CA2134264C

CA2134264C - Process for production of molybdenum-99 and management of waste therefrom

Info

Publication number: CA2134264C
Application number: CA002134264A
Authority: CA
Inventors: William T. Hancox; Jean-Pierre Labrie; Richard J. Harrison; Deonaraine Singh
Original assignee: Individual
Current assignee: Atomic Energy of Canada Ltd AECL
Priority date: 1994-10-25
Filing date: 1994-10-25
Publication date: 2000-09-12
Anticipated expiration: 2014-10-25
Also published as: WO1996013039A1; CA2134264A1; AU2559295A; ZA958980B

Abstract

A process is taught for production of Mo-99 from uranium-235 and for managing waste produced therefrom. The process includes separation of the Mo-99 from irradiated uranium and solidification of the waste remaining after such separation. Because of the great reduction in the amounts of waste produced fromthe process over previous methods, the process is of particular use in the large-scale production of Mo-99. Further, the present process results in a reduced process time over previous processes and thereby offers a reduction in the loss of Mo-99 by decay.

Description

~13~~64 MANAGEMENT OF WASTE THEREFROM
Field of the Invention This invention is directed to the production of molybdenum-99 and, in particular a process for production of molybdenum-99.
Background of the Invention Molybdenum-99 (Mo-99) is the parent nucleus to technetium-99m (Tc-99m). Tc-99m is used in nuclear medicine for liver, kidney, lung, blood pool, thyroid and tumour scanning. Tc-99m decays to a stable isotope, technetium-99, emitting a low energy gamma ray which can be detected outside the body and used to reconstruct the image of an organ. Tc-99m is preferred over many other radio isotopes for nuclear medicine because of its short half-life of approximately 6 hours which results in reduced radiation exposure of organs relative to the exposure given by most other imaging radio isotopes.
Because of its short half-life Tc-99m must be produced just prior to administration. Tc-99m can be produced from its parent nucleus Mo-99 which has a half-life of approximately 66 hours. Mo-99 is produced by nuclear fission of uranium-235 (U-235). Production techniques for Mo-99 have been developed which yield a suitable product for use in nuclear medicine. However, current production techniques are complex and time consuming and result in considerable decay losses. In addition, current production techniques create large quantities of high level radioactive liquid waste, thus increasing production costs and reducing the suitability of such processes for large scale commercial production of Mo-99.
A
process for production of Mo-99 is required which reduces the amount of waste produced.

A target for use in Mo-99 production having high heat transfer will allow irradiation at high fluxes so that a high rate of fission is obtained. Targets having high heat transfer have been proposed incorporating uranium embedded in an aluminum matrix typically containing 79% by weight of aluminum and 21 % by weight of uranium. However, the use of aluminum in the target presents serious disadvantages in the production of Mo-99. The need to dissolve the aluminum matrix in order to obtain the uranium requires a considerable period of time, adding several hours to the production process. During this time, the radioactive materials are decaying and therefore final product is being lost. Moreover, the presence of dissolved aluminum in the solution complicates the separation steps and renders it difficult to obtain pure products. Mercury is required as a catalyst in the process to remove aluminum. Mercury is of course toxic, and thereby adds to process hazard.
The relatively high volume of solution needed for dissolution of the large mass of aluminum results in corresponding large volumes of radioactive waste solution.
This is difficult and expensive to store, and cannot easily be disposed of in a safe way.
Other targets are known consisting of closed cylinders in which uranium oxide or metal is electroplated about the inner surface. The cylinder is made from stainless steel or zirconium alloy (zircaloy) and allows for a direct exposure of the irradiated uranium for processing. However, such targets are useful only in low power reactors since heat transfer is a problem at higher powers.
Summary of the Invention A process has been invented for production of Mo-99 from uranium which is suitable for large-scale operation and provides a means for waste management.
In accordance with a broad aspect of the present invention, there is provided a process for producing Mo-99 comprising: irradiating a target containing aluminum-free uranium or uranium oxide, extracting the irradiated uranium or 2.134264 uranium oxide by dissolution in a nitric acid solution, separating the Mo-99 from the nitric acid solution, evaporating the solution and calcining to form solid uranium oxide.
Brief Description of the Drawings A further, detailed, description of the invention, briefly described above, will follow by reference to the following drawings of specific embodiments of the invention, which depict only typical embodiments of the invention and are therefore not to be considered limiting of its scope. In the drawings:
Figure 1 shows a flow diagram of a process according to the present invention; and, Figure 2 shows a perspective, cutaway view of a target useable in the inventive process; and, Figure 3 shows a perspective, cutaway view of another target useable in the inventive process.
Detailed Description of the Present Invention The process of the present invention comprises a process for producing Mo-99 which comprises irradiating a target containing substantially aluminum-free uranium or uranium oxide containing a portion of U-235, dissolving the uranium or uranium oxide in a nitric acid solution and separating the Mo-99 from the nitric acid solution. The waste remaining after the separation is managed by evaporating the solution and calcining to form solid uranium oxide.
Referring to Figure 1, a flow diagram of a preferred process for production of Mo-99 and management of the waste produced therefrom is shown.

Steps 1 to 4 pertain to the irradiation of uranium oxide and recovery of Mo-99.
Steps 5 to 8 pertain to a process for management of a waste stream after Mo-99 recovery.
Mo-99 is produced by placement of a target containing uranium-235 into the irradiation zone of a nuclear reactor, particle generator or neutron particle source. The target can be any suitable target containing uranium or uranium oxide which is substantially free of aluminum, or alternatively, a target as described herein.
After a suitable period of irradiation, such as up to about 21 days, the target is removed and cooled for a suitable period such as, for example, for 2 to 16 hours.
The target is punctured to release off gases such as Xe-133 and I-131 such gases are collected and retained for decay. However, such puncturing is not essential to the process, since gases can be collected during decladding.
The Mo-99 is recovered by a process comprising opening the target to expose the uranium and dissolving the uranium or uranium oxide in nitric acid solution. Dissolution requires at least stoichiometric equivalents of nitric acid for each gram of uranium-235 irradiated. However, this may be increased depending on the form of uranium or uranium oxide used. For example, 5 to 40 ml of 2 to N nitric acid are required to dissolve each gram of U-235, depending on the form of U-235, with powder forms of uranium oxide requiring the least amount of nitric acid.
Where it is necessary to submerge the target, amounts greater than this may be required. To reduce the amount of waste produced the volume of acid used should be as little as conveniently possible to provide dissolution. Immersion in the acid is maintained until the layer is dissolved. The time for dissolution is not rritir~l ~~a should be optimized on a cost benefit analysis in terms of amount dissolved versus time spent. Gases released during exposure of the uranium or uranium oxide layer and dissolution thereof are collected for off gas treatment.
After the uranium or uranium oxide has dissolved, the target is removed from the acid solution and is managed as low level waste. Mo-99 is recovered from the acid solution by contacting with an adsorbent. In an embodiment, the acid solution is passed at least once through an alumina column.
The alumina column useful in the preferred method is prepared by dissolving aluminum oxide in 1 N nitric acid to form a slurry. A column packed with 150 ml to 250 ml of wet aluminum oxide is sufficient to absorb 100 to 2000 six day Ci of Mo-99. The alumina column containing adsorbed Mo-99 is passed to treatment for removal of Mo-99.
After recovery of Mo-99, waste acid solution remains which contains uranium nitrite. Such waste is passed to a process wherein it is converted to solid uranium oxide. In a preferred embodiment, the waste acid solution is allowed time for decay of short half-life radioisotopes. However, this step is not essential to the process and the waste can move directly to a process including de-watering, such as for example, by boiling, and heating to about 500°C in the presence of oxygen to allow oxidation and calcination.
In one embodiment, waste solution is passed to an evaporation cell, wherein it is boiled to remove the water, and then to a calciner where it is further heated to about 500°C in the presence until solid uranium oxide and calx thereof is formed. Alternately, the waste solution is passed directly to a calciner where the steps of evaporation and calcination are combined.
Any suitable calciner can be used such as an in-pot calciner where temperatures will be increased from 400°C to 650°C, or a rotary calciner where calcination can be affected at temperatures of 400°C to 500°C.
Waste in the form of stable, ceramic-like uranium oxide calx is obtained by the process and is suitable for long term storage in sealed canisters.
A target, having high heat transfer, which is useful in the present invention comprises a first wall member and a second wall member which sandwich n a layer of uranium or uranium oxide therebetween. The layer can be in the form of uranium oxides such as, U02 or 0308, in powder form, uranium metal foil, uranium metal foil oxidized to U02 or electrodeposited U02 or U308. In a preferred embodiment, the uranium or uranium oxide is highly enriched. The outer wall members are in contact with the uranium or uranium oxide layer such that the target has effective heat transfer during fission.
Referring to Figure 2 there is shown a view of a target 10 useful in the present invention, cutaway to reveal its inner contents. Target 10 comprises a first wall member 12, a second wall member 14 and a layer of uranium 16 therebetween.
Wall members 12 and 14.are rolled to be in intimate contact with layer 16 to provide for effective heat transfer and to stabilize the uranium within the target.
Edges 17 of wall members 12 and 14 are then sealed such as by welding.
Referring to Figure 3 there is shown a view of another target 110 useful in the present invention. Target 110 comprises an inner wall member 112, an outer wall member 114 and a layer of uranium oxide 116 therebetween. End caps 118 are provided to seal a gap formed between the wall members 112, 114 during loading of the uranium oxide.
Wall members are produced from any suitable material for use in nuclear reactor environments, such as, for example zirconium alloy. Stainless steel can be used but is not preferred because of its high neutron absorption when compared to zirconium alloy. To provide close contact between the uranium or uranium oxide layer and the wall members and, thereby, effective heat transfer during fission, the members are preferably compressed about the layer, such as by rolling or swaging. In an alternate embodiment, the uranium or uranium oxide is in close contact with at least one member and the target is helium filled to provide for heat transfer. However, it is to be noted that helium filling provides good heat transfer across small gaps, such as less than about 1 mm. Heat transfer by means of helium filling is diminished substantially as the space between the wall members of _7_ the target is increased. The outer wall members are adapted to facilitate exposure and dissolution of the layer after irradiation. For example, where uranium foil is used, the zirconium alloy surfaces are anodized prior to application of the foil to facilitate removal of the foil after irradiation.
The uranium or uranium oxide is loaded between the wall members in a thin layer and in an amount to give the desired power level such as for example about 100 mg/cm2 and, thereby, the desired Mo-99 production. In a preferred embodiment, an annular target, generally as shown in Figure 3, is 470 mm in length having an inner diameter of 13 mm and an outer diameter of 15 mm and has loaded therein about 20 g of uranium oxide.
In an embodiment, uranium oxide in the form of a finely divided powder is vibration packed into an annular gap formed between the wall members. In an another embodiment, a film of uranium oxide is electrodeposited onto the wall members. In still another embodiment, uranium metal or oxidized uranium metal is disposed between the wall members.
To produce a target having a packed powder layer of uranium oxide, the wall members are positioned such that a uniform annular gap of between about 0.10 and 0.20 mm is formed between the members. The edges of the wall members are sealed to contain the powder, such as by insertion of end caps or welding, and the powder is vibration packed into the gap such as, for example, by use of a Syntron vibrator. The outer walls are then rolled or swaged to compress the uranium oxide to the desired density of about 6.5 to 11 g/cm3 and to cause the wall members to be in intimate contact with the uranium oxide.
A target is produced using electrodeposition by first washing one or both wall members in preparation for electrodeposition of the uranium oxide.
The uranium oxide is electrodeposited over the surface of the wall members such that it will be disposed between the wall members in the assembled target and such that _$_ a total amount of about 100 mg/cm2 is disposed between the walls. Such electrodeposition is affected by any known r~iethod suitable for uranium loading. For example, the uranium oxide can be electrodeposited by use of a bath containing 0.042 M uranyl nitrate and 0.125 M ammonium oxalate, the pH being adjusted to 7.2 with NH40H. Uranium oxide is electrodeposited to suitable thicknesses by use of current of 0.9 amperes, 1.5 volts and a temperature of about 93°C. The wall member having the electrodeposited layer thereon is then heated to 500°C.
After electrodeposition the edges of the walls are dipped in nitric acid to remove a portion of the uranium oxide to allow for sealing. The walls are positioned in close relation, such that the space between the walls is preferably less than 0.2 mm, and sealed at the edges. The walls are then pressed such as by rolling or swaging or, alternatively, the space between the walls is helium filled, to provide for good heat transfer.
A target having uranium metal or oxidized uranium metal foil therein is prepared by placing the foil between wall members which have, preferably, been anodized. The members are then rolled or swaged to provide intimate contact between the metal and the walls. The edges are sealed by any suitable means such as by welding.
The target can be of any suitable shape which will allow heat transfer through each wall member such as, for example, a plate assembly, as shown in Figure 2, an annular assembly, as shown in Figure 3, or other suitable shapes that provide for direct heat transfer from the uranium or uranium oxide through the walls to a heat sink or cooling fluid. As an example, some targets generally as described in relation to Figure 3, have been successfully irradiated at target powers of 18.2 kW/g of U-235.

_g_ Examples Four targets, generally as shown in Figure 3, containing 18.5 g of U-235/target in the form of highly enriched uranium oxide powder were irradiated for days at a target power of 60.7 kW. Similarly, sixteen pencil targets containing 5 2.4 g of U-235/target in the form of aluminum-uranium alloy (79% AI, 21 % U) were irradiated for 10 days at 15.5 kW.
After irradiation, the targets were cooled and processed to recover Mo-99. The targets containing uranium oxide were opened and treated with 2 N
nitric acid until completely dissolved. The targets containing aluminum-uranium alloy was 10 dissolved in 2 N nitric acid containing Hg(N03)2 until completely dissolved. The resulting solutions were passed though -an alumina column to recover the Mo-99.
Liquid waste remaining after the recovery was allowed decay time followed by evaporation and calcination. Results are shown in Table I.
Table I
Process Parameters AI-U
Target power (kW/g U-235) 6.46 3.28 Mo-99 yield from irradiation (Ci/g U-235) 229 140 Process time (hours) 28.5 21.0 Mo-99 yield from processing (Ci/g U-235) 153 101 Volume liquid waste/g U-235 (ml) 200 1 g Volume of calcined waste/g U-235 (ml) 13 1.3 The process of the present invention is simplified over previous processes and offers faster process time. Thus, less Mo-99 is lost due to decay during the process. In addition, the process provides a great reduction in the amount of waste produced by the process over other processes and such waste is in stable form for long-term storage.
It will be apparent that many other changes may be made to the illustrative embodiments, while falling within the scope of the invention and it is intended that all such changes be covered by the claims appended hereto.

Claims

1. A process for producing Mo-99 comprising: irradiating a target containing aluminum-free uranium or uranium oxide whereby to create an admixture of uranium oxide or uranium and Mo-99, dissolving said admixture in a nitric acid solution, separating the Mo-99 from the nitric acid solution, evaporating the solution and calcining the residue to form solid uranium oxide.

2. The process as defined in claim 1, wherein the Mo-99 is separated from the nitric acid solution by adsorption on an alumina column.

3. The process as defined in claim 1, wherein the calcining comprises heating at 400 - 650°C in the presence of oxygen.

4. The process as defined in claim 1, wherein the calcining comprises heating at 400 - 500°C in the presence of oxygen.

5. The process as defined in claim 1 further comprising providing time for decay of radioactive isotopes from the nitric acid solution prior to the evaporation of the nitric acid solution.

6. The process as defined in clam 1, wherein the target comprises a first outer wall member and a second outer wall member having a layer of the uranium disposed therebetween, such that heat produced by fission of the uranium is transferred directly to the first and the second outer wall members.