: | " MASTER ORNL=-3298 UC-10 — Chemical Separations Processes for Plutonium and Uranium TID-4500 (17th ed.) BY FLUORINATION OF FUSED FLUORIDE SALTS e G. |. Cathers R. L. Jolley OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. 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Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or thot the use of aony information, opporatus, method, or process disclosed in this report may not infringe privately owned rights; o Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report. As used in the obove, ‘‘person acting on behalf of the Commission'’ includes any employee or contractor of the Commission to the extent that such employee or contractor prepares, handles or distributes, or provides access to, any information pursuant to his employment or contract with the Commission. ORNL-3298 Contract No. W-7405-eng-26 CHEMICAL TECHNOLOGY DIVISION Chemical Development Section A RECOVERY OF PuF6 BY FLUORINATION OF FUSED FLUORIDE SALTS G. |. Cathers and R. L. Jo||ey- DATE ISSUED BCT 8 - 195% OAK RIDGE NATIONAL LABORATORY - Oak Ridge, Tennessee Operated by UNION CARBIDE CORPORATION U. S. ATOMIC ENERGY COMMISSION ABSTRACT . Fused salt fluorination tests were conducted at 600°C to determine the feasibility of recovering plutonium as PuFg in the fused salt-fluoride volatility process. Recoveries and material balances were good, although the initial plutonium concentration in 50-50 mole % NaF~ZrF4 or 31-24-45 mole % LiF- NaF-ZrF4 salt was only 2 ppm. The volatilization reaction appeared to be approximately first-order with respect to the plutonium concentration in the salt. Results of absorption of the volatilized PuFg4 on beds of LiF, CaFy, or NaF indicate that this is a possible method of trapping the material in fluoride volatility processes, possibly separately from UFg. CONTENTS Page 1.0 INTRODUCTION 1 2.0 FUSED SALT-FLUORIDE VOLATILITY METHOD ] 2.1 Retention of Plutonium in Fluorination of Fused Salts 5 2.2 Behavior of PuF70% of UF6 I REACTOR FUEL ' . I | - PRODUCT | » 100°C —Y»{ 100-400°C L * Dissolvent Salt ‘ | UFg# PuF, —— _ . 6 _ (LiF-NoF—ZrF4 or -BeF2) ' q Absorbent Bed NoF Absorbenr S ‘ ‘ = K : ystem - Y 7 Y for PUF6 “for UFé (2 Beds) . ! ‘ ] [ . 1 I —_——— [—_—t—— e Salt Transfer -1 - _— - Y - 2 L T .~ PRODUCT - - - - - PuF = -' _ Recycle | 6 Dissolution Fluorination 500-600°C _ 500°C Salt Waste (>99% FP Fluorides) Fig. 1. Processing of U-Pu-bearing reactor fuel byAFQsed Salt-Fluoride Volatility method. The used NaF is transferred to the molten salt fluorinator. Further absorption steps to purify the plutonium from fission product- ~activities may be necessary as in the case of UFé. | : Table t. Summary of PuFg Volatilization Tesis with Fused Salt 2 ppm Pu in 50 g 31-24-45 mole % LiF-NaF-ZrF, at 600°C with exceptions noted in runs |, 2, 3, 5, 6, and 11 Flow R?fe, - Pu Retenticn in Fused Salt, % Run ml/min No. F, He Special Remarks ‘ 1 hr 2hr 3hr 4hr 5hr 6 hr 7hr 8hr 10hr 12hr 14 hr 16 hr 18 hr 20 hr 1 100 -- 50-50 mole % NaF-ZrF, 80.4 285 216 -- 14.0 -- -- - - -- -- -- -- - 2 100 -- . 50-50 mole % NaF-ZrFy -- -~ -- -- -- 15.3 -- .- -- -- -- -~ -- -- 3 100 -- 50-50 mole % NoF~ZrF4 -- -- -- -- -- 27.2 -- - -- -- -- -- -- - 4 100 -- : -- -- -- -- - 37.7 -- -—- == -- -- -- -- -- 5 100 -- 50-50 mole % NaF-ZrFy -- -- 39.1 -- - 16.0 -~ -- -- -- - - - -- 6 100 -- 50-50 mole % NaF-ZrFy4 756 70.5 564 417 282 -- -- -~ - -- -- -- -- -- 7 100 - 92.3 744 577 48.1 404 -- -- - - -- -- -- -~ - 8 100 -- -- 69.3 - 52,6 -- 34.6 -- 23.1 154 105 7.6 4.3 2,7 1.4 9 50 -- 85.9 74.4 686 73.0 60.2 53.8 -- -— -- -- -- -- -- -- 10 150 - 67.9 48.7 423 33.3 205 16.7 -- - -- -- -- -- -- -- 1 100 --- 500°C 75.6 65.4 545 449 39.7 34.0 -- - .- -- -- -~ -- -- 12 100 -- 7 mole % UF 4 in F5 stream 77.0 62.0 48.0 385 31.7 30.1 -- - - - - -- -= -- (5.3 hr) 13 25 75 75.6 70.2 795 87.2 654 65.4 -- - - -- -- -- - - 14 100 -- 2.98%UF4 in original salt 74.4 40.4 35.2 269 27.% 20.5 -- -— -- -- -- -- - -- 15 50 50 ) 53.2 47.5 40.4 41.0 46.2 34.0 -- - = -- -- -- -- -- 16 100 -- 40.8 3483 288 19.1 159 - -- - -- -- -- -- - - 17 100 -- 13.5-mil-i.d. capillary gos inlet 65.4 52.6 48.1 30.8 29.5 20.5 -- - -- -- -- -- - -- 18 150 -- 30-mil-i.d. copillary gas inlet 73.1 53.2 4.0 295 23.7 21.2 -- - -- -- -- -- -- -- Table 2. Summary of PuFg Absorption Results and Material Balances (See Table 1 for Summary of Conditions in Fused Salt Fluorination Tests} Pu Distribution, % of total Pu Material Balance, % Run Descripticn of Absorption Beds Fused Salt Tubing Bed Bed Bed Bed Of Of No. No. 1 No. 2 No. 3 Salt Samples Walls No.1- No.2 No.3 Wallsd Total Pu Volatile Pu 19 200 g NaF, 25°C - -- 12.4 -- 1.9 - -- .- == -- - 2 8 g NoF, 25°C 8 g NaF, 25°C - 27 -- 0.8 46.6 0.2 -- 1.3 76.0 67.1 39 8.g LiF, 25°C 8 g NaF, .25°C -- 23.0 -- 1.3 58.4 1.4 -- 7.4 97.5 96.5 4b 8 g Lif, 25°C "~ 8 g NaF, 25°C -- 37.7 “- 3.2 62.6 0.1. -- 5.1 109. 114.. 5¢ 8 g Cofy, 25°C 8 g NaF, 25°C -- 15.0 13.1 6.1 48.1 0.1 -- 4.5 87.9 83.0 6 8 g LiF, 25°C 8 g NaoF, 25°C -- 23.2 7.5 1.2 52.8 0.4 ©-- 6.7 96.8 95.2 7 8 g NaF, 25°C 8 g NaF, 25°C -- 40.4 © 6.5 - 55.1 1.7 -- -- 104. 107. 8 8 g CaFq, 25°C 8 g NaF, 25°C -- 1.4 5.6 2.5 929 ~0 - -- 102. 103. 9 8 g NoF, 25°C 8 g NaF, 25°C - 53.8 8.9 4.8 28.8 5.1 -- -- 101. 104. 10 8 g CaF,, 100°C 8 g CaFy, 25°C - 15.7 - 62 20 73.0 0.1 -- -- 98.0 97.5 LR 8 g CaF 4, 400°C 8 g CaFy, 50°C -- 34.0 8.7 4.2 59.6 0.5 -- -- 107. N2 12 8 g CaF,, 100°C 8 g NaF, 25°C 25 g NoF, 25°C 3C.1 7.3 3.7 62.2 0.1 ~Q -~ 103. 105. 13 8 g NoF, 400°C 8 g NaF, 70°C -- . 654 9.8 8.3 14.7 0.7 -- -- 98.9 95.5 14 8 g LiF, 40C°C 8 g NaF, 100°C 8 g LiF, 100°C 2C.5 | 5.6 12.2 3.7 51.9 1.3 -- 95.2 93.7 15 8 g LiF, 40C°C 8 g LiF, 100°C -- 34.0 5.7 37.3 0.7 153 -- -- 93.0 88.5 16 8 g LiF, 10C°C 8 g NaF, 100°C, -= 15.9 6.1 15.1 43.8 0.2 .- -- 81.1 75.9 17 5 g NoF, 1C0°C 5 g LiF, 100°C -- 20.5 10.1 18.5 1.9 0.2 -- -- 61.2 44,1 18 "5 g NaF, 100°C 5 g NaF, 25°C - 212’ 4.2 27.3 19.2 0.1 - -- 72.0 62.5 “Analytical method No. 1 (Sect. 4.2). bAno|yfico| method Ne. 2 (Sect. 4.2). CAnclytical method No. 3 used in runs 5-13 (Sect. 4.2). dMainly absorbent. dust. -5~ of plutonium volatilized as PuFg. The material balances for many of the tests were in the range 90-100% despite the use of only 0.1 mc of Pu-239 per run. In two tests with uranium present the possibility of separating recovered plutonium and uranium was indicated. 2.1 Retention of Plutonium in Fluorination of Fused Salts In all tests the plutonium retained in the fused salt decreased during fluorina- tion, indicating volatilization of PuFg (Table 1). The rate of disappearance of plutonium from the fused salt had an approximately first-order dependence on the concentration in the salt. Deviations from the first-order dependence could be due to inhomogeneities in the gas or salt mixing. In three lesls {(runs 7, 9, and 1U) with fluorine flow rates of 50, 100, and 150 ml/min, the volatilization rate constants (assuming a first-order dependence) were approximately proportional, being, respectively, 0.11, 0.18, and 0.31 hr~ (Fig. 20). The corresponding half-value times in these tests were 6.5, 3.8, and 2.2 hr. It was concluded from this that the amount of plutonium transfer or vola- tilization depends approximately on the total amount of gas passed through the : salt. ' A special long-duration test of 20 hr (run 8) demonstrated that the initial plutonium concentration of 2 ppm could be reduced to 1.4% or 0.028 ppm, with no indication that this was a lower limit (Fig. 2b). The half-value time in the initial part of this run, 4.0 hr, duplicated the result in run 7 under about the same condi- tions. The curvature of the plotted data indicates that the fractional volatilization rate increased v some extent as the experiment proceeded. Tests at 600°C with 50-50 mole % NaF-ZrF4 salt (Fig. 3) instead of with 31-24-45 mole % LiF-NaF~ZrF 4 salt gave little indication that salt composition was @ major variable. The half-value time (run 6) was about 3.4 hr. There was no significant change in the volatilization rate at 500°C. Fluorination with fluorine gos diluted with helium gave anomaloys data. With a 50/50 F»/He gas mixture, there was initially a rapid decrease of plutonium in the salt, but then the rate of disappearance decreased so that the final salt concentra- tion of 34.0% (of initial level) in run 15 compares closely to the 34.6% in run 8. A 25/75 Fo/He gas mixture definitely gave a slower plutonium disappearance rate. Both runs were characterized by abnormally erratic data (Fig. 4). Evidence was obtained that the rate of plutonium transfer from the fused salt is enhanced by increasing the degree of dispersion of the F5 in the salt. In the first 16 runs the gas bubble size was the result of using a 1/4-in. tube immersed in the salt. In runs 17 and 18, capillary fluorine inlets were used, giving half-value times of about 2.5 hr, with little difference noted between fluorine flow rates of 100 and 150 ml/min (Fig. 5). 100 Fz‘ur 50 ml/min 50— Pu REMAINING IN SALT, % 150 mb/min 100 UNCL ASSIFIED ORNL-LR-DWG, 67522 4 . é 0 TIME, hr Fig. 2. Volatilization of PuFg from fused 31-24-45 mole % LiF-NaF-ZrF, at 600°C. (a) 6~hr runs at different fluorine Flowrc’res, (b) 20-hr run at fluorme flowrate- of 100 ml/min. Pu REMAINING IN SALT, % UNCLASSIFIED ORNL-LR-DWG. 67519 UNCLASSYFIED ORNL-LR-DWG. 67520 100 100 50-50 mole % NcF-ZrF4, 600°C [ ] B 31-24-45 mole % LiF'-NoF-'ZrF4, 500°C ® T 50 mi/min Fy + 50 ml/min He A ® 4 @ A R 5 — -— < S [F4] Z = A . A O ° 4 Z F'y B A - A = w [-*4 < . A A 25 ml/min F, + 75 ml/min He \ 20 | | 20 { | I l g 4 6 0 4 6 TIME, hr TIME, hr Fig. 3. Comparative PuFg volatilization rate from 50-50 mole % NaF-ZrF 4 at 600°C and 31-24-45 mole % LiF-NaF-ZrF 4 at 500°C. Fluorine flowrate 100 ml/min. Fig. 4. Effect of dilution of fluorine with helfum in volatilization of PUF6 from 31-24-45 mole % LiF-NaF- ZrF4 salt at 600°C, UNCLASSIFIED ORNL-LR-DWG, 87521 100 -i.d. capillary Pu REMAINING [N 5ALT, % 20 TIME, hr Fig. 5. Volatilization of PuFg from 31-24-45 mole % LiF-NaF-ZrF4 at. 600°C with capillary fluorine inlet tubes. ' 2.2 Behavior of PuF, Volatilized in Fused Salt Tests In all except the first run the data indicated that the bulk of the volatilized PuF ¢ was trapped in dry fluoride beds consisting of NaF, LiF, or CaFy (Table .2). These materials appeared equally effective in the 25-100°C temperature. range. CaF 5 was effective also at 400°C (run 11). There was some indication of a plu- tonium breakthrough with NaF at 400°C (run 13), and there was detinite nonsorp- tion on LiF at 400°C (runs 14 and 15). 2.3 Overall-Plutonium Material Balances A good material balance was obtained in most of the runs, not only for the total plutonium used in the test, but also for the part that was volatilized from the salt (Table 2). The latter was calculated on the basis of the final fused salt concen- tration, corrected for the amount of plutonium removed in fused salt samples. The material balances obtained appear to be reasonable in view of the small amount of initial plutonium used (100 pg) and of the large number of samples that had to be analyzed. The data show that some plutonium was retained on all wall surfaces within the system. This was expected since such a small amount of plutonium was used. The experience of other workers with PuF4 indicates that such loss is insignificant, on a percentage basis, when handling 50~100 g quantities. ' ~9- 2.4 Absorption as a Separation Method for PuF .~UF ., Mixtures In one test the fluorine used in the volatilization step contained 7 mole % UF4 (run 12). The absorption results (Table 3) show that it is possible to effectively separate PuF, and UFg4. In this test with a 7% UF4-F5 mix the final fused salt con~ tained less than 100 ppm of uranium. In a second test the initial fused salt contained 2.26% uranium (as UF4) in addition to the usual plutonium spike (run 14). With a LiF bed at 400°C, however, the PUF6 broke through to the following NaF bed. Similar plutonium behavior was evident in run 15 where no uranium was present. Tuble 3. Relative Absorption Effects? tor PuF s and UF P 6 6 Amount Absorbed Plutonium, pg Uranium, g Run Bedb BedP BedP Bed Bed Bed No. Walls 1 2 3 Walls ] 2 3 12 3.7 62 0.1 <0.1 0 0.05 6.02 17.9 14 12 3.7 52 1.3 <1073 <1073 1. <1073 OSeporcfion factors in run 12 for plutonium on bed 1 = 450 and for uranium on - beds 2 and 3 = 330. See Table 2 for description of the absorption beds used in these experiments. . 3.0 DISCUSSION 3.1 Fused Salt Volatilization The conditions under which the above results were obtained do not duplicate the conditions that might be expected in actual processing of reactor fuels. For example, irradiated low-enrichment UO9 might be expected to have a plutonium content of 5000 g/tonne after use as power reactor fuel. ‘When this fuel is dissolved in fused salt, a reasonable uranium concentration would be about 5% with a plutonium concentration of 250 ppm, which is far above the level of 2 ppm used in this work. However, the adequate volatilization and recovery obtained at the 2 ppm level indicate that little trouble would be encountered at the higher level. The PuF¢ volatilization process appears to be primarily a sweep-out or sparging action, as is also the case with UF¢ volatilizations at low concentrations (<1%). -10- In a typical run at a fluoride flowrate of 100 ml/min, the plutonium transfer from salt to gas in the first minute of operation (assuming a first-order rate effect) was ~0.3 pug. At a fused salt concentration of 250 ppm in actual fuel processing, the initial transfer value would be increased to 37.5 pg. This is still well below the value .of about 500 g in the first 100 ml of fluorine gas obtained by using data for the equilibrium PuF, + Fp— PuFg at 150°C (2,3). The data presented indicate that PuF4 volatilization from fused salt is slow, but this does not mean that it is impractical as a processing technique. The super- ficial linear velocity of the fluorine gas in the reactor (1 in. dia) varied from 4 to 12 in./min, to give half-times of 2.5-4 hr. Probably shorter half-times would be achieved by increasing the gas flowrates to a superficial linear velocity of as much as 500 in./min. Flows of this magnitude have been used in the HF sparging of salt in the Oak Ridge Volatility Pilot Plant (5). 3.2 Absorption of PuFg The existence of chemical complexes of PuFg with LiF, NaF, or CaF5 is indi- cated by the results. Although less than 100-ug quantities were used, they were trapped by these materials. It is dubious that this was due only to surface adsorption, to a hydrolytic mechanism, or to simple filtration of entrained material. The chemical complex concept, however, is consistent with the behavior of UF, with NaF, forming the complex UFg-3NaF (6). An adsorption mechomsm is unlikely due to the fact that the specific surface areas of these materials are 1 m /g or less. The hydrolytic mecha- nism-is discounted because of the large excess of fluorine. The similar behavior of PuFg to that of UFg in forming chemical complexes or compounds is supported by similar reactions of NaF with other hexafluorides, e.g. MoF¢, TcFg, and NpFg (7). The dissociation pressures of the UFg-NaF and MoF 4= NaF complexes over a wide temperature range have been studied, and similar work is needed on the other compounds. The UF,-NaF, MoF,~NaF, and probably the NpF¢-NaF complexes appear completely reversible. The behavior of PuF¢ with NaF and CaF, at 400°C indicates that these complexes might be irreversible under prac- . tical conditions. The breakthrough of PuFg in a LiF bed at 400°C, in contrast to the behavior at lower temperatures, shows that this complex might be more easily re- versible than the others. UF, does not complex with LiF or CaF,, whereas PuF g apparently does. ‘ - In the one test (run 12) with both PuF¢ and UF,4 entering the absorption bed train, the absorption of PuF, in the presence of a large excess of UF¢ further supports the view that a PuF4~CaF o complex was formed. If surface adsorption had occurred, it is reasonable to assume that the UF ¢ gas would have "washed off" the PuFg4 since the condensation and vapor properties of the two materials are similar: sublimation temperature of UFg 56.5°C, boiling point of PuF, 62.3°C. -11- 4.0 EXPERIMENTAL TEST EQUIPMENT AND PROCEDURE 4.1 Fluorination Work The equipment for the plutonium work was mainly nickel vessels connected by 1/4-in. copper tubing with compression-type tube fittings (Fig. 6). The nickel fluorinator was constructed from 1-in.-dia tubing and was 6 in. long.: A 1/2-in. entry port was provided for introduction of salt and plutonium spike solution. The outlet was 1/4-in. nickel tubing about 6 in. long. In runs 1-16 the fluorine inlet was a 1/4-in. dip tube, welded into the side of the vessel, which extended down to about 1/4 in. from the reactor bottom. In runs 17 and 18 special capillary inserts were attached to the end of the dip tube before insertion and welding. The 8-g absorption traps were made from 1/2-in.-dia nickel tubing and were about 5 in, long. " Nickel-wool plugs were used to retain the absorbent (12-20 mesh) in the trap. The 5-g absorption traps were slightly shorter and made with a level cut in UNCL ASSIFIED ORNL-LR-DWG, 67518 Poper Filter for a Activity Thermocouple Attochment Port for Sompling and Addition of Salt and PU(NO3)3 Spike Solution Absorbent Sodo Lime F2 Trep F2, He, ond UF, Nichrome Wire Furnace ' / Fig. 6. Schematic of experimental equipment. -12- a V-form to eliminate the necessity of using nickel wool retainer plugs. The procedure consisted in inserting the LiF~NaF-ZrF4 or NaF-ZrF 4 salt, " broken up into pea-size lumps, into the fluorinator, after which 100 ul of PuO2(NO3)o solution. (~1 g Pu/liter) was placed directly on the salt, with care to avoid contact with the metal walls of the reactor. The reactor was then inserted in the furnace and connected to the gas tubing system. Heating was carried out slowly and carefully with a helium sparge to decompose the aqueous plutonium spike solution. After the reactor had reached the operating temperature of 500-600°C, the helium flow was stopped and fluorine flow was started through the by-pass circuit to condition the tubing, vessel walls, and absorption material, This was continued for about 1 hr. All the opporctus was at operating temperature during the conditioning period. Salt sc:mples were taken at intervals during fluorination, Each sampling was preceded by a short helium sparge, and this time was not counted in the total fluori- nation time. The salt samples (~0.5 g each) were taken with a 1/8-in.~dia nickel rod by the quick-freeze technique, i.e. by quickly inserting the cold rod and with- drawing it before the frozen salt could remelt. Experience with uranium and radio- activity determinations has shown that this is a reliable method of sampling since the frozen salt does not have a porous structure and the time lnvolved (2-4 sec) is short. A large safety trap containing ~1 kg of soda lime was placed at the end of the gas system to absorb the fluorine and to ensure that no plutonium would leave the system and contaminate the external working area. This worked efficiently, only one replacement being made over the entire series of tests. No plutonium a activity was ever detected on a paper gas filter placed at the exit of this trap. The fluorine gas used in these tests was supplied by the Oak Ridge Gaseous Diffusion Plant, It was passed through NaF to remove 3-5% HF before use. The purity after this treatment was in the range 93-97% Gas flowrates were controlled and measured with 50-mil-dia capillary flow- meters, using 0.25 psi input differential pressure instruments to measure the Ap. 4.2 Analytical Methods Suitable fluoride salt dissolution procedures had to be developed during the test runs because of difficulties in initial analytical tests in achieving reproducible results., Aluminum nitrate solution (1 M) was used initially to dissolve fluorination salt samples as well as absorbent bed fluorides. Erroneous and erratic results were obtained in using LaF3 precipitation followed by TTA extraction to measure the plutonium a activity. Consultations with F. L. Moore and J. H. Coopcr at ORNL indicated that the low plutonium analytical recoveries were due to APt and F~ interference in LaF3 precipitations and to ZrF4 interference in counting due to its extraction by TTA. ' -13- A satisfactory analytical method found was to use dilute aqua regia as the solvent (method 3, below). However, even in this case the presence of dissolved ZrF4, NaF, and LiF in the dilute aqua regia affected the plutonium a determinations. The percentage recovery appeared reproducible, however, and method 3 was therefore used with all types of material fo obtain comparative but not absolute values. Method 1: Filter Paper Technique. In the first fluorination tests the fluoride salts were dissolved in T M AI(NO3)3 solution (I g of salt in ~20 ml of solution). Aliquots of these solutions were used successfully with a filter paper technique since results could be quickly obtained and the fluorinations could be monitored as they proceeded. Comparison of the filter paper results with LaF3-TTA results indicated greater.reliability of the former over the latter, and hence the filter paper results were used exclusively in the first three runs. The filter paper technique consisted simply in slowly dropping 0.050 m! of the 1 M AI(NO3)3 solution onto a piece of 5-cm filter paper and allowing it to air dry at 80-90°. The plutonium a activity was then counted with a scintillation counter at 41% geometry. Method 2: HNOs3 Dissolution. Dissolution of some of the absorbent bed materials (run 4) in 4 M HNO3 gave accurate data by a standard LaF3-TTA analyti- cal method. However, since the method was not suited for use with fluorination salt samples, no further work was done with it, Method 3: Dilute Aqua Regia Dissolution. Dilute aqua regia (4 M HNO3- 4 M HCI) at 95°C, in polythene containers, was suggested by C. J. Shipman as a general dissolvent for all the fluorides. In 10 analytical runs in which an aliquot spike of the standard Pu(NO3)3 solution in a synthetic salt solution (1 g of 31-24-45 mole % LiF~NaF-ZrF4 in 25 ml of dilute aqua regia) was used, plutonium recoveries were 74.7, 77.9, 83.2, 81.5, 85.5, 74.3, 92.2, 95.6, 83.4, and 87.6%. The average was 83.6%, with a standard deviation of 6.65. The counting level in these tests was in the range 700-900 cpm/ml. The average recovery again demonstrated that some interference (probably from ZrF4) was recurring in the analytical LaF3 precipitation- TTA extraction procedure; however, the recovery was much higher than when aluminum nitrate solution was used, and the variation in error was low. In fused salt runs 6-10, inclusive, a statistical test of the zero-time values for plutonium in the salt after spiking, melting, and helium sparging with duplicate sampling showed the same average recovery value of 83.6% with a standard devia-~ tion of 6.35. The values obtained were 77.5, 90.4, 79.1, 95.2, 89.3, 89.3, 78.6, 81.3, 77.0, and 78.6%. ~14- 5.0 REFERENCES . B. Weinstock, E. E. Weaver, and J. G. Malm, "Vapor Pressurés of NpF¢-and PuF: Thermodynamic Calculations with UF4, NpF 4 and PuF¢," J. Inorg. Nucl Chem., 11: 104-11 (1959) . L. E. Trevorrow,"W. A. Shinn, and R. K. Stenenberg, "Thermal Decomposition of - Plutonium Hexafluoride," J. Phys. Chem., 65: 398-403 (1961). . J. Fischer, L. Trevorrow, and W. Shinn, "Kinetics and Mechanism of the Thermal Decomposition of Plutonium Hexofluorlde," J. Phys. Chem., -65: 1843-6 (1961) G. I Cathers, "Uranium Recovery for Spent Fuel by Dissolution in Fused Salt and Fluorination, " Nucl. Sci. Eng., 2 768-77 (1957) . W. H. Carr, S. Mann, and E. C, Moncnef Uramum—erconium Alloy Fuel Proc~ essing in the ORNL Volatility Pilot Plant," Preprint 150, Symposium on Volatility Reprocessing of Nuclear Reactor Fuels, 54th Annual Meeting, A.l1.Ch.E., December 2- 7 1961, H. J. Emeleus and.A. G. Sharp, eds., "Advances:in Inorganic Chemistry and Radio- chemistry," Vol 2, pp.: 214-215, Academic.Press, Inc., New York, 1960. G. |. Cathers, "Dissociation Pressure of MoF ¢~NaF Complex and the Interaction of Other Hexafluorides with-NaF, " Paper presented at 140th Meeting of American. Chemical Society, -September 1961. -15- ORNL-3298 UC-10 — Chemical Separations Processes for Plutonium ond Uranium TID-4500 (17th ed.) INTERNAL DISTRIBUTION I.—‘E-U€Z>O§-IZ§-!—€%€1‘F".’°T'>T§"I’° . B. Lindauer . G. MacPherson Mann E. McNeese . B. Meservey W. Miles P. Milford C. Moncrief P. Murray (K-25) . R. Musick P. Nichols . W, Pitt B. Ruch H. Shaffer . J. Skinner . F. Soard A. Swartout . M. Watson . M. Weinberg . E. Whatley . R. Whitson . L. Katz (consultant) J. Katz (consultant) H. Pigford (consultant) . Worthington (consultant) Atomic Energy Commission, Washington 1. Biology Library 46, 2-3. Central Research Library 47, 4, Laboratory Shift Supervisor 48, 5. Reactor Division Library 49, 6. ORNL — Y=12 Technical Library 50. Document Reference Section 51, 7-26. Laboratory Records Department 52, 27, Laboratory Records, ORNLR.C. 53. 28. M. R. Bennett 54, 29. R. E. Blanco 55. 30, J. C. Bresee 56. 31. R. E. Brooksbank 57. 32. K. B. Brown 58. 33. W, H. Carr 59. 34. W. L. Carter 60. 35. G. I. Cathers 61, 36. F. L. Culler 62. 37. D. E. Ferguson 63. 38. H. E. Goeller 64. 39. A. T. Gresky 65. 40. C. E. Guthrie 66. 41, R. W. Horton 67. 42, R. L. Jolley 68, 43. R. G. Jordcn (Y-12) 69. 44, P. R. Kasten 70. 45. C. E. Larson EXTERNAL DISTRIBUTION 71.'E. L. Anderson, Jr., 72, O. E. Dwyer, Brookhaven National Laboratory 73. L. P. Hatch, Brookhaven National Laboratory 74, S. Lawroski, Argonne National Laboratory 75. O. Roth, Atomic Energy Commission, Washington 76. R. C. Vogel, Argonne National Laboratory 77. R. H. Wiswall, Brookhaven National Laboratory 78. Division of Research and Development, AEC, ORO 79-586. 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