CA2134263A1 - Target for use in the production of molybdenum-99 - Google Patents
Target for use in the production of molybdenum-99Info
- Publication number
- CA2134263A1 CA2134263A1 CA002134263A CA2134263A CA2134263A1 CA 2134263 A1 CA2134263 A1 CA 2134263A1 CA 002134263 A CA002134263 A CA 002134263A CA 2134263 A CA2134263 A CA 2134263A CA 2134263 A1 CA2134263 A1 CA 2134263A1
- Authority
- CA
- Canada
- Prior art keywords
- uranium
- target
- walls
- oxide
- members
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Abandoned
Links
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
- G21G1/00—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes
- G21G1/02—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
- G21G4/00—Radioactive sources
- G21G4/04—Radioactive sources other than neutron sources
- G21G4/06—Radioactive sources other than neutron sources characterised by constructional features
- G21G4/08—Radioactive sources other than neutron sources characterised by constructional features specially adapted for medical application
Abstract
A target is taught for use in the production of Mo-99 from aluminum-free uranium. The target is formed so that the uranium is disposed between a pair of walls and thereby provides for efficient heat transfer during fission of the uranium.
The target can be used in high power reactors where efficient heat transfer is essential.
The target can be used in high power reactors where efficient heat transfer is essential.
Description
Field of the Invention This invention is directed to the production of molybdenum-99 and, in particular a target for production of molybdenum-99.
5 Backqround 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 10 recol1sl,.lct the image of an organ. Tc-99m is prefer,ed 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 Tc99m must be produced just prior to 15 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 dcv010ped which yield a suitable product for use in nuclear medicine. However, current production techniques are complex and time consuming and result in considerable decay 20 losses. In addition, current production techniques create large quantities of high level radioactive liquid waste, thus increasi"g 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.
213~263 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 weight5 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 ll,ererore final product is being lost. Moreover, the presence of 10 dissolvcd aluminum in the solution complicates the separation steps and renders it diffficult 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 cor,esponding large volumes of radioactive waste solution. This 15 is diffficult and expensive to store, and cannot easily be disposed of in a safe way.
Other targ~l~ are known consisting of closed cylinders in which uranium oxide or metal is electroplated about the inner surface. The cylinder ismade from stainless steel or zirconium alloy (zircaloy) and allows for a direct exposure of the irradiated uranium for processing. However, such targets are useful 20 only in low power reactors since heat transfer is a problem at higher powers.
Summar,v of the Invention A target has been invented for production of Mo-99 having effective heat transfer without the use of aluminum and which is suitable for use in high power reactors.
In accordance with a broad aspect of the present invention there is provided a target for the production of Mo-99 comprising: a first outer wall member;
a second outer wall member; and, a layer of aluminum-free uranium or uranium 213~263 oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and the second outer wall members.
In accordance with a further broad aspect of the present invention, there is provided a process for producing a target for the production of Mo-99 5 comprising: loading aluminum-free uranium or uranium oxide between a pair of walls such that the uranium or uranium oxide is in intimate contact with walls, and sealing the uranium or uranium oxide within the walls.
In accordance with a further broad aspect of the present invention, there is provided a target for the production of Mo-99 comprising: a first tubular 10 member; a second tubular member arranged concentrically with the first memberand a layer of aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and second members.
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 perspective, cutaway view of a target according to the present invention; and, Figure 2 shows a perspective, cutaway view of another embodiment of a target according to the present invention; and, Figure 3 shows a flow diagram of a process for using the target of the present invention.
213426~
_ -- 4 --Detailed Descri, liGn of the Present Invention The target of the present invention comprises a first wall member and a second wall member which sandwich a layer of uranium or uranium oxide 5 therebetween. The layer can be in the form of uranium oxides such as, U02 or U3O8, in powder form, uranium metal foil, uranium metal foil oxidized to U02 andelectrodeposited U02 or U3O8. In a preferred embodiment, the uranium or uranium oxide is highly enriched. The outer wall members are in contact with the uraniumor uranium oxide layer such that the target has effective heat transfer during fission.
Referring to Figure 1 there is shown a view of a target 10 according to 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 14 and 16 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 14 and 16 are then sealed such as by welding.
Referring to Figure 2 there is shown a view of another target 110 according to 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 closecontact with at least one member and the target is helium filled to provide for heat 21~4263 l,al1sfer. 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 the target is increased. The outer wall members are adapted to facilitate exposure 5 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 2, 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 9 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 isdisposed 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 about0.10 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 compressthe 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. Theuranium oxide is electrodeposited over the surface of the wall members such thatit 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 5 electrodeposition is affected by any known method 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 is elect,odeposited to suitable thicknesses by use of current of 0.9 amperes, 1.5 volts and a temperature of about 93C. The wall member 10 having the ele~,tlodeposited layer thereon is then heated to 500C.
After electrodeposition, the walls are dipped in nitric acid to remove a portion of the uranium oxide such that a portion of the wall is exposed for sealing the target. The walls are then positioned in close relation and sealed at the edges.
The walls are then pressed such as by rolling or swaging or, alternatively, the space 15 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 20 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 1, an annular assembly, as shown in Figure 2, or other suitable shapes that provide for direct heat l,ansrer from the uranium or uranium oxide through the walls 25 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.
213~263 _ -- 7 -Referring to Figure 3, 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 5 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 according to the present invention or, alternatively, any suitable target containing uranium or uranium oxide which is substantially free of 10 aluminum. 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 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 15 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 16N nitric acid are required to dissolve each gram of U-235, depending on the form U-235 with powder forms of uranium oxide requiring the least amount of nitric acid.
20 Where it is necess~ry 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 critical and should be optimized on a cost benefit analysis in terms of amount dissolved versus 25 time spent. Gases rele~sed during exposure of the uranium or uranium oxide layer and dissolution thereof are collected for off-gas treatment. In an embodiment, the target is punctured to release fission products such as Xe-133 and 1-131 prior to target decladding and dissolution.
213~26~
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.5 The alumina column useful in the preferred method is prepared by dissolving aluminum oxide in 1N 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 p~ssed to a process wherein it is converted to solid uranium oxide. The process includes de-watering, such as for example, by boiling, and heating to about 500 C in the presence of oxygen to allow oxidation and calcindlion. In a preferred embodiment, suitable time is provided prior to 15 evaporation for decay of isotopes having a short-half life.
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 500C in the presence until solid uranium oxide and calx thereof is 20 formed. Alternately, the waste solution is passecl directly to a calciner where evaporation and crllcin~liGn are combined.
Any suitable calciner can be used such as an in-pot calciner where temperatures will be increased from 400C to 650C, or a rotary calciner where calcination can be affected at temperatures of 400C to 500C. Waste in the form25 of stable, ceramic-like uranium oxide calx is obtained by the process and is suitable for long term storage in sealed canisters.
Examples Four targets, generally as shown in Figure 2, containing 18.5 9 of U-235/target in the form of highly enriched uranium oxide powder were irradiated for 10 days at a target power of 60.7 kW. Similarly, sixteen targets containing 2.4 9 of U-235/target in the form of aluminum-uranium alloy (79% Al, 21% U) were irradiated 5 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 l,ealed with 2 N nitricacid until completely dissolved. The targets containing aluminum-uranium alloy was dissolved in 2 N nitric acid containing Hg(NO3)2 until completely dissolved. The10 resulting solutions were passed though a 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 1.
Table I
Plocess Parameters Al-U UO2 Target power (kW/g U-235) 6.46 3.28 Mo-99 yield from irradiation (Ci/g U-235) 229 140 Plocess 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 18 Volume of calcined waste/g U-235 (ml) 13 1.3 The target of the offers faster process time over previous Al-containing targets. Thus, less Mo-99 is lost due to decay during processing.
It will be apparent that many other change-s may be made to the 25 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.
5 Backqround 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 10 recol1sl,.lct the image of an organ. Tc-99m is prefer,ed 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 Tc99m must be produced just prior to 15 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 dcv010ped which yield a suitable product for use in nuclear medicine. However, current production techniques are complex and time consuming and result in considerable decay 20 losses. In addition, current production techniques create large quantities of high level radioactive liquid waste, thus increasi"g 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.
213~263 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 weight5 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 ll,ererore final product is being lost. Moreover, the presence of 10 dissolvcd aluminum in the solution complicates the separation steps and renders it diffficult 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 cor,esponding large volumes of radioactive waste solution. This 15 is diffficult and expensive to store, and cannot easily be disposed of in a safe way.
Other targ~l~ are known consisting of closed cylinders in which uranium oxide or metal is electroplated about the inner surface. The cylinder ismade from stainless steel or zirconium alloy (zircaloy) and allows for a direct exposure of the irradiated uranium for processing. However, such targets are useful 20 only in low power reactors since heat transfer is a problem at higher powers.
Summar,v of the Invention A target has been invented for production of Mo-99 having effective heat transfer without the use of aluminum and which is suitable for use in high power reactors.
In accordance with a broad aspect of the present invention there is provided a target for the production of Mo-99 comprising: a first outer wall member;
a second outer wall member; and, a layer of aluminum-free uranium or uranium 213~263 oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and the second outer wall members.
In accordance with a further broad aspect of the present invention, there is provided a process for producing a target for the production of Mo-99 5 comprising: loading aluminum-free uranium or uranium oxide between a pair of walls such that the uranium or uranium oxide is in intimate contact with walls, and sealing the uranium or uranium oxide within the walls.
In accordance with a further broad aspect of the present invention, there is provided a target for the production of Mo-99 comprising: a first tubular 10 member; a second tubular member arranged concentrically with the first memberand a layer of aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and second members.
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 perspective, cutaway view of a target according to the present invention; and, Figure 2 shows a perspective, cutaway view of another embodiment of a target according to the present invention; and, Figure 3 shows a flow diagram of a process for using the target of the present invention.
213426~
_ -- 4 --Detailed Descri, liGn of the Present Invention The target of the present invention comprises a first wall member and a second wall member which sandwich a layer of uranium or uranium oxide 5 therebetween. The layer can be in the form of uranium oxides such as, U02 or U3O8, in powder form, uranium metal foil, uranium metal foil oxidized to U02 andelectrodeposited U02 or U3O8. In a preferred embodiment, the uranium or uranium oxide is highly enriched. The outer wall members are in contact with the uraniumor uranium oxide layer such that the target has effective heat transfer during fission.
Referring to Figure 1 there is shown a view of a target 10 according to 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 14 and 16 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 14 and 16 are then sealed such as by welding.
Referring to Figure 2 there is shown a view of another target 110 according to 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 closecontact with at least one member and the target is helium filled to provide for heat 21~4263 l,al1sfer. 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 the target is increased. The outer wall members are adapted to facilitate exposure 5 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 2, 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 9 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 isdisposed 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 about0.10 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 compressthe 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. Theuranium oxide is electrodeposited over the surface of the wall members such thatit 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 5 electrodeposition is affected by any known method 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 is elect,odeposited to suitable thicknesses by use of current of 0.9 amperes, 1.5 volts and a temperature of about 93C. The wall member 10 having the ele~,tlodeposited layer thereon is then heated to 500C.
After electrodeposition, the walls are dipped in nitric acid to remove a portion of the uranium oxide such that a portion of the wall is exposed for sealing the target. The walls are then positioned in close relation and sealed at the edges.
The walls are then pressed such as by rolling or swaging or, alternatively, the space 15 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 20 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 1, an annular assembly, as shown in Figure 2, or other suitable shapes that provide for direct heat l,ansrer from the uranium or uranium oxide through the walls 25 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.
213~263 _ -- 7 -Referring to Figure 3, 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 5 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 according to the present invention or, alternatively, any suitable target containing uranium or uranium oxide which is substantially free of 10 aluminum. 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 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 15 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 16N nitric acid are required to dissolve each gram of U-235, depending on the form U-235 with powder forms of uranium oxide requiring the least amount of nitric acid.
20 Where it is necess~ry 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 critical and should be optimized on a cost benefit analysis in terms of amount dissolved versus 25 time spent. Gases rele~sed during exposure of the uranium or uranium oxide layer and dissolution thereof are collected for off-gas treatment. In an embodiment, the target is punctured to release fission products such as Xe-133 and 1-131 prior to target decladding and dissolution.
213~26~
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.5 The alumina column useful in the preferred method is prepared by dissolving aluminum oxide in 1N 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 p~ssed to a process wherein it is converted to solid uranium oxide. The process includes de-watering, such as for example, by boiling, and heating to about 500 C in the presence of oxygen to allow oxidation and calcindlion. In a preferred embodiment, suitable time is provided prior to 15 evaporation for decay of isotopes having a short-half life.
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 500C in the presence until solid uranium oxide and calx thereof is 20 formed. Alternately, the waste solution is passecl directly to a calciner where evaporation and crllcin~liGn are combined.
Any suitable calciner can be used such as an in-pot calciner where temperatures will be increased from 400C to 650C, or a rotary calciner where calcination can be affected at temperatures of 400C to 500C. Waste in the form25 of stable, ceramic-like uranium oxide calx is obtained by the process and is suitable for long term storage in sealed canisters.
Examples Four targets, generally as shown in Figure 2, containing 18.5 9 of U-235/target in the form of highly enriched uranium oxide powder were irradiated for 10 days at a target power of 60.7 kW. Similarly, sixteen targets containing 2.4 9 of U-235/target in the form of aluminum-uranium alloy (79% Al, 21% U) were irradiated 5 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 l,ealed with 2 N nitricacid until completely dissolved. The targets containing aluminum-uranium alloy was dissolved in 2 N nitric acid containing Hg(NO3)2 until completely dissolved. The10 resulting solutions were passed though a 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 1.
Table I
Plocess Parameters Al-U UO2 Target power (kW/g U-235) 6.46 3.28 Mo-99 yield from irradiation (Ci/g U-235) 229 140 Plocess 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 18 Volume of calcined waste/g U-235 (ml) 13 1.3 The target of the offers faster process time over previous Al-containing targets. Thus, less Mo-99 is lost due to decay during processing.
It will be apparent that many other change-s may be made to the 25 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 (25)
1. A target for the production of Mo-99 comprising:
a first outer wall member;
a second outer wall member; and, a layer of substantially aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and the second outer wall members.
a first outer wall member;
a second outer wall member; and, a layer of substantially aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and the second outer wall members.
2. The target of claim 1 wherein the uranium is uranium metal.
3. The target of claim 2 wherein the uranium metal is oxidized.
4. The target of claim 1 wherein the uranium oxide is a powder and is compressed between the tubular members by swaging.
5. The target of claim 1 wherein the uranium oxide is electrodeposited onto at least one of the members.
6. The target of claim 1 wherein the members are formed of zirconium alloy.
7. A target for the production of Mo-99 comprising:
a first tubular member;
a second tubular member arranged concentrically with the first member; and, a layer of substantially aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and second members.
a first tubular member;
a second tubular member arranged concentrically with the first member; and, a layer of substantially aluminum-free uranium or uranium oxide disposed therebetween, such that heat produced by fission of the uranium or uranium oxide is transferred directly to the first and second members.
8. The target of claim 7 wherein the uranium is uranium metal.
9. The target of claim 8 wherein the uranium metal is oxidized.
10. The target of claim 7 wherein the uranium oxide is a powder and is compressed between the tubular members by swaging.
11. The target of claim 7 wherein the uranium oxide is electrodeposited onto at least one of the members.
12. The target of claim 7 wherein the members are formed of zirconium alloy.
13. A process for producing a target for the production of Mo-99 comprising:
loading substantially aluminum-free uranium or uranium oxide between a pair of walls such that the uranium or uranium oxide is in intimate contact with at least one of the walls, and sealing the uranium or uranium oxide within the walls.
loading substantially aluminum-free uranium or uranium oxide between a pair of walls such that the uranium or uranium oxide is in intimate contact with at least one of the walls, and sealing the uranium or uranium oxide within the walls.
14. The process of claim 13 wherein the uranium oxide is a powder and is loaded between the walls by means of vibration packing.
15. The process of claim 14 wherein the walls are compressed about the uranium oxide powder.
16. The process of claim 13 wherein the uranium oxide is electrodeposited onto at least one of the members.
17. The process of claim 16 wherein the walls are compressed about the uranium oxide.
18. The process of claim 13 further comprising helium-filling the target.
19. The process of claim 13 wherein the uranium is uranium metal.
20. The process of claim 19 wherein the uranium metal is oxidized.
21. The process of claim 13 wherein the walls are formed of zirconium alloy.
22. The process of claim 21 wherein the walls are formed as plates.
23. The process of claim 21 wherein the walls are formed as tubular members and are arranged concentrically.
24. The process of claim 13 wherein the uranium is sealed within the walls by means of end caps.
25. The process of claim 13 wherein the uranium is sealed within the walls by means of welding the walls together at their edges.
Priority Applications (4)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CA002134263A CA2134263A1 (en) | 1994-10-25 | 1994-10-25 | Target for use in the production of molybdenum-99 |
| PCT/CA1995/000332 WO1996013038A1 (en) | 1994-10-25 | 1995-06-07 | Target for use in the production of molybdenum-99 |
| AU25591/95A AU2559195A (en) | 1994-10-25 | 1995-06-07 | Target for use in the production of molybdenum-99 |
| ZA958981A ZA958981B (en) | 1994-10-25 | 1995-10-24 | Target for use in the production of molybdenum-99 |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CA002134263A CA2134263A1 (en) | 1994-10-25 | 1994-10-25 | Target for use in the production of molybdenum-99 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CA2134263A1 true CA2134263A1 (en) | 1995-10-13 |
Family
ID=4154517
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA002134263A Abandoned CA2134263A1 (en) | 1994-10-25 | 1994-10-25 | Target for use in the production of molybdenum-99 |
Country Status (4)
| Country | Link |
|---|---|
| AU (1) | AU2559195A (en) |
| CA (1) | CA2134263A1 (en) |
| WO (1) | WO1996013038A1 (en) |
| ZA (1) | ZA958981B (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US6160862A (en) * | 1993-10-01 | 2000-12-12 | The United States Of America As Represented By The United States Department Of Energy | Method for fabricating 99 Mo production targets using low enriched uranium, 99 Mo production targets comprising low enriched uranium |
Families Citing this family (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| RU2494484C2 (en) * | 2008-05-02 | 2013-09-27 | Шайн Медикал Текнолоджис, Инк. | Production device and method of medical isotopes |
| US9431138B2 (en) * | 2009-07-10 | 2016-08-30 | Ge-Hitachi Nuclear Energy Americas, Llc | Method of generating specified activities within a target holding device |
| AU2015200445B2 (en) * | 2010-07-29 | 2016-11-03 | The State Of Oregon Acting By And Through The State Board Of Higher Education On Behalf Of Oregon State University | Isotope production target |
| WO2012015974A1 (en) * | 2010-07-29 | 2012-02-02 | The State Of Oregon Acting By And Through The State Board Of Higher Education On Behalf Of Oregon State University | Isotope production target |
| CN114540828A (en) * | 2022-03-23 | 2022-05-27 | 中国原子能科学研究院 | Method for electrodepositing uranium on metal surface |
| CN115449764B (en) * | 2022-09-14 | 2023-09-01 | 中国工程物理研究院材料研究所 | Actinide alloy gradient film and preparation method thereof |
Family Cites Families (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3940318A (en) * | 1970-12-23 | 1976-02-24 | Union Carbide Corporation | Preparation of a primary target for the production of fission products in a nuclear reactor |
| US4839133A (en) * | 1987-10-26 | 1989-06-13 | The United States Of America As Represented By The Department Of Energy | Target and method for the production of fission product molybdenum-99 |
-
1994
- 1994-10-25 CA CA002134263A patent/CA2134263A1/en not_active Abandoned
-
1995
- 1995-06-07 WO PCT/CA1995/000332 patent/WO1996013038A1/en active Application Filing
- 1995-06-07 AU AU25591/95A patent/AU2559195A/en not_active Abandoned
- 1995-10-24 ZA ZA958981A patent/ZA958981B/en unknown
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US6160862A (en) * | 1993-10-01 | 2000-12-12 | The United States Of America As Represented By The United States Department Of Energy | Method for fabricating 99 Mo production targets using low enriched uranium, 99 Mo production targets comprising low enriched uranium |
Also Published As
| Publication number | Publication date |
|---|---|
| AU2559195A (en) | 1996-05-15 |
| WO1996013038A1 (en) | 1996-05-02 |
| ZA958981B (en) | 1996-05-23 |
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| EEER | Examination request | ||
| FZDE | Dead |