EP4390972A1 - Installation comprising at least one nuclear reactor and a thermal storage pit at least partly arranged above the reactor and connected to a heat network - Google Patents

Abstract

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

The invention essentially consists of positioning a PWR type reactor at the bottom of a thermal storage pit delimiting a thermocline and thermally coupling them with a view to supplying heat to a network.

Description

Technical area

The present invention relates to the field of nuclear power plants, in particular those comprising pressurized water nuclear reactors (PWR). More particularly, it concerns the field of small so-called small or medium power reactors or SMRs in English (acronym for “Small Modular Reactor”), with a calogenic vocation.

The main objective of the invention is thus to greatly simplify the operation of a heat-producing PWR reactor which operates at low pressure, typically less than 15 bar and which is intended to provide a relatively low thermal power, of the order of a few tens of MWth.

In particular, the invention aims to achieve a coupling between a nuclear reactor and a heat storage to smooth the heat production of the reactor, until it can operate all year round at 100% of its power using the storage to manage the intermittent consumption of a low temperature heat network, typically below 100°C, for which the reactor is intended. It can be an urban, rural or industrial heating network.

By “SMR reactor”, we mean here and in the context of the invention, the usual technological meaning, namely a nuclear fission reactor, of smaller size and power than those of conventional PWR reactors, which is manufactured in a factory and transported to a nuclear installation site for installation.

Although described with reference to an SMR reactor with heat generation vocation with natural convection of fluids in the primary and secondary circuits, the invention applies to any type of PWR reactor capable of being immersed in a water basin with its primary circuits and secondary to natural or forced convection.

By “SMR reactor”, we mean here and in the context 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 factory and transported to a nuclear installation site for installation.

By "reactor block", we mean here and in the context, the vessel, called the reactor vessel as well as all the components and part of the fluidic circuit, in particular the core of the reactor creating heat by nuclear fission reactions , which is housed inside the reactor vessel.

By “heat-producing”, we mean here and in the context 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-producing reactor can be 100% to provide heat. A small part of its power can still be used to provide electricity.

By “electric-generating” is meant here and in the context 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 an electrogenerating reactor can be 100% to provide electricity. A small part of its power can still be used to provide heat.

Prior art

One of the current development topics for nuclear reactors concerns so-called heat-generating reactors intended to provide a level of thermal power of a few tens of MWth, for the purposes mainly of supplying so-called urban heat, that is to say in networks urban, for cities/conurbations of several hundred thousand inhabitants.

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

In fact, boiling water reactors (BWR) are intended to produce steam in a primary circuit directly operated in a turbo-alternator group in order to produce electricity.

Fourth generation fast neutron reactors (FNR), provide heat at temperature levels beyond the necessary requirements, and have the main disadvantages of construction and operating costs which make them incompatible with the exclusive supply of heat. urban.

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

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

We recall that a pressurized water nuclear reactor (PWR) comprises three cycles (fluidic circuits) whose general principle of normal operation is as follows.

High pressure water from a primary circuit draws the energy provided, in the form of heat, by the fission of uranium nuclei, and where applicable plutonium, in the reactor core.

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

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.

The design principles of PWR reactors according to these three cycles have been essentially the same since the start of commissioning of the first ones to be operated.

• un bâtiment réacteur 1 assurant différentes fonctions dont notamment une contribution à la fonction de sûreté de confinement,
• une cuve de réacteur 20, implantée au centre du bâtiment 1, logeant le coeur C du réacteur,
• un circuit primaire 2 en eau pressurisée comprenant la cuve 20.

The main elements of a primary REP circuit are shown in figure 1 :

• a reactor building 1 ensuring different functions including in particular a contribution to the containment safety function,
• a reactor vessel 20, located in the center of building 1, housing the reactor core C,
• a primary circuit 2 in pressurized water comprising the tank 20.

These main elements are therefore common, their constitution and the number of components varying depending on the power of the reactor.

Typically, the envelope of the reactor building 1 can be made up of several thicknesses. For example, as illustrated in figure 1 , a reactor building 1 can consist of an exterior reinforced concrete wall 12, an interior concrete wall prestressed 10 separated from the exterior 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.

• une cuve de réacteur 20,
• des boucles primaires 21 comprenant chacune une pompe primaire 22 et un générateur de vapeur 23,
• un unique pressuriseur 24.

As illustrated on the figure 2 , from publication [1], the primary circuit 2 is made up of the following main components:

• a reactor vessel 20,
• primary loops 21 each comprising a primary pump 22 and a steam generator 23,
• a single pressurizer 24.

In addition, we distinguish on this figure 2 , the reactor core control rod and control cluster mechanisms 25.

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

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

There Figure 3 illustrates the energy transfer cycle (heat then electricity) of a PWR reactor. On this Figure 3 , we can distinguish in particular the distribution of the positioning of the components in relation to the reactor building 1, which provides the function of a third containment barrier.

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

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

Among the solutions already studied for heat-producing PWR reactors, we can distinguish three main categories 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 Anglo-Saxon acronym SMR ( “Small Modular Reactor”), with an electrogenerating vocation. We can refer to publications [2] and [3].

A pool reactor has generally been implemented on an experimental basis with a low power, typically 10 MWth: the pressure in the primary circuit can be close to atmospheric pressure, implying both moderate neutron fluences within the core, and the temperature of the primary circuit is limited and close to 100°C maximum. In this pool reactor, the height of liquid water above the core, however, makes it possible to slightly increase the primary pressure within the fissile core, 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 an enclosure subject to pressure, the concrete vessel well and its watertight internal 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 tank shaft and its internal watertight casing. The reactor vessel itself is assembled on site, and must of course withstand all extreme external attacks, such as major earthquakes and plane 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 at present due to the strong constraints of demonstration of quality of construction and on-site control 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 the mechanical equipment of the nuclear islands of PWR reactors. A self-supporting metal tank, made in the factory, is preferred, and also allows rise to higher primary pressures.

An example of a pool reactor is represented by the SLOWPOKE project carried out in the 1970s-1980s by Canada (AECL) concerning NHP reactors (Anglo-Saxon acronym “Nuclear Heating Plants”). The figure on page 13 of publication [3] illustrates this demonstration reactor, with a thermal power output of 10 MWth, whose water-filled vessel well contains the entire primary circuit and the exchangers between the primary circuit and the secondary circuit. , the whole being closed by a slab at ground level. In this basin reactor, the circulation of the primary fluid is carried out by natural convection, which thus simplifies electrical loss transients, and general maintenance of the systems. The reactor also has significant thermal inertia, due to the importance of the water inventory in the primary circuit in relation to the power of the reactor, which also brings a gain in general safety with respect to -vis incidental transients, and driving. Finally, the thickness of the tank well and its absence of lateral or lower crossings ensures by design that it is impossible to drain the primary circuit, nor its depressurization. Such a pond reactor therefore meets many criteria of simplicity of design, safety in design, and ease of control.

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 envelope, typically of the order of one meter, coated in the internal part of 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 on site, which are paramount, are respected.

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

An example of a heat-producing loop reactor is the Chinese HAPPY 200 reactor from the company SPIC, dedicated to district heating in the city of Beijing. With a unit power of 200 MWth, it is designed in batches of two units, thus equaling the performance of the aforementioned CNNC DHR400 project. Unlike the latter, the reactor vessel is a self-supporting steel structure, designed in the factory and assembled on site, although with complete welding of the primary loop, thus requiring heavy site work. The thermal power is evacuated from the core via two primary loops supplying plate exchangers, by forced convection using two pumps. A pool of water surrounds the entire reactor vessel, but without direct contact due to a double envelope. In an accidental transient situation, this double envelope 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 of 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 primary and secondary circuits, which also allows a low thermal gradient between temperatures of the primary circuit and the secondary circuit.

The last category is that of so-called integrated reactors, which include a block delimited by a reactor vessel entirely produced in the factory and transported to site, and which houses the primary circuit in its entirety, and in particular the exchangers between 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 electrogenerating function, namely a configuration based on the integration of the steam generator, or even of 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.

The main advantages of SMR reactors compared to existing PWRs are that they allow simplification of systems, mainly for safety purposes, and an increased capacity for modularity through significant manufacturing of components in the factory for transport to the construction site.

In addition to the gain in compactness, integrated SMRs have the advantage of no longer requiring overhead fluid lines in pressurized water, which considerably reduces the risk of accidents and associated consequences linked to the rupture of the primary circuit lines. Thus installation on site is greatly facilitated by limiting itself to secondary piping connections, apart from the connections of the volumetric and chemistry system of the primary circuit, which are of small diameter.

For example, the nuclear power plant project with the acronym NUWARD ^™ , is a power plant with an electrogenerating vocation, made up of two integrated SMRs, with a unit power equal to 170MWe, each of which includes a block housing all the components of the primary circuit at inside the reactor vessel.

Other integrated SMR projects for electricity generation are in development or have been studied, among which we can cite the SCOR project with a power of 150 to 200 MWe in the name of the Applicant or the ACP100 project with a power equal to 100 MWe.

We represented on the figure 4 an example of an integrated SMR currently in project. Such an integrated SMR reactor whose block is generally designated under the numerical reference 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.

A precursor concept for an integrated heat-producing reactor is the THERMOS project carried out jointly by the Applicant and the company Technicatome: [4]. The reactor according to this project had a thermal power of 100 MWth and was intended to supply urban 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 tank reactor, necessary to provide urban heat close to 120°C. With a diameter of 5m and a height of 9m, the reactor vessel which was proposed was thus entirely assembled in the factory, and notably housed the exchangers between primary and secondary circuits in its upper part. The low temperature gradient in the core required the presence of primary fluid pumping groups, 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. power of the customer heating network.

Recent studies are dedicated to heat-generating reactors in Finland, carried out mainly by the VTT organization, for district heating in 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 confinement enclosure. close. The circulation of the primary fluid takes place by natural convection in nominal operation, and the secondary circuit is a loop liquid 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 precisely, the IRWST pool includes a pit forming the vessel well inside which the assembly consisting of the reactor vessel and the containment vessel is partially immersed. This IRWST pool also serves as a cold source for sizing accidents. This type of integrated reactor presents the same advantages of factory buildability and modularity, as those of integrated SMRs with generator function, in particular from the company NuScale Power.

Like other energies, there would be a considerable advantage in coupling a heat-generating nuclear reactor with a heat storage system in order to smooth out the heat production of the reactor, until it could be operated all the time. year at 100% of its power by using storage to manage the intermittent consumption of a heating network.

Patent applications/patents can be cited here CN210069996U , CN113790469A , CN205388908U , US20120314829A1 , JP2003270383A , US4294311 , FR2322242B2 , FR2257869B1 , CN210123170U And CN105509121B which raise the possibility of such a coupling.

In particular, the patent US4294311 evokes the possible storage of heat from a nuclear reactor of the PWR type in a pit filled with a media, of the PTES type (Anglo-Saxon acronym “Pit thermal Energy Storage”) with layers of different temperatures separated by membranes in polyurethane. Not only does such a pit seem expensive to implement but, moreover, no real details are given on the coupling between the reactor and the PTES pit.

There is therefore a need to improve the installations with a reactor envisaged as a heat-generating reactor and with a thermal storage system for the heat produced by the reactor, in order to overcome the disadvantages mentioned above.

The aim of the invention is therefore to respond at least in part to this need.

Presentation of the invention

• au moins un réacteur nucléaire à vocation calogène comprenant une cuve de réacteur;
• un puits de cuve logeant la cuve de réacteur, un circuit primaire et une partie d'un circuit secondaire;
• un circuit tertiaire comprenant :
  • une fosse de stockage thermique, creusée dans le sol, remplie d'eau et dans le fond de laquelle le puits de cuve est excavé ou posé, la fosse étant configurée pour réaliser une stratification thermique verticale entraînant la formation d'une thermocline délimitée entre le fond de la fosse à une température dite froide et le haut de la fosse à une température dite chaude,
  • au moins un échangeur de chaleur avec le circuit secondaire dont au moins une entrée est reliée fluidiquement à une zone de la fosse en-dessous de la thermocline et au moins une sortie est reliée fluidiquement à une zone de la fosse au-dessus de la thermocline de sorte que l'eau de la fosse circule en convection naturelle ou forcée entre l'entrée et la sortie de l'échangeur de chaleur;
• un réseau de chaleur dont au moins une entrée est reliée fluidiquement à une zone de la fosse en-dessous de la thermocline et au moins une sortie est reliée fluidiquement à une zone de la fosse au-dessus de la thermocline.

To do this, the invention relates, in one of its aspects, to a nuclear installation comprising:

• at least one heat-producing nuclear reactor comprising a reactor vessel;
• a vessel shaft housing the reactor vessel, a primary circuit and part of a secondary circuit;
• a tertiary circuit comprising:
  • a thermal storage pit, dug in the ground, filled with water and at the bottom of which the tank well is excavated or placed, the pit being configured to achieve vertical thermal stratification leading to the formation of a thermocline delimited between the bottom of the pit at a so-called cold temperature and the top of the pit at a so-called hot temperature,
  • at least one heat exchanger with the secondary circuit of which at least one inlet is fluidly connected to an area of the pit below the thermocline and at least one outlet is fluidly connected to an area of the pit above the thermocline so that the water in the pit circulates in natural or forced convection between the inlet and outlet of the heat exchanger;
• a heat network of which at least one inlet is fluidly connected to an area of the pit below the thermocline and at least one outlet is fluidly connected to an area of the pit above the thermocline.

By “vessel pit placed in the bottom of the pit”, we mean here and in the context of the invention, that the pit is not excavated as such to modify the altimetry of the different components of the reactor and the related exchangers so as to allow circulation by natural convection of tertiary water. Thus, a tank well placed on the ground involves forced convection circulation of tertiary water.

In order to better thermally insulate the pit from the ambient air, the pit is advantageously covered with a thermally insulating cover.

According to an advantageous alternative embodiment, the pit is artificially dug into the ground and includes a sealing coating, preferably thermally insulating, covering its bottom and its side walls.

Preferably, the covering is a membrane adapted to match the shape of the bottom and side walls of the pit.

The thermally insulating blanket or sealing coating may comprise one or more thermal insulation material(s) chosen from stainless steel, a polymer such as high density polyethylene (HDPE), or an elastomer.

• au moins une tuyauterie reliée à l'entrée du circuit tertiaire de l'échangeur de chaleur dont l'extrémité formant un collecteur de soutirage débouche directement dans la zone de la fosse en-dessous de la thermocline,
• au moins une tuyauterie reliée à la sortie du circuit tertiaire de l'échangeur de chaleur dont l'extrémité formant un collecteur d'injection débouche directement dans la zone de la fosse en-dessus de la thermocline,

According to an advantageous embodiment, the installation comprises:

• at least one pipe connected to the inlet of the tertiary circuit of the heat exchanger, the end of which forming a withdrawal collector opens directly into the zone of the pit below the thermocline,
• at least one pipe connected to the outlet of the tertiary circuit of the heat exchanger, the end of which forming an injection manifold opens directly into the zone of the pit above the thermocline,

• au moins une tuyauterie reliée à l'entrée du réseau de chaleur dont l'extrémité formant un collecteur d'injection débouche directement dans la zone de la fosse en-dessous de la thermocline,
• au moins une tuyauterie reliée à la sortie du réseau de chaleur dont l'extrémité formant un collecteur de soutirage débouche directement dans la zone de la fosse en-dessus de la thermocline.

chaque collecteur de tuyauterie comprenant un dispositif de répartition de débit hydraulique adapté pour élargir le débit d'eau entrant ou sortant dans les zones de la fosse au-dessous ou en-dessus de la thermocline.According to another advantageous embodiment, the installation comprises:

• at least one pipe connected to the inlet of the heat network, the end of which forming an injection manifold opens directly into the area of the pit below the thermocline,
• at least one pipe connected to the outlet of the heat network, the end of which forming a withdrawal collector opens directly into the area of the pit above the thermocline.

each pipe manifold including a hydraulic flow distribution device adapted to expand the flow of water into or out of areas of the pit below or above the thermocline.

• une tuyauterie reliant directement la partie de circuit secondaire logée dans le puits de cuve à une entrée de l'échangeur avec le circuit tertiaire;
• une tuyauterie reliant directement la partie de circuit secondaire logée dans le puits de cuve à une sortie de l'échangeur avec le circuit tertiaire.

According to yet another advantageous embodiment, the installation comprises:

• piping directly connecting the part of the secondary circuit housed in the tank well to an inlet of the exchanger with the tertiary circuit;
• piping directly connecting the part of the secondary circuit housed in the tank well to an outlet of the exchanger with the tertiary circuit.

The excavated or installed tank pit is preferably closed by a removable metal cover forming the separation wall with the volume of the pit.

The excavated or installed tank pit may include a concrete wall forming the separation wall with the ground and the volume of the pit.

The height He of the tank well excavated below the bottom of the pit is advantageously at least equal to 15m.

Advantageously, the reactor is a calogenous SMR type reactor.

Thus, the invention essentially consists of positioning a PWR type reactor at the bottom of a thermal storage pit delimiting a thermocline and thermally coupling them.

To do this, the tertiary circuit of the reactor is made up of the pit itself and one (or more) exchanger(s) with the secondary circuit whose inlet is fluidly connected in a cold temperature zone below the thermocline and the outlet is fluidly connected in a hot temperature zone above the thermocline so as to ensure natural or forced convection of the water in said tertiary circuit, whatever the season of the year and the call power of the heating network whose inputs and outputs are preferably connected respectively to the cold and hot temperature zones.

The heating network itself includes a circuit for pumping and discharging water and heat exchange with the network.

• du fait du couplage entre le réacteur calogène 4 et la fosse de stockage thermique de type PTES, la possibilité de faire fonctionner le réacteur à 100% de sa puissance toute l'année, à l'exception des phases de rechargement/déchargement du coeur, durant lesquelles la fosse permet de continuer à alimenter le réseau de chaleur. Le rendement économique de l'installation est donc maximal. L'adaptation de la production à l'intermittence de consommation de chaleur d'un réseau de chaleur de type urbain est gérée par le stockage thermique dans l'eau au sein de la fosse, dont le dimensionnement est adapté pour gérer cette intermittence, en particulier lors de changements de saison. La fosse de stockage thermique permet le décalage temporel d'une grande quantité de chaleur, typiquement plusieurs GW.h, produite l'été par le réacteur vers une consommation pendant l'hiver,
• une sûreté accrue du réacteur contre les agressions externes du fait de son positionnement en dessous de la fosse de stockage thermique, typiquement sous plusieurs mètres d'eau,
• la suppression des situations incidentelles de type « perte de refroidissement » avec la présence d'une source froide passive et quasi-infinie constituée par l'eau tertiaire au sein de la fosse, au-dessus du réacteur,
• une protection radiologique accrue par la présence d'une hauteur d'eau tertiaire importante au-dessus du réacteur, typiquement sous quelques mètres d'eau, de préférence au moins 4 mètres d'eau,
• le positionnement du réacteur nucléaire au moins en partie en dessous de la fosse de stockage thermique avec une hauteur d'eau importante au-dessus du réacteur permet, en cas d'évènement extrêmes conduisant à une perte prolongée de moyens électriques et de source froide normale, du type de l'accident survenu à Fukushima Daishi, d'avoir la proximité directe d'un réservoir d'eau (fosse remplie d'eau tertiaire) illimité en regard des besoins en refroidissement du réacteur en fonctionnement de puissance résiduelle. A titre d'exemple, l'eau à 1 bar de pression a une enthalpie de vaporisation de 2257 kJ/kg, soit pour un volume d'eau dans la fosse de l'ordre de 800 000 m³, une énergie thermique potentiellement évacuable par évaporation de 500 GW.h. En première approximation, en considérant une puissance résiduelle du réacteur de 1% de MWth, soit 200kWth, l'énergie potentielle de 500GW.h offre une capacité de refroidissement d'une durée d'environ 280 ans. Plus raisonnablement, en considérant une loi de puissance résiduelle pénalisante pour le coeur du réacteur dont la puissance nominale est de 20 MWTh, l'énergie thermique totale dégagée par les produits de fissions nucléaires pendant une période de 1000 ans est d'environ 28 GW.h, soit 6% de la réserve disponible d'une fosse remplie d'eau à l'ébullition.
• la possibilité de mutualiser les coûts d'excavation et de génie civil entre la fosse de stockage thermique et le puits de cuve du réacteur, et de minimiser l'empreinte au sol totale de l'installation,
• une augmentation de la hauteur motrice disponible pour la circulation par convection naturelle de l'ensemble des circuits du réacteur (primaire, secondaire et tertiaire), ce qui minimise voire annule de fait la puissance de pompage nécessaire dans ces circuits,
• la démonstration de l'élimination pratique de l'accident de fusion du coeur du réacteur à vocation calogène REP, dont l'origine proviendrait d'un manque prolongé de moyens de refroidissement du coeur du réacteur, car il est agencé au moins en partie sous une fosse de stockage thermique remplie d'eau,
• l'optimisation du fonctionnement neutronique du réacteur et de son pilotage grâce à son fonctionnement possible à 100% de sa puissance toute l'année, ce qui permet de supprimer les paliers de puissances intermédiaires qui sont nécessaires en cas de fonctionnement réduit puis relancé. Notamment, il n'y a plus d'effet xénon à contrôler, plus de contrainte liée à la prise en compte du phénomène d'Interaction Pastille Gaine (IPG) dans le pilotage du réacteur.
• une minimisation des fuites thermiques du réacteur puisqu'une partie d'entre elles peut être récupérée dans la fosse de stockage thermique, ce qui optimise d'autant l'efficacité énergétique d'ensemble de l'installation.

The advantages of the invention are numerous, among which we can cite:

• due to the coupling between the heat reactor 4 and the PTES type thermal storage pit, the possibility of operating the reactor at 100% of its power all year round, with the exception of the core reloading/unloading phases, during which the pit allows the heat network to continue to be supplied. The economic return of the installation is therefore maximum. The adaptation of production to the intermittency of heat consumption of an urban type heat network is managed by thermal storage in the water within the pit, the dimensioning of which is adapted to manage this intermittency, in especially during seasonal changes. The thermal storage pit allows the temporal shift of a large quantity of heat, typically several GW.h, produced in the summer by the reactor for consumption during the winter,
• increased safety of the reactor against external attacks due to its positioning below the thermal storage pit, typically under several meters of water,
• the elimination of incidental situations such as “loss of cooling” with the presence of a passive and almost infinite cold source constituted by tertiary water within the pit, above the reactor,
• increased radiological protection by the presence of a significant height of tertiary water above the reactor, typically under a few meters of water, preferably at least 4 meters of water,
• the positioning of the nuclear reactor at least partly below the thermal storage pit with a significant height of water above the reactor allows, in the event of extreme events leading to a prolonged loss of electrical means and normal cold source , of the type of the accident which occurred at Fukushima Daishi, to have the direct proximity of a water reservoir (pit filled with tertiary water) unlimited with regard to the cooling needs of the reactor in residual power operation. For example, water at 1 bar pressure has an enthalpy of vaporization of 2257 kJ/kg, i.e. for a volume of water in the pit of around 800,000 m ³ , potentially evacuable thermal energy by evaporation of 500 GW.h. As a first approximation, considering a residual power of the reactor of 1% of MWth, or 200kWth, the potential energy of 500GW.h offers a cooling capacity lasting around 280 years. More reasonably, considering a penalizing residual power law for the reactor core whose nominal power is 20 MWTh, the total thermal energy released by the products of nuclear fission over a period of 1000 years is approximately 28 GW. h, or 6% of the available reserve of a pit filled with boiling water.
• the possibility of pooling excavation and civil engineering costs between the thermal storage pit and the reactor vessel shaft, and of minimizing the total footprint of the installation,
• an increase in the driving height available for circulation by natural convection of all the reactor circuits (primary, secondary and tertiary), which minimizes or even eliminates the pumping power necessary in these circuits,
• the demonstration of the practical elimination of the melting accident of the core of the PWR heat-producing reactor, the origin of which would come from a prolonged lack of means of cooling the reactor core, because it is arranged at least in part under a thermal storage pit filled with water,
• the optimization of the neutron operation of the reactor and its control thanks to its possible operation at 100% of its power all year round, which makes it possible to eliminate the intermediate power levels which are necessary in the event of reduced operation then restarted. In particular, there is no longer any xenon effect to control, no more constraints linked to taking into account the Pellet Sheath Interaction (IPG) phenomenon in reactor control.
• minimization of thermal leaks from the reactor since part of them can be recovered in the thermal storage pit, which further optimizes the overall energy efficiency of the installation.

The installation as planned with its heat reactor is intended to supply urban heating networks, supply industrial processes such as desalination, desiccation, transformation of food products.

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 ultimate need, complementary to the conventional battery park solutions generally used. For reasons of simplicity of design and maintenance, it is not planned to install active means classified as emergency, of the diesel combustion type.

Other advantages and characteristics of the invention will become clearer on reading the detailed description of examples of implementation of the invention given for illustrative and non-limiting purposes with reference to the following figures.

Brief description of the drawings

• [ Figure 1 ] there figure 1 is a schematic perspective and partial section view of an existing PWR type nuclear reactor.
• [ Figure 2 ] there figure 2 is a schematic view of a primary circuit of a PWR type nuclear reactor according to the state of the art in a configuration with three primary loops.
• [ Figure 3 ] there Figure 3 is a schematic view of the three cycles of a PWR type nuclear reactor according to the state of the art.
• [ Figure 4 ] there Figure 4 is a schematic perspective view of an integrated type SMR reactor as it is currently envisaged.
• [ Figure 5 ] there Figure 5 is a schematic perspective schematic view of an SMR reactor with a heat-producing purpose according to the invention.
• [ Fig 5A ] there figure 5A is an exploded perspective view of the parts of the calogen SMR reactor vessel according to the invention.
• [ Figure 6 ] there Figure 6 is a longitudinal sectional view of the reactor according to the figure 5 , and which illustrates the circulation by natural convection of the water in the primary circuit and the secondary circuit.
• [ Fig 6A ] there Figure 6A is a detailed schematic view of a control cluster assembly, the control rod, and the rod control mechanism of a heat-producing SMR reactor according to the invention.
• [ Fig 7] [Fig 8 ] THE figures 7 and 8 are detailed perspective views showing the valve for regulating the water flow of the circuit respectively in an intermediate position and the fully open position.
• [ Figure 9 ] there Figure 9 is a perspective view of part of the reactor according to the figure 5 , and which shows in detail the installation of a valve for regulating the water flow of the secondary circuit.
• [ Fig 10A] [Fig 10B] [Fig 10C ] THE Figure 10A, 10B and 10C are perspective views showing an example of a valve for regulating the water flow of the secondary circuit 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 .
• [ Figure 11 ] there Figure 11 is a partial view in longitudinal section of the upper part of the reactor according to the figures 5 And 6 showing a pressurizer according to the invention.
• [ Figure 12 ] there Figure 12 is a schematic view in longitudinal section and in transparency of a nuclear installation according to the invention with a heat-generating reactor according to the figures 5 And 6 in a watertight tank well arranged below the bottom of a PTES type thermal storage pit filled with water which constitutes the water of the tertiary circuit.
• [ Figure 13 ], [ Figure 14 ] THE figures 13 And 14 illustrate alternative embodiments of flow distribution devices within the thermal storage pit of the installation according to the invention.
• [ Figure 15 ] there Figure 15 is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to the figures 5 And 6 In a watertight tank well, arranged below the bottom of a thermal storage pit whose water constitutes part of the tertiary circuit in natural convection.
• [ Figure 16 ] there Figure 16 is a schematic perspective and transparency view of a nuclear installation according to the invention with a heat-generating reactor according to the figures 5 And 6 in a watertight tank well, arranged on the bottom of a thermal storage pit whose water constitutes part of the tertiary circuit in forced convection.
• [ Fig 16A ] there Figure 16A is a detailed view of the Figure 16 showing the bottom of the tank well with the installation of a water circulation pump for the thermal storage pit.

detailed description

Throughout this application, the terms “vertical”, “lower”, “upper”, “lower”, “high”, “below” and “above” are to be understood by reference in relation to an SMR nuclear reactor , as it is provided in vertical operating configuration and whose tank well is on the bottom or excavated below the thermal storage pit according to the invention.

By “primary water”, “secondary water”, “tertiary water”, we mean the water which constitutes the fluid respectively of the primary, secondary and tertiary circuit.

THE figures 1 to 4 have already been detailed in the preamble, they will therefore not be commented on below.

Please note that the different temperatures, powers, volumes, flow rates, etc. indicated are for information purposes only. For example, other temperatures can be considered depending on the configurations in particular SMR reactor power, volume of water in the secondary basin, power requirement for the heating network, etc.

It should also be noted that the figures are not necessarily to scale. In particular, in the figure, the reactor pit 100 as well as the reactor 4 which is housed there are in reality much smaller relative to the thermal storage pit 9.

We describe with reference to figures 5 And 6 , a pressurized water type nuclear reactor 4, according to an integrated SMR type primary circuit configuration according to one embodiment of the invention.

This reactor 4 has a unit power of 20 MW thermal, for heat generation purposes, that is to say 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.

The reactor 4 of central axis vertical cylinder. This reactor vessel consists of a fixed compartment 40 and a removable compartment 41, above the reactor core for the fuel handling or maintenance phases of the reactor internals. This removable compartment 41 is a cover in the form of a dome 5 whose central chimney integrates a rolling valve 64 adapted for cooling the reactor pressurizer as detailed below.

The core C of the reactor comprises a set of fuel assemblies 42 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 locations lacking fuel rods, replaced by absorbent rods which can move up or down in the assembly to control the reaction. Data from preliminary studies carried out by the Applicant consider a number of 52 assemblies and a lifespan of 10 years, with a fissile height of 1.5 m.

The metal envelope 40 houses in its lower part a cylinder 43 supporting a basket of assemblies usually designated under the name "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.

Two flanges 45 are bolted between the fixed compartment 40 and the dome 41. The seal between a flange 45 and each of the two compartments 40, 41 is advantageously ensured by a metal seal. The dismantling of this flange 45 allows the complete handling of the fuel assemblies during the core reloading phases. The studies carried out by the Applicant provide for outages for fuel reloading scheduled for ten-yearly inspections, without intervention on the core between these periods.

Above the core C, rod cluster guides 46 allow the insertion of nuclear reactivity control rods 42, in a manner similar to what is usually encountered in conventional PWR reactors. The control bars 42 are rods made of neutron absorbing material.

The free volume above the reactor core C allows the completely extended positioning of the control rods 42, as well as the waiting position of so-called emergency absorbent rods, dedicated to the safety shutdown of the nuclear reaction.

Above the control bars 42, a perforated plate with holes 48 is fixed, allowing the free passage of the hot primary fluid coming from the core through the holes 480, as well as the rods for controlling the movement of the rolling ring 481.

At the periphery of the plate 48, a regulating valve 481 for the water flow of the primary circuit, also called flow lamination, is arranged. This valve is in the form of a rolling ring 481 which matches the interior periphery of the compartment 45 of the tank and extends over a height between the plate 48 and the primary water outlet openings, making it possible to adjust the flow rate of the latter.

This rolling valve 481 has the function of regulating the natural circulation flow of the water from the primary circuit passing through the openings 400 which constitute the inlets of the primary water collectors of the exchangers 49 between primary and secondary circuit. The positioning control of this regulation valve is carried out by an electric motor controlling the vertical movement of the control rod 482 linked to the crown 481, which is advantageously identical to those used by the control rods of the reactivity control bars 46 .

In an intermediate position, as shown in Figure 7 , the rolling ring 481 leaves the openings 400 partially clear, thus regulating the flow of primary water intended for the exchangers 49.

In the event of an electrical failure or emergency stop triggering, the gravitational fall of the bars 46 mechanically coupled to the absorbent rods 42 also triggers the gravitational fall of the primary fluid regulation valve 481. In the extreme low position, as illustrated in there figure 8 , this regulation valve 481 allows the water from the primary circuit to completely pass through the openings 400, thus making it possible to maximize the heat exchange in the exchangers 49 between primary and secondary circuits, in order to evacuate the residual power and cool the primary circuit. This operation by gravity fall of this regulation valve guarantees reliability and safety in the event of electrical loss or reactor failure.

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 along the different outlet openings 400 corresponding to the inlet collectors of the exchangers 49 between primary and secondary circuit and in an upper central portion of the core, in the form of a chimney. This non-detailed central chimney, called a riser, contains in addition to the control rod control mechanisms, the core parameter instrumentation sensors, and above the movement motors 47 of the reactivity control rods 42.

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 surrounding the core C inside the envelope 40, while the primary water at hot temperature, heated by circulating upwards in the core C, is found in the central portion upper part of the heart.

Above the outlet openings 400, within the reactor 4, a separation plate 7 separating the interior of the dome 41 from the tank containing a pressurizer, from the riser. This separation plate 7 contains a plate with through holes 70 making it possible to ensure the functions of thermal insulation and pressure differences of the integrated pressurizer. This separation plate can be of the type already described in the request for patent WO2012/158929 A3 .

The part above integrating the reactor pressurizer will be detailed later with reference to the Figure 11 .

After cooling through the exchangers 49 in a descending manner, the water from 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 towards the lower part of the reactor core . First on the periphery of the shell 40, in a zone conventionally called "down corner", then turns around at the bottom of the tank, crosses the core support basket 43, and will heat up through the core C. The circulation in closed circuit P only by natural convection of primary water is symbolized by the white arrows in figures 5 And 6 . The driving force of the primary circuit in natural convection is controlled by the difference in height between the average altimetric position of the exchangers 49 (primary cold source), and the average height of the fissile zone of the core C (primary hot source).

As already mentioned, the adjustment of the primary water pressure is carried out by the rolling valve 481 whose control mechanisms 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 conditions of neutron fluence, that is to say the thermal power of the core, by the positions of the reactivity control bars 42 in the core, and to the conditions of temperature and saturation pressure 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 valve for rolling the primary water flow which sets the conditions of exchange circulation between primary and secondary circuits in relation to the power produced at the core. In fact, the operation of the heat reactor can be controlled simply by controlling the pressure and temperature at the level of the pressurizer using the primary steam cooling system 6 and the heaters 8, and adjusting the circulation flow rate. primary natural evacuating the thermal power of the core by the rolling ring 481, and the adjustment of the power of the core by all of the control bars 42.

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

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

As shown in Figure 6 , the secondary circuit of this reactor 4 is not a closed loop circuit as in conventional PWR reactors, but includes a water basin B, as shown schematically in Figure Figure 11 . This basin B is contained in the space of the reactor well reactor forming the third containment barrier, and the reactor vessel 4 is immersed therein.

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

With such a pool of liquid water 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 improbable possibility of rupture of the primary water inlet or outlet pipes, thus causing a large diameter breach and loss of the latter, does not have significant accidental consequences for the reactor. . In fact, the liquid water pool 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.

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 on contact with the primary water within the exchanger 49, by natural suction from its inlet collector 490 at the bottom to their outlet collector 491 at the top thereof. The secondary water then creates a greater volume at 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 under the name of thermocline, as symbolized under the name thermocline 1 in Figure 6 .

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 submerged and the top of the basin at the warm temperature. It is the height of the thermocline layer which will set the cooling flow of the secondary circuit through the exchangers 49. The closed circuit circulation S only by natural convection of the secondary water is symbolized by the gray arrows in figures 5 And 6 .

The flow rate by natural convection of the secondary water is adjusted by regulating valves 5 integrated in each of the outlet collectors 491 of the exchangers 49, as illustrated in Figure 9 . The cold secondary water temperature is governed by the temperature conditions of the secondary water pool B. The hot temperature is set by the control valves in the outlet collectors 491, and by the heat exchange within the exchangers 49 themselves.

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

Advantageously, the end of the shaft 51 opposite that linked to the butterfly 50 is linked to an offset flyweight 53. As shown in Figure 10A , in the event of an electrical failure or emergency stop triggering, the gravity fall of the flyweight 53 places the valve 5 in its fully open position so as to circulate the maximum flow of secondary water from basin B.

During the start-up phase of the nuclear reactor, thermocline 1 is completely merged with the upper free level of secondary water basin B. The driving height of secondary water circulation is then maximum due to the maximum weight of the water column cold supplying the inlets of 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 temperature of the primary water leads to an average cooling of the moderator in the core, thereby inducing an increase in the reactivity of the core, and therefore an increase in its thermal power. Maximum conditions for thermal heating of the secondary water volume are accompanied by a natural increase in core power, reactor 4 is therefore naturally stable. As already mentioned, the position of the primary water rolling valve, 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 well.

Conversely, when thermocline 1 drops in level, this implies an elevation of the secondary hot water layer, and therefore a drop in the driving height of secondary water 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 exchange thermal between primary and secondary circuits. At constant neutron power in core C, the reduction in power evacuation leads to an increase in the average temperature of the primary water, and therefore an increase in the average temperature of the moderator in the core. There is therefore a decrease in reactivity through expansion of the moderator, and the neutron and thermal power produced decreases. The reactor is therefore naturally stable for evacuation and thermal storage towards the volume of secondary water defined by basin B.

The volume of secondary water is dimensioned 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, the volume of secondary water is of the order of 300 to 400 m ³ for the cold zone, and 100 to 150 m ³ for the hot zone, i.e. a total volume for basin B of between 400 and 550 m ³ .

The position of thermocline 1 can only be maintained at a fixed position on the condition that there is a continuous withdrawal of a quantity of the secondary water at its warm temperature and replacement by the same quantity of water secondary to its cold temperature. This continuous sampling is detailed below in relation to the Figure 12 , so that there is a balance between the thermal power produced by the reactor core, and its extraction by this sampling and replacement.

There is therefore an adjustable pumping system making it possible to transport, to a heating network, the power corresponding to the customer's demand, that is to say required by the heating network.

According to the invention, as detailed below, the heat evacuated by the reactor can be stored, before its transport to a heat network, by a thermal storage pit.

In the event of an untimely stoppage of this heat evacuation, or a fortuitous stoppage 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 and its temperature. cold and at its hot temperature, and by lowering the level of thermocline 1. After several minutes of operation, the continued evacuation of the thermal power produced by the reactor without an 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 thermal power of 50 MW, with secondary water supply at 90°C and return at 45°C, requires pumping of 270 liters per second, or 970 m ³ per hour from the hot water layer above above thermocline 1 and a return of the same quantity at the bottom of the reactor well. Preferably, this evacuation and this 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 lateral mechanical strength problems restricting expansion. or earthquake resistance.

The presence of this piping must not harm the transport of the entire reactor block, as detailed below for its exit from the reactor shaft using heavy handling means.

As previously stated, the primary circuit of the reactor operates solely by natural convection, that is to say without a pumping group.

Consequently, the inventors were faced with the problem of producing a pressurizer whose cooling and condensation part of the vapors of the primary fluid cannot be designed with a liquid water sprinkling/injection device, from a sampling of the primary circuit as according to the state of the art (due to the absence of primary pumping groups).

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, taking advantage of the fact that the metal envelope 60 of the cover 41 is thin, typically between 10 and 20 mm.

Thus, as illustrated in Figure 11 , the steam cooling and condensation part comprises a double-walled dome 6 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 around 1 to 3 cm.

In normal operation, the level of the thermocline 1 is fixed sufficiently above the pressurizer, in particular so as to be located above the central evacuation chimney 62, as illustrated in Figure 11 .

Thus, as illustrated in Figure 11 , 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 1, will condense the saturated vapor of the primary circuit inside of 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 50°C, which allows effective and rapid cooling of the dome 6, and in particular of the internal wall 60 forming the enclosure mechanical resistance to the pressure of the primary circuit, and thereby the primary water vapor underlying it.

The central chimney 62 integrates within it a regulation valve 64 or in other words rolling valve which makes it possible to adjust the flow of secondary liquid water which circulates in space E and therefore to regulate the liquid cooling as such. Indeed, in a completely closed position of the valve 64, the layer of water is trapped and stratified in the space E. Conversely, in a position of opening, in particular completely, of the valve 64, the water hot rises naturally in space E then through the central chimney 62 and will join the upper hot water layer of basin B, while the cold water from basin B is sucked in through the entry in the low position of the double wall 60, 61.

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

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

The external wall 61 of the dome 6 is advantageously covered with a cap 63 housing within it a thermal insulator.

According to an advantageous variant, the reactor 4 comprises, as a passive heat sink, a plurality of cooling fins 65 arranged inside the internal wall 60, being preferably distributed uniformly on the surface of the latter, of preferably 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.

Furthermore, the heating part of the pressurizer comprises a plurality of electrical resistances 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 allowing the functions of thermal insulation and pressure difference of the integrated pressurizer to be ensured. Such a portion with holes 70 is for example as according to the device described in the application patent WO 2012/158929A3 . The electrical resistances 8 can be of the type of those described in the patent US4135552 .

According to the invention, in order to smooth the heat production of reactor 4, until it can operate all year round at 100% of its power, the inventors have provided a thermal coupling between the latter and a thermal storage pit 9 filled with water delimiting a thermocline which will make it possible to manage the intermittent consumption of a heating network.

In addition, the inventors planned to position the reactor well 100 of the reactor 4 below the bottom of an artificially dug thermal storage pit 9, of the PTES type (Anglo-Saxon acronym for “Pit Thermal Energy Storage”), filled of water. As illustrated in Figure 12 , the positioning of the excavated tank pit 100 is in the middle of the pit 9 which is symmetrical in shape. We can refer to publication [6] which describes the creation of such a pit 9.

The bottom 90 and the side walls 91 of the pit are covered with a waterproofing coating in the form of a membrane (liner). For example, this membrane can be made of high density polyethylene (HDPE).

The pit 90 is covered with a thermally insulating cover 92 to limit thermal losses into the atmospheric air.

Such a PTES type pit 9 makes it possible to achieve vertical thermal stratification leading to the formation of a thermocline delimited between the bottom 90 of the pit at a so-called cold temperature and the top of the pit at a so-called hot temperature. This thermocline is designated as thermocline 2 in Figure 12 .

As an indicative example, we can consider a volume of the order of 800,000m ³ for a WPS of pyramid geometry with a truncated rectangular base with a small base (bottom of the storage) of 100m*100m, a large base (top of the storage) of 180m* 180m and a height of water of 40m. The cold and hot temperatures of the heating network can be respectively equal to 45°C and 90°C.

• une dynamique rapide de charge et de décharge thermique,
• lorsqu'elle est remplie avec de l'eau, une stratification significative et une densité de stockage élevée (64 kWh/m³ par exemple pour des températures chaude et froide de respectivement 90°C et 35°C).
• un coût de construction faible quand les caractéristiques géologiques sont réunies.

The main advantages inherent in such a thermal storage pit 9 can be summarized as follows:

• rapid thermal charge and discharge dynamics,
• when filled with water, significant stratification and high storage density (64 kWh/m ³ for example for hot and cold temperatures of 90°C and 35°C respectively).
• a low construction cost when the geological characteristics are combined.

Within pit 9 filled with water, the position of thermocline 2 changes depending on the season and the related atmospheric and ground temperatures. More precisely, the position of thermocline 2 evolves rather to the first order as a function of the energy loaded and discharged voluntarily from the reactor 4 and/or to the heat network 300. Then, to the second order, it degrades naturally due to first of all the thermal conduction between the hot layer and the cold layer of thermocline 2, and also because of thermal losses to the outside which can actually depend on the season and the temperatures of the atmosphere. We specify here that the ground temperature varies very little and only has a marginal effect.

We can refer to the publication [7] which details the progress of a thermocline in a thermal storage tank. Thus, at the end of the summer period, the layer of cold water within pit 9 is minimal and the thermocline is at its lowest, as symbolized by the dotted line on the Figure 12 .

According to the invention, as shown in Figure 9 , the tank well 100 excavated below the bottom 90 of the thermal storage pit 9 comprises a peripheral wall 101 of prestressed concrete internally coated with a metallic coating, preferably in the form of a stainless steel membrane. Such a covering has the function of serving as a rigid structure for the reactor vessel 40, 41, 42 forming the third confinement barrier, and the prestressed concrete wall 101 has the function of responding to the pressure and temperature conditions of secondary water, and ensuring the physical integrity of the third barrier.

As also shown in Figure 12 , the assembly consisting of the reactor vessel 40, 41, 42 forming a reactor block 4 to which the exchangers 49 are fixed is supported by a metal base 102, preferably made of stainless steel, or of black steel coated with an anti- corrosion. Preferably, this base is held in the bottom of the tank well 100 by a mechanical locking system, not shown, adapted to prevent its movement and its lifting in the event of an earthquake. However, during the fuel handling or component replacement phases, the base 102 and the reactor block 4 which it supports can be lifted and brought to the top of the reactor well. To do this, the mechanical locking system of the base must be able to be unlocked in a simple manner using tools accessible from the top of the tank well 100.

The base 102 which supports the reactor block 4 is rigidly connected to a foundation slab 103, and the metal covering with which the prestressed concrete wall 101 is covered is fixed rigidly and watertight to the slab 103. As already mentioned, the reactor 4 has a total physical impossibility of the occurrence of a serious accident, with significant melting of the core and perforation of the primary circuit tank. There is therefore no specific corium recovery device at the bottom of the reactor pit.

The tank well 100 is closed by a removable and waterproof cover cap 104, adapted to resist the pressure of the secondary circuit. Preferably, the cover cap 104 is a metal slab, mechanically welded, more preferably cellular or formed of boxes.

A gas sky 105, preferably nitrogen, is delimited by the free level of the volume of secondary water in the tank well 100. This gas sky 105 makes it possible to control the secondary water pressure, to adapt the free level variations and secondary water expansion, and to prevent the presence of oxygen. The mechanical connection between the metal lining 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 tightness.

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

According to the invention, the tertiary circuit of the reactor is constituted by the pit 9 itself and at least one exchanger 200 with the secondary circuit whose inlet 201 is fluidly connected in a cold temperature zone below the thermocline and the outlet 202 is fluidly connected in a hot temperature zone above the thermocline.

And a heat network 300 has at least one inlet 301 which is fluidly connected to an area of the pit below the thermocline and at least one outlet 302 fluidly connected to an area of the pit above the thermocline. These taps make it possible to guarantee stability of the layers of tertiary water at cold and hot temperatures and therefore of the thermocline within pit 9.

More precisely, as illustrated in Figure 12 , a pipe 210 is connected to the inlet of the tertiary circuit of the heat exchanger 200 and its end 211 forming a withdrawal collector opens directly into the zone of the pit below the thermocline.

A pipe 220 is connected to the outlet of the tertiary circuit of the heat exchanger and its end 221 forming an injection manifold opens directly into the zone of the pit above the thermocline.

By considering the altitude of the withdrawal collector 211 in the zone below the thermocline, that is to say the lowest cold layer of the water of the pit 9, whatever the season, we guarantee a minimum driving height of water necessary for the establishment of natural convection of water in pit 9, that is to say in the tertiary circuit of the installation. The closed circuit circulation T only by natural or forced convection of tertiary water is symbolized by the dotted arrows in Figure 12 . It is specified that this circulation T is shown in pit 9 for the purposes of clarity but that in reality the rising branch of the convection passes through the pipe 220 and that the descending branch does not exist as such, since it is a natural piston which advances over the entire section of thermocline 2.

In addition, a pipe 310 is connected to the inlet 301 of the heat network and its end 311 forming an injection manifold opens directly into the area of the pit below the thermocline.

A pipe 320 is connected to the outlet 302 of the heat network and its end 321 forming a withdrawal collector opens directly into the area of the pit above the thermocline.

• une tuyauterie 230 débouche directement dans l'eau secondaire du puits de cuve 100 et est reliée à une entrée 203 de l'échangeur 200;
• une tuyauterie 240 reliée à une sortie 204 de l'échangeur 200 débouche directement dans l'eau secondaire du puits de cuve 100.

For the part of the secondary circuit which is connected to the exchanger 200, the installation includes:

• a pipe 230 opens directly into the secondary water of the tank well 100 and is connected to an inlet 203 of the exchanger 200;
• a pipe 240 connected to an outlet 204 of the exchanger 200 opens directly into the secondary water of the tank well 100.

The reactor installation configuration in storage pit 9 as shown in figures 12 And 15 allows circulation by natural convection. To do this, as shown, the tank well 100 is installed partly below the bottom 90 of the swimming pool 9, that is to say with a tank well 100 excavated to a height He such that we can do without a pump, the pipe 210 supplying water from the pit 9 at its cold temperature opening into the bottom 90 of the pit.

For a reactor with a power equal to 20MW, with a thermal storage pit 9, the excavation height He can be between 16 and 18m.

We can also consider an installation configuration of the reactor in pit 9 as shown in Figures 16 and 16A according to which the tank well is in some way placed on the bottom 90 of the thermal storage pit 9. Such a configuration involves circulation by forced convection of the tertiary water by means of pumps 270, advantageously arranged in the concrete wall 101 of tank well 100.

The inventors carried out thermo-hydraulic pre-sizing calculations on all of the primary, secondary and tertiary circuits of the installation according to the Figure 12 from thermomechanical calculation software. It may be a classic thermo-hydraulic software known under the name CATHARE:[8].

Considering a pinch of 15°C for the exchangers 49 between primary and secondary circuit and an identical pinch for the secondary/tertiary exchanger 200, and a supply of hot water to the network 300 at 90°C, the calculations give a temperature average at the outlet of reactor core C of 120°C.

We recall here that an exchanger pinch 49 is the minimum temperature difference between the primary water and the secondary water at a point of the given exchanger.

With a temperature gradient of 40°C between the inlet and outlet of the reactor core, we obtain the thermo-hydraulic operating parameters making it possible to pre-size the exchangers 49, and an estimate of the positioning of the latter vis-à-vis of the heart.

The following table 1 summarizes these different parameters for a 20 MWTh reactor. [Table 1] Settings Values Primary water pressure 4 bars Core inlet temperature C 80°C Core outlet temperature C 120°C Primary water flow 118.6 kg/s Secondary water pressure 5 bars Temperature T1 at exchanger inlet 49 65°C Temperature T2 at exchanger outlet 49 105°C Secondary water flow 119.1 kg/s Driving height of water balance between the core and the exchangers 49 3.16m

From these data, an estimated volume of exchangers 49 can be established, with installation at a minimum height of 3.16 m relative to the average altitude of the fissile zone. The primary water pressure of 4 bars is determined such that the boiling margin with respect to the average core outlet temperature 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 integrating 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.

A number of three identical exchangers 49 is retained with for each a useful heat exchange volume of approximately 1.2 m ³ and an exchange height between primary circuits and secondary of 2m. The three exchangers 49 are fixed with their collectors 400, 401, 490, 491 at 120° from each other to the compartments 40, 42 of the reactor vessel.

In fact, the thickness of the reactor vessel is not dictated by considerations of resistance to pressure since it is immersed in the secondary water of basin B, but rather by constraints of mechanical rigidity, resistance to buckling and heart support. An average thickness of around 20 mm is therefore retained. We recall here that the primary pressure is lower than the secondary pressure.

The material envisaged for the reactor vessel 4 and the exchangers 49 is stainless steel.

Computer-aided design (CAD) gives a secondary water volume of approximately 200 m ³ for its cold temperature T1 and approximately 100 m ³ for its hot temperature T2.

The average temperature of secondary water is around 85°C, which is lower than the boiling temperature of water in the atmosphere.

The volume of primary water is of the order of 30 m ³ , or approximately a factor of 10 lower than that of the volume of secondary water.

This therefore results in a very significant thermal inertia contributing strongly to the operational safety of the reactor block 4, 102, both in terms of absorption of possible heating of the primary water, but also with regard to the rise in primary water pressure in the event of a failure to evacuate the residual power.

In addition, the consequences of a Loss of Primary Cooling Accident (APRP) are greatly reduced by the configuration of the secondary water completely enveloping the primary circuit.

The following table 2 illustrates the characteristics of the thermo-hydraulic sizing of secondary and tertiary circuits. The CAO indicates a number of three 200 interchanges, which could be reduced to two, for reasons of economic optimum. [Table 2] Settings Values Temperature T1 at exchanger inlet 49 65°C Temperature T2 at exchanger outlet 49 105°C Secondary water flow 119.1 kg/s Secondary water pressure 5 bars Temperature T3 at inlet 201 of exchangers 200 35°C Temperature T4 at outlet 202 of exchangers 200 90°C Tertiary water flow 86.9 kg/s Natural convection equilibrium driving height S of secondary water 2.7m

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

A number of three identical exchangers 200 are retained, each with a useful heat exchange volume of approximately 1.1 m ³ and an exchange height between secondary and tertiary circuits of 1.7 m.

The pinch in an exchanger 200 is increased compared to that of an exchanger 49 due to the temperature difference between the layer of cold water at 35°C in pit 9, and the layer of hot water at 90° vs.

The driving height of natural convection of secondary water is at least 2.7 m, which sets the maximum altitude position of thermocline 1. We can consider a height of approximately 3.5 m in order to take margins of sizing concerning pressure losses in piping, and having adjustment margins by modulating the flow of tertiary water.

The minimum driving height necessary for the establishment of natural convection of tertiary water corresponds to the difference in altimetry between the median position of an exchanger 200, and the position of the separation thermocline between cold layer and hot layer in pit 9. This thermocline evolves seasonally and at the end of the summer period, before the supply of heat to the network 300 is started, the cold water layer is minimum and as is already apparent from the above, the conservative value chosen is the altitude of the withdrawal collector 211 of the tertiary water. The pre-sizing carried out is based on the minimization of pressure losses allowing the establishment of natural convection at moderate driving height. The minimum calculated height is of the order of 2 m, which allows compact installation, despite the cumulative conditions of natural convection of the primary, secondary and tertiary circuits which lead to relatively deep burial of the reactor block 4.

The level of the invert 103 of the tank pit is located approximately 50 m above the ground, when the tank pit 100 is completely excavated, that is to say buried completely below the bottom 91 of the pit 9. In the CAD design, the bottom of pit 9 is 32 m from the ground surface. In the hypothesis of a deeper pit 9, for example 40 m, the depth of the raft would then be closer to 60 m, because of the need to establish natural convection of tertiary water.

Water injections into thermal storage pit 9 and withdrawal to reactor 4 are planned to have constant flow rates all year round corresponding to operation of the reactor at 100% power.

The outward and return flow rates of the heating network 300 are variable depending on the power demand of the network. Variations in hydrostatic head in pit 9 during the year may require the addition of a device to regulate the withdrawal flow to the reactor to keep the latter constant.

• si le débit vers le réseau de chaleur 300 est inférieur au débit d'injection/soutirage du réacteur 4, alors la thermocline se déplace vers le fond 90 et le stockage thermique est en mode de charge, c'est-à-dire que l'énergie thermique qui y est stockée augmente,
• si le débit vers le réseau de chaleur 300 est supérieur au débit d'injection/soutirage du réacteur 4 alors la thermocline se déplace vers le haut et le stockage est en mode de décharge, c'est-à-dire que l'énergie thermique qui y est stockée diminue,
• si le débit vers le réseau de chaleur 300 est égal au débit d'injection/soutirage du réacteur 4, alors la thermocline ne se déplace pas, et l'énergie thermique stockée reste constante.

The flow differential between the injection/withdrawal of the reactor 4 and the back and forth of the heat network 300 in the pipes 310, 320 leads to the movement of the thermocline towards the top or towards the bottom of the pit 9, as follows:

• if the flow rate to the heat network 300 is lower than the injection/withdrawal flow rate of the reactor 4, then the thermocline moves towards the bottom 90 and the thermal storage is in charge mode, that is to say that the the thermal energy stored there increases,
• if the flow rate to the heat network 300 is greater than the injection/withdrawal flow rate of the reactor 4 then the thermocline moves upwards and the storage is in discharge mode, that is to say that the thermal energy which is stored there decreases,
• if the flow rate to the heat network 300 is equal to the injection/withdrawal flow rate of reactor 4, then the thermocline does not move, and the stored thermal energy remains constant.

To optimize the stability of thermocline formation, the different collectors for withdrawal 211 or injection 221 of tertiary water or withdrawal 321 and injection 311 of the heat network 300 can be produced according to a preferred configuration.

It may advantageously be a hydraulic flow distribution device adapted to expand the flow of incoming or outgoing water, of the type known under the designation of a pavilion as illustrated in figures 13 And 14 for injection of tertiary water at hot temperature.

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.

Other variants and embodiments can be considered without departing from the scope of the invention.

On the Figure 12 , for the sake of clarity, a single exchanger 200 has been illustrated. It goes without saying that the number of exchangers 49 between primary and secondary circuits and exchangers 200 between secondary and tertiary circuits is only illustrated for information purposes and will be adapted according to the configurations of the installation required for the supply of hot water to a heating network 300.

Also, for the sake of simplification and clarity, the exchanger 200 between secondary and tertiary circuit is shown in Figure 12 inside the pit 9. Preferably, the exchanger(s) 200 is(are) arranged outside the secondary water volume, more preferably behind the tank well 100 in the concrete wall 101, as illustrated in Figures 16 and 16A , with an isolation valve 212 on each inlet and outlet pipe, in order to be able to isolate the secondary circuit and complete the sealing of the third radioactive containment barrier.

Furthermore, if the heat-producing reactor 4 described is provided with a circulation of primary water and secondary water only by natural convection, we can envisage any PWR reactor with water flow rates in the primary and secondary circuits generated by pumps, that is to say according to forced convection or by a thermosyphon may be suitable for coupling by its tertiary circuit in natural convection with a thermal storage pit filled with water as described previously.

List of cited references

• [1]: The World Nuclear Industry Status Report 2017. https://www.worldnuclearreport.org/IMG/pdf/20170912wnisr2017-en-lr.pdf .
• [2]: IAEA-TECDOC-397 “Potential of Low-temperature Nuclear Heat Applications” .
• [3]: IAEA-TECDOC-463 “Small Reactors for Low-temperature Nuclear Heat Applications” .
• [4]:https://inis.iaea.org/collection/NCLCollectionStore/Public/10/494/10494922.pdf
• [5]:https://www.ecosmr.fi/wp-content/uploads/2021/06/Leppanen EcoSMR_15062021.pdf
• [6]: “Design Aspects for Large Scale Aquifer and Pit Thermal Energy Storage for District Heating and Cooling”. International Energy Agency Technology Collaboration Program on District Heating and Cooling including Combined Heat and Power - Integrated Cost-Effective large-scale thermal energy storage for smart district heating and cooling. Draft - September 2018 .
• [7]: Kaloudis E. & al., “Large eddy simulations of turbulent mixed convection in the charging of a rectangular thermal storage tank”. International Journal of Heat and Fluid Flow 44 (2013) 776-791 .
• [8]: G. Geffraye et al. “CATHARE 2 V2.5 2: A single version for varying applications” Nuclear Engineering and Design 241 (2011) 4456-4463 .

Claims (13)

Nuclear installation including: - at least one heat-generating nuclear reactor (2) comprising a reactor vessel (40, 41, 42); - a vessel well (100) housing the reactor vessel, a primary circuit and part of a secondary circuit; - a tertiary circuit including: • a thermal storage pit (9), dug in the ground, filled with water and at the bottom of which the tank well is excavated or placed, the pit being configured to achieve vertical thermal stratification resulting in the formation of a thermocline (thermocline 2) delimited between the bottom of the pit at a so-called cold temperature and the top of the pit at a so-called hot temperature, • at least one heat exchanger (200) with the secondary circuit of which at least one inlet (201) is fluidly connected to an area of the pit below the thermocline and at least one outlet (202) is fluidly connected to a area of the pit above the thermocline so that the water in the pit circulates in natural or forced convection between the inlet and outlet of the heat exchanger; - a heat network (300) of which at least one inlet (301) is fluidly connected to a zone of the pit below the thermocline and at least one outlet (302) is fluidly connected to a zone of the pit above above the thermocline. Nuclear installation according to claim 1, the pit being covered with a thermally insulating cover (92). Nuclear installation according to claim 1 or 2, the pit being artificially dug into the ground and comprising a sealing covering, preferably thermally insulating, covering its bottom (90) and its side walls (91). Nuclear installation according to claim 3, the coating being a membrane adapted to match the shape of the bottom and side walls of the pit. Nuclear installation according to one of claims 2 to 4, the thermally insulating cover or the sealing covering comprising one or more thermal insulation material(s) chosen from stainless steel, a polymer such as high density polyethylene (HDPE), or an elastomer. Nuclear installation according to one of the preceding claims, comprising: - at least one pipe (210) connected to the inlet of the tertiary circuit of the heat exchanger, the end of which forming a withdrawal collector (211) opens directly into the zone of the pit below the thermocline, - at least one pipe (220) connected to the outlet of the tertiary circuit of the heat exchanger, the end of which forming an injection collector (221) opens directly into the zone of the pit above the thermocline, Nuclear installation according to one of the preceding claims, comprising: - at least one pipe (310) connected to the inlet of the heat network, the end of which forming an injection collector (311) opens directly into the area of the pit below the thermocline, - at least one pipe (320) connected to the outlet of the heat network, the end of which forming a withdrawal collector (321) opens directly into the area of the pit above the thermocline. A nuclear installation according to claim 5 or 6, each pipe manifold comprising a hydraulic flow distribution device adapted to expand the flow of water into or out of areas of the pit below or above the thermocline. Nuclear installation according to one of the preceding claims, comprising: - a pipe (230) directly connecting the part of the secondary circuit housed in the tank well to an inlet (203) of the exchanger with the tertiary circuit; - a pipe (240) directly connecting the part of the secondary circuit housed in the tank well to an outlet (204) of the exchanger with the tertiary circuit. Nuclear installation according to one of the preceding claims, the excavated or installed reactor pit being closed by a removable metal cover (104) forming the separation wall with the volume of the pit. Nuclear installation according to one of the preceding claims, the excavated or installed reactor pit comprising a concrete wall (101) forming the separation wall with the ground and the volume of the pit. Nuclear installation according to one of the preceding claims, the height He of the reactor shaft excavated below the bottom of the pit being at least equal to 15m. Nuclear installation according to one of the preceding claims, the reactor being a calogenous SMR type reactor.

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