MALRTIN MARIETTA ENERGY SYSTEME LIBRARIES DTN 5k 031350 b ORNI~2661 Contract No. W-TLOS-eng-26 CHEMICAL TECHNOLOGY DIVISION Chemical Dévelapment Section A THE FUSED SATT-FLUORIDE VOIATILITY PROCESS FOR RECOVERING URANIUM ¢. T. Cathers, M. R. Bennett, R. L. Jolley DATE ISSUED APR 11959 OAK RIDGE NATIONAL LABORATORY Osk Ridge, Tennessee Operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION DA 3 445k 0361350 & ABSTRACT A fluoride volatility process for recovering enriched uranium from UF)-NaF-ZrF) melts, produced by the dissolution of reactor fuel elements in molten NaF-ZrF), through hydrofluorination, is described. The UF) is fluorinated to UFg, which volstilizes, and is separated from volatile fission products in a NaF absorption- desorption cycle. In laboratory studies gross bets and gamms decontamination factors of 10” with a urenium loss of less than 0.1% were secured. The product UFE may dbe conveniently returned to UFYy or metal for further use as reactor fuel. Chemical studies, supplementary to process development work, showed that impurities, possibly oxides, have a much grester effect in the fused salt-fluorination step than such factors as use of nitrogen with the fluorine, or the method of introducing the fluorine into the fluoride melt. The colloidal behavior of NiFo, a slightly soluble corrosion product formed in the fluorination, indicated that this materisl would not interfere in molten salt handling if thbe fused salt was not allowed to stand without agitation for prolonged periods. The absorption of UFg in NaF was found to be due to the formation of a UFg-NaF complex. The egquilibrium between gasecus UFg and sclid complex was established for the temperature range of 80 to 320°C. Decomposition of the UFg~-NalF complex to a UF5wNaF complex dees not lesd to appreciable uranium loss in the process 1f specified process conditions are maintained. CONTENTS Page 1.0 Immmcmm L 2.0 DESCRIPTION OF PROCESS h 3.0 PROCESS DEVELOPMENT STUDLES 6 3.1 Fused Salt-Fluorination Work 6 3.1.1 Fluorine Efficiency 6 3.1.2 Corrosion Studies 9 3.1.3 Recovery Yield of UFg | 1L 3.1.4 Behavior of NiF, in Corrosion | 15 3.2 NaF Decontemination Step | 19 3.2.1 Operation iu = Temperature Range of 100-400°¢C 19 3.2.2 Operation st a ‘I‘émperature Above 600°C 23 4.0 BASIC CHEMICAL STUDIES | 26 L.1 NaF Cepacity for UFg é6 4.2 Pressure of UFg Above UFg-+3Na¥ Complex 26 4.3 Decomposition of UFge3NaF Complex 29 5.0 RECOMMENDATIONS 32 6.0 REFERENCES | 33 1.0 INTROLUCTION A new approach to processing of enriched uranium civiliau power reactor fuels by a flvoride volatility method bas been reportedfilfg The mathod consists in three steps:; dissolution of the metal or elloy in a fluoride melt by hydroflueorination, volatilization of UFg from the molten selt through fluorination, and fipal purification of the UFg from volatile or eptrained fission product fluorides by absorption or distillation. A typical salt composition is 50-50 mole % NeF-ZrF),, with a melting point of about 510°C. The second step eppears feasible with use of either elementel fluorine or bremine pentafluoride. The ORNL development program on this process has been directed toward use of elemental fluorine in the second step and absorption of the UFg on Ne¥ as a mesns of completely deconteminating the URg product from fission product sctivity. A practiceble process flowsheet for the fused selt-fluorination and NaF decontemination steps is described in this report along with the results of laboratory process test studies. 1In addition, the status of some of the more basic development work carried on con~ currently with the process studies is presented. This supporting work bhas included further study of the fluorination step, some of the chemistry involved in the WeF decontamination process, apd exploratory work on the corrosion problem. 2.0 DESCRIFTION CF PROCESS The reccmmended flowsheet for the fused salt-volatility process (see Fig. 2.1) hes the following festures: Fused Salt-Fluorination Step 1. A Fo/UF), mole ratio of 6/1 results in essentially complete UFg volatilization at 600°C. The wvolatilization is over 90% complete at a mole ratio of 3/1. If the fused salt contains an vwnusiually bigh oxygen content (oxymcompounds) a ccupensatory emount of Fo is needed. 2. The optimum fluorination pericd is probably about 2 hr. For 8. 10-kg batch of uranivm a Fo rate of 1.7 scfm is required. A bhigher flow rate over a shorter period would incresse the Fo efficiency at the expense of encountering excesgive mechanical entrairment of the salt and a lower decontamination effectiveness. A lower Tlow rate would only leasd to magnification of the corrosion problem. UNCLASSIFIED ORNL-LR-DWG 34376 F.P COLD TRAP < 80 % Ru activity 2 o i c F o CJJ FLUORINE FEED 2 x| DISPOSAL < Bwt % inw 50-50 mole % NaF-ZrFy4 @ © <7 \ lst NaF BED 2nd NaF BED FL UORINATION L - S om | UF~+F, Absorption: . Fp/UFq mole mg‘;’oifrf” 570-630°C 4 SF mig2 | 160°C, 2hr g Desorption: UFs COLD TRAP F - = E fi I00-400°C BD.F ~ 105 | x 3 f hour » D.F. ~s103 ¥ Desorption: @ O WASTE 100-400°C e < 0005wt % U i hour > 99 % gross beta and | > UFg STORAGE gamma activity — — WASTE D.F~Decontaomination Factor <002 wt % U F.P-Fission Products >80 % Nb activity Fig. 2.1, Flowsheet for Fused Salt-Fluoride Volatility Process. NaF Absorption~Desorption Cycle 1. A N&F/U‘Weight ratio (U being the total uranium being pro- cessed) of 2/1 to 3/1 is needed in each of the two NaF beds. The recommended grede of NeF is 12-20 mesh, prepared from pelletized material (Harshaw Chemical Company). The NaF beds should be preconditioned with a slow flow of Fo for 1 hr at 400°C before pProcess use. 2: The absorption cycle, using only the first bed, is carried out at gpproximately 100°C. Overstion on a large scale will result in the bed temperature rising 50° or more due to heat of abscrption. There is no reason why the temperabure could not be closer to the triple point of UFg (65°C) initially to partly compensate for this effect, 2. Descrption of UFg from the first bed through the second bed requires about 1 hr, using the same ¥, flow rate employed in the absorption cycle. Ths desorption cycle consisted, in labora- tory tests, of raising both NaF beds simultaneously from 100 to LOOBC inm about 0.5 hr, at which time the transfer of UFs to the cold trap system was essentially complete. 3.0 PROCESS DEVELOFMENT STUDIES 3.1 Fused Sali-Flucrination Work 3.1.1 Fluorine Efficlency When fluorine is introduced into fused Na¥F-ZrFi-UF), the uranium content drops sharply as the Fo/U mole ratio incresses (Fig. 3.1). Fluorine utilization efficiency is highest when 90% or.more of the UFg has been volatilized {Fig. 3.2). The efficiency decrasses thereafter, with the Fo acting essentially as a sweep gas. In an ideal case, the amount of UFg volatilized would be stoichiocmetrically equivalent to the amount of Fp introduced up to a Fo/U mole ratioc of 1, thereafter decreasing hypserbolically. The Fg/U mole ratio required for volstilization of more than 99% of the UFg was decreased by the elimination of impurities, but it was not significantly affected by the concentrstion of uranium in the initial fused salt (Fig. 3.3). Use of No with the Fp, the method of gas intrcduction into the melt, and the rate of gas flow hes some effect on the Fg/U’lee UNCLASSIFIED UNCLASSIFIED QRNL.A R-DWG. 1C827R1 ORNL-L2fW5. 10B2BRT $00 1.0 50 . \_ — S 0.8 “ ------- - —te 20 8 \ 5 2 3 2 N 0 = \ = = =z \ . o ? , ; w06 fe- \ S CESV I S s e \ | - : 0 \ 2 o —t— W a \ z 2 N \ z Joa | \ _ = B N Z a \. g > \ a " e N 2 o \ 0.5 o ® |_ é 02 L . . .“, [ __k e 2 o 0.2 = - / /‘ 04 * 0 ,——.—Jw——. l 0 1 2 3 4 5 0 1 2 3 4 5 6 MOLE RATIO OF FLUDRINE TO TOTAL. URAN'UM MOLE RATIO OF Fp INTRODUCED TO TOTAL URANIUM Fig. 3.1. Amount of UF4 Remaining in 375 g of NaF~ZrF4-UF4 (50-46-4 mole %) Fluorinated at 600°C at a Rate of 100 ml/min Fig. 3.2. Efficiency of UFg Volatilization as a Function of Amount of Fluorine Intro- as a Function of Amount of Fluorine Introduced. duced. UNCLASSIFIED ORNL-LR-DWG., 12611 { / ! r i / | ’ 80 S . | | | £ | | 0 60 i ‘ ™ | b | 'L<'£ ; 2 Q 40 | ‘ © l | W ! | 20 x \ | [ ’ O l. e e 1 2 3 4 Fo /U MOLE RATIO RUN U CONTENT NO. MOLTEN SALT (mcle %) 1 AS-RECEIVED NafF-7rF,~UF, (50-46-4 mole %) 4 6 NaF - ZrF, - UF, (50-46-4 mole %), HYDROFLUJORINATED 4 hr AT 800°C 4 7 PREFLUORINATED NofF-ZrF, (50/46 mole RATIO) PLUS UF, (76.29%% URANIUM) 4 8 PREFILUCRINATED NaF-Zrk, (50/46 mole RATIO) PLUS UF, (76.2 %% URANIUM), AIR-SPARGED FOR 2hr BEFORE FLUORINATION 4 o PREFLUORINATED NaF-Zrf, {(50/46 mole RATIO) PLUS UF, (72.99%% URANIUM) 4 3 PREFLUORINATED NaF - ZrF, (50/46 mole RATIO) PLUS UF, (72.9 %" URANIUM) 4 4 PREFLUORINATED NarF-ZrF, {50/46 mole RATIO) PLUS Uk, (72.9%" URANIUM) o 5 PREFIUCORINATED NoF~ZrF, (50/46 mole RATIO) PLUS UF, (72.9 %" URANIUM) 1 TTHEORETICAL = 75.8 % Fig. 3.3. Effect of Uranium Concentration and Impurities of the Fused Salt Fuel Mixtures on the Fluorine~to-Uranium Mole Ratio Required for UF, Volatilization., Conditions: 100 m] of F, per min; 1/16~in,~dia sieve plate on dip tube thaéinfroduced fluorine to melt. ratio, but the results could not be correlated with the known variables (Fig. 3.4). These experiments were performed at 600°C in a 2-in.-dis nickel reactor with a 375-g charge of NaF-ZrF,-UF,. In some of the tests the salt was made by the addition of UFy to Na¥ -ZrF), (50/46 mole ratio) that was believed to be relatively free of oxide impurities as a result of previous use in a fluorination run. Uranium tetrafluoride concentrations of 1, 2, and 4 mole % were used to study the effect of concentration. In other tests NaF-ZrF)-UFy, (50-46-U4 mole %) was used as received. Data were obtained by direct sampling of the salt at intervals during the fluorination., The curves were extrapolated to the 100% volatiliza- tion point for comparison, but usually a sharp break was observed in the curve between 95 and 100%, which extended the curve to higher Tluorine~to~uranium mole ratios for volatilization of the last traces of UF¢. Volatilization of more than 99% of the UFg from as-received NaF-2rF),-UF), required a fluorine-to-uranium mole ratio of about 3.1/1,'which was reduced to about 2.2/1 by sparging with HEF for I hr before fluorination. In two tests with the fuel mixture synthesized by adding UF}, with a uranium content of only 72.9% (theoretical, 75.8%) to prefluorinated NaF-ZrF), the fluorine-to- uranium mole ratio required for more than 99% UFg volatilization was about 2.4/1. When very pure UF|, uranium assay of 76.2%, was used, the fluorine-to-uranium mole ratio was 1.4/1, which represents a fluorine utilization efficiency of about T0%. A quick~Treeze sampling technique was used in all of the fluori- nation work. A comparison of sampling methods showed that agreement to within 3% was obtained in uranium analyses of samples taken ,during the course of fluorination experiments with the use of a dip ladle, immersion of a solid rod into the fused salt to obtain a quick~freeze sample, and samples taken after fluorination by grinding and sampling the entire batch of salt. This study was made with three different uranium concentratlions in the Nal’-ZrF) salt, 8, 2, and 0.5 wt %. 3.1.2 Corrosion Studies The corrosion of nickel test couponsg and of a nickel vessel was fairly low after 20 fused salt-fluorination runs at 650°C, confirming previous work.” Since conditions changed continually during the runs and since the various components of the vessel were attacked to different degrees, a calculated over-all corrosion rate would have no significance. However, it appears that a large nuiber of fluorination runs can be made in one reaction vessel before the 100 UF, VOLATILIZED (%) -10~- UNCLASSIFIED ORNL-LR-DWG, 12612 80 e F, /U MOLE RATIC RUN FLOW RATE (ml/min) NO. Fo No 10 100 0 5 40 200 4 100 G 2 {00 200 3 150 0 4 150 150 & 300 o 7 100 0 8 100 200 9 150 O 1 100 Q 12 100 O GAS DISPERSION DEVICE ON END OF /,-in.~DIA DIP TUBE NONE SIEVE PLATE, 3/g, -in.-DIA HOLES SIEVE PLATE, %4 -in -DI& HOLES SIEVE PLATE, ¥4 -in -DIA HOLES SIEVE PLATE, 3/ga-in.-DIA HOLES SIEVE PLATE, 34 -in-D!A HOLES SIEVE PLATE, 3ga-in-DIA HOLES SIEVE PLATE, /g ~in.-DIA HOLES SIEVE PLATE,'/yg -in.-DIA HOLES SIEVE PLATE, ¥/4g -in.—DIA HOLES THREE SIEVE PLATES, 3/gq-in.-DIA HOLES, '/ in. APART PERCCLATOR DRAFT TUBE Fig. 3.4. Effect of Sparge Gas Flow Rate and Method of Intreduction of Fluorine into the Melt on UF 6 Volatilization. -11- corrosion is 00 severs. A summary of the resistance of various metals to fluorination conditions st high temperatures has besn reported elsevhere. The "A" nickel reaction vessel was 2 in. in diameter. The three fest coupons were mounted in an upright position at the bottom of the reaction vessel, as shown in Fig. 3.5, in suck & way that one-third the surface area of each coupon extended from the liquid into the gas phase. The coupons were 3 in. long, 3/ in, wide, apd 1/4 in. thick. Two of the coupons were "A" nickel (nominel composition: 99.4% Ni, 0.05% C), and one of them was cat longitudinally and welded. The third coupon, which was "L" nickel (nominal composition: 99.4% Ni, 0.01% C), was also cut longitudinally and welded. Each run was made with 200 g of NaF-ZiF)-UF) (50-46-% mole %). The time for a run varied from 4.58 to 0.83 hr, the resction vessel and the coupons belng exposed to process copditions for a totel of 30 br. The fluorine flow rate varied from 50 to 300 mi/min and was regulated so that 9.4 moles of fluorine was used per mole of uranium in each run. Corvosion of the welded coupons (both "A" and "L" nickel) was scmewhat grester than that of unwelded ones, but in bolh cases the corrosion was of the solution type (Fig. 3.6), and there was fairly uniform surface removal. Dimensional and weight-change analyses also showed that corrosion may have been slightly grester in welded than in urwelded coupons (Tsble 3.1). The most severe attsck was on the outer surface of the flucorine gas inlet tube in the vapor zone (Fig. 3.7). It is very likely that this attack, about 2 in. above the salt surface, was due to the freguent admission of atmos- pheric molsture and oxygen Intec the reactor possibly producing an aqueous HF and oxidation attack when if was ab an elevated tempera- ture. The same type of attack 4did not occur at the resctor wall. Corrosion on the Fo Inlet tube in the liquid zore was more uniform and varied from 4.0 to 7.5 mils in depth. The reaction vessel showed nonuniform attack of 5 to0 9 mils in both the liquid and gas zones; in the region in contact with molten salt the atiack was of a solution nature {Fig. 3.8). | 3+.1.3 Recovery Yleld ofVUFé Uraniuwm hexafluoride recovery was more than 99,0% in the 20-run corrosion series (Sec. 3.1.2). The recovery of uranium was nigh in all runs (Table 3.2). The wrenium loss in the waste salt wes con- sistently lowest in the 50-min runs at the highest fluorine flow rate. This result was possidbly due to & smeller loss of fluorine in corrosion -12- UNCLASSIFIED Y-14847 Fig. 3.5. Cross Section of Assembled ''A" Nicke! Reaction Vessel. The pitting on the fluorine gas inlet tube may be seen at point A, B UNCLASSIFIED ! Y-15140 | v * * Fig. 3.6. Cross Section of Welded "'L"" Nicke! Test Coupon Exposed to Melten Salt in a Nickel Reaction Vessel. Note uniformity of attack. Etched with KCN plus (NH,),$,0g. 12X. Table 3.1. Welght Loss of Nickel Corrcosion Coupons Tested in Laborataryw8¢ale Flucrination Runs Original Final Weight Change Weight Welght - Type of Coupon (g) (g) (2) (%) Welded "L" nicksl 83.9878 80.3760 3.6118 L.3 Welded “A" nickel 86,3445 82,7515 3. 5930 Yoo Unwelded "A" nlckel 82,6071 80.2160 £.3911 2.9 ~14- UNCILASSIFIED Y-14850 Fig. 3.7. Outer Surface Attack of Fluorine Gas Inlet Tube in Vapor Zone of Reaction Vessel, Section taken at point A of Fig., 3.5, Etched with KCN plus (NH,),5,0g. 20X, e UNCLASSIFIED P : Y-15139 Fig. 3.8. Inner Surface of Specimen of ""A'' Nickel Reaction Vesse! Taken From Region Exposed to Na F—ZrF4"UF4 Fuel. Note nonuniform surface attack. Ftched with KCN plus (N H4)25208' 250X, -15- in the short runs than in the long runs. Out of 3935 g of salt, 341 g of uranium was recovered as UFg, which corresponds to an initlal uranium content of 8.66%. Analyses of this batch of fuel ranged from 8.30 to 8.76% uranium. Even if the higher velue is assumed, recovery was 99.0%. 3.1.4 Behavior of NiFs in C@frosion The behavior of NiFy in molten NeF-ZrF) and Ne¥F-ZrF|-UF) systems was studied to determine if the presence of this corrosion product would form sludges which would interfere with salt transfers in the fluoride volatility process. Although WNiFo has been reported to be fairly insoluble in this type of salt (approximately 0.2 wi % as Ni at 600°C), at higher concentrations NiFo readily forms a viscous dispersion which settles slowly. Based on this observation, it was concluded that NiF, concentrations up to 2 wt % would not interfere with salt transfers unless the molten salt were permitted to stand, unagitated, for long periocds of time. Anhydrous NiFp was added to molten NaF-2ZrF) (50«50 mole %); the mix was heated until a clear sclution was obtalned and then cooled until turbidity reappeared. Solubility values (Table 3.3) estimated by the disappearance of turbidity were in fairly good agreement with those determined electrochemically on 53 mole % NaF-47% ZrF) salt. Additicn of as much as 6 wbt % NiF, to molten NaF-ZrF) (50-50 mole %) at 600°C resulted in the formation of a viscous dispersion which was falrly stable. Although some settling of NiFo was evident after only 0.5 hr with an initial nlckel concentration of 2 wt %, complete settling had not occurred even after 72 hr (Table 3.4). With 1 wt %, settling was more nearly complete at T2 hr since the Ni¥o concenbration and viscosity at the bottom could not increase as much. The results for the Nalf-ZrF|~UF) system appear very similar to those for the uranium-free system (Table 3.5). However, with 2 wt % Ni and with uranium present the settling was less after 2 hr than in the test with no uranium. Since the solubllity of NiFo is represented by the lower 1limit of nlckel concentration encountered in the settling tests, it appears to be approximately the same in uranium~-bearing snd uranius-free salt. The tests were made by dry mixing the required smount of salt (~ 30 g), and melting in a 1/2-in.-i.d. nlckel tube. MNitrogen was used initially for agitation and then as a blanket while the material was kept at 600°C for varicus times. The tube was quickly guenched with cold water at the end of the test to fix the NiFo concentration at varicus heights in the tubs. The tube was then cut into 1/2m1n0~ long sections and the salt was analyzed for nickel. ~16- Table 3.2. Uranium Losses in laboratory-Scale Fluorination Runs Uranium Loss Number of Duration Fluorine Flow Rate in Waste Runs (nr) (m1/min) (% of Total) 1 .58 55 0,11 5 2,50 100 0.02 to 0,16 5 1.25 200 0.06 to 0.23 9 0.83 300 0.01 to 0.0k Table 3.3. Visual Determinations of Solubility of NiFs. in NaF-ZrF) (50-50 mole %) Temperature Solubility (°c) (wt % NiFy) 640 0.7 670 1.0 685 1.3 -17- Table 3.4. Sedimentation of NiF, in Molten NaZrF s Temperature: 600°C il PR Ni Concentration (wt %) Relatlive Position of ' Initial After After After After Ssxple Content 0.5 hr 2 hr 8 hr T2 hr Initial Ni Content® - 2 wt % 1 (top) 1.60 - 0.28 0,20 - 2 1.74 0.73 0.30 0.22 0.20 3 1.72 1.72 2,20 1.40 0.21 L 1.78 1.86 2.62 3.48 1.23 5 (bottom) 2.13 2,02 3.02 3.54 2.99 Initial Ni Content® - 1 wt % 1 (top) --- --- ~—- - 0.22 2 ——— ——— 0.30 - 0,16 3 - - 0.31 —— 0.17 b - — 1.71 - 0.18 5 (bottom) - - 3.06 - - aThe nickel was added as NiFeo 18~ Table 3.5. Sedimentation of NiFo in Molten NaF-ZrF)-UF), (48-48-4 mole %) Temperature: 600°¢ Ni Concentration (wt %)- Relative Initial Ni, 2 wt % Initial Ni, 1 wt % Initial Ni, 0.5 wt % Pogition of After After After After After Sample 2 hr 48 hr 2 hr 48 hr 48 hr 1 (top) 1.26 - ——— - e - 2 1.23 0.25 0.36 0.18 0,24 3 2,12 0.34 0.40 0,20 0.33 ) 2.78 2,75 2,53 0. 47 0.22 5 {bottom) 2.81 6.94 3.08 - 0.85 -19- 3.2 HNaf¥ Dacontaminafiion Step 3.2.1 Qperation in a Temperature Range of 100-k00°¢ The flowsheet (”ig, 2.1) for the volastiiity process provides for volatilizing UFg from molien fluoride salt with fluorine, shsorbing the UFg on NeF at 1009C, then desorbing with fluorine at 100-400°C and pessing the desarbea UFg through & second WaF bed to the finel cold trap. In the case of long-decayed uranlum resctor fuel, most of the volaiile achbivity 1n the UWFg stresm from the molten salt step is due to ruthenium and nicbium, both of which form wvolatile pentafluorides. Laboratory tests have demonstrated that ruthenium is not absorbed very much on NaF at 1009C, but effectlvely passes throvgh the first NaF bed with the excess Tluorine used in the molten salt step. Nioblum, on the other hand, is absorbed on the NaF with the UFg. This sbsorpiion is predominantly irrveversible siunce the nicblum remains for the most part with the Na¥ during U’5 desorption. Use of the second NaF bed appears essential to prevent cross contamination and to achieve effective decontamination of the UFg (particulerly from any ruthenium revolatilized from the end of the first bed) in processing consecutlve batches. Much of the rubthenium remelining in the first Na? bed is "plated out” over all of the metal surface; including that wuear the outlet., This preblem has also been encountered in dlstillistion work. The effective absorpiion of ruthenius activity on NaF et high temperatures (see Sec. 3.2.2) is perheaps partly responsirle for the efficiency of the secan& bed. in six tests of the double-bed procedure (using 2040 g uranium), the mcbtlivity of the product Ufg was less than the Ui)~-UXp activity normel in nsturel uranium. Four of the tests were mede consecubively with the same 60-ml Naf beds end showed that the decontemination effectiveness of the system does not decrease with use (able 3.6). The over-ell beta- or gamma-decontamination factor in each of the six runs was no less than 105 with 102 veing sttributeble to the fused salt-fluorination stepl and 103 to the gbsorption-desorption process. The low product activity made calcilation of specific decontamination factors imp“actlcal. The effectivensss of the double bed system wss shown in the six tests by the distribution of the volatilized activity (Teble 3.7). An, over-all beta- or gamma»dficonuamination factor of gresler than lQ% was obbtained by the asbsorption-desorption procedure using 200 ml WeF 1n s single bed (Table 3.8). Calcuwlatlon of speeclfic decontamination factors was possible because the producht UFg was radicactive in excess of the Uk -Up level. The typlesl behavior of the ruthenimn and rdobium activitles was also observed iam this 1un. D0 - Table 3.6. Summary of Four Consecutive Runs in Two-Bed Fused Salt Fluoride-Volatility Process Conditions: 128 g of uranium in NaF-Zr¥)-UF), (52-44-U4 mole %) with gross beta activity per milligram of uranium of 5 x 10° counts/min. Each run fluorinated with 1/1 Fp-Ny mixture for 1.5 hr and then with pure Fs for 0.5 hr; UFg in Fo-No gas stream absorbed on NaF; UFg desorbed at 100-400°C through second NaF bed into a cold trap Average Fp/U mole ratio in absorption period: L4/1 Average Fo/U mole ratio in desorption period: 2/1 Absorbent beds: 60 ml of 12~ to 4O-mesh NaF in l-in.-dia tubes NaF/U weight ratio after four runs: 1/1 Uranium Retention (%) Product Gamms Activity Product First First Second per Milligram Yield Cold NaF NaF Waste of Uranium® Run (%) Trap Bed Bed Salt (cts/min) 1 83 0.01 0.02 3.6 2 35.2 0,10 0.05 3.1 3 151 0.07 0.08 1.0 Ly 3.8 0.08 0.02 2ol Over-all 70.1 0.06 0.5 5.1 0.0L 2.5 & Gamme, activity per milligram of natural uranium is 8 cts/min. Table 3.7. Distribution of Volatilized Activity in the Two-Bed NaF Procedure Runs 1, 2; 3, and b: Consecutive tests with two 60-ml NaF beds (12-40 mesh) in 1-in.-dis tubes. Total NeF/U welght ratio for both beds in each run: 4/1. NeF/U ratio over four rums: 1/1 Runs 5 and 6: Single-batch runs with two 90-ml NaF beds (12-40 mesh) in l-in.-dis tubes. Total NaF/U weight ratlo for both beds: 6/1 FPercent of Total Volatilized Activity Runs 1, 2, 3, snd b Run 5 Run 6 Fission First S@c@nd' Figsion First Second Figsion Plrst Becond Product NaF Na¥ Product NaF NaF - Product NaP NeF i Activity Cold Trap Bed Bed Cold Trap Bed Bed Cold Trap Bed Bed §? drogs bats, 81 158 0.85 51 W 0.8 59 26 17 Gross gamms, 11 89 Dells | 3 G7 J.07 7 93 0.02 Ru gamms, 97 1.6 1.1 81 18 0,9 86 1 Very low Zr-Nb gamms, 2.2 98 0.0k Dk w100 0,0h 0.8 99 0,02 Tybal rars-earth beta 4.3 o2 3.6 3 57 0ol 3 97 Very low Table 3.8. Decontamination of UFg in the Single-Bed Fused Salt Fluoride-~Volatility Process UFG in Fo-Np gas stream from fluorination of NaF-ZrF)-UF), (gross beta activity per milligram of uranium in salt = 5 x 105 cts/min) at 600°C; absorbed on NaF at 100°C, desorbed with excess Fo by increasing the temperature from 100 to L00OC Absorbent: 200 ml of 12- to LO-mesh NeF in 1l-in.-dia bed Total Fo/U mole ratio: =~ § NaF/U weight ratio: =~ 6 Product yield: 87% Decontamination Factors Over-all, Including Activity Absorption® Desorption® Fluorination Gross beta 2.1 40 1.2 x 104 Gross gamma 1.2 310 1.k x 10% Ru gamma 2.k L6 1100 Zr-Nb gamms, 1,0 1600 5.9 x 10% Total rare-earth beta 1.0 5 ®Based on activity not absorbed with UFg on NaF but passed into cold trap. bBased on activity remaining on NaF after desorption of UFg. -23- The uranium retention on the two Nal’ beds In the four-run series was respectively 0.5 and 5.1% (Table 3.6)., This excessive loss was due to some back-pressure buildup, which was evlident in all runs, as the result of either partial plugging of the NaF beds or a stoppage in the UFg cold trap. It is essential to avold any prolonged restriction of the gas flow, which would result in a major part of the UFg remsining in the bed as it approaches 400°C (see discussion of decomposition effect, Sec. %4.3). For example, three successive complete cycles with UFg in a two-bed NaF system, where little plugging occurred, resulted in retentions of 0.04 and 0.01%, respectively, in the two beds. In a single-cycle test with the same equipment, the uranium retentions were 0.05 and 0.01% respectively. The per cent of loss, therefore, does not appear to depend on the number of cycles, 3.2,2 Operation at a Temgérature Above 600°C Fair decontamination was obtained with NaF in either a 9- or an« 18-in.-long bed at 650°C when the NeF/U welght ratio was 3/1 (Table 3.9, Runs 1, 2, 3) or 6/1 (Table 3.10, Runs 1 and 3). When the 9-in.-long bed was re-used (Table 3.9; Runs 3 and 4), so that the over-all NaF/U weight ratio was 1.5/1 for both runs, deconbami- nation factors for gross beta, gross ganme, and ruthenium dbeta decreased sharply. With the 18-in.~-long bed the same effect was observed, although here the over-all NaF/U weight ratio for the two runs was 3/1 (Table 310, Runs 1-2 and 3-4). The imperfect decontamlnation of the UFg 1n any single run and the ruthenium absorbed in one run leads to cross-contamination cver consecutive runs (illustrated,by sharp decrease in ruthenium gamms-decontami- netion factors in subsequent runs). However, the high effectiveness of NaF in removing rutheniwm st > 600°C was in contrast to the much smaller ef?ect observed in a UFg absorption cycle at 100°C (see Secs 3.2.1). Use of NaP at 600-650°C was discontinued because of the excessive corrosion, the diffieulty with which the tempersture was maintained, and the great variation of uranium retention on the bed. The best results were secured with a high flow rate for the UFg-Fo gas stream Trom the molten salt step. When the flow rate was decreased 1n order to secure more decontamination, the uraniwas retentlion became very high (see Sec. 4.3)e Actual sinbering of the NaF bed usually occurred at high uranium retentions, thus accounting for some of the excessive corrosion. Table 3.9. Decontasmination in a 9~in.~long NaF Absorbent Bed UFg-Fo gas stream from fluorination of NaF-ZrF)-UF) (52-44-l mole %) fuel at 600 to 650°C passed through l~in.-dias NaF bed (12-40 mesh) with temperature of 650°C in hottest portion; same bed used in Runs 3 and b Fo/U mole ratio: Run 1 - 3.7 Run 3 - 8.2 Run 2 - 3.6 Run b - 4.8 NaF/U weight ratio in ebsorber: 3/1 for Runs 1, 2, and 3; over-all ratio for Runs 3 and 4 = 1.5/1 Decontamination Factors Activity Run 1 Run 2 Run 3 Run I Over-all Gross beta 1.3 x 10% 5800 1.0 x 103 2900 Gross gomma, 3.2 x 10% 2.17x 10 2.4 x 10 14500 Ru ganma 1700 1600 1.0 x 10% 200 r-Nb gamma 3.2 x 107 T.h x10% 7.0 x10% 1.0 x 107 Total rare-earth beta 5.2 x 105 5.0 x 10% 4.9 x 10% 1.1 x 107 Across Absorbent? Gross beta Gross gamma, Ru gamms Zr-Nb gamma Total rare-earth beta 340 200 35 930 4100 220 9lo 1400 62 900 1.2 x 10% 2000 8 17 8 aCa,lculated on basis of activity found in absorbent and final product. ~25- Table 3.10. Decontamination in an 18-in.-long NaF Absorbent Bed UFg-Fo gas stream from fluorinstion of NaF-ZrFL-UFy (52-44-4 mole %) fuel at 800°C passed through l-in.-dis NaF bed (12-%0 mesh) with hottest polnt at 670°C; same bed used in Buns 1 and 2 end in Runs 3 and L Fo/U mole ratio: NaF/U welght ratio in absorber: Run 1 - 8.1 Run 3 - 8.4 Run 2 = 9.9 Run b - 10,2 for Runs 1 and 2 and Runs 3 and 4 was 3/1 6/1 for Runs 1 and 3; over-all ratioc Decontamination Factors Activity Run 1 Run 2 Run 3 Run k4 Over~all Gross beta 3900 1600 4300 2400 Gross gamma 9700 2700 2,0 x 104 1000 Ru gamma 1500 140 6700 450 | Zr-No gauma 3.5 x 10% 2.9 x10% 9.8 x 1ot 5.2 x 10 Total rare-earth beta 3.7 x 10“ 1.0 x 102 Across Absorbent® ~ Gross beta 2 1k Gross gamma 4o 8g Ru géumna 5 31 Zr-Nb gamms, 250 TOO %Calculated on basis of activity found in absorbent and finsl product. ~26- 4.0 BASIC CHEMICAIL STUDIES 4.1 NaF Capacity for UFg The absorption capacity of Nal for UFg was about 0.9 g of uranium per gram of salt for one lot of Harshaw Chemical Company meterial under a variety of conditions (Table 4.1). The capacity of Bsker and Adamson Company reagent-zrade NaF was 1.89 g of uranium per gram of salt. Absorption of UFE on Naf under an initial vacuum indicated a capacity range of 0.80 to 0.90 for different mesh sizes, while absorption values obtained when the excess UFg was removed with fluorine as a sweep gas varied from 0.72 to 0.86. Practically the same capacity values were obtained at 70, 100, and 150°C and at two different pressures. The capacity of 0.9 for the Harshew Chemical Company materisl is independent of the particle size to which it is degraded. The 1.89 wvalue for the capacity of the Baker and Adamson Company material corresponds closely to the molecular ratio in the complex UFS+3Nalf'. A complex of this nature has been reported by other investigators. As indicated, the original 1/8-in. pellets (Harshaw Chemical Company) had practically the same capacity as the same material degraded to 12-40 mesh. Examination showed that the lemon-yellow color, characteristic of the complex, was uniform inside the pellet. Although the material was therefore fairly porous, the limited capacity of approximastely 474 of theoretical and a surface area of only about 1 mc per gram indicated that the crystallite size was sbout 103 sngstroms. Severe conditioning of the NaF; such ag heating to 100-400°C with either vacuum or an atmosphere of fluerine to eliminate HF or moisture, did not affect the capacity measurements. Welght-loss tests after prolonged heating showed thet the HF or molisture content was no more than 0305%0 4.2 Pressure of UFg Above UPg-3NaF Complex Vapor-pressure data for the UFgG-NaF complex were obtained by the transpiration method over the temperature range 80 to 320°0¢ (see Fig. 4.1). The remction involved is described by the equation UFg*3NeF (s) — UFs(g) + 3NaF(s). (1) The data are fitted with the analytical expression log By = 10.88 - (5.09 x 103/1), Tavle 4.1. Absorption Capscity of NeF for UFg -27- Material Conditions Fg Pressure (psia) & Removal of Excess @FG Capaalty (g of U per g of NelF) Harshaew Chemical o, 12 to 40 mesh 15% 160 Not removed 0.86 15% 100 By vacuum 0,56 To58 100 By vacuam 0,80 158 TC By vacuum 0,88 is58 150 By wvecuun 0.90 155 100 By fluorine 0.86 15P 100 By fluorine 0,86 8 to 12 mesh 150 100 By fluorine 0.81 1/8«1n, pellets 150 100 By fluorime 0.72 159 100 By vacumm 0,61 Baker and Adamson Co. Powder 152 100 By vacwum 1.89 aUF6 introduced o N&F under vacuum. hUFs introduced st atmospheric pressure 1o displace fluorine. UFg VAPOR PRESSURE {(mm Hg) 1 O I ._.'."__ B A R 10 ~28- UNCLASSIFIED ORNL-LR-DWG. 12613 0 100 mi/min & 50 ml/min & 20 mi/min T T Ny FLOW RATE: T T T Tl 47 - T T T T r R i* i e 1" - " log A, =10.88 % - 5.09 X 103/ 7— e 1.6 1.9 2.2 A Tk (X4103) Fig. 4.1. Dependence of UFé-NdF Complex Vapor Pressure on Temperature, -29.. wvhere T is the absolute temperature, Use of the Clausius~Clapeyron formula with this equation gave a value of +23.2 kcal per mole of UF, for the enthalpy change of Eq. 1. The data were obtained by passing nitrogen at a flow rate of 100 ml/min, or less, through a prepared bed of the UFg-NaF complex at any desired temperature, trapping out the UFg in the nitrogen stream in a dilute A1(NO3)3 solution, and measuring the total volume of nitrogen with a wet-test meter. The UFg hydrolysis samples were analyzed by colorimetric or fluorimetric methods +o an accuracy of better than +5%. Temperature control of the bed was maintained always to within +0,29C. The UFg-NaF' complex was prepared by saturating a 30-g bed of 12-20 mesh Harshaw NaF in a l-in.-dia vertical nickel reactor with UFg at 100°C. Over 90% of bed saturation was maintained throughout the tests. Check runs, made at various No flow rates, showed no flow-rate effect. Crude adiabatic experiments were made with 100-ml batches of NaF to show that the reaction heat of 23.2 keal per mole of UFg produces a large temperature rise, approximately 130°C, if total saturation with UFg (preheated to about 100°C) is carried out guickly in a period of a few minutes. 4.3 Decomposition of UFg-3NaF Complex A study of the decorwosition of the UFg+3NaF complex at temper- atures of 245°C and higher has confirmed the belief that uranium retention on the NaF bed will be excessive in the NaF desorption step 1f the temperature and sweep-gas flow rate are not properly controlled. The retention results from decomposition of the UF o= 3Nal’ complex to a complex of NaF with UFs5, which is not volatile.”? A maxirum decomposition rate of about 0,01, 0.09, and 0.5% per minute is incurred at 250, 300, and 350°C, respectively, in the absence of fluorine if all the uranium is assumed to be 1n the form of the solid complex UFg-3NalF. Under optimum condltions, UFg desorption from the NaF bed (see Secs. 3.1.1 and 4,2) competes favorably with the decomposition effect, resuliting in a small uranium loss. Fluorine appears essential to inhibit the decomposition reaction, and possibly to promote refluorination of the nonvolatile U{V) com- pound, formed in the decompesition. The possibility of any uranium retention by the decomposition mechanlsm in the absorption step at 100°C appears to be insignificant, even over extended periods. The temperature-dependence of the rate of decomposition of UFge3NalF was determined in a series of runs over the temperature -30- range 245-355°C (Fig. 4.2). The probable reaction involved is UFge3NaF (s0lid) ——» UFg«xNaF (solid) + O.5F; (gas). The dependence of decomposition rate on temperature is log r = 6.09 - 5.22 x 103/T where r is the fractional decomposition rate in reciprocal ninutes and T is the absolute temperature. The rate was calculated on the basis of an absorption capacity of 1.33 g of UFg per gram of NaF. The energy of activation was calculated as +23.9 kcal/mole of UFg-«3Nal complex. It is possibly significant that this energy change is approximately the same as the enthalpy change of +23.2 kcal per mole involved in the volatilization of UFE from the U¥Fg-3NaF complex. The decomposition data were obtained with 4~ to 5-g samples of Na¥ (Harshaew Chemical Company pellets classified to 12-20 mesh) held in a U tube (1/2-in.-dia stainless steel tubing), through which gaseous UFg was passed at atmospheric pressure. An oll bath was used Tor manually controlling the temperature 1o iBOC during the course of each experiment. The runs were ended by removing the cil vath and rapidly cooling the sample. The UFge-3NalF was formed at the beglnning by saturating the NaF at a high UFg flow rate, after which the flow rate was decreased to 0.1-1 g/min for the remaining time, The length of the runs at various temperatures was adjusted so as to obtain a U(V) content in the final product of 1 to 10%. The excess UFg still absorbed on the NaF at the end of each test was not desorbed because of the difficulty of achieving this without increasing the U(V) content. The amount of UFg«3NaF complex affected by the reaction was determined from U(V) and U(VI) analyses. A temperature sbove 355°C was not used since it would be difficult to maintain saturation of the NaF with UFg without use of a pressurized system. In preliminary work on the decomposition reaction, a U(V) content of 20-26% represented a limit which could not be exceeded in one cycle of saturation of the NaF with UFg at 100°C followed by heating as a closed system to 350-400°C. Generally, in these runs, the NaF weight increase verified the asssumption that the decomposition product is a complex of UFs with Na¥. X-ray crystal- lography data indicated that the UF5’ ag complex in the decomposi- tion product has an orthorhombic strycture with cell dimensions a0 = 4.90 &, bg = 5.47 &, ¢g = 3.87 A. The x-ray pattern of y- or B-UF5 was not observed in the material. -31- UNCLASSIFIED ORNL-LR-DWG. 156894 2 o) ~ l | - log r=6.09-5.22x103/T - e ' UFg" 3NaF ~——= UF5- x NaF +, Fp c £ }_ - o . l& » o = o - * &L 1073 % | ® O | . & L) i) | .| 2 N = ) N '_ &