Table V.
A m Distribution in Multikilogram Electrorefining Runs
.Yo.of Electrorefining Runs 19
Inzt. Concn., Salt Mole 7‘ Ka PuC13 2 10 190 f 20 PUCI? 0 49 3 220 & 20 PuF~ 2.10 2 65 f 1 PuF~ 0.49 20 110 i 10 a K = parts Amlmillion parts Pu (salt phase after r u n ) parts Am/million parts P u (metal product)
ride is more readily available and less hygroscopic than the chloride.) The use of mixed fluoride-chloride was very unsatisfactory when alumina cells were used. The aluminum concentration was increased from 40 to 450 p.p.m. Acknowledgment
The authors are indebted to A. N. Morgan, C. A. Emery, and C . A. Rendell, Group CMB-11, who assisted in performing the experiments described in this paper. literature Cited
(1) Blumenthal, B., Brodsky, M. B., ”Preparation of High Purity
110 grams of salt (2.10 mole % PuC13) gave a value of K = 227 a t 740’ C. Thus the value of ii‘ given in Table V for comparable conditions indicates that the Am concentration in the electrorefined metal is about the equilibrium amount. Effect of Materials of Construction on Purity of Producf
During the development phase of the process, the following modifications Mere introduced, primarily to improve the product purity: Electrorefined metal was not allowed to come in contact Lvith tantalum cell components. Instead, tungsten parts were used. This resulted in a decrease in refractory metal content of from about 500 p.p m. of tantalum (72) to about 30 p.p.m. of tungsten. The stirrer was changed from tantalum to ceramic. Although magnesia is more desirable from the compatibility viewpoint, alumina is satisfactory in all-chloride electrolytes. Alumina was replaced by magnesia as the principal material of construction of the cell. This altered the impurity content in the product metal from about 40 p , p , m ,of aluminum to less than 5 p.p.m. of magnesium. \Vhen magnesia was the principal material of construction for the cell, plutonium fluoride could be used in place of the plutonium chloride electrolyte. (Generally, plutonium fluo-
Plutonium,” “Plutonium 1960,” p. 171, Cleaver-Hume Press, Ltd.. London. 1961. (2) Curtis, M. H.. Hopkins, H. H., Jr., Electrochem. Technol. 2, 239-44 (1964). (3) Ellinger, F. H., Land, C. C., Struebing, V. O., J . .Vuclear .Mater. 12, 226-36 (1964). (4) Glassner, X., U. S. ‘4t. Energy Comm., Rept. ANL-5750 11957). (3)‘ Leaiy, J. A . et ai., “Pyrometallurgical Purification of Plutonium Fuels,” Peaceful Uses of Atomic Energy, Vol. 17, p. 376, 1958. (6) Mullins, L. J., Leary, J. X.:“Proceedings of First Australian Conference Electrochemistry,” p. 923, Perganion Press, Oxford, England, 1964. (7) Mullins. L. J., Leary, J . X.,Bjorklund, C. \I/., U. S. At. Energy Cornm., Rept. LAMS-2441 (1960). (8) Mullins, L. J., Leary, J . A,, Morgan, A. N., Zbid., LA-2758 (1962). (9) Ibid., LA-2981 (1963). (10) Zbid., LA-3029 (1963). (11) Mullins, L. J.. Leary, J. .A,, Morgan. A. K.. Maraman, \V, J . , IND.Esc. CHEM. PROCESS DESICSDEVELOP. 2, 20 (1963). (12) Mullins. L. J., Leary, J. A , , Morgan, A. N.. Maraman, \V. J., U. S. At. Energy Comm., Kept. LA-2666 (1962). (13) Schonfeld, F. \V,, Crarner, E. M.: Miner, \V. N., Ellinger, F. H., Coffinberry, A. S., Progr. .Vucl. Energy, Ser. V , 2 (Metallurgy and Fuels), 585 (1959). RECEIVED for review January 14, 1965 ACCEPTEDJune 7, 1965 Division of Nuclear Chemistry and Technology, 148th Meeting, ACS, Chicago, Ill., September 1964. IVork done under auspices of the L.S. Atomic Energy Cornmission.
CONVERSION OF IRRADIATED FUEL SWARF T O A STABLE CERAMIC SUITABLE FOR LONG-TERM STORAGE D . A.
H I L T O N A N D J.
H. BUDDERY
Central Electriczty Generating Board, Berkeley LYuclear Laboratories, Berkeley, Gloucestershire, England
x
years considerable effort has been devoted to the of nuclear waste disposal and the methods in general use have been well reported in the literature (7, 3-7). T h e treatment of the highly active fuel swarf (metal debris from cutting operations such as milling, drilling, and turning), obtained from cutting operations associated M ith the postirradiation examination of fuel elements from British Power reactors, presents a special case in waste disposal. These reactors use natural uranium fuel canned in magnesium alloy (Magnox A180, 99.27, Mg, 0.87, AI) and the swarf RECENT
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400
l&EC P R O C E S S DESIGN AND DEVELOPMENT
obtained from such fuels (which can be mechanical mixtures of uranium and Magnox or uranium alone) presents a considerable fire hazard. I t is necessary, therefore, from both operational and safety aspects, to convert such waste material into a chemically stable form which can be safely dispatched for long-term storage. To develop a process suitable for the treatment of irradiated fuel swarf, the following was desirable: Operations should be mechanically simple and involve the minimum amount of handling, since they will be carried out remotely.
A method i s described for the conversion of irradiated fuel swarf, of the type obtained in the postirradiation examination of uranium-Magnox fuel elements, into a stable ceramic by reaction with a 5 weight % cupric oxide-32 weight % alumina-63 weight lead borate glass mixture. The process requires the minimum of handling cind i s carried out by a three-stage heating sequence for periods of 1 hour each at 350', 500", and 700' C. Losses of fission products throughout the reaction are less than 20 p.p.m. of the gross @ and y activities and the product formed i s suitable for long-term storage. Brief details of the resistance of the ceramic to water leaching and radiation are given.
yo
T h e process sequence should preferably take place in one piece of equipment. The product should be in a form \vhich is conveniently transportable and has 110 dust losses during handling. Losses of fission products throughout the process should be negligible. Although the produiit was intended for storage in a waterproof silo, the water leaching property was, nevertheless, investigated, since a t some future date direct ground disposal of this type of waste might be considered. A number of possible schemes were examined using uranium swarf alone (this being the more active and hazardous material). These included dissolution, fixation in the more radiation-resistant organic plastics, direct fixation in glass, and gaseous and solid state oxidation. Qf these, solid state oxidation appeared to he the most attractive method for converting uranium swarf to a stable compound. Process Development Based on Solid State Oxidation
T h e basis of this method is the reaction of uranium with a metal oxide less stable thermodynamically than uranium dioxide. I n this way the uranium is converted to oxide and the metal oxide reduced to the metal. Several criteria are desirable for a reaction of this type : The temperature of the reacting mass should be as low as possible to reduce losses,of volatile fission products. The oxide should have sufficient bulk to cover all the swarf. The metal formed by reduction should be reasonably stable chemically.
A series of experiments designed on the basis of these considerations showed that a smooth controllable conversion of uranium sivarf (bulk density 0.35 gram per cc.) to uranium dioxide could be effected by heating the swarf to 350' C. bvith cupric oxide (representing 1257, of the stoichiometric quantity required to react with all the uranium) mixed with several times its own weight of alumina (in the form of C60 po\vder supplied by Thermal Syndicate, Ltd.). Alumina is more stable thermodynamically than uranium dioxide and was used to give the oxide powder both sufficient bulk to cover the low bulk density uranium swarf and to control the temperature rise (which was not greater than 100' C. for the treatment of 100 grams of uranium) during the reaction. T h e product was a completely unsintered powder which, although chemically stable, would have presented a serious dust hazard if transported in this form. Methods for the consolidation of the pavder by incorporation of a powdered loiv melting glass in the mixed cupric oxide-alumina powder were therefore examined. I t \vas subsequently shown that a simple 90 weight Yc lead oxide-10 Lveight 7 , boric oxide glass (which melts a t approximately 500' C.) mixed with the oxide powder in the weight ratio of 1.7 glass to 1 cupric oxide-alumina poivder gave a suitable product. After heating at 700' C. for 1 hour the reaction product was highly sintered and had no possibility of
dust losses during handling. The composition of the additive finally chosen was 5 weight % cupric oxide, 32 weight % alumina, and 63 weight % glass, the amount of powder required for reactions being based on 1257, of the stoichiometric quantity of cupric oxide required to oxidize all the uranium. Incorporation of Magnox in Process
T h e possibility of using the process for mixed uraniumMagnox swarf was investigated by heating together mixtures of the cupric oxide-alumina glass powder and Magnox swarf (bulk density 0.16 gram per CC.). A reaction occurred at 500' C., and with high proportions of Magnox the reaction temperature reached 1200' C. In some cases particles of Magnox caught fire. However, provided the ratio of Magnox to oxideglass powder was not greater than 1 to 40 by weight, the maximum temperature recorded \vas 850' C . and the reaction was smooth and controllable. T h e product was a stable solid which had lead globules distributed throughout, indicating that a t least part of the reaction was between the magnesium and the lead borate glass. Further tests showed that mixtures of uranium and Magnox could be successfully converted to a stable solid of average density 3.4 grams per cc., within the limits already detailed. T h e process took place in three well defined stages: reaction of the uranium with cupric oxide a t 350' C., reaction of the Magnox with some of the lead borate glass a t 500' C . , and final sintering at 700' C. Process Scale-up
The full scale process is operated in a shielded cell designed for handling up to 1000 m.e.v. curies of y activity (see Figure 1). T h e cell is maintained a t a negative pressure of several inches of water gage, and the extract (which gives 30 air changes per hour) is filtered initially in the cell itself and then through a second filter outside the cell. Handling operations inside the cell are carried out remotely by two Hobson No. 7 heavy-duty manipulators. As a precaution against fire, a nozzle points from the roof between the manipulators and is directly connected to a chloride eutectic powder extinguisher, operated by argon pressure, outside the cell. For ease in maintenance the reaction furnace and drying oven stand on platforms attached to the cell door. The oxide-glass polvder (in 250-gram quantities, of particle size less than 250 microns) is weighed into a 1-inch diameter polythene "sausage," which is then fed into the cell in sealed lengths through a hole a t the top of the front face. The sxvarf, drained of cutting fluid and contained in a mild steel gauze basket, enters the cell via a railway from the cutting cells. After being Iveighed on a spring balance, it is transferred to a mild steel pot (4 inches in diameter, 2 inches deep, '/*-inch wall thickness) which is used as the reaction vessel. The oxideglass powder is poured over the sivarf (after piercing the polythene sausage), and the two are then roughly mixed. The mixture is dried in the oven a t 200' C. and then transferred to the reaction furnace for the three-stage heating cycle a t 350', 500', and 700' C. (see Figure 2). If the sivarf is extremely fine, an excessive amount of fuming can occur during the oven drying due to retention of large amounts of cutting VOL. 4
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4 OCTOBER 1965
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only during periods of routine maintenance. O n the present scale about 1 kg. of mixed uranium-Magnox swarf can be processed in a 40-hour week, and this capacity is in excess of the current production rate of the swarf. Fission Product Losses during Process
T h e losses of fission products by volatilization during the process were assessed, using sintering temperatures of 700°, BOO", and 900" C., by processing a sample of irradiated swarf (1200 magawatt-days per ton burnup, 150 days' cooling) in a crucible inside a silica tube in a horizontal furnace. The ends of the silica tube were cooled and closed by glass wool plugs. After the reaction was complete and the furnace had cooled, the volatilized fission products were dissolved from the silica tube and end plugs and determined by normal procedures. T h e losses of gross /3 and y activities are shown in Table I . These activities were due to the fission product isotopes of ruthenium and cesium. The results show that fission product losses were less than 20 p.p.m. of the total gross /3 and y activities at the sintering temperature of 700" C., but that they can rise by over an order of magnitude if a sintering temperature of 900" C. is used. Figure 1 ,
Irradiated fuel swarf disposal cell
Microscopic Examination
7. Intercell railway
1. 2.
Reaction furnace Drying oven 3. Air outlet filter 4. Air inlet 5. Reaction pots 6. Spring balance
8.
Irradiated fuel transporter
9. Fire extinguisher 10. 11. 12.
Power point Thermocouple point Pasting port
700
/-
y 600 u a a
2
500
U
n W
5
400
IY
U
a 300
2
IL 3
200
0
*
s:
100
W
a
O O
I
3
2
TIME
5
4
b
(HOURS)
Figure 2. Reaction chart for fixation of uranium-Magnox swarf using mixed oxide-lead borate glass mixture A. 6.
From drying oven Uranium-copper oxide reaction, U 2CuO UOr 2Cu. Reaction controlled by alumina diluent and lead borate glass C. Magnox oxidation. Reduction of some lead borate glass to lead D. Sintering process. Uranium and magnesium oxides converted to stable ceramic
+
+.
+
fluid by the sivarf even after draining. T o prevent this, the swarf is degreased by spraying with cold trichloroethylene and allowed to drain prior to weighing and mixing with the reaction powder. After the final sintering a t 700' C. the furnace is allowed to cool and the processed swarf vessels are transferred by the railvay to another cell to be sealed in epoxy resin-painted mild steel cylinders prior to final underground storage in waterproof silos. The process has been operated satisfactorily on a semiroutine basis for over 2 years, and the cell has been out of commission 402
I&EC PROCESS D E S I G N A N D DEVELOPMENT
To study further the nature of the ceramics produced a number of specimens were sectioned, mounted in plastic, ground, and polished for detailed examination. Samples of the lead borate glass and a glass-alumina mixture were heated at 850" C. for 30 minutes and examined. There was no visible attack of the alumina by the glass. Examination of sections of sinters prepared from unirradiated uranium, Magnox, and mixtures of the two showed that in each case the predominant phase was alumina. With the uranium-based sinters, no lead should be present if the uranium-cupric oxide reaction is completed before the glass melts. It was confirmed that the only phases present were sharply defined alumina particles, copper, and a glassy phase filling the interstices. The Magnox-based sinters, besides having the alumina particles and the glassy phase, had two types of metallic phase present: globules of lead from which copper had precipitated on cooling and bright areas of copper free of lead. As would be expected, the general features of a section through a mixed uranium-Magnox sinter were similar to the sinter based on Magnox. Microscopic examination did not define the distribution of uranium in the ceramics, but electron probe microanalysis showed that uranium-rich areas were always associated with areas rich in lead glass, indicating the formation of a mixed uranium dioxide-lead borate glass. Autoradiographic studies further showed that the (unirradiated) uranium was considerably segregated throughout the ceramics. The distribution of the reacted Magnox was also not clear, since it was not possible to examine a material of such low atomic number with the electron probe microanalyzer used.
Table 1.
Loss of Volatile Fission Products during Sintering
Sintering Conditions 700' C. for 1 hour 800' C. for 1 hour 900' C. for 1 hour
Loss of Volatile Fission Products, P . P . M . of Total Gross p and y Activities of Original Sample Gross p Gross y 15 3 48
125
32 70
leaching Tests
Polished surfaces of sinter were examined microscopically at regular intervals after leaching in cold deionized water. Some areas of lead borate glass were attacked after as little as 3 hours’ leaching, but this rapid removal then ceased and thereafter little further change was noted except a gradual removal of lead borate glass and dulling of the bright metallic areas due to corrosion. T h e initial rate of loss was 0.00370 of the gross y activity per \veek and this rate decreased by an order of magnitude within 6 months. Cesium-1 37 was the principal fission product leached from the sinter. Irradiation Tests
Samples of the sinters and lead borate glass were subjected for long periods to high intensity ?-radiation. No significant change was observed in samples irradiated to a total dose of loio rads. Discussion
T h e process developed for the fixation of irradiated fuel swarf takes place in several well defined steps. T h e first stage involves the oxidation of uranium by cupric oxide at 350’ C. In this case both alumina and the lead borate glass act as inert diluents and there is negligible reaction of the hfagnox. \$’hen the temperature is raised to 500’ C., the Magnox reacts with some of the lead borate glass. This reaction increases the boric oxide content of the glass but does not affect its final sintering properties, since the softening point of an 84 weight % lead oxide-boric oxide glass (which in this process represents the maximum quantity of lead replaced by magnesium) is approximately the same as that of the original glass (5). During the final sintering a t 700’ C. the uranium dioxide dissolves in the lead borate glass and some copper dissolves in lead. T h e product formed, although heterogeneous, is a stable solid which has no dust losses during handling. T ~ v ofurther secondary reactions involving the oxidation of Magnox by the residual cupric oxide and the oxidation of Magnox by alumina may take place a t the temperature reached by the Magnox-glass reaction. T h e former reaction does occur,
since copper-lead alloys and free copper are present in sinters formed by reaction of Magnox only. However, although in the absence of uranium approximately half of the Magnox could be converted by this mechanism, only a small proportion of Magnox would be oxidized when mixed swarf was processed. Microscopic examination has shown that the Magnox-alumina reaction is unlikely, since the alumina suffers no serious attack. Because of the low sintering temperature the losses of fission products by volatilization during the process are less than 2@ p.p.m. of the gross fl and y activities and since the product contains crystalline grains of alumina in a glassy matrix, it is more akin to a porcelain than a glass. Although it is relatively water-insoluble, it cannot be seriously compared with the very water-resistant glasses which are suitable for direct ground disposal of fission product wastes. These glasses do, however, have a considerable loss of volatile fission products which can cause operational problems, since the sintering temperatures are generally above 1000’ C. (2). There is no evidence to date to suggest that y-irradiation will seriously affect the general properties of ceramics prepared in this way. Acknowledgment
T h e authors thank R. J. Pearce for carrying out the electron probe microanalysis of the swarf ceramics and M. T. AMcCahill for help with some of the experimental work. This paper is published by permission of the Central Electricity Generating Board. literature Cited
(1) Amphlett, C. B., “Treatment and Disposal of Radioactive IVastes,” Pergamon Press, Oxford, London, New York, and Paris, 1961. (2) Bancroft, A. R., Can. J . Chem. Eng. 38,19 (1960). (3) Collins, J. C., “Radioactive IVastes, Their Treatment and Disposal,” E. and F. N. Spon, Ltd., London, 1961. (4) Glueckauf, E., “Atomic Energy IVastes,” Interscience, New York, Butterworths, London, 1961. ( 5 ) Levin, E. M., McMurdie, H. F., “Phase Diagrams for Ceramists,” Vol. 11, American Ceramic Society, New York, 1960. (6) Saddington, K., Templeton, I V . L., “Disposal of Radioactive IYastes,” G. Newnes, London, 1958. (7) Symposium on Treatment and Storage of High Level Radioactive It’astes, International Atomic Energy Agency Proceedings, Vienna, 1963. RECEIVED for review February 17, 1965 ACCEPTED June 28, 1965
KINETICS OF CATALYTIC VAPOR-PHASE HYDROGENATION OF NITROBENZENE
TO ANILJN E D . N . R I H A N I , T. K. N A R A Y A N A N , A N D
L. K . D O R A I S W A M Y
National Chemical Laboratory, Poona, India HE catalyst employed in the vapor-phase hydrogenation T o f nitrobenzene t o aniline is generally a metal supported on a suitable carrier. Copper is by far the most widely used metal, and has been used as: a mixture of copper and 10% calcium oxide ( 7 4 , copper carbonate and a 50% solution of sodium silicate impregnated o n lumps of pumice (a), cuprammonium nitrate (75), and copper precipitated on pumice (78, 20).
Nickel in finely divided form, obtained by the ignition of nickel nitrate ( 2 ) ,has also been used. Quantitative conversion using a thiophene-poisoned nickel catalyst has been reported (23). High conversion to aniline is also obtained by using a nickel catalyst prepared from nickel nitrate containing 1.5 to 2% nickel sulfate (22). But in general nickel catalysts are too active for this reaction, and are therefore used mostly in combination with other metals such as copper or aluminum ( 7 ) . VOL. 4
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