the primary circuit by natural convection between the exchanger(s) and the interior of the reactor vessel by passing through the fuel assemblies defining the core of the reactor.

[0072] According to an advantageous embodiment, the water basin is 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 F(es) 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 in natural convection between F(es) exchanger(s) and the thermocline.

[0073] Preferably, the cold water comes from at least one heat exchanger between the secondary circuit and the tertiary circuit.

[0074] According to an advantageous construction variant, the installation further comprises a casing arranged around the reactor vessel and adapted to guide a flow of water from the basin around the reactor vessel, by natural convection.

[0075] Preferably, the thermocline level is set so that it is above the reactor vessel.

[0076] According to an advantageous configuration, the heat exchanger is fixed outside the reactor vessel, the reactor vessel being pierced with inlet and outlet openings opening respectively into the inlet and outlet manifold of the part of the primary circuit in the exchanger.

[0077] In an advantageous arrangement, the heat exchanger comprises an inlet manifold and an outlet manifold for the pool water, the level of the thermocline being fixed so as to be located above the outlet of the outlet manifold.

[0078] According to an advantageous embodiment, the installation comprises a pool for reloading fuel assemblies, intended to be inserted into the core of the SMR reactor, the reactor pit 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. According to this mode, and an advantageous variant embodiment, the installation comprises at least one heat exchanger suspended by the removable cover of the reactor pit, adapted to exchange heat between a closed circuit of water coming from the bottom of the pool and the water of the secondary circuit of the basin, the water coming from the bottom of the pool flowing by gravity into the suspended exchanger and rising by natural convection to join the pool.

[0079] According to an advantageous embodiment variant, the reactor block comprises a plate with holes, one part of which individually houses a control rod for reactivity control rods and at least one hole of which houses a valve for regulating the flow of primary water circulating in the reactor.

[0080] Advantageously, the structure delimiting the water basin comprises a bottom configured to support the reactor vessel.

[0081] The invention also relates to an assembly comprising a heat network and a nuclear installation as described above, the SMR reactor for calogenic purposes of which is connected to the heat network.

[0082] Thus, the invention essentially consists of immersing an SMR type reactor block as well as the heat exchangers between the primary and secondary circuits in a pool of liquid water which is part of the secondary circuit.

[0083] Until now, 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 general dimensioning of the vessel and the exchanger components themselves.

[0084] This classic design is explained by the interest in eliminating any possibility of a significant primary breach compared to a PWR reactor with a classic primary loop, essentially for applications where the reactor is intended for electricity generation and for which the primary circuit is subjected to significant pressures, of the order of 150 bars or even higher.

[0085] However, to address the construction, maintenance and safety issues, the inventors analyzed that by considering a calogenic reactor with a secondary water circuit comprising a secondary water basin, contained inside the reactor pit forming 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 breach of the primary circuit were greatly reduced and any risk of core dewatering 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.

[0086] Also, the inventors have wisely thought of arranging the exchangers between 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 makes the assembly and mounting phases more complex, and also complicates the general dimensioning of the reactor vessel and the exchanger components.

[0087] This so-called secondary water basin, which surrounds the reactor vessel, is advantageously at a slightly higher pressure than the latter to prevent the flow of the primary circuit from escaping in the event of a breach. The secondary water basin thus constitutes a passive defense barrier for radiological containment.

[0088] Responding to the need for flexibility, one advantage is that the secondary water basin is used as a buffer storage basin, providing high thermal inertia and making it possible to absorb a large part of the power variations on the heat network to which the SMR reactor for calogenic purposes is connected.

[0089] Also meeting the need for simplicity of intervention and operation, a subsequent advantage with a secondary water basin, which does not require connection / disconnection of secondary water circulation pipes, lies in the possibility of easily replacing or moving the entire reactor block, particularly for fuel reloading phases. During these phases, the reactor block can be completely moved, in order to greatly facilitate the handling of the fuel assemblies, in particular avoiding long lengths of handling pole as according to the state of the art and the associated risks of blockage or fall. In the event of a major failure of components of the reactor block, it is also possible to easily replace the reactor vessel and / or the exchangers between primary and secondary circuits, to limit downtime.Also meeting the need for safety in the event of a strong external attack such as an earthquake, the absence of a primary or secondary piping connection 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 pit, freely immersed in the secondary water basin, without fixed fluid transport connections to evacuate the thermal power.

[0090] Advantageously from the point of view of safety and from the point of view of thermal efficiency, a vertical thermal stratification is established in the secondary basin so as to constitute a thermocline, the water at the hot temperature being at the top of the basin while the water at the cold temperature is at the bottom of the basin. This cold water advantageously comes from at least one exchanger between the secondary circuit and the tertiary circuit.

[0091] Secondary water circulation is achieved by natural convection between the exchangers immersed in the basin and the thermocline.

[0092] This natural convection can be improved by the presence of a jacket arranged around the reactor vessel which allows the flow of pond water to be guided in natural convection around the latter. The arrangement of the jacket can be carried out a few centimeters around the reactor vessel. This flow will allow the exchange of thermal power between the primary circuit and the secondary circuit through the reactor vessel. This additional exchange surface makes it possible to reduce the size of the submerged heat exchangers and to recover the thermal losses through the reactor vessel. The jacket guiding the flow ensures that the thermal power thus exchanged through the reactor vessel does not destabilize the thermocline of the basin.

[0093] In an accident situation, the residual power can be evacuated by natural convection through the 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 need to cool the volume of water in the secondary circuit. Due to the possibility of a large thermal inertia by the volume of secondary water which can be significant, typically several days before any external intervention, the evacuation of the residual power from the reactor core is facilitated and simplified both through the exchangers dedicated to normal operation, but also directly by conduction by the secondary water through the walls of the reactor vessel.In addition to the removal of residual power through the primary exchangers, the jacket guiding the water flow around the reactor vessel also allows, under extreme accident conditions, to cool the core by direct thermal conduction with the vessel thanks to the secondary volume of cold water surrounding this jacket. This ultimate power removal mode makes it possible to overcome a possible common mode defect concerning the removal through the exchangers, by judiciously using the submerged situation of the reactor vessel in the secondary cold source.

[0094] In order to be able to correctly balance the flow rates by natural convection in primary and secondary water, throttling valves are provided and judiciously arranged.

[0095] The primary water throttling valve is advantageously integrated within the reactor vessel and its control mechanism is slaved to the control system of the reactivity control rods. The adjustment of the primary water flow rate is thus dependent on that of the neutron and therefore thermal power level of the core. More precisely, the height adjustment of the position of the throttling valve makes it possible to adjust the primary water flow rate and thus the heat exchange between the primary circuit and the secondary circuit, using rod followers and mechanical rod positioning systems identical to those dedicated to reactivity control. Preferably, the passive safety system makes it possible to release both the set of reactivity control rods to ensure the gravitational descent of the rods into the core and that of the throttling valve.In its extreme low position, the opening of the primary water flow is maximum, thus promoting the evacuation of residual power towards the secondary water volume.

[0096] In addition, the operation of the reactor by natural convection in primary and secondary water meets the need for adaptability to the customer's heating network insofar as it makes it intrinsically stable with respect to the thermal power demands of the heating network for which the installation is intended. This contributes to operational safety, and in fact simplifies the control and management constraints. For example, in the event of a sudden shutdown or drop in thermal power withdrawal from the secondary water volume, the supply of secondary water at its hot temperature by the exchangers between the primary and secondary circuits quickly leads to a thickening of the secondary water layer at its hot temperature compared to the water layer at its cold temperature surrounding the reactor block. This leads to a lowering of the thermocline altitude, thereby causing a drop in flow rate by natural convection in the secondary side exchangers.At constant nuclear power output, the heat exchange between the primary and secondary circuits therefore decreases, which in fact induces an increase in the average temperature of the primary fluid. With regard to safety, the neutron counter-reactions induced by the expansion of the primary fluid in the core on the one hand, and in the nuclear fuel by the Doppler effect on the other hand, generate an immediate drop in the reactivity of the core, thus counteracting the increase in the average temperature of the core.

[0097] There is therefore a natural or passive stable reaction of reduction of the neutron power of the core, without modification of the position of the reactivity control rods, during a reduction in the supply of thermal energy to the customer network, a guarantee of safety and operational security.

[0098] In addition to the core power adjustment by all 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 benefit is to be able to operate this calogenic reactor in a passive and simplified manner, by eliminating all the following accidental transients, which can usually be encountered in pressurized water reactors for power generation according to the state of the art, namely:

[0099] - a Primary Coolant Loss Accident due to the impossibility of significant loss of the primary water inventory and therefore the risk of core flooding,

[0100] - a Water Pipe Break and Steam Pipe Break, due to the absence of steam generation and a secondary water supply by direct immersion,

[0101] - a Reactivity Accident due to dilution of clear water, since the reactor operates without primary soluble boron,

[0102] - a loss of primary flow, due to primary natural circulation,

[0103] - a loss of secondary flow, due to secondary natural circulation,

[0104] - a loss of external electrical network, due to the passive nature of nominal operation.

[0105] Generally speaking, the supply of thermal heat around 90 to 110°C for a heat network leads to sizing the cold and hot temperatures of the primary circuit between 110 and 140°C, and a little below for the hot temperature of the secondary circuit. The return to the cold stop 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 rupture of the pipe or the tank well cover plug, cannot lead to a massive boiling of the secondary water and a significant loss of water inventory.

[0106] 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 that at its hot temperature, is less than 100°C, and thus cannot boil spontaneously in the event of sudden depressurization, thereby preventing the risk of flooding of the reactor core.

[0107] The invention can be implemented with secondary water pressures in the basin of up to approximately 12 bars, subject to appropriate sizing of the walls of this basin and in certain specific configurations of confinement of the secondary volume explained below.

[0108] Responding to the need for modularity and factory construction, compared to a conventional solution of an installation with a state-of-the-art PWR reactor and a basin, the solution according to the invention has the advantage of a primary circuit design in the factory (SMR reactor) and brought to the site, for better mechanical realization in accordance with the rules for the construction of nuclear equipment (RCCM). The development of the structure delimiting the secondary basin according to the invention can be done either on site, or also in the factory and transported to the site, with fewer constraints of realization and control compared to those of the primary circuit.

[0109] Still with the aim of reinforcing safety, an advantageous configuration consists of installing the reactor vessel well below the bottom of a fuel assemblies reloading pool, intended to be inserted into the SMR reactor core. 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 height of the IRWST pool, approximately 8 to 10 meters, ensures biological protection during the handling of fuel assemblies, which requires the dismantling of the vessel to access the reactor core and the submerged storage in the pool of the removable part of the vessel 41, 45.This configuration is also provided 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 metal liner forming the tank well.

[0110] This configuration allows a natural flow by gravity of part of the water present in this fuel reloading pool, to cool the secondary metal tank detached from the tank well and itself ensuring resistance to secondary pressure.

[0111] During normal reactor operation, a low flow of pool water flows naturally down a self-supporting metal tank inside the secondary reactor pit, and returns to the basin once heated, by natural convection.

[0112] In the event of a transient reactor incident, such as a power outage preventing the normal evacuation of the nuclear power produced, a greater flow of water from the pool can flow, which maximizes the cooling of the thick vessel, and therefore the volume of secondary water contained inside. 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.

[0113] The installation as planned with its calogenic reactor is intended to supply urban heat networks, industrial processes such as desalination, drying, and food product processing.

[0114] Advantageously, a complementary electrical production unit can also be added, using an organic Rankine cycle, to obtain local emergency electrical production in the event of an ultimate need, in addition to the conventional battery bank solutions generally used. For reasons of simplicity of design and maintenance, the installation of active means classified as emergency, such as diesel combustion, is not planned.

[0115] Other advantages and characteristics of the invention will become more apparent upon reading the detailed description of examples of implementation of the invention given by way of illustration and not limitation with reference to the following figures. Brief description of the drawings

[0116] [Fig 1] Figure 1 is a schematic perspective and partial sectional view of an existing PWR nuclear reactor.

[0117] [Fig 2] Figure 2 is a schematic view of a state-of-the-art PWR nuclear reactor primary circuit in a three-primary loop configuration.

[0118] [Fig 3] Figure 3 is a schematic view of the three cycles of a state-of-the-art PWR nuclear reactor.

[0119] [Fig 4] Figure 4 is a schematic perspective view of an integrated type SMR reactor as currently envisaged.

[0120] [Fig 5] Figure 5 is a schematic perspective view of a calogenic SMR reactor according to the invention.

[0121] [Fig 5A] Figure 5A is an exploded perspective view of the parts of the calogenic SMR reactor vessel according to the invention.

[0122] [Fig 6] Figure 6 is a longitudinal sectional view of the reactor according to Figure 5, and which illustrates the circulation by natural convection of the water in the primary circuit and the secondary circuit.

[0123] [Fig 6A] Figure 6A is a detailed schematic view of a control cluster assembly, the control rod, and the rod control mechanism of a calogenic SMR reactor according to the invention.

[0124] [Fig 7] [Fig 8] Figures 7 and 8 are detailed perspective views showing the primary circuit flow control valve in an intermediate position and the fully open position respectively.

[0125] [Fig 9] Figure 9 is a perspective view of part of the reactor according to Figure 5, and which shows in detail the installation of a valve for regulating the flow of water in the secondary circuit.

[0126] [Fig 10A] [Fig 10B] [Fig 10C] 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.

[0127] [Fig 11] Figure 11 is a partial longitudinal sectional view of the upper part of the reactor according to Figures 5 and 6 showing a pressurizer according to the invention.

[0128] [Fig 12] Figure 12 is a schematic view in longitudinal section and in transparency of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6 in a sealed reactor pit.

[0129] [Fig 13] Figure 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.

[0130] [Fig 14] Figure 14 is a schematic perspective and transparency view of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6 in a sealed reactor pit and the various components (pipes, heat exchangers, pumps, valves, hydraulic distribution block) of the primary, secondary and tertiary circuits.

[0131] [Fig 15] Figure 15 shows the hydraulic distribution block section with the two exchangers between the secondary and tertiary circuits which can be connected in series or in fluid parallel.

[0132] [Fig 16] Figure 16 is a perspective view showing a variation of the hydraulic components of Figure 14.

[0133] [Fig 17] Figure 17 is a schematic perspective and transparency view of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6 in a sealed reactor pit, arranged on the bottom of a handling pool (IRWST) whose water constitutes the safety cold source.

[0134] [Fig 17A] Figure 17A is a schematic perspective and transparency view of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6, during the handling phase, with the reactor pit open, and with the top of the reactor block dismantled and moved to an IRWST pool storage station.

[0135] [Fig 18] Figure 18 is a schematic and transparent view of a nuclear installation according to the invention with a calogenic 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 pit.

[0136] [Fig 19] Figure 19 is a schematic perspective and transparency view of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6, with a metal 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.

[0137] [Fig 20] Figure 20 is a schematic perspective and transparency view of a nuclear installation according to the invention with a calogenic reactor according to Figures 5 and 6, comprising a thick metal tank filled with secondary water, which is separated from the reactor pit, and is dedicated to applications at higher secondary water pressure and temperature.

[0138] [Fig 20A] Figure 20A is a schematic view of the thick metal tank filled with secondary water according to Figure 20.

[0139] Detailed description

[0140] Throughout the present application, the terms "vertical", "lower", "upper", "bottom", "top", "below" and "above" are to be understood with reference to an SMR nuclear reactor, as provided in a vertical operating configuration and whose reactor pit is arranged at least partly 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.

[0141] By "primary water", "secondary water", "tertiary water" we mean the water which constitutes the fluid of the primary, secondary and tertiary circuits respectively.

[0142] Figures 1 to 4 have already been detailed in the preamble, so they will not be commented on below.

[0143] It is specified that the different temperatures, powers, volumes, flow rates, etc. indicated are for information purposes only. For example, other temperatures may be considered depending on the configurations, in particular the SMR reactor power, the volume of water in the secondary basin, and the power requirement for the heat network. A pressurized water type nuclear reactor 4 is described with reference to Figures 5 and 6, according to a primary circuit configuration of the integrated SMR type according to one embodiment of the invention.

[0144] This reactor 4 has a unit power of 20 MW thermal, for calogenic purposes, i.e. dedicated to the supply of hot water, typically at 90°C. Its unit power can however vary upwards or downwards, in a range of approximately 10 MW to 100 MW, and the hot water supply temperature can also change depending on the customer's heating network.

[0145] The reactor 4 with central axis X in Figure 5 comprises a block delimited by a vessel body 40, an intermediate body 45, and a metallic dome 41 preferably made of stainless steel, with a thickness of the order of 10 to 20 mm, and formed of a hemispherical vessel bottom and a vertical cylinder. This reactor vessel consists of a fixed compartment 40 and a removable compartment 45 and 41 as indicated in Figure 5A, above the reactor core for the fuel handling or maintenance phases of the reactor internals. The compartment formed by the bodies 41 and 45 is removable to allow the handling of the assemblies contained in the vessel body 40. The removable compartment 41 is a cover in the form of a dome 6 whose central chimney incorporates a valve 64, suitable for cooling the reactor pressurizer as detailed below.

[0146] 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 which can move up or down in the assembly to control the reaction and form the control rods 42. Data from preliminary studies carried out by the Applicant consider a number of 52 assemblies and a cycle duration of 10 years, with a fissile height of 1.5 m.

[0147] The vessel body 40 houses a cylinder 43, supporting an assembly basket usually referred to as a “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. A set of flanges is bolted between the fixed compartments 40 and the removable compartment 45 and the dome 41. The seal between the flanges of the compartments 40 and 45 on the one hand, 45 and 41 on the other hand, is advantageously ensured by a metal seal. The dismantling of the bolted flange located between the compartments 40 and 45 allows the complete handling of the fuel assemblies during the core reloading phases. The block formed by the set of compartments 45 and 41, with the fixed exchangers 49, is completely removed to directly access the reactor core during the handling phase.The purpose of removing the bolted flange connecting compartments 45 and 41 is to gain access to the upper internals of the core, the control rod mechanisms and the flow throttling ring. The pressurizer associated with compartment 41 and its internal elements can also be separated from the rest of the vessel block for intervention and maintenance.

[0148] The studies carried out by the Applicant provide for shutdowns for fuel reloading scheduled during ten-year inspections, without intervention on the core between these periods.

[0149] Above core C, control rods 46 of the control cluster allow the insertion of nuclear reactivity control rods 42, in a manner similar to that usually found in conventional PWR reactors. The control rods 42 are rods made of neutron absorbing material.

[0150] The free volume above the reactor core C allows the control rods 42 to be positioned fully extended, as well as the standby position of so-called emergency absorber rods, dedicated to the safety shutdown of the nuclear reaction. The control rods 46 are individually controlled vertically using the rod control mechanisms 47.

[0151] Above the control mechanisms 47, a plate 48 is fixed with holes 480 which allow the passage of the hot primary fluid leaving the core in the central part called "riser". Peripheral holes also allow the passage of the control rods of the throttling valve 481.

[0152] At the periphery of the plate 48 is arranged a flow control valve 481 for regulating the flow rate of the water in the primary circuit, called a flow control valve. 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 enabling the primary water outlet openings to be covered, which enables the flow rate of the latter to be adjusted.

[0153] This throttle valve 481 has the function of regulating the natural circulation flow of the water of the primary circuit passing through the openings 400 which constitute the inlets of the primary water collectors of the exchangers 49 between the primary and secondary circuits. The positioning control of this regulating valve is carried out by a motor or bar control mechanism, with the control rod 482 linked to the crown 481, which is advantageously similar to those used by the control rods of the reactivity control bars 42.

[0154] In an intermediate position, as illustrated in Figure 7, the rolling crown 481 leaves the openings 400 partially clear, which determines the flow rate of primary water which passes through it towards the exchangers 49.

[0155] In the event of a power failure or emergency shutdown, the gravitational fall of the control rods 46 also triggers the gravitational fall of the primary fluid control valve. In the extreme low position, as illustrated in Figure 8, this control valve 481 allows the water from the primary circuit to pass completely through the openings 400, thus maximizing the flow rate in the exchangers 49, in order to evacuate the residual power and cool the primary circuit. This operation by gravitational fall of this control valve guarantees reliability and safety in the event of power loss or reactor failure.

[0156] In reactor 4 of figures 5 and 6, in normal operation, the thermal power created by the nuclear chain reaction within the reactor core is evacuated by the fluid of the primary circuit which rises by natural convection in an upward manner, to arrive in the upper part, where it can then flow according to the different outlet openings 400 corresponding to the inlet collectors of the exchangers 49 between the primary and secondary circuits and in an upper central portion of the core, in the form of a chimney. This central chimney (not detailed), called a riser, contains, in addition to the control rod piloting mechanisms, the instrumentation sensors for the core parameters.

[0157] Thus, the separation envelope 44 of the core C makes it possible to separate the water, the fluid of the primary circuit, in its so-called cold and hot temperatures. Thus, the primary water at cold temperature surrounds the core C outside the envelope 44, while the primary water at hot temperature, heated by circulating upwards in the core C, is found in the upper central portion of the core.

[0158] 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 making it possible to ensure the thermal insulation and pressure difference functions of the integrated pressurizer. This separation plate may be of the type already described in patent application WO2012 / 158929 A3.

[0159] The upper part integrating the reactor pressurizer will be detailed later with reference to figure 11.

[0160] After its cooling through the exchangers 49 in a downward manner, the water of the primary circuit passes through the openings 401 which constitute the primary water outlet collectors of the exchangers 49 then returns in a closed circuit to the lower part of the reactor core for a new heating phase. The circulation in closed circuit P solely by natural convection of the primary water is symbolized by the white arrows in figure 6. The driving force of the primary circuit in natural convection is controlled by the difference in height between the median altitude of the exchangers 49, and the average height of the fissile zone of the core defined by the core C.

[0161] As already mentioned, the adjustment of the primary pressure losses and therefore the flow rate are carried out by the throttling valve 481 whose positioning control rods are housed in one of the holes 480 of the plate 48. The inlet and outlet temperature of the primary water is adjusted thanks to the neutron flux conditions, that is to say the thermal power of the core, by the positions of the reactivity control rods 42 in the core, and to the saturation temperature and pressure conditions in the pressurizer. Here, due to the circulation only in natural convection, that is to say in the absence of active means of pumping primary water, it is the primary water flow throttling valve and the thermo-hydraulic parameters (hot and cold temperatures) which set the circulation and exchange conditions between primary and secondary circuits in relation to the power produced in the core.In fact, the control of the operation of the calogenic reactor can be carried out simply with, in addition to this adjustment of primary water pressure, the adjustment of the core power by the set of control rods 42. The exchangers 49 between primary and secondary circuits are preferably plate exchangers, advantageously made of stainless steel, and designed to withstand the pressure of the water in the primary circuit. Advantageously, these exchangers 49 are manufactured by stacking consisting of grooved metal plates assembled together either by hot isostatic pressing (HIP) or by hot uniaxial pressing (HUC) so as to obtain diffusion welding between the metal plates, or by brazing.

[0162] Within a 49 exchanger, the flow is downward for primary water, and upward for secondary water.

[0163] As shown in Figure 6, the secondary circuit of this reactor 4 is not a closed loop circuit as in conventional PWR reactors, but comprises a water basin B, as shown diagrammatically in Figure 12. This basin B is contained in the space of the reactor vessel well forming the third containment barrier, and the reactor vessel 4 is immersed therein.

[0164] This secondary circuit with liquid water basin B is an open environment delimited by the tank well, without a circulation pump.

[0165] With such a liquid water basin B for the secondary circuit, the exchangers 49 are not integrated into the reactor vessel 40, 41 but arranged and fixed outside it. Such an arrangement is possible because the unlikely event of rupture of the primary water inlet or outlet pipes, then causing a large diameter breach of loss of the latter, 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 envelops the reactor vessel 4, and an accident of this type cannot lead to a risk of dewatering of the core, endangering the physical integrity of the reactor core.

[0166] The internal circuit within an exchanger 49 which is part of the secondary circuit of the reactor 4 therefore sees a flow of liquid water pass as a secondary fluid which heats up in contact with the primary water within the exchanger 49, by natural suction from its inlet manifold 490 at the bottom to their outlet manifold 491 at the top thereof. The secondary water then creates a volume greater than a so-called hot temperature. The separation layer between a so-called cold temperature and the so-called hot temperature of secondary water is designated as a thermocline, as symbolized under the name thermocline in figures 6, 11 and 12.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 the hot temperature. It is the height of the thermocline layer which will fix the cooling flow rate of the secondary circuit through the exchangers 49. The circulation in closed circuit S solely by natural convection of the secondary water is symbolized by the gray arrows in Figure 6.

[0167] The natural convection flow of the secondary water is regulated by control valves 5 integrated in each of the outlet manifolds 491 of the exchangers 49, as illustrated in Figure 9. The cold temperature of the secondary water is governed by the temperature conditions of the secondary water basin B. The hot temperature is set by the control valves 5 in the outlet manifolds 491, and by the heat exchange within the exchangers 49 themselves.

[0168] An example of integration of a control valve 5 in the form of a butterfly valve 50 in an outlet manifold 491 is shown in Figures 10A, 10B, 10C which show the valve in a fully open position respectively in which the maximum secondary water flow from the basin can pass, intermediate, and a fully closed position in which no flow can pass. The butterfly valve 5 is rotated by the output shaft 51 of an electric motor 52.

[0169] Advantageously, the end of the shaft 51 opposite that linked to the butterfly 50 is linked to a remote weight 53. As shown in figure 10A, in the event of an electrical failure or triggering of an emergency stop, 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 the basin B.

[0170] During the nuclear reactor start-up phase, the thermocline is completely merged with the upper free level of the secondary water basin B. The driving height of the secondary water circulation is then maximum due to the maximum weight of the cold water column feeding the inlets of the exchangers 49. The thermal power demand towards the primary circuit is then maximum, and the average temperature of the primary water drops. The lowering of the primary water temperature leads to an average cooling of the moderator in the core, inducing an increase in the reactivity of the core, and therefore an increase in its thermal power. Maximum conditions of thermal heating of the secondary water volume are accompanied by a natural increase in the power of the core, reactor 4 is therefore naturally stable.As already mentioned, the position of the primary water throttling valve 480, combined with the positions of the reactivity control bar 42, however, makes it possible to limit the increase in the reactivity of the core, to remain within the temperature rise range of the entire reactor block 4 and its reactor pit.

[0171] Conversely, when the thermocline level drops, this implies a rise in the secondary hot water layer, and therefore a drop in the secondary water driving height through the exchangers 49 since the height of the cold water column decreases. Then, the circulation of secondary water by natural convection decreases, thereby reducing the heat exchange between the primary and secondary circuits. In core C, the drop in power removal leads to a rise in the average temperature of the primary water, and therefore to a rise in the average temperature of the moderator in the core. There is therefore a drop in reactivity due to expansion of the moderator, and the neutron and thermal power produced decreases. The reactor is therefore naturally stable for removal and thermal storage towards the volume of secondary water defined by basin B.Typically, a difference in altitude between the median plane of core C and the median plane of exchangers 49 of approximately 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 approximately 1100 Pascal of driving pressure necessary to overcome the pressure losses of core C, of ​​exchangers 49, and of the rest of the primary circuit including the additional adjustable pressure loss formed by the rolling crown 480.

[0172] The secondary water volume is sized 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 hand. Typically, secondary water volumes are in the order of 200 to 300 m 3 for the cold zone, and from 100 to 150 m 3 for the hot zone, i.e. a total volume for basin B of between 300 and 450 m 3Typically, a difference in altitude of approximately 4 meters between the median plane of the exchangers 49 and the position of the secondary thermocline separating a secondary cold water layer at 65°C from the hot water layer at 105°C allows the establishment of the natural circulation of a cold secondary fluid at 65°C at the inlet of the exchanger 490, and hot at 105°C at the outlet of the exchanger 491, generating approximately 1000 Pascal of driving pressure necessary to overcome the pressure losses due to the crossing of the exchangers 49, from the inlet 490 to the outlet 491, including the pressure losses adjustable using the secondary flow control valves 5.

[0173] The position of the thermocline can only be maintained at a fixed position on the condition that there is a continuous withdrawal of a quantity of secondary water at its hot temperature and a replacement by the same quantity of secondary water at its cold temperature. There is therefore an adjustable pumping system making it possible to transport, to a heating network, the power corresponding to the customer's demand, i.e. required by the heating network. In the event of an untimely shutdown of this heat evacuation, or an accidental shutdown of the pumping system, the stability conditions described above make it possible to temporarily store the power produced by the reactor core by modifying the ratio between the secondary water at its cold temperature and at its hot temperature, and by lowering the level of the thermocline.After several minutes of operation, the continuous evacuation of the thermal power produced by the reactor without external escape, leads to the shutdown of the reactor, to evacuate only the residual power, through specifically dedicated residual power evacuation systems. Typically, a continuous power of 20 MW thermal, with a tertiary water supply at 90°C and a return at 45°C by the customer heating network, requires pumping of 123 liters per second, or 442 m. 3per hour from the hot water layer at 105°C above the secondary thermocline and a return of the same quantity at the bottom of the tank well at 65°C. Preferably, this evacuation and return can be implemented by means of piping coming from the upper part of the tank well, in order to avoid lateral connections likely to cause leaks or problems of lateral mechanical strength restricting expansion or earthquake resistance. The heat transfer of 20 MW of thermal energy from the secondary water layer pumped to the tertiary water circuit of the customer network is carried out by means of one or more heat exchangers sized to transfer 20 MW of thermal energy 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, in order to have operational redundancy in the event of a breakdown or the need for intervention. The presence of this piping must not hinder the transport of the entire reactor block, as detailed below for its removal from the reactor pit using heavy handling equipment.

[0174] As previously stated, the primary circuit of the reactor operates solely by natural convection, i.e. without a pumping unit.

[0175] Consequently, the inventors were faced with the problem of producing a pressurizer whose cooling and condensation part of the primary fluid vapors 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.

[0176] The inventors then thought of modulating the thermal losses by conduction through the dome 6, to control the depressurization of the primary steam of the pressurizer, by taking advantage of the fact that the metal casing 60 of the cover 41 is of low thickness, typically between 10 and 20 mm.

[0177] Thus, as illustrated in Figure 11, the steam cooling and condensation part comprises a dome 6 with double walls 60, 61 spaced apart from each other forming a space E inside which liquid water from basin B can circulate from the bottom to the top of the dome forming a central evacuation chimney 62. Typically, space E has a constant height of the order of 0.5 to 2 cm.

[0178] In normal operation, the thermocline level is set sufficiently above the pressurizer, in particular so as to be located above the central exhaust stack 62, as illustrated in Figures 11 and 12.

[0179] Thus, the liquid water which thus circulates 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 liquid water entering the space E at the bottom of the dome 6 is around 65°C, which allows efficient and rapid cooling of the dome 6, and in particular of the internal wall 60 forming the enclosure of mechanical resistance to the pressure of the primary circuit, and thereby of the primary water vapor which is underlying it, and to a saturation temperature of the primary pressure around 3.5 bar, or approximately 140°C. Typically, such a device allows the extraction by natural circulation of the order of 0.3 MW thermal, and thus condenses approximately 0.15 kg / s of steam to saturation.The saturation steam inventory in the pressurizer, during normal operation, is of the order of a few kilograms, depending on the required pressurizer volume. The primary depressurization power is therefore fully compatible with the primary pressure control requirements.

[0180] The central chimney 62 incorporates within it a regulating valve 64 which makes it possible to regulate the flow rate of secondary liquid water which circulates in the space E and therefore to regulate the liquid cooling as such. Indeed, in a position of total closure of the valve 64, the layer of water is trapped and stratified in the space E. Conversely, in an open position, in particular total, of the valve 64, the hot water rises naturally in the space E then through the central chimney 62 and will join the upper layer of hot water of the basin B, while the cold water of the basin B is sucked in by the inlet in the lower position of the double wall 60, 61.

[0181] Valve 64 may be a butterfly valve such as secondary flow valve 5 shown in Figures 10A, 10B, 10C.

[0182] The two walls 60, 61 of the dome 6 are metallic, preferably stainless steel.

[0183] The external wall 61 of the dome 6 is advantageously covered with a cap 63 housing within it a thermal insulator to prevent the dome from cooling when the valve 64 is in the closed position.

[0184] According to an advantageous variant, the reactor 4 comprises, as a passive heat sink, a plurality of cooling fins 65 arranged inside the inner wall, preferably being distributed uniformly over the surface of the latter, preferably being welded or brazed. These fins 65 thus increase the contact surface with the steam of the primary circuit and therefore make it possible to improve the heat exchange by conduction between said steam and the dome 6. In the example illustrated, these fins 65 are rectilinear 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 cm and a length of a few tens of cm along the inside of the wall 60.

[0185] Furthermore, the heating part of the pressurizer comprises a plurality of electrical resistors 8 wrapped in an electrical insulator and powered by electrical cables, arranged inside the dome, preferably on the separation plate 7 which in its center comprises a portion with holes 70 making it possible to ensure the thermal insulation and pressure difference functions of the integrated pressurizer. Such a portion with holes 70 is for example as according to the device described in patent application WO 2012 / 158929A3.

[0186] The electrical resistors 8 may be of the type described in patent US4135552.

[0187] As shown in Figure 12, the reactor vessel 40, 41, 45 of the reactor block 4 with the exchangers 49 is supported by a metal base 102, preferably made of stainless steel, or black steel coated with an anti-corrosion deposit. Preferably, this base is held in the bottom of the reactor pit 100 by a mechanical locking system, not shown, adapted to prevent its displacement and lifting in the event of an earthquake. However, during fuel handling or component replacement phases, the base 102 and the reactor block 4 that it supports can be lifted and brought to the top of the reactor pit. To do this, the mechanical locking system of the base must be able to be unlocked simply by tools accessible from the top of the reactor pit 100.

[0188] The base 102 which supports the reactor block 4 is rigidly connected to a foundation raft 103, and the metal cladding with which the prestressed concrete wall 101 is coated, is rigidly and tightly fixed to the raft 103. As already mentioned, the reactor 4 has a total physical impossibility of occurrence of a serious accident, with significant melting of the core and piercing of the primary circuit vessel. There is therefore no specific device for recovering the corium at the bottom of the reactor pit, the latter being confined in all imaginable physical configurations inside the reactor vessel, in total immersion in a liquid permanently.

[0189] 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 metal slab, mechanically welded, preferably still honeycomb or formed of boxes.

[0190] A gas ceiling 105, preferably nitrogen, is delimited by the free level of the secondary water volume in the tank well 100. This gas ceiling 105 makes it possible to control the secondary water pressure and to adapt the variations in free level and expansion of secondary water. The mechanical connection between the metal coating of the tank well and the cover plug is constituted by a metal seal or a high-temperature elastic seal, in order to maintain total sealing.

[0191] If necessary, as a cooling system for the tank well 100, a pipe not shown with water circulation therein can be embedded in the concrete wall 101, at a distance close to the interior and the metal coating forming the third barrier.

[0192] The reactor block 4, 102 only has connections of the instrumentation and core reactivity control type, in the form of electrical connectors, which are easily removable. Additional flexible or semi-rigid connectors, also removable, allow the permanent withdrawal and return of a small fraction of the primary circuit water, for chemical treatment and adjustment of the total volume of primary water. Thus, the reactor block 4, 102 is easy to disconnect and move for its possible total replacement.

[0193] Forced convection can be used to create the tertiary circuit of the installation.

[0194] An example of such a forced convection configuration is shown schematically in Figure 13.

[0195] The tertiary water circuit intended to supply heat to a heat network, comprises 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.

[0196] The part of the secondary circuit which is connected to the exchanger 200 includes:

[0197] - a pipe 230 which 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;

[0198] - a pipe 240 connected to an outlet 204 of the exchanger 200 which opens directly into the secondary water of the tank well 100 at the cold temperature Tl below the thermocline.

[0199] A pump 250 ensures forced circulation in a closed loop of the secondary water from the hot temperature T2 through the pipe 230, the exchanger 200, the pipe 240 respectively, up to the cold temperature TL. Figures 14 and 15 show an advantageous embodiment of the configuration according to Figure 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 fluidically 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 have countercurrent circulation.

[0202] As shown in Figure 15, different pipes connect each of the exchangers 200.1, 200.2 to the hydraulic distribution block 260 to achieve this connection in series or in parallel as follows:

[0203] - pipe 261.1 filled with secondary water at hot temperature connects block 260 to the hot inlet of exchanger 200.1;

[0204] - pipe 262.1 filled with secondary water at cold temperature connects the cold outlet of exchanger 200.1 to block 260;

[0205] - pipe 263.1 filled with tertiary water at cold temperature connects block 260 to the cold inlet of exchanger 200.1;

[0206] - pipe 264.1 filled with tertiary water at hot temperature connects the outlet of exchanger 200.1 to block 260;

[0207] - pipe 261.2 filled with secondary water at hot temperature connects block 260 to the hot inlet of exchanger 200.2;

[0208] - pipe 262.2 filled with secondary water at cold temperature connects the cold outlet of exchanger 200.2 to block 260;

[0209] - pipe 263.2 filled with tertiary water at cold temperature connects block 260 to the cold inlet of exchanger 200.2;

[0210] - pipe 264.2 filled with tertiary water at hot temperature connects the outlet of exchanger 200.2 to block 260.

[0211] Series operation of the exchangers 200.1, 200.2 makes it possible to divide the thermal exchange power in two in order to reduce the power supply to a heating network while maintaining the same temperature operating parameters, or for maintenance work on one of these two exchangers, in half-power operation. Parallel operation of the exchangers 200.1, 200.2 makes it possible to maintain nominal power evacuation operation, with one of the two exchangers 200. or 200.2 fluidically isolated on which work can be carried out. In this case, the exchangers 200.1 or 200.2 are each sized at 100% of the thermal power, while the exchangers 200.1 and 200.2 in series are sized at 50%.

[0212] For series operation, block 260 includes a series of isolation valves and connections between the secondary inlet and outlet pipes 210 and 220 on the one hand, and 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 pipes 263.1, 264.1, 263.2, 264.2 on the other hand.

[0213] For a series connection, block 260 connects piping 220 to piping 261.1, piping 262.1 to piping 261.2, and piping 262.2 to piping 210. Block 260 also connects piping 240 to piping 263.1, piping 264.1 to piping 263.2, and piping 264.2 to piping 230.

[0214] For parallel connection, block 260 connects one or the other of the exchangers, while the other is isolated. For example, for operation of exchanger 200.1, block 260 connects pipe 220 to the hot inlet pipe of exchanger 261.1. Block 260 also connects outlet 262.1 to pipe 210. On the tertiary side, block 260 connects pipe 240 to the cold inlet pipe 263.1, and block 260 connects the hot outlet 264.1 to pipe 230.

[0215] Advantageously, as shown in Figure 14, for forced convection circulation of the tertiary water, a pump 270 can be installed, for example on the cold inlet 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.

[0216] Figure 16 illustrates a variant with a duplication of the pumps of the secondary and tertiary circuits which are placed in fluid parallel for an intervention and maintenance of operation during this intervention.

[0217] Thus, two pumps 250.1, 250.2, each with an associated isolation valve 242.1, 242.2, are connected in parallel on the pipe 240. And two pumps 270.1, 270.2, each with an associated isolation valve 212.1, 212.2, are connected in parallel on the pipe 210.

[0218] Figure 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 figure 17A.

[0219] The tank well 100 containing the secondary water basin is sealed by a removable metal cover 104 forming the separation wall with the bottom 90 of the pool 9.

[0220] The exchangers shown in Figures 14, 15, and 16 between the secondary and tertiary circuits are housed in the nuclear facility near the IRWST pool. The secondary suction pipes 230 and secondary discharge 240 are also illustrated in Figures 17 and 17A.

[0221] Pool 9 contains a volume of water necessary to completely cover reactor block 4 in the open handling position, as illustrated in Figure 17A, for cooling the fuel assemblies on the one hand, and biological protection and radiological containment on the other hand.

[0222] In addition, the quantity of cold water contained in pool 9, which is protected from external aggression, serves as a safety cold source for accidental situations where normal cooling systems, when stopped, cannot operate.

[0223] As mentioned previously, it is possible to provide, as a cooling system for the tank well 100, a pipe (not shown) with a circulation of water within it from a water sample from the pool 9, embedded in the concrete wall 101, at a distance close to the interior and the metal coating forming the third barrier.

[0224] Another system is illustrated in Figure 18, with two redundant safety exchangers 500, of the drum exchanger type, fixed in the upper part of the tank well 100 at the level of the secondary hot water layer. As symbolized by the arrows in Figure 18, the cold water from the pool 9 descends by gravity to the inlet manifold 501 of each exchanger, heats up through the tubes of the exchanger in contact with the secondary water volume B to its hot temperature, and exits into the pool 9 through the outlet manifold 502 forming a discharge chimney. Between each of the inlet manifolds 501 and outlet manifolds 502 and the exchanger 500 to which they are connected is arranged an isolation valve 503, 504 in order to isolate the exchanger systems 600 in normal operation and therefore to dedicate their operation only to accidental phases. The positioning of these exchangers 500 in the vessel well 100 does not prevent handling or the complete removal of the reactor block 4.

[0225] Another system that can be considered consists of two redundant exchangers 600 in the form of panels, which are each suspended, directly by the cover plug 104 of the tank well, with a closed circuit of water coming from the bottom 90 of the pool 9, as shown in Figure 19. As symbolized by the arrows in Figure 18, the water from the pool 9 flows by gravity into each suspended exchanger 600, from an inlet collector 601 configured for lateral withdrawal at the bottom 90 of the pool 9, then through a central coaxial pipe 605, and rises by natural convection towards an outlet collector 602 forming a central discharge chimney, after heating to reach the pool 9.Between each of the inlet 601 and outlet 602 collectors and the exchanger 600 to which they are connected, an isolation valve 603, 604 is arranged in order to isolate the exchangers 600 in normal operation and therefore, to dedicate their operation of these exchangers only to accidental phases. A thermal insulator, not shown, inside the metal dome 104 isolates the bottom of the pool 9 from the temperature conditions of the secondary water. During the handling phases, all of the components 600 to 605 and the dome 104 from which they are suspended are removed, which does not harm the cooling of the secondary water volume in direct contact with the water of the IRWST pool 9.

[0226] Another ultimate complementary system consists of a direct connection by opening a valve in order to directly flow the water from pool 9 into the secondary water volume of pool B, after depressurization and balancing of the latter with the pressure of pool 9. This ultimate complementary system is intended to be implemented only for ultimate accident sequences where all the other previous systems are not operational, pool 9 then having to be kept watertight by a watertight closing cover, in order to maintain the integrity of the third containment barrier. It is possible to provide, in place of or in addition to the pool cover, for the building housing pool 9 to be waterproofed.

[0227] In the example illustrated in Figure 12, as shown, the tank well 100 is delimited by a prestressed concrete wall 101 coated with a liner-type metal coating. This configuration is suitable when the maximum temperature of the secondary water is less than 100-110°C and the pressure is less than 5 bars.

[0228] When pressure and / or temperature conditions of the secondary fluid are higher, a thick-walled metal tank type structure 300, independent of the concrete structure of the reactor pit 100, can be envisaged, as shown in Figure 20. This thick metal tank 300 then forming the third containment barrier of the reactor, is detached from the internal concrete wall 101 of the reactor pit 100 but placed on the bottom thereof, and fixed by welding or bolting with a metal flange 301 installed directly in the base of the reactor pit. In the upper part, the tank 300 is also welded to the upper flange 302, and this flange 302 is placed on the flange 303 installed directly on the upper face of the reactor pit 101. Sealing is also achieved by bolting with crushing of a metal seal, not shown.An inlet airlock 106 at the bottom of the tank well ensures, during installation and during the ten-yearly inspection visits, that the connection of the tank bottom between 300 and 301 is properly sealed. In normal operation, the airlock 106 must be perfectly sealed to avoid loss of inventory from the pool 109. An intermediate space between the internal wall of the tank well 100 and the external diameter of the tank 301 allows a flow by gravity of part of the water from the pool 9, as indicated in Figure 20 by the arrows of the circulation loop R. Thus, the water from the pool 9 can flow around the secondary water volume to maintain an acceptable temperature for the concrete of the wall 101 of the tank well but with a flow rate to be limited to avoid too many thermal losses. This flow rate can be increased in an accident situation to serve as a means of evacuating the residual heat.

[0229] It is possible to arrange shutters for regulating the flow rate of the water circulation in the pool 9 at the inlet or outlet of the space between the thick metal tank and the internal wall of the tank well, at the penetrations 105 of the tank well, in order to be able to regulate the cooling flow rate of the thick tank, protecting the concrete wall from excessive temperatures.

[0230] A coating on the inner wall 101 of the tank well prevents direct contact between the water in the pool 9 and the concrete of the tank well. Such a coating may be a paint or an epoxy resin. In order to limit heat losses from the secondary water volume, a thermal insulator may be added to the inner wall of the thick tank 300, provided that it does not prevent the control and inspection of the tightness of the thick tank.

[0231] Preferably, the thick tank should not be wider than 7 meters, in order to allow its transport by road or sea, road transport by exceptional convoy. This thick tank is manufactured in the factory and assembled either in one module or in several modules to be assembled vertically on site, in a simple way for example by automatic welding at the periphery. The vertical and horizontal expansion of the thick tank is ensured by the flexibility of its walls, with a thickness of 10 to 20 mm, and whose geometric shape is optimized so as not to exceed the maximum thermomechanical constraints at the level of the rigid connection welds.

[0232] 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 resistance of the concrete of the wall 101 of the tank well, and in particular the bottom of the well.

[0233] Preferably, the circulation of the cooling flow of the thick tank by a part of the water taken from the bottom of the pool can be adjusted according to two modes:

[0234] - in the case of normal operation of reactor 4, a low flow of water from pool 9 flows by gravity along the thick tank, and returns by natural convection once heated, into the pool. Typically, a flow of the order of 2 m 3 / h is sufficient to ensure thermal protection of the tank well;

[0235] - in the event of incidental transient operation of reactor 4, such as an electrical failure preventing, for example, the normal evacuation of the nuclear power produced, a complete opening of the control shutters can then maximize the cooling of the thick vessel, and therefore the volume of secondary water contained inside. The normal cooling system of the pool is sized to evacuate to the outside the heat input from the minimum cooling flow rate during normal operation of the reactor. In the incident or accident phase, the overall thermal inertia of pool 9 is sufficient to ensure cooling of the reactor for at least 7 days, and infinitely in extreme conditions where boiling of pool 9 is authorized.The primary water within the reactor vessel 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 49 which continue their exchange function. The primary water flow control valve 481 as well as each secondary flow control valve 5 are then automatically brought into their fully open position by gravity drop of the crown 481 and the weight 53 respectively in order to maximize this exchange. In the fully open position, the aforementioned control flaps can allow a flow of water from the pool 9 to pass by natural convection greater than 0.5 1 / s, up to approximately 100 1 / s.

[0236] The thermal power evacuation 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, we obtain a rapid decrease in the average temperature of the total volume of secondary water, i.e. of basin B.

[0237] This type of cooling by direct thermal conduction of the metal tank 300 is similar to the system presented in figure 18, with cooling panels installed in the upper wall of the tank well 100, with the difference that the wall of the tank 300 acts as an exchanger instead of the exchangers 500 shown in figure 18. The performances, in terms of exchange surface and thermal extraction, are substantially equivalent.

[0238] The inventors carried out thermo-hydraulic pre-sizing calculations on all the primary, secondary and tertiary circuits of the installation according to figure 12 using thermomechanical calculation software. This can be conventional software, such as the pre-sizing code known under the name CATHARE: [6].

[0239] Considering a pinch of 15°C for the exchangers 49 between the primary and secondary circuits, and a supply of hot water to a heating network at 90°C, the calculations give an average temperature at the outlet of the reactor core C of 120°C.

[0240] It is recalled here that an exchanger pinch 49 is the minimum temperature difference between the primary water and the secondary water at a given point of the exchanger.

[0241] 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 their positioning with respect to the core.

[0242] Table 1 below summarizes these different parameters. [Table 1]

[0243] From these data, an estimate of the volume of the exchangers 49 can be established, with an installation at a minimum height of 3.16 m relative to the average altitude of the fissile zone. A design margin of 4 m is recommended in order to keep the possibility of adding additional pressure losses via the throttling valve 481. The primary water pressure of 4 bars is determined so that the boiling margin with respect to the average temperature at the core outlet is 20°C. The reactor vessel 40, 41 of the reactor block 4 has an overall height of approximately 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.

[0244] A number of three identical exchangers 49 is retained, each with a useful heat exchange volume of approximately 1.2 m 3and an exchange height between primary and secondary circuits of 2m. The three exchangers 49 are fixed with their collectors 400, 401, 490, 491 at 120° from each other to compartment 45 of the reactor vessel as illustrated in Figure 5A. 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, buckling resistance and core support. It is recalled here that the primary pressure is lower than the secondary pressure.

[0245] An average thickness of approximately 20 mm is therefore retained. The material considered for the reactor vessel 4 and the exchangers 49 is stainless steel.

[0246] Computer-aided design (CAD) gives a secondary water volume of approximately 200 m 3 for its warm temperature T2 and about 100 m 3for its cold temperature TL The secondary thermocline height is set at 4 meters, depending on the operating conditions in secondary natural circulation allowing the evacuation of 20MW thermal at nominal power, and an overall pressure drop on the secondary side of approximately 1000 Pa.

[0247] The average temperature of secondary water is around 80°C, which is lower than the boiling temperature of water in the atmosphere.

[0248] The primary water volume is around 30 m 3 , i.e. a factor of approximately 10 lower than that of the secondary water volume. This therefore results in a very significant thermal inertia which contributes greatly to the operational safety of reactor block 4, both in terms of absorption of any heating of the primary water, but also with regard to the rise in pressure of the primary water in the event of a failure to evacuate the residual power.

[0249] In addition, the consequences of a Primary Cooling Loss Accident (PCLA) are greatly reduced by the configuration of the secondary water completely surrounding the primary circuit.

[0250] Table 2 below illustrates the characteristics of the thermo-hydraulic dimensioning of the secondary, tertiary and quaternary circuits (heat network). [Table 2]

[0251] The operating conditions are therefore substantially identical to those governing the circulation of primary water.

[0252] A number of two identical exchangers 200.1 and 200.2 is retained, each with a useful heat exchange volume of approximately 2.4 m 3and an exchange height between secondary and tertiary circuits of 2m, for the parallel configuration, i.e. each evacuating 100% of the nominal power. In the series configuration, each exchanger 200.1 and 200.2 is sized at 50% of the nominal power, and the exchange length is then divided by two, i.e. approximately 1m. The secondary hydraulic pumping power, corresponding to the pumping groups 250.1 and 250.2 in Figure 16, is estimated at approximately 1kW, which leads to an electrical power of approximately 1.7 to 2kW per pump. This power is very dependent on the sizing in diameter and length of the pipes 230 and 240 considered. The power of the tertiary pumping group is very dependent on the quaternary circuit, i.e. the customer heating network, and cannot be estimated.

[0253] The driving height of natural convection of secondary water is at least 4 m, which fixes the maximum altitude position of the thermocline.

[0254] The level of the reactor pit floor 103 is approximately 25 m above ground level. In the CAD design, the bottom of pool 9 is 10 m above ground level. The reactor building is therefore completely buried, with the top of the IRWST pool corresponding to ground level.

[0255] The invention is not limited to the examples which have just been described; in particular, it is possible to combine characteristics of the illustrated examples within non-illustrated variants.

[0256] Other variants and embodiments may be envisaged without departing from the scope of the invention.

[0257] List of cited references

[0258] [1]: The World Nuclear Industry Status Report 2017. https: / / www.worldnuclearreport.org / IMG / pdf / 20170912wnisr2017-en-lr.pdf.

[0259] [2]: IAEA-TECDOC-397 «Potential of Low -temperature Nuclear Heat Applications».

[0260] [3]: IAEA-TECDOC-463 «Small Reactors for Lo -temperature Nuclear Heat Applications».

[0261] [4]: https: / / inis.iaea.org / collection / NCLCollectionStore / _Public / 10 / 494 / 10494922.pdf

[0262] [5]:https: / / www.ecosmr.fi / wp-content / uploads / 2021 / 06 / Leppanen_EcoSMR_15062021.pdf [6]: G. Geffraye et al. “ CATHARE 2 V2.5 2: A single version for various applications” Nuclear Engineering and Design 241 (2011) 4456-4463.

Claims

Claims 1. Nuclear installation comprising: - at least one SMR reactor block (4) for calogenic purposes delimited by a reactor vessel (40, 41, 42), the reactor block comprising all of the components and part of the fluidic circuit, including the core of the reactor creating heat by nuclear fission reactions, which is housed inside the reactor vessel; - a tank well (100) delimiting a basin filled with water (B) forming part of a secondary circuit, in which the reactor tank 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 pressure of the water in the secondary circuit 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 one of the preceding claims, the SMR reactor being configured to, when in operation, circulate the water of the primary circuit by natural convection between the exchanger(s) and the interior of the reactor vessel by passing through the fuel assemblies defining the core of the reactor.

5. Nuclear installation according to 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, the 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 a casing 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 located above the reactor vessel.

9. Nuclear installation according to one of the preceding claims, the heat exchanger (49) being fixed to the outside of the reactor vessel, the reactor vessel (40, 42) pierced with inlet (400) and outlet (401) openings opening 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) for the water from the basin, the level of the thermocline being fixed so as to be located above the outlet of the outlet manifold.

11. Nuclear installation according to one of the preceding claims, comprising a pool (9) for reloading fuel assemblies, intended to be inserted into the core of the SMR reactor, the reactor pit (100) being arranged below the bottom (90) of the pool (9) and being closed by a removable metal cover (104) forming the separation 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 reactor pit, adapted to exchange heat between a closed circuit of water coming from the bottom (90) of the pool (9) and the water of the secondary circuit of the basin, the water coming from the bottom of the pool flowing by gravity into the suspended exchanger and rising by natural convection to join the pool.

13. Nuclear installation according to one of the preceding claims, the reactor block comprising a perforated plate (48) of which a part individually houses a reactivity control rod control rod (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 one of the preceding claims, the structure delimiting the water basin comprising a bottom configured to support the reactor vessel.

15. Assembly comprising a heat network and a nuclear installation according to one of the preceding claims, the SMR reactor for calogenic purposes of which is connected to the heat network.

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