Patent Application
20240371535
The present invention relates to the field of solid-fuel nuclear reactors cooled with one or more molten salt(s) or liquid metal coolant(s), in particular with liquid sodium, referred to as fast neutron reactors (FNR), which are part of the Generation IV family of nuclear reactors.
More particularly, the invention relates to a simplification of the architecture of these nuclear reactors while ensuring reliable removal of both the nominal heat and the decay heat in normal and accident shutdown situations.
The invention applies to small modular reactors (SMR), and more specifically to micro modular reactors (MMR), typically with an operating power of less than 20 MWth.
It will be recalled here that the decay heat of a nuclear reactor is the heat produced by the core after the nuclear chain reaction has been shut down and which consists mainly of the energy of the decay of the fission products.
The decay heat corresponds to a fraction of the nominal heat and decreases over time. In any event, it must be taken into account during the normal reactor shutdown phases, during handling phases, and also during accident situations, in order to size heat removal systems.
Although the invention is described with reference to a nuclear reactor cooled with one or more molten salt(s), it also applies to reactors cooled with liquid sodium or any other liquid metal, such as lead, used as a coolant in a nuclear reactor primary circuit.
Likewise, although the invention is described with reference to a fast neutron reactor, it also applies to thermal and/or epithermal spectrum reactors.
The fast neutron reactor system was developed to allow improved management of the nuclear fuel, in particular through sustainable management of the plutonium stock and the ability to recover the stock of isotope 238 of uranium, which isotope cannot be recovered in thermal neutron reactors.
The uranium-plutonium breeding cycle is only possible with high-energy neutrons. The resulting fast neutron spectrum is the key to the sustainable management of the plutonium stock and to the possibility of recovering stocks of uranium-238 not used by the more conventional system of light-water cooled thermal neutron reactors.
Solid-fuel fast neutron reactors are based on a physical separation between the solid fuel and the coolant, by means of a barrel forming a physical barrier (cladding). The fuel itself is made up of materials that are solid at the target operating temperatures (in particular in the form of oxides, silicon carbides (SIC) or nitrides of fissile materials, or directly in the form of a metal alloy).
Operationally, the physical barrier separating the fuel and the coolant makes it possible to:
In this context, the main solid-fuel fast neutron reactor projects were developed around concepts implementing liquid metals or gases to perform the coolant function, and with a solid fuel physically separated from the coolant by a barrier.
Mention can be made of FNRs cooled with liquid sodium or lead, or high-temperature gas-cooled reactors.
The system of FNRs cooled with sodium (Na) is the most mature technology, in particular integral reactors, that is, with a primary sodium circuit positioned inside the reactor vessel.
The feedback derived from the operation of a number of FNR—Na reactors both in France and abroad [1] has highlighted the following advantages:
However, integral FNR—Na reactors have the following drawbacks:
Lead-cooled reactors (FNR—Pb) make it possible to eliminate the risks linked to the exothermic interaction between sodium and water and to produce high-temperature heat by means of a solid fuel, while increasing the coolant boiling margin, as the boiling temperature of lead is approximately 1,750° C.
Nevertheless, lead-cooled reactors have problems of erosion/corrosion of the components, which are worse at high temperatures and with very high coolant densities, which requires them to operate at temperatures limited to 500-550° C. [2].
Two other known reactor technologies are intrinsically capable of overcoming or at the very least attenuating the risks linked to the exothermic reaction between sodium and water and the boiling of the coolant in the event of an accident, and of producing high-temperature heat (>550° C.) by means of a solid fuel.
These are gas-cooled fast reactors (GFR), very high temperature gas reactors (VHTGR) and solid-fuel reactors cooled with a molten salt, referred to as MSFR.
VHTGRs are characterized by solid fuel the very high thermal inertia, low power density and very high melting temperature of which improve safety compared to sodium reactors. Reference can be made to FR2956773B1 or publication [3]. A pressurized gas is used to cool the fuel, which makes it possible to avoid the risks linked to a liquid coolant boiling in the event of an accident. In addition, using a gas makes it possible to go up to very high temperatures provided that the structural materials are compatible, typically up to 1,000° C. at the core outlet, and to achieve significant efficiency levels. Among the gases that have been considered as coolants, helium has often been selected as the main candidate due to its thermal properties (greater conductivity and specific heat than other gases) and its chemical inertia.
However, having a gas (helium) as the coolant involves the following different drawbacks:
In order to overcome the problems resulting from the use of a liquid metal or a gas as a coolant, a liquid salt can be used as a coolant.
A very compact nuclear reactor architecture using a liquid salt as a coolant, with low power of approximately 120 MWth, is described in publication [6], and mainly illustrated in
This architecture of a liquid salt-cooled solid-fuel reactor solves all of the problems linked to the use of helium set out above. Although the operating temperature mentioned in publication [6] is approximately 500-550° C., the boiling/dissociation temperature of the salt is sufficiently high (>1,000° C.) to make it possible to operate at higher temperatures without any risk of boiling.
However, the reactor architecture according to publication [6] has the following drawbacks:
There is therefore a need to improve solid-fuel nuclear reactors cooled using liquid metal or liquid salt(s), in particular in order to overcome the aforementioned drawbacks.
In general, there is a need to improve the safety of solid-fuel nuclear reactors cooled using liquid metal or liquid salt(s), by complying with all of the points of specifications that can be defined as follows:
The aim of the invention is to at least partly meet this need/these needs.
In order to achieve this, the invention relates, in one of its aspects, to a nuclear reactor cooled using liquid metal or one or more molten salt(s), comprising:
Advantageously, the stock of PCM(s) makes it possible, due to the energy required for the phase change (latent heat) to absorb all of the decay heat from the core for a predetermined duration of three days. There is therefore a sizing margin linked to the sensible heat of the molten salt in the liquid state that would make it possible, well before it boiled, to absorb additional decay heat and therefore leave a grace period before intervention of more than three days.
“Sensible heat” is given to mean the heat that the material, in a given state, can absorb without changing phase.
Here and in the context of the invention, “shutdown situations” is given to mean a normal reactor shutdown and not a reactor shutdown in the event of an accident (accident situations).
“Inert liquid salt” is given to mean a liquid salt coolant that does not comprise any fissionable elements or breeder elements and does not react chemically with water or air.
Advantageously, the height of PCM(s) between the shell and the secondary vessel is greater than the height of inert liquid salt(s) between the primary vessel and the shell. This ensures that the leaking of liquid salt coolant through the primary vessel is limited or prevented in the event of an accident. Likewise, preferably, the height of the liquid metal between the primary vessel and the shell is greater than the height of inert liquid salt(s) between the primary vessel and the shell.
According to one advantageous structural variant, the primary and second vessels and the shell are right cylinders arranged concentrically.
Preferably, the inert liquid salt is selected from chlorine-based salts, optionally enriched with chlorine-37 to reduce the formation of radioactive Cl36, more preferably NaCl, KCl, MgCl2, CaCl2, ZnCl2 or a mixture thereof, in particular a molten salt mixture NaCl—MgCl2, NaCl—MgCl2—KCl or NaCl—MgCl2—KCl—ZnCl2.
According to one advantageous variant embodiment, the closed circuit comprises a serpentine coil, the periphery of which is preferably provided with heat dissipating fins, the serpentine coil being arranged between the primary vessel and the shell, in a spiral around the shell. Preferably, the serpentine coil can be fixed, in particular by welding to the shell.
The liquid metal of the bath between the primary vessel and the shell is advantageously an aluminium alloy, preferably A0850.0, as disclosed in [7].
Preferably, the PCM(s) between the shell and the secondary vessel is/are in the form of a powder.
More preferably, the PCM(s) between the shell and the secondary vessel is/are selected from pure aluminium or an aluminium alloy, preferably A0850.0.
More preferably, the primary vessel is made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC).
More preferably, the secondary vessel and the shell are made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the planned operating conditions.
The nuclear reactor according to the invention can be a thermal spectrum, epithermal spectrum or fast neutron reactor.
According to a first alternative, the primary vessel is thus devoid of moderating material so that the reactor operates using fast neutrons.
According to a second alternative, the reactor core houses at least one moderating material so that the reactor operates using thermal neutrons or epithermal neutrons. The coolant can advantageously act as the moderator (chloride salt containing lithium, or fluoride salt).
In the context of the invention, “moderating material” is given to mean any material that makes it possible to slow the neutrons. In the usual sense, the kinetic energy of a fast neutron is greater than 1 eV, while the kinetic energy of a thermal neutron is less than 1 eV, typically of the order of 0.025 eV. Reference can be made to publication [8], and particularly to
Reference can be made to FR3025650B1, which describes the insertion of moderating materials into a fuel assembly of a fast neutron reactor.
According to one advantageous embodiment, the solid nuclear fuels are nuclear fuel assemblies and/or fuel particles, referred to as TRISO, and/or fuel pellets individually housed in separate cells of a plate.
The solid nuclear fuels can be based on depleted, low enriched, preferably with enrichment of <5%, or reprocessed (URT) uranium dioxide (UO2), and/or on plutonium dioxide (PuO2), or based on enriched uranium U235 (HALEU, or high-assay low-enriched uranium), preferably with enrichment of between 5% and 20%.
Preferably, the nuclear reactor comprises a reactivity control system made up either of control rods inside the primary vessel, or by rotating drums outside the primary vessel.
The nuclear reactor described above is particularly intended to have a power of less than 20 MWth.
The invention therefore essentially consists of producing a solid-fuel nuclear reactor using liquid metal or molten salt primary vessel coolant that simultaneously ensures:
Some of the many advantages of the invention are:
The preferred applications of the invention are small reactors in the Gen IV family, in particular reactors cooled using sodium, lead or liquid salt.
Further advantages and features of the invention will become more clearly apparent on reading the detailed description of exemplary embodiments of the invention given by way of non-limiting illustration, with reference to the following figures.
Throughout the present application, the terms “vertical”, “lower”, “upper”, “bottom”, “top”, “below” and “above” are to be understood with reference to a primary vessel filled with inert liquid salt of a fast neutron nuclear reactor according to the invention, in its vertical operating configuration.
Such a reactor 1 comprises a primary vessel 10 or reactor vessel, filled with inert liquid salt, referred to as primary salt S, and which houses the core 11 in which are immersed a plurality of fuel assemblies, not shown, which generate thermal energy through nuclear fission of the fuel.
The inert liquid salt S is therefore the coolant of the primary circuit: it stores and transports the heat from the core 11 and exchanges heat through the wall of the primary vessel 10. The primary salt is provided with physical and chemical properties that make it possible to ensure that it remains in a liquid state at atmospheric pressure in the normal and accident operating temperature range of the core 11. It is inert from a radioactive point of view, as it does not contain any breeder elements or fissionable elements, and it is also chemically inert vis-à-vis the PCMs and liquid metal bath described in detail hereinafter, and vis-à-vis all of the structures of the reactor.
The primary salt S is selected so that it has good thermal and physical properties to promote natural convection and heat exchanges with the core 11 and through the wall of the primary vessel 10.
The primary salt is thus advantageously selected from NaCl, KCl, MgCl2, CaCl2 or ZnCl2 enriched with chlorine-37 or a mixture thereof, in particular a molten salt mixture NaCl—MgCl2, NaCl—MgCl2-KCl or NaCl—MgCl2-KCl—ZnCl2. For example, NaCl—KCl—MgCl2, which has a melting point of less than 500° C., good thermal capacity (Cp) and a good thermal expansion coefficient, is advantageous for improving natural convection.
The solid nuclear fuels can be based on depleted, low enriched, preferably with enrichment of <5%, or reprocessed (URT) uranium dioxide, and/or on plutonium dioxide (PuO2), or based on enriched uranium U235 (HALEU, or high-assay low-enriched uranium), preferably between 5% and 20% enriched.
The fuel assemblies can comprise SiC fuel cladding, in order to withstand very high temperatures. The claddings of the fuel assemblies form the first containment barrier while the vessel 10 forms the second containment barrier for the radioactive materials contained in the core 11.
The primary vessel 10 supports the weight of the liquid salt of the primary circuit and the internals.
The core 11 is supported by a welded structure referred to as the diagrid 12 in which the feet of the fuel assemblies 11 are positioned.
The core 11 is surrounded by a separating barrel 13 provided with a peripheral neutron reflector intended to ensure that the neutron flux is retained in the core. This separating barrel 13 of the core 11 makes it possible to separate the cold and hot primary salt. The cold primary salt thus surrounds the core 11 inside the primary vessel, while the hot primary salt, heated by circulating upwards in the core 11, is in the upper central portion of the core.
Typically, the diagrid 12 is made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions.
As shown, the primary vessel 10 is a right cylinder with a central axis X. Typically, the primary vessel 10 is made from AISI 316L stainless steel, preferably with a very low boron content in order to guard against the risk of cracking at high temperature. Its external surface is preferably rendered highly emissive by a pre-oxidation treatment, which is carried out in order to facilitate the radiation of heat to the outside during the decay heat removal phase. The primary vessel 10 can also be made from or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions.
The reactor vessel 10 is divided into two distinct zones by a separating structure made up of at least one shell 14 arranged inside the reactor vessel 10. This separating device is also known as a redan.
As illustrated in
The redan 14, and more specifically its bottom shell 142, can be welded to the diagrid 12 as shown in
The redan 14 further comprises through-openings 143 made through the bottom shell 142.
As shown by the arrows in
Typically, the redan 14 is made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions.
Advantageously, as shown in
The central chimney form of the redan 14 improves the circulation by natural convection of the liquid salt S.
A removable plug 15 referred to as the core head plug is arranged directly above the core 11 and closes the primary vessel 10 in order to contain the liquid salt S, act as a barrier between the liquid salt S and the external environment, and also form, with the primary vessel, the second containment barrier for the materials contained in the core 11.
Like the primary vessel 10, the core head plug 15 is chemically inert vis-à-vis the PCMs and liquid metal bath described in detail hereinafter, and vis-à-vis all of the structures of the reactor.
The core head plug 15 is provided with feedthroughs for the components of the control rods 16 and for the elements for controlling and monitoring the core 11 and the stock of liquid salt S, not shown. The core head plug 15 is therefore a plug that can be removed in an inert atmosphere and which carries all of the handling systems and all of the instrumentation necessary for monitoring the core and comprising the control rods, the number of which depends on the type of core and the power thereof, as well as the thermocouples and other monitoring devices. A system for maintaining the temperature of the plug will be provided in order to limit the risks of deposits of salt aerosols.
Typically, the core head plug 15 is made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions. Also typically, the material used for the control rods is B4C.
In one variant, instead of the control rods, rotating drums, also made from B4C, can be arranged outside the primary vessel 10, so as to advantageously eliminate the feedthroughs of the plug 15 and therefore the risks of seizing associated with the salt aerosols.
The reactor vessel 10 comprises a plenum, usually referred to as cover gas plenum, filled with an inert gas, such as argon or helium, above the molten salt liquid fuel. This plenum makes it possible to absorb the thermal expansion of the liquid within the reactor vessel, when it undergoes a change in level, and to recover the gaseous fission products generated by the nuclear fission within the fuel.
A support, containment and thermal insulation assembly between the primary vessel and the external environment E is arranged around the primary vessel 10.
More specifically, as shown in
The reactor pit 20 is a block with a generally cylindrical external shape that supports the weight of all of the components inside it. The reactor pit 20 has the functions of providing biological protection and protection against external attack, and also of providing cooling of the external environment in order to maintain low temperatures. Typically, the reactor pit 20 is a block of concrete.
The layer of thermally insulating material 21 ensures the thermal insulation of the reactor pit 20. Typically, the layer 21 is made of a polyurethane or silicate-based foam.
The secondary vessel 22 ensures that the liquid salt S is retained in the event of a leak from the primary vessel 10 and protects the reactor pit 20. The secondary vessel 22 also contains a volume 23 of solid-liquid phase-change material(s) (PCM).
The volume 23 of PCM(s), preferably in the form of powder, is capable of melting while storing by latent heat at least part of the decay heat emitted by the core in accident situations, for a predetermined duration. The physical and chemical properties of the PCM(s) thus make it possible to ensure that it/they remain in the solid state at atmospheric pressure in the normal operating temperature range of the core and that they change phase when the core 11 enters an accident situation, for a predetermined duration.
As shown in
The PCM(s) is/are chemically inert vis-à-vis the liquid metal bath and all of the structures of the reactor.
Typically, the PCM is a pure aluminium alloy or an A0850.0 aluminium alloy. As a variant, the metal PCM can be replaced by a salt PCM, for example MgCl2.
The secondary vessel 22 bears against the reactor pit 20 and its top part is welded to the reactor closure 17.
Typically, the secondary vessel 22 can be made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions.
An annular removable closure 24 around the core head plug 15 closes the secondary vessel 22 in order to contain the volume 23 of PCM(s) and act as a barrier between this volume 23 and the external environment E.
Like the secondary vessel 22, the closure 24 is chemically inert vis-à-vis the PCMs and liquid metal bath described in detail hereinafter, and vis-à-vis all of the structures of the reactor.
The closure 24 is provided with feedthroughs for the elements for controlling and monitoring the stock of PCM(s), not shown, and meets the sealing requirements of the second containment barrier. In the handling phase, the operations requiring the removal of this closure must take place in an inert atmosphere.
Typically, the closure 24 can be made from AISI 316L stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions.
The nuclear reactor 1 further comprises a system 3 for removing the heat both during nominal operation and in situations in which the nuclear reactor is shut down.
This system 3 firstly comprises a cylindrical shell 30 arranged concentrically between the primary vessel 10 and the secondary vessel 22.
This shell 30 thus defines a volume with the primary vessel 10 filled with a liquid metal bath 31.
Typically, the shell 30 can be made from AISI 316 stainless steel, or a nickel-based alloy, or silicon carbide (SiC), depending on the operating conditions, and is designed to chemically and mechanically withstand the liquid metal bath 31 and the volume of PCM 23.
A closed circuit 32, referred to as the secondary circuit, is filled with a coolant and is capable of removing the heat removed by conduction through the primary vessel 10 and transferred by the liquid metal bath 31, to an energy conversion system and/or a heat network, not shown.
The liquid metal bath 31 thus improves the heat transfer by conduction from the primary vessel 10 to the secondary circuit 32.
Typically, the liquid metal bath is made from an A0850.0 aluminium alloy in liquid form, preferably with a different chemical composition from the A0850.0 used as a PCM.
The liquid metal is chemically inert vis-à-vis the liquid salt S, the PCMs and all of the structures of the reactor.
The closed circuit 32 is preferably made up of a serpentine coil arranged in a spiral around the primary vessel 10, preferably welded to the inner wall of the shell 30.
The serpentine coil 32 has a diameter that depends on the diameter of the primary vessel 10 and a height that is sufficient to have the surface area necessary for the heat removal sought.
In other words, the total number of turns, the spacing and the diameter of these turns that make up the serpentine coil 32 are dependent on the diameter of the primary vessel 10 and on the power of the nuclear reactor core 11. For example, the pitch of the turns of the serpentine coil 32 can be equal to 10 cm, which is a good compromise between manufacture and conductive heat absorption by the liquid metal bath.
Again for example, the outer diameter of the serpentine coil 32 is of the order of 5 to 10 cm with a turn pitch of the order of 10 to 15 cm, so as to minimize pressure drops, reduce the footprint of the pipes and maximize the surface area exposed to the primary vessel 10. The thickness of the serpentine coil 32 depends on the mechanical stresses applied by the internal liquid metal and by its weight.
The material of the serpentine coil 32 must have good emissivity properties. Typically, the material of the serpentine coil is selected from AISI 316L stainless steel, ferritic steels, and nickel-based alloys. This material depends on the internal fluid used in the closed circuit 32.
This internal coolant that circulates in the serpentine coil 32 is a chemically stable, low viscosity liquid metal that is a good conductor and carrier of heat, chemically compatible with all of the pipework of the circuit 3 and able to operate in natural or forced convection in a temperature interval of 150-600° C. Typically, the liquid metal of the circuit 3 can be selected from an Nak alloy, a Pb—Bi alloy, sodium or one of the ternary alloys of the liquid metals, etc.
In order to improve the heat exchange between the liquid metal bath 31 and the pipe forming the serpentine coil 32, the serpentine coil can be provided with heat dissipating fins 33, in particular with a straight shape, that extend radially from the pipe, as shown in
The operation of the nuclear reactor 1 will now be described with respect to the different normal, planned shutdown and accident situations.
During normal operation of the reactor, all of the heat generated by the fission reactions of the core 11 is removed by heat exchange when the salt S in the liquid state passes through the core, and solely by natural convection. The liquid salt S exchanges this heat through the wall of the primary vessel 10 when it falls back down between the redan 14 and the primary vessel 10. This heat is then transferred by the liquid metal bath 31 to be removed by the secondary circuit 32 to a heat network and/or an energy conversion system. During normal operation, the volume 23 of PCM remains in the solid state.
In the event of a planned shutdown caused by an operator, the same heat exchanges take place. The decay heat from the shut down core 11 is also removed by heat exchange when the inert salt S in the liquid state passes through the core 11, and solely by natural convection. Then, the salt S exchanges this heat through the wall of the primary vessel 10 when it falls back down between the redan 14 and the primary vessel 10. This heat is then transferred by the liquid metal bath 31 to be removed by the secondary circuit 32 to a heat network and/or an energy conversion system. During this planned shutdown, the volume 23 of PCM remains in the solid state.
During operation in an accident situation, in particular in the event of a station blackout (SBO), which corresponds to a complete loss of electrical power supply, as in the Fukushima accident, the decay heat from the shut down core 11 is removed by heat exchange when the salt S in the liquid state passes through the core, and solely by natural convection. Then, the salt S exchanges this heat through the wall of the primary vessel 10 when it falls back down between the redan 14 and the primary vessel 10. This heat is absorbed by the initially solid PCM, which undergoes a transition from the solid state to the liquid state. The stock of the volume 23 of PCM makes it possible, due to the energy required for the phase change (latent heat) to absorb all of the decay heat from the core for a predetermined duration, typically three days.
The inventors have conducted sizing studies to demonstrate the feasibility of a nuclear reactor 1 as described above, and to propose orders of magnitude relating to its characteristic elements.
The studies conducted cover the 1 to 20 MWth range in normal and accident operation, with circulation of the primary salt S solely by natural convection.
The cold temperature of the inert liquid salt S, as the primary fluid, is set at 600° C.
The hydraulic path of the salt S is determined by the analytical study shown in the tables below. The thermal power range is defined within which the nuclear reactor 1 is intended to operate solely by natural convection, that is, without pumps or heat exchangers within the primary vessel.
The relevant properties are summarized in Table 1 below.
It will be noted here that the properties of the salt in question are as set out in publication [11].
These input data, for a variable power of between 1 and 20 MWth, were used by the inventors for the preliminary calculations using thermal calculation software such as COPERNIC: [9], [10].
These sizing calculations were carried out according to two sequential steps, as follows.
Step 1/: a core configuration is selected with the lowest possible resistance with respect to thermal hydraulics. A first core sizing is necessary to calculate the total pressure drops of the primary hydraulic circuit. The result of this calculation forms the resistive load that the circulation must overcome to allow flow solely by natural convection of the coolant salt S.
Step 2/: a flow of the primary salt S solely by natural convection within the primary vessel 10 and a heat transfer that takes place solely by natural convection through the wall of the primary vessel 10.
In step 1/, a preliminary core design with nuclear fuel pin assemblies was determined for a power interval of 1 to 20 MWth. In this calculation, the hydraulic flow area was maximized by applying the physical quantities given in Table 2.
Taking into account the physical quantities in Table 2, the hydraulic flow area for the salt S is maximized by fissile height values of the fuel and a number of fuel pins in the core. Having a maximized hydraulic flow area improves the flow of the fluid in natural convection operation as the greater the area for a given flow rate, the more the core pressure drops are reduced.
The preliminary sizing calculations for the core 11 are summarized in Table 3, for thermal powers of 10 and 20 MWth respectively.
It will be noted that the fuel assemblies in question have hexagonal sheaths.
The results of Table 3 give the input data for the sizing calculation of the primary circuit. This determines the temperature difference at the inner wall of the primary vessel 10 necessary to remove the heat carried by the primary salt S.
The operating point for each thermal power value of the reactor is found by applying the pressure drops equal to the driving force supplied by natural convection, which is equivalent to:
ΔPdriving force=ΔPhydraulic resistance
where
the difference in density between the salt in the cold collector and the salt in the hot collector (core outlet);
gravitational constant;
density of the primary sans, variable depending on the local temperature of the fluid;
At this stage the acceleration pressure drops and the local pressure drops are not taken into account (as they are not known at this stage of the sizing).
Solving the equation above makes it possible to calculate the temperature difference ΔT of the salt S, between its hot flow zone and the cold flow zone delimited by the redan 14, which is capable of initiating circulation solely by natural convection in the primary circuit, and the minimum temperature difference to remove all of the heat produced through the wall of the primary vessel 10.
Table 4 below shows the results of operating calculations for the reactor on the basis of the calculations above.
It can be seen from the results of Table 4 that, for thermal powers of up to 20 MWth, operation with circulation solely by natural convection (without pumps) can be envisaged in the primary circuit and is sufficient to be able to transfer the heat to the wall of the primary vessel, provided that the temperature of the wall of the primary vessel 10 is at least 171° C. lower than the mean temperature of the salt at its hot temperature because:
Pth,transferred to the wall of the PV=SHsalt,wallΔTsalt,wall
where
heat transfer coefficient between the primary salt S and the inner wall of the primary vessel 10;
This temperature difference determines the temperature of the coolant that will flow in the serpentine coil 32 on the outside of the primary vessel 0.
The greater the temperature difference between the primary salt S and the inner wall of the primary vessel 10 (ΔTsalt,wall), the lower the temperature of the coolant in the serpentine coil 32 will be, with a resulting deterioration of the electric conversion efficiency of the energy conversion system to which the serpentine coil 32 is connected.
The invention is not limited to the examples that have just been described; features of the illustrated examples can in particular be combined together within variants that are not illustrated.
Further variants and embodiments can be envisaged without departing from the scope of the invention.
| Number | Date | Country | Kind |
|---|---|---|---|
| 23 01895 | Mar 2023 | FR | national |
