3,825,649 United States Patent Office Patented July 23, 1974 1. 2

l, ...... 3,825,649 Thorium-232 - I - Thorium-233 --> PROCESS FORSEPARATION OF PROTACTINUM, 23.5 min. ... THORIUM AND. URANIUM FROM NEUTRON IRRADIATED THORIUM Protactinium-233 -> Uranium-233. Alan T. Gresky, Jouko E. Savolainen, and William T. 5 27.4 day McDuffee, Jr., Oak Ridge, and Russell P. Wischow, Thorium may be subjected to neutron bombardment in Nashville, Tenn., assignors to the United States of varying types of reactors. For example, thorium metal America as represented by the United States Atomic may be inserted as aluminum-encased slugs into a hetero Energy Commission geneous reactor, or slurries of thorium oxide may be cir * * * * * Filed Aug: 7, 1956, Ser. No. 602,686 0 culated about a homogeneous reactor core of an aqueous ?“ '.' : , nt. CI. C01g 56/00; C22b 61/04 uranyl sulfate solution enriched, beyond natural abund U.S. C. 42 . . . . . 21 Claims ance, with regard to uranium-235. The chemical processing of neutron bombarded tho - ABSTRACT OF THE DISCLOSURE rium is of prime importance, for any product material lost Protactinium, uranium and thorium are separated from in the chemical processing, in effect, increases the de an aqueous nitric acid solution of neutron-irradiated tho mands upon the efficiency of the reactor system. Further rium containing these elements and fission products by more, in reactor fuel processing, contrary to most chem contacting, under net nitrate deficient conditions, the ical processing operations, relatively great amounts of acid solution with an organic solution of a trialkyl phos unreacted material must be separated from relatively phate in an inert organic diluent, thereby preferentially small amounts of products. This arises from the fact that extracting uranium and thorium into the organic phase nuclear fission products of high neutron-absorption cross while confining protactinium and fission products to the sections compete with the fuel for fission-released neu aqueous phase. After scrubbing the organic phase with a trons. Unless such fission products are removed from the aqueous solution of an inorganic nitrate salt to remove reactor, the maintenance of the chain reaction itself may small amounts of protactinium and fission products, the be threatened. Thus, in actual practice, the fuel and fertile two-phases are separated and thorium and uranium are material must be periodically removed from the reactor separately recovered from the organic phase. for decontamination long before the fuel and fertile ma terial are consumed. In addition to extremely high recov ery of fissionable uranium-233, ideally approaching Ouriyention relates to a process for the decontamina 30 100%, the chemical processing should also achieve ex tion of neutron-irradiated thorium, and more particularly cellent decontamination of uranium-233 and thorium to a process for the separation of protactinium-233, from highly radioactive fission products before prepara thorium and uranium-323 from neutron-irradiated tho tion for reuse in reactors. This is essential for both per 1. sonnel safety and maintenance of good neutron economy. A major factor in the cost of generating electricity Perhaps the most perplexing of all problems associated from nuclear fission is the cost of the fuel. Factors which with the chemical processing of neutron-irradiated tho contribute to low fuel cost and towards which reactor rium is the handling of the highly radioactive protactin designs, seek to approach are low, cost fabrication of fuel ium-233, the parent of uranium-233. This isotope usually elements, high burn-up of fuel before reprocessing is re accounts for greater than 95% of the beta-gamma ac quired, low cost reprocessing, and high thermal efficiency. 40 tivity in the irradiated thorium at the time of withdrawal A concurrent approach in reducing the unit cost of gen from the reactor. The relatively short half-life of protac erating electricity from nuclear fission is to obtain by tinium (27.4 days) would argue for prolonged cooling products of high value which can be credited against other of the irradiated thorium prior to any chemical process 45 ing to minimize losses of potential uranium-233. It is generation costs. A principal effort in this direction is estimated that a cooling period of about 250 days would towards the regeneration of fissionable material from normally permit uranium-233 losses of less than 0.1% “fertile" materials concurrent with the consumption of and would allow decay of the 24.1-day thorium-234 ac nuclear: fuel. Reactors designed for fuel regeneration as tivities which otherwise limit thorium-product purification. well as power production are commonly known as "dual Furthermore, the extreme radioactivity of protactinium, purpose" or “breeder" reactors and the regenerative fis 50 with its consequent shielding and handling problems, pre sionable materials produced by such reactors are the well sents additional argument for longer cooling before chem known plutonium (from uranium-238) and also uranium ical processing. Overcoming all these arguments in favor 233 (from thorium). Depending upon the neutron econ of longer cooling period, nonetheless, is the single, crucial ofny of a particular reactor (the number of neutrons 55 fact of the high inventory charges against fissionable ma available for radiative capture by a fertile material be terials. Thus, the precious and expensive fissionable ura yond the requirements of maintaining the chain reaction) nium-233 and the fertile thorium cannot be permitted to remain dormant and unproductive. Furthermore, and as much or more fissionable material may be produced apart from a uranium-233 breeder program, protactinium as is consumed. Such a breeding program may make re itself is required for basic academic studies, as a tracer actor-produced power- competitive with conventional 60 and as a concentrated beta-gamma source for a host of radiation purposes. Therefore, the chemical process for ply of precious fissionable material will be conserved. In recovering uranium-233 must be prepared to deal with fact, since the world supply of thorium is greater than relatively short-cooled feed material, e.g. 40 days and even the world supply of uranium, the potential exists for ac less, as well as possess flexibility for treating longer-aged tually, increasing the amount of fissionable material by material. - conversion of thorium to fissionable uranium-233, which, The separation of protactinium, thorium and uranium upon recovery, may be used to convert additional thorium presents problems of unprecedented severity. For exam to uranium. : ...... " ple, thorium, protactinium, and uranium are immedi Uranium-233 is obtained by the neutron bombardment 70 ately adjacent neighbors in the actinide rare earth series of naturally occurring thorium-232, essentially by the fol of the Periodic Chart of the Elements. Although recog llowing principal nuclear. reactions: nizable differences are present among the rare earths, 3,825,649 .. 3. 4. they are notoriously chemically similar, since they differ trate salt, separating said protactinium- and fission prod only in the number of electrons in their deep, underlying ucts-containing aqueous phase from said uranium- and shells, rather than in their valence electrons which nor thorium-containing organic phase, and thereafter separat mally govern chemical reactions. Furthermore, there is ing said extracted uranium and thorium from each other. scanty and unreliable information available concerning The practice of our invention achieves an excellent sep the basic chemistry of protactinium. Ideally, a protac 5 aration of protactinium, thorium and uranium in a single, tinium recovery process should provide for its separa relatively simple, continuous solvent extraction cycle. A tion relatively early to permit the subsequent chemical single extractant, trialkyl phosphate, in proper volumetric separation and decontamination of thorium and uranium proportion in an inert organic diluent, in combination 233 to be conducted under less shielding and with re O with an aqueous scrub Solution of an inorganic nitrate duced radiation hazards. salt, sharply and efficiently extracts thorium and uranium There are presently available continuous solvent ex from an aqueous nitric acid solution of neutron irradiated traction processes for accomplishing the two-way sepa thorium, while confining protactinium and the preponder ration of plutonium and uranium from neutron-irradiated ance of fission products to the aqueous phase, the net ex uranium. A representative process of this nature is de traction and scrub conditions being nitrate ion deficient scribed in Ser. No. 303,691, filed Aug. 1, 1952 in the The protactinium may be thereafter separated from fission names of T. C. Runion, W. B. Lanham, Jr. and C. V. Elli products or may be permitted to decay to, uranium-233, son for "Process for Separation of Plutonium, Uranium and the fission product solutioni readily concentrated to and Fission Product Values.” In brief, this process consists relatively small volume for convenient storage or recov of the extraction of uranium and plutonium from an aque- is ery of individual radioisotopes. By first separating-protac ous solution with an organic solvent while confining the tinium, which accounts for approximately 95% of the ra fission products to the aqueous solution, followed by pref dioactivity of short-cooled thorium, from thorium and erential stripping of the plutonium and then of the ura uranium-233 in a single solvent extraction step, the subse: nium from the organic extract with aqueous solutions. quent separation of these two elements in the resulting or The extraction of plutonium into the organic solvent is 2 5 ganic extract may be made under conditions of greatly critically dependent upon its maintenance in the tetrava reduced radiation. ?, - ???????????????????????????????????????????? lent state, while the subsequent stripping of plutonium re We find that such subsequent processing, when com lies upon its reduction to the trivalent state. bined in a single, continuous process with the protactinium Until now, however, there has not been available a separation, consistently obtains uranium-233 recoveries satisfactory solvent extraction process for the separation 30 ranging up to approximately 99.7%. This valuable product of protactinium from neutron-irradiated thorium. Nor affords the opportunity for vastly increasing the supply has there been available a solvent extraction process for of fissionable material in abundance far beyond the poten the immensely difficult three-way separation of protac tial extractable natural uranium-235, and providesa sig tinium, thorium and uranium from each other and fission nificant credit against the unit cost of generating electricity products. by nuclear fission means, thereby bringing closer the An object of our invention, consequently, is to pro dream of economically competitive nuclear power. vide a process for the separation of protactinium, ura The term "fission” is used herein in its generally ac nium and thorium from neutron-irradiated thorium. cepted meaning as referring to the splitting of an actinide Another object is to provide a process adapted for such element, notably uranium and plutonium, into a plurality a separation from an aqueous solution of neutron-irradi 40 of parts upon the capture of a neutron of appropriate en ated thorium in high yield. ergy, and the term "fission products" refers to the immedi Another object is to provide a liquid-liquid solvent ex ate product nuclei from fission as well as to their radio traction process for such separation. active decay products. (See Glasstone, Principles of Nu Still another object is to provide a continuous solvent clear Reactor Engineering, especially pages 105-128). The extraction process for the individual separation of protac closely similar statistical fission product yields of U-233; tinium, thorium and uranium from fission products from U-235 and Pu-239 are shown in Stevenson, Introduction to an aqueous solution of neutron-irradiated thorium. Nuclear Engineering. . . . . Again another object is to provide such a process Considering our invention now in its broader aspects, wherein a single extraction with a single extractant fol the Present process comprises first dissolving the neutron lowed by a pair of simple stripping operations achieves 50 irradiated thorium metal, thorium. oxide, thorium oxy distinct and complete separation of the three compo carbonate, or other thorium compounds. Perhaps the most mentS. connon form of thorium utilization, at the present stage A further object is to provide such a process wherein of the breeder program, is as an aluminum-clad thorium the protactinium is the first of the three components sep metal slug in heterogeneous reactors, the aluminum serving arated. 55 largely to contain charged fission fragments within the A still further object is to provide such a process suffi slug. Such aluminum-clad slugs may be dissolved, for ex ciently versatile to handle neutron-irradiated thorium of ample, by removing the aluminum jacket with a caustic. varying ages. sodium nitrate Solution, and then dissolving the thorin Yet a further object is to provide a process appropriate with an aqueous mineral acid, for instance aqueous nitric for large scale operation in a continuous manner. 60 acid. In a preferred simultaneous dissolution of aluminum. These and additional objects and advantages of our clad slugs, the slug is dissolved with aqueous nitric acid invention will become apparent to those skilled in the Containing tiny amounts of both fluoride and mercuric art from the following detailed description and the claims ions, the fluoride ion catalyzing thorium dissolution and appended hereto. the mercuric ion catalyzing the aluminum dissolution. Al In accordance with our present invention, protactinium, though the mechanism of neither catalytic action is clear, uranium and thorium may be separated from an aqueous it 1s. Suggested, in the case of the mercuric ion, that the nitrate acid solution of neutron-irradiated thorium by con on Is reduced to the metal by aluminum metal, after which tacting, under net nitrate ion deficient conditions, said. So it amalgamates the aluminum surface and prevents forma lution with an organic solution of a trialkyl phosphate in tion of passive aluminum oxide films. While the presence an inert organic diluent, thereby preferentially extracting of fluoride ion is essential for thorium dissolution, it poses thorium and uranium into the resulting organic phase while O a corrosion problem in Subsequent processing. We find that confining protactinium and fission products to the result this problem may be reduced by incorporating at least an ing aqueous phase, Scrubbing any small amounts of ex equi-normal amount of aluminum in the dissolver solution tracted protactinium and fission products from said or to complex the fluoride. When aluminum-clad slugs are ganic phase with an aqueous solution of an inorganic ni dissolved, extra aluminum may not have to be provided. 3,825,649 5 6 The aluminum nitrate, very conveniently, is beneficial in ration-digestion step, the nitric acid condensate being re the obtainment of a nitrate ion deficient feed solution and coverable for reuse in the next dissolution cycle. This treat also serves as a salting action for thorium and uranium ment provides solution and temperature conditions which in solvent extraction. , promote formation of stable silicas that are not deleteri When aqueous nitric acid solutions of thorium are con ous in the following extraction cycle. Another distinct tacted with the organic trialkyl phosphate solution, more 5 advantage of this treatment is the convenient obtainment than 90% of the protactinium and certain fission product of nitrate ion-deficient feed conditions. A further benefit species such as ruthenium are unfortunately extractable, is that it serves to deal with perhaps the most notoriously along with thorium and uranium-233. A cardinal feature troublesome of all fission products, ruthenium. Ruthe of our process, then, is the critical discovery that such O nium shows an uncanny ability to exist simultaneously in deleterious protactinium and ruthenium extraction may various valence states, as well as in different forms of be substantially suppressed, without concomitant Suppres molecular association, such as complexes and polymers, sion of thorium and uranium extraction, by providing net the result of which is extreme difficulty in seeking to con nitrate ion deficient feed and scrub solutions. Thus, the fine it to a single phase during extraction. Nitrate ion feed or the scrub solution may be acid (not nitrate ion deficient conditions strongly suppress ruthenium extract deficient), provided the net extraction and scrub condi ability, and digestion also achieves less extractable ruthe tions are nitrate ion deficient through the nitrate ion de nium species. ficient solution overbalancing the acid solution; however, After the digestion of the feed solution, it may then be it is preferred that the feed and scrub solutions be each contacted with the trialkyl phosphate-organic diluent so nitrate ion deficient. lution. The trialkyl phosphate employed should be a liq As understood in this specification and in the appended uid at the ambient atmospheric temperature and should claims, nitrate ion deficiency is a relative term to indicate preferably comprise approximately 3-6 carbon atoms that a solution of a nitrate salt of a metal of a given among each of its alkyl radicals (that is, from tri-propyl molarity will not register as high an acidity as a solution to tri-hexyl phosphate). The most suitable extractant is of the normal nitrate salt of the same metal molarity, or tri-n-butyl phosphate (hereinafter referred to as “TBP'). in other words, this is a measure of a stoichiometric de The organic diluent should be an inert hydrocarbon and ficiency of nitrate ion, which stoichiometric deficiency is have a density distinctly different than that of water, in made up by hydroxyl ion supplied through hydrolysis order to permit adequate countercurrent flow type of con rather than by addition of other anions like sulfate; in tacting without requiring excessive pump capacity. Petro this respect the solution is acid deficient in anions other 30 leum cuts, especially kerosene fractions, are particularly than hydroxyl. Thus, a 0.1 normal nitrate ion-deficient suitable diluents. solution of thorium, uranium and aluminum nitrate con -Upon contacting of the nitrate ion-deficient aqueous tains that much less nitrate ion than a solution of the feed solution with the organic extractant, the thorium and same thorium, uranium and aluminum molarity. The So uranium-233 preferentially pass into the organic phase, lution will still register an acid pH, although less acid confining substantially all the protactinium and fission than a solution of the normal salt. In the case of alu product values to the aqueous phase. The mechanism pro minum nitrate, nitrate ion deficient solution may be moting preferential extraction of thorium and uranium thought of as a solution of a basic aluminum nitrate Salt into the organic phase, while confining protactinium and (e.g. Al(OH)NO), and such a salt is nitrate ion defi fission products to the aqueous phase, is not completely cient compared to a solution of normal aluminum nitrate 40 understood and we do not wish to be bound to any par of the same aluminum molarity. Nitrate ion deficient ticular theory. It is suggested, however, that organic thorium and aluminum nitrate solutions may be conven soluble TBP complexes of thorium, uranium, and to lesser -iently achieved by dissolving additional thorium or alu extent, nitric acid, are formed, as represented by minum metal in aqueous solutions of the normal salt, by boiling off nitric acid as nitrogen oxides, or in the case of aluminum, by directly employing a basic salt. Gener UO(NO).2TBP and HNOTBP. The confinement of ally, a net nitrate ion deficiency of approximately 0.1- the protactinium and fission products to the aqueous phase 0.6 normal is satisfactory, while approximately 0.3 nor is explained by their failure to complex with TBP under mal is preferred. This may raise the question: Why can't the nitrate ion deficient conditions of the aqueous phase, nitrate ion deficiency be brought about by partial neutrali 50 various hydrolysis effects occurring instead, producing un zation of an acidic solution by direct addition of a base, extractable ionic species of protactinium and fission prod since, in effect this is the result? Two crucial considera CS. tions advise against, although not totally prohibiting, such To further enhance the sharpness of the separation, procedure. Firstly, an undesirable increase of non-volatile we find that an aqueous scrub solution containing nitrate bulk salt concentration would obtain (even with ammo ion serves to drive any small amounts of extracted pro nium hydroxide). Secondly, there would be real risk of tactinium and fission products from the organic phase. precipitation particularly of Al(OH3)] by formation of While a number of inorganic nitrate salts such as sodium localized base concentration gradients. An analytical pro nitrate, may be used to provide nitrate ion, we prefer to cedure for determination of nitrate ion deficiency will be use nitrate ion deficient, aqueous aluminum nitrate solu described later. 60 tion. We further find that the provision of a small amount When an aluminum-clad thorium metal slug is the tho of phosphate ion in the scrub solution is unexpectedly rium form employed for neutronic bombardment, addi effective in decontaminating the organic phase of pro tional process problems arise from troublesome metal tactinium. Under the described operating conditions de lurgical impurities commonly contained therein, Such as contamination factors from protactinium of 102 are ob beryllium, silicon, calcium, magnesium, niobium, iron, tained without the inclusion of phosphate ion; with phos chromium, and nickel and compounds thereof. Especially phate ion, decontamination factors of 10 are obtainable. undesirable is silicon, since siliceous materials are par If the feed solution is of simultaneously dissolved alumi ticular offenders as emulsion promoters in solvent extrac num-jacketed thorium slugs, the inclusion of a small tion contactors and highly refractory claylike materials, amount of ferrous ion beneficially prevents extraction of which are distributed randomly throughout the process any oxidized chromium impurities into the organic phase. equipment, tend to become surface-active carriers of radio The aqueous stream from the extraction cycle contains activity. It is, therefore, highly desirable that the aqueous virtually all the protactinium and fission products. The nitric acid feed solution be treated to minimize these ef protactinium may then be separated from the fission prod fects. ucts, if its individual recovery is desired, or it may be We find that this may be accomplished by an evapo 5 permitted to decay to uranium-233 which is then sepa 3,825,649 7 rated from the fission products by solvent extraction, as in thorium. As only about 50% of the nitriç acidis con; the above extraction step. One method for the separation sumed by reaction, the final dissolver solution has the of the protactinium from fission products in the aqueous approximate primary composition 1.0 M.Th. (NO3)4.04 waste stream, in the absence of interfering ions, is by M Al(NO3)3 and 6.5 MHNO3. Upon completion of the selective sorption on common inorganic adsorbents such 5 dissolution, the dissolver, solution may be slightly cooled, as silica gel, followed by elution therefrom. Another meth say to 90-100° C., and then transferred to a feed adjust od is by solvent extraction means with organic solutions ment tank...for evaporation and digestion. During distilla of a trialkyl phosphate or a diisoalkyl carbinol in an tion, the excess nitric. acid and a portion of free acid, inert hydrocarbon diluent. The preferred method for pro formed by hydrolysis of thorium and aluminum.nitrates, tactinium recovery comprises precipitating aluminum are evaporated, and the condensate collected for further from the aqueous stream as an aluminum chromate pre dissolving cycles. It is postulated that during distillation, cipitate, which precipitate selectively carries protactinium. the major components of the system apparently undergo The organic extract from the extraction column, which stepwise dehydration; first the nitric acid, is dehydrated contains thorium and uranium-233, substantially decon and distilled; and this is followed by dehydration and par taminated of fission products and protactinium, is con tial denitration of aluminium nitrate. It appears that no tacted with an aqueous stream of dilute nitric acid under dehydration of denitration of thorium occurs under these carefully controlled flow and acid conditions to preferen conditions (the aluminum nitrate apparently decomposes tially strip the thorium from the organic solution, while at 130-140° C., while thorium nitrate does not until confining uranium to the organic solution. In conjunction 157-160° C.). If this hypothesis is correct, nitrate ion de with this, it is found beneficial to scrub any small amounts : ficiency of the feed solution is due to the aluminum rather of uranium-233 from the stream with fresh organic solu than the thorium contribution...... tion of the character employed in the extraction cycle. The distillation residue reaches a maximum acidity The aqueous stream, containing the bulk product of the (6.75 M HNOs) at a thorium concentration of about 1.33 overall process, thorium, is substantially decontaminated molar and then decreases linearly with increasing thorium and is sent to an evaporator for concentration. Careful 25 concentration, as shown by Table I, below. This maxi control of conditions in the thorium separation column is mum is thought to be a region with no "free water," only necessary since the distribution coefficients of both urani water of hydration remaining. . um and thorium favor the organic phase. ? ? ? The organic stream, containing the uranium-233, is then TABLE.J. COMPOSITION OF RESIDUE, AND DISTILLATE contacted with slightly acidified water to strip the uranium 30 FRACTIONS DU RING EVAPO RATION CYCLEæ” product. The resulting aqueous uranium solution may then Calculated residue be passed through a cation exchange column for further concentration " . . .": purification from trace amounts of corrosion products « « Liquid and any thorium or fission products. The uranium-de Distillate acidity Th HNO temperature pleted organic stream is introduced into a solvent recovery (M) (M) (N) : ( . C.) 0. 6.07. 15 column where the decomposition products of the TBP 1.05 6.30 6 are removed by washing with an aqueous sodium car 6.47 17 1.18 6.60 18 bonate solution, after which the recovered solvent is re 25 6.70 9 cycled in the process. 1.33 6.75 120 ?.43 6.72 20 Having completed a general description of our process, i.54 6.60 12 a detailed description will now be undertaken, in con 67 6.40 122 182 6.02 123 junction with the single accompanying drawing which 2.00 5.52 124 represents a preferred schematic flowsheet of our process. 2.22 4.82 26 2.50 3.88 129 The main process flow is indicated by the heavy lines. Re 2.86 2.68 13? turning now to the dissolution of neutron-irradiated alumi 3.33 .25 42 3.64 0.47 1. nun-jacketed thorium slugs, such slugs, which commonly 400 ?? 0? 45 15 contain 3 moles of thorium per mole of aluminum, may 4.45 -67? :1 85 5.00 3.30-? 177 be charged into a metal dissolver tank, preferably of stain 5.72 ?5.75 85 less steel, as is all process equipment. There it may be 6.66 -9.30 90 dissolved at a temperature of approximately 110-130 ; *100% nitric acid excess 10 ml.-distillate cuts. C. with a stoichiometric excess of an aqueous solution of "Minus values refer to nitrate ion deficiency. concentrated nitric acid containing small amounts of mercuric ion (aluminum reaction catalyst) and fluoride The evaporation should be continued until the feed ion (thorium reaction catalyst). The exact stoichiometric solution reaches the approximate concentration 4-4.2 excess employed depends upon the irradiation history of 5 5 molar thorium, 1.6-1.8 molar aluminum and 0.2-0.4 the thorium; short term material (two weeks of irradia normal nitrate ion deficient. This concentration can be tion) dissolves in a 50% excess, while long term material reached without deleterious crystallization of thorium dissolves incompletely in a 75% excess and requires a nitrate or oxide or of alumina. The residual solution is 100% stoichiometric excess. Although the quantitative then diluted with water to yield, in addition to various composition of the dissolver solution is subject to con 60 metallurgical, corrosion, and fission product impurities, a siderable variation within the scope of our invention, par feed solution of approximate composition 0.4–0.6 molar ticularly since the resulting solution is subjected to a feed aluminum nitrate, 0.1-0.2 normal nitrate ion deficiency, adjustment step, an aqueous solution of approximate com 1.0-2.0 molar thorium nitrate, 0.02-0.1 molar fluoride ion position 13 N HNO, 0.04 M F- and 0.003 M Hg2+ is and 0.002-0.02 molar mercuric ion and small concentra preferred. The quantity of nitric acid used is calculated 65 tions of protactinium and uranium-233, depending upon on the basis of 10.0 moles per mole thorium and 7.5 moles the age and irradiation history of the slug. Characteristic per mole aluminum charged to the dissolver. During the uranium concentrations are about 0.0013 molar, or about dissolution, which takes several hours (up to four hours 3 gms. /liter, and characteristic protactinium concentra of reflux are required for the dissolution of difficultly tions are about 0.00006 molar. More acid feed solution soluble "blue” thorium oxide impurities in thorium can be tolerated if compensatory increase in nitrate ion metal), the off-gases, consisting mainly of Na and NO, deficiency of the scrub solution are made. A typical feed with lesser amounts of NO2, N2O and H2 are contacted Solution of 80 day-cooled thorium may have a practical with water in a condenser to produce additional nitric beta-activity spectrum in counts per minute per milliliter acid, which may then be recycled to the dissolver. The at 10% geometry: 3X100 protactinium-233; 4x108 total total off-gas volume is about 50 liters per kilogram of rare earths, 2X 10. Zirconium; 2X107 niobium; and 3,825,649 10 5X106 ruthenium. These activities constitute the primary mately 30%-60% trialkyl phosphate, by volume, and the decontamination problems of the process. remainder inert diluent. From the practical view point of The benefits of the evaporation-digestion cycle are providing sufficient extractant capacity for the bulk prod numerous. Silicious impurities, dangerous emulsion pro uct, thorium, we prefer to employ a solution of approxi moters and radioactivity adsorbents, are rendered dehy mate volumetric composition 42% (or 1.5 M) TBP and drated and non-surface active. The digested silicious ma 5 58% diluent, which solution has a specific gravity of less terials do not affect extraction efficiency, and need not be than 0.9 gm./cm3. It is understood, however, that the removed from the feed solution, affording further opera composition of the extractant may be varied, provided tional simplicity. Any remaining undissolved thorium compensatory adjustments of thorium concentration in oxide constituents are dissolved. Finally, the resulting ni O the feed solution and/or relative flow rates or contact vol trate ion deficient solution permits variation in the op umes of feed to extractant are made, withut seriously af erating conditions of the extraction column (when col fecting process operability. umn contacting means are employed) without disruption As might be expected in the contacting of organic so of steady state operation. Perhaps the greatest benefit of lutions of such complex character as petroleum cuts with the nitrate ion deficient feed solution, in addition to sup 5 aqueous acids, certain degradation products are inevitably pression of protactinium extractability, is the great re formed. Particular offenders seem to be olefinic and duction in organic extractability of ruthenium. Ruthenium aromatic contributions to the kerosene fraction and traces distribution coefficients (organic/aqueous) in the extrac of acids, alkali and suspended materials. Traces of aro tion step decrease from approximately 10-2 to 10-4 in matics contribute to the formation of a second organic in passing from acidic to nitrate ion deficient feeds, as phase by extraction of a polymerized TBP complex of shown in Table II, below. This table also indicates that if thorium. These undesirable constituents can be removed the feed adjustment is conducted in glass ware, rather from the diluent by a pretreatment to yield a substantially than in the preferred stainless steel, a small amount of saturated paraffinic diluent. In one pretreatment method, ferrous ion is beneficial. the diluent is washed with a 400 volume of chromyl chlo TABLE II-RUTHENIUM DECONTAMINATION FACTORS AND DISTRIBUTION COEFFI CIENTS OBTAINED IN EXTRACTION COLUMNASA RESULT OF FEED ADJUSTMENT STEP . Distribution coefficient (organiclaqueous) Decontamination 8th scrub Nitric factor stage (top of 5th ex- acid in

: - extraction traction feed * Gross i - Ru " column) stage (molar) Remarks 127 8 0.86 2X10-3 0.56 Acid feed-no adjustment. 96 8 1.03 0.016 0.60 D0. 2.8X103 238. 0.74 3X10-4 -0.15" Nitrate ion deficient feed, feed adjust net. 2.2XOS 24? 0.58 2x10 ?0.44 Do. 1.5X103 160 . . . 0.62 3X10 -0.46 IDO. 1.33X10 740 0.74 5X10-4 -0.4 F???djustment in presence of 0.01 M. ?? ?? ? ?? .etik 1.31X10 823 0.84 6X10-4 -0.20 Do. 1.77x10. 875 0.79 4X10-3 0.28-? F??????djustment in presence of 0.005 M catt. ... , 1,03X10 : 940 1.37 5X10-4 -0.06 F??? adjustment in presence of 0.0025M et.. 1.68x10 1010 1.04 3X10-4 -0.45 Feed adjustment in presence of type 309 SN b stainless steel. Negative values indicate nitrate ion deficiency. The nitrate ion deficiency of our feed and scrub Solu ride, filtered, washed with caustic and then with water. Al tions may be determined, in one way, by titration with though this method gives a highly stable hydrocarbon standardized alkali, after complexing polyvalent metal diluent, it is not suitable for large scale use due to the ions with oxalate. The reagents are a saturated potassi corrosive nature and the expensiveness of chromyl chlo um oxalate solution, 0.1N NaOH standardized against po ride. In a more highly regarded pretreatment, the diluent tassium acid phthalate, and 0.1N HCl standardized is mixed with %-Ao volume of fuming sulfuric acid, against the foregoing NaOH. An aliquot of sample is agitated for one hour, the phases separated, the diluent pipetted into a titration vessel and a small magnetic stir washed with water, neutralized with 0.1-1.0 molar sodium ring bar placed into the vessel. If less than 5 ml. of a carbonate and then given a final water wash. This treat 0.1N NaOH solution will be required to neutralize the ment may be used in conjunction with a silica gel contact estimated acidity of the sample, pipet an NCl spike into 5 5 ing, for silica gel displays a tendency to adsorb olefins and the titration vessel. Next, pipet 10 ml. of the potassium aromatics. oxalate to solution into the vessel, buffer a Beckman The tributyl phosphate extractant also has certain hy automatic titrator and set the pH dial to read 7.0 and drolysis products, di- and monobutyl phosphate, which titrate with the NaOH. The calculation to give the tota tend to strongly complex thorium. The thorium-mono milliequivalents of nitrate ion deficiency in the sample is: 60 butyl phosphate complex is apparently not extracted from the aqueous phase and remains as an emulsifying, col (ml. of basexN of base) – (ml. of spikeXN of spike) loidal precipitate, whereas the thorium-dibutyl phosphate . . Following the adjustment step, the resulting feed solu complex also appears as a colloidal precipitate but tends tion is contacted with the organic trialkyl phosphate solu to follow the organic phase. These TBP-hydrolysis prod 65 ucts may be removed, in one satisfactory method, prior tion. As indicated previously, the most suitable diluents to process use, by washing with a 4 volume of 1.0 molar are petroleum hydrocarbon fractions, especially the satu sodium hydroxide solution followed by three 4 volumes rated hydrocarbons (paraffins and naphthenes). Partic of 0.1 molar sodium carbonate or sodium hydroxide. ularly suitable are kerosene fractions having a specific These pretreatments, in addition to removing potential gravity of about 0.75 gm./cm., a boiling range of 300 emulsifying degradation products and preventing thorium 400° F. and a flash point of about 120° F. Such diluents loss, render the diluent more stable to destructive nuclear are sold under the trade names "Varsol," "Esso 107," radiations, and increase thorium and uranium decon “Shell HFMS," “Gulf BT,” “Atlantic Ultrasene” and tamination from fission products in the extraction cycle. “Shell Sol 72." The "Amsco' class of diluents find highest For example, the iodine decontamination factor goes favor, “Amsco 125-82” being preferred. A Satisfactory 5 from 2-8 with untreated extractant to about 200 with composition range of the organic extractant is approxi pretreated extractant. 3,825,649 11 12 To effect the extraction, the organic extractant is in important for decontamination from the rare earth fission timately, and preferably countercurrently, contacted products, which would be extracted in the presence of a with the aqueous feed solution. Virtually any conven large excess of TBP and which are normally found to tional solvent-extraction contacting means, such as sep undergo extensive reflux in the lower section of the ex aratory funnels, mixer-settlers, packed columns or the traction column. - like may be employed. Remarkably efficient for large The IAP stream may be subjected to a wide variety scale operation are pulse columns (i.e., a vertical column of treatments, depending upon the product desired. For spanned by a plurality of horizontal perforated stainless protactinium recovery, a number of alternative recovery steel plates; the column contents are periodically, sequen schemes are available. One involves the direct extraction tially surged upwardly and downwardly, being thereby of protactinium from an acidified IAP stream. Although turbulently admixed upon jetting through the plate perfo () protactinium is not extractable from nitrate ion deficient rations and being provided with fresh contacting surfaces aqueous solution, it may be selectively extracted from for extraction beyond that expected from simple counter aqueous acidic solutions by an organic solution of a tri current operation). It should be apparent that varying alkyl phosphate or diisoalkyl carbinol, such as diisopropyl flow rates may be employed in column operation while or diisobutyl carbinol, in an inert diluent of the character yet achieving efficient separation, provided compensatory previously described. The protactinium may then be adjustments in column length, contacting time, and feed, stripped from the organic extract with slightly acidic TBP and scrub solution concentrations are made. It is water or preferably, with an aqueous alkali fluoride so generally preferred however, that for extraction, the flow . lution, for instance a sodium fluoride solution. Another rate of the organic extractant should exceed that of the scheme involves the adsorption of protactinium by vari aqueous feed by several times. Generally, deviations of ous solid inorganic adsorbents likesilica gel. Protactini approximately --20% in flow rates may be very satis um appears to adsorb quantitatively on the adsorbent in factorily practiced under the preferred process conditions the absence of interfering ions such as iron, niobium, zir stated below, but for optimum product recovery and de conium and chromium, and resolution of the adsorbed contamination, the exact values should be employed. The protactinium from any adsorbed fission products by se relative flow rates of the various process streams will, 2 5 lective elutriants affords a means of obtaining high con for convenience, be based on a value of 1.0 for the feed centrations of the constituent. Satisfactory elutriants are stream, where units may be in milliliters per minute, liters aqueous acidic solutions; aqueous carboxylic acid solu per hour, gallons per day or relative contact volumes (in tions are particularly efficient, aqueous oxalic acid being batch countercurrent systems). The term "volume flow preferred. ratio" is used as a convenient expression of flow relations 30 The preferred chromate precipitation method for throughout the process system. Considering that the or protactinium recovery comprises adjusting the IAP stream ganic extractant is of specific gravity less than 1, as it is to approximately 0.03-0.1 molar sodium chromate. This with the preferred 42% TBP-58% diluent system, such is concentrated by evaporation to approximately 2.5 that the organic streams tend to rise in columns while the molar aluminum nitrate, 0.5-1.3 normal nitrate ion de aqueous streams descend, the preferred column operation ficient and 0.1-0.5 molar chromate. An aluminum outlined in the flowsheet may be readily appreciated. The chromate precipitate, tentatively identified as aqueous feed solution (IAF stream on the flowsheet) is introduced near the middle of IA column at a volume Ala (CrO) (NO)12HO, flow ratio of approximately 1.0, while the extractant, 42% 40 then forms, carrying the protactinium. Interestingly, this TBP-58% Amsco (IAX stream) is introduced at the bot precipitate will form only in nitrate ion deficient solutions, tom of the column at a flow ratio of approximately 5.0 and not in neutral or acid solutions. The amount of and flows upwardly through the column, thereby effect protactinium carried by the precipitate appears to vary ing extraction of thorium and uranium-233 in the lower with the length of time the precipitate is allowed to re part of the column. An aqueous scrub solution (IAS) of main in contact with the fission-product containing super approximate composition 0.55 molar aluminum nitrate, natant solution; one hour yields an 80% protactinium ad 0.3 normal nitrate ion deficient, 0.01 molar ferrous Sulfate Sorption, while extending the time to 4-6 hours gives and 0.003 molar phosphoric acid enters at the top of the 85%-95% adsorption. A protactinium concentration fac column at a volume flow ratio of approximately 1.0. The tor of about 50 is attained, owing to the relatively small aqueous scrub flows downwardly in intimate contact with volume of the carrier precipitate. The precipitate is sepa the upflowing organic extract, thereby scrubbing the ex rated from the supernatant solution by centrifugation or tract, and upon reaching the feed point, mixes with the other suitable means, and the supernatant is disposed of aqueous feed flowing downwardly through the upflowing as a permanent waste, or as a source of radioisotopes. stream of organic extractant. About a half dozen counter The chromate precipitate can be stored as a source of current scrub stages are all that are required. Naturally, isotopically pure uranium-233, or, for protactinium re if an extractant of greater specific gravity than the feed 5 5 covery, readily dissolved in dilute nitric acid, extracted solution were employed, the points of introduction in the therefrom with an organic extractant of the type pre column would be inverted. The thorium and uranium-233 viously described, and stripped from the extract with containing organic extract (IAU) which is substantially dilute nitric acid or aqueous sodium fluoride. decontaminated of protactinium and fission products, is If uranium-233 rather than protactinium recovery is 60 of primary concern, which of course is normally the case, continuously withdrawn from the top of the column at the aqueous stream may be stored to permit protactinium a volume flow ratio of approximately 5.0 and the aqueous decay (about 10 half-lives are considered, for practical product stream (IAP) of approximate concentration 0.5 purposes, to constitute complete decay) after which the molar aluminum nitrate and 0.3 normal nitrate ion de uranium can be readily recovered by solvent extraction. ficient, and containing virtually all the protactinium and A number of distinct advantages flow from this, ap over 95% of the fission products, is continuously with proach. Higher uranium and thorium losses could be drawn from the bottom of the column at a volume flow tolerated in the extraction step (generally higher fission ratio of approximately 1.8. product decontamination factors are obtainable at a - Under these preferred volume flow ratios, the TBP ca slight cost of product recovery), since this procedure, in pacity of the organic stream provides 5.0 moles of the effect, amounts to a second extraction cycle; less shielding TBP per mole of thorium nitrate, and it may be con would be required; all the uranium would be recovered; sidered that the organic stream becomes about 80% satu and very short-cooled material could be processed with rated with thorium; however, near the feed plate this out fear of loss as protactinium. Minor disadvantages value may reach 95% to 100%, owing to a degree of would be the provision of storage facilities and a final reflux in the scrubbing section. This characteristic is very product not as isotopically pure as uranium derived from

3,825,649 17 18 The IC column had a total length of 36 ft. to provide a avoid any of the unit operation treatments of the prod minimum of five stages. ucts, say the uranium ion exchange purification, or if Irradiated aluminum-jacketed thorium slugs, cooled for products of still higher purity are desired, or if more 210 days, and containing about 1000 grams uranium per ... highly radioactive, short cooled material is processed. ton thorium were dissolved in an aqueous solution of Therefore, our invention should be limited only as is approximate composition 13 NHNO, 0.04 NF-, 0.003 M 5 indicated by the appended claims. Hg and 0.04 M Alt-3. (In view of the long cooling Having thus described our invention, we claim: period, the uranium contained only 0.1%-0.2% Pa, and 1. A process for the separation of protactinium, ura so, after separation, no Pa recovery was attempted. With nium and thorium from an aqueous nitric acid solution 90 day cooling, the Pa concentration would have been of neutron-irradiated thorium containing said elements 5%-10%.) The final dissolver solution had an approx () together with fission products, which comprises contact imate composition of 10 M Th(NO3), 0.4 M A1(NO3)3 ing, under net nitrate ion deficient conditions, said solu and 6.5 M HNO3. The dissolver solution was transferred tion with an organic solution of a trialkyl phosphate in to a feed-adjustment tank, where the excess nitric acid an inert organic diluent, thereby preferentially extract and a portion of free acid were evaporated and the con ing uranium and thorium into the resulting organic phase densate collected for recycle. The residual nitrate ion while the confining protactinium and fission products to deficient thorium and aluminum nitrate solution was di the resulting aqueous phase, scrubbing any small amounts gested for about 1 hour at 155 C. and the resulting of protactinium and fission products from said organic solution adjusted to IAF stream conditions. phase with an aqueous solution of an inorganic nitrate The radiochemical composition of the IAF streams, Salt, separating the said protactinium-containing aqueous indicating the decontamination problems faced, are shown phase from said uranium and thorium-containing or ganic phase and thereafter separating said thorium and in Table VI, below. Said uranium in the separated organic phase. . TABLE WI-RADIOCHEMICAL COMPOSITION OF AF 2. The process of claim 1, wherein said scrub solution STREAMS is an aluminum nitrate solution provided with a small Activity (cts.fmin.fml.) amount of phosphate ion. Example Example Example 3. The process of claim 1, wherein said trialkyl phos Constituent 2 3. 4. phate contains approximately 3-6 carbon atoms among 69X103 1.67X103 1.38X10 each of its alkyl groups. 4.24X106 5.3LX106 6.0X105 8.04X107 1. OX103 7.8x107 30 4. The process of claim 1, wherein said trialkyl phos 2.6X107 1.73X107 1.24x107 phate is tri-n-butyl phosphate. 1.01X107 1.06X10 7. 13X10" 5. The process of claim 1, wherein said inert organic 2.5X106 3.9X106 5.0x108 diluent is a saturated hydrocarbon diluent. 6. The process of claim 1, wherein the organic solu Table VIII below, shows the excellent average overall tion is of approximate volumetric composition 30%- decontamination factors achieved and the radioactivity 35 60% tri-n-butyl phosphate and the remainder an inert, in the products for the three examples. Saturated hydrocarbon diluent. TABLE VII.-DECONTAMINATION FACTORS AND ACTIVETY 7. A process for the separation of protactinium from IN THORIUMAND URANIUM-233 PRODUCTS BASED ON an aqueous nitric acid solution of neutron-irradiated 100 GM UITON THAT 210 DAYS COOLING 40 thorium containing said elements, fission products, and Overall decontamina tion factors, (Activity uranium, which comprises adjusting said solution to ni in IAF stream) trate ion deficient conditions, countercurrently contacting SS Activity in products the resulting feed solution with an organic solution of tri (Activity in product) (cts.finin.fml.) n-butyl phosphate in an inert, saturated hydrocarbon dilu Constitulent Uranium Thorium Uralium Thoriumn 45 ent, at a volume flow ratio in which said organic solution Gross------5. 52X10 ------flow is several times greater than said feed solution flow, 1.3X06 1. OSX10+ l... 43X105 2.31X104 6.6X10 3.6x103 6.62x10 ... 08X103 thereby preferentially extracting uranium and thorium into 8.0x108 1.9X104 E. 13X10 1. 02X1?? the resulting organic phase while confining protactinium 1.7x107 4.1X103 4.31x103 1. 75X1?? and fission products values to the resulting aqueous phase, 6.7x108 2.57x103 7.94x103 1.28X10 175 ------50 Scrubbing any small amounts of extracted protactinium - ? - ? - ? - ? ? 463 ------and fission products from said organic phase with an aque ous Solution of nitrate ion deficient aluminum nitrate pro Table VIII, below, shows the very small uranium and vided with a small amount of phosphate ion at a volume thorium losses to the IAP stream (protactinium-fission flow ratio approximately equal to that of said feed solu products) and the very small IB column uranium loss tion, and separating said protactinium and fission products to the IBT stream (thorium product). 5 5 containing aqueous phase from said uranium and thorium containing organic phase. TABLE VIII.-URANIUMAND THORIUM LOSSES 8. The method of claim 7, in which the feed solution: IA columni IB c0]um?In organic Solution: aqueous scrub solution volume flow ratio Percent Percent Percent is approximately 1:5:1. U lost Th l0St U l(ost 60 9. The process of claim 7, wherein said protactinium is recovered from said separated protractinium and fission 0.039 0.19 ... O (). O6 0.14 0.4 products-containing aqueous phase by acidifying said 0.09 0.09 U?? solution and contacting the resulting solution with an or ganic solution of an extractant selected from the group consisting of trialkyl phosphate and diisoalkyl carbinol in The foregoing examples are merely illustrative and aninert, Saturated hydrocarbon diluent, separating the re should not be construed as limiting our invention. In sulting protactinium-containing organic phase from the re particular, it should be understood that changes may Sulting fission products-containing aqueous phase, and be made in process variables by those skilled in the art without departing from the spirit of our invention. Even stripping said protactinium from said organic phase with if any such changes result in products of decreased de an aqueous Solution. contamination, this may be remedied by additional Sol 10. The process of claim 7 wherein said proctinium is vent extraction cycles. In the same vein, additional Sol separated from said fission products in said separated vent extraction cycles employing our herein disclosed aqueous phase by providing said solution with chromate conditions may be performed if it is desired either to ion, and separating the resulting protactinium-carrying

3,825,649 21 22 while confining said protactinium and fission products to 21. The process of claim 20 wherein said neutron-irra the resulting aqueous phase, scrubbing any extracted pro diated thorium is initially in the form of aluminum-jack tactinium and fission products from said organic phase eted thorium metal and is dissolved in an aqueous acidic with an aqueous scrub solution comprising approximately solution of approximate composition 13 molar nitric acid, 0.5 molar aluminum nitrate, 0.3 nitrate ion deficient and 0.04 molar fluoride ion and 0.003 molar mercuric ion. 0.003 molar phosphate ion at a volume flow ratio of 5 approximately 1.0, separating said protactinium and fission References Cited products containing aqueous phase from said uranium and thorium-containing organic phase, contacting the sepa UNITED STATES PATENTS rated organic phase with an approximately 0.2 molar nitric 10 2,897,046 7/1959 Bohlmann ------423-10 acid solution at a volume flow ratio of organic: aqueous 2,894,806 7/1959 ElSon ------423-10 X phases of approximately 6:5, thereby preferentially strip 2,789,878 4/1957 Peppard -.-.-.-.-.-.-.-.-.-.-.- 23-14.5 ping said thorium into the resulting aqueous phase while OTHER REFERENCES confining said uranium to the resulting organic phase, Gresky, "Proceedings of the International Conference countercurrently scrubbing any extracted uranium from 5 on the Peaceful Uses of Atomic Energy,” vol. 9, pp. 505 said aqueous phase with additional of said organic solu 510, held in Geneva Aug. 8-20, 1955, United Nations, tion, separating said thorium-containing aqueous phase N.Y. from said uranium-containing organic phase, countercur Kleinberg AEC Document AECD-3674, pp. 12-6, rently contacting the separated organic phase with an Sept. 10, 1954, declassified Sept. 20, 1955. approximately 0.05 molar aqueous nitric acid solution at 20 a volume flow ratio of organic: aqueous phases of approxi Thompson, AEC Document AECD-1897, pp. 1-4, mately 6:3, thereby stripping said uranium into said aque February 1948, declassified Apr. 15, 1948. ous phase and separating the resulting uranium-containing Ward, AEC Document AECD-2524, p. 6, Aug. 7, 1947, aqueous phase and the resulting uranium-depleted organic declassified Mar. 11, 1949. phase; concentrating said separated, thorium-containing aqueous phase by evaporation; contacting said separated, CARLD. QUARFORTH, Primary Examiner uranium-containing aqueous phase with a comminuted R. L. TATE, Assistant Examiner organic cation exchange resin bed characterized by a plu lU.S. Cl. X.R. rality of nuclear sulfonic acid groups, and selectively elut ing and collecting the resulting adsorbed uranium from 30 423-7, 10, 19, 250, 252, 253; 252-301.1 R. said bed with an aqueous acetate solution. UNITED STATES PATENT OFFICE CERTIFICATE OF CORRECTION Patent No. 3,825,649 Dated July 23, 1974 Inventor(s) A. T. Gresky et al. It is certified that error appears in the above-identified patent and that said Letters Patent are hereby corrected as shown below: Column l, line 33, "323" should read - -233--.

Column 8, line 43, "21.25" Should --12.25--; line 46, "14" Should read --149--; line 47, "15" Should read --157--. Column 9, line 59 "tota" should read --total.--. Column 10 , l ime 1 1 , 'withut" Should read --Without--. Column 5, line 50, "Steam" should read --Stream--; 'rate" should read --Ratio- -; ime 63, º IBX º Should read --IB X--. Column l6, line 15, "IAP" Should read --IAU--; line 19, "IAU" Should read --IAP--. Column 17, line 39, "100 GM" Should read --1000 GM--. Column 18, 1 ine 16, "the" should be deleted; line 72 "proctinium" Should read -- protactinium--. Column 19, line 13, "rate" should read --ratio--; line 24 "separate" should read -- separated.--; line 26 "ntirate" should read --nitrate-- eigned and Sealed this fourteenth D ay of October 1975 SEAL Attest

RUTH C. MASON C. MARSHALL DANN At testing Officer Connissioner of Patents and Trademarks

UNITED STATES PATENT OFFICE CERTIFICATE OF CORRECTION Patent No. 3,825, 649 Dated July 23, 1974 Inventor(s) A. T. Gresky et al. It is certified that error appears in the above-identified patent and that said Letters Patent are hereby corrected as shown below: Column li, li me 33, "323" should read - -233--.

Column 8, line 43, "21.25" should --l2.25- - ; line 46, "14" should read --l49--; line 47, "l5" should read --57--. Column 9, line 59 "tota" should read --total.--. Column 10 , l ime l . , "withut" Should read --Without--. Column l5, line 50, "Steam" should read --Stream-- "rate" should read --Ratio- -; line 63, "IBX " Should read -- IBX--. Column 16, 1 ine 15, "IAP" Should read --IAU--; 1 ine 19, "IAU" Should read --IAP--. Column l 7, line 39, "100 GM' Should read --1000 GM--. Column 18, 1 ine 16., "the" should be deleted; 1 ine 72 "proctinium" Should read -- protactinium- - . Column 19, line 13, "rate" Should read --ratio--; ime 24 "separate" should read -- separated--; line 26 "ntirate" should read --nitrate-- eigned and eealed this fourteenth Day of October 1975 SEAL Attest:

RUTH C. MASON C. MARSHALL DANN At testing Officer Commissioner of Patents and Trademarks