Patent Application
20250128960
The invention relates to a target for the manufacture of 99Mo (also referred to as Mo-99) and a method of manufacturing such a target, of particular but by no means exclusive application in maximizing efficiency and minimizing the production of unwanted by-products.
The radioisotope 99Mo is produced for its decay product, 99mTc, which is of value in certain nuclear medicine diagnostic procedures. An existing method of producing 99Mo involves the fission of 235U by neutron irradiation in a nuclear reactor. This method employs highly enriched uranium. (Natural uranium is approximately 0.71% 235U by mass, with a 235U to 238U [mass]ratio of approximately 0.0072; the term highly enriched uranium typically implies a 235U enrichment of greater than 20%.)
Enriched uranium targets of approximately 20% 235U enrichment are also employed for the manufacture of 99Mo via the fission method, but the maximizing of 99Mo output per unit time, in conjunction with the use of such targets, has led to increasing volumes of solid waste created from the dissolving of uranium targets. (Note that uranium with a 235U enrichment of approximately 20% may be described as low enriched uranium (LEU); the term “low enriched uranium” generally implies a 235U enrichment of greater than that of natural uranium but less than or equal to 20%.)
Reusable targets have been proposed but their realization has had a number of problems, including fission product build-up (which can lead to greater impurity levels), the incompatibility of targets with existing chemical extraction processes, the greater design and manufacture costs of reusable targets, and the presence of an extraction medium in the target (which could suffer degradation due to prolonged radiation damage, and give rise to complications when resealing and testing the target prior to re-irradiation).
Additionally, plutonium in the form of PuO2 is a by-product of the irradiation of the 238U, which reduces efficiency and leads to waste that creates both proliferation and disposal concerns.
It is an object of the present invention to provide a UO2 target for use in the manufacture of 99Mo, and a method of manufacturing such a target.
According to a first aspect of the invention, there is provided a UO2 target for use in the manufacture of 99Mo, the target comprising:
It should be appreciated that only 235U and 238U are considered in any detail in the present disclosure. Owing to the very small quantities of other isotopes (principally 234U) found in naturally occurring uranium, the presence of such isotopes is considered to fall within the precision as quoted herein of the parameters pertaining to the disclosed embodiments.
In a first particular embodiment, there is provided a UO2 target for use in the manufacture of 99Mo, the target comprising:
That is, the particles comprise UO2 and the UO2 comprises uranium with a 235U to 238U ratio of less than 3% 235U enrichment.
Thus, the target comprises UO2, as UO2 is impervious to the effects of the typical fluids used to extract the 99Mo (such as super critical CO2 or an alkaline chemical). For example, an alkaline solution can be passed through the matrix of UO2 (to remove the 99Mo), obviating the need to manage hydrogen gas. A porous matrix allows the produced 99Mo to be more readily released and extracted, such by flushing the matrix or pores thereof with a solution in which 99Mo is soluble.
Methods of extraction that may oxidize the target should be avoided, as conversion of a quantity of the UO2 into U3O8 will compromise the target. To prevent oxidization, the target is desirably housed in a sealable target container to isolate it from the surrounding environment; optionally, the container may be backfilled with helium gas. The latter reduces oxidization (cf. the backfilling with helium of nuclear fuel rods), facilitate conduction of heat from the target and reduce the distance that the ejected 99Mo travels.
A suitable sealable target container is advantageously thin-walled to maximize neutron transparency, has a valve and mesh filter at one or both ends, and a closure (such as a snap-fitting) at one or both ends. Suitable models for such a target container are anion exchange columns of the type provided by Hamilton Company of Reno, Nevada, U.S.A.
According to one aspect of the invention, there is provided a method of manufacturing the particles of UO2 for the matrix, the method comprising:
The polymer template may be in the form of PAN beads.
Herein, reference to uranium oxide/hydroxide is intended to refer to a mixture of uranium oxide and uranium hydroxide. Likewise, reference to cerium oxide/hydroxide is intended to refer to a mixture of cerium oxide and cerium hydroxide. The ratios will depend on the application, but in many cases the mixture may contain more of the oxide than of the hydroxide.
The polymer template may thus be removed and the uranium oxide/hydroxide (or uranyl nitrate) infiltrated in the polymer template converted to U3O8 concurrently, preferably by heating the infiltrated polymer template to a maximum temperature of 400° C. or 600° C. (It is envisaged that, in some applications, still higher calcination temperatures may improve bead stability, but run the risk of reducing porosity.)
The reduction of the U3O8 to UO2 is preferably at a maximum temperature of 1000° C.
Nitrate salts have the advantage of being highly soluble in water, which facilitates the uranyl nitrate's incorporation into the template (such as by soaking PAN in an aqueous solution of uranyl nitrate). If the template comprises beads, the beads desirably have a size (viz. mean diameter) selected to be—or to result in—the desired ultimate size of the particles. The solution of uranyl nitrate (or other precursor) comprises uranium with a 235U to 238U ratio of the desired 235U enrichment. The concentration of the solution and the volume infiltrated into the PAN beads are selected, in combination with the desired size of the particles, such that the final density of UO2 in the matrix is the desired density.
The porous matrix may then be manufactured by, for example, sintering the particles of UO2, or compressing the particles of UO2 within a suitable container. In this and other target manufacturing methods of this invention, if a later step-such as sintering or compression-changes the volume or density of the particles or matrix, that change in volume or density should be taken into account and allowed for when manufacturing the particles and/or matrix, so that the target, once manufactured, has the desired characteristics.
In this and other aspects in which the polymer template is in the form of PAN beads, the beads may be manufactured according to the following method. The method involves preparing a (e.g. 5 wt %) solution of PAN in dimethyl sulfoxide (DMSO), pressurizing the PAN solution (such as in a CS-1560 Loctite (trade mark) pressure chamber using a HP-2.0, 2HDD air compressor), passing the PAN solution under pressure into a (e.g. PTFE) nozzle with needle outlets (in one example with 21 gauge needles of length 1-4 mm), and vibrating the nozzle so as to emit droplets of the PAN solution. The method may include controlling regularity of the droplets (which in due course constitute the beads) by controlling the period of vibration of the nozzle. The period of vibration of the nozzle can be monitored, such as with an oscilloscope.
The method includes collecting the droplets in a container (such as a beaker), advantageously containing water and a structure directing agent (such as Pluronic F127), resulting in formation of PAN beads. The beads are advantageously washed, such as with water, to remove the structure directing agent. This may involve washing until the washings are ˜pH 7, indicating removal of the Pluronic (trade mark) F-127.
The method then includes cross-linking the PAN beads (such as in petri dishes in an evaporation chamber). In one example, this step is performed with a controlled flow of air, temperature (e.g. ˜35° C.) and humidity (e.g. ˜45% RH), such as for 3 days. The method includes subsequently drying the beads (such as in air for 24 hours followed by under vacuum for 3 hours).
According to another aspect of the invention, there is provided a method of manufacturing the particles of UO2 for the matrix by nanocasting or ‘repeat templating’, such as by creating a template comprising polymer beads, and infiltrating the beads with UO2, and calcinating the infiltrated beads. The matrix can then be formed by sintering or compressing the calcinated beads.
In this alternative aspect, the method may optionally be controlled to provide the matrix with a hierarchical porosity, with pores that are progressively smaller (or the density progressively greater) from the centre of the matrix to the periphery of the matrix. In such a configuration, the matrix may be uniform in the axial direction, but have a hierarchical porosity radially. Desirably, the matrix at or constituting the peripheral walls of the target has a lower density than the average density, to facilitate ejection of 99Mo from the particles and minimize the likelihood that the recoil distance for ejection of some of the 99Mo will be excessive.
Hierarchical porosity can be achieved, for example, by dropping droplets of PAN solution (with their size controlled/regulated via oscillation) into water containing a surfactant, which causes the beads to form. The removal/evaporation of the solvent (e.g. water/DSMO) in the formed bead results in the macroporosity. The meso/micropores exist as interparticle meso/micropores when the UO2 is introduced.
Thus, a porous matrix can be synthesized with a desired average density (as discussed above) and, optionally, hierarchical porosity.
The reusable uranium target makes use of the property of fission recoil whereby, when a fission occurs, the fission fragments have an initial energy that is dispersed via movement. The recoil energy (90 MeV) penetration range of 99Mo is about 7.15 μm in UO2 and 21.2 μm in H2O so, if the UO2 target has a particle size of 6±1 μm, the 99Mo will be ejected into the surrounding target medium—provided there is enough distance between the uranium particles so that the 99Mo does not implant itself into a neighbouring UO2 particle. The 99Mo can be chemically extracted from the target once the details of distribution of the UO2 particles in the matrix, the minimum particle separation distance, the absorption of the matrix, the radiation properties, and the efficiency of 99Mo extraction have been determined.
In principle, the UO2 matrix could contain other materials, but it is generally advantageous (with a specific exception discussed below) that the matrix and the target contain little or no other materials, as these can complicate both the neutronics (i.e. neutron transport) and 99Mo extraction. It will also be understood that the porosity of the target may have implications for the transfer from the target of the heat generated by neutron irradiation and the consequent nuclear fission and decay-such as reducing the ability of the heat to dissipate from the target (such as by conduction to a target cladding or to a heat transfer medium). However, this potential problem is ameliorated by the relatively low 235U enrichment of the target and/or intended irradiations times (of from 3 to 7 days). Indeed, it is envisaged that—in some examples—the heat will be just sufficient to at least partially reverse radiation damage (such that the target may be self-annealing to some degree and thereby reduce the risk or extent of pore collapse).
In one example, the matrix has an average density of less than or equal to 75% of the density of the UO2 (viz. approximately 8.23 g/cm3, depending on the 235U enrichment).
The matrix is typically of approximately uniform average density.
The density of UO2 per se is approximately 10.97 g/cm3, although this will vary to a small degree with 235U enrichment. The more porous the matrix (in this example with an average density of less than or equal to 75% of the density of UO2), the easier the 99Mo extraction, but this also reduces the total amount of 235U for any particular enrichment and target dimensions. Hence, the average density of the UO2 matrix will generally be selected so as to provide sufficient total yield of 99Mo and/or subsequently allow efficient 99Mo extraction, in a manner that balances these considerations, in the context of available reactor time, waste minimization goal, 235U enrichment, 99Mo demand and target dimensions.
In an example, the matrix has an average density of less than or equal to 65% of the density of the UO2 (viz. approximately 7.13 g/cm3, depending on the 235U enrichment). In another example, the matrix has an average density of less than or equal to 55% of the density of the UO2 (viz. approximately 6.03 g/cm3, depending on the 235U enrichment). In a further example, the matrix has an average density of less than or equal to 50% of the density of the UO2 (viz. approximately 5.49 g/cm3, depending on the 235U enrichment). In a still further example, the matrix has an average density of less than or equal to 45% of the density of the UO2 (viz. approximately 4.94 g/cm3, depending on the 235U enrichment).
In a particular example, the matrix has an average density of less than or equal to 40% of the density of the UO2 (viz. approximately 4.39 g/cm3, depending on the 235U enrichment). In another example, the matrix has an average density of less than or equal to approximately 2.5 g/cm3. In an example, the matrix has an average density of approximately 2.5 g/cm3, and in another an average density of approximately 2.0 g/cm3.
A lower average density (e.g. between 50% and 70% of the density of the UO2) may be advantageous in some applications in order to reduce waste, even at the expense of yield.
In an example, the average density is between 50% and 60% of the density of the UO2.
The average density may be an initial average density (that is, before the first use of the target for the manufacture of 99Mo).
It will be noted that depleted uranium may be employed. As will be appreciated, this may be less desirable in some applications, as—at lower 235U enrichments—yield will be reduced (all other parameters being equal). However, this effect can be at least somewhat compensated for by increasing average density.
In an example, the UO2 comprises uranium with a 235U to 238U ratio of between 0.3% and 3% 235U enrichment (i.e. 0.3%<235U enrichment <3%). In an example, the UO2 comprises uranium with a 235U to 238U ratio of between 0.5% and 3% 235U enrichment (i.e. 0.5%<235U enrichment <3%). In an example, the UO2 comprises uranium with a 235U to 238U ratio of between 0.7% and 3% 235U enrichment (i.e. 0.7%<235U enrichment <3%). In an example, the uranium has a 235U enrichment of <2.8%. In an example, the uranium has a 235U enrichment of <2.5%, and in another example, the uranium has a 235U enrichment of <2%. In an example, the uranium has a 235U enrichment of <1.8%. In an example, the uranium has a 235U enrichment of <1.6%. In a certain example, the uranium has a 235U enrichment of <1.4% and in another <1.2%.
In certain example, the uranium has a 235U enrichment of >0.75%, and in another example, the uranium has a 235U enrichment of >0.8%. In still another example, the uranium has a 235U enrichment of >0.9%.
It should be understand that this particular embodiment also includes examples with any combination of these upper and lower 235U enrichments. For example, examples with the following 235U enrichments are envisaged:
In an example, the uranium has a 235U enrichment of approximately 1%.
The 235U to 238U ratio may be an initial 235U to 238U ratio (that is, before the first use of the target for the manufacture of 99Mo).
In an example, the target is configured to yield a maximum amount of 99Mo and a maximum amount of burnup from a lowest initial amount of 235U, thus minimizing 235U waste.
In an example, the target is configured to maximize a sustainability index Starg, where:
where AT is a predefined amount of 99Mo desired to be produced in the irradiation, 235UT is the total amount of 235U in the target before the irradiation, and 235Ub is the amount of 235U burned up in the irradiation. The parameters 235UT and 235Ub may be established empirically or by modelling, such as before or after the irradiation. Though principally intended for a single irradiation, this relationship is also valid for plural irradiations—in which case AT would represent the total desired 99Mo yield, 235UT is the total amount of 235U in the target before the first irradiation and 235Ub the total amount of 235U burned up in all of the irradiations. Extensive modelling has shown that a change in volume does not substantially affect sustainability, such that volume changes—if any—could be neglected in the analysis of target performance.
The sustainability index Starg for one or more (n≥1) irradiations may alternatively be expressed as:
where ATi is the 99Mo yield of the i-th irradiation, 235UTi is the amount of 235U in the target before the i-th irradiation (or equivalently the amount of 235U in the target after the (i−1)-th irradiation, when i>1), and 235Ubi is the amount of 235U burned up in the i-th irradiation.
The UO2 target may be of any suitable dimensions, but is typically of a size dictated by the dimensions of the core of the reactor that is to be used to irradiate the target, including being able to fit the irradiation position or target holder within the reactor. For example, the height of the target is, in one example, less than or equal to the height of the core. That is, if the reactor core has a height of height of 60 cm, the target may be sized with a height of less than or equal to 60 cm.
As mentioned above, the UO2 matrix may contain other materials, provided they do not unduly complicate the neutronics or the 99Mo extraction. However, the target may be doped with one or more minor actinides in order to reduce proliferation concerns arising from 239Pu build-up (see Peryoga et al., (2005)). Suitable dopants (e.g. 237Np or a mixture of Np, Am and Cm) and amounts of doping (e.g. approximately 1% by mole relative to the 235U content) may be ascertained from Peryoga et al. (2005), which is incorporated herein by reference.
A major further example of the inclusion of another material arises from the fact that the crystal structures of cerium(IV) oxide (CeO2, also referred to as cerium dioxide or ceria) and UO2 are similar, as are their molar densities. Hence, CeO2 may be—in effect—substituted for at least some of the 238UO2. In principle, CeO2 may be substituted for substantially all of the 238UO2 (such that the particles comprise essentially only 235UO2 and CeO2, with possibly trace amounts of 238UO2), but it is expected that this would be needlessly or prohibitively expensive.
Thus, according to a second particular embodiment of this aspect of the invention, there is provided a UO2 target for use in the manufacture of 99Mo, the target comprising:
Generally, the matrix comprises enriched UO2 mixed with CeO2, in which the UO2 comprises uranium with a 235U to 238U ratio of less than or equal to 20% 235U enrichment (viz. low enriched uranium), but higher enrichments are possible and contemplated in order to further minimize the 238U content of the target. As mentioned above, the matrix may comprise 235UO2 and CeO2 only, but it may not be convenient or possible to obtain pure 235UO2. Even if 235UO2 is available, it may be more cost-effective to use a matrix that comprises 235UO2 or highly enriched UO2 mixed with natural UO2 and CeO2.
In certain examples, the molar ratio of 235U to Ce and 238U is between 0.3% and 3%, or between 0.5% and 3%, or between 0.7% and 3%, or between 0.75% and 2.8%, or between 0.8% and 2.0%, or between 0.9% and 1.4%.
In a particular example, the molar ratio of 235U to Ce and 238U is approximately 1%. If the molar ratio of U:Ce is 50%, this example corresponds to a UO2 feedstock with an 235U enrichment of approximately 2%.
In another example, the matrix comprises 50% UO2 and 50% CeO2 by mass, wherein the UO2 comprises uranium with a 235U enrichment of between 1.5% and 5.6%, or of between 1.6% and 4.0%, or of between 1.8% and 2.8%, or of approximately 2%. In these examples, therefore, the matrix comprises, respectively, between 0.75% and 2.8% 235UO2, between 0.8% and 2.0% 235UO2, between 0.9% and 1.4% 235UO2, and approximately 1% 235UO2, by mass (ignoring trace amounts of 234UO2).
In each example of this particular embodiment, the CeO2 typically comprises natural Ce. Natural Ce is predominantly (88.4%) 140Ce, so CeO2 comprising natural Ce is generally the least expensive form of CeO2. It will be appreciated, however, that other isotopes of Ce may be used, especially one or more of the naturally occurring isotopes.
The second particular embodiment shares the advantages of the first particular embodiment. In addition, a number of advantages arise from the use of cerium in this manner. For example, this particular embodiment effectively substitutes cerium for at least some of the 238U, and the thermal neutron absorption cross section of natural Ce is 0.63 barns whereas the thermal neutron absorption cross section of 238U is 2.68 barns. Hence, the production of plutonium in the form of PuO2 (from the irradiation of the 238U) can be substantially reduced. This also leads to greater efficiency, as fewer neutrons will be absorbed by the target so fewer neutrons are required in the production of 99Mo. (For example, it has been found that, when there are no Mo plates in the Australian Nuclear Science and Technology Organisation's OPAL reactor, the reactor uses 5% more fuel. This is because the Mo plates comprise LEU so generate their own neutron flux, in essence acting like fuel.)
The target of the second particular embodiment behaves much as does the target of the first particular embodiment, so each of the optional features disclosed above in the context of the first particular embodiment are likewise optional features of the second particular embodiment, though with CeO2 substituted for at least some of the 238UO2 of the first particular embodiment and with consequent adjustment of various parameters as required.
In certain examples of the second particular embodiment, the matrix has a porosity such that an average density of the matrix is less than or equal to 50% of the density of the UO2 and CeO2 content.
Cerium dioxide (if comprising natural cerium) has a density of approximately 7.215 g/cm3 whereas, as mentioned above, the density of UO2 depends on its 235U enrichment; with the naturally occurring isotopic abundances, density of UO2 is approximately 10.97 g/cm3. (The densities of 235UO2 and 238UO2 are approximately 10.850 g/cm3 and 10.972 g/cm3 respectively.) Consequently, in an example in which the matrix comprises essentially only 235UO2 and CeO2, with a molar ratio of 235U to Ce of just under 3%, the UO2 and CeO2 content has an average density of approximately 7.32 g/cm3. Hence, an average density of the matrix of less than or equal to 50% of the density of the UO2 and CeO2 content equates to an average density of less than or equal to approximately 3.66 g/cm3.
In another example, the matrix has a porosity such that an average density of the matrix is less than or equal to 50% of the density of the UO2 and CeO2 content, but non-235UO2 content has a molar ratio of 50% 238UO2 and 50% CeO2, again with a molar ratio of 235U to Ce and 238U of just under 3%. CeO2 has a density of about 41.9 mmol/cm3, and 238UO2 a density of about 40.6 mmol/cm3, so the density of the combined CeO2 and 238UO2 is approximately 41.25 mmol/cm3, implying a density of 235UO2 of approximately 1.256 mmol/cm3.
The particles, porous matrix and target of this particular embodiment may be manufactured as described above in the context of the first particular embodiment of the first aspect of the invention, varied to incorporate the CeO2, such that—in effect—some of the UO2 is replaced with CeO2 and the resulting matrix comprises a desired molar ratio of 235U to Ce and 238U. According to another aspect of the invention, there is provided a method of manufacturing the particles, comprising:
In one example, this method comprises forming the particles of UO2 and the particles of CeO2 sequentially, in which case the method results in two sets of particles (those comprising UO2 and those comprising CeO2) which are then mixed.
The particles (whether one or two sets) are formed into the porous matrix by, for example, sintering the particles, or compressing the mixed sets of particles within a suitable container. Again, to prevent oxidization, the target is desirably housed in a sealable target container, optionally backfilled with helium gas.
The ratio of cerium and uranium can be controlled as desired, such as by controlling the ratio of the sizes of the first and second sets of particles, and/or by controlling the amount or amounts of infiltration of the cerium salt and uranyl nitrate.
If the template comprises PAN beads, the beads are selected to have a size (viz. mean diameter) to be or result in the desired size of the particles, and the solution or solutions having a concentration or concentrations and a volume or volumes such that the resulting matrix comprises a desired molar ratio of 235U to Ce and 238U and, in combination with the desired size of the particles, such that the final density of UO2 in the matrix is a desired density.
According to another aspect of the invention, there is provided a method of manufacturing particles of UO2 and CeO2 for a porous matrix of a target for use in the manufacture of 99Mo, the method comprising:
Subsequently, the size of the particles will depend on (and be controlled by) for how long and/or at how high a temperature sintering is performed when forming the porous matrix.
According to another aspect of the invention, there is provided a method of manufacturing, comprising nanocasting or ‘repeat templating’, such as by creating a template comprising polymer (e.g. PAN) beads and infiltrating the beads with cerium and uranium (as described above), and calcinating the infiltrated beads. The matrix can then be formed by sintering or compressing the calcinated beads. Optionally, the method may be controlled to provide the target with a hierarchical porosity, as described above, wherein meso/micropores exist as interparticle meso/micropores when UO2 and/or Ce is introduced.
The cerium for infiltration may be in any suitable form, such as a cerium salt (e.g. cerium(III) nitrate (Ce(NO3)3), cerium(III) oxalate (Ce2(C2O4)3), or cerium(III) acetylacetonate (Ce(C5H7O2)3(H2O)x)). As mentioned above, nitrate salts are highly soluble in water, which facilitates cerium nitrate's incorporation into the template (such as by soaking PAN in an aqueous solution of uranyl nitrate and cerium nitrate).
The ratio of infiltrated cerium and uranium and the enrichment of the uranium (in whatever form is employed) are selected to provide the desired ultimate molar ratio of 235U to Ce and 238U.
Targets according to this particular embodiment may also be doped with one or more minor actinides (e.g. 237Np or a mixture of Np, Am and Cm) in order to reduce proliferation concerns. Suitable dopants (e.g. 237Np or a mixture of Np, Am and Cm) and amounts of doping (e.g. approximately 1% by mole relative to the 235U content) may be ascertained from Peryoga et al. (2005).
According to a second aspect of the invention, there is provided a method of producing 99Mo (or use of a UO2 target to produce 99Mo), the method comprising:
In an embodiment, the method includes a delay between an instance of step (a) and a next instance of step (a) (such as before and/or after step (b)), sufficient to allow—in combination with the time required to perform step (b)—one or more by-products (such as 135Xe) in the target to decay to a predefined level. In one example, the predefined level is less than 50% of the amount of a specified by-product (e.g. 135Xe) present at the end of step (a). In another example, the predefined level is less than 25% of the amount of a specified by-product present at the end of step (a), and in another less than 12.5% of the amount of a specified by-product present at the end of step (a).
As mentioned above, the relatively short irradiation time has the advantage of minimizing target heating and hence the risk of target damage. In addition, this effect—as well as the low 235U enrichment—reduces the production or build-up of the by-product, 135Xe. As will be appreciated by the skilled person in this field, 135Xe has a much higher neutron absorption cross-section than does 235U, so reduces the neutron flux available for the production of manufacture 99Mo. Short irradiation times minimize 135Xe build-up and, as 135Xe has a half-life of 9.1 h, the time required to extract the 99Mo from the target (and any further optional delay) allows time for significant 135Xe decay (as well as decay of its daughter, 135Cs).
In one embodiment, the method includes performing steps (a) and (b) 3 or more times. In another embodiment, the method includes performing steps (a) and (b) 4 or more times. In still another embodiment, the method includes performing steps (a) and (b) 2 to 6 times.
In a further embodiment, the method includes performing steps (a) and (b) 3 to 5 times (i.e. the target is re-irradiated and re-processed to extract 99Mo—after a first irradiation and processing—2 to 4 times).
Generally, the maximum number of times the target is irradiated and the 99Mo yield extracted depends on how many times the target can be profitably used. This maximum may correspond to the 99Mo yield's becoming too low to justify the expense of operating the reactor, and/or to justify the expense of performing 99Mo extraction, and/or to justify the waste generated by the method, and/or to satisfy 99Mo demand/requirements.
In an embodiment, the irradiation time is between 4 and 6 days. In one embodiment, the irradiation time is between 4.5 and 5.5 days. In a particular embodiment, the irradiation time is approximately 5 days.
The irradiation may be performed with, for example, a nuclear reactor that includes a heavy water reflector vessel with a UO2 core (e.g. a reflector vessel with a diameter of 200 cm and a height of 120 cm, and a UO2 core with a diameter of 30 cm and a height of 60 cm).
It should be noted that any of the various features of each of the above aspects of the invention and of the embodiments detailed below can be included or combined, as suitable and desired, in each of those aspects.
In order that the invention be better understood, embodiments will now be described, by way of example, with reference to the accompanying drawing in which:
Reflector vessel 20 has a diameter of 200 cm and a height of 120 cm. UO2 core 30 has a diameter of 30 cm and a height of 60 cm.
For reactor model 10 to simulate a practical reactor, the amount of uranium in UO2 core 30 is adapted to allow a self-sustaining nuclear reaction. The sustainability of a nuclear reaction is given by the reactor's effective neutron multiplication factor, keff:
where keff>1 indicates supercriticality: the number of neutrons produced by fission is greater than the number lost;
To determine the density of UO2 in UO2 core 30 that will produce a keff of approximately 1, a number of different densities of UO2 core 30 were modelled using the KCODE function in MCNP6 (trade mark), a Monte-Carlo radiation transport code that can be used to track different particle types over a broad range of energies and has user-definable variables such as geometries and timeframes.
Reactor model 10 was created with an initial value for keff of 1.0, and 5000 neutrons per cycle were generated. A total of 250 cycles were run, with data accumulation commencing after the first 50 cycles, resulting in approximately 200 million neutron collisions. These numbers were chosen to make the computing time practical.
In order for 99Mo to be ejected from the UO2 particles in reusable target 40 and into the surrounding material, the density of the UO2 needs to be adjusted downwards to allow for the presence of other materials or voids that will be used to contain the 99Mo prior to chemical extraction. MCNP6 was used to model different UO2 densities, with reactor model 10 at 20 MW and using the BURN function of MCNP6. When using the BURN function, the fission products produced are grouped into three tiers. Tier 1 includes the isotopes: 93Zr, 95Mo, 99Tc, 101Ru, 131Xe, 134Xe, 133Cs, 137Cs, 138Ba, 141Pr, 143Nd, 145Nd. Tier 2 and tier 3 contain progressively more and more isotopes (which are listed in MCNP6 User's Manual). For calculation simplicity Tier 1 was used with the additional inclusion of 99Mo and 135Xe, as MCNP6 allows the addition of user-selected isotopes to the output. To compare the properties of targets with different 235U to 238U ratios, two types of targets were modelled using MCNP6: 20% enriched, and 1% enriched.
Firstly, reusable target 40 was modelled with a 20% 235U enrichment, as shown in Table 1:
235U Enrichment (%)
It will be noted from
Secondly, reusable target 40 was modelled with a 1% 235U enrichment, as shown in Table 2:
235U Enrichment (%)
Compared with the 20% enriched target, the 1% enriched target had a relatively linear relationship between activity and density from 1 g/cm3 to 10.97 g/cm3, which is higher than that over the density range of 5 to 6 g/cm3 for the 20% enriched target—consistent with the idea that, as UO2 density increases, the amount of fissioning that occurs per 235U atom decreases. Tables 3 compares the amount of 99Mo produced with a UO2 density of 6 g/cm3, with 20% 235U enrichment and 1% 235U enrichment respectively:
Hence, the amount of 99Mo produced is only 7.5-8.6 times higher with the 20% enriched target as compared to the 1% enriched target, despite the fact that the amount of 235U in the 20% enriched target is 20 times greater than in the 1% enriched target. That is, when considering 99Mo produced per quantity of 235U present in the target, the 1% enriched target was found to be 2.3-2.7 times more productive than the 20% target, according to the MCNP6 model used.
Another parameter to be considered in designing reusable target 40 is the amount of waste produced, which depends on the target efficiency. Target efficiency εtarg can be expressed as the total activity of 99Mo produced per total mass of 235U in the target:
Target efficiency εtarg was thus calculated for both the 20% enriched UO2 target and the 1% enriched UO2 target, for 2, 5 and 10 day irradiations and with UO2 densities ranging from 1 to 10.97 g/cm3. The results are plotted in
Another consideration in target design is the amount of 235U burnup, as burnup affects the waste produced and the number of times a target can be reused. Firstly, typical waste from fission based uranium targets is spent uranium containing an isotopic ratio of approximately 19.7% 235U/238U due to the 2-3% burnup for 99Mo production. A target with a burnup greater than 2-3% thus implies reduced nuclear waste.
Secondly, as the amount of 235U reduces with target burnup (owing to the destruction of 235U atoms), the amount of 99Mo produced with each subsequent irradiation is reduced. Eventually, 99Mo production is too low to warrant an additional irradiation.
The burnup percentage of 235U in the 20% and 1% 235U targets was modelled for irradiations of 2 days, 5 days, 10 days, four×5 days and ten×5 days, for UO2 densities ranging from 1 to 10.97 g/cm3 using the BURN function of MCNP6. The four×5 (=20) day and ten×5 day (=50) day irradiations were modelled to simulate a target being irradiated, 99Mo extracted and the target re-irradiated multiple times, to obtain an indication of how times a target can be profitably reused.
The results are shown in
These simulations suggest that, for high efficiency and reusability, reusable target 40 advantageously has these characteristics:
However, as the overall yield produced with this target design is lower than with a 20% enriched target, a balance must be struck between (a) efficiency and reusability, and (b) total yield, such as by suitable selection of target size and volume, ideally to approach the yield that can be obtained with a 20% enriched target.
To identify a suitable balance, the maximum output AT produced per gram of 235U burned up was examined—which would allow 99Mo producers to reduce the generation of nuclear waste.
Current methods of 99Mo production are characterized by the formula:
which is commonly expressed in GBq per week. When designing a target with this formula in mind it is understandable to pack as much 235U into the target as possible to ensure the maximum number of total fissions per unit time. In such cases, the 235U is in a state of saturation as there is significantly greater quantities present in the target than will ever fission. However, the efficiency of 99Mo target 40 may be expressed as the amount of activity produced per gram of 235U burned up, or 235Ub, rather than—as discussed above—per gram of 235U initially in the target. Hence:
A further parameter is then introduced to take into account the total output (AT), a parameter termed ‘target quality’ or Qtarg, where:
Thus, a target with a high Qtarg would produce the highest 99Mo output for the most 235U burned. Next, it is desirable to consider the total amount of 235U originally in the target before irradiation, 233UT, because the amount remaining in the target after the target's use should—all things being equal—be minimized, and the amount remaining is the difference between 235UT and the 235Ub. Hence, a target sustainability index Starg is proposed,
where:
Hence, a reusable target 40 with high 99Mo Starg would produce the maximum output with the highest burnup from the lowest initial amount of 235U, thus minimizing 235U waste.
MCNP6 was again used to model both 235U burnup in grams and AT of 99Mo produced. The modelling was conducted with UO2 target densities of 0.2 to 8 g/cm3 in 0.2 g/cm3 intervals, irradiation times of 2, 3, 4, 5, 6, 7, 8, 9, 10, 15 and 20 days, and target enrichments (% 235U/238U) of 1%, 3%, 7% and 10%.
From
In a commercial context, a program for the manufacture of 99Mo will commonly be expressed in terms of the amount of 99Mo to be produced in a specific period. For example, the 99Mo manufacturing plant of the Australian Nuclear Science and Technology Organisation was designed to produce 3000 curie (=111 TBq) per week. Hence, in practical applications it may be important to determine the most sustainable process (viz. with the highest sustainable index) that produces a specified total activity (e.g. AT=111 TBq) in a specified target irradiation time (e.g. 4≤t≤7 days: cf. the simulations discussed above).
From
It will be noted that plutonium production decreases essentially monotonically with increasing 235U enrichment.
The simulation employed a UO2 density of 10.97 g/cm3, and SRIM's standard stopping energies. The average longitudinal range (that is, in the +z direction) of the Mo ions was found to be 7.16 μm with a straggle of 6489 A. The average radial range of the Mo ions was 1.20 μm with a straggle of 5983 A.
These plots simulate the travel of the 99Mo within, and hence likelihood of ejection from, UO2 and CeO2, respectively. It may reasonably be expected that the range of the 99Mo in a mixture of UO2 and CeO2 would be essentially a linear combination of the individual ranges. For example, a target with a UO2 to CeO2 ratio of 50:50 may be expected to have a 99Mo range that is approximately the average of the two shown in these plots.
It is evident from these simulations that Mo ions travel further and deviate less in CeO2 than in UO2, as might be expected in view of the lower density of CeO2. Channeling and other effects are expected to be essentially the same, owing to the similar crystal structures of UO2 and CeO2. Thus, from this perspective there should be no disadvantage to the use of CeO2 in conjunction with UO2, and the greater range of the Mo ions in CeO2 will—all things being equal—increase the proportion of 99Mo that will be ejected.
However, reusable target 50 comprises a porous matrix of particles that comprise a mixture of UO2 and CeO2 (of natural cerium) in a U:Ce molar ratio of 50%. The particles have a size (viz. mean diameter) of 6 μm. In this example, the molar ratio of 235U to Ce and 238U is approximately 1%, so the target contains 235U, 238U and Ce in the (molar) proportions of approximately 1:49:50. This corresponds to a UO2 feedstock with an 235U enrichment of approximately 2%.
Target 50 is thus comparable in performance to a UO2 target of like characteristics (but omitting cerium) of 1% 235U enrichment, such that 235U and 238U are present in the molar ratio of approximately 1:99. However, owing to what is, in effect, the substitution of 49/99=49.5% of the 238UO2 with CeO2, the density of target 50 is approximately 17% lower than the density the comparable UO2 only target—with the benefit of facilitating 99Mo ejection, as discussed above.
It is evident that plutonium production can be substantially reduced by, in effect, substituting CeO2 for 238UO2. It will be noted that—with 1% 235U and 99% Ce and hence no 238U—plutonium production is effectively eliminated.
Referring to
By heating the infiltrated polymer template, the uranium oxide/hydroxide is converted into U3O8 (cf. step 68) and, concurrently, the polymer template is removed (cf. at step 70). The method then continues at step 72, where the U3O8 is reduced to UO2 via heating (such as at a maximum temperature of 1000° C.) in a reducing atmosphere (such as 3.5% hydrogen in nitrogen gas).
It will be understood that effecting steps 68 and 70 concurrently (and other pairs of steps described and claimed herein as performed concurrently) does not imply that both steps will commence simultaneously (once heating commences) or reach completion simultaneously.
If, after step 62, the method continues at step 66, then by heating the infiltrated polymer template, the uranyl nitrate is converted into U3O8 (cf. step 66) and, concurrently, the polymer template is removed (at step 74).
The uranyl nitrate may be converted into U3O8 (see step 66) by removing the nitrate by, for example, direct denitration. For example, this can be done by heating the sample (e.g. to >300° C., thereby also effecting the concurrent template removal of step 74) in a rotary kiln or a fluidized bed reactor. The rotary kiln is harsher, and may crush the beads owing to their fragility, so it is envisaged that a fluidized bed reactor is likely to be more advantageous in that regard.
The method then continues at step 72.
Steps 70 and/or 74 may comprise heating the infiltrated polymer template to a maximum temperature of 400° C.
Referring to
If, after step 82, the method instead continues at step 86, by heating the infiltrated further polymer template (such as in a fluidized bed reactor), the cerium salt is converted into CeO2 (cf. step 86) and, concurrently, the further polymer template is removed (cf. step 92).
Thus, if the particles of UO2 and the particles of CeO2 are formed sequentially, they can then be mixed in readiness for forming the matrix. In addition, the method can include controlling the ratio of cerium and uranium by controlling the amount or amounts of infiltration of the cerium salt (at step 82) and uranyl nitrate (at step 62).
Referring to
At step 104, a gaseous base or other alkali chemical (such as gaseous ammonia) is introduced to the uranium and cerium nitrate infiltrated polymer template, causing co-precipitation of the uranium oxide/hydroxide and cerium oxide/hydroxide. By heating the infiltrated polymer template, the uranium oxide/hydroxide and cerium oxide/hydroxide are converted to respectively U3O8 and CeO2 (cf. step 106) and, concurrently, the polymer template is removed (cf. step 108).
At step 110, the U3O8 and CeO2 is reduced to a UO2/CeO2 system (UxCe1-xO2, where x is the initial molar mixing ratio of uranium and cerium) in a reducing atmosphere (such as 3.5% hydrogen in nitrogen gas).
The particles of UO2 and CeO2 are formed non-sequentially (concurrently), and method 100 can include controlling the amount or amounts of infiltration of the uranyl nitrate and cerium nitrate (step 102) to achieve a desired molar ratio.
Subsequently, the size of the particles will depend on (and be controlled by) for how long and/or at how high a temperature sintering is performed when forming the porous matrix.
Manufacture of Porous UO2 and UxCe1-xO2 Targets Using Nanocasting
The synthesis of porous CeO2 (acting as a UO2 simulant), UO2 and UxCe1-xO2 were investigated using nanocasting, with the object of making a porous UO2 system for 99Mo production with a particular focus on materials with a lower density and higher volume compared to conventional smaller volume, high density 99Mo production targets.
In the following examples, simultaneous thermal gravimetric and differential scanning calorimetry analysis (TG-DSC) data were recorded using a Netzsch STA449F3 (trade mark) heating at a rate of 5° C. min−1 under a flow of either Ar or 20% O2 in N2 at 30 cm3·min−1. Gas adsorption studies were carried out using a Quantachrome (trade mark) Autosorb MP instrument and high purity nitrogen gas (99.999%). Surface areas were determined using Brunauer-Emmett-Teller (BET) calculations.
A Zeiss Ultra Plus (trade mark) scanning electron microscope (SEM, Carl Zeiss NTS GmbH, Oberkochen, Germany) operating at 15 kV equipped with an Oxford Instruments X-Max (trade mark) 80 mm2 SDD X-ray microanalysis system was used to check the crystal morphology and electron dispersive spectroscopy (EDS) calibrated with a Cu standard for the determination of key elements.
Synthesis of UxCe1-xO2 Beads
An aqueous solution was prepared by dissolving known amounts of UO2(NO3)2·6H2O and Ce(NO3)3·6H2O in H2O to achieve a desired molar ratios (100% uranium, 100% cerium, 5% uranium in cerium). This solution was then used for infiltration into polyacrylonitrile (PAN) beads. The PAN beads were synthesized using the method described by J. Veliscek-Carolan et al. (2015). The infiltration was achieved by heating the PAN-U/Ce solution in an oven at 60° C. overnight (Ibid).
Upon infiltration, the beads were removed from the U/Ce solution and vacuum dried at room temperature for 60 minutes. After drying, the beads were placed in an evaporating dish alongside a separate dish containing a solution of a base (e.g. for 100% cerium and 5% uranium in cerium: a 20% ammonia solution). The evaporating dish was covered and left overnight. The beads were collected the next day and washed with H2O three times over three hours and left to air dry.
Air Calcination of UxCe1-xO2 Beads
The beads were heated in air at a rate of 1° C./min to 400 or 800° C. and held at this temperature for 5 hours before being cooled to room temperature.
Pyrolysis of UxCe1-xO2 Beads
Uncalcined beads were first heated in air at a rate of 1° C./min to 230° C. and held at this temperature for 3 hours before being cooled to room temperature. The beads were then heated under argon at a rate of 1° C./min to 800° C. or 1200° C. and held at this temperature for 3 hours before cooling to room temperature.
The synthesis of the uranium-cerium containing beads was achieved in a two-stage process. The first stage involves infiltrating the PAN beads with the desired molar ratio of U/Ce in a concentrated aqueous solution containing known amounts of UO2(NO3)2·6H2O and Ce(NO3)3. The need for an aqueous solution is evident by the incompatibility of PAN with concentrated amounts of nitrate i.e., a melt reaction.
Upon infiltration, to convert the NO3 species to their oxide counterparts, the U/Ce is precipitated as UxCe1-xO2 via vapour diffusion of a base such as NH3 using a covered evaporating dish. Removal of the nitrate species was achieved by washing the precipitated UxCe1-xO2@PAN with water. The molar ratios explored so far are UO2, CeO2 and U0.05Ce0.95O2, with precipitation using gaseous NH3 used for both the CeO2 and U0.05Ce0.95O2 samples.
Upon precipitation, the PAN was removed by heating the samples under a controlled atmosphere. Two atmospheres have been explored so far. The use of an air atmosphere can be used to completely remove the PAN, with the resulting porous material existing entirely as UxCe1-xO2. The alternative option is to use an Ar atmosphere to pyrolyze the material, resulting in decomposition of the PAN without completely removing the carbon. The purpose of leaving the carbon within the structure is to ideally make the beads more robust and mechanically stable, so that they are suitable for use as a reusable 99Mo production system. Explored below is the characterization of the materials under both atmospheres.
TG-DSC was performed on the CeO2@PAN to determine the temperature at which the PAN can be removed from the structure, and to ensure the remaining CeO2 remained thermally stable past this point. As the equipment is located in a non-active area, only characterization of the inactive CeO2 material has been performed thus far.
The stable mass and return of the DSC trace back to zero after this loss of PAN suggests a completed reaction, and thus the amount of PAN within the material can be totaled as ˜30% of the total mass. Additionally, this confirms that 400° C. is the minimum target temperature to remove the PAN from the porous beads, leaving behind just CeO2.
SEM-EDS was the primary method chosen to examine the UxCe1-xO2 beads after calcination under both an air and Ar atmosphere, focusing on determining (a) whether the porous structure remains intact upon removal of the PAN, and (b) the resulting pore widths and hierarchical porosity.
SEM of CeO2 after Air Calcination
These images reveal an intact bead exhibiting clear hierarchical porosity throughout. The pore wall thicknesses were examined, revealing that the walls were progressively thicker closer to the centre of the bead. The pores near the outer edges of the beads had wall thicknesses of 3.5-4.5 μm (see
SEM of UO2 after Air Calcination
The same calcination procedure was applied to porous UO2 beads, but the SEM results confirmed that the internal structure of the bead was not intact after PAN removal.
Consequently, weaker gasesous bases are proposed, to allow the base to infiltrate the bead further before precipitation of the UO2.
SEM of CeO2 after Ar Calcination
The achievability of these pore wall thicknesses achievable established that the desired properties for a target could also be achieved under pyrolytic conditions.
SEM-EDS of U0.05Ce0.95O2 after Ar Calcination
Synthesis and subsequent characterization of a 5% uranium in cerium (U0.05Ce0.95O2) bead was also performed using the gaseous NH3 precipitation method, and subsequently characterized with SEM-EDS.
Mechanical Stability of CeO2 Beads after Air and Ar Calcination
These observations suggest that the air calcined material is much more delicate than the same material after Ar calcination. The air calcined beads appear unable to be handled with tweezers without extreme care, whilst the Ar beads are much more robust, increasing in apparent structural integrity as the calcination temperature is increased.
Porosimetry was performed on two samples: CeO2 beads after air calcination at 400° C. and CeO2 beads after Ar calcination at 1200° C. Thus,
The N2 isotherms, measured at 77 K, reveal that both samples remained porous upon calcination and removal of the PAN. The air calcined CeO2 beads were calculated to have a BET surface area of 57.823 m2/g, which dropped to 11.212 m2/g in the Ar calcined beads. One possible reason for this decrease is the carbon remaining in the Ar calcined material, which would reduce the accessible pore space compared to the purely CeO2 samples made under the air calcination. However, even with the lower observed surface area, the beads remain porous so constitute a reusable platform for 99Mo production.
It is to be understood that, if any prior art is referred to herein, such reference does not constitute an admission that the prior art forms a part of the common general knowledge in the art in any country.
In the claims which follow and in the preceding description of the invention, except where the context requires otherwise owing to express language or necessary implication, the word “comprise” or variations such as “comprises” or “comprising” is used in an inclusive sense, i.e. to specify the presence of the stated features but not to preclude the presence or addition of further features in various embodiments of the invention.
| Number | Date | Country | Kind |
|---|---|---|---|
| PCT/AU2022/050052 | Feb 2022 | WO | international |
This application is based on and claims the benefit of the filing date of International Patent Application no. PCT/AU2022/050052, filed 2 Feb. 2022, the content of which as filed is incorporated herein by reference in its entirety.
| Filing Document | Filing Date | Country | Kind |
|---|---|---|---|
| PCT/AU2023/050068 | 2/2/2023 | WO |
