This application claims priority to Canadian Patent Application No. 3,011,398, filed on July 16, 2018.

FIELD OF THE INVENTION

This invention provides a process and apparatus for a permanent destruction of long- lived radioactive nuclear waste.

BACKGROUND

The destruction of nuclear waste is an issue of enormous importance not only to individual countries but to the whole human civilization. There are three important factors: 1) long-lived radioactive materials can retain their toxicity for millions of years; 2) the toxicity of nuclear materials exceeds the toxicity of any known chemical poison; 3) nuclear materials are capable to contaminate and make non-habitable vast areas of land. Potentially, nuclear waste can make the whole planet non-habitable for millions of years.

The worldwide inventory of bumed-up nuclear fuel was 266,000 metric tons in 2014. This inventory increases by 70,000 metric tons every year (World Nuclear Association,

2015). Currently, an economically acceptable method for the destruction of the radioactive waste simply does not exist. The main, under consideration, way of dealing with the nuclear waste is the "geological depository" approach. With the "geological depository" approach, the nuclear waste will be deposited in stable geological formations, while waiting for the emergence of a technology capable to transmute the radioactive isotopes into stable elements. For example, Finland, which gets 26% of its electrical energy from nuclear reactors, is planning to build a 70 km depository tunnel in the Onkalo Area. That area would then be sealed for the next 100,000 years (“How to dispose of the nuclear waste”, 2017). In the USA, the Yucca Mountain is under consideration for a similar depository. Expecting political and geological stability for the next 10 million years is a subject of concern to the general public.

In recent years, five technologies for the transmutation of nuclear waste (Trans- Uranic Residue, TRU) were proposed:

1. A sub-critical, accelerator driven reactor (ADS). This method will be very costly, as the cost of a proton accelerator with the energies in the GeV range (needed for this process) is likely to be in the range of a billion dollars. The irradiated TRUs would still require multiple remote re-processings. The costs of remote reprocessing are very high, often about $1300 per kilogram of the re-processed material. The accelerator-driven sub-critical installation is not expected to be very efficient, particularly when the processed nuclear materials will include Plutonium. Computer modeling studies by Coates indicated, that contrary to the common opinions, the ADS reactor will generate Plutonium and the minor actinides up to certain equilibrium concentrations (Coates, 2010).

2. Recycling TRUs using a Reduced moderation Boiling Water Reactor (RBWR). This option was investigated by Lindlay (Lindlay, 2014), who calculated that an RBWR can be loaded with as much as 35% of TRU. However, since the RBWR uses solid fuels (packed fuel rods), the reprocessing costs will still be very high, as the TRU material must be recycled several times and each time the cost of the reprocessing can be more than $1300 per kilogram (Lindlay, 2014: Table 7.6). A practical reprocessing technology is currently not available: Lindlay declared that the "industrial re-processing of Thorium-irradiated fuel needs to be developed " (Lindlay, 2014: 198).

3. Thorium fast reactor, sodium cooled, with recycled TRU (Th-TRU-SFR). This reactor could potentially transmute 225 kg of TRU per year (Lindlay, 2014). The separation and recycling of the TRU isotopes will still be a problem. As with any other fast reactor, this reactor will be still facing safety and control issues, such as, for example, an accidental contamination with a moderator, or a meltdown. A meltdown of a fast reactor is likely to be much more dangerous than a meltdown of a thermal reactor.

4. Thorium Molten Salt Reactor (Th-MSR). This type of a nuclear reactor is known to convert the TRU materials into the fission products (FP). Therefore, this type of reactor could potentially be used for the transmutation of the separately loaded TRU materials. However, recycling still represents a problem, and the Protactinium produced from Thorium needs to be separated on a continuous basis. The technology of the Th-MSR reactors is still not very well developed. In particular, the issue of the continuous separation of Protactinium, Lanthanides and the Fission Products represents a significant problem.

5. Advanced Neutron Source (ANS) from Oak Ridge National Laboratories. This design is based on a reactor with a core having high neutron flux and a blanket of heavy water. The TRU isotopes will be dissolved in the blanket. These dissolved TRU-isotopes will be destroyed by the thermal and "hybrid“ neutrons, while the TRU-isotopes accumulating in the blanket (mostly Curium), will be separated, converted into solids and transferred into the core of the reactor. In this reactor, the reprocessing of the blanket and the core will still be needed, and the separated Curium would have to be converted to the solid form, to be inserted as rod- filling in the core of the reactor. So, the ANS-ORNL Reactor will still not solve the reprocessing problems (Iwasaki and Harakawa, 1995).

None of the technologies presented above deals effectively with the aspects of closing the material loop, such as repeated separations and the need for several reprocessing stages of the same material. Therefore, there is still a strong need for a simple and low-cost process, which will solve the reprocessing as well as safety problems related to the destruction of transuranic isotopes and the problems related to the cost, handling and storing of radioactive wastes.

SUMMARY OF THE INVENTION

The invented method is based on using a new type of a nuclear reactor with reduced moderation, called here: "The Reduced Moderation Homogeneous Water Reactor (RM- HWR)", which is based on salts of actinides dissolved in water. The content of the reactor will be semi-continuously processed, to recycle to the reactor the transuranic elements (TRU), U-233, Thorium and Plutonium. The fission products (stable and short-lived isotopes) will be removed and the Protactinium will be sent to a quarantine for spontaneous conversion to U-233.

The RM-HWR reactor will be connected with an on-site reprocessing-separation unit (R-SEP). The reduced moderation will be achieved by maintaining low concentration of water in the reactor and providing some moderator-free space (voids, sparging, and radiolysis gases) in the core of the reactor.

The concentration of water in the reactor will be approximately from 20 to 30 wt. %, or 25 to 30 wt.%. Optionally, an additional reduction of neutron moderation will be achieved by adding a miscible salt with low affinity to neutrons, such as lithium nitrate, which has a melting point of 55 °C. The addition of lithium nitrate can significantly reduce the amount of water needed in the RM-HWR reactor.

The low concentration of water will reduce the neutron moderation, thus increasing the number of fast and "hybrid" (hyperthermal) neutrons available in the reactor. The fast neutrons and the hybrid neutrons are more effective in fragmenting of minor actinides (TRU) than the thermal neutrons.

The radioactive waste materials to be destroyed can be optionally located in separated compartments within the reactor, where the active volume of these compartments could be controlled, so some of the compartments will function in a way similar to control rods. The control of the volume could be achieved by inserting or retracting metal cylinders into the parts of the pipes located in the core of the invented reactor, or by draining or partial draining of some of the selected pipes. Standard control rods typically used in nuclear reactors will also be used, but their role will be limited to serve as a back-up elements of the control system.

The main concept of the invention is to continuously maintain approximately the same, high concentration of the TRU materials within the reactor. The main operating goal of the installation is to keep the TRU materials at the highest technically acceptable

concentration at all times. At the startup, the metal content of the nuclear materials in the reactor will be close to the metal content used in the molten salt thorium reactor (shortly after the startup situation). With time, the enriched uranium from the startup stage will be replaced with Thorium, and the concentrations of the Major Actinides and U-233 will be approaching the dynamic, non- equilibrium steady state, analogous to that of the Marshalkin’s Thorium Reactor (Marshalkin, 2016, RU2634476C1), but with much higher content of the TRU materials, as the TRU materials will be continuously added.

Marshalkin calculated that a self-sustaining, water-moderated Thorium Reactor (with solid nuclear fuel encased within the fuel rods) would proceed to an equilibrium, where the metals would have the following concentrations (expressed in wt.% of total metals):

Thorium: 89 - 83%,

Uranium isotopes: 10%,

237-Np: 0.2%,

Plutonium: 0.5 % (includes 0.33% of Pu-238),

Americium: about 0.1%,

Curium about 0.1 %.

In the Marshalkin's Thorium Reactor, one metric ton of bumed-up thorium nuclear fuel will generate 54 kg of fission products. The reactor of the present invention will generate more fission products, because of the continuous presence of the TRU materials at a high concentration. Similarly to the Marshalkin’s reactor, the reactor, which is a part of this invention, will operate in a self-sustaining mode (it will require only natural Thorium as a feed) and, similarly as the Thorium Molten Salt Reactor (Th-MSR) or the Marshalkin's reactor, it will even produce some excess of Uranium-233. The Marshalkin’s reactor was not solving the problem of the separation of the Fission Products, as solid nuclear fuel would be used, and the separations according to Marshalkin would be carried out stepwise, as “campaigns”.

The present invention proposes a solution that is different from both the Thorium Molten Salt Reactor and from the Marshalkin's Reactor. The main differences are:

1. The nuclear materials will be used as salts dissolved in water.

2. The composition will be semi-continuously adjusted, approximately daily, every second day, every 20-40 hours, or every 25-35 hours, to compensate for the transmutations and fissions and to maintain approximately the same concentrations of all actinides in the reactor, at all times. This will assure the high rate of fissions and transmutations, which, in turn, will assure a high rate of destruction of the radioactive actinides.

3. Since the core of the reactor will have a high flux of neutrons and the blanket will assure high utilization of generated neutrons, the potential processing capability of the reactor of this invention will be similar to the ANS-ORNL. The capacity can be scaled up to approximately 3.2 metric tons of TRU materials per year. For the ANS-ORNL reactor, the destruction rate was estimated to be 240 kg of TRU material per year, which would correspond to the nuclear waste generated by a PWR with the total power of 8 GWe (a typical PWR, at 1 GWe, generates about 30 kg/year of the TRU actinides).

4. The semi-continuous composition adjustments will be based on draining small batches of the reactor's content. The small batches of the dissolved reactor salts, preferably without any conversion or with minimal adjustments, will be taken for batch-wise or continuous separations, which will separate the radioactive minor actinides (TRU) and U-233 for recycling, back to the reactor. The TRU materials, with the exception of Protactinium, will be recycled back to the reactor, to the minor actinides compartments or to the main volume; Protactinium will be transferred to the quarantine vessels. Thorium, Uranium-233 and Plutonium, will be sent back to the main part of the reactor, while the separated Fission Products (FP) will be sent for a long-term storage or and additional transmutations or fissions. The Fission Products (FPs) require between 100 to 300 years of "cooling" to have their radiation level reduced to the level of the background radiation. The 300-years of "cooling" is currently considered as an acceptable time for a safe repository.

Alternatively, the Fission Products could be separated into radioactive and stable isotopes. The radioactive isotopes (separated from the Fission Products) can be further irradiated to transmute them into stable isotopes.

It is expected that the concentration of Uranium-238, remaining from the burned- up nuclear fuel in the RM-HWR reactor will be very low, due to the pre-processing removal of Uranium. Optionally, Plutonium can be separated to be destroyed in designated installations using the present invention, but dedicated specifically to destroying the TRU materials combined with Plutonium.

5. The reactor can be operated at relatively low temperatures, such as, for example, 100- 200 °C, or l50-200°C, to generate power (in a pressurized system) or even at 60- 80 °C, for a non-pressurized system. The non-pressurized system will generate a significant amount of low-grade heat, and it will be less costly to build and maintain. The efficiency of thermal energy conversion into mechanical and electrical power for the RM-HWR will be significantly lower than the efficiency of the regular nuclear reactors. For example, for the pressurized system, assuming T(hot) = 150 °C and T(cold)= 60 °C, the efficiency of the ideal Carnot cycle would be 21.3%.

DETAILED DESCRIPTION OF THE INVENTION

An embodiment of this invention is presented in Fig. 1. The processing line will consist of a Pre-Processing unit (U-l), the Reactor unit (U-2), the Reactor's Separation unit (R-SEP), the Heat Exchangers (U-3), the Quarantine (U-4) and the Fission Products Unit (FP). The pre-processing unit will preferably remove Uranium isotopes from the burned up material to be processed. In one embodiment, illustrated in Fig. 5, the commercial UREX process will be appropriate for the pre-processing. This can be followed by the TRUEX, a commercial process for the separation of the TRU isotopes. The separated TRU isotopes will be sent to the reactor for the fragmentations. An alternative way of pre-processing of burned- up nuclear fuel will be to use the Reactor's Separation Unit (R-SEP), which is illustrated in Fig. 4. In this case, the Reactor's Separation Unit will be a large scale chromatography installation, preferably with the use of supercritical carbon dioxide.

The startup of the reactor will be similar to a startup of the Thorium-MSR, that is, the starting materials will be the salts of Thorium, Plutonium (optional) and enriched Uranium. However, according to the present invention, the salts will be preferably the nitrates or sulfates rather than fluorides, as in the classical MSR. The Reactor-Separation System will separate the transmuted and fragmented materials from the start-up stage, leading to the point where the criticality will be maintained only by the addition of natural Thorium (together with the TRU or the pre-processed bumed-up nuclear fuel to be destroyed), and, optionally, Plutonium. In an embodiment, the reactor will be used to generate power. If this is the case, the reactor would operate at temperatures exceeding 100 °C, and at pressures exceeding atmospheric pressure. Although the nitrate solutions in the presence of nitric acid are stable at the temperatures exceeding 200 °C (Lane, 1958), in an embodiment the reactor temperature is 150 °C, to maintain a safety margin. Under these conditions, the reactor will generate steam, which can be used to drive a turbine-generator producing electricity. Alternatively, the steam can be used for heating of districts in the cities and for industrial processes using heat.

The proposed reactor can be designed to be potentially most likely safer than most of reactors currently in use. Since the proposed RM-HWR will operate on the same principles as the known homogeneous aqueous reactors, it will be self-controlling and will be capable to withstand very large increases in reactivity. These characteristics were verified by the Atomic International Company in the late l940s, where in a so called“Kinetic Energy Experiment”, such a reactor was subjected to a stability test. In that test, the control rods were rapidly removed from the reactor (in a few milliseconds time), which resulted in a power surge. No problems were observed in this experiment (Wikipedia, 2018).

In the pressure version of the reactor, the reactor's core may consist of 3 to 6 main pipes and 3 to 6 smaller pipes, which will be in close proximity to each other, as shown in Fig· 2, which shows the cross section of the core of the reactor. In the implementation presented in Fig. 2, "1" represents the main pipes, containing Thorium, Plutonium and Uranium-233, while the smaller pipes "2" contain concentrated solutions of Curium and other isotopes from the minor actinide group, that will be separated from the blanket“3”, where they will tend to accumulate.

Beyond the core section, the pipes may be bent away from the center of the reactor. These pipes will be connected to the external heat exchangers "4" (Fig. 3). The circulation of the reactor material through the heat exchangers may be gravitational, due to the differences in the density of the fluids. The density differences will be caused by heat, gases generated due to radiolysis of water and by sparging with Helium. One method of sparging with Helium will be to use ejectors, acting as spargers and also as non-mechanical pumps. In another embodiment, the core of the reactor will consist of a single pipe or a vessel, surrounded with a blanket. The blanket section will contain dissolved TRU materials at a low concentration (e.g. 1 gram/L).

In one embodiment, the reactor core will have several smaller diameter pipes ("2", Fig. 2), filled with the solutions of separated TRU salts or just with Curium and similar hard to destroy isotopes. This approach will reduce the volume of the materials to be re- processed for the removal of the Fission Products (FP). Controlling the quantities of the TRU isotopes in the small tubes will permit fine tuning of the criticality conditions of the reactor.

As a provision for overheating, the reactor-pipes may have sub-critical vessels ("51", Fig. 3), connected with the reactor-pipe via rupture disks ("50"). On overheating, the rupture disks may yield and the content of the pipes discharged to the sub-critical vessels ("51").

Optionally, the reactor can be operated at low temperature, under conditions similar to a standard, homogeneous water reactor. The temperature in this case would be about 60- 90°C, or about 80 °C, and the generation of electricity would have to be similar to the technique used in the geothermal electricity generation from the low-grade heat sources. The advantage in this case would be in:

very high stability of the reactor,

proven and well known design,

ability to use a high volume reactor, not restricted by the pressure requirements,

lower risk of leaks from the process.

In the pressure-version of the reactor, the vapors may go into a packed (or shelved) rectification column ("5"), from which the bottoms, containing an azeotrope of water and nitric acid, may be returned to the reactor pipes, while water, separated at the top, may be metered back to the reactor, and that metering will provide an additional way to control the reduced moderation of the neutrons. The radioactive gas, Xenon-l35, may be recycled with the sparging gas to avoid its decomposition to Cesium-l35, which is a long-lived

radioisotope.

The Helium sparging gas may be used to dilute the water-radiolysis gases to the concentrations below the explosion level and to control the moderation, by creating voids in the liquid reactor mixture containing dissolved salts. The sparging gases and the radiolysis gases, after passing through the reactor and the distillation column, may be passed through a catalytic combustion bed ("6"), where the hydrogen and oxygen will react to form water. The heat from the catalytic combustion can be used to generate additional power.

The Separation Unit (R-SEP) is an important element of this invention, as it is more simple and it has higher resolution compared to the standard recycling processes. All currently proposed standard processes for the destruction of nuclear waste are encountering very significant problems with the recycling of the TRU isotopes back to the reactor. The costs of separation and recycling of the nuclear materials are so high that the industrial-scale destruction of the nuclear waste was not implemented, so far, in any country. However, proposed herein is a new method of processing for a low-cost destruction of nuclear waste, by integrating the nuclear reactor with a reprocessing installation. This integration is based on using the nuclear fuel and the nuclear waste as liquids. The liquid materials may be handled remotely. The remote handling of liquids is a mature technology and it is a standard practice in the chemical industry. This invention eliminates the practice of: making solid pellets of nuclear fuel, filling the rods with the pellets, cutting the rods with the bumed-up nuclear fuel and conversion of the oxides or carbides of the bumed-up nuclear fuel into nitrates for reprocessing. In the "Standard Processes", such as BMWR, ADS (Accelerator Driven Sub-Critical Reactor) and the Th-TRU-SFR, the strongly radioactive recycled materials have to be converted back to solids (oxides, carbides or nitrides), pelletized and inserted into the rods. In the ANS-ORNL (Advanced Neutron Source from the Oak-Ridge National Laboratories), only the blanket is filled with dissolved TRU materials, while the core contains solid nuclear fuel. In the case of Th-MSR (Molten Salt Reactor), the fluorides would have to be extracted into the organic phase with ligands or converted into nitrates for the separations and, after the separation, converted back to fluorides. Some separations (for example ion exchange-type separations) could be carried out directly on fluorides or nitrates without extractions into organic phases containing ligands.

Optionally, the excess of the processing capacity of the Reactor's Separation Unit can be used to provide the required pre-processing of the bumed-up nuclear fuel or to prepare fresh, enriched, uranium-based nuclear fuel from bumed-up nuclear fuel or even from natural uranium. The separation unit of the present invention can also be used to produce other high purity isotopes.

The Separation Unit (R-SEP) may be based on chromatographic systems, such as those described by Elchuk (Elchuk, 1991), Tanaka (Tanaka, 1986) and Kumar (Kumar,

2013). The chromatographic columns can provide up to 500,000 separation steps per column, and are known to be able to separate isotopes. The separation of isotopes is known to occur in liquid chromatography (Fukuda, 2010; Tanaka, 1986), ion-chromatography and gas chromatography (Wilzbach, 1957; Mohnke, 1989; Matucha, 1991 and Hufnagel, 1983). The isotope separations can be enhanced with the use of crown ethers or other sterically-sensitive ligands, liquid crystals, liquid crystal ligands (or their derivatives or combinations) and photo-activation at room temperature (Khudyakov, 1993). Additional enhancement can be achieved by the application of a magnetic field , such as 1.5 Tesla combined with liquid crystals and photo-activation (the Tuno Effect), or other isotope-specific activations. For the Turro Effect, a reversible photo-reaction will be particularly suitable for the separation, as such a reaction will enable use of multiple separation steps on a single separation column. Reversible photo-reactions are known in photochemistry, for example compounds from the class of fulgides and spiropyranes undergo such reactions. A review of isotope separations by chromatography was published by Hufhagel (Hufhagel, 1984).

Supercritical carbon dioxide chromatography will be the most prefened

chromatography system for this invention. This is because supercritical carbon dioxide chromatography is a technique demonstrating an exceptionally high processing capacity (due to low viscosity of the supercritical C0 ₂ ), very high separation power and can use any column packing used in liquid chromatography. Large scale supercritical carbon dioxide chromatography installations are used commercially (Subramanian, 1995). For the purposes of this invention, the supercritical separation system can use organic ligands, such as crown ethers, fluorinated beta-diketones (US Pat.5606724A, W09533541 Al), liquid crystals (or their combinations with the ligand groups), magnetic field and photo-excitation, as well as the methods of isotope-specific excitation (NMR, Nuclear Acoustic Resonance and other methods) described by Rygas in Canadian Pat. Appl. 2,908,999, incorporated herein by reference. For the separation of Uranium isotopes, particularly preferred will be the method based on using gel permeation column chromatography, combined with using Uranium complexed with liquid crystal ligands and enhanced with the NMR or Nuclear Acoustic Resonance isotope-specific excitation, as described in Canadian Pat. Appl. 2,908,999.

Suitable gel permeation chromatography techniques are described by Deyl (Deyl, 1984).

To use the organic ligands, the nitrates or sulfates will be removed from the reactor and the metals will be extracted from them into an organic phase containing the ligands. After the separations, a change in the pH will allow the re-extraction of the separated isotopes, back into the water phase.

The supercritical chromatographic separations are particularly suitable for the separation of the elements and even for the separation of different isotopes of the same elements. For example, 20% increase in the concentration of U-235 (the“column separation factor” was 1.20) was observed using supercritical carbon dioxide and complexed Uranium, on a short laboratory rectification column, with approximately 20 physical plates (Boris et al., 2017, RU2606973).

This high separation factor was observed when there was a temperature gradient on the column, with the bottom held at 40 °C while the top was maintained at 35 °C. Assuming the linearity of this effect at higher degrees of isotopic separations, the estimated number of recirculations of the isotopic mixture would be only about twenty six, to achieve 80 wt.% isotopic purity of a desired isotope from the starting material containing only 0.7 wt.% of the isotope.

For a chromatography system using supercritical carbon dioxide, it is technically feasible to get a resolution equivalent to 500,000 theoretical plates. Therefore, the supercritical column chromatography system with a temperature gradient will easily separate isotopes of many elements.

In the operation of the nuclear reactor of this invention (RM-HWR), the

concentrations of the TRU isotopes will be maintained at high levels, far away from the equilibrium concentrations. Typical loads of the TRU isotopes will be about 15 to 35 wt.% of the total metals. However, if the RM-HWR reactor would be permitted to achieve an equilibrium, then the concentrations of TRU metals (wt. % of total metals) would be expected to be close to the steady-state concentrations in the Marshalkin's Thorium Reactor

(Marshalkin, 2016, RU2634476C1):

- 237-Neptunium: 0.2%,

Plutonium: 0.5 % (includes 0.33% of Pu-238),

Americium: about 0.1%,

Curium: about 0.1%.

The Reactor's Separation Unit (R-SEP) may be remotely operated and fully automated. In one embodiment, a Pre-Processing Unit can be based on the commercial separation systems, such as UREX, TRUEX, etc., but modified by elimination of processes where solids are processed or produced.

In the case of a separations based on supercritical carbon dioxide (a preferred separation process), the one mode of operation will be to have the system sealed with only the inlet and the final fraction collection connected with the rest of the equipment. The high- pressure in the carbon dioxide system can be achieved by cooling the recycled C0 ₂ , collecting the liquid CO2 in high pressure vessels or long pipes and then heating those vessels or pipes to make the supercritical carbon dioxide for the separation process.

Examples of the embodiments of the Reactor's Separation Unit are:

1. An adaptation of a commercial separation system by combining elements of the processes such as TRUEX, TALSPEAK, DIAMEX, PUREX and SANEX. A schematic of such an adaptation is illustrated in Fig. 5. These processes are described in the publication "Reprocessing and Recycling of Spend Nuclear Fuel" (Taylor, 2015), and they can be utilized in this invention. These processes are based on liquid-liquid extraction processes, using tributyl phosphate, beta-diketones or similar complexing agents. The most preferred equipment for the extractions will be pulsing towers.

2. Separation systems based on liquid or supercritical chromatographic column arrangements, with the packings being standard, reversed phase or other suitable material, with an optional recirculation within each of the chromatographic column systems. A general schematic for these chromatography systems is described in Fig. 6. A production scale chromatographic system is a prefened method for implementation of the Separation Unit of the present invention. All chromatographic systems can use the Simulated Moving Bed (SMB) technique or, optionally, the continuous annular chromatography (CAC) system, known in the art of chromatography. The SMB is a preferred configuration of the columns, optionally with a gradient of the eluent and/or temperature gradient on some columns.

The packing of the columns may include borosilicate glass beads or glass beads with encapsulated gallium or a similar neutron-capturing materials to reduce the damage due to the neutron exposure; if needed, lead-glass beads could be used to absorb the gamma radiation. Some columns can be packed with microporous, size exclusion chromatography packings. These columns will be suitable for isotope separation, for example enrichment of uranium in U-235. Supercritical carbon dioxide as an eluent, as illustrated in Fig. 4, will be particularly suitable for this process. The separation of isotopes can also be achieved using isotope specific activation (NMR) and the Bakhmutov effect, as described in Canadian Pat.

Application 2,908,999, incorporated herein by reference.

The chromatography separation system will enhance the separation factors, making it possible to achieve separation of not only the elements, but also the individual isotopes of the actinides. In the case of supercritical carbon dioxide, the separation can be as exceptionally high as being equivalent to 500,000 single separations per 1 meter of column length.

Therefore, the "Reactor's Separation Unit" of this invention will be capable to produce enriched Uranium isotopes, such as U-235, at the level required for the nuclear fuel and even at the higher level of enrichment. Therefore, the Reactor's Separation Unit of this invention will also be suitable for a highly efficient enrichment of isotopes from natural uranium or for the isotope separation of other elements. The interdependence between the single-step separation factor, the required number of separation steps and purity, are illustrated in Table 4. The following chromatography systems are suitable for the Reactor's Separation Unit:

2. A. A preferred chromatography system is the system based on supercritical carbon dioxide. For this process, the streams SI, S2 and S3 will be first extracted with a complexing agent in an organic solvent. Suitable extracting agents include tributyl phosphate and derivatives, acetyloacetone, trifluoroacetyloacetone, dibenzoylmethane and similar complexing agents. Examples of the complexing agents are listed in Canadian Patent Application No. 2,908,999 (Tables 2 and 3: 5-hexadecyloxytropolone, b-diketones - tropolone ligands, b-diketones - dithiol ligands, complexes of phthalocyanines, monodentate nitrogen-containing ligands, salicylideneamine metal complexes (Schiff bases),

dithiocarboxylate complexes ferrocene and other organometallic compounds, various classes of compounds forming metallomesogens, supermolecular liquid crystals, 2-(2- hydroxyphenyl)oxazoline-derived compounds, bis[l-(2,3-didecyloxyphenyl)-3-(2,3,4- tridecyloxyphenyl) - 1,3 -propane-dionate] complexes, substituted b-diketonates, polycatenar elongated ligands derived from imino-pyridine or imino-bipyridine, 2’-fluoro-4’-((S)-(+)-7- methyl nonyl)-phenyl)ethene-l,2-dithiolate, an azine with a chiral carboxylato intermetallic bridge, chiral saliciladiminates, dinuclear complexes derived frorm Schiff bases with a halogen intermetallic bridge, ferrofluids based on liquid crystals, nanometer-sized ferromagnetic platelets suspended in a nematic liquid crystal, 3,3’-bis- {[4- (2diphenylphosphanyl-ethyltio-phenylimino]-methyl}[l,rbinaph thylenyl-2,2’-diol). The organic solvent can be, Petroleum Ether (with a boiling point of about 120 -150 °C), hydrogenated kerosene, or Magiesol 40 ^® solvent. The column packing can be silica gel or silanized silica gel.

The extracted complexes of the nuclear isotopes will be dissolved in supercritical carbon dioxide and fractionated using a chromatography system illustrated in Fig. 4. From the fractions, carbon dioxide will be separated by decompression and the nitrates will be recovered through back-extraction, after adjustment of the pH value for the process of the back-extraction. The diluted water fractions from the back-extractions can be concentrated in a separate evaporator. Alternatively, the water-diluted nitrates can be pumped directly into the reactor and the excess of water can be removed from the top of the distillation tower, because the distillation tower is connected to the top of the reactor. The complexing agent, carbon dioxide and water, will be recycled back to the separation process.

2.B. A reversed phase chromatography system based on silanized silica, with the remotely controlled valves and an optional circular flow of the components to be separated, as in Fig. 4.

2C. A system based on ion chromatography (a valved system). There are numerous ion- exchange resins, such as Dowex™ 1 or Dowex™ 50, suitable for the required separations. 3. Standard partition chromatography system (valved). For high resolution, this system would require a reversible conversion of the nitrate salts into separable organic complexes, as described in the section related to the supercritical carbon dioxide separation. After separations, the organic complexes are re-extracted, to provide water solutions of nitrates.

The separated fractions, depending on their composition, will be sent to the quarantine, reactor, or the depository, as appropriate.

The method of destruction of nuclear waste and the separation of isotopes described in this disclosure, describes merely exemplary embodiment of the present invention. One skilled in the art will readily recognize from such discussions and from accompanying drawings and claims that various changes, modifications and variations can be made therein, without departing from the spirit and scope of the invention, as described in the claims that are listed below.

PATENT LITERATURE

Marshalkin, V.E., Povyshev,V.M. (2016). RU2634476C1. Method of operating nuclear reactor in thorium fuel cycle with production of uranium isotope U-233.

Turro, N.J., Tenafly, N.J. (1982). US Pat. 4,351,707. Methods employing magnetic isotope effect.

Boris, B.V., Ivanova, S.F., Kazarinov, Y.G., Neklyudov, I.M., (2017). RU 2,606,937 C2. Method for supercritical fluid extraction of uranium complexes.

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Xu, H.Y. st al. (1986) Determination of the 235U/238U single stage isotope separation factors (alpha) a for Uranium (VI) extraction by macrocyclic polyether. Journal of Nuclear and Radiochemistry. 1986-03. Table 1. List of abbreviations

Table 2. Definitions of terms and definitions of effects used in this document

Table 3. Typical separation factors for isotopes.

Table 4. Estimated number of separation steps required to achieve the isotopic purity of 99 wt.%. The estimation is based on the assumption that the separation factor remains unchanged, even at the high levels of enrichment*.

* Linearity of the single-stage separation factor, for the U235/238 system in centrifuge separations was maintained up to about 60% of isotopic purity (Choppin, 2016).

** Literature values (Choppin, 2016, p. 37). Table 5. Listing of Equipment

Table 6. Material Streams

Table 7. List of drawings.