Thorium chloride salts in molten salt nuclear reactors

Abstract

A set of salt compositions containing thorium tetrachloride, for use in molten salt nuclear reactors, is disclosed. The first subset embodiments disclosed involve mixtures of ThCl4 with alkali chlorides NaCl, KCl, and RbCl, alkaline earth chlorides MgCl2, CaCl2, or mixtures thereof, in different proportions. The most useful composition ranges are specified. A second subset of embodiments further includes non-thorium actinide chlorides such as UCl3, UCl4, PuCl3, and the minor actinide trichlorides. The most preferred composition domains are also provided for this subset of embodiments, and were chosen to fit within a design envelope constrained by melting point, boiling point, viscosity, and fission product retention requirements. Additionally, the composition space includes embodiments which are useful in reactors intended to breed new fuel or instead to burn synthetic transuranic elements. In all cases, the presence of ThCl4 confers reduced corrosivity towards engineering alloys and a reduced tendency to degrade ceramics in nuclear reactor environments, compared to salts with a similar design envelope but without ThCl4.

Description

[0001] Title: Thorium chloride salts in molten salt nuclear reactors

[0002] Field of the Invention

[0003] The invention relates to salt mixtures comprising Thorium chloride and their use in a molten salt nuclear reactor systems. The invention further pertains to the use of thorium chloride- containing salt mixtures in reducing the corrosivity of the molten salt towards (transition) metals and / or reducing ceramic degradation in molten salt nuclear reactors.

[0004] Background of the invention

[0005] With the world’s increasing demand for energy, there remains a need for nuclear energy despite the disadvantages associated with early generation nuclear plants. New generations of nuclear plants (4thand 5thgeneration) are being developed that specially aim to supply safe and clean abundant amounts of energy and at the same time take a conscious approach to the proliferation of nuclear material and the generation of nuclear waste.

[0006] One of these ‘waste materials' is plutonium, a long living isotope of which large supplies have been accumulated over the years and that has a large risk of proliferation as it can be used also for non-peaceful activities. Attempts have been made to decrease the volume of plutonium.

[0007] For instance, plutonium is added to present day mixed oxide (MOX) fuels, yet a large fraction of the initial loading remains. The substantial remaining plutonium which can be recovered from spent MOX fuel is enriched in non-fissile isotopes, and reprocessing is uneconomical. Furthermore, only a small percentage (< 5%) of minor actinides may be added to MOX fuels. The complexity and costs of MOX fuel production, mainly lies in the tight quality assurance and control required for solid fuels, the (re-)qualification of composition variations, for example (potential) heterogeneity or isotopic content variations, the limits of introducing minor actinide (<5%), in combination with the radiological hazards of handling and managing radiotoxic powders and dust.

[0008] One of the newer generation nuclear plants are so-called molten salt reactors.

[0009] A molten salt reactor (MSR) is a type of nuclear reactor where the primary coolant, or even the fuel or target material itself is a molten salt mixture. There have been many designs put forward for this type of reactor and a few prototypes built. The early concepts and many current ones rely on nuclear fuel dissolved in molten fluoride salt. The fluid would reach criticality by flowing into a core where a moderator such as graphite can be present. Many known concepts rely on fuel flowing through channels in a graphite matrix with the molten salt providing low pressure, high temperature cooling. Some newly developed concepts exclude the use of a moderator to create a fast neutron spectrum system characteristic, or adopt specific moderation to generate dedicated neutron spectra (even locally).

[0010] Even though molten salt reactors have been successfully designed, built and operated in the past, the knowledge and experience required for this reactor type to be redeveloped is limited. The key advantages of molten salt reactor systems concerning safety, minimal waste streams, and maximized resource efficiency, have been elaborated in many instances, and with increasing frequency in the last five to 10 years.

[0011] In molten salt reactors, the coolant fuel composition can be adjusted relatively conveniently. Even though fluoride salts have been the most used and considered so far, largely due to the fact that fluorine is mono-isotopic and does not require enrichment, molten chloride salts have been proposed since very early on in Molten Salt Reactor (MSR) development. For instance, Bulmer, J. J. Fused salt fast breeder: reactor design and feasibility study. Vol. 56. No. 8-204. United States Atomic Energy Commission, Technical Information Service Extension, 1957 and Dumaire et al. (CALPHAD: Computer Coupling of Phase diagrams and Thermochemistry 79 (2022) 102496 describe chloride salts as fuel carriers in MSRs, focusing on the NaCI- ThCU-PuCh fuel salt solution as well as NaCI-ThCL-UCh and NaCI-ThCL-PuCh-UCh-UCL fuel salt systems. Ocadiz Flores et al. Thermo 2021, 1(2) 122-133 wrote a study about the thermodynamic description of ACI-ThCL (A= Li, Na, K) systems.

[0012] WO2017 / 106509 described salt compositions for molten salt reactors based on UCh, which may be in the combination of UCIS-PuCh-and discloses the use of ThCL as a fuel salt.

[0013] WO2017 / 106509 teaches that melting point reduction can be achieved by UCI4. As the presence of UCI4 can lead to increased corrosivity of the structural alloys in view of the electrochemical potential, it is suggested to replace UCI4 with ThCL as it can provide a comparable reduction of the melting point but without the corrosive effects of UCI4.

[0014] US 2016 / 0189813 also describes molten salt reactors, focuses on the use of uranium as a fissile material and fuel embodiments including end-members such as UCI3, ThCL, UCI4, and PuCh, alkali chlorides, and alkaline earth chlorides.

[0015] Molten salt fuels, unlike solid ones like MOX, do not need to maintain thermomechanical properties or mechanical integrity, complex solid fuel requalification for each new composition and a complex, expensive and hazardous powder step in the manufacturing process. To be able to design, operate, and license an MSR, both the thermodynamic and transport properties of the fuel must be known as a function of temperature and composition. In fact this is true of any nuclear fuel, with the advantage that the properties of liquid fuels are state functions. That is, they do not depend on manufacturing history or are influenced by radiation damage, as is the case for solid fuels. Summary of the invention

[0016] The present inventors have now, in essence, found a set of chloride salt compositions suitable for use in molten salt reactors. The compositions are useful in a burning mode, i.e. burn synthetic transuranics, these being the major actinide plutonium and the minor actinides neptunium, americium, and curium. The compositions are also useful in a breeding mode, i.e. generate fissile isotopes (e.g.233U and239Pu), for the subsequent generation of energy. The compositions of the invention combine low melting temperatures, relatively high boiling points, reduced redox potential, and reduced solubility for transition metal chlorides, notably those of chromium, nickel, and iron which aids in reducing corrosivity.

[0017] Thus, in a first aspect the invention pertains to a salt mixture for a molten salt reactor comprising thorium tetrachloride (ThCU) and one or more of NaCI, KCI, RbCI, MgCh, CaCh, wherein the ThCU mole fraction in the mixture is from 0.05 to 0.70.

[0018] A second aspect relates to a molten salt nuclear reactor comprising the salt mixture of the invention.

[0019] A third aspect to the use of thorium chloride-containing salt mixtures in reducing corrosivity of the salt mixture towards metals and reducing ceramic degradation in molten salt nuclear reactors.

[0020] Brief description of the drawings

[0021] Fig. 1 shows a quasi-ternary composition diagram displaying a domain , having vertices with composition coordinates 1a: 0.95MCIx-0.05ThCl4, 1 b: 0.3MCIx-0.7ThCl4, 1c: 0.57MCL- 0.03ThCl4-0.4AnClyand 1d: 0.18MCIx-0.42ThCl4-0.4AnCly. Another preferred domain, with coordinates 2a: 0.90MCIx-0.10ThCI4, 2b:0.35MCIx-0.65ThCI4, 2c: 0.585MCIx-0.065ThCI4- 0.350AnClyand 2d: 0.228MCIx-0.422ThCl4-0.350AnCly. Further preferred isa domain with coordinates3a: 0.85MCIx-0.15ThCI4, 3b: 0.5MCIx-0.5ThCI4, 3c: 0.567MCIx-0.100ThCI4- 0.333AnCly, 3d: 1 / 3MCIx-1 / 3ThCl4-1 / 3AnCly. MCIXdenotes either an alkali chloride, alkaline earth chloride, or mixture thereof, while AnClydenotes an actinide chloride other than ThCk Fig. 2 shows a pseudobinary phase diagram between Na2ThCle and an actinide trichloride mixture made of NpCh-PuCh-AmCh in a 1 :1 :1 ratio, up to X(AnCh) = 0.5.

[0022] Fig. 3 is a solubility curve of CrCh in different low-melting chloride mixtures.

[0023] Detailed Description of the invention

[0024] In a first aspect the invention pertains to a salt mixture for a molten salt reactor comprising thorium tetrachloride (ThCU) and one or more of NaCI, KCI, RbCI, MgCh, CaCh, wherein the ThCU mole fraction in the mixture is from 0.05 to 0.70, preferably from 0.10-0.65, more preferably from 0.15-0.50. The salt mixture of the present invention contains thorium chloride in combination with one or more alkali (Na, K, Rb) or alkaline earth metal (Mg, Ca) chlorides or a combination of alkali (Na, K, Rb) and alkaline earth metal (Mg, Ca) chlorides. These can be in binary, ternary or quaternary mixtures, i.e. ThCU in combination with one, two or three further salts selected form NaCI, KCI, RbCI, MgCh, and / or CaCh. The ThCU mole fraction in the mixture is from 0.05 to 0.70, preferably from 0.10-0.65, more preferably from 0.15-0.50.

[0025] The embodiments disclosed herein are described in a chemical composition state prior to irradiation in a nuclear reactor. Those skilled in the art know that during operation, fission, activation, and corrosion products will also be present.

[0026] The choice of the salt mixtures of present invention present certain advantages over the art. For molten salt reactors historically the focus was on fluoride salts. Hence much remains to be determined related to chloride salts for molten salt reactor applications. The inventors have found compositions which are useful in the design of molten salt reactors and aims to close the current nuclear fuel cycle (i.e. make maximum use of nuclear materials resources, while minimizing the amount of long-lived radiotoxic elements generated, or even reduce the amount of long-lived radiotoxic elements). In combination with an earlier published invention of the inventors (WO2020 / 225156) focused at a modular molten salt reactor and the possibility of the use of thorium in many embodiments, the thorium fuel cycle (conversion of Th to II) may also be begun in parallel. The salt compositions of the invention are rather specific and are selected to meet performance requirements in several thermophysical characteristics such as melting point and boiling point. Chloride salts can be more easily produced from reprocessing facilities, can accommodate higher actinide concentrations which are favourable for fast spectrum applications, provide lower melting temperatures leading to more manageable system operating temperatures and lower heat load, while the enrichment technology has been developed to levels where enriched chlorine is both available and affordable.

[0027] Molten salt systems containing natural actinides have certain advantages. Thorium is one such actinide. Its key advantages, used in molten salt reactor systems or in other nuclear reactor systems, include: i) the relative abundance of thorium on the Earth’s crust, especially when compared to that of uranium; ii) its widespread availability as a by-product of rare earth mining, iii) it does not need enrichment; iv) the hard gamma emitters in its fuel cycle, desirable from a safeguards perspective; v) U-233 fuel bred from Th-232 has favorable fission characteristics; vi) a reduced tendency to produce transuranics. The latter characteristic makes it especially preferred as a component in nuclear fuels designed to produce a net decrease in the inventory of transuranics.

[0028] The compositions presented herein satisfy several constraints chosen by the inventors to make them useful in molten salt reactors.

[0029] Thus, in preferred embodiments of invention, the salt mixture is a mixture of ThCU with NaCI, KCI, or RbCI, wherein the ThCU mole fraction in the salt mixture is from 0.05 to 0.70, , preferably 0.10-0.65, more preferably from 0.15-0.50.

[0030] In one preferred embodiment, the second salt is NaCI, more preferably the salt mixture comprises NaCI-ThCU and preferably the mole fraction of ThCU in the salt mixture is 0.20 - 0.55, preferably 0.22-0.50, more preferably 0.25-0.46. In one embodiment, a third salt may be present, a mixture of ThCU with NaCI and RbCI . In this preferred embodiment, the third salt is RbCI, more preferably the salt mixture comprises NaCI-RbCI-ThCU and preferably the mole fraction of RbCI in the NaCI-RbCI mixture is 0.40-0.70, more preferably 0.45-0.65, even more preferably 0.50 - 0.60.

[0031] In one preferred embodiment, the second salt next to ThCU is KCI, and preferably the mole fraction of ThCU in the salt mixture is 0.40-0.60, preferably 0.43-0.55, more preferably 0.45- 0.52, even more preferably 0.46-0.50.

[0032] In one preferred embodiment, the second salt is RbCI, and the mole fraction of ThCU in the total salt mixture is 0.40-0.60, preferably 0.43-0.55, more preferably 0.45-0.52, even more preferably 0.46-0.50.

[0033] In one preferred embodiment, the second salt is CaCh, and the mole fraction of ThCU in the total salt mixture is 0.40-0.60, preferably 0.43-0.55, more preferably 0.45-0.52, even more preferably 0.46-0.50.

[0034] In preferred embodiments of the salt mixture comprising a second and a third salt next to ThCU, the mole fraction of ThCU in the salt mixture is 0.40-0.60, preferably 0.43-0.55, more preferably 0.45-0.52, even more preferably 0.46-0.50.

[0035] In this preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU with, in addition to ThCU, two or more salts selected from the group consisting of NaCI, KCI, RbCI, MgCh, or CaCh. A preferred mixture is thorium chloride in combination with a salt combination selected from the group consisting of NaCI-KCI, NaCI-RbCI, KCI-RbCI, NaCI- CaCh, NaCI-MgCh, KCI-MgCh, or RbCI-MgCh.

[0036] In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and a NaCI-KCI mixture wherein the mole fraction of KCI in the NaCI-KCI mixture is 0.35-0.65, preferably 0.40-0.60, more preferably 0.45 - 0.55. In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and a KCI-RbCI mixture wherein the mole fraction of RbCI in the KCI-RbCI mixture is 0.50-0.70, preferably 0.52-0.65, more preferably 0.54 - 0.62.

[0037] In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and an NaCI-CaCh mixture wherein the mole fraction of CaCh in the NaCI-CaCh mixture is 0.40- 0.60, preferably 0.45-0.58, more preferably 0.48-0.56, even more preferably 0.50 - 0.54.

[0038] In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and an NaCI-MgCh mixture wherein the mole fraction of MgCh in the NaCI-MgCh mixture is 0.30- 0.55, preferably 0.35-0.50, more preferably 0.38-0.48, even more preferably 0.40 - 0.46.

[0039] In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and a KCI-MgCh mixture wherein the mole fraction of MgCh in the KCI-MgCh mixture is 0.20-0.40, preferably 0.25-0.35, more preferably 0.28 - 0.33.

[0040] In one preferred embodiment of the salt mixture, the salt mixture is a mixture of ThCU and an RbCI-MgCh mixture wherein the mole fraction of MgCh in the RbCI-MgCh mixture is 0.20- 0.35, preferably 0.22-0.33, more preferably 0.24 - 0.28.

[0041] In one preferred embodiment wherein the salt mixture is a mixture of ThCk- NaCI-KCl-MgCh, in the NaCI-KCl-MgCh mixture, i. the mole fraction of NaCI is 0.15-0.25, preferably 0.17-0.23, more preferably 0.19 - 0.21 ; ii. the mole fraction of KCI is 0.45-0.60, preferably 0.47-0.55, more preferably 0.52 - 0.53; iii. the mole fraction of MgCh is 0.20-0.32, preferably 0.25-0.30, more preferably 0.27 - 0.28;

[0042] In one preferred embodiment, the salt mixture is a mixture of ThCk with three salts selected from the group consisting of NaCI, KCI, RbCI, CaCh. A preferred salt mixture is, in addition to ThCI4, KCI-RbCI-MgCh.

[0043] In one preferred embodiment wherein the salt mixture is a mixture of ThCk- KCI-RbCI-MgCh, in the KCI-RbCI-MgCh mixture, i. the mole fraction of KCI is 0.40-0.50, preferably 0.41-0.47, more preferably 0.42 - 0.43; ii. the mole fraction of RbCI is 0.22-0.25, preferably 0.23-0.26, more preferably 0.25 - 0.27; iii. the mole fraction of MgCh is 0.25-0.40, preferably 0.27-0.37, more preferably 0.30 - 0.35;

[0044] In one preferred embodiment wherein the salt mixture is a mixture of ThCk- KCI-RbCI-MgCh, in the KCI-RbCI-MgCh mixture, i. the mole fraction of KCI is 0.40 - 0.55, preferably 0.45-0.50, more preferably 0.47 - 0.49; ii. the mole fraction of RbCI is 0.14-0.20, preferably 0.15-0.19, more preferably 0.16 - 0.18; iii. the mole fraction of MgCh is 0.25-0.45, preferably 0.28-0.42, more preferably 0.32 - 0.40.

[0045] Table 1 lists examples of useful embodiments within the composition spaces discussed above, satisfying melting point, boiling point, and viscosity requirements.

[0046] Melting point

[0047] Within the scope of the embodiments, the inventors have found melting points lower than 670 K ( -400 degrees Celsius). This is lower than what has been reported to date in the art for similar actinide-bearing chloride salts. Additionally, the inventors have found broad areas in the vicinity of the eutectics in which the liquidus surface rises slowly with respect to changes in composition. The preferred salt mixtures of the invention share that the salt mixture has a melting point below 600 degrees Celsius, preferably below 500 degrees Celsius, more preferably below 400 degrees Celsius.

[0048] Boiling point A core meltdown scenario is not possible in molten salt reactors, since the fuel is already in the liquid state. The main criterion for composition selection in WO 2017 / 105609 A1, for example, is the melting point. However, the boiling point is just as crucial, as a plant shutdown scenario in which the decay heat from fission products is not removed at an adequate rate, due to loss of secondary coolant for instance, may lead to temperature increases which could bring the fuel salt to boil.

[0049] The increases in pressure associated with such boiling may result in breach of containment, just like a meltdown does. This outcome is to be avoided by any nuclear plant in any scenario, be it during normal operations or accidental conditions. In the quasi-ternary phase diagram of in Fig. 1 , the composition domain is chosen because the boiling point is more than about 1300 K (-1027 degrees Celsius) everywhere. Combined with the aforementioned low melting points generally found within the domain, fuel compositions with a large solidification- to-boiling margin are found.

[0050] The preferred salt mixtures of the invention share in preferred embodiments that the salt mixture has a boiling point higher than 1000 degrees Celsius, preferably higher than 1100 degrees Celsius or even 1200 degrees Celsius, even more preferably higher than 1300 degrees Celsius (see Table 1). The salt mixtures of the invention are hence advantageously suitable as salt mixture for molten salt nuclear reactors.

[0051] Viscosity

[0052] The principal domain in the phase diagram in Fig. 1 and hence the salt mixture of the invention is also advantageous because of a favorable viscosity. The viscosity found for the salt mixtures of the invention is found to have an upper bound of about 30 mPa.s near the lowest eutectic temperatures within the domain (e.g. T < 400 degrees Celsius), and of about 5 mPas (see Table 1) at widely foreseen operating temperatures for molten salt reactors, e.g. above 500 degrees Celsius. During operation, the viscosity of the fuel is expected to remain within these bounds with fuel burnup, as a great portion of the fission products will be lanthanide trichlorides with similar viscosities as those of the actinide trichlorides. Alkali chlorides and alkaline earth metal chlorides, which will also be soluble in the fuel matrix, in general have a lower viscosity at a given temperature than both lanthanide and actinide trichlorides, and so are also not expected to raise the upper bounds in viscosity. On the other hand, insoluble fission products such as noble gases and metals will have a negligible effect on the liquid fuel viscosity, as they will plate and bubble out of solution, respectively.

[0053] The salt mixture of the present invention can be used in a breeding mode. In breeding, in a nuclear reactor, the salt mixture comprising thorium is subjected to a neutron flux allowing thorium to be converted into fissionable product such as uranium. The salt mixture containing uranium can be used as a fuel. The decay daughters of U-232, present in the uranium vector bred from Th-232, emit hard gamma radiation which might give the fuels and blankets disclosed herein strong nonproliferation characteristics. Indeed, such an intense gamma radiation makes unauthorized handling of these salt mixtures very difficult to conceal from relevant safeguards authorities such as Euratom or the International Atomic Energy Agency. To improve the non-proliferation characteristics of these salts, it may also be desirable to add natural or depleted uranium as UCh or IICI4, such that the U-233 content in the uranium vector becomes diluted. Such an addition may be in the order of 1-20 mol% of the ThCU mole fraction.

[0054] Thus in certain embodiments II, preferably in the form of salts can be added, preferably in the form of natural or depleted uranium, i.e. rich in the U-238 isotope.

[0055] In a further embodiment of the invention, the salt mixtures of the invention are combined with an actinide chloride (other than thorium chloride). Thus the salt mixture is made up of three and up to four chloride salt families: thorium tetrachloride, an alkali chloride and / or alkaline earth metal chloride, or a mixture of alkali chloride and alkaline earth chloride, and a further actinide chloride. This kind of embodiment affords more flexibility and allows the use of thermal, epithermal, and fast reactor designs and to burn the major actinide plutonium, and the minor actinides neptunium, americium, and curium.

[0056] In a preferred embodiment, the use of the salt composition in a nuclear reactor is aimed at the burning of plutonium and / or other minor actinides. Thus in a preferred embodiment, the salt mixture, in addition to the thorium and the alkali and alkaline earth chlorides, further contains other actinide chlorides than thorium. These actinide chlorides can be one or more chosen from the group consisting of UCh, UCI4, NpCh, PuCh, AmCh, and CmCh. In embodiments wherein the emphasis is to burn plutonium and the minor actinides, PuCh, NpCh, AmCh, and CmCh are preferred as the main actinide components. Such a salt mixture can also be used as an actinide burning blanket.

[0057] In certain embodiments , 1-20% of the ThCU mole fraction can be replaced by a natural or depleted uranium salt, preferably UCh and / or IICI4. For example, in an initial salt with the following composition in mole fractions: 0.60NaCI-0.20ThCI4-0.20PuCI3, 20% of the ThCI4 fraction is replaced with LICI3 (20% of 0.2 is 0.04), the overall composition ends up being 0.60NaCI-0.16ThCI4-0.04UCI3-0.20PuCI3. One reason for the addition of a uranium salt is to dilute the U-233 vector bred from ThCU with other uranium isotopes, as described above. Another is to breed Pu-239 from U-238 in order to increase time between fuel loadings or even to fuel other reactors in a reactor fleet (i.e. a constellation of multiple reactors). In certain embodiments the amount of UCU is at most 5% of the uranium mole fraction calculated on the total amount of IICI3 and UCU, so as to limit the redox potential of the fuel, as well as its volatility.

[0058] An example thereof is present in Fig 1 which presents a quasi-ternary composition diagram. One end-member is an alkali chloride, alkaline earth chloride, or a mixture of both, denoted by MClx. Another end member is thorium tetrachloride (ThCU). Yet another member is an actinide chloride (not ThCU), either pure or mixed with other actinide trichlorides. KCI-ThCU- NpCh is thus a system which belongs to this notation, for example, but so is (Naa, Mgb)Cla+2b- ThCI4-(Pu, Am)CI3.

[0059] The salt mixture of the invention can also be characterised as a composition that, preferably in a ternary phase diagram, lies between the following mole fractions: 0.95MCIx-0.05ThCl4, 0.7MCIx-0.3ThCI4, 0.57MCIx-0.03ThCI4-0.4AnClyand 0.18MCIx-0.42ThCI4-0.4AnCly.

[0060] In a preferred embodiment, the salt mixture of the invention can be characterised as a composition that, preferably in a ternary phase diagram, lies between the following mole fractions: 0.90MCIx-0.10ThCI4, 0.65MCIx-0.35ThCI4, 0.585MCIx-0.065ThCI4-0.350AnClyand 0.228MCIx-0.422ThCI4-0.350AnCly.

[0061] In a more preferred embodiment, the salt mixture of the invention can be characterised as a composition that, preferably in a ternary phase diagram, lies between the following mole fractions: 0.85MCIx-0.15ThCI4, 0.5MCIx-0.5ThCI4, 0.567MCIx-0.100ThCI4-0.333AnCly, 1 / 3MCIx-1 / 3ThCl4-1 / 3AnCly. In these embodiments M is an alkali or alkaline earth selected from the group consisting of Na, K, Rb, Mg, Ca and An is a non-Thorium actinide.

[0062] FIG. 1 displays a shaded domain, having vertices with composition coordinates 0.85MCIx- 0.15ThCI4, 0.5MCIx-0.5ThCI4, 0.567MCIx-0.100ThCI4-0.333AnCly, 0.333MCIx-0.333ThCI4- 0.333AnCly, where AnClydenotes an actinide chloride other than thorium tetrachloride. That is, in preferred embodiments, the global ThCU mole fraction desired is 0.15< X(ThCl4) 0.5, while the remaining actinide chloride mole fraction is preferably between 0< X(AnCly) < 0.333. As an example, some of the preferred embodiments may be chosen from the KCI-ThCU- PuCI3system, within the domain of the invention (FIG. 1), such as compositions in the vicinity of 0.49KCI-0.37ThCI4-0.14PuCI3and 0.56KCI-0.18ThCI4-0.26PuCI3. In related embodiments, AmCI3might be chosen to be 5 mol% of the overall (non-Th) actinide trichloride mole fraction, leaving PuCI3as the remaining 95 mol%, such that the global mole fractions become 0.49KCI-0.37ThCI4-0.133PuCI3-0.007AmCI3and 0.56KCI-0.18ThCI4-0.247PuCI3-0.013AmCI3, respectively.

[0063] Another preferred embodiment might be chosen from the quasi-binary system formed by the phase Na2ThCle and a 1 :1 :1 mixture of NpCI3-PuCI3-AmCI3. The quasi-binary phase diagram as calculated by the inventors is shown in Fig. 2, in the range 0< X(AnCh) 0.5. It can be seen that the eutectic point at an actinide trichloride mole fraction close to X(AnCh) = 0.27 is below 600 Kelvin (32 degrees Celsius) , which makes it a very attractive composition for actinide burning in a molten salt reactor. To visualize this in FIG. 1 , the composition for this embodiment would have coordinates X(NaCI)= 0.593, X(ThCl4) = 0.297, X(AnCh) = 0.11. Other preferred embodiments in this quasi-ternary system would include 0.579NaCI- 0.315ThCI4-0.107AnCI3, 0.638NaCI-0.257ThCI4-0.104AnCI3, and 0.515NaCI-0.399ThCI4- 0.086AnCI3.

[0064] The salts of the invention can be used in a molten salt reactor, in the ‘breeding’ configuration and / or in the ‘burning’ configuration, as explained herein elsewhere. A suitable configuration can be a conventional molten salt reactor. It can also be (preferably) a modular core reactor as described in WO2020 / 225156. In WO2020 / 225156 the configuration is made up of a constellation of independent salt-loaded cartridges. In embodiments, the salt cartridges individually are subcritical, but together form a critical constellation to sustain an fission reaction. It can be envisioned that the central cartridge, where the neutron flux is highest, could be loaded with such an actinide blanket to perform the transmutation. Alternatively, such a blanket could be used on the outermost cartridges of the core, using a longer residence time to achieve the same level of transmutation.

[0065] The vast potential of molten salt mixtures of the invention lies in the fuel cycle flexibility they can afford, namely in fissioning and thus extracting value from synthetic actinides which otherwise constitute a radiotoxic burden to be managed at great expense for generations. These synthetic transuranic actinides include the major actinide plutonium and the minor actinides neptunium, americium, and curium. They comprise the bulk of what is termed long- lived nuclear ‘waste’.

[0066] Retention of fission products

[0067] In a molten salt reactor, the chemical speciation of fission products may be classified as saltsoluble, metallic, and noble gases. Within the salt soluble products, there are also those with a high volatility or tendency to evaporate, and their tendency to remain in the fuel salt may not be taken for granted, as in WO 2017 / 105609 A1 , where it is declared that ‘fission products in molten fuel salts are chemically bound and physically frozen’. Csl is expected to be one a fission product with a tendency to vaporize from the salt. Due to the relatively high fission yield of both cesium and iodine, the relatively long half-life of Cs-137 (30 years), and the possibility iodine has to accumulate in human biology, it is important to design fuel mixtures which are able to retain Csl as much as possible. The salt mixtures of the invention express a higher retention of Csl than other viable MSR fuels, such as those based on the LiF-ThF4 system. Thermodynamic calculations in the scientific literature point to a vapor pressure of 25.7 Pa for the 75.4LiF-23.6ThF4-1.0Csl mixture at 884 K, whereas the vapor pressure near the eutectics in the chosen domain has an upper bound of about 2 Pa at the same temperature for the same, realistic content of 1 mol% Csl. Hence, the retention of Csl in the chosen salt mixtures of the invention is improved with respect to other relevant MSR fuel families.

[0068] Reduced corrosivity via mass transport

[0069] Iron, nickel, and chromium make up a large fraction of typical steels and alloys used and foreseen in the nuclear industry. One of the issues with nuclear reactors lies in corrosion. It is well established from thermodynamic considerations and static and dynamic experiments that the most susceptible metal to corrosion in chloride media is chromium.

[0070] Metal chlorides and particularly CrCh will dissolve in the salt matrix, such that a structural alloy will not be protected from corrosion by passivation. The inventors have found that the solubility curves for CrCh in different low-melting salt media: LiCkKCI = 1 :1 , 0.75NaCI- 0.25ThCI4, 0.632NaCI-0.171 MgCh -0.196PuCI3, and 0.470NaCI-0.354ThCI4-0.176PuCI3(Fig. 3). The solubility is expressed as grams of CrCh per 100 g of solvent as a function of temperature. It can be seen that at all temperatures, the solubility of CrCh is reduced in the ThCI4-bearing solvents, as is the solubility gradient, compared to non-Th containing solvents (molten salts). In the case of a binary mixture, CrCh is less soluble in 0.75NaCI-0.25ThCI4than in LiCI-KCI, while in the case of ternary mixtures, CrCh is less soluble in 0.470NaCI- 0.354ThCI4-0.176PUCI3than in 0.632NaCI-0.171MgCh -0.196PuCI3.Thus, the use of ThCI4in salt mixtures of the invention reduces corrosivity of the resulting mixtures and the solubility gradient. The invention thus also involves the use of the salt mixture of the present invention comprising ThCI4in the reduction of corrosivity and in the reduction of the solubility of transition metal chlorides, especially CrCh, in molten salt reactors. Since the mass transport of metal from the hot section of a molten salt circuit to its cold leg is proportional to the solubility gradient between the sections, the reduced solubility of the transition metal chlorides in the salt mixtures of the invention is an important feature and advantage of the present invention, especially because large thermal gradients are desired to maximize the power extracted from a reactor. In the tables below, corrosion rates are given in micrometers per year.

[0071] Table 2. Corrosion rates of binary mixtures of NaCI with selected chloride salts.

[0072] Table 3. corrosion rates of NasPuCh mixed with 15, 33, and 50 mol% of ThCI4, MgCh and UCI3.

[0073] Table 4. corrosion rates of 0.9NaCI-0.1 PuCh mixed with 15, 33, and 50 mol% of ThCI4, MgCh, and UCI3.

[0074] It can be seen that for the binary mixtures and a given mole fraction (Table 2), the corrosion rates are lower for ThCU-containing mixtures.

[0075] In Table 3, a clear trend can be observed, with corrosion rates being lowest when ThCI4is present, followed by IICI3 and finally MgCh.

[0076] In Table 4, the corrosion rates again develop in the order ThCI4< UCI3 < MgCh. .

[0077] In all cases, the presence of ThCI4has a beneficial effect on the corrosion rate.

[0078] Use of salt mixtures of the invention preserving or maintaining the integrity of ceramic structural materials

[0079] Promising ceramic materials, chiefly silicon carbide, be it in monolithic or fiber composite form with SiC or carbon fibers, are vulnerable to attack by molten salt media by attack of any free silicon in the matrix, for example by tetravalent uranium: 4UCI4+ Si = UCI3+ SiCI4

[0080] The combination of actinide trichlorides with ThCI4and optional minor additions of UCI4minimize said attack and the production of volatile silicon tetrachloride. First, because the fission of the actinide trichlorides mostly results in a combination of noble metals or gases and lanthanide trichlorides which are able to hold chlorine largely to the same extent as the original salt: the net valence state changes little. In parallel, burnup calculations by the inventors have shown that thorium breeds more uranium slowly. Whatever uranium-233 is bred has favorable fissile characteristics, keeping the overall inventory low. Finally, whatever uranium tetrachloride is able to form at a given level of redox potential has its chemical activity diminished, in effect being dissolved by ThCI4. Other ceramic materials such as carbon fiber composite, graphite, AI2O3, BeO, MgO, silicon nitride, silicon oxide, Y2O3- stabilized zirconia, ZrO2, may also be corrosion-resistant towards mixtures in the chosen region (Fig. 1). Thus the salt mixture of the invention can also be used to reduce deterioration of ceramic materials in molten salt reactors.

Claims

CLAIMS1. Salt mixture for a molten salt reactor comprising thorium tetrachloride (ThCL) and one or more of NaCI, KCI, RbCI, MgCh, CaCh, wherein the ThCL mole fraction in the mixture is from 0.05 to 0.70.

2. Salt mixture according to claim 1 , wherein the salt mixture comprises NaCI-ThCL and the mole fraction of ThCL in the total salt mixture is 0.20 - 0.

553. Salt mixture according to claim 2, wherein the salt mixture further comprises RbCI and the mole fraction of RbCI in the NaCI-RbCI mixture is 0.40-0.70.

4. Salt mixture according to claim 1 , wherein the mole fraction of ThCL in the total salt mixture is 0.40-0.60 and the salt mixture further contains a salt or salt mixture selected from the group consisting of (a) to (I): a. KCI; b. RbCI; c. CaCh; d. NaCI-KCI mixture wherein the mole fraction of KCI in the NaCI-KCI mixture is 0.35-0.65; e. KCI-RbCI mixture wherein the mole fraction of RbCI in the KCI-RbCI mixture is 0.50-0.70; f. NaCI-CaCh mixture wherein the mole fraction of CaCh in the CaCh-RbCI mixture is 0.40-0.60; g. NaCI-MgCh mixture wherein the mole fraction of MgCh in the NaCI- MgCh mixture is 0.30-0.55; h. KCI-MgCh mixture wherein the mole fraction of MgCh in the KCI- MgCh mixture is 0.20-0.40; i. RbCI-MgCh mixture wherein the mole fraction of MgCh in the RbCI- MgCh mixture is 0.20-0.35; j. NaCI-KCl-MgCh wherein, in the NaCI-KCl-MgCh mixture, i. the mole fraction of NaCI is 0.15-0.25; ii. the mole fraction of KCI is 0.45-0.60; iii. the mole fraction of MgCh is 0.20-0.32; k. KCI-RbCI-MgCh wherein, in the KCI-RbCI-MgCh mixture, i. the mole fraction of KCI is 0.40-0.50; ii. the mole fraction of RbCI is 0.22-0.25; iii. the mole fraction of MgCh is 0.25-0.

40. l. KCI-RbCI-MgCh wherein, in the KCI-RbCI-MgCh mixture, i. the mole fraction of KCI is 0.40 - 0.55; ii. the mole fraction of RbCI is 0.14-0.20;iii. the mole fraction of MgCh is 0.25-0.45.

5. Salt mixture according to any of the previous claims, further comprising a nonthorium actinide preferably a non-thorium actinide is a non-thorium actinide chloride, more preferably a synthetic actinide chloride, preferably selected from the group consisting of plutonium chloride, neptunium chloride, americium chloride and curium chloride.

6. Salt mixture as defined in any of the previous claims, wherein 1-20 mol% of the ThCU mole fraction is replaced by a uranium salt, preferably UCh and / or IICI4.

7. Salt mixture according to claim 6, wherein the amount of IICI4 is at most 5 mol% of the uranium mole fraction calculated on the total amount of UCh and IICI4.

8. Salt mixture according to claim 5, wherein the synthetic transuranic salts are present in an amount of up to a mole fraction of 0.01-0.4 calculated on the total salt mixture.

9. Salt mixture according to claim 8, wherein the salt mixture has a composition that in a ternary phase diagram lies between the following mole fractions: 0.95MCIx- 0.05ThCI4, 0.3MCIx-0.7ThCI4, 0.57MCIx-0.03ThCI4-0.4AnClyand 0.18MCIx- 0.42ThCl4-0.4AnClywherein M is an alkali or alkaline earth selected from the group consisting of Na, K, Rb, Mg, Ca and An is a non-Thorium actinide.

10. Salt mixture according to any of the previous claims, wherein the salt mixture has a melting point below 600 degrees Celsius and / or a boiling point higher than 1000 degrees Celsius.11 . Use of a salt mixture as defined in any one of the previous claims in a molten salt reactor, preferably a modular nuclear salt reactor.

12. Use of a salt mixture as defined in any one of the previous claims in reducing temperature gradient corrosive degradation and / or ceramic degradation.

13. Molten salt nuclear reactor comprising a salt composition as defined in any of the claims 1-10.

14. Molten salt nuclear reactor according to claim 13, wherein the molten salt nuclear reactor is a modular nuclear salt reactor comprising a plurality of modules.

15. Molten salt nuclear reactor according to claim 14, wherein optionally each module contains a sub-critical amount of nuclear material.

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