NL2011311A

NL2011311A - Extracting method of radioactive 99mo from low-enriched uranium target.

Info

Publication number: NL2011311A
Application number: NL2011311A
Authority: NL
Inventors: Chang-Kyu Kim; Ho Jin Ryu; Jong-Man Park; Moon-Soo Sim; Hyun Seok An; Se-Jung Jang; Sun-Chil Kwon; Yong Jin Jeong; Eung-Soo Kim
Original assignee: Korea Atomic Energy Res
Priority date: 2012-08-16
Filing date: 2013-08-16
Publication date: 2014-02-18
Also published as: FR2994574B1; NL2011311B1; KR101460690B1; KR20140023514A; FR2994574A1

Description

EXTRACTING METHOD OF RADIOACTIVE 99Mo FROM LOW-ENRICHED URANIUM TARGET

BACKGROUND

1. Field of the Invention

The present invention relates to a method of extracting radioactive molybdenum-99 (99Mo) from low-enriched uranium target.

2. Description of the Related Art

Technetium-99 (symbolized as ,99mTc' ) takes up approximately 80% of radioisotope consumption for medical diagnostic purpose, and it is an important medical radioisotope used for the nuclear medical diagnosis of diseases. 99mTc is the artificial element that is not naturally found, and also a daughter nuclide created by the nuclear decay of 99Mo. 99Mo is the one and only parent nuclide of the medical radioisotope (99mTc) and can be mainly produced by the following two methods.

The first method extracts Mo from the fission products of uranium decay, and the second method produces 99Mo by neutron activation of 98Mo. Considering many shortcomings of the second method such as difficulty of obtaining raw material (i.e., enriched 98Mo) which is expensive, but yields a lower specific activity, the first method is most widely used. Currently, the high-enriched uranium (HEU) with density of 90% or above is mainly used as the target for producing 99Mo by nuclear fission.

Most large-scale 99Mo producing facilities produce uranium targets using alloys of high-enriched uranium and aluminum. The alloys of aluminum added with uranium are present in the form of microstructure in which UA13 and UA14 phases with excellent neutron stability are precipitated and dispersed in A1 matrix during quenching, and which has uranium content of approximately 1.5 g-U/cc.

The pool type research reactors generally perform neutron irradiation for the purpose of uranium target production, as these provide easy loading and unloading, and operability at low temperature. Since the targets are exposed to neutron irradiation, these are cooled under water for about 6 hr before being transported by the underwater channel or special-purpose containers into the hot cells of the 99Mo extraction facility. After transportation, the uranium targets are disintegrated, and charged into a dissolver where these are dissolved in alkaline solution containing NaOH. Along the process, Al, in the UA1X intermetallic compounds, is dissolved and contained in a liquid state, U forms oxide or hydroxide and exists in a slurry state, so that solid U sludge is generated as filtered therethrough. After that, 99Mo extraction separation and purification from the filtered solution are performed by ion exchange, etc.

In a research to find low-enriched uranium (LEU) targets as a replacement for the highly-enriched uranium (HEU) with high nuclear proliferation risk, U.S. Department of Energy's (DOE) Argonne National Laboratory developed high density targets as thin uranium metal sheets ("foils") which are approximately 120 ~ 150 jum in thickness inserted in two aluminum discs. Since the temperature at the center of the uranium foil targets is relatively low, the targets do not suffer swelling from the short irradiation period which spans approximately 5 days, and thus is evaluated to be usable.

A metal target is generally subject to deformation due to irradiation defect and nuclear fissile products, when irradiated in the reactor, and it may suffers anisotropic deformation. Among the fission products, solid elements have volume expansion as the number of atoms increases, but the gas elements have greater volume expansion and deformation due to generation of bubbles. Because higher temperature accelerates atomic diffusion and bubble generation, the targets in the form of foil can efficiently release the heat generated from the nuclear fission, using an aluminum material with excellent thermal conductivity for the cladding or dispersion matrix.

After transported to the hot cell facility, the uranium foil targets are eliminated of aluminum tubes which have wrapped around the uranium foils, so that the uranium foils are taken out and put into the dissolver to be dissolved into nitric acid (HNO3) . However, since the highly radioactive nuclear fission gases can be generated and contaminate the interior of the hot cells when the neutron-radiated uranium foil targets are taken out of the dissolver, extra precautions are required to prevent leak outside the hot cells. Since uranium is hardly dissolved in alkaline solution such as NaOH, nitric acid has to be used, in which case the solution left after separation of 99Mo by ion adsorption from the uranium solution dissolved in nitric acid is categorized into nuclear waste that has to be accordingly dealt with, as it contains uranium. However, since the liquid nuclear waste with uranium dissolved therein has low critical mass for nuclear fission chain reaction, this has disadvantages such as need for large space, corrosion of nearby equipments and facilities due to acid fume.

To address the above-mentioned disadvantages, the Argonne National Laboratory researches a method that can utilize the conventional process of extraction-purifying 99Mo from alkaline solution by dissolving uranium foils in alkaline solution and taking out uranium as solid sludge. However, as explained above, considering the fact that uranium metal hardly dissolves in alkaline solution such as NaoH, an option of dissolving in NaHCCh solution by electrolysis was developed, in which the uranium foils are used as anodes. However, this method also suffers a shortcoming because the uranium foils fall into pieces as these are almost dissolved, causing power failure and discontinued dissolution.

Producers of large-scale uranium targets are also actively involved with researches to increase uranium density of the targets which are Δ1 substrate with UA1X dispersion. For example, it has been suggested that uranium density be increased by preparing UAI2 powder with higher purity and higher U content than conventional UA1X dispersed in the uranium alloy target, and mixing the same with A1 powder and performing compressing and rolling. However, although the above suggestion was able to increase the uranium density from 1.5 g-U/cc to 2.7 g-U/cc, the U density was less than approximately 7.0 g-U/cc, the density necessary to compensate the reduction of uranium enrichment from 90% to 20%.

Accordingly, while searching for the ways to prepare high-density uranium targets using low-enriched uranium, the present inventors could develop a method for separating and extracting 99Mo with a conventional alkaline solution processing, by preparing high-density target by mixing metal uranium powder and aluminum power and compressing and rolling the same, and altering, by thermal treatment, the metal uranium insoluble in alkaline solution into uranium aluminide that is soluble in alkaline solution, and thus completed the present invention.

SUMMARY

Exemplary embodiments of the present inventive concept overcome the above disadvantages and other disadvantages not described above. Also, the present inventive concept is not required to overcome the disadvantages described above, and an exemplary embodiment of the present inventive concept may not overcome any of the problems described above.

Accordingly, a technical object is to provide a method for extracting radioactive 99Mo from low-enriched uranium target.

According to an embodiment of the present invention, a method for extracting radioactive 99Mo from low-enriched uranium target is provided, which may include the steps of: heating high-density low-enriched uranium target containing low-enriched uranium alloy particles (step 1); dissolving the high-density uranium target heated at step 1 in alkaline solution (step 2); and extracting 9Mo from the solution in which the uranium target is dissolved at step 2 (step 3).

According to a method for extracting radioactive 99Mo of the present invention, high-density low-enriched uranium target is thermally treated, according to which the metal uranium is converted into uranium aluminide which can dissolve in alkaline solution. As a result, extraction process using alkaline solutions as conventionally used, can be used. Furthermore, use of low-enriched uranium target does not compromise yield of 99Mo, while an amount of radioactive waste can also be reduced.

BRIEF DESCRIPTION OF THE DRAWINGS

The above and/or other aspects of the present inventive concept will be more apparent by describing certain exemplary embodiments of the present inventive concept with reference to the accompanying drawings, in which: FIG. 1 schematically illustrates plate type uranium targets; FIG. 2 schematically illustrates annular type uranium targets; and FIGS. 3 and 4 are SEM/EDS analysis on low-enriched uranium target after heating at step 1 according to Examples 1 to 3.

DETAILED DESCRIPTION OF EXEMPLARY EMBODIMENTS

Certain exemplary embodiments of the present inventive concept will now be described in greater detail with reference to the accompanying drawings.

However, the description is provided hereinbelow only to help understand certain embodiments of the invention and the invention should not be limited to any specific embodiment.

The expression "low-enriched uranium (LEU)" as used herein refers to uranium isotope containing 20% or less 235U which is fissile. Uranium (U) exists as isotope such as 238U and 235U, 238U which is hardly fissile makes up approximately 99.3 % of natural uranium, while fissile 235U makes up approximately 0.7 %. Compared to high-enriched

uranium (HEU) containing 90% or more 235U, the LEU

containing relatively less U content is more suitable for the purpose of preventing nuclear proliferation.

The expression "target" as used herein refers to a material containing uranium which has nuclear fission by the neutron irradiation, and which contains fission product including 99Mo .

The method for extracting radioactive 99Mo from the LEU according to the present invention will be explained in greater detail with step by step.

According to one embodiment, step 1 includes heating high-density LEU target containing LEU alloy particles .

As explained above, although many suggestions have been made for the preparation of uranium targets using LEU to deal with limited fabrication and use of uranium targets using HEU, the currently available LEU uranium targets are mostly the uranium aluminide dispersed in aluminum matrix which has low uranium density of approximately 2.6 to 2.7 g-U/cc, and shortcoming such as relatively low yield than HEU target and more waste generation. Additionally, preparing uranium target by dispersing high-density uranium alloy or uranium metal on aluminum matrix also has the problem of low dissolution in alkaline solutions and accordingly, cannot be prepared using a conventional 99Mo extraction process that involves use of alkaline solutions.

According to an embodiment, step 1 includes heating high-density uranium target containing LEU alloy particles for reaction of the uranium alloy particles into uranium aluminide, so that the uranium target containing high density of LEU is dissolved in alkaline solution.

Accordingly, at step 1, the metal uranium particles dispersed in the high-density target are chemically reacted with aluminum by the heating, and as a result, converted into uranium aluminide which is soluble in alkaline solution .

As a result, it is possible to use LEU target containing same level of 235U as HEU, in the conventional alkaline solution processing.

The uranium target of step 1 may include plate type or annular type uranium target as the ones illustrated in FIGS. 1 and 2.

The plate type uranium target includes aluminum matrix with uranium alloy powder dispersion, in which the aluminum matrix and the uranium alloy powder are reacted with each other during heat treatment of step 1, thereby converting uranium alloy into uranium aluminide.

The annular type uranium target includes inner and outer aluminum tubes and uranium alloy foil disposed therebetween, in which the uranium alloy foil and the inner and outer aluminum tubes are reacted with each other during the heat treatment of step 1 to yield uranium aluminide.

However, the uranium target is not limited to the example such as plate type or annular type, and accordingly, ant uranium target can be used as long as the target produces 99Mo by neutron irradiation.

The uranium target of step 1 may have uranium density of 3 g-U/cc or above, or more preferably, 8 g-U/cc or above.

This is higher uranium density compared to the density of the currently-commercialized uranium target using low-enriched uranium aluminide which is approximately 2.7 g-U/cc. Accordingly, 235U density increases, and 99Mo production yield can be improved.

Since the proportional relationship between uranium density and 235U content of the uranium target, the 99Mo production yield can increase as the uranium density increases. However, considering difficulty of fabricating targets containing high density, it is preferred that the uranium density of the uranium target is limited to above 3 g-U/cc, and the uranium target at step 1 has uranium density of 8 g-U/cc or above to have the same level of 235U

content as HEU target, but embodiments are not limited thereto. Considering fabrication process, the uranium target with proper uranium density of 3 g-U/cc or above may be applied.

Meanwhile, the uranium target at step 1 has high density, because the uranium target contains uranium alloy instead of uranium aluminide.

Referring to density of uranium compounds/alloys as Table 1 below, in the case of uranium aluminide (UA1X, 2^x^5), the average uranium density is approximately 4.5 g-U/cc, while in the case of uranium metal (U) , the density is 19.0 g-U/cc which is considerably higher than that of uranium aluminide.

That is, while it is difficult to fabricate high-density target with uranium aluminide which is currently used as the low-enriched uranium target, it is possible to fabricate high-density uranium target by using uranium alloy.

[Table l]

However, as explained above, since uranium target containing uranium alloy is not dissolved in alkaline solution such as NaOH, incompatibility with the generally used alkaline solution process for 99Mo production has to be resolved.

Meanwhile, unlike uranium alloy, uranium aluminide is dissolved in alkaline solution and therefore, appropriate for 99Mo extraction by alkaline solution process .

Accordingly, step 1 may include reacting uranium alloy contained in the uranium target into uranium aluminide by heating, in which the heating of step 1 may be performed at a temperature between 500 and 1200 °C , or preferably, between 700 and 800 °C .

If the heating is done at a temperature lower than 500 °C in step 1, reaction into uranium aluminide occurs slowly, which will reduce the duration of using 99Mo which has short half-life (i.e., 66 hr), while if the heating is done at a temperature exceeding 1200 °C, equipment to carry out such heating is necessary, which is quite difficult.

The heating at step 1 may preferably be performed between 10 min and 12 hr. If the heating is performed for less than 10 min in step 1, only a trace of uranium may react into uranium aluminide, while if the heating is performed for longer than 12 hr, the duration of using 99Mo with short half-life is shortened.

Furthermore, the heating at step 1 may preferably be performed in vacuum or inert atmosphere. Because the uranium target contains fission gas therein, the uranium target is preferably heated within sealed equipment which is in vacuum or inert atmosphere to ensure that fission gas is captured, but the heating conditions of step 1 are not limited to any specific example only.

By the heating at step 1, the low-enriched uranium alloy particles are reacted into uranium aluminide, and depending on the conditions of the heating, 80% or more, or preferably 90% or more uranium alloy contained in the uranium target are reacted into uranium aluminide. That is, most uranium alloy in the uranium target are reacted into uranium aluminide which is soluble in alkaline solution, and as 80% or more uranium alloy is reacted into uranium aluminide, uranium target is dissolved in alkaline solution to yield 99Mo .

Considering solubility in alkaline solution, the complete reaction into uranium aluminide is preferred. However, because it can take a lengthy time depending on heating conditions until uranium is completed reacted, in one embodiment, the rate of reacting uranium alloy into uranium aluminide is set to 80% or above, and the rate may vary within the above-mentioned range depending on the conditions of heating.

If the uranium alloy is reacted into uranium alumnide at a rate that is under the above-mentioned range, it may be difficult to dissolve the uranium target in alkaline solution.

According to one embodiment, a method for extracting radioactive 99Mo from low-enriched uranium target includes step 2 for dissolving in alkaline solution the high-density uranium target which is heated at step 1.

As the heating continues at step 1, the uranium alloy within the uranium target is converted into uranium aluminide. That is, the heating at step 1 causes the uranium alloy insoluble in alkaline solution to change into uranium alminide which is soluble in alkaline solution, and step 2 involves dissolving such high-density uranium target in alkaline solution and thereby preparing alkaline solution containing uranium aluminide and fission products therein .

The process of extracting 99Mo is largely categorized into alkaline solution extraction process and acidic solution extraction process. The acidic solution extraction process has shortcoming such as large amount of waste and disposal of the same. Accordingly, the alkaline solution extraction process is widely used for dissolving target and extracting 99Mo, as this process provides advantage such as solid waste which is relatively smaller in amount, readily storable and convenient to handle.

As a result of step 2, the high-density uranium target heated at step 1 is dissolved in alkaline solution, thus yielding fission product such as 99Mo, etc.

The alkaline solution at step 2 may preferably be NaOH solution, but not limited thereto. Accordingly, any alkaline solution may be used, provided that the solution can dissolve uranium aluminide.

According to one embodiment, a method for extracting radioactive 99Mo from low-enriched uranium target includes step 3 of extracting 99Mo from the solution in which uranium target is dissolved at step 2. The alkaline solution containing uranium target dissolved therein contains fission product such as 99Mo. Step 3 extracts 99Mo dissolved in alkaline solution with extraction method such as, for example, adsorption, chromatography, precipitation separation, or ion exchange, among others .

The method for extracting radioactive 99Mo according to various embodiments generates uranium aluminide which is soluble in alkaline solution, by heating high-density low-enriched uranium target, which is prepared by using uranium alloy, into uranium aluminide which is soluble in alkaline solution, to thus extract 99Mo by using a conventional extraction process using alkaline solution.

Despite concerns over nuclear proliferation and active efforts to produce uranium target using low-enriched uranium, the density of uranium in the currently-commercialized uranium target is not more than approximately 2.7 g-U/cc which cannot yield high amount of 99Mo . Further, while it is possible to fabricate high density target with uranium alloy, such target is not compatible with the conventional process that requires use of alkaline solution process. Further, extracting 99Mo by dissolving target made with uranium alloy in acidic solution has the shortcoming of large amount of waste and difficulty of disposing the waste.

The present invention overcomes the above-mentioned problems occurring in the prior art, by heating high-density low-enriched target made with uranium alloy into uranium aluminide, since it is possible to use the conventional alkaline solution process and also to increase density of uranium which in turn increases production yield of 99Mo. Accordingly, the present invention produces high yield of 99Mo by using the extraction method provided herein, and use of low-enriched uranium and prevention of nuclear proliferation issue.

According to the present invention, the radioactive 99Mo extracted by the extraction method explained above is provided.

99Mo recovered by the extraction method explained above is medical radioactive 99Mo and used for the production of 99mTc which takes up approximately 80% of the medical diagnostic radioactive isotope consumption. This artificial element which is not found in nature is a daughter nuclide produced by 99Mo radioactive decay. Accordingly, the 99Mo according to the present invention is applicable as radioactive 99Mo to produce 99mTc.

The present invention will be explained below with reference to Examples. However, the Examples are provided below only for the purpose of explanation and illustration, and shall not be construed as limiting.

<Example 1> Extraction of 99Mo from high-density low-enriched uranium target

Step 1: Low-enriched uranium (LEU) metal powder was prepared with a raw material of LEU metal ingot by the apparatus for preparing nuclear fuel powder as disclosed in Korean Patent No. 10-279880 by centrifugal atomization. The LEU metal powder was mixed with aluminum powder at a ratio of 84 parts aluminum powder to 16 parts uranium, so that the core of the final dispersion target was 3 g-U/cc. The mixed powder was pressed by a press into a cylindrical compact which was 10 mm in diameter, and 2 mm in height.

The compact was then charged into a pair of plate type aluminum frames in sandwiched manner, rolled at 450 °C three times, so that the frame assembly was reduced from 4 mm to 1.5 mm in thickness. After the rolling, neutron is emitted to the compact to induce fission, and as a result, low-enriched uranium target was prepared.

The uranium target was heated at 700 °C under vacuum atmosphere, for 1 hr, to be reacted into uranium aluminide.

Step 2: The uranium target heated at step 1 was dissolved in sodium hydroxide solution.

Step 3: After separation by alumina column from the sodium hydroxide solution in which uranium target was dissolved therein at step 2, 99Mo was extracted using ammonium hydroxide.

<Example 2> Extraction of 99Mo from high-density low-enriched uranium target 2 99Mo was extracted in the same manner as Example 1, except for a difference that the final dispersion target core was 6 g-U/cc at step 1.

<Example 3> Extraction of 99Mo from high-density low-enriched uranium target 3 99Mo was extracted in the same manner as Example 1, except for a difference that the final dispersion target core was 9 g-U/cc at step 1.

<Example 4> Extraction of 99Mo from high-density low-enriched uranium target 4 99Mo was extracted in the same manner as Example 1, except for a difference that the heating at step 1 was performed at 700 °C, for 2 hr.

<Example 5> Extraction of 99Mo from high-density low-enriched uranium target 5 99Mo was extracted in the same manner as Example 1, except for a difference that the heating at step 1 was performed at 700 °C, for 4 hr.

<Example 6> Extraction of 99Mo from high-density low-enriched uranium target 6 99Mo was extracted in the same manner as Example 2, except for a difference that the heating at step 1 was performed at 700 °C , for 2 hr.

<Example 7> Extraction of 99Mo from high-density low-enriched uranium target 7 99Mo was extracted in the same manner as Example 2, except for a difference that the heating at step 1 was performed at 700 °C, for 4 hr.

<Example 8> Extraction of "Mo from high-density low-enriched uranium target 8 99Mo was extracted in the same manner as Example 3, except for a difference that the heating at step 1 was performed at 700 °C, for 2 hr.

<Example 9> Extraction of 99Mo from high-density low-enriched uranium target 9 99Mo was extracted in the same manner as Example 3, except for a difference that the heating at step 1 was performed at 700 °C, for 4 hr.

The process was performed in the same manner as

Example 1, except for a difference that the heating was omitted at step 1.

Comparative Example 2>

The process was performed in the same manner as

Example 2, except for a difference that the heating was omitted at step 1.

Comparative Example 3>

The process was performed in the same manner as

Example 3, except for a difference that the heating was omitted at step 1.

experimental Example 1> SEM/EDS analysis

The microstructure of LEU target heated at step 1 according to Examples 1 to 3 of the present invention was analyzed under SEM/EDS, and the results were observed as shown in FIGS. 3 and 4.

Referring to FIG. 3, the LEU target, prepared according to Examples 1 to 3, were reacted so that the uranium metal was converted into uranium aluminide (in white in the image) as the heating continued. It was also observed that the target of Example 3, which is relatively denser than the others, produced largest amount of uranium aluminide due to the largest uranium metal content therein.

Further, referring to FIG. 4, as a result of energy-dispersive X-ray spectroscopy (EDS) on the uranium aluminide of the LEU target of Example 1, the uranium aluminide had UAI4.4 composition of uranium (15 at.%), aluminum (84 at.%). Accordingly, it was confirmed that the uranium metal of the high-density low-enriched uranium target was reacted into uranium aluminide according to the extraction method of the present invention.

From the above findings, it was confirmed that the extraction method of the present invention converted uranium alloy into uranium aluminide which is soluble in alkaline solution by heating the LEU target containing uranium alloy, and it was therefore confirmed that 99Mo can be extracted by the alkaline solution process.

The foregoing exemplary embodiments and advantages are merely exemplary and are not to be construed as limiting the present invention. The present teaching can be readily applied to other types of apparatuses. Also, the description of the exemplary embodiments of the present inventive concept is intended to be illustrative, and not to limit the scope of the claims.

In the Figures 1 and 2: 1 is aluminum plate 2 is uranium metal particles 3 is aluminum plate 4 is A1 tube 5 is uranium metal foil 6 is A1 tube.

Claims

A method for extracting radioactive "Mo from low-enriched uranium (LEU) target, which method comprises: heating a high-density LEU target comprising particles of LEU alloy (step 1; dissolving the LEU target with a high density heated in step 1 in an alkaline solution (step 2; and extracting "Mo from the alkaline solution in which the LEU target is dissolved in step 2 (step 3).

The method of claim 1, wherein the LEU target of step 1 comprises uranium alloy target of the platelet or ring type.

The method of claim 1, wherein the LEU target of step 1 contains a uranium density of 3 g-U / cc or more.

The method of claim 1, wherein the LEU target of step 1 contains a uranium density of 8 g-U / cc or more.

The method of claim 1, wherein the heating in step 1 is carried out at a temperature between 500 and 1200 ° C.

The method of claim 1, wherein the heating in step 1 is carried out at a temperature between 700 and 800 ° C.

The method of claim 1, wherein the heating in step 1 is performed for 10 minutes to 12 hours.

The method of claim 1, wherein the heating in step 1 is carried out in vacuum or in an inert atmosphere.

The method of claim 1, wherein the LEU target after heating in step 1 is present in the form of uranium aluminide in a proportion of 80% or more.

The method of claim 1, wherein the alkaline solution in step 2 is a sodium hydroxide solution.

The method of claim 1, wherein extracting "Mo in step 3 is performed in a manner selected from the group consisting of adsorption, chromatography, ion exchange, and precipitation separation.

A radioactive "Mo extracted from LEU target with the method of claim 1.

Use of the radioactive "Mo according to claim 12, for the purpose of medically radioactive" Mo.