Note : Les descriptions sont présentées dans la langue officielle dans laquelle elles ont été soumises.
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PROCESS FOR PRODUCTION OF MOLYBDENUM-99 AND
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.
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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
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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.
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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
16
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
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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
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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
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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
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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.
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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
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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.
