Patent Application
20060233685
This invention pertains generally to reducing risk in the disposition of spent nuclear reactor fuel, and more particularly to the reduction of spent nuclear fuel into chemical constituents, achieving a physical separation of the radioactive elements from the inert, often valuable, non-radioactive or low level radioactive, transportable majority of the material. More particularly, this invention relates to unique separation methods and apparatuses which make use of the greater vapor pressures at elevated temperature of the chemical species generated from the many neutron interactions than that of the remaining unreacted spent uranium dioxide.
From successful operation of the world's first nuclear reactor in December 1942 it was realized that an efficient separation process was urgently needed. Of primary importance at that time was collection of plutonium, produced by the neutron interaction with natural uranium U238. The new method developed at that time for plutonium recovery was based on wet chemistry and dissolution of the spent fuel in nitric acid, together with an organic phase solvent-extraction method. The “PUREX” (Plutonium/URanium EXtraction) is still used exclusively in all the world's spent nuclear fuel reprocessing plants. The PUREX process requires a large central chemical facility equipped with expensive hot cells, extensive liquid storage and requires transport of the spent fuel from the security of the individual nuclear power plants. A major drawback of the PUREX process is the production of large volumes of low level liquid radioactive wastes. This invention achieves the chemical separations required and does so with no liquid or chemical additions.
Nuclear power plants are in use worldwide and currently produce ˜16% of the world's supply of electricity; 20% of U.S. electricity. Nuclear power plants function by the violent process of neutron induced fission, predominantly of the U235 isotope of uranium. Approximately one-third of the heat produced from the fission process in the reactor is converted from heat to electricity via a conventional steam cycle. Natural uranium contains only 0.711% of the fissionable isotope U235. Certain types of reactors can operate for a limited time fueled with natural uranium. The uranium fuel employed in the most common reactor designs has been enriched in the fissionable isotope. The natural uranium is enriched to be 3% to 5% U235 in a gaseous diffusion process. The use of enriched fuel enables the nuclear power plant to operate for an extended period, i.e. until the U235 content is spent and has fallen to an inoperable level of about 0.92%. The enriched fuel allows the typical nuclear power plant to release approximately 33,000 thermal megawatt-days of heat energy from a single metric ton of enriched uranium. A nominal 1,000 megawatt-electric nuclear power plant is about 33.3% thermally efficient in converting heat to work. The nominal plant is therefore seen to operate at 3,000 thermal megawatt heat release rate and will generate one metric ton of spent uranium fuel every 33,000/3,000 or eleven days of operation. This calculates to be 8.35 pounds of spent nuclear fuel generated each hour of plant operation. A nuclear plant typically refuels every eighteen months. If for example the reprocessing of the spent fuel is carried out during the month dedicated to refueling the reactor, the design spent fuel reprocessing rate would be about 150 pounds per hour. It will be shown that this invention will readily allow operation at this flow rate to allow location of the separation unit on the premises of the power plant.
Approximately fifty tons of spent uranium fuel is removed from the reactor during reprocessing and the reactor refueled with fifty tons of fresh enriched uranium fuel. The fifty tons of highly radioactive spent fuel is stored underwater adjacent to the power plant facility and is allowed to cool as the radiation level decays with time and awaits transport without further processing to a permanent storage repository. This refueling method is employed at every one of the 104 nuclear power plants now operating for four decades in the U.S. This design concept is termed “the once-through fuel cycle,” as no attempt is made to recover the chemical constituents contained in the spent fuel.
One such recoverable constituent is plutonium. Approximately one percent of the spent fuel is plutonium, a fissionable element equivalent in fission energy to U235. One gram of plutonium has the fission energy equivalent to 14.1 barrels of oil and could rightly be so valued if recovered and blended into fresh nuclear fuel and done so within cost. The slight enrichment left in the spent uranium would also reduce natural uranium feed required to the enrichment plant.
Eleven of the twelve “nuclear nations” are reprocessing the spent nuclear fuel and are recycling the recovered plutonium and uranium. Only the U.S. does not reprocess the spent fuel and is currently committed to the “once through fuel cycle” strategy. This decision is understood considering the prevailing economics unique to the U.S., uncertain regulatory cost associated with transporting spent fuel across country, reprocessing plant siting and uncertain long term storage costs of the large volumes of low level nuclear wastes that is generated. Additionally, the current low cost of gaseous diffusion operation needed and to buy and enrich the additional natural uranium outweighs the uncertain net cost benefit of the plutonium recovered. The additional apprehension factor against reprocessing unique to the U.S. is summed up in a recent paper by a prominent government scientist saying “It is difficult to imagine that any form of chemical reprocessing would be more proliferation-resistant than leaving the plutonium mixed with the highly radioactive fission products in the solid fuel matrix.” This invention describes the “difficult to imagine process” that offers even more proliferation-resistance.
Each of the owners/operators of the 104 multi-billion dollar nuclear power plants would store the spent fuel in the adjacent cooling pools until such time as a permanent repository can be identified. Final identification of an acceptable repository has not occurred and the country's spent fuel cooling pools are rapidly filling up. It is agreed by many analysts that regardless of the single choice of the disposal site obtaining nationwide public acceptance will be difficult for the transportation of the spent fuel across the U.S. road and rail network. This perception arises from considering normal transportation accident rates together with the potential for terrorist hijacking and the resulting liability.
After a seventeen year nine billion dollar U.S. Department of Energy (DOE) effort spent on establishing the permanent depository at Yucca Mountain, Nevada, law suits filed in federal courts on behalf of the state of Nevada may stop the DOE effort completely. The cost of the Yucca Mountain project in year 2000 prices is estimated to be $58 billion dollars. The storage capacity will equal the year 2010 total U.S. production of spent fuel, 70,000 tons, such that no room exists for additional fuel generated thereafter, at this expense level. This invention offers an alternative to the Yucca Mountain project and the first practical long-term solution to disposition of spent nuclear reactor fuel.
This invention makes use of the well known fact that approximately 95.5% of the spent fuel is not involved in the fission process and does not require long term storage. The amount of material to be stored according to the prior art is thus correspondingly higher than it would be if the 4.5% involved in the fission process were the only separated and stored material. The spent fuel consists of approximately 95.5% spent depleted uranium in the form of uranium dioxide, and 4.5% other chemical species resulting from the fission process and neutron capture. these other chemical species are believed to be in part in the oxide form. Many of the oxides decompose at elevated temperature. Plutonium dioxide, which is ˜1% of the spent fuel, is the major object of nuclear proliferation concern. The most secure point for the handling of spent fuel is inside the nuclear power plant. Processing during the one month refueling time produces ˜1.6 cubic feet of plutonium dioxide, which can be secured within the facility indefinitely in a manner that makes it secure over the lifetime of the power plant.
Therefore, there is a demonstrated need worldwide for a more economic and effective method for processing the spent nuclear fuel produced in the operation of nuclear reactors. This new method proposed herein would significantly reduce the cost of the world's electricity and effectively end the possibility of the proliferation of plutonium. The present invention details a unique approach to advancing the state of the art and to reduce the associated risk.
This invention makes possible more economic and secure processing of spent nuclear reactor fuel within the security of the nuclear power plant site. This is made possible because neither liquid solvents nor wet chemistry or storage for large volumes of radioactive liquids are used nor is any waste generated, and the system can operate with no moving parts. This invention processes the spent fuel by solvent extraction using the mother of all solvents, heat. The low vapor pressure at elevated temperature of the uranium dioxide (95.5% of the spent fuel) relative to the remaining (4.5%) neutron-induced chemical species in the spent fuel mix allows the physical separation to happen. How best to effect this is the subject of this invention.
There are some 50 chemical species generated by the neutron interactions and each exerts its own vapor pressure response to temperature. The melting point of uranium dioxide (UO2) is 2,847° C. the boiling point is 4131° C. One component of the spent fuel is Strontium which melts at 769° C. and boils at 1384° C. If spent fuel were melted the Strontium would, given enough time, evolve and escape (boil off) the liquid surface as a gas. Further, plutonium dioxide (1% of the fuel) boils at 2800° C. and so would also leave the melted spent fuel as a gas.
Twenty percent of the fission products make up the volatile fraction at room temperature. The volatility of nearly all the chemical species in the spent fuel is much greater than the UO2 at all temperatures. The chemical species will diffuse through and evolve, driven by their individual vapor pressure, from the solid surface of the UO2. If the UO2 is in a molten liquid state the transport to the surface of volatile chemical species is at a greater rate by many orders of magnitude.
The principle this invention is based on is perhaps better understood by analogy. The analogy is to carbon dioxide (CO2) gas dissolved in cold liquid water under pressure. When the pressure is released the higher vapor pressure CO2 (M.P. 9° C., B.P. −36° C.) begins to evolve and eventually the water goes “flat.” In a like manner, the 4.5% of non-UO2 chemical species are dissolved and dispersed in the solid spent UO2. Minimum drop size and maximum temperature achievable can both increase the rate at which the dissolved gas is evolved from the liquid phase. If for instance, again by way of the example, the same CO2 saturated in cold water is heated to 100° C. and the water “atomized” in a spray, the rate of CO2 gas evolution would then, intuitively, be more or less instantaneous. With this understanding of vapor pressure effect, this invention goes on to answer two questions needed to achieve unique processing of spent nuclear fuel: 1) how, in a practical manner, to heat the solid spent fuel to the liquid state; 2) how to produce the smallest practical liquid drop size.
As to the first question, a plasma spray process is preferred, according to this invention. Plasma electric arc technology is known in the art to vaporize and deposit a refractory material heated to vaporization on specific surfaces such as gas turbine blade coatings. This technique is so potent that any material, including tungsten (M. P. 3380° C., B.P. 4889° C.) can be turned to a molten/vapor stream and the sprayed material deposited as a coating on a selected surface. A chemically inert flow of helium and or argon gas is used as the electric conducting plasma gas and carrier of the vaporized or melted material. The ultra effective heating device is referred to as a plasma torch and is in commercial operation worldwide. The material to be melted is fed into the carrier gas and into the plasma arc as a powder typically in the range of 100-micron particle size. This invention employs a plasma torch to heat the carrier gas flow and powdered spent nuclear fuel to the melting point and above, to approximately 3000° C. This is a practical limit considering the refractory materials available for containment. This is the temperature range of the tungsten filament temperature in a light bulb.
To answer the second question, a minimum drop size of the melted spent fuel is achieved by preferably submitting the powdered UO2 to an additional ultra fine grind to the talcum powder micron range. A one micron particle of UO2 is approximately 3000 molecules in diameter. 4.5% or 135 of the molecules across the diameter of the spent fuel are the chemical species with very high vapor pressure at 3000° C. In this case, upon melting of spent nuclear fuel, a “popcorning” effect will be seen. A typical carrier gas velocity into the torch (10 feet/second and thickness of the plasma arc of say 0.1 feet) means the residence or heating time is in the order of 10 milliseconds—the liquid drop will be shattered. In this short time span the solid particle at ambient temperature will have been heated to 3000° C. and the collective vapor pressure of the chemical species will have generated a gas volume many times the size of the liquid drop. This gas production is at very rapid rates. This explosive effect enhances the transport of the volatile chemical species from the liquid to the helium/argon carrier gas. In which case, the separation of the chemical species (4.5%) from the spent uranium (95.5%) has been achieved.
In addition to the gasified chemical species the 3000° C. plasma gas flow now contains a mist of the ultra small liquid drops of the pure spent UO2. The mist is preferably coalesced by use of a cyclone separator. Cyclone separators work on the difference between the density (grams/cubic centimeter) of the liquid drop and density of the carrier gas. In this case there is an extreme difference. The UO2 is 10.96 g/cm3 and the carrier gas is (0.00018 g/cm3) at ambient temperature and 1/10 this value (0.000018 g/cm3) at 3000° C. The cyclone separator, with no moving parts, achieves the required demisting. This means the plasma gas phase will follow and turn with the flow volume allotted to it in a cyclonic manner and the ultra dense liquid drop will continue forward largely ignoring the centrifugal forces and collide directly with the wall of the vessel, will coalesce and in a circular manner flow down the walls under the influence of gravity.
The cyclonic flow imparted to the plasma gas phase will ensure that no liquid drops remain in the gas phase as the plasma gas, free of the spent UO2, exits the separator in the vertical central pipe provided. The liquid, substaintially 100% spent UO2, free of the chemical species, swirls down the walls to the drain in the bottom of the vessel.
Upon cooling, the solidified UO2 (95.5% of the discharge from the system) would present no increased cross country transportation risk. This spent uranium would still be enriched in U235 to 0.92%. This enrichment would offer a more economic feed to the gaseous diffusion enrichment plant than natural uranium 0.711% and would reduce somewhat the natural uranium requirement. The depleted uranium tailings produced from this spent uranium would join the current storage of the previous tailings.
Alternative Methods to Cause Separation of the Chemical Species.
As an alternative embodiment, other systems could be utilized to provide the heat for separation according to this invention. In the broadest sense, evolving the chemical species from spent nuclear fuel requires only that the temperature of the spent fuel be raised above about 2,847° C. A readily understood description of a demonstration of the basic concept is a puddle of liquid spent nuclear fuel contained on a tungsten saucer that is heated by electrical resistance to this melting point temperature or slightly above. A small hole in the bottom of the dish gravity drains the dish at the same rate UO2, preferably as a talc sized powder, is being added to the surface of the liquid. The more volatile chemical species are distilled from the surface of the liquid. A low temperature condenser creates a vacuum and pulls out the evolved vapor. The degasification could be on a continuous basis or in a batch mode.
The heat source could be electric resistance heating filaments inside the enclosure and adjacent the dish. The UO2 filled dish would simmer until the chemical species had been boiled off and the purified UO2 is then released to storage. In this degassing process no attempt is envisioned to segregate the chemical species. The entire quantity of condensed chemical species produced in the 18 months of operation of the 1000 MWe nuclear power plant and 30 days of processing would fit in a cube 2 feet on a side.
Disposition of the Chemical Species in the Helium/Argon Effluent.
This invention permits separation of the gaseous chemical species after they leave the enclosure. In the process of the preferred embodiment, using a plasma torch and a cyclone separator, the helium/argon plasma gas effluent contain all the chemical species in the vapor state at 3000° C. Since the xenon and krypton fission products condense several hundred degrees below 0° C. the plasma gas must be cooled to this temperature to liquefy and recover these radioactive gases. The temperature of the plasma gas will be reduced in stages. The dew points of the individual species will dictate when condensation occurs at each stage. Cyclonic flow will preferably be induced at each stage of cooling, thus the condensed phase will be separated and the liquid formed can be drained from each stage vessel. It is understood that a perfect separation of the individual chemical species will not be achieved. The helium or other plasma gas from the final stage is routed back to the plasma torch for reuse.
This partial separation technique is repeated until the cryogenic stages are reached. The stage the plutonium dioxide condenses will be given most careful design attention. This is because the plutonium fraction (about 22% of the non-UO2 chemical constituents) is the largest component of the chemical species. Also, and most important, the manner with which the plutonium is physically managed and custody secured will establish the degree of risk of undesirable nuclear material proliferation.
Material Considerations.
The melting point of Uranium Dioxide (UO2) is about 2,847° C. The 4.5% contaminants in the spent UO2 fuel lowers the melting point some 170° C. This invention is predicated on processing the spent fuel at or above these very high temperatures. Few refractory materials are available that can contain the process at these temperatures. One such material is considered to be tungsten. The tungsten filament standard light bulb operates on the order of 3000° C. The physical properties, metallurgy and fabrication techniques of tungsten are widely and thoroughly understood. Tungsten pipes are routinely fabricated, usually with plasma arc deposition on a dissolvable surface. While the process temperature targeted is some 380° C. below the melting point, the metal's residual tensile strength is near zero at these high temperatures.
To add strength to the tungsten enclosure, two concentric tungsten pipes with an insulating layer between are provided to contain the plasma gas at 3000° C. The insulating layer is preferably thoria, (ThO2, M.P. 3390° C.). The inner pipe carries near zero stress the thoria reduces the temperature some 200° C. for the outside pipe of tungsten, allowing the outer pipe to provide all the support. The pipe is boxed on four sides with flat pieces of pyrolytic graphite whose phenomenally low thermal conductivity will insulate the pipe to the degree required by the load carrying outer tungsten pipe. This suggested conceptual design allows the plasma gas to be conveyed within the pipe for an extended length beyond the plasma torch allowing the spent fuel drop suspended in the plasma gas to remain in the liquid state with sufficient time to degas.
The alternate method of effecting separation of the more volatile chemical species operates at the same temperatures. The simmering spent fuel in the electrically heated tungsten boat requires the same multi layer design considerations. The insulated dome above the simmering liquid UO2 will not allow condensation on the inner surface of the dome by the boiled off effluent. The evolved gaseous chemical species will therefore exit the dome and enter the condenser. Pyrolytic graphite can be fabricated in a hemispherical shape and may, in addition to maintaining the temperature level, be able to also carry the load of the enclosure and related structure.
Accordingly, a primary object of the present invention is to separate chemical constituents within spent nuclear reactor fuel from each other so that the separate chemical constituents can be most effectively separately handled.
Another object of the present invention is to provide a spent nuclear fuel chemical constituent separator which does not create additional radioactive waste.
Another object of the present invention is to provide a spent nuclear fuel chemical constituent separator which does not utilize liquid solvents.
Another object of the present invention is to provide a spent nuclear fuel chemical constituent separator which can achieve substantially complete separation of individual chemical constituents.
Another object of the present invention is to provide a method for separating spent nuclear fuel chemical constituents which is simple to operate, has a minimum of moving parts, and which can be effectively operated and controlled remotely with secure containment of the spent nuclear fuel during separation.
Another object of the present invention is to provide a separator for separating UO2 from other chemical constituents within spent nuclear reactor fuel.
Another object of the present invention is to provide a cost effective alternative to long term storage of mixed spent nuclear reactor fuel.
Another object of the present invention is to provide a method for recycling UO2 from spent nuclear reactor fuel for enrichment and reuse.
Another object of the present invention is to provide a method for extracting plutonium from spent nuclear reactor fuel for secure custody thereof by authorized personnel with low proliferation risk.
Other further objects of the present invention will become apparent from a careful reading of the included drawing figures, the claims and detailed description of the invention.
Referring to the drawings, wherein like reference numerals represent like parts throughout the various drawing figures, a non-aqueous method for separating chemical constituents in spent nuclear reactor fuel is described according to a preferred embodiment and various alternative embodiments. The spent nuclear reactor fuel F (
In essence, and with particular reference to
Most preferably, these gaseous chemical constituents of the fuel F are further separated by one or more additional separators in the form of condensation vessels. Each one of the separators 50, 60, 70, 80, 90 (
More specifically, and with particular reference to
To separate these chemical constituents according to this invention, the mixture of chemical constituents making up the spent nuclear fuel F must be heated to at least a melting temperature for UO2. While any heating method capable of heating this mixture to at least the melting point of UO2 (2,847° C.+/−30° C.) could be utilized, most preferably a plasma torch 30 is utilized to achieve such heating.
Before the plasma torch 30 can be effectively utilized, it is desirable to grind the fuel F in the form of a rod into fuel F in the form of fine particles. This particalization preferably grinds the fuel F into particles of one micron diameter or smaller. Such particalization facilitates passage of the solid particles through the plasma torch 30 where they are heated and boiled into a gas, or at least a liquid in the case of UO2. If other forms of heating are utilized, such particalization can optionally be eliminated.
After particalization of the fuel F, the fuel F is preferably delivered into a hopper 20 for storage before separation according to this invention. The hopper 20 preferably includes an inlet 22 for receiving the particalized fuel F and an outlet 24 for delivering the particalized fuel F out into a supply tube 32 feeding the plasma torch 30.
The supply tube 32 preferably has a flow of plasma gas, such as helium or argon delivered thereinto (along arrow E of
This supply tube 32 leads (along arrow A of
Details of the influence this heating has on the spent fuel are depicted in the illustration provided in
A discharge 36 downstream from the electrodes 34 of the plasma torch 30 extends into a UO2 separator enclosure 40. The UO2 separator enclosure 40 can have various different configurations, with a most preferred configuration shown in
The UO2 separator enclosure 40 includes a side wall 42 which is preferably generally cylindrical oriented about a vertical central axis. A lower portion of the enclosure 40 is preferably tapered along a taper 43 with a conical form leading down to a drain port 44 at a bottom of the enclosure 40. A UO2 collector 46 is preferably located below the drain port 44 where liquid UO2 is collected.
Preferably, the UO2 is allowed to solidify within the UO2 collector 46 by lowering the temperature thereof to below the UO2 melting temperature. The UO2 collector 46 can be configured to cast the liquid UO2 into desired shapes for convenience in handling once the UO2 solidifies, or otherwise be converted into an optimum form for handling in a solid phase.
The UO2 separator enclosure 40 also includes a gas outlet preferably in the form of a stand pipe 48 aligned with a centerline of the enclosure 40. An entrance 49 at a bottom of the stand pipe 48 receives gaseous non-UO2 chemical constituents of the fuel F for removal from the UO2 separator enclosure 40.
While the UO2 separator enclosure 40 could operate merely on the principal of gravity separation, preferably separation is further enhanced by generating cyclonic flow within the UO2 separator enclosure 40. Specifically, the discharge 36 of the plasma torch 40 preferably enters the UO2 separator 40 at an inlet which is offset laterally from the centerline of the UO2 separator enclosure 40 and generally horizontally. Hence, the gas emitted from the discharge 36 and inlet into the UO2 separator enclosure 40 is caused to swirl circumferentially within the generally cylindrical UO2 separator enclosure 40.
Because the gas outlet from the UO2 separator 40 is in the form of the entrance 49 at the bottom of the stand pipe 48, this circumferential flow tends downward so that the general path followed by the gaseous chemical constituents of the fuel F follows a helical path (along arrow B of
Preferably, the side wall 42 of the enclosure 40 is maintained at approximately 3,000° C. with this temperature being sufficiently low to cause the UO2 to condense into a liquid phase. If the UO2 never boils into a gas, it exists as a fine droplet due to the particalization of the fuel F before passing through the plasma torch 30. These fine droplets or condensing gases of UO2 will tend to collect on the walls of the enclosure 40, as heat transfer will tend to occur out through the walls 42 to a surrounding environment, causing the side wall 42 to be the coldest portion of the UO2 separator enclosure 40. This condensing UO2 liquid will travel by gravity down the side wall 42 to the taper 43 and down to the drain port 44 and into the UO2 collector 46 (along arrow C of
Preferably, the liquid UO2 travels sufficiently slowly that any non-UO2 gases entrained within the liquid UO2 will have a sufficient opportunity to boil out of the liquid UO2. Beneficially, the vapor pressures of UO2 and the vapor pressures of other chemical constituents of the fuel F are such that substantially all of the non-UO2 chemical constituents will boil out of the UO2 liquid before the UO2 liquid passes through the UO2 drain port 44 and into the UO2 collector 46. Preferably, the drain port 44 is sufficiently small and the flow of UO2 sufficiently high that the drain port 44 is filled with liquid and prevents gaseous chemical constituents of the fuel F from passing into the UO2 collector 46 through the drain port 44.
Another advantage of the orientation of the stand pipe 48 for the gaseous outlet from the UO2 separator 40, in addition to generating cyclonic flow for enhanced separation, is its tendency to avoid gas stratification within the UO2 separator enclosure 40. In particular, one gaseous chemical constituent within the UO2 separator enclosure 40 is plutonium oxide (PuO2). PuO2 and other transuranic chemical constituents of the fuel F are relatively heavy gases at this temperature and so would tend to settle toward a lower portion of the UO2 separator 40.
If the gaseous outlet from the UO2 separator 40 were at a top of the UO2 separator enclosure 40, a tendency might exist for PuO2 or other transuranic gaseous chemical constituents to stagnate near a lower portion of the UO2 separator enclosure 40. By spacing the inlet into the UO2 separator enclosure 40 in the form of the discharge 36 near an upper portion of the UO2 separator enclosure 40, and with the gaseous outlet in the form of the entrance 49 into the stand pipe 48 spaced vertically from each other and with somewhat turbulent cyclonic flow therebetween, the gases within the UO2 separator enclosure 40 remain sufficiently stirred that such stratification and stagnation of gases within the UO2 separator enclosure 40 is resisted. Also, the stand pipe 48 discourages liquid UO2 escape therethrough should the level of UO2 liquid rise to an unacceptably high level.
With particular reference to
Outside of this second tungsten layer, preferably a gas region 106 is placed to provide further insulation for the enclosure 40. This gas region 106 is preferably filled with helium or other non-reactive gases to prevent degradation of the tungsten or other metals forming the separator, and also to exhibit little performance degradation in the high radiation environment involved.
Preferably, this gas space is contained adjacent the enclosure 40 by an outer enclosure 108 of pyrolytic graphite. Pyrolytic graphite has a low thermal conductivity and thermal radiation properties which further discourage heat transfer out of the separator enclosure 40. The containment structure 100 both acts to maintain the fuel F while undergoing separation, while also discouraging heat transfer beyond a desirable rate out of the UO2 separator enclosure 40, and while also providing some degree of radiation shielding for a surrounding environment. This containment structure 100 provides a preferred form of containment structure, with various different modifications being acceptable provided that desired rates of heat transfer are provided and appropriate levels of radiation shielding are provided.
With particular reference to
In particular, gas flow occurs along arrow D of
The gas pipe 56 is preferably oriented generally as a stand pipe aligned with a vertical centerline of the vessel wall 52 with an entrance 58 at a lowermost end of the gas pipe 56. Thus, cyclonic flow within the second separator 50 can beneficially enhance separation within the second separator 50.
Gas flow into the entrance 58 follows arrows D to a third separator 60 in the preferred embodiment of this invention. The number of separators provided can vary depending on the degree of separation desired. Most preferably, a third separator 60 has a similar form to the second separator 50. Hence, the third separator 60 includes a vessel wall 62, collection region 64, gas pipe 66 and entrance 68. One or more chemical constituents H having a lower condensation temperature than the chemical constituents G collect within the collection region 64 of the second separator 60.
A fourth separator 70 is preferably located downstream from the third separator 60. The fourth separator preferably similarly includes a vessel wall 72, collection region 74, gas pipe 76 and entrance 78. The fourth separator 70 thus causes one or more chemical constituents I to collect in the collection region. The chemical constituents I would be characterized by having a condensation temperature below that of the chemical constituent H.
Preferably, a fifth separator 80 is located downstream from the fourth separator 70. The fifth separator 80 is similar to the other separators 50, 60, 70 such that it includes a vessel wall 82, collection region 84, gas pipe 86 and entrance 88. The collection region 84 collects one or more chemical constituents J which are characterized by having a condensation temperature below the condensation temperature for chemical constituents I.
Finally, according to the preferred embodiment, a sixth separator 90 is preferably provided downstream from the fifth separator 80. The sixth separator 90 preferably similarly includes a vessel wall 92, collection region 94, gas pipe 96 and entrance 98. One or more chemical constituents K collect within the collection region 94 with the chemical constituents K characterized by having a condensation temperature lower than that of the chemical constituents J. Preferably, only the plasma gas remains in gaseous form within the sixth separator 90. This plasma gas exits through the gas pipe 96 (along arrow E of
With reference to
The boat 120 preferably includes a floor 122 surrounded by a rim 124 to contain the liquid UO2. The boat 120 can have any of a variety of particular shapes, with the shape depicted in
The entire enclosure is sufficiently hot that none of the other chemical species driven out of the liquid UO2 condense on the sidewall 112. The gaseous chemical species thus are contained within the enclosure 110 until they escape out of the enclosure 110 through the top vent 150.
The boat 120 and UO2 therein can be heated in many ways. Most preferably, resistance heating electrodes 130 are provided for such heating. These electrodes 130 can be in contact with the boat 120 for conduction heating, or be adjacent, but spaced from the boat 120, so that heating is primarily by heat radiation. The boat 120 is made of a material, such as tungsten, which can remain solid and with sufficient strength at the temperatures involved (2,847° C.+/−30° C.).
The spent fuel is preferably ground to a powder, such as to particle sizes of one micron or less, before being fed into the enclosure 110 through a spent fuel feed pipe 140 and into the boat 120. Alternatively, the spent fuel could be provided in larger particle sizes, or even in rod form, provided the boat 120 or other liquid UO2 holder can handle the larger sizes. This alternative heater could operate continuously, or in a batch fashion. The top vent 150 is preferably coupled to a series of chemical species separators 160. These separators 160 preferably are arranged similarly to the various separators provided downstream of the separator enclosure 40 of the preferred embodiment.
Once the various different chemical constituents of the spent nuclear reactor fuel F have been separated from each other, they can be separately handled in a manner which is optimal for each separate chemical constituent. For instance, the UO2 could beneficially be reused in nuclear reactors after reprocessing, such that less uranium would need to be mined to power nuclear reactors. The plutonium dioxide could be processed for use as a fuel in nuclear reactors, or beneficially transmuted into other isotopes, or safely stored in a secure area for non-proliferation.
Some of the remaining chemical constituents can be grouped by associated levels of radioactivity and half lives, such that those chemical constituents having relatively short half lives can be stored at progressively less sensitive disposal sites over time. Those chemical constituents having longer half lives can be placed in more long term storage, with only a small amount of such storage then being necessary because only the appropriately long half life chemical constituents require such storage. Also, many chemical constituents may have a valid market therefore, such as to power fluoroscopes and medical diagnostic equipment, for use in radiological medicine, for beneficial use in research, and in other applications where such chemical constituents are in demand.
It is contemplated that when a nuclear power plant is undergoing refueling, that space would be set aside at each nuclear power facility where the separation equipment for practicing the separation method of this invention would be utilized. This equipment could be permanently provided on site or be provided in a mobile fashion for use at various different nuclear power plants or other nuclear facilities. Most preferably; common teams of operators would utilize mobile equipment on site to separate the spent nuclear reactor fuel on site at the various spent nuclear reactor fuel generating facilities. Such handling experts would then deliver constituents having market value into the marketplace, distribute low value constituents into properly sequestered storage on site at the facility, and deliver chemical constituents such as the uranium dioxide and plutonium dioxide to facilities where reprocessing and/or secure storage can occur.
This disclosure is provided to reveal a preferred embodiment of the invention and a best mode for practicing the invention. Having thus described the invention in this way, it should be apparent that various different modifications can be made to the preferred embodiment without departing from the scope and spirit of this invention disclosure. When structures are identified as a means to perform a function, the identification is intended to include all structures which can perform the function specified. When structures of this invention are identified as being coupled together, such language should be interpreted broadly to include the structures being coupled directly together or coupled together through intervening structures. Such coupling could be permanent or temporary and either in a rigid fashion or in a fashion which allows pivoting, sliding or other relative motion while still providing some form of connection, unless specifically restricted.
