Process for separating fission product molybdenum from an irradiated target material

A process for the separation and collection of molybdenum-99 from an irradiated uranium-containing target material utilizes thermal chromatographic separation. The irradiated target material containing the molybdenum-99 is heated in an oxidizing atmosphere to form an oxidized target material and gaseous molybdenum-99 trioxide. The gaseous molybdenum-99 trioxide is carried by the oxidizing atmosphere along with other vaporized materials to a cooling zone for progressive condensation and collection of the molybdenum-99 trioxide and the other materials in the form of separate deposits.

BACKGROUND OF THE INVENTION 
This invention relates to a process for extracting and separating fission 
product molybdenum from an irradiated uranium-containing target material. 
Technetium-99m is an extremely useful tool for medical applications and 
diagnosis. This radioisotope is used in a variety of applications in 
medical diagnosis. It is well suited for brain, thyroid, liver, lung blood 
pool and tumor scanning. It is preferred over the other radioisotopes 
because of the selective uptake by specific organs, its short half-life 
and low radiation dose rate which reduces the exposure of the patient to 
radiation. In addition, technetium-99m can also be used in industrial 
applications, such as measurement of flow rates, process control and the 
like. 
Since the radioisotope technetium-99m has a short half-life (6 hours), it 
is common practice to use a molybdenum-99-technetium-99m generator to 
provide a supply of the technetium-99m. Basically, such a generator is 
made by sorbing the molybdenum-99 parent radioisotope, which has a 66-hour 
half-life, on an anion exchange material such as alumina. Subsequent decay 
of the molybdenum-99 produces the technetium-99m which can be selectively 
separated as needed from the generator by elution with a saline solution 
(which yields the technetium-99m as sodium pertechnetate). 
Conventionally, technetium generators have been charged with molybdenum-99 
which has been obtained by neutron bombardment of natural molybdenum or 
enriched molybdenum-98. A minor proportion of the molybdenum-98 is 
converted to its radioisotope molybdenum-99 by neutron capture. 
Radioactive molybdenum so prepared is referred to as "neutron product 
molybdenum". 
Molybdenum-99 can also be obtained as a fission product from neutron 
bombardment of uranium-235. This "fission product molybdenum" is available 
in the form of sodium molybdate or ammonium molybdate solutions having a 
much higher specific activity than the maximum attainable from neutron 
product molybdenum. By using the more highly active fission product 
molybdenum for loading a technetium generator, a generator can be prepared 
which yields sodium pertechnetate eluates of exceptionally high 
radioactivity and which are consequently very desirable for certain 
medical applications. 
The production of fission product molybdenum usually is done using a target 
comprising a mixture of aluminum and uranium, generally about 75 percent 
aluminum by weight with the balance being uranium highly enriched with 
uranium-235. This production requires an extraction process for extracting 
the molybdenum-99 from the target. The process currently used for 
extracting fission product molybdenum (molybdenum-99) from an irradiated 
aluminum-uranium target involves dissolving it in a strongly alkaline 
aqueous solution, such as sodium hydroxide. The resulting solution is 
filtered to separate the uranium and insoluble fission products from the 
molybdenum-bearing alkaline solution. The solution is acidified with 
sulfuric acid, and molybdenum-99 is separated from the solution by 
extraction with bis (2-ethyl-hexyl) phosphoric acid. This yields a 
molybdenum-99 containing extract from which residual soluble fission 
products are removed by stripping with organic solvents. This process is 
effective in extracting the molybdenum-99 from the irradiated target, but 
it produces a large volume of highly radioactive liquid waste. For 
example, approximately 33 liters of liquid waste is generated in the 
production of 4000 Ci of molybdenum-99. This presents a costly liquid 
waste disposal problem necessitating the use of difficult handling 
procedures. Accordingly, it is desirable to provide a process for 
extracting the molybdenum-99 from the target material and simultaneously 
separating the molybdenum-99 from the other fission products in a manner 
avoiding the creation of large volumes of liquid wastes. 
Various processes have been devised for the separation of fission products 
from an irradiated target material. In Nuclear Science and Engineering, 
Volume 29, Number 2, August, 1967, at pages 159-164, A. W. Castleman, Jr. 
and I. N. Tang describe the results of an experimental study of the 
behavior of fission products released into helium and air from target 
materials of metallic uranium and a uranium-3.5% molybdenum alloy. This 
experiment used a thermochromatographic technique for investigating the 
nature of low-concentration gas species and this was applied to the 
release from these target materials of fission products barium, lanthanum, 
cerium, and molybdenum. Barium, lanthanum, and cerium were found not to be 
released into air in significant quantities because of the low volatility 
of their respective oxides, but molybdenum was released as MoO.sub.3 in 
air. The same authors also report the chemical nature of fission products 
iodine and cesium vaporized from irradiated specimens into helium and air 
in the Journal of Inorganic & Nuclear Chemistry, Volume 32, Number 4, 
April, 1970, at pages 1057-1064. The same approach was used by E. A. 
Aitken et al for an investigation described in an article in the 
Transactions of the American Nuclear Society, Volume 14, Number 1, pages 
176-177. In this article, the reactions of fission products in 
urania-plutonia fuels held in stainless steel fuel cladding were 
investigated. 
Kenji Motojima et al disclose a method of separating molybdenum-99 from 
irradiated uranium dioxide by sublimation in the International Journal of 
Applied Radiation and Isotopes, Volume 27, pages 495-498 (1976). The 
irradiated uranium dioxide is converted to U.sub.3 O.sub.8 by heating at 
about 500.degree. C. in an oxygen atmosphere and then the nuclides (Mo-99, 
Te-132, and Ru-103) are separated from the U.sub.3 O.sub.8 by heating at 
1200.degree. C. under vacuum. 
However it remains desirable to provide a process in which the separation 
of the molybdenum is conducted simultaneously with the oxidation of the 
target material. 
SUMMARY OF THE INVENTION 
This invention comprises a process for the separation and collection of 
molybdenum-99 from an irradiated enriched uranium-containing target 
material. The process is carried out in a thermochromatographic zone 
comprising a heating zone and a cooling zone. The irradiated target 
material is heated in the heating zone in the presence of a flowing 
oxidizing atmosphere at a temperature sufficient to oxidize the oxide 
forming constituents of the target material, drive off the fission product 
gases and sublime the sublimable materials either initially present in the 
target material or formed during the heating step. The flowing oxidizing 
atmosphere carries the vaporized and sublimed materials and passes from 
the heating zone through the cooling zone. The temperature in the cooling 
zone decreases from that of the heating zone to near room temperature so 
the temperature of the oxidizing atmosphere is controllably decreased as 
the atmosphere passes through the cooling zone. The sublimed and vaporized 
materials in the atmosphere are progressively condensed and collected at 
successive locations in the cooling zone (and thus separated from one 
another in the form of deposits) as the temperature of the atmosphere is 
decreased. The separated materials, including molybdenum-99 trioxide, are 
then recovered from the cooling zone. 
OBJECTS OF THE INVENTION 
It is an object of this invention to provide a process for separating 
molybdenum-99 from an irradiated target material without generating large 
volumes of radioactive liquid waste and encountering the cost of disposing 
of such waste. 
Another object of this invention is to provide a process for separating 
molybdenum from an irradiated target material utilizing the oxidizing 
atmosphere for two functions, namely for oxidizing the irradiated target 
material and for transporting the gaseous products emanating from the 
oxidized target material to a lower temperature location for collection of 
sublimates from the gaseous products. 
Still another object of this invention is to provide a process achieving 
very high recovery of molybdenum-99 from an irradiated target material in 
an efficient and economical manner.

DESCRIPTION OF THE INVENTION 
Referring now to FIG. 1, an apparatus (thermochromatographic column) 10 is 
comprised of a non-reactive tube 12 (such as quartz or high purity 
alumina) having removable plugs 14 and 16 at its ends. Oxidizing gas inlet 
tube 18, provided with control valve 22, opens through plug 14 and gas 
exit tube 20 opens through plug 16. A source (not shown) of an oxidizing 
gas, such as oxygen or air, is connected to inlet tube 18. A movable 
quartz-clad closed-end tube 24 housing a thermocouple is positioned in 
exit tube 20 for use in measuring the temperature at different points 
along the length of tube 12. 
A heating means 26 is provided at one end of the quartz tube, and a 
preferred heating means 26 is in the form of an electrical resistance 
heater. The heated end of tube 12 constitutes the heating zone 11. The 
remainder of tube 12 constitutes the cooling zone 13 that is cooled by 
heat lost to the surrounding atmosphere. The heat loss is such that a 
temperature profile for the tube 12 is developed as shown in FIG. 2 for a 
maximum temperature in the heating zone of about 1500.degree. C. 
The tube 12 and plugs 14 and 16 define a thermochromatographic zone when 
the heating means 26 is in operation. 
During the practice of the process of this invention, an irradiated 
uranium-containing target material is introduced to the heating zone 
portion of the thermochromatographic zone. In FIG. 1 the container 28 is 
shown holding the target material 30 in particulate or powder form. 
Container 28 is typically made of a refractory material, such as platinum 
or alumina, suitable for use in elevated temperature furnaces. After 
insertion of the target material, a flow of an oxidizing gas successively 
through the heating zone and the cooling zone is started and heating means 
26 is turned on. The irradiated target material is thus heated in the 
presence of an oxidizing atmosphere to a temperature sufficient to oxidize 
the oxide-forming components of the target material, drive off the fission 
product gases, and sublime the sublimable materials either initially 
present in the target material or formed during the heating step. A 
temperature of at least about 1000.degree. C. is required and preferably a 
temperature in the range of about 1000.degree. C. to about 1600.degree. C. 
is used. One sublimable oxide formed is molybdenum-99 trioxide from the 
fission product molybdenum present in the target material. The flow of the 
oxidizing gas through the heating zone 11 carries the gaseous oxides and 
any gaseous fission products emitted from the irradiated target during 
heating into cooling zone 13. 
In cooling zone 13, the gaseous products susceptible of forming sublimates 
do so according to their respective vapor pressures as the gas temperature 
decreases according to the temperature profile in FIG. 2. In this manner 
there is a thermochromatographic separation and collection of the 
sublimates in the form of deposits in different locations on the inner 
surface of tube 12. In greater detail the temperature gradient in the 
cooling zone 13 permits effective separation of the various sublimed 
materials to provide maximum localization of the sublimates in the form of 
separate deposits. Typical sublimates collected are illustrated in FIG. 1. 
Deposit 32 represents a small amount of U.sub.3 O.sub.8, deposit 34 
represents a deposit of Cs.sub.2 O, deposit 36 represents a deposit of 
MoO.sub.3 and deposit 38 represents a deposit of iodine. Iodine is 
transported to the coolest part of the cooling zone since it, of those 
materials condensible, exhibits the highest vapor pressure. Molybdenum, 
the desired product in the form of MoO.sub.3 will be deposited on the tube 
12 at a specific location between the target material 30 and the plug 16. 
The gaseous fission products, such as xenon and krypton, do not condense 
at all at atmospheric temperature and pressure, and are carried in the gas 
stream exiting the apparatus 10 for collection. 
The process of this invention is modified when the target material is in 
the form of pellets. In such a case, a two stage heating step is employed. 
The pellets are first heated to and maintained at a temperature sufficient 
to convert the pellets into a powder, i.e., a temperature in the range of 
about 400.degree. to about 600.degree. C. for a time in the range of about 
1 to about 3 hours. The target material is reduced to a powder in order to 
facilitate substantially complete release of the molybdenum in the form of 
gaseous molybdenum trioxide. Then the temperature is raised to at least 
about 1000.degree. C., preferably from about 1000.degree. to about 
1600.degree. C. to achieve oxidation of the target material and 
sublimation of the sublimable materials. 
The recovery of the molybdenum from tube 12 can be facilitated by making 
the portion of the tube 12 removable where the deposit of molybdenum 
trioxide occurs. Recovery of the MoO.sub.3 from the tube 12 is 
conveniently accomplished by dissolving the molybdenum trioxide in a 
caustic solution or by mechanical removal such as brushing the deposit 
from tube 12. 
It is possible to mix the oxidizing gas (e.g., air or pure oxygen) with a 
carrier gas. Typical carrier gases are non-reactive gases such as nitrogen 
and the inert gases, such as argon, helium and neon or other oxidizing 
gases such as carbon dioxide. In practice carrier gases comprising from 20 
percent up to about 80 percent by volume can be used in a mixture with the 
oxidizing gas. 
Typical gas flows for the oxygen-containing gas through the chromatographic 
zone in the process of this invention are in the range of 1 cubic 
foot/hour to 10 cubic feet/hour. 
Several different compositions can be used for target material prior to 
irradiation. One preferred composition is uranium dioxide in which 
uranium-235 comprises about 93% by weight of the uranium in the uranium 
dioxide. Another composition is comprised of a comparably enriched pure 
uranium metal. The process of this invention has the advantage of being an 
anhydrous process. The process does not produce any significant amount of 
radioactive liquid wastes and has no liquid waste disposal problem. 
The utilization of the oxidizing gas to transport the gaseous products from 
the oxidized target material in the heating zone into the cooling zone 
achieves more rapid and efficient separation and collection of the 
molybdenum-99 from the irradiated target material. The process of this 
invention offers an advantage of very high release rates for molybdenum 
from the target material and in practice releases rates of 99% by weight 
or more are achieved. 
Those skilled in the art will gain a further understanding of this 
invention from the following illustrative, but not limiting, example of 
this invention in which the percentages are based on weight, unless 
otherwise stated. 
EXAMPLE 
Starting Material Preparation 
About 50 grams of ceramic grade unenriched uranium dioxide powder of 
particle size less than 325 mesh were placed in a container. From a first 
portion of this powder, four cylindrical control pellets each weighing 
about 10 grams, 1/2 inch in diameter and 1/2 inch in height were pressed. 
The remaining portion of uranium dioxide powder was divided into four 
generally equal lots to which various additions of natural molybdenum 
powder of particle size less than 325 mesh were added to produce four 
different UO.sub.2 powder lots containing respectively, 50 parts per 
million (p.p.m.), 100 p.p.m., 300 p.p.m. and 700 p.p.m. natural 
molybdenum. Natural molybdenum is comprised of the following: 15.84% 
Mo-92, 9.04% Mo-94, 15.72% Mo-95, 16.53% Mo-96, 9.46% Mo-97, 23.78% Mo-98 
and 9.63% Mo-100. The four resulting uranium dioxide-molybdenum mixtures 
were separately blended for uniformity in a Specs Industries blender, and 
4 pellets from each lot weighing either 2 grams or 10 grams were pressed. 
All the pellets prepared, both UO.sub.2 control and UO.sub.2 -Mo mixture 
pellets, were sintered under an atmosphere of dry hydrogen (having less 
than 10 p.p.m. by volume H.sub.2 O) at 1750.degree. C. for 4 hours. The 
molybdenum content of each pellet was determined by chemical analysis of 
the powder from which the pellet was pressed using the Dithiol method. 
Molybdenum Release Test 
A group of five pellets was selected from those prepared as described 
above, including one control pellet and one pellet each from the four 
different molybdenum-containing lots. This group was given the arbitrary 
description of pellet lot A. 
Each pellet was crushed to a powder and the powders were placed in separate 
platinum crucibles and a small portion of the powder from each crucible 
was taken for chemical analysis. The crucibles containing the remainder of 
the powders were simultaneously treated according to the process of this 
invention in the same furnace. The process was carried out in an 
atmospherically controlled electrically heated tubular furnace 
manufactured by the Heraeus Company. The furnace had a high purity 
cylindrical alumina tube 30 inches in length, one inch in outer diameter 
and 1/8 inch in wall thickness. The resistance heating element surrounded 
approximately 15 inches of the tube and was positioned in the middle of 
the tube leaving approximately 71/2 inches of cooling zone cooled by loss 
of heat to the surrounding atmosphere. 
The furnace containing all the crucibles was heated from room temperature 
at a rate of 7.degree. C. per minute to a maximum temperature of 
1500.degree. C. and held at that temperature for 18 hours. A gas 
atmosphere of pure oxygen was passed through the furnace at a rate of one 
cubic foot per hour, flowing first into and through the heating zone and 
then through the cooling zone. 
Upon examination after completion of the heating, the oxidized powders were 
noted to have volatilized some of the uranium oxide. A deposit of what 
appears to be uranium oxide was found within the furnace tube downstream 
from the heating element at a point where the temperature was 
approximately 1200.degree. C. The deposit was not analyzed but was 
believed to be U.sub.3 O.sub.8. A yellow deposit determined by x-ray 
fluoresence to be molybdenum trioxide was noted at another point further 
downstream in the cooling zone at a location indicative of 500.degree. C. 
.+-. 50.degree. C., and this deposit was easily removed from the tube. 
Table II gives the pellet identity by attempted molybdenum addition, the 
molybdenum content of the crushed pellets prior to oxidation, and the 
molybdenum content of the residual powder in the crucibles removed from 
the furnace. 
TABLE I 
______________________________________ 
LOT A 
Pellet Identity 
Mo Content Mo Content 
by Attempted Mo 
Prior to After 
Addition (p.p.m.) 
Oxidation (p.p.m.) 
Oxidation (p.p.m.) 
______________________________________ 
0 3.8 1.0 
50 35.9 1.7 
100 77.7 4.5 
300 256.8 1.0 
700 545.7 6.1 
______________________________________ 
This demonstrates that release of more than 99% by weight of the molybdenum 
contained in a material can be achieved by the practice of the process of 
this invention. 
It is to be understood that although the invention has been described with 
specific reference to particular embodiments thereof it is not to be so 
limited, since changes and alterations therein may be made which are 
within the full intent and scope of this invention as defined by the 
appended claims.