Reactor and safety method for a reactor against a core meltdown event

Abstract

The invention relates to a safety method for a reactor (1) comprising at least: a first circuit (100) of water-based first level fluid intended to circulate therein; a second circuit (200) of water-based second level fluid intended to circulate therein, the second circuit (200) being hydraulically isolated from the first circuit (100) and comprising at least one steam generator (210). In the event of a core (30) of the reactor (1) melting and forming a pool (71) of core melt (70) in a bottom (12) of a vessel (10), in response to detecting the formation of a layer (72) of liquid metal at a surface of the pool (71) of core melt (70): the method provides for placing the second circuit (200) in fluid communication with the first circuit (100) so that the second level fluid follows the first circuit (100) to flow inside the vessel (10) past the layer (72) of liquid metal of the pool (71) of core melt (70).

Description

Technical Field

[0001] This invention relates to the field of nuclear power plant safety, and more specifically to the management of severe accidents in pressurized water nuclear reactors (PWRs). More specifically, the invention applies to the management of accidents in which a pool of molten core forms at the bottom of the container, following a severe accident, within the context of strategies for holding or retaining the molten core within the container. Background Technology

[0002] PWR type nuclear power plants, such as Figure 1 and Figure 2 As shown.

[0003] Typically, such a reactor station includes a containment vessel 600, within which reactor 1 is housed. Reactor 1 (the reactor) includes a vessel 10, which, together with a cap 20, forms a tight outer shell. This tight outer shell houses the fuel assemblies of the reactor core 30. Figure 2 In the image, for clarity, the reactor core is represented by a dashed area.

[0004] The container 10 also includes at least one so-called cold fluid inlet 13 connected to the first-stage loop 100 and at least one so-called hot outlet 14 also connected to the first-stage loop 100. Pressurized water circulates in the first-stage loop 100.

[0005] Therefore, during normal operation of reactor 1, the first-stage loop 10 ensures the transfer of heat from the core 30 to the second-stage loop 200, and water also circulates within the second-stage loop.

[0006] The second-stage loop 200 includes one, and preferably multiple, steam generators 210. Heat exchange between the first-stage loop 100 and the second-stage loop 200 occurs within the steam generators 210. Figure 1 In the example shown, loop 214 of the first-stage loop 100 is shown in the steam generator 210.

[0007] Figure 1 and Figure 2 The simplified diagram illustrates other perfect classic elements of the first-stage loop 100 and reactor 1 core 30, such as: control rod 40, hydraulic pump 102 located at the inlet of container 10 to circulate water in the first-stage loop 100, pressurizer 110 located at the outlet of container 10, containment 600, and structures forming a threshold 601 and wall 602 that define the well 603 of container 10.

[0008] During normal operation of reactor 1, as it enters steam generator 210, a portion of the water in the second-stage loop 200 evaporates and reaches an energy conversion device, such as a turbine 220 movably driven by the depressurization of steam. Subsequently, the mechanical energy at the output of turbine 220 is converted into electrical energy by generator 500 and transformer 510 before being transferred toward a consumption location.

[0009] At the outlet of turbine 220, the fluid is condensed in condenser 230 before being re-injected into steam generator 210 by pump 240.

[0010] To condense the steam from turbine 220, condenser 230 is cooled by an open loop 300 supplied by a water source 310, such as a river. The loop 300 also includes booster pumps 320 and 330 and a cooling tower 340.

[0011] like Figure 1 The diagram clearly shows that the water in the first-stage loop 100 is designed to extract the calories generated by reactor 1 so that they can be transferred to the second-stage loop 200 to be converted into electrical energy using energy conversion systems such as turbine generators. During normal operation, the core 30 is critical, and the generated thermal power is discharged by the first-stage loop 100. Monitoring of the core's criticality is ensured by the position of control rods 40 within the core 30, but also by the content of water-soluble boron in the water of the first-stage loop.

[0012] Different accident scenarios, such as large-break first-stage coolant loss accidents, combined with aggravating factors such as failure of the graded emergency response system, may lead to serious accident situations.

[0013] Subsequently, with the formation of core melt, the core 30 transitions to a molten state. Core melt is part or all of the molten fuel stock within the vessel's internal components 10, which is in a molten state. Under the influence of gravity, the core melt flows into the bottom 12 of the vessel 10 and forms a molten pool therein. Therefore, the core melt comprises all or part of the molten fuel feedstock, containing all solid fission products and their associated residual heat energy.

[0014] Figures 3A to 3C The image illustrates a severe accident involving a PWR container in a very illustrative manner, showing the formation of a molten core pool followed by the container being punctured.

[0015] exist Figure 3A The diagram illustrates the depolymerization and partial melting of the core 30, the formation of the core melt 70, and the appearance of a molten pool 71 of the core melt 70 at the bottom 12 of the vessel 10. In a known manner, during the formation of the molten pool 70, a surface liquid metal layer 72 appears on the free surface of the pool 71. A portion of the thermal power of the molten pool 70 is transferred to this layer 72.

[0016] Figure 3B The appearance of the metal layer 72 is shown in a very schematic manner (its thickness is intentionally exaggerated for clarity). The area marked with reference numeral A shows the metal layer 72 beginning to penetrate the container 10. This penetration of the container 10 is caused by a phenomenon commonly referred to as the "focusing effect." The metal layer 72 is a good thermal conductor and absorbs most of the heat energy of the molten core pool 71. The focusing effect corresponds to the situation where the metal layer 72 transfers a portion of its heat to a small surface of the sidewall 11 of the container 10 through conduction. This heat power focused on the small surface can cause the penetration of the wall 11 of the container 10.

[0017] Figure 3C The completion of puncturing container 10 is shown. The molten core 70 then overflows into well 603 of container 10. Furthermore, an explosion known as a steam explosion (sudden decompression of the water and expansion of the steam) occurs when the molten core comes into direct contact with the cooling water contained in well 603, producing hydrogen gas. These consequences are unacceptable considering the risk of rupture of the third containment barrier formed by the reactor shell 600. In the case of a saturated well 603 of container 10, the risk of container puncture should be eliminated. Furthermore, the uncontrolled spread of molten core 70, a highly radioactive material with very high thermal power, must be avoided. The spread of molten core 70 at the threshold level 601 poses a risk of puncturing the latter and propagating downwards into potential saturation areas.

[0018] Depending on the designers of the PWER nuclear boiler, there are two modes of managing the core melt in the event of a severe accident. A first-range solution, known as Ex Vessel Retention, involves allowing the core melt 70 to puncture vessel 10—where the primary loop is depressurized—and flow into areas used for spreading the core melt and managing its cooling. A drawback of this type of solution is the need for an impressive platform to receive and spread the core melt, known as a regenerator, which significantly increases the weight of the structure to be built and the infrastructure costs of the reactor shell. Another major drawback involves proving the integrity of the third containment barrier formed by the reactor shell to avoid environmental contamination, since the containment barrier formed by the primary loop has been breached.

[0019] The second range of solutions, known as In Vessel Retention or its acronym IVR, involves deploying a system to retain the core melt 70 within the vessel 10 by avoiding puncturing it.

[0020] In solutions within this range, one strategy aims to retain the core melt 70 within the vessel 10 and extract its remaining power through the walls of the vessel 10 via external cooling, particularly by setting up a (natural or forced) convection loop after soaking the well 602 of the vessel 10.

[0021] For example, this type of solution is described in document WO2009 / 053322. In this document, a pump located at the bottom of the container allows for increased forced convection of water in the container well 603 that is in contact with the outer wall of the container 10.

[0022] It is also possible to set up a natural circulation system for water that comes into contact with the container and then evaporates and condenses again at the top of the outer shell. (At Westinghouse) TM In the event of a serious accident involving the company's AP1000 reactor, this type of solution is recommended for managing the discharge of residual power. This cooling device is entirely passive compared to the solution described in document WO2009 / 053322.

[0023] Furthermore, based on the assumptions considered regarding the thickness of the metal layer 72 at the origin of the focusing effect, the possibility that insufficient cooling performance may not be achieved to avoid container puncture still exists. This is why, in general, even in the context of IVR management for severe accidents, supplementary devices for handling hydrogen risks are described, along with studies on the limitations of the consequences of vapor explosions, and additional means are provided to handle the worst-case scenario of container puncture.

[0024] The IVR strategy described in document FR2763168 involves providing a device for recovering the molten core material at the bottom of the vessel. The molten core material flows from the core to the regenerator under gravity. Additionally, it is provided for gravity-assisted injection of water through an additional tank connected to the vessel well.

[0025] The drawback of this solution is the need for an additional system formed by the regenerator, which increases the weight and size of the reactor vessel. Furthermore, the device for injecting water into the vessel formed by a tank directly connected to the first-stage loop is merely a supplement to existing backup injection devices for handling dimensional accidents. The residual first-stage pressure present in the event of a severe accident necessitates providing a considerable gravitational height for this tank, significantly increasing the weight of the structure to be constructed.

[0026] Therefore, all the few solutions proposed for managing the cooling of the molten core have shortcomings. There is a need for a solution that limits and potentially mitigates at least some of these shortcomings. This invention aims to achieve this objective.

[0027] Another object of the present invention is to reduce and possibly eliminate the risk of puncturing the container through the focusing effect. Summary of the Invention

[0028] To achieve at least one of these objectives, the present invention provides a safe method for a pressurized water nuclear reactor, wherein, in the event that the reactor core is at least partially melted and a core melt pool is formed, the reactor during the operation phase comprises at least:

[0029] • A first-stage loop in which the water-based first-stage fluid is intended to circulate is configured such that the first-stage fluid permeates into the reactor vessel and through the core contained within the vessel in order to extract heat generated by the core.

[0030] • A water-based second-stage fluid is intended to circulate in a second-stage loop, which is hydraulically isolated from the first-stage loop and includes at least one steam generator. The second-stage loop is configured to absorb heat from the first-stage loop and convert it at least partially into steam in the steam generator.

[0031] The method includes at least the following steps. In response to an event that at least partially melts the reactor core, forming a core melt pool at the bottom of the vessel and a liquid metal layer at the surface of the core melt, the method provides to configure a second-stage loop in fluid communication with a first-stage loop such that the second-stage fluid follows the first-stage loop, preferably via a cold branch side, to flow through the liquid metal layer in the core melt pool within the vessel.

[0032] Therefore, the pressurized second-stage fluid, typically water at saturation pressure and temperature, contained in the second-stage loop and especially in the steam generator, flows into the bottom of the container above the metal layer.

[0033] Therefore, this coolant injection is performed passively because the water in the second-stage circuit is at a higher pressure than that in the first-stage circuit.

[0034] Typically, the pressure of the water contained in the steam generator is in the range of 60 to 70 bar, while in the event of a severe accident, the pressure of the water in the first-stage circuit is usually below 20 bar.

[0035] Within a very short time, the water in the second-stage loop begins to come into contact with the overmolten liquid metal layer, causing a sharp reduction in heat flow at the puncture origin via the focusing effect. All or part of the water from the second-stage loop flows out advantageously along the volume of the reactor vessel on the cold branch side, commonly referred to as a downcomer, and evaporates upon contact with the liquid metal layer and the molten core pool.

[0036] The water from the second-stage loop flows for a sufficiently long time to cool the liquid metal layer, at least for the entire time period, provided that the liquid metal layer is thin enough to pose a risk of puncturing the container. Therefore, this flow can be considered monitored. However, the flow rate through the liquid metal layer does not need to be very high.

[0037] One way to manage this flow rate and the duration of water injection from the second-stage loop into the container is to accurately calibrate the cross-section of the open gap between the second-stage loop and the first-stage loop.

[0038] As a non-limiting example, the flow rate of the second stage of water entering the container corresponds to a slit with a diameter of about 20 mm, and more typically between 10 and 30 mm.

[0039] Then, all the water contained in one or more steam generators (GVs) is gradually flowed into the reactor vessel. A typical liquid flow rate from the second-stage loop to the first-stage loop is approximately 5 kg / s. As a non-limiting example, this flow rate is more generally between 2 kg / s and 10 kg / s. This value is significantly lower than all safe injection water intended for saturating the reactor core. Therefore, the method according to the invention allows for the long-term transfer of water from the GV to areas that could potentially puncture the vessel through a focusing effect, and also significantly limits the steam generated by the water / liquid metal interaction, thereby limiting overpressure in the first-stage loop.

[0040] Again, as a non-limiting example, the duration of water injection from the secondary loop into the container ranges from three hours, and more generally from 30 minutes to 5 hours. In contrast, in a solution that soaks the reactor core by pouring water from the secondary loop into the container, the injection time would be approximately one minute to several minutes.

[0041] Depending on the unforeseen circumstances, for a reactor with a power of 1300 MWe (French Palier N4 type), the available water capacity per steam generator varies between 29 tons and 70 tons per steam generator.

[0042] Regarding the timescale of a severe accident, particularly the timescale of core meltdown and the formation of the core melt, when implementing the method according to the invention, the time for water from the second-stage loop to flow through the core melt pool is very short.

[0043] However, in the context of the development of this invention, it has been observed that this duration is sufficient to significantly reduce or even eliminate the risk of puncturing the container through the focusing effect.

[0044] In fact, it has been noted that the timeframe during which the focusing effect may lead to vessel puncture is relatively limited. It has been observed that the puncture risk due to the focusing effect corresponds to a time interval in which the molten core pool has a relatively thin layer of liquid metal on its surface, typically a few centimeters thick. In this configuration, most of the thermal power from the molten core pool is transferred to this thin metal overlay, and then this power is transferred through contact with the inner wall of the vessel, causing the latter to gradually puncture.

[0045] As time progresses, the increasing stock of molten metal within the gradually melting vessel continues to supply the molten core pool, and the surface layer of liquid metal thickens. The contact surface between the liquid metal layer and the vessel's inner wall increases. The heat transferred to the latter is then distributed across the thicker liquid metal layer, and the power to puncture the vessel becomes less concentrated. The risk of vessel puncture thus decreases.

[0046] Once all the internal metal stock has melted, the thickness of the surface metal layer makes the heat power at the periphery and in contact with the container insufficient to puncture it. Typically, at this level, external cooling of the container is sufficient to effectively dissipate the heat power from the molten pool. Puncture of the container is then definitively stopped.

[0047] For example, if 3 MWh is transferred to a metal layer with a thickness of 10 cm and a perimeter of 12 m, and there is no possibility of heat dissipation from the upper surface, then there will be 3 / 1.2 = 2.5 mW / m. 2 The heat flow is applied to the container. When the container is submerged in the molten water, this flow cannot be discharged by conventional external cooling. The excess heat power transferred to the container is then converted into the molten heat of the container itself, resulting in a puncture from the inside. Once the liquid metal layer reaches two or three times its original thickness, the flow decreases by such a large margin that external cooling becomes sufficient to stop the puncture process, as all the heat power from the molten pool is discharged to the external water without causing the container metal to melt.

[0048] When the second-stage liquid water comes into contact with the liquid metal layer from the cold branch of the first-stage loop, preferably at the puncture site of the container, i.e., around the downcomer, as will be described in detail later, each kilogram of water evaporated per second causes a cooling of the liquid layer of approximately 2 to 3 MWth. Cooling the liquid metal layer at a flow rate of approximately 5 kg / s over a period of approximately 3 hours corresponds to the removal of approximately 10 to 15 MWth of heat power over 3 hours, which greatly reduces or even prevents container puncture.

[0049] Furthermore, once the molten core pool reaches a given height, the surface layer of liquid metal will come into contact with the core support plate, such as... Figure 2 and 3AAs shown in Figure 17, the temperature of the liquid metal is superheated by the thermal power of the core melt and then significantly reduced due to the substantial melting of the lower portion of the core support plate. In fact, the melting point of the metal is much lower than the melting point that the liquid metal layer might reach, thus leading to a focusing effect.

[0050] Therefore, in the context of the development of this invention, it has been noted that by injecting a certain amount of water into the container at a relatively low flow rate, container puncture due to the focusing effect can be effectively counteracted. The volume of water contained in the steam generator is then large enough to effectively cool the liquid metal layer throughout the entire time period when the liquid metal layer has a thickness small enough to cause container puncture due to the focusing effect.

[0051] In a particularly advantageous manner, the present invention does not require any additional components in the reactor, such as a regenerator, or additional systems in the container well, such as forced convection devices.

[0052] Furthermore, this invention does not require any additional tanks placed at a considerable height. However, by utilizing the pressure of the second-stage circuit, this invention allows for very rapid water injection.

[0053] However, in configurations that keep the core melt inside the container, it is necessary to provide a cooling container for the outer wall and to cool it by fully immersing the container.

[0054] Therefore, the proposed solution allows for passive operation, i.e., no pumps pour coolant onto the core melt pool, enhancing the strategy for managing the core melt within the vessel by counteracting the effects of focusing effects that could jeopardize the success of preventing puncture of the reactor vessel.

[0055] Furthermore, the present invention does not provide additional components or connectors directly on the reactor vessel, which could potentially reduce the safety and reliability of the reactor.

[0056] Therefore, the present invention allows for a significant improvement in the safety of pressurized water reactors when the core melts and a molten pool is formed.

[0057] The present invention also relates to a pressurized water core reactor, which comprises at least:

[0058] • A container for housing the reactor core, the container comprising at least one inlet and at least one outlet.

[0059] A first-stage loop, with at least one end connected to the inlet of the container and at least one end connected to the outlet of the container, allows a first-stage fluid, preferably water-based, circulating in the first-stage loop to permeate into the reactor container through the inlet and exit through the outlet, while simultaneously passing through the reactor core to extract heat generated by the core.

[0060] • A second-stage loop is fluidly isolated from the first-stage loop, the water-based second-stage fluid is intended to circulate in the second-stage loop, and the second-stage loop includes at least one steam generator, the second-stage loop being configured to absorb heat from the first-stage loop and convert it at least partially into steam in the steam generator.

[0061] The reactor includes a safety system comprising a safety device configured to form a channel, preferably only one channel, which inhibits fluid insulation between the second-stage loop and the first-stage loop and sets the second-stage fluid present in at least one steam generator in fluid communication with the first-stage loop, such that the second-stage fluid contained in the steam generator flows at the bottom of the container while passing through the first-stage loop beforehand.

[0062] Therefore, if the safety device is activated when a molten pool forms in the core and a layer of liquid metal appears on the surface of the pool, the safety device is activated and the pressurized second-stage fluid contained in the steam generator is poured into the first-stage loop, advantageously into the cold branch, and then into the container. This fluid cools the liquid metal layer, thereby reducing the focusing effect. Thus, container puncture is avoided.

[0063] In a particularly advantageous manner, it should be noted that the proposed solution does not provide direct tapping of the second-stage loop on the reactor vessel, which improves reactor safety.

[0064] This device possesses the advantages mentioned above regarding the method of the present invention. In particular, it allows for a significant improvement in reactor safety without complicating or reducing the latter's reliability level.

[0065] According to a non-limiting example, the component is selected from the following valves: valves, manually operated valves, and remotely controllable valves.

[0066] S breach It is the minimum cross-section of the channel, which is capable of injecting the second-stage fluid present in at least one steam generator into the first-stage loop.

[0067] According to the non-restrictive example, S breach Less than 20cm 2 (10 -2 rice).

[0068] Based on an example, S breach Greater than 2cm 2 And preferably S breach Greater than 3cm 2 According to an example, S breach Between 2cm 2 Up to 20cm 2 between.

[0069] Based on an example, S breach Between 0.2cm 2 Up to 20cm 2 Between, preferably between 0.8cm 2 Up to 20cm 2 Between, more preferably between 2cm 2 Up to 7cm 2 between.

[0070] The inlet cross-section on the cold branch side of the reactor vessel is much larger than the cross-section of the rift formed between the second-stage loop and the first-stage loop.

[0071] Typically, based on non-restrictive examples, S breach <0.05*S inlet Preferably S breach <0.01*S inlet Preferably S breach <0.005*Sinlet, and preferably S breach <0.001*S inlet S inlet It is the minimum cross-section of the passage leading from the first-stage fluid in the first-stage loop to the container.

[0072] If the first-stage loop includes multiple inlets in the vessel, i.e., there are multiple steam generators, then the cross-section S inlet It is the sum of the cross sections from the first-level loop up to the container inlet.

[0073] Typically, the cross-sectional area of ​​the first-stage fluid channel on the cold branch side is 6000 cm². 2 Within the range.

[0074] Typically, when S inlet When it has a circular cross-section, S inlet The diameter is between 800 and 900 mm (10 -3 Between m).

[0075] Therefore, the cross-section through which the second-stage fluid flows into the first-stage loop is much smaller than the cross-section through which the first-stage fluid typically flows within the container. Cross-section S breach and S inlet This ratio allows for the injection of a second-stage fluid flow into the first stage and thus into the container over a relatively long period of time. More specifically, a sufficiently long period of time, for the duration during which the thickness of the liquid metal layer covering it is small enough to pierce the container wall through a focusing effect.

[0076] These features allow for a reduction and may suppress the risk of puncturing the container through the focusing effect. Attached Figure Description

[0077] The objects, objectives, features, and advantages of this invention will become more apparent from the detailed description of embodiments thereof, which are illustrated in the following figures, wherein:

[0078] Figure 1 The PWR type nuclear power plant is described.

[0079] Figure 2 The diagram schematically shows a vertical cross-section of a PWR reactor vessel in its vessel well under operating conditions excluding serious accidents.

[0080] Figures 3A to 3C This schematically illustrates the process leading to partial or complete core melting, the formation of a molten core pool, and subsequent puncture through a focusing effect. Figure 2 Different stages of a severe accident in the vessel of the reactor shown.

[0081] Figure 4 The invention is illustrated schematically in which the core melt pool is cooled by injecting water from the second-stage loop through the first-stage loop to the bottom of the container.

[0082] Figure 5 The container portion and parameters that allow the fuse to be positioned along the mother face (busbar) of the container are shown in a very schematic manner.

[0083] Figure 6 An example of a portion of a conventional steam generator is shown.

[0084] Figure 7 The hydraulic connection between the container and the steam generator according to a first embodiment of the present invention is shown in an enlarged view.

[0085] Figure 8 yes Figure 7 A schematic cross-sectional view of a variant of the illustrated embodiment.

[0086] Figure 9 The arrangement of a power plant with an integrated safety system according to a third embodiment of the present invention is illustrated schematically.

[0087] The accompanying drawings are provided by way of example only and do not limit the invention. They represent block diagrams intended to facilitate understanding of the invention and do not necessarily reflect the scale of actual applications. In particular, the relative dimensions of the different components of the station, especially the relative dimensions of the reactor and its piping, the liquid metal layer, and the components of the different parts of the station, do not represent reality. Detailed Implementation

[0088] Before beginning a detailed examination of embodiments of the invention, it should be remembered that the invention according to its first aspect may particularly include optional features that can be used in combination or alternatively below.

[0089] According to one example, the detection of the formation of a liquid metal layer at the surface of the core melt pool is performed using at least one fuse disposed on the container wall, the at least one fuse being configured to melt when the liquid metal layer reaches it.

[0090] According to one example, at least one fuse has a melting temperature of the fuse that is higher than or equal to a temperature threshold Tf, wherein Tf ≥ 400°C, preferably Tf ≥ 500°C, and preferably Tf = 600°C.

[0091] According to one example, the reactor includes a plurality of fuses distributed along at least one parent surface of the vessel wall, such that two adjacent fuses define a vessel section with a volume V. slice They are the same.

[0092] According to one example, a setup is provided to connect a second-stage circuit to a first-stage circuit based on the detection that the temperature of the inner wall of the container is higher than a temperature threshold Tf, wherein Tf is higher than 400°C, and preferably higher than 500°C.

[0093] According to one example, the curve depicting the evolution of the height of the liquid metal layer in the container is determined by a fuse disposed on the inner wall of the container, and preferably disposed according to at least two mother faces of that wall. The time point at which the second-stage circuit is configured to be in fluid communication with the first-stage circuit is determined based on this curve.

[0094] Preferably, a series of fuses are disposed on the inner surface of the vessel wall. More preferably, the fuses are placed on the mother surface of the vessel bottom and sidewalls. The progression of the core melt level rise and the onset of vessel degradation by the liquid metal layer at its surface are detected by the continuous deactivation of the fuses on this mother surface. Fluid communication is initiated from a determined core melt height.

[0095] According to one example, the reactor includes at least one fuse on the vessel wall. The fuse is configured such that when the liquid metal layer reaches the fuse, the liquid metal layer melts the fuse. For example, the melting temperature of the fuse is higher than or equal to a temperature threshold Tf, where Tf ≥ 350°C, preferably Tf ≥ 450°C, and preferably Tf = 600°C.

[0096] According to one example, the reactor includes a plurality of fuses distributed along at least one mother-face of the vessel wall. The fuses are arranged along the mother-face such that if the volume of the liquid metal layer increases at a constant rate, the time interval between two consecutive fuses separating the liquid metal layer reaching the mother-face remains constant.

[0097] According to one example, at least the liquid metal layer has a thickness e that is small enough to at least partially pierce the inner wall of the container.72 Throughout the entire time period, the second-stage fluid flows through the liquid metal layer within the container.

[0098] According to one example, the second-stage fluid flows through the liquid metal layer within the container for at least thirty minutes, and preferably at least one hour, and preferably at least two hours.

[0099] As an example, the second-stage water flow passes through a channel in the first-stage loop, the minimum cross-section of which is S. breach Between 0.2cm 2 (0.2.10 -4 m 2 The value is between 0.8 cm and 20 cm², and preferably between 0.8 cm². 2 Up to 7cm 2 Between. If the minimum cross-section S of the channel breach If it is circular, its diameter is between 5mm and 50mm, and preferably between 10 and 30mm. Typically, the diameter is about 20mm.

[0100] According to one example, the second-stage fluid operates at a rate below 10 kg / s (10 3 The flow rate is (g / s) and preferably less than 7 kg / s within the container (10).

[0101] According to one example, the reactor includes an inner sheath located inside the vessel, which surrounds the core and, together with the inner wall of the vessel, defines an annular volume called a downcomer, which is configured such that during normal operation of the reactor:

[0102] The inlet opens to the outside of the envelope and into the downcomer, allowing the first-stage fluid from the inlet to be directed to the bottom of the container.

[0103] The outlet leads into the envelope, allowing the first-stage fluid present in the reactor core to exit the reactor through the outlet.

[0104] The reactor is configured such that when the explosive device creates at least one channel that inhibits fluid insulation between the second-stage circuit and the first-stage circuit, the second-stage fluid contained in the steam generator then flows into the bottom of the container, while passing through the container's inlet before passing through the downcomer.

[0105] This implementation facilitates the outflow of the second-stage water onto the inner wall of the container. This allows for more efficient cooling of the liquid metal layer compared to the case where the second-stage fluid flows from the container outlet or the hot branch side and therefore does not permeate into the container through the downcomer. In effect, by flowing out through the downcomer side, the evaporation of the second-stage water and thus the cooling of the liquid metal layer occur in the area where the container is punctured. To achieve this implementation, the connection on the second-stage side should be made on the cold branch side of the steam generator, i.e., on the heat exchanger side of the cold return (cold water tank) where the first-stage temperature corresponds to the first-stage circuit.

[0106] According to one example, the reactor includes at least one fuse disposed on the container wall, the fuse being configured such that when a liquid metal layer reaches the fuse, the liquid metal layer melts the fuse, the melting temperature of the fuse being higher than or equal to a temperature threshold Tf, wherein Tf ≥ 400°C, preferably Tf ≥ 500°C, and preferably Tf = 600°C.

[0107] According to one example, the reactor includes a plurality of fuses distributed along at least one parent surface of the vessel wall, such that two adjacent fuses along the parent surface define a vessel section with a volume V. slice They are the same.

[0108] According to one example, a steam generator includes a sheath enclosing a second-stage fluid and a first-stage fluid, which encloses an insulating material that fluidly isolates the second-stage fluid and the first-stage fluid. A safety system is configured to suppress the insulation between the second-stage fluid and the first-stage fluid within the sheath of the steam generator, thereby forming the channel. Suppression of this insulation corresponds to a slit with a limited and monitored diameter (typically 20 mm).

[0109] • According to one example, the steam generator includes an outer enclosure comprising a first portion enclosing a first-stage fluid and a second portion enclosing a second-stage fluid, the first and second portions being fluidly isolated from each other.

[0110] The safety system includes at least one pipe located outside the steam generator, forming the passage, and having at least:

[0111] -The first end leading to the second section of the confined second-stage fluid.

[0112] - The second end of the branch located between the steam generator and the vessel, leading to the first-stage loop.

[0113] The safety device includes at least one component, which is mounted on the pipeline and selectively has:

[0114] - A closed configuration in which at least one component prevents fluid from passing through the passage.

[0115] - At least one of the components allows fluid to pass through the pipe, thereby enabling the second stage fluid of the steam generator to flow through the pipe to join the first stage loop and then to the open configuration of the container.

[0116] • As an example, the second end of the pipe forms a tap on the piping system line that connects to the branch of the first-stage loop.

[0117] • According to one example, the branch of the first-stage loop extends between the steam generator and the inlet of the vessel.

[0118] • According to one example, the reactor includes a device selected from a safety injection line (IS) and a volume and chemical control loop (RCV), the device being configured to lead to a first-stage loop at the second end of the pipeline.

[0119] • According to one example, a steam generator includes a first section enclosing a first-stage fluid and a second section enclosing a second-stage fluid, the first section and the second section being fluidly isolated from each other.

[0120] The reactor also includes an RRA device for cooling the reactor when it stops. The RRA device includes at least one first loop, which includes a heat exchanger and branches that fluidly connect the heat exchanger to one or more portions of the first loop.

[0121] According to one example, the safety system includes at least one pipe located outside the steam generator, forming the passage and having at least:

[0122] -The first end leading to the second section of the confined second-stage fluid.

[0123] - The second end of the branch of the first circuit leading to the RRA device.

[0124] The safety device includes at least one component, which is mounted on the pipeline and selectively has:

[0125] - A closed configuration in which at least one component prevents fluid from passing through the channel.

[0126] - At least one of the components allows fluid to pass through the pipe, thereby enabling the second stage fluid of the steam generator to flow through the pipe to join the branch of the first circuit, then the first stage circuit, and then the open configuration of the container.

[0127] This implementation method has the advantage of avoiding any taps in the first-stage loop. This further improves the safety of the reactor.

[0128] • According to one example, the safety device is configured such that the second-stage fluid contained in the steam generator flows in the container at a flow rate between 4 and 5 kg / s, and the pressure of the steam generator is in the range of 68 bar.

[0129] Therefore, even with a limited amount of cooling fluid, the focusing effect can be mitigated, preventing container puncture.

[0130] The terms “basically,” “approximately,” and “within the range” mean taking into account both manufacturing and / or measurement tolerances, and may specifically correspond to “within 10%.”

[0131] In the following description, normal operation of reactor 1 or the station refers to the operation phase without accidents or serious accidents. Accidents such as injection of first-stage coolant loss, large cracks, or extra-large cracks do not represent the normal operation phase of reactor 1.

[0132] Now refer to Figures 4 to 9 The present invention will be described in detail.

[0133] Figure 4 Describe reactor 1, for example, with reference Figures 2 to 3A The same type of reactor described.

[0134] Reference Figures 2 to 3A All the features described apply Figure 4 The embodiment shown. In this Figure 4 In this reactor, the core 30 is molten or partially molten. A molten core pool 71 has formed in the bottom 12 of the container 10. On the surface of the molten pool 71, a layer of liquid metal 72 has formed or is about to form.

[0135] Before the liquid metal layer 72 begins to pierce the container 10 or quickly after the partial piercing begins, cooling fluid is poured into the bottom 12 of the container 10 and thus onto the layer 72.

[0136] The cooling fluid comes from the inlet 13 and / or outlet 14 of the first-stage loop 100.

[0137] As will be explained in detail later, the cooling fluid consists of water flowing from the second-stage loop 200 in the first-stage loop 100.

[0138] It should be noted that, typically, in the event of a severe accident, the depressurization of the first-stage circuit 100 is triggered. This can be achieved by opening a specific valve located, for example, at the top of the pressure booster 110. This depressurization of the first-stage circuit may be triggered once a threshold temperature is reached, such as the temperature of the component sleeve reaching 650°C or higher. This depressurization of the first-stage circuit 100 results in its pressure being lower than that of the second-stage circuit 200. Typically, the pressure in the depressurized first-stage circuit 100 is below 20 bar. The pressure difference between the first-stage 100 and the second-stage 200 circuits causes the connection of these circuits 100 and 200 to result in a rapid injection of fluid from the second-stage circuit 200 into the first-stage circuit 100.

[0139] According to a particularly advantageous embodiment, fluid 800 from the second-stage loop 200 overflows into the first-stage loop 100 and reaches the interior of the vessel 10 through inlet 13. Preferably, the reactor 1 includes an inner sleeve 15 located inside the vessel 10, surrounding the core 30, and defining an annular volume, commonly referred to as a downcomer 16 (the downcomer portion), together with the inner wall 11 of the vessel 10. This inner sleeve 15 is configured such that during normal operation of the reactor 1 (i.e., in the absence of a serious accident):

[0140] • Inlet 13 leads to the outside of the sleeve 15 and to the downcomer 16, so as to guide the cold fluid from inlet 13 upward to the bottom 12 of the container 10.

[0141] • Outlet 14 leads to the interior of the envelope 15, allowing the hot fluid present in the core 30 to exit from the reactor 1 through outlet 14.

[0142] Therefore, during normal operation of reactor 1, the cold fluid of the first-stage loop 100 permeates into reactor 1 through inlet 13; descends by gravity in downcomer 16 to reach the bottom 12 of the vessel, rises within the sleeve 15 as it passes through the perforated plate commonly referred to as support plate 17; passes through core 30 to extract heat from fission and exits from reactor 1 through outlet 14.

[0143] In the context of this invention, the cooling fluid from the second-stage loop 200 and permeating into container 10 via loop 100 also descends along the wall 11 of container 10 and reaches the liquid metal layer 72. Thus, this safe cooling fluid follows the natural path of the water in reactor 1. The cooling fluid contacts the surface of the liquid metal layer 72. More specifically, the cooling fluid reaches the liquid metal layer 72 at the most critical point, i.e., at the interface between the latter and the wall 11 of container 10. Therefore, the cooling fluid ensures cooling function over the entire perimeter of the inner wall 11 of container 10 where the liquid metal layer 72 might puncture through a focusing effect. Therefore, the supply of cooling fluid from the downcomer 16 provides a particularly effective solution to reduce the risk of puncturing the container through a focusing effect.

[0144] Advantageously, this mode of contact between liquid water and the overmolten metal layer 72 is achieved by flowing out from the inner wall of the vessel 10, which is much softer than injecting large amounts of water onto the molten core pool. Injecting large amounts of water onto the molten core pool can cause steam shock, which is detrimental to the integrity of the reactor vessel.

[0145] exist Figure 4 In this context, the cooling fluid is represented by a body 802 spread out on the free surface of the metal layer 72. Naturally, the cooling fluid evaporates upon contact with the metal layer 72 before the latter has been sufficiently cooled.

[0146] It should be noted that the pressure-reducing valve of the first-stage circuit 100 is preferably kept open to discharge vapors generated when the cooling fluid from the second-stage circuit 200 comes into contact with the liquid metal layer 72. Furthermore, the pressure reduction in the first-stage circuit 100 facilitates the injection of cooling fluid from the second-stage circuit into the container 10.

[0147] Therefore, this cooling fluid allows the liquid metal layer 72 to have a thickness e that is thin enough to concentrate the thermal power of the molten pool 72 on a very small surface. 72 This allows it to pierce the inner wall 1110 of the container while cooling the liquid metal layer 72.

[0148] Continue pouring this cooling fluid until the thickness e of the liquid metal layer 72 is reached. 72 The power is large enough to allow the thermal power of the layer 72 to be transferred over a larger surface area, so that the power per unit surface area is low enough to prevent puncture of the inner wall 11 of the container 10.

[0149] like Figure 4 As shown, it is also used to cool the outer wall of container 10. For this purpose, well 603 of container 10 can be soaked, i.e., water can be injected or poured between container 10 and well 603. This cooling is usually sufficient in the case of IVR (entrapment within container) accidents, but naturally not if a focusing effect occurs.

[0150] In fact, cooling from outside the container via well 603 in immersion container 10 allows for the extraction of, for example, 1 megawatt per square meter (1 MW / m²). 2 Under the focusing effect, this cooling is no longer sufficient, because it is necessary to extract 1.5 MW / m³ from the area where the focusing effect causes the vessel 10 to puncture. 2 Even 2MW / m 2 .

[0151] According to a non-limiting example, for soaking the well 603 of container 10, water contained in a tank, such as water from a fuel loading pool, can be used. This tank can be used in the construction of the reactor or externally to the latter. Preferably, at least a portion of the tank should be arranged high enough relative to container 603 to allow water to flow into the container by gravity. In most cases, at least a portion of the tank should be located above the cap or cover 20 of reactor 1.

[0152] According to one embodiment, the flow rate of the cooling fluid from the second-stage loop 200 is not monitored. Instead, once the pressure of the second-stage loop 200 and optionally the pressure of the first-stage loop 100 are known, the flow rate can be easily modeled. This mainly involves the initial water level in the steam generator 210 and the cross-sectional area S of the passage between the second-stage loop 200 and the first-stage loop 100. breach The cooling duration was determined. Calculations showed that a fairly limited portion of the total stock in the steam generator 210 was sufficient to adequately cool the overmolten metal layer 72 and avoid puncturing the container 10, while the metal layer 72 at the surface thickened sufficiently.

[0153] Typically, the safety device is configured such that the channel cross-section S of the passage between the second-stage circuit 200 and the first-stage circuit 100 is... breach This allows the second-stage water to enter container 10 at a flow rate of less than 10 kg / s, and preferably less than 7 kg / s. Typically, this flow rate is between 4 and 5 kg / s for the initial pressure in a steam generator (GV) in the range of 68 bar (i.e., before one or more channels are opened toward the first-stage loop 100).

[0154] This allows the liquid metal layer to be cooled sufficiently for a long enough period of time to avoid puncturing the container.

[0155] According to a non-limiting example, in order to monitor this cooling time, a cross-section S can be provided. breach Fine calibration.

[0156] Based on an example, S breach Less than 20cm 2 (10 -2 (meters). Preferably, S breach Greater than 2cm 2 According to an example, S breach Between 2cm 2 Up to 20cm 2 Between. Advantageously, it lies between 2cm. 2 Up to 7cm 2 between.

[0157] S inletIt is the minimum cross-sectional area of ​​the passage between the first-stage loop 100 and the inlet 13 of the vessel 10. Therefore, it typically consists of the minimum cross-sectional area through which the first-stage fluid passes during normal reactor operation. For example, S inlet This corresponds to the cross-section of inlet 13 in the container. This cross-section is located in... Figure 8 As shown in the diagram. If container 10 has multiple entrances to the first-level loop 100, for example, Figure 7 As shown, S inlet It is the sum of all the entries in container 10.

[0158] S breach It is the cross-section of the channel, or the sum of the cross-sections of the channels when multiple channels exist, thereby setting the second-stage fluid present in at least one steam generator 210 to be in fluid communication with the first-stage loop 100.

[0159] The inlet cross-section on the cold branch side of the reactor vessel is much larger than the cross-section of the opening formed between the second-stage loop and the first-stage loop. Typically, according to a non-limiting example, S breach <0.05*S inlet And preferably S breach <0.01*S inlet And preferably S breach <0.005*S inlet .

[0160] Typically, the cross-sectional area of ​​the first-stage fluid channel on the cold branch side is 6000 cm². 2 Within the range.

[0161] Typically, the pressure of the water contained in the steam generator 210 is in the range of 60 to 70 bar. Consequently, the first-stage loop 100 is depressurized. In practice, in the event of a severe accident, a device for opening valves at the level of the pressurizer 110 is actuated to depressurize the first-stage loop 100. This allows for the prevention of pressure ejection of core fission products in the event of a vessel puncture. Furthermore, this depressurization of the first-stage loop allows for the facilitation of the injection of the second-stage loop within the vessel 30.

[0162] In most scenarios that lead to serious accidents, the second-stage loop is shut down and isolated, partly by closing the steam injection line to the turbine and partly by using an atmospheric vent valve.

[0163] According to one example, the safety device is triggered by the operator. To determine the point in time when the fluid in the second-stage loop 200 should be poured into the first-stage loop 100, it is advantageous to estimate the height of the core melt pool and, preferably, the evolution curve of that height.

[0164] For this purpose, one or more fuses 900 may be provided on the wall of the vessel 10. These fuses are configured to melt when a threshold temperature Tf is applied to them. Typically, this temperature Tf is reached when the core melt forms in the vessel 10 and comes into contact with the fuse 900. When the temperature inside the vessel corresponds to normal operation of the reactor, the fuse 900 will not melt. According to one example, Tf > 400°C, preferably Tf ≥ 500°C, and preferably Tf ≥ 600°C.

[0165] When fuse 900 melts, it prevents electrical signals from passing through. Therefore, the resistance of the circuit integrated into this fuse is infinite.

[0166] A fuse consists of a core made of conductive material and an electrically insulating sheath. This prevents a short circuit between the metal container and the conductive core.

[0167] For example, the core is made of metals such as aluminum or antimony, which have a melting point close to 600°C. For example, the insulating sheath is made of ceramic.

[0168] For example, the fuse forms a cable with two ends connected to the safety device and a bend located between these two ends. The bend corresponds to the lowest point of the fuse. Therefore, when the fuse changes from a conductive configuration where current flows from one end to the other in the core (resistance R1) to a non-conductive configuration where current no longer flows from one end to the other in the core (resistance R2 > R1, preferably infinite R2), it means that the core melt pool has melted the bend. Therefore, it can be deduced that the height of the free surface of the core melt pool corresponds to the height of the bend of the fuse 900 relative to the bottom of the container 10.

[0169] It has been proven that using fuses is more robust and reliable than using temperature sensors.

[0170] Preferably, the fuse is positioned on the inner wall 11 of the vessel 10. This allows for enhanced reliability in detecting the focal effect. In practice, by placing the fuse on the outer wall of the vessel 10, temperature measurements would primarily depend on the boiling temperature of the water in contact with the vessel wall, which is ineffective in detecting the rise of the molten core pool and the formation of the focal effect layer.

[0171] Preferably, the safety device includes a series of fuses 900 positioned along at least one parent surface of the inner wall 11 of the container 10. Preferably, the fuses are positioned along at least two parent surfaces. Thus, if a molten core stream occurs along a parent surface, the fuses positioned in the upper portion can be reached and do not indicate a gradual rise of the molten core pool at the bottom of the container.

[0172] Preferably, for each mother surface, a series of fuses 900 are disposed on the mother surface of the hemispherical portion forming the bottom 12 of the container 10, and the latter and another portion of the fuses are disposed on the side wall of the container 10.

[0173] These fuses allow for the determination of the time point at which the core melt pool 71 begins to form and the time point at which water from the second stage should be injected into the first stage loop 100.

[0174] For example, based on the evolution curve of the core melt pool height, which is estimated based on the signal sent by the fuse, the optimal time to trigger the injection of water from the secondary loop into the primary loop 100 can be determined through simulation.

[0175] The temperature evolution curve also allows for the detection of rising core melt levels in vessel 10. The curve also allows for the detection of the onset of the latter being punctured by the liquid metal layer 72.

[0176] For example, fuses 900 with the same facet can be provided, or fuses 900 with two different facets having different melting temperatures can be provided. For this purpose, different materials can be provided for the fuse sheath and / or core. Preferably, a device that can be completely disassembled and replaced, for example, during a ten-year inspection, is provided, so that a set of safety-critical devices can be available with a service life not exceeding 10 years of reactor operation.

[0177] According to a particularly advantageous example, it is desirable to position the core melt pool height detector so that the fuse reveals the rate of rise of the core melt pool. This allows for more accurate monitoring of the time points at which focusing effects may occur and the time points at which fluid from the steam generator 210 should be injected into the first-stage loop 100. Therefore, this positioning of the fuse 900 is performed such that the volume V of the core slice located between two consecutive or adjacent fuses 900 is... slice It is constant.

[0178] More specifically, a container slice is defined by two vertically distributed adjacent fuses. This container slice is defined on one hand by the inner wall 11 of the container 10, and on the other hand by two vertical planes, each passing through one of these adjacent fuses 900. At least some of the volumes of the container slice are preferably all identical. Preferably, the vertically distributed adjacent fuses are arranged according to the same curve that preferably forms the parent surface of the container 10.

[0179] Figure 5 A portion of container 10 and a fuse 900 disposed according to the mother face G are shown in a very schematic manner. The fuse 900 defines a portion having substantially equal volume V. slice Slices.

[0180] Therefore, if the formation rate of the core melt pool (and thus the formation rate of the core melt volume) is constant, then the fuses on the same parent plane are reached by the core melt pool with the same time interval between two consecutive fuses on that parent plane.

[0181] refer to Figure 5 An example of calculating the fuse location will now be described.

[0182] In this example, the fusible detector is placed along two to four mother faces that are evenly distributed along the axis surrounding the hemispherical bottom of the container 10. Therefore, for mother faces that are equal to 2, 3, and 4, these mother faces will be separated by angles of 180°, 120°, and 90°, respectively.

[0183] It is assumed that the bottom of container 10 is formed by a spherical portion with a diameter of 4 meters.

[0184] The desired placement corresponds to 3, 6, 9, 12, and 15m. 3 A core molten material overflow detector. This allows for the determination of the rising curve of the core molten material pool. By clearly understanding the changes in the rising rate of the cemented carbide molten pool, the operator (or automatic safety device) can determine the most suitable time to trigger the water injection of the steam generator 210 into the first-stage loop 100.

[0185] The total filling volume of the hemispherical bottom is 16.75m³. 3 .

[0186] exist Figure 5 The following parameters are referenced in the text:

[0187] ·R = radius of the hemispherical part of the container,

[0188] h = the height of the fuse relative to the bottom of the container.

[0189] ·L arc = Length along the container wall, between the bottom of the container and the location of the fuse

[0190] ·r = the distance between the cylindrical axes of the cylindrical portion of container 10 (i.e., the axis passing through the center of the sphere and perpendicular to plane P).

[0191] Plane P corresponds to the junction between the hemispherical portion of the container and the sidewalls of the container that extend from the column.

[0192] The volume V of the hemispherical part of the cube cap The following formula can be used for calculation:

[0193]

[0194] The location of the fuse can be calculated based on the value L according to the following formula.arc Sure:

[0195]

[0196] If it is desired to install five fuses on each mother surface, the fuses can be arranged in the following manner to establish a correspondence between the rise rate of the molten core pool and the melting of these fuses:

[0197] • Fuse No. 1: Height h = 0.738m; L arc =1.776m;

[0198] • Fuse No. 2: Height h = 1.079m; L arc =2.185m;

[0199] • Fuse No. 3: Height h = 1.362m; L arc =2.492m;

[0200] • Fuse No. 4: Height h = 1.617m; L arc =2.756m;

[0201] • Fuse No. 5: Height h = 1.860m; L arc =3.002m.

[0202] Alternatively, the connection between the second-stage loop 200 and the first-stage loop 100 can be controlled based on a timeline starting from the point in time when the fuse 900 detects the formation of the core melt pool 71.

[0203] Through simulation, the duration D1 between the beginning of core melt pool formation 71 and the start of liquid metal layer formation 72 at the point of puncture of container 10 can be determined very early. Naturally, this duration varies depending on the reactor. For some reactors, this duration D1 is approximately one hour. The operator should activate the safety device at time t1, where t1 = t0 + (D1 - k1), where k1 is a safety factor to ensure that the second-stage water is poured onto the liquid metal layer 72 sufficiently early to avoid significant weakening of container 10 and, preferably, to prevent the initiation of puncture. For example, k1 is between -5 minutes and 15 minutes.

[0204] Through simulation, the thickness e from the point where the core melt pool 71 begins to form to the liquid metal layer 72 can also be determined. 72 The duration D2 between the point t2 when it is large enough to prevent it from piercing container 10.

[0205] The cross-sectional dimensions of the channel between the second-stage loop 200 and the first-stage loop 100 are designed to allow a flow rate Qmin sufficient to cool the liquid metal layer 72 to be poured between time points t1 and t2.

[0206] As described above, it is highly advantageous that the fluid 800 from the second-stage circuit and poured onto the overmolten metal layer 72 flows down by exiting along the wall 11 of the container 10. However, alternatively or in combination with this embodiment, the fluid 800 can be made to reach the interior of the container 10 by permeating it into the container 10 via the outlet orifice 14.

[0207] Figures 5 to 7 Different embodiments are shown that allow fluid from the second-stage circuit 200 to be injected into the container 10 via the first-stage circuit 100.

[0208] The solution in all these implementations lies in setting the second-stage circuit 200 in communication with the first-stage circuit 100. This communication is achieved because the safety system is configured to intentionally break the containment barrier that isolates the two circuits 100 and 200. It should be remembered that during normal operation, that is, in the absence of accidents and during the power generation phase at the station, the first-stage circuit 100 and the second-stage circuit 200 are fluidly insulated from each other.

[0209] Before describing some examples of the solutions in detail, the following paragraphs refer to Figure 6 A conventional steam generator 210 is described. The steam generator 210 forms a housing that encloses the fluid of the second-stage circuit 200 and simultaneously encloses at least one pipe 214 through which the fluid of the first-stage circuit 100 circulates. Therefore, the fluid of the second-stage circuit 200 is in contact with the outer wall of the pipe 214.

[0210] Figure 6 The lower portion of a steam generator 210 is shown. The steam generator 210 includes a sleeve 260 having a generally cylindrical shape. The sleeve 260 defines an upper portion 211 having openings 211b, 211b' communicating with a second-stage circuit 200 and a lower portion 212 having openings 212b, 212b' communicating with a first-stage circuit 100. The upper portion 211 and the lower portion 212 are separated by a plate 213 having pipes. The lower surface of the plate 213 with pipes, together with the lower portion 212, defines a volume forming a water tank. Preferably, the water tank has a hemispherical shape. It is divided into two portions 212a, 212a' by a partition 2121.

[0211] Section 212a has an opening 212b that is hydraulically connected to the inlet 13 of container 10. Section 212a' has an opening 212b' that is hydraulically connected to the outlet 14 of container 10. Therefore, sections 212a and 212b, as well as pipe 214, are part of the first-stage circuit 100.

[0212] At outlet 14 of container 10, pressurized water at high temperature from the core 30 of nuclear reactor 1 permeates into section 212a' of the water tank and then circulates in pipes 214 of the piping network of steam generator 210. In practice, plate 213 with pipes has multiple pipes 214, one end of which leads to section 212a' and the other end to section 212a. Typically, these pipes 214 are inverted "U" shapes. These "U"-shaped pipes are immersed in water in the second-stage loop 200 present in the upper section 211 of steam generator 210.

[0213] Therefore, the pressurized hot water circulation, starting from section 212a', proceeds from bottom to top until the top of the "U" bend, and then from top to bottom to reach section 212a of the water tank. Throughout this route, the water circulating in pipe 214 transfers heat to the fluid in the second-stage loop 200 present in the upper section 211 of the steam generator 210. Once it reaches section 212a of the water tank, the water can escape via outlet 212b and return to inlet 13 of vessel 10 to be reheated by the core 30.

[0214] It should be noted that each part 212a, 212a' of the water tank is provided with a hatch opening 2122 that is closed by a plug 2123. The hatch opening 2122 is large enough to allow a person or robot to enter the interior of the water tank.

[0215] Now refer to Figure 7 and 8 The first embodiment of the present invention is described.

[0216] In these embodiments, a safety system capable of intentionally disrupting the hydraulic containment between the first-stage circuit 100 and the second-stage circuit 200 includes at least one passage formed by a conduit 120 connecting the closed second-stage water portion 211 of the steam generator 210 to the first-stage circuit 100. This conduit 120 may include a safety injection line (commonly referred to as the "IS line") or an RCV (an acronym for Volume and Chemical Control Circuit) tap, or a tap on the cooling circuit at the RRA stop. It should be noted that the cross-section of the IS line or RCV tap is typically smaller than the sum of the cross-sections of the inlets 13 in the vessel 10. However, as described above, the invention operates perfectly by injecting water from the second stage into the first-stage circuit 100 via a small cross-section.

[0217] More specifically, the upper portion 211 of the steam generator 210 has an orifice connecting to one end 121 of the conduit 120. The other end 122 of the conduit 120 is connected, for example, via a tap on the first-stage circuit 100 or an existing tap such as an IS line, RCV line, or RRA line. Naturally, the orifice 121 is located in the bottom of the upper portion 211 to facilitate the gravity flow of all the second-stage water contained in the steam generator 210.

[0218] The pipe 120 is equipped with a device that allows the following:

[0219] • Under normal operating conditions of reactor 1, that is, in the absence of any accidents, prevent any circulation of the second-stage fluid within it.

[0220] • In the event of abnormal operation of reactor 1, the container 10 may be punctured or at risk of being punctured, causing the second-stage fluid to circulate in the pipe from steam generator 210 to the first-stage loop 100.

[0221] For this purpose, the device is configured to selectively allow fluid to pass through. According to one example, the device includes a valve 219. In a hydraulically insulated configuration, valve 219 prevents any circulation of fluid within the conduit 120.

[0222] When the safety system is activated, valve 219 suppresses this hydraulic insulation.

[0223] According to the first embodiment, valve 219 is intended to be remotely controlled between a closed configuration and an open configuration. The alternation between the closed and open configurations is performed by actuating valve 219. In this first embodiment, valve 219 can be remotely triggered to open when the safety device detects a risk of puncture or impending puncture of container 10.

[0224] According to the second embodiment, the switch from the closed configuration to the open configuration is manually performed by the operator. In this second embodiment, when the safety device detects a risk of puncture or imminent puncture of the container 10, it can trigger a signal that the notification valve 219 should be opened.

[0225] In this embodiment, with a dedicated pipe 120 between the sleeve 15 of the steam generator 210 on the cold first-stage side and the piping system of the first-stage circuit 100, the presence of the "U" branch between the steam generator 210 and the container 10 does not actually cause a problem of filling this "U" branch. In fact, it can be ensured that water flows from the second-stage water collection in the steam generator 210 to the tap 122 of the pipe 120 by gravity, thus ensuring that water enters the main cold branch 13 of the container 10 and then into the downcomer 16, without any booster pump. This is as follows... Figure 8 As shown.

[0226] The low elevation of the steam generator 210 is above the elevation of the hot and cold branches of the container 10 and their respective orifices 13, 14, thereby enabling gravity flow from the casing of the steam generator 210 to the top of the downcomer 16. When valve 219 is opened (or when pipe 214 ruptures), the second-stage pressure is significantly higher than the first-stage pressure, and adiabatic expansion occurs at the level of the containment barrier rupture, causing a portion of the second-stage liquid water to decompose into steam, while the other portion remains liquid and then flows by gravity after the initial propulsion phase caused by the steam power generated by the expansion.

[0227] In these embodiments, the entry of the second-stage fluid into the first-stage loop 100 is achieved via pipe 120, with cross-section S. breach Equal to and preferably less than the minimum cross-section of the pipe 120. Of course, if multiple pipes 120 are configured to pour secondary water into container 10, then S breach It is equal to the sum of the equivalent crack portions entering container 10.

[0228] In one embodiment, a conduit 120 may be provided, for example, via a tap 120', which is also in fluid communication with the boron tank.

[0229] It should also be noted that the same end 122 of pipe 120 can be connected to several steam generators 210. For this purpose, a portion 123 of pipe is supplied by several branches 120, 120', each of which is connected to a different steam generator. This has the advantage of drawing water captured in several steam generators 210 into the first-stage loop 100 while performing a tap 122 only once on the first-stage loop 100, thereby reducing the safety implications of the solution.

[0230] exist Figure 7 In this example, the pressurizer 110 is located between the steam generator 210 and the outlet 14 of the container 10, as is the case normally. In this example, the tap 122 is located between the steam generator 210 and the container 10.

[0231] exist Figure 8 In this example, the pressurizer 110 is located between the steam generator 210 and the inlet 13 of the container 10. In this example, the outlet 122 is located between the pressurizer and the container 10.

[0232] Figure 9 One embodiment is shown in a very simplified manner, wherein the pouring of second-stage water into the first-stage loop 100 is performed by adding a pipe to the steam generator 210, as shown in... Figure 7 and 8The same applies to pipe 120 in the illustrated embodiment. This pipe extends between inlet 121 and outlet 122.

[0233] However, the pipe is tapped on the device 170 (which is typically abbreviated as RRA) used to cool the reactor when it stops, in order to avoid drilling into the pipes of the first-stage loop 100.

[0234] The RRA device 170 of the station includes:

[0235] The first loop includes a heat exchanger 130, whose inlet 131 and outlet 132 are connected to a first-stage loop 100, preferably to cold inlets, each of which is connected to a different steam generator 210. In the illustrated example, for clarity, inlet 131 and outlet 132 are respectively connected to the hot and cold branches of the same first-stage loop 100. Preferably, the first loop of the RRA device 170 is connected to the cold inlets of several, typically two, steam generators 210. Thus, branch 131 is connected to the cold inlet of the first steam generator 210, and branch 132 is connected to the cold inlet of the second steam generator 210.

[0236] • The second circuit 140 forms a hydraulic loop, which is connected to heat exchanger 130 on one side and to an additional heat exchanger 150 on the other.

[0237] • A third external loop 160, which is connected to an additional heat exchanger 150 and includes a cold source.

[0238] As in the previous example, the pipeline is equipped with selective devices for hydraulic insulation or hydraulic connection, such as valve 219. Valve 219 can be controlled manually or remotely.

[0239] Therefore, the advantage of this embodiment is that the safety system of the present invention does not require additional tapping of the first-stage loop 100. Thus, this embodiment allows for the avoidance of the need to introduce additional constraints in terms of safety, while providing an effective solution against container puncture via focusing effects.

[0240] Based on the foregoing description, it is clear that the present invention provides a reliable and robust solution to significantly improve the safety of PWR-type nuclear reactors, especially in the event of coolant loss from the first-stage loop.

[0241] Advantageously, and as described above, the connection point between the first and second stage loops is chosen such that the propagation of water from the second stage to the first stage loop occurs almost entirely on the cold branch side of the first stage loop. Thus, the water from the steam generator, flowing under gravity in the downcomer—where the generating vessel is punctured by a focusing effect—will follow the first stage cold branch before reaching the first stage.

[0242] This invention is not limited to the embodiments previously described and extends to all embodiments covered by the claims.

[0243] Naturally, the present invention is not limited to having Figure 4 The reactor shown has the structure shown, and various modifications of the reactor are possible without departing from the scope defined by the claims.

[0244] Specifically, only one inlet 13 and one outlet 14 are shown in the figure. Preferably, reactor 1 includes multiple inlets and multiple outlets. Preferably, it also includes multiple steam generators 210.

Claims

1. A safety method for a pressurized water nuclear reactor (1), wherein, in the event that the core (30) of the pressurized water nuclear reactor (1) is at least partially melted and forms a core melt (70) molten pool (71), during the operation phase, the pressurized water nuclear reactor (1) comprises at least: • A first-stage loop (100) in which a water-based first-stage fluid is intended to circulate, the first-stage loop (100) being configured such that the first-stage fluid permeates into the vessel (10) of the pressurized water core reactor (1) and through the core (30) contained within the vessel (10) in order to extract heat generated by the core (30). A second-stage loop (200) is intended to circulate a water-based second-stage fluid, the second-stage loop (200) being hydraulically isolated from the first-stage loop (100) and including at least one steam generator (210). The second-stage loop (200) is configured to absorb heat from the first-stage loop (100) and convert it at least partially into steam in the steam generator (210). The security method is characterized by comprising at least the following steps: • In response to the detection of an event that at least partially melts the core (30) characterizing the pressurized water nuclear reactor (1) and forms a core melt (70) pool (71) in the bottom (12) of the container (10) and a liquid metal layer (72) is formed on the surface of the core melt (70) pool (71), the second stage loop (200) is configured to be in fluid communication with the first stage loop (100) such that the second stage fluid follows the first stage loop (100) to flow through the liquid metal layer (72) of the core melt (70) pool (71) inside the container (10).

2. The security method according to claim 1, wherein, The detection is performed using at least one fuse (900) disposed on the wall (11) of the container (10), the at least one fuse (900) being configured to melt when the liquid metal layer (72) reaches it.

3. The security method according to claim 2, wherein, The at least one fuse (900) has a melting temperature of the fuse (900) that is higher than or equal to a temperature threshold Tf, wherein Tf ≥ 400 °C.

4. The security method according to claim 3, wherein, Tf≥500℃.

5. The security method according to claim 3, wherein, Tf=600℃.

6. The security method according to claim 2 or 3, wherein, The pressurized water core reactor includes a plurality of fuses (900) distributed according to at least one mother face (G) of the wall of the vessel (10), such that two adjacent fuses (900) define a volume V. slice A slice of container, the volume of which is V slice They are the same.

7. The security method according to claim 1, wherein, During a period of at least thirty minutes, the second-stage fluid flows through the liquid metal layer (72) inside the container (10).

8. The security method according to claim 7, wherein, During a period of at least one hour, the second-stage fluid flows through the liquid metal layer (72) inside the container (10).

9. The security method according to claim 7, wherein, During a period of at least two hours, the second-stage fluid flows through the liquid metal layer (72) inside the container (10).

10. The security method according to claim 1, wherein, The second-stage fluid flows inside the container (10) at a flow rate of less than 10 kg / s.

11. The security method according to claim 10, wherein, The second-stage fluid flows inside the container (10) at a flow rate of less than 7 kg / s.

12. A pressurized water core reactor (1), said pressurized water core reactor (1) comprising at least: • A container (10) for accommodating the core (30) of the pressurized water nuclear reactor (1), the container (10) comprising at least one inlet (13) and at least one outlet (14). • A first-stage loop (100), at least one first end of which is connected to the inlet (13) of the container (10), and at least one end of which is connected to the outlet (14) of the container (10), such that the first-stage fluid circulating in the first-stage loop (100) permeates through the inlet (13) into the container (10) of the pressurized water nuclear reactor (1) and exits through the outlet (14), while penetrating through the reactor core (30) to extract the heat generated by the reactor core (30), S inlet It is the cross section of the first-stage loop (100) up to the at least one inlet (13) of the container (10). A second-stage loop (200) is fluidly isolated from the first-stage loop (100), a water-based second-stage fluid is intended to circulate in the second-stage loop, and the second-stage loop includes at least one steam generator (210), the second-stage loop (200) being configured to absorb heat from the first-stage loop (100) and convert it at least partially into steam in the steam generator (210). The pressurized water nuclear reactor (1) is characterized in that it includes a safety system comprising a safety device configured to form at least one channel that inhibits fluid isolation between the second-stage loop (200) and the first-stage loop (100), and sets the second-stage fluid present in the at least one steam generator (210) in fluid communication with the first-stage loop (100), such that the second-stage fluid contained in the at least one steam generator (210) can flow in the container (10) while passing through the first-stage loop (100) beforehand, the channel having a minimum cross-section S. breach , so that: S breach <0.05 S inlet 。 13. The pressurized water core reactor according to claim 12, wherein, The safety device is configured to form only one channel.

14. The pressurized water core reactor according to claim 12, wherein, S breach <0.01 S inlet 。 15. The pressurized water core reactor according to claim 12, wherein, S breach <0.001 S inlet 。 16. The pressurized water core reactor according to claim 12, wherein, Section S breach Between 0.2 cm 2 Up to 20 cm 2 between.

17. The pressurized water core reactor according to claim 16, wherein, Section S breach Between 0.8 cm 2 Up to 20 cm 2 between.

18. The pressurized water core reactor according to claim 16, wherein, Section S breach Between 2 cm 2 Up to 7 cm 2 between.

19. The pressurized water nuclear reactor according to claim 12 or 16, comprising at least one fuse (900) disposed on the wall (11) of the container (10), the fuse being configured such that when a liquid metal layer (72) reaches the fuse, the liquid metal layer melts the fuse, the melting temperature of the fuse (900) being higher than or equal to a temperature threshold Tf, wherein Tf ≥ 400°C.

20. The pressurized water core reactor according to claim 19, wherein, Tf≥500℃.

21. The pressurized water core reactor according to claim 19, wherein, Tf=600℃.

22. The pressurized water nuclear reactor according to claim 19, comprising a plurality of fuses (900) distributed along at least one mother face (G) of the wall of the vessel (10), the fuses (900) being distributed such that two adjacent fuses (900) along the mother face define a volume V. slice A slice of container, the volume of which is V slice They are the same.

23. The pressurized water core reactor according to claim 12, wherein, The pressurized water nuclear reactor (1) includes an inner sleeve (15) located inside the vessel (10), surrounding the core (30) and defining an annular volume called a downcomer (16) together with the inner wall (11) of the vessel (10), the downcomer (16) being configured such that during normal operation of the pressurized water nuclear reactor (1): The inlet (13) leads to the outside of the inner sleeve (15) and to the downcomer (16), so that the first-stage fluid from the inlet (13) is guided to the bottom (12) of the container (10). The outlet (14) leads to the interior of the inner casing (15), allowing the first-stage fluid present in the core (30) to exit from the pressurized water nuclear reactor (1) through the outlet (14). The pressurized water reactor is configured such that when the safety device forms at least one channel that inhibits fluid isolation between the second stage loop (200) and the first stage loop (100), the second stage fluid contained in the steam generator (210) then flows into the bottom (12) of the container (10), while passing through the inlet (13) of the container (10) before passing through the downcomer (16).

24. The pressurized water core reactor according to claim 12, wherein, The steam generator (210) includes a casing (260) comprising a first portion enclosing a portion of the first-stage fluid and a second portion enclosing a portion of the second-stage fluid, the first portion and the second portion being fluidly isolated from each other. The safety system includes at least one pipe (120) forming the passage, located outside the steam generator (210), and having at least: • A first end, the first end leading to the second portion that encloses the second-stage fluid, • The second end, which leads to a branch of the first-stage circuit (100) located between the steam generator (210) and the container (10), The safety device includes at least one component (219), which is mounted on the pipe (120) and selectively has: • In the closed configuration, the at least one component prevents fluid from passing through the pipe (120). • Open configuration, in which at least one component allows fluid to pass through the pipe (120), thereby allowing the second stage fluid of the steam generator (210) to flow in the pipe (120) to join the first stage loop (100) and then to the container (10).

25. The pressurized water core reactor according to claim 24, wherein, The second end of the pipe (120) forms a tap on a branch of the first-stage circuit (100).

26. The pressurized water core reactor according to claim 24 or 25, wherein, The branch of the first-stage circuit (100) extends between the steam generator (210) and the inlet (13) of the container (10).

27. The pressurized water nuclear reactor of claim 24, comprising means selected from a safety injection line and a volume and chemical control loop, the means being configured to lead to the first stage loop (100) at the second end of the conduit (120).

28. The pressurized water core reactor according to claim 12, wherein, The steam generator (210) includes a first portion enclosing the first-stage fluid and a second portion enclosing the second-stage fluid, the first portion and the second portion being fluidly isolated from each other. The pressurized water core reactor also includes an RRA device (170) for cooling the pressurized water core reactor when stopped. The RRA device (170) includes at least one first loop, the at least one first loop including a heat exchanger (130) and branches fluidly connecting the heat exchanger (130) to one or more portions of the first stage loop (100). The safety system includes at least one pipe (120) located outside the steam generator (210), forming the passage and having at least: • A first end, the first end leading to the second portion that encloses the second-stage fluid, • The second end, the second end leading to the branch of the first circuit of the RRA device (170), The safety device includes at least one component (219), which is mounted on the pipe (120) and selectively has: • In the closed configuration, the at least one component prevents fluid from passing through the pipe (120). • Open configuration, in which at least one component allows fluid to pass through the pipe (120), thereby allowing the second stage fluid of the steam generator (210) to flow in the pipe (120) to join the branch of the first circuit, then join the first stage circuit (100), and then join the container (10).

29. The pressurized water core reactor according to claim 24, wherein, The component is selected from the following valves: manually operated valves and remotely controlled valves.

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