TRILAU IRYLAMIME EXTRACTION OF NEPTUFWJM AND PLUTONIUM FROM PUREX PROCESS WASTE W. W. SCHULZ Pacijic Northwest Laboratory, Battelle Memorial Institute, Richland, Wash.
A batch solvent extraction process, used to recover Np237from a radioactive waste solution generated during processing (of irradiated uranium in the Hanford Purex Plant, also recovers the small amount of plutonium lost to the waste stream in Purex process operation. Trilaurylamine diluted with kerosine is the extractant. Hydrazine added to the acidic waste establishes both neptunium and plutonium in the + 4 valence state for maximuin extraction. More than 95% of both elements i s extracted in a single batch contact of equal volumes of :;olvent and waste. Both elements are stripped from the solvent effectively b y contact with dilute hydroxylamine sulfate or sodium sulfate-ferrous sulfamate solution. A satisfactory feed solution for final purification of the recovered neptunium and plutonium b y the tributyl phosphate solvent extraction procedures used in the Hanford Purex Plant can b e prepared from either strip solution. Trilaurylamine solutions have limited chemical and radiolytic stability when in contact with the waste solution.
irradiation of natural uranium in Hanford’s
N nuclear reactors to produce Pu239also produces gram
Process Concept
quantities of Np237. Neptunium-237 is the starting material an isotopic power source recently used for production of PU?:~, in space vehicles. Technology and equipment now used in the Hanford Purex Plant to recover and purify NpZ3’have been described recently (9, 77).
The TLA batch extraction process for NpZ37recovery is compared in Figure 1 to the recovery scheme now used in the Hanford Purex Plant.
EUTRON
I n this latter scheme, after dissolution of fuel elements in nitric acid, neptunium as Np(1V) is extracted into 30% tributyl phosphate (TBP)-kerosine solvent along with uranium and plutonium in the H A column. [In normal operation about 15% of the neptunium and 0.1% of the plutonium are lost to the aqueous raffinate (HAW) from the H A column and are not subsequently recovered.] The pregnant solvent is scrubbed for fission product removal in the HS column. Plutonium is then partitioned into an aqueous phase in the 1BX column; ferrous sulfamate is used to reduce the plu-
This paper describes a new batch solvent extraction process developed as an alternative way of recovering NpZ37in the Hanford Purex Plant. In this scheme all the Np237 is routed to an aqueous waste !solution and subsequently extracted into trilaurylamine (TLA) diluted with kerosine. Small amounts of plutonium present in the waste are also extracted and recovered. Present R o u t e
8 5 % NP -
t
TLA Route
Feed Preparation
Column
Column
4 100% Np
4 Waste Evaporator
c 15% Np Lost t o Waste
t 1WW Solution
4 B a t c h TLA
t Evaporator
pT+ 2E C o l
Column
Figure 1.
TLA
Aqueous to Underground Storage
Orqanic Batch
Aqueous Strip
Hanford Purex Plant neptunium recovery schemes VOL. 6
NO. 1
JANUARY 1 9 6 7
115
tonium to the inextractable Pu(II1) state. Neptunium and uranium are costripped from the solvent in the I C column by using dilute nitric acid. The I C product is concentrated and uranium is re-extracted in the 2D column. By control of uranium concentration in the organic phase in this step, neptunium is forced into the 2D raffinate stream which is concentrated for backcycle to the HA column. Approximately one third of the backcycle stream is continuously processed through the new 3A column, where the neptunium is extracted into the organic phase. (Plutonium associated with the neptunium is forced into the 3AW stream and thence to the backcycle stream for eventual recovery in the HA column.) The neptunium is stripped in the 3B column and returned to the 3A column. Neptunium is thus accumulated in the 3A and 3B systems and is periodically decontaminated and removed to the new batch ion exchange unit (the 3X column) for final purification and loadout. All the neptunium would be forced into the HAW stream in the TLA extraction scheme, if the HA column was operated with a high saturation of the solvent with uranium and sodium nitrite added to establish neptunium in the inextractable quinquevalent state. This technique is used in Purex process operation at the U. S. Atomic Energy Commission Savannah River Plant for neptunium recovery (76). After appropriate blending with various other Purex process aqueous wastes, the HAW solution is evaporated to yield a small volume of highly acid and highly radioactive waste (1WW solution). T o recover neptunium and plutonium from 1WW solution, hydrazine is added to establish both elements in the quadrivalent state. Neptunium and plutonium are then coextracted into the TLA-kerosine solution and subsequently costripped into an aqueous phase with any of several stripping reagents. This aqueous phase, after acidification with nitric acid, may be routed to the 3A-3B column system, just as in the present recovery scheme, for final purification and concentration of the neptunium. The benefits which might accrue in the Hanford Purex Plant from the ability to recover neptunium and plutonium from the 1WW stream by a TLA extraction process include: More flexibility in control of the main-line Purex process if neptunium were recovered separately. Improved fission product decontamination performance in the first cycle of the Purex process without net loss of plutonium or neptunium by operating a t higher levels of uranium saturation of the TBP phase. More flexibility in waste rework. Extensive rework operations have a t times been necessary to recover product losses associated with operational difficulties and equipment failures. Improved over-all plant recovery of neptunium and plutonium with concomitant economic gain.
(IV) > (VI) >> (111) and for neptunium as (IV) > (VI) >> (V). Amine extraction, according to Coleman (6), can be expressed as either anion exchange or adduct formation. These alternative, but thermodynamically equivalent, descriptions of the extraction of nitric acid and of plutonium(1V) nitrate from nitric acid solutions by a tertiary amine are illustrated in Equations 1 and 2 and Equations 3 and 4, respectively. Underlined terms indicate organic phase components.
+ H'N03- + HzO e RaNHOH + H'N03- e R3NH+N03- + HzO (1) R3N + H'N03R3NH'NOa(2) 2RaNH+NOs- + Pu(NO~)B-'e ( R ~ N H ) Z P U ( N O+ ~)B
R3N -
2N032RsNH+N03-
+ P U ( N O J ~e ( R ~ N H ) ~ P u ( N O J B(4)
Nitric acid is extracted into hydrocarbon solutions of trialkyl amines to an extent which exceeds the quantity required by the salt formation reaction (Equation 1 or 2) (2). Keder and Wilson (72) find extraction of excess acid into tertiary aminediluent solutions to be directly proportional to the hydrogenbonding propensity of the particular acid but also dependent upon the diluent. Process Flowsheet
A chemical flowsheet for the TLA extraction process is presented in Figure 2. Principal features are: Plutonium is present in 1WW solution largely as Pu(1V); neptunium is split between Np(1V) and Np(V). Both elements are established in the quadrivalent state by reducing Np(V) to Np(1V) with small concentrations of hydrazine a t 25' C. Contact of the 1WW solution a t 25' C. with an equal volume of 0.3M TLA-kerosine extracts 97 to 99% of both the nep-
Amine Extraction Chemistry
The use of high molecular weight amines as extractants dates from the work of Smith and Page in 1948 (27). Since that time myriad applications of this class of extractants have been made in the development of analytical methods (75) and in hydrometallurgical separations (3, 7). Currently, descriptive and physical chemistry of amine extraction and its use as a physicochemical probe are being explored in laboratories throughout the world. Coleman (6) has recently reviewed the current status of amine extraction with emphasis on reaction mechanisms and separations and recovery applications. Of particular significance is the high extraction power of long-chain tertiary amines for plutonium(1V) from nitric acid solutions (79). Advantage of this property is taken in this work as well as in the process recently developed by French workers for final purification of plutonium nitrate ( 5 ) . Neptunium(1V) also extracts strongly into trialkyl amines from nitric acid solutions, but extraction coefficients are nearly an order of magnitude lower than those for plutonium(1V) (6). Only the +4 states of both actinides are highly extractable from nitric acid solutions by tertiary amines. Weaver and Horner (23) list the order of extraction of plutonium in different oxidation states as 116
I&EC PROCESS DESIGN A N D DEVELOPMENT
(3)
1. A d d N 2 H 4 t o 1 W W t o 0 . 0 2 M 2. S t a n d 3 0 M i n u t e s a t 25 OC 3. C o n t a c t w i t h T L A a t 2 5 OC f o r 30 M i n u t e s
Aaueous
-
To W a s t e
I
Recycle TLA
TLA P h a s e 0.05M (NHzOH12. H 2 S O 4 F l o w : 100
II Strip Contact 1. C o n t a c t 3 0 M i n u t e s a t 5 0 'C 2. A l l o w 3 0 M i n u t e s f o r P h a s e s t o D i s e n g a g e
Aqueous
-
(99% PU. 97% Npl 3 A - 3 6 TBP C y c l e
Figure 2.
-
TO
TLA process flowsheet
tunium and plutonium. Slight extraction of fission product cerium, zirconium, and ruthenium also occurs. Neptunium and plutonium are costripped a t 50' C. into an equal volume of 0.05M (NH20H)2 H2S04. T h e resulting aqueous solution is a satisfactory feedstock for final TBP recovery and purification in the 3A-3B column system. Reagents used in the process do not increase the heavy metal ion concentration of the 1WW solution. This is a desirable feature. Present waste management plans at Hanford envision solvent extraction of Sr" and fission product rare earths from denitrated 1WW solution. Proposed flowsheets for this purpose involve sequestering of heavy metal ions [Fe(III), Ni(II), etc.] with a complexing agent such as citric or tartaric acid (78). Any increase in metal ion concentration would increase the amount of complexant required and hence the cost of recovering Sr" and other isotopes. Specification of T L 4 for this process rather than some other tertiary amine merits comment. Several commercially available tertiary amines were tested, Only a single organic phase was obtained when solutions of TLA in paraffin diluents were brought in contact with 1WW solution. Two organic phases resulted n i t h such solutions of all other amines tested. Formation of a second organic phase is a common phenomenon in amine extraction systems and can often be avoided by using aromatic diluents such as diethylbenzene or Solvesso 150 (Humble Oil Co.). Aromatic diluents, however, are not regarded as suitable for use in the Hanford Purex Plant. Experimental
Materials. Some experiments were made with Hanford Purex Plant 1WW solution. I n the majority of the experimental studies, however, laboratory-prepared IWW solution of the composition shown in Table I was used. This composition is typical of current plant waste, although some variation occurs from time to time because of minor changes in plant operating conditions. The concentration of certain of the inert and radioactive fission products in plant waste was approximated by addling inert strontium, barium, ruthenium, zirconium. and various rare earths to the laboratory-prepared waste solution. Insoluble phosphates and sulfates precipitate when waste of the composition shown in Table I is allowed to stand. Two stock solutions, each stable to precipitation of solids, were made. As required, 1WW solution, initially free of solids, was prepared by mixing appropriate volumes of the stock solutions. Appropriate amounts of stock solutions of either plutonium(1V) nitrate or neptunium(V) nitrate obtained from the Hanford Purex Plant and purified by anion exchange technique, were added at this time also. Plutonium was present as a mixture of the 238,239,240, and 241 isotopes; only Np237 was present in the neptunium stock. Unless otherwise noted, the extractant was 0.3M TLASoltrol-170 prepared by diluting as-received TLA (Archer Daniels Midland Co. Adogen 363) with Soltrol 170 (Phillips Petroleum Co.). Other trialkyl tertiary amines
-
Table I.
Constituent"
Typical Composition of 1WW Solution
M' 6.9
Constituent
Ni +2 104-3
a
Fe +3
0.35
~ 1 + 3
0.13
Cr +3
0.025
NP Pu
M 0.015 0.015 10-5 to 10-6 10-5 to 10-6
Also radioactive and inert fission products at concentrations in the
70-3 to 10-6 M range are present.
tested include triisooctylamine (TIOA) (Union Carbide Chemical Co.), tri-n-octylamine (Distillation Products Industries, Eastman No. 7723), Alamine 336 (General Mills, Inc.), and Adogen 364 and Adogen 368 (Archer Daniels Midland Co.). I n studies of the effects of irradiation on extractant stability, TLA from Distillation Products Industries (Eastman No. 7727) and General Mills, Inc. (Alamine 304) were also used to prepare extractant. Other diluents used in these irradiation studies were n-dodecane (Humphrey Wilkinson, Inc., olefin-free grade) and a mixture of normal paraffin hydrocarbons, principally Clo through (213, sold by the South Hampton Co. and identified as NPH in this report. TBP was purchased from the Commercial Solvents Corp. Hydrazine was obtained from the Olin Mathieson Chemical Co. as an 11M stock solution. All other chemicals were of reagent grade quality. Except in experiments with actual plant 1WW solution, decontamination of neptunium and plutonium was followed with the aid of the radioactive isotopes Srsj, Fej9, Crjl, Y88, Cs137, Ru'o6, Zrgj-Nbgs, Ce144-Pr144, and Eu'j2-'54. Yttrium-88 was obtained from Nuclear Science and Engineering Co. ; the other isotopes were obtained from Oak Ridge National Laboratory. I n some experiments NpZ39, a 2.35-day gamma emitter isolated and purified from recently irradiated uranium, was added to synthetic 1WW solution. The use of this isotope greatly facilitated study of the extraction of neptunium by irradiated TLA solutions. Extraction-Strip Contacts. A standard 10-minute contact period (mechanical stirring) at 25' to 26' C. was allowed for attainment of equilibrium under extraction conditions. Variable times and temperatures, as appropriately noted, were used in strip contacts. Phases were separated by either centrifugation (synthetic 1WW) or gravity settling (plant 1WW). Equal volumes of aqueous and organic phases were used throughout. Distribution ratios are listed in some tables. The distribution ratio is defined as the ratio of the concentration of the element of interest in the less dense organic phase to the concentration of the element in the more dense aqueous phase after equilibration of the two essentially immiscible liquids. Mixer-Settler Runs. Effects on Purex process H A and HS column performance of the presence of small amounts of TLA in the Purex process extractant were studied in mixersettler runs. The 19-stage units used were Hanford-designed versions of a type developed at the Knolls Atomic Power Laboratory by Coplan and others (8). Aqueous feed for these runs was prepared from laboratory chemicals and spiked with Hanford Purex plant dissolver solution to provide a source of fission product activity and plutonium. Mixer-settlers were operated with the particular feed, scrub, and organic solutions required until a steady state in the effluent streams was reached. Flow rates were maintained either with metering pumps or by a syringe-drive feed system. Samples of the effluent streams were taken hourly and analyzed to determine when steady state was reached. Waste losses and decontamination factors were computed from analyses of steady-state effluent streams. Solvent Irradiation Studies. Irradiation studies were performed in a Coco radiation facility where the dose rate was 1.1 X 106 rad per hour as determined by Ce(SOJ2 dosimetry. Initially 0.3M TLA-diluent solutions were irradiated at 25' to 30' C. in the absence of an aqueous phase for 4- to 72-hour periods. I n further experiments equal volumes of 0.3M TLAdiluent solutions and either 4M "01 or synthetic 1WW solution containing 0.02 mole per liter of hydrazine were in contact (with agitation) continuously a t 25' to 30' C. for as long as 72 hours while in the irradiation facility. Finally, in one experiment equal volumes of 0.3M TLA-Soltrol-170 and synthetic 1WW were in contact 1 hour in the C060 facility, The resulting organic phase was in contact outside the radiation field for 30 minutes at 50' C.with an equal volume of 0.05M (NH2OH)z. H2S04 solution. This irradiation-strip cycle was repeated 100 times, each time with a fresh portion of synthetic 1WW solution. I n all cases, capacity of the irradiated solvents to extract neptunium(1V) was determined with a fresh, unirradiated portion of synthetic 1WW solution containing hydrazine and both NpZ37and NpZ39. VOL. 6
NO. 1
J A N U A R Y 1 9 6 , 7 117
Analyses. NeptuniumZ37 and plutonium in aqueous solutions were determined by sequential T I O A and thenoyltrifluoroacetone extractions and alpha counting (77). Neptunium and plutonium in TLA solutions were determined by direct alpha counting of aliquots. Alpha pulse height analyses were also performed whenever required. Gamma activity of solutions which contained only a single radioisotope was counted with a gamma scintillation counter. Gamma energy pulse height analyses were performed on samples derived from plant solutions to determine concentrations of individual radioisotopes. Results and Discussion
Valence Adjustment, Extraction Contact. From its manner of preparation and composition and from known aqueous actinide chemistry, Purex plant 1WW solution might be expected to contain plutonium and neptunium in the +4 and +5 states, respectively. This expectation is realized for plutonium (Table 11). But, although some Np(V) is apparently present in plant l W W , a sizable fraction of the neptunium exists in a form, presumably the $4 state, readily extractable into TLA solutions. Russian workers have noted that the addition of sulfate ions to a nitric acid solution of Np(1V) considerably increases its stability (70) ; perhaps this effect accounts for the presence of Np(1V) in plant 1WW solution. I n any case, Np(V) in 1WW solution must be reduced to Np(1V) before 96 to 98y0of the neptunium can be extracted in a single batch contact. This is readily accomplished by adding hydrazine to the 1WW solution and allowing the resulting mixture to stand 30 to 60 minutes at about 25' C. Hydrazine, like most reductants-e.g., ferrous sulfamatewhich reduce Np(V) to Np(IV), is thermodynamically capable of reducing Pu(1V) to Pu(II1). Although hydrazine rapidly reduces Np(V) to Np(IV), the rate of reduction of Pu(1V) to Pu(II1) in 1WW solution a t 25' C. is very low. [Burney ( 4 ) has also noted the low rate of hydrazine reduction of Pu(1V) in nitric acid solution.] Advantage can thus be taken of the difference in reduction kinetics to establish both elements in the quadrivalent state. A concentration of hydrazine in the range 0.02 to 0.05M is satisfactory. Batch contacts of TLA and 1WW solutions in plant equipment can be performed satisfactorily a t equal phase volumes. Extraction of more than 95y0of both neptunium and plutonium in a single batch contact at this volume ratio requires the extractant to be at least 0.3M TLA. (Since neptunium and plutonium recovery were adequate at 0.3M TLA, extractionstrip studies were not made a t higher TLA concentrations.) At all TLA concentrations tested, and even with hydrazine added to the 1WW solution, plutonium extracted better than Table II. Typical Results in TLA Extraction of Np and Pu
(0.3M TLA unless otherwise indicated) 1WW Source
Synthetic
M
Hydrazine Addition Time stood, min.a
Plant
0.0 0.0
Plant -
0.05
Synthetic Synthetic Plant Synthetic Synthetic* Svntheticd Sinthetic Synthetic
0.05
0.022 0.022 0.022 0.022 0.022 0.011 0.011
0 0
30 30 30 30 60 30 30 30 60
70 Not Extracted Np 50.5 12.0 2.6 2.7 3.4 3.3 3.2 5.8 21.7 3.2 7.1
Pu
Table 111.
Stripping of Neptunium and Plutonium from TLA Phase
(Single batch contact of 0.3M TLA-Soltrol-170 solution containing Pu and Np)
0.7 0.7
1.2 0.5 1.5 0.8 1.6 1. b e
3.4c 0.9
At 25' to 26' C. after hydrazine addition. b 0.2M T L A . In absence of hydrazine. d 0.lM TLA. Q
119
neptunium. Because of the high concentration of sulfate, however, plutonium and neptunium extracted less efficiently from 1WW solution than from nitrate solutions of the same acidity (6). Other results (not shown in Table 11) demonstrated that neptunium and plutonium extraction recoveries were only slightly changed by a hundredfold increase in the uranium concentration of the 1WW solution or a twofold decrease in the nitric acid concentration. Neptunium and plutonium extraction were significantly better, however, when the 1WW solution was 0.2M rather than 0.8M sulfate. Additional experiments also demonstrated that 90 to 95% of the equilibrium extraction of neptunium and plutonium was attained in less than a minute of vigorous agitation. T o ensure adequate mixing of phases in present Hanford Purex Plant equipment, however, a 30-minute extraction contact time is proposed (Figure 2). Mixtures of TLA extractant and either plant or synthetic 1WW solution disengaged readily even without centrifugation. Disengaging behavior was not affected by the presence of finely divided solids (sulfate and phosphate precipitates and siliceous material) in plant 1WW solution. Pu-Np Strip Contact. Aqueous solutions which contain complexants (sulfate, oxalate, etc.) for Np(1V) and Pu(1V) effectively strip these elements from TLA solutions. Plutonium(1V) may also be removed from the TLA phase by reducing it to the less-extractable Pu(II1). Thus, depending on conditions, plutonium and neptunium may be either costripped or stripped consecutively. Costripping was emphasized in this work. Most efficient costripping was obtained (Table 111) by contact of the TLA phase 30 to 60 minutes at 50' C. with an equal volume of 0.05 to 0.1M (NH20H)z. H2S04. Hydroxylamine sulfate combines in a single reagent a reductant and a complexant. Reductive stripping of plutonium is much faster at 50' C. than a t 25' C. Because of decreased viscosities and densities, phase separation at 50' C. is also much faster than at 25' C. Flash point and vapor pressure data for typical Purex process kerosine diluents indicate that stripping at 50' C. can be conducted safely in plant equipment. Ferrous sulfamate solutions strip plutonium well but are ineffective in stripping neptunium. Sulfuric acid (0.05M) and 0.05M Na2S04 strip neptunium as efficiently as 0.05M ( N H 2 0 H ) 2 . H z S 0 4but, of course, are less efficient for plutonium removal. However, combination of sodium sulfate with ferrous sulfamate leads to a highly efficient reagent for stripping both elements. This combination is a satisfactory alternative to hydroxylamine sulfate solutions. Most satisfactory partition stripping of plutonium and neptunium in a single batch contact was accomplished a t
16EC P R O C E S S D E S I G N A N D D E V E L O P M E N T
Strib Conditions Time, T:mj., min. C. -
Strip Comjosition 0 . 0 5 M (NH2OH)z.HzS04 0.05M (NHz0H)z.HzSO4
0.05M (NHzOH)z,HzSOa 0.10M (NH;OH'&. HiSOi 0.05M HzSOa 0.05M NazSOa 0.05M ferrous sulfamate
0. I O M
ferrous sulfamate 0.05M Na2S04-0.05M ferrous sulfamate 0.10M ferrous sulfamate1 .OM " 0 3
1
30 30 60 30 30 30 30
25 50 50 50 50 50
% Not Stripped Np Pu 3.5 3.6 2.4 1.9 2.7
50
IO
25
2.9 15
6.4 0.70 0.79 0.23 8.6 8.2 0.34 48
30
50
2.9
0 24
60
50
97
4.0
50' C. by using nitric acid solutions containing 0.1M ferrous sulfamate. At 1.OM H N 0 3 the neptunium distribution ratio was sufficiently high that only 3% of the neptunium costripped with 96y0of the plutonium. The 0.3M TLA phase. after contact with the 1WW solution, per liter. About contains approximately 0.62 mole of " 0 3 half sf this represents RaNH+NOa- extracted according to Equations 1 and/or i!. T h e other 0.32 mole represents excess nitric acid extracted into the organic phase. Essentially all the excess nitric acid strips into the 0.05M (NH20H)z HzS04 solution along with the plutonium and neptunium. Decontamination Performance. Excellent decontamination of neptunium and plutonium from inert and radioactive contaminants is realized in the extraction contact with TLA. Such a result is to be anticipated, of course, from the experience of previous workers with tertiary amine extraction systems (7, 5). The distribution ratio of ruthenium is about tenfold higher than that of cerium in both plant and synthetic 1WW solutions (Table IV). Isotopes of these elements extract considerably more than any of the other contaminants of neptunium and plutonium. Ruthenium may extract as a nitrosyl species; extraction of nitrosylruthenium by TLA in nitrate systems has been extensively studied by Skavdahl and Mason (20). Cerium may be extracted as C ~ ( N O S ) by ~ - a~ mechanism similar to that shown in Equation 3. Essentially all the extracted cerium transfers to the aqueous phase along with the neptunium and plutonium during the strip contact with O.05M ( N H Q O H ) H2S04 ~ solution. However, some further decontamination from ruthenium and zirconium-niobium is obtained in this contact. A liter of the final 0.05hii ( S H 2 0 H ) 2 H 8 0 4 strip solution obtained from plant l\WV solution is expected to contain about 1 curie each ' G -Rh106 and Ce144 -Pr144 and about 0.1 curie of of RuO Zr95 --Nb95. Final N p Purification Procedures. Neptunium recovered by the TLA process must be separated and purified from plutonium and other inert and radioactive contaminants. A convenient purification scheme is that now used in the Hanford Purex Plant--namely, a TBP solvent extraction cycle in the 3A and 3B columns and a final anion exchange purification step (Figure 1). Detailed chemical flowsheets for the accumulation and decontamination phases of the TBP extraction cycle (77) are suitable for recovery of neptunium from either hydroxylamine sulfate or sodium sulfate-ferrous sulfamate strip solutions. Batch contact data show that sulfate ion in such solutions impairs TBP extraction of Np(1V) but does not prevent preparation of satisfactory extraction column feeds. Normal TBP cycle fission product decontamination was ob-
served in other batch contacts with feeds prepared from TLA strip solutions. Available literature (4) indicates that anion exchange resin purification techniques should be applicable to purifying neptunium in TLA strip solutions. Thus, some consideration was given to bypassing the 3A-3B system and, instead, routing the strip solution to the plant neptunium ion exchange purification unit recently described by Duckworth and LaRiviere (9). This approach proved untenable because the shielding provided in the installed unit is not sufficient to provide adequate protection from the gamma radiation from Ru106-Rh106and Ce144-Pr144 in the strip solution. Also, the present unit is designed to operate with feeds having a Np-Pu ratio of 10 to 100. I n the TLA strip the Np-Pu ratio is 1 or less. At this ratio Burney's results (4) indicate that neptunium and plutonium could not be adequately separated in a single separation cycle. Compatibility of TLA with Purex Process. Incorporation of a TLA extraction process in the Hanford Purex complex would convert the plant from the traditional one-solvent to a two-solvent plant. Mixer-settler runs were made to study the effects on Purex process H A and H S column performance of the presence of TLA in the 30 volume % TBP-Soltrol-170 extractant. U p to 5 volume % TLA in the TBP extractant had no significant effect on uranium extraction or on decontamination from Zrgj-Nbgj (Table V). Plutonium extraction actually improved with increased TLA concentration. Decontamination from ruthenium, as might have been anticipated from the results discussed earlier, decreased sharply when TLA was added to the Purex process extractant (Figure 3). Decreased decontamination from ruthenium was reflected in decreased decontamination from gross gamma activity. The TLA process is a low-volume batch operation which, as currently projected, would be performed in a section of the plant isolated from the main-line process solvent. Loss of the entire quantity of 0.3M TLA extractant to the TBP solution would be improbable. However, even in such an unlikely event, the Purex process solvent would contain only 0.5 to lyOTLA. I t is concluded from Table V and Figure 3 that Purex process performance would not be seriously impaired at this TLA concentration. Any loss in ruthenium decontamination across the first extraction cycle could probably be made up in later cycles. Also, the presence of 1% TLA in the TBP extractant would not be expected to affect reductive stripping of plutonium in the IBX column. Solvent Degradation, Radiolytic a n d Chemical. The 0.3M TLA extractant receives radiation at an exposure rate of about l o 6 rad per hour when in contact with the highly radioactive plant IWW solution. Solvent radiolysis is thus a
Decontamination Performance of TLA Process (Extraction and strip contact made with equal volumes of aqueous
Table V.
Effect of Contaminating Purex Process Solvent with TLA Mixer-settler runs. 11 extraction, 8 scrub stages
Distribution Ratio Extraction Synthetic 7WW Plant 7WW
Organic. 30% TBP-Soltrol-170 containing indicated vol. % TLA Flows. Feed = 100, scrub = 54, organic = 400
Table IV.
and organic phases acc:ording to conditions shown in Figure 2)
Contaminant
Ru'Oa R u lOS-Rh 106
Ce144Lpr144 ZrQS-Nb96
Fe Y EU Cr Sr cs
0.044
0.0061 0 . 00019 O.ClOl0 0.00061
0.077 0.074 0.0072 0.00053
Strip, plant 1WW
Feed.
HzS04-0.008M
1.8
1.9 0.002 0.29
0
0.5 1 .o 2.0 5.0
0 . Cl0050 0 ,00023 0 ,00014
0.00003
1.3M U O Z ( N O ~ ) Z - ~"03-0.025M .~M
a
0.80 1.I6
1.6
0.55
0.27 0.45
0,68 0.78
0.21 Other decontamination factors gioen in Figure 3.
VOL. 6
NO. 1
>loo0 1000
1800 1200 >550
J A N U A R Y 1967
119
I
.L
0
m U
U
e
\
.-0 #
m
Gross Gamma
c ._
E m
e
0
a u
n
I
I 0
I
/
I
3
2
1
/
4
I
I
5
P e r c e n t TLA i n P u r e x Process,Solvent
Figure 3. Effect of TLA on Purex process fission product decontamination Mixer-settler conditions listed in Table V
matter of prime importance in this system. Baroncelli and coworkers (7) have noted that scarce and sometimes conflicting information exists in the literature on the radiation stability of amine-diluent solutions. A study of chemical as well as radiolytic degradation of TLA-diluent solutions, when exposed to plant and synthetic 1WW solutions, is still in progress. Preliminary results are discussed in this paper. T h a t irradiation conditions highly influence the extent of radiation damage of amine-diluent solutions has been reported by both Italian and French workers (7,5). O u r results (Figure 4) are in agreement with this observation, in that neptunium extraction capacity of irradiated TLA solutions depends strongly on the presence or absence of an aqueous phase during irradiation. Thus, radiolytic destruction of TLA (as measured by the neptunium distribution ratio) is apparently a first-order reaction when the extractant is irradiated while in contact with I W W solution. When an aqueous phase was not present, however, the neptunium extraction capacity fell by about a factor of 2 on even small irradiation (ca. 5 x 106 rad) ; further irradiation to an exposure of 8 x IO7 rad (216 watt-hours per liter) resulted in only small additional decreases in neptunium
401 30
while in contact with IWW solution
F 20
-
9
12
24
36
48
60
72
04
irradiation Dose, rad ( x iC7)
Figure 4. Extraction of neptunium by irradiated TLA-paraffin diluent
0.3M
Average distribution ratios obtained with extractants prepared from various TLA and diluent sources
120
I&EC PROCESS DESIGN A N D DEVELOPMENT
distribution ratio. The beneficial effects of the presence of an aqueous phase during irradiation of TLA-diluent solution have been noted by Italian workers (7). TLA from the various sources tested all appeared to degrade at about the same rate. Also, substitution of one paraffin diluent for another had no effect on the rate or degree of TLA radiolysis. Finally, the extent of radiolytic destruction of TLA when in contact with 1WW solution was not altered by contact of the TLA phase intermittently outside the radiation field with hydroxylamine sulfate strip solution. White solids precipitated from all irradiated TLA-paraffin diluent solutions. White solids did not form when either 0.3M TLA-benzene or 0.3M Alamine 336-Soltrol-170 solutions were irradiated to an exposure of about 4 X lo7 rad (108 watt-hours per liter), Solids did precipitate, however, in irradiated 0.3M TLASoltrol-170 solutions which contained initially 2 volume % ' n-octyl alcohol. The latter reagent is often used to prevent formation of a second organic phase in amine extraction systems. Formation of solids in irradiated TLA solutions was also noted recently by Martin (74) of this laboratory and, apparently, in 1958 by Wagner and Towde (22). T o the author's knowledge, however, a detailed account of the formation, composition, and properties of these solids has not been given. White solids produced in these experiments have not been identified. Infrared analyses demonstrate that they are not a simple tertiary amine oxide or a secondary amine, two likely possibilities. Our preliminary results indicate that the white material is a polymeric substance whose composition depends on irradiation dose and the presence or absence of an aqueous phase during irradiation. In this latter connection, infrared analyses indicate the presence of nitroso groups in solids obtained from TLA solutions irradiated while in contact with "01 or synthetic 1WW solutions. These particular solids may be dimeric (or higher polymeric) nitrosoamines produced from the reaction of nitrous acid and secondary amines. Dilaurylamine is known ( 5 ) to be one of the primary radiolysis products of TLA. Chemical, as well as radiolytic, degradation of TLA-dihent solutions can result in formation of solids. This was demonstrated in experiments in which equal volumes of synthetic 1WW solution and either 0.3M TLA-Soltrol-170 or 0.3M TLA-NPH were in contact continuously for one week a t 30' C. in the absence of a radiation field. The organic solutions thus obtained were initially clear, but, on standing, white solids began to precipitate from the TLA-Soltrol-170 solution after about 9 days and from the TLA-NPH solution after about 3 weeks. Solids continued to precipitate thereafter as both solutions were allowed to stand a total of 3 months. Coincident with precipitation of solids, the neptunium distribution ratio for both TLA solutions fell from an initial value of about 24 immediately after the contact with 1WW solution to 8 to 10 after standing 3 months. For about 1 month after contact with 1WW solution, the neptunium distribution ratio for both solutions remained a t about 24. Solids isolated from these solutions appear to contain nitroso groups and are believed to be identical to those obtained from TLA solutions irradiated while in contact with 1WW solutions. Finally, complete loss of TLA from the organic phase occurred when equal volumes of 0.3M TLA-Soltrol-170 and synthetic 1WW solutions were in contact 48 hours at 50' C . Although white solids similar to those discussed previously were not observed, a thick scum, possibly containing the missing TLA in some form, was present at the aqueous-organic phase
interface. British workers (73) have also noted that chemical degradation of tertiary amines by nitric and/or nitrous acids becomes more severe at elevated temperature; in their work, contact of tricaprylamine with 3M H N 0 3 at 100’ C. for 120 hours completely (destroyed the amine. They suggest that the reactions with nitrous acid involve decomposition of tertiary amine to secondary amines and conversion of secondary amines to nitrosoamines. O u r results appear in line with this suggestion. These results, although to some extent still regarded as preliminary, demonstrate that both radiolysis and/or chemical reaction with constituents of 1WW solution can cause loss of TLA from the extractant. Such loss naturally constitutes a disadvantage to the proposed extraction process, but not, it is believed, an insurmountable one. Thus, extractant losses can be minimized by controlling the temperature a t 25’ C. during the extraction step and, particularly, by limiting exposure to 1WW solution to minimum times consistent with extraction kinetics and contactor design. Contactors having short residence times, such as a single-stage centrifugal mixer-settler of a typr recently described (24),may be applicable to this system. Use of NPH as the diluent may also be of value in improving solubility of TLA degradation or reaction products. Effects, if any, of periodic washing of the extractant with sodium hydroxide solutions, as i:; done in the Italian Eurex process ( Z ) , on the radiation stability of the TLA solvent have not been determined. Acknowledgment
The contributions of D. G. Bouse, L. C. Neil, and G. E. Smedberg to this work are gratefully acknowledged. T h e author thanks S. J. Beard, S. M. Nielson, and R. C. Forsman of Isochem, Inc., and E. C. Martin of the Pacific Northwest Laboratory for many stimulating and helpful discussions.
(5) Chesne, A,, Koehly, G., Bathellier, A., Nucl. Sci. Eng. 17, 557 (1963). (6) Coleman, C.F., Zbid., 17, 274 (1963). (7) Coleman, C. F., Kappelmann, F. A., Weaver, B., Zbid., 8, -5n7 - .( i~9 m ) .
(8) Coplan, B. V., Davidson, J. K., Zebroski, E. L., Chem. Eng. Proer. 50. 403 (1954). (9) Suckwbrth, J. P., LaRiviere, J. R., IND.ENG.CHEM.PROCESS DESIGN DEVELOP. 3, 306 (1964). (10) Gel’man, A. D., Moskvin, A. I., Zaitsev, L. N., Mefod’eva,
N. P., “Complex Compounds of Transuranium Elements,” trans. by Nigel Turton and T. I. Turton, p. 3, Consultants Bureau, Inc., New York, 1962. (11) Isaacson, R. E., Judson, B. F., IND.ENG. CHEM.PROCESS DESIGN DEVELOP.3, 296 (1964). (12) Keder, W. E., Wilson, A. S., Nucl. Sci. Eng. 17, 287 (1963). (13) Lane, E. S., Pilbeam, A., Fletcher, J. M., “Reprocessing with Amine-Ether Systems. Part I. Choice of Solvent System,” U. K. Atomic Energy Auth. Rept. AERE-R4440, Pt. I(1965). (14) Martin, E. C., Pacific Northwest Laboratory, Battelle Memorial Institute, Richland, Wash., private communication, 1964. (15) Moore, F. L., “Liquid-Liquid Extraction with High Molecular-Weight Amines,” National Academy of Sciencies, Nuclear Ser. WAS-NS-3101 (1960). (16) Poe, W. L., Joyce, A. W., Martens, R. I., IND.ENG.CHEM. PROCESS DESIGN DEVELOP. 3,314 (1964). (17) Schneider, R. A., Harmon, K. M., eds., “Analytical Technical Manual,” U. S. Atomic Energy Comm. Rept. HW-53368 (1961). (18) Schulz, W. W., “Solvent Extraction of Strontium, Cerium and Rare Earth with D2EHPA. Part 1. Laboratory Studies,” U. S. At. Energy Comm., Rept. HW-79762, Pt. 1 (1964). (19) Sheppard, J. C., “Extraction of Neptunium( IV) and Plutonium( IV) from Nitric Acid Solutions with Tri-n-octylamine,” U. S. At. Energy Comm., Rept. HW-51958 (1957). (20) Skavdahl, R. E., Mason, E. A., “Solvent Extraction of Nitrosylruthenium by Trilaurylamine in Nitrate System,” U. S.At. Energy Comm., Rept. MITNE-20 (1962). (21) Smith, E. L., Page, J. E., J . SOC.Chem. Znd. (London) 67, 48 (1948). (22) \Vagner, R. M., Towle, L. H., “Radiation Stability of Organic Liquids, Semiannual Report No. 4, July 1 to December 31, 1958,” U. S. Atomic Energy Comm., Rrpt. AECU-4045 (1959). (23) Weaver, B., Hornet-, D. E., J . Chem. Eng. Data 5, 260 (1960). (24) .Williamson, C. L., Ward, J. F., Webster, D. S., “Centrifugal Mixer-Settler,” U. S. At. Energy Comm., Rept. DP-370 (1962).
Literature Cited
(1) Baroncelli, F., Calleri, G., Moccia, A., Scibona, G., Zifferero, N., h’ucl. Sei. Eng. 17, 298 (1963). (2) Baroncelli, F., Scibona, G., Zifferero, N., J.Znorg. Nucl. Chem. 24,405 (1962). (3) Brown, K. B., et al., Proc. 2nd Intern. Conf. Peaceful Uses Atomic Energy, Geneva 3,472 (1.948); P/509. (4) Burney, G. A,, IND.ENG. CHEM PROCESS DESIGNDEVELOP. 3, 328 (1964).
RECEIVED for review April 11, 1966 ACCEPTED September 12, 1966 Division of Nuclear Chemistry and Technology, 149th meeting ACS, Detroit, Mich., April 1965. Work performed under contracts AT(45-1)-1350 and AT(45-1)-1830 between the U. S. Atomic Energy Commission and the General Electric Co. and Battelle Memorial Institute, respectively.
VOL. 6
NO. 1
JANUARY 1967
121
