HW-60116CHEMISTRY -- SEPARATION PROCESSES FOR

PLUTONIUM AND URANIUM

DESCRIPTION OF PUREX PLANT PROCESS rIFts CoeicFFR y

ByE. R. Irish

May 19, 1959[Date Declassified]

Hanford Atomic Products Operation

General Electric CompanyRichland, Washington

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UNITED STATES ATOMIC ENERGY COMMISSIONTechnical Information Service

Work performed under Contract W-31-109-Eng-52.

LEGAL NOTICEThis report was prepared as an account of Government sponsored work. Neither the UnitedStates, nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accu-racy, completeness, or usefulness of the information contained in this report, or that the useof any information, apparatus, method, or process disclosed in this report may not infringeprivately owned rights; or

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This report has been reproduced directly from the bestavailable copy.

Printed in USA. Price $0.50. Available from the Office ofTechnical Services, Department of Commerce, Washington25, D. C.

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DESCRIPTION OF PUREX PIA1T PROCESS

I . INTRODUCTION

The Purex process is a continuous solvent extraction process for irradiateduranium for the separation and decontamination of plutonium and uranium fremeach other and from fission products. Much information has been documented(2,11) in the unclassified literature in regard to the process bases, chemistry,and possible flowsheets based on laboratory and pilot plant work. However,since the Purex Plant has been in production operation at the Hanford AtomicProducts Operation since 1956, many revisions have been made to the processflowscheme and details. These process improvements have been the result ofcoordinated in-plant and laboratory development programs. The purpose of thisdocument is to present a brief summary, with reference literature for details,of pertinent and important process flowsheet conditions (12, 19) which are inuse in the Purex Plant.

II. OVER-ALL PROCESS DESCRIPTION

The Purex Plant process can be divided into seven major steps or unit oper-ations:

1. irradiated uranium dissolution (including jacket removal)and feed preparation;

2.. gross decontamination and recovery of uranium and plutoniumfrom fission products;

3. partitioning of the uranium and plutonium;

4. final decontamination and recovery of plutonium (includinganion exchange tail-end treatment);

5. final decontamination and recovery of uranium;

6. solvent recovery; and

7. nitric acid recovery (including waste concentration anddisposal).

In addition, several auxiliary steps are required for the coordinated oper-ation of the plant.

Initially, the solvent extraction portion of the plant consisted of: a) acodecontamination cycle for gross decontamination of plutonium and uraniumfrom fission products; b) a partition cycle for further decontamination fromfission products and separation of plutonium and uranium from each other;and c) final decontamination cycles for both plutonium and uranium. Finalproducts were concentrated by boil-down to concentrated nitrate solutions

1

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for further processing. Wastes from each cycle were concentrated; nitric acidwas recovered by distillation for re-use in the process, and the wastes wereneutralized and stored. Used solvent was chemically washed and recycled tothe process.

As discussed in Reference 11, the above flowscheme is only one of severalpossible arrangements of the Purex process. A major improvement in the pro-cessing scheme has been incorporated in the Purex Plant by conversion from athree-cycle to a two-cycle solvent extraction flowsheet employing the back-cycling of waste (2). Figures 1, 2, and 3 show the details of the over-allprocess flowsheet, and Table I summarizes the design criteria for the solventextraction columns (6, 12). Over-all process performance with these flowsheetsand process equipment has been demonstrated on a plant scale as follows:

Separation of U from Pu >107

Separation of Pu from U 106

Decontamination of F.Ps from Pu > 10First Cycle IF = 2 x 10Second Cycle IF = 2 x 102Anion Exchange IF =>3

Decontamination of F.Ps from U 107First Cycle IF =2 x 10Second Cycle IF = 5 x 102

Plutonium & Uranium Recovery 99.9%

Acid Recovery - Dissolving 70%- Solvent Extraction 95o

Solvent Recovery 99.7%

The major items contributing to the successful transition from the three-cycleto the two-cycle flowsheet include:

1. improved pulse column technology;

2. improved solvent treatment methods;

3. new technology for anion exchange purification of plutonium;and

4. increased process "know-how".

In the following sections each of the major unit operations will be describedbriefly in conjunction with reference to appropriate flowsheet Figures 1, 2,and 3. Notes on these figures define the flow bases, temperatures and otherprocess conditions.

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III. UNIT OPERATIONS IESCRIPTIONS

A. Uranium Dissolution and Feed Preparation

The uranium dissolving operation is a two-step batch process: coatingremoval and slug dissolution. Aluminum-canned, irradiated uranium slugsare charged into a dissolving vessel on top of an equal weight of bareuranium slugs (a "heel" left from the prior operation) which are allcovered with a sodium nitrate solution. In the dissolution of the aluminumjackets, concentrated sodium hydroxide is added to the dissolver slowly,limited only by the reaction rate. In the presence of nitrate ion, hydrogenformed by the reaction of aluminum and hydroxide ion is converted to ammoniagas which is diluted with air and discharged to the off-gas system. At theend of the coating-removal operation, usually after two to three hours di-gestion, the solution is jetted from the dissolver; water rinse of thedissolver is made to remove residual chemicals. The resulting coating-waste solution is stored indefinitely in large mild-steel undergroundtanks (15).

The uranium charge is normally dissolved in two steps. Concentrated nitricacid is next added to the dissolver vessel in a ratio of approximately 3.5moles of acid per mole of uranium to be dissolved. The solution is heatedto boiling, and dissolving occurs for three to six hours until the solutionspecific gravity reads 1.75 at the boiling point. At this time the solution,still containing free acid, is cooled to prevent further acidity reduction(which would permit polymerization of plutonium to an inextractable form)and jetted from the dissolver to another tank preparatory to further pro-cessing. The process is repeated to remove a second batch of solution,leaving a residual "heel" of uranium. Finally, the dissolver is rinsedwith water to minimize loss of plutonium and uranium in the subsequentcoating-removal operation.

During the uranium dissolution process, the oxides of nitrogen are passedthrough a water-cooled, downdraft ref lux condenser which returns significantacid to the dissolver. Unreacted oxides leave the condenser, are heatedto 190 C, and pass through a silver reactor (17) for removal of iodine,are filtered through a glass fiber filter (17), and are discharged to astandard nitrogen oxide recovery unit. Approximately seventy percent ofthe acid is recovered. All equipment prior to the final acid absorbersis located in the remotely-maintained and operated canyon building (3, 8).

The dissolver solution containing plutonium, uranium, and fission productsis next centrifuged (1,000 G's with a ten-minute hold-up) to removesiliceous solids from the solution, primarily aluminum-silicon compoundsfrom the slug-jacket bond. If radio-iodine suppression is needed, 5x10 4Mmercuric nitrate is added to minimize the evolution of iodine during centri-fugation and further processing. All process vessels (excluding dissolvers)are vented through a common silver reactor for removal of radio-iodine andfiltered through a common glass fiber filter prior to discharge to theatmosphere.

Hw-6116

Following centrifugation, the solution is adjusted to the proper uraniumand acid concentration for solvent extraction. Normally, no plutoniumvalence adjustment with sodium nitrite is necessary because of the pre-sence of adequate nitrite ion resulting from radiolysis of nitrate ion inthe dissolver solution.

B. Gross Decontamination from Fission Products

With reference to Figures 1 and 2, and solvent extraction process de-scriptions elsewhere (2, 11), stepwise descriptions of the solvent ex-traction flowsheets are unnecessary. However, major facets will bementioned. In regard to the HA-HS Columns, essentially the HS Column isan extension of the HA Column scrub section, but being separate, it may beoperated at 70 C with the HA Column operated at 35 C for maximum decon-tamination as discussed in Reference 2. This procedure has not beenadopted in the Purex Plant, however, because all solvent solutions aremaintained below MC for safety reasons with existing equipment andsolvent.

The HAO stream is a solvent stream consisting of backcycled IBSU (returningfission products removed in the IBS Column)and H00 recycled solvent usedto minimize uranium reflux in the HA and HS Columns. Sodium nitrite maybe added to oxidize any entrained ferrous ion from the IBS Column, but thisis not usually necessary.

The 3WB stream, as shown in Figure 3, is a concentrated solution of secondcycle wastes being backcycled (2) for recovery of plutonium and uraniumlosses from these cycles. These waste acid solutions are steam-strippedthrough six bubble-cap trays for removal of entrained solvent prior to con-centration. Most of the nitric acid entering the HA Column enters viathis solution.

Both the HA and HS Columns have special cartridges defined in Table I andare operated with the solvent phase continuous. Thus, the aqueous-organicinterfaces are located at the bottom ends of the columns; highly radio-active solids which accumulate on these interfaces are entrained with theaqueous phases leaving the columns and thus do not follow the product-bearing solvent phases leaving the columns. The combination of the specialcartridges and the bottom-interface locations resulted in a 5- to 10-foldimprovement in fission product decontamination compared with top-interfacelocations.

Operation of the columns with bottom interfaces was dependent on twomajor developments. Satisfactory plastic plates with adequate radiationstability (integrated exposure greater than 2x109 R) were made fromlinear polyethylene. Satisfactory interface control (under pulsed columnand slightly emulsified conditions) was achieved by use of a float-typeinterface detector (7) for remote locations.

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C. Partitioning of Plutonium and Uranium

Two columns, IBX and IBS, are utilized for partitioning of plutonium anduranium. Use of two short columns rather than one long column permitseasier installation of the columns in canyon cells. Also, such a flow-scheme permits the IBSU backcycle which provides additional fission pro-duct decontamination for both uranium and plutonium.

D. Final Plutonium Decontamination

As shown in Figure 2, the plutonium is further decontaminated from fissionproducts by solvent extraction in the 2A and 2B Columns and from bothfission products and uranium Dy an anion exchange purification step (16).The 2A Column is operated with the solvent phase continuous.

All equipment downstream of the partitioning step is designed to be geo-metrically safe for plutonium solutions. This feature is important from thestandpoint of preventing a criticality excursion when processing quiteconcentrated plutonium solutions. Two other concepts are used for criticalmass control elsewhere in the plant: (a) batch-size control where feasible,and (b) safe solution concentration elsewhere. The primary critical masscontrol feature, however, is maintenance of chemical conditions (1) toprevent the precipitation and accumulation of a critical mass of plutonium.In the acidic Purey process, the problem exists primarily only where plu-tonium concentrations become high.

Solvent for the Final Plutonium Cycle is shown to be from the No. 2 SolventSystem, the low-activity solvent used also in the Final Uranium Cycle, formaximum fission product decontamination. If the No. 1 Solvent System, thehigh-activity solvent used in the First Cycle, provides adequate decon-tamination of the solvent from fission products, 2AX can be supplied fromthis system; such has proved to be possible in the plant when the alkaline-permanganate solvent washing process (described later) is used. In thisway the 2BW can be backcycled to IBXF, but caution must be used to preventrouting of excessive nitrite ion (very extractable) to the IBXF andoxidizing the IBX Column solutions.

The anion exchange process is performed in Higgins (9, 10) continuous anionexchange columns using Permutit SK resin (20 - 40 mesh). A 0.045 M_ H2SO4solution is used in 2BX in order to complex Pu IV to permit a low 2BP flow;thus, the XAF flow rate is maintained at a minimum to maximize the hold-uptime in the XA Column for anion resin adsorption, kinetics of adsorptionbeing the controlling factor. Following elution of the plutonium from theXC Column, the solut ion is concentrated in a titanium concentrator tominimize introduction of impurities from corrosion; tantalum would alsobe suitable for this purpose.

E. Final Uranium Decontamination

As shown in Figure 2, the ICU stream is first steam stripped (through sixbubble-cap trays) for removal of entrained solvent and then concentrated.

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(Steam pressure is limited to a maximum of 30 psig for safety reasons.)Ferrous sulfamate is added to the 2IF to reduce trace amounts of plu-tonium so that the plutonium will not be extracted in the 2D Column. The2D Column is operated with a high degree of saturation, resulting inapproximately five percent of the uranium in the 2D Column feed beingbackcycled to 3WB, in order to maximize fission product decontamination.The 2EU is concentrated in equipment identical to that used for ICU and3WB and transferred to a storage tank for further processing by calcin-ation to the oxide.

The 2D Column is operated with the solvent phase continuous and a bottominterface similar to the HA Column. A twenty-fold increase in fissionproduct decontamination was effected by the conversion of this column fromaqueous to solvent phase continuous even though the "solvent extraction"decontamination factor should not have been different. This significantimprovement resulted because decontamination in this column is limited byparticulate rather than soluble fission products (primarily Zr-Nb).

The plastic plates in the 2D Column are fluorothene because they were in-stalled a year before the new linear polyethylene plates used in theColumn. The fluorothene plates have adequate radiation resistance (10 R)for use in this low-activity field.

F. Solvent Recovery

Two solvent systems, low-activity (No. 2) and high-activity (No. 1) systems,are used in the Purex Plant. The low-activity system simply employs adilute sodium carbonate wash of the solvent in the 20 Column, followed bycentrifugation or decanting for removal of solids and aqueous solution.Such a procedure normally maintains solvent of adequate quality. Sinceapproximately 0.3 percent of the solvent processed in the plant is lostby chemical reaction, entrainment, and volatilization, fresh solvent isoccasionally added to the No. 2 system, and solvent from the No. 2 systemis then added to the No. 1 system for make-up solvent.

With the much more severe conditions of exposure of the No. 1 systemsolvent, a more comprehensive washing procedure is used as shown in Figure3. The washing consists of good contacting of the solvent with analkaline-permanganate solution first, where the permanganate is reducedto manganese dioxide, a good scavenger for diluent degradation productsand particulates. The solvent is next washed with dilute acid for removalof residual manganese dioxide and finally washed with dilute sodium car-bonate alone..

The contactors for all three operations in the plant are different becauseof the presence of original equipment in the plant. However, all threecontacts could be made in identical equipment achieving adequate contact-ing and separation of 15 to 30 minutes. No centrifugation is requiredwith the above process.

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Chemical solutions used for solvent washing are combined with the neutral-ized IWW from solvent extraction processing, concentrated, and stored inlarge underground mild-steel tanks. Infrequently, gross amounts ofsolvent nave been discarded to separate storage areas because of contam-ination with grease or paint resulting from leaks out of process equip-ment. No solvent which has remained in routine process use has requireddiscarding.

G. Nitric Acid Recovery and Waste Disposal

As shown in Figure 3, the only acid waste from solvent extration is HAW.Nitric acid is recovered from this waste by a double-distillation, ab-sorption, and fractionation according to the flowsheet specified. No steam-stripping of this acid is performed; the entrained solvent is decanted andreturned to process, and the dissolved solvent is distilled or hydrolyzed.The acid product recovered from these operations has been decontaminatedfrom fission products by a factor of lx10 and is re-used in the processwherever possible without introducing excessive mission products. Noproblem of ruthenium volatilization has resulted from the acid recoveryprocess.

The IWW waste solution containing essentially all of the fission productsis neutralized with sodium hydroxide and stored in large, underground mild-steel (SAE-1020) tanks under boiling conditions as described in Reference15. The concentration of these wastes must be maintained below 8 M sodiumion to prevent precipitation of sodium salts. The precipitates (iron saltsprimarily) in the tank scavenge most of the Zr-Nb, Ce, and Sr fission pro-ducts leaving Cs and Ru in the supernate. Because of this tendency forconcentration of the high heat-generating Zr-Nb and Ce fission products inthe solids, large air-lift recirculation tubes are used to turn the solu-tion in the tanks over at rates of several thousand gallons per minute toalleviate a tendency for localized over-heating, which would result inuneven boiling. The vapors from the tank are condensed, and most of thecondensate is returned to the tank. The remainder of ths condensate isreturned to the plant and used for acid distillation water prior to dis-posal to the ground via caverns k15).

Use of the formaldehyde process in the Purex Plant is contemplated fordenitration of nitric acid in IWW (5, 14). This process has many economicadvantages in addition to its main justification of reducing the solidscontent (six-fold) and volume (three-fold) of high-level wastes forultimate storage.

IV. SPECIAL FEATURES

A. Process Control

Many of the process control techniques and devices used in the plant arestandard for chemical plants. However, special equipment is used for

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sampling the radioactive solutions, and special handling is required foranalyses. Several in-line analytical instruments (18) are used for con-tinuous measurement of pH, plutonium and uranium concentrations, andgamma activities of solutions. These instruments provide much assist-ance for detecting off-standard conditions at an early time in theirdevelopment and for aiding in the understanding of the process duringnormal and test operations.

Continuous flow streams are generated either by pumps (4) or gravity-flow,and flow measurement and control (4) are achieved by remotely-operatedrotameters and diaphragm-operated valves, respectively.

The evaluation of the allowable departures from flowsheet and the sensi-tiveness of process variables is largely dependent upon which type of flow-sheet is being evaluated. For example, a flowsheet employing extensivewaste backcycle can be extremely insensitive to deviations from flowsheetas compared to a process involving no backcycle. Sensitiveness of processvariables is also dependent upon specific details, such as: 1) interfaceposition, 2) cartridge design, 3) critical mass considerations, 4) pro-duct specifications, 5) how close the plant is being pushed to capacity,6) frequency of analytical checks for off-standard conditions, 7) degreeof technical process supervision, 8) the existence of secondary indi-cations of off-standard conditions, and 9) rate of process response orturnover.

In general, instrumentation ani analytical techniques should be capable oftwo percent precision. An equally nigh degree of accuracy is not generallynecessary, except in isolated cases such as nuclear materials accountability.Operating and process control personnel rapidly develop a "feel" for thevarious biases in tne data.

HAPO experience has shown tnat 5 percent variation in composition or flowrate of most streams can be tolerated. However, certain key streams shouldbe controlled more closely. For example, in the uranium extraction columns,close control of solvent uranium saturation is required to optimize decon-tamination without incurring excessive plutonium losses. Temperaturesshould be controlled with a precision of 2*C as wider fluctuations canresult in column upsets due to changes in phase behavior.

A thorough evaluation of allowable flowsheet departures cannot be madesufficiently general to be realistic, and such an evaluation should bemade specifically for the proposed flowsheet with consideration given tothe points outlined above.

B. Waste Rework

Off-standard aqueous streams are reworked occasionally to recover uraniumor plutonium. Such streams usually consist of concentrated acid wastes,solvent recovery wastes, cell drainage, flushes and spill solutions. Theprocedure for handling all aqueous streams is generally the same. The

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rework stream is acidified, if necessary, to 4 - 7 M nitric acid and thenboiled or ref luxed until all TBP and its degradation products are hydro-lyzed to phosphoric acid and the diluent boiled off. The time requiredto destroy TBP depends on temperature, the concentration of nitric acid,salts present, and on the TBP concentration. Unfortunately, due todifficulties involved in measuring TBP concentrations in waste streams,no fixed formula for reflux time has been obtained. Rather, a trialmethod is used to determine completion of the hydrolysis. A small portionof the rework solution is blended into the First Cycle feed at volume ratiosof about 1 to 20, and the recovery efficiency, decontamination performance,and stability of the first extraction column is observed. If the batch isprocessed satisfactorily, the remainder of the rework is blended off in thesame manner. However, if the trial is not successful, the rework solutionis refluxed for a longer period. To date, concentrated acid wastes whichcontain a minimum of entrained TBP and which have boiled more than 50 hourshave been processed successfully by blending into the feed for First Cyclefeed make-up solution. Other aqueous rework solutions mentioned abovehave required 12 to 22 days of refluxing before they could be processed.In refluxing rework material, as in concentrating any Purex solution, caremust be taken not to exceed 135*C (the de-nitration temperature of "Red Oil").

Recently processes have been developed to rework aqueous waste solutions,particularly concentrated acid waste, through anion exchange resins for therecovery of plutonium. Several problems remain to be resolved in this pro-cess, one of which is solids removal. The composition of concentratedacid waste is being studied, and the ion exchange method of recoveryappears promising. One item of interest learned from plant experience isthat a solid iron salt of dibutylphosphate is formed in concentrated acidwastes which has a tendency to coagulate and plug lines. This salt isslurryable in dilute (5 weight percent) sodium hydroxide.

The processing of off-standard organic solution, notably large volumes ofspilled organic which have dissolved paint and grease from cell floors,has not been always successful. Amercoat paint used at Hanford acts asan emulsifying agent and upsets the extraction columns so that blendingis not possible.. If the organic contains recoverable product and thevolume is such that it cannot be conveniently distilled and hydrolyzed,then batch contacting is used. Usually carbonate or caustic solutionswill adequately strip the product from the solvent so that the strippedsolvent can be discarded. The stripped solution is then treated as off-standard aqueous rework solution.

V. REFERENCES

1. Brunstad, A., "Polymerization and Precipitation of Plutonium IV in

Nitric Acid", HW-54203 (Unclassified), December 17, 1957.

2. Cooper, V.R., and M. T. Walling, Jr., "Aqueous Processes for Separa-tion and Decontamination of Irradiated Fuels", Proceedings of the

Second International Conference on the Peaceful Uses of Atomic Energy,United Nations, Paper 2409 (1958).

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3. Courtney, J. J., and B. E. Clark, Jr., "An Introduction to the PurexPlant", HW-32413 (Secret), July 15, 1954.

4 . Dunn, J., and H. M Jones, "Flow Generation, Measurement,and Control",Symposium on the Reprocessing of Irradiated Fuels, Brussels, Belgium,TID-7534, Book 3, pp. 569 -91, (1957).

5. Evans, T. F., "The Pilot Plant Denitration of Purex IWW with Formalde-hyde", HW-58587, (Unclassified), February 23, 1959.

6. Geier, R. G., "Application of the Pulse Column to the Purex Process",Symposium on the Reprocessing of Irradiated Fuels, Brussels, Belgium,TID-753 , Book 1, pp. 107 - 129,(1957).

7. Hahn, K. J.,and H. 4. Jones, "A Remotable Float-Type Liquid InterfaceController", HW-55166 (Unclassified), February 27, 1958.

6. Harty, W. M., "The Design Philosophy of Remote Operation and Maintenanceof Separations Facilities", Chemical Engineering Progress SymposiumSeries, No. 13, Volume 50, pp. 115 - 121, (1954).

9. Higgins, I. R., "Mechanical Features of the Higgins Continuous Ion Ex-change Column", ORNL-1907 (Unclassified), September 29, 1955.

10. Higgins, I. R., "Countercurrent Liquid-Solid Mass Transfer Method andApparatus", U.S. Patent No. 2,815,322, December 3, 1957.

11. Irish, E. R., and W. H. Reas, "The Purex Process - A Solvent ExtractionReprocessing Method for Irradiated Uranium", Symposium on the Reprocess-ing of Irradiated Fuels, Brussels, Belgium, TID-7534, Book 1, pp. 83 -106, (1957).

12. Irish, E. R., "Process Specifications for Operational Control - PurexPlant (Revision No. 1)", HW-56933 (Secret), November 17, 1958.

13. Irish, E. R., "Purex Plant Production Capability Derived from UnclassifiedInformation - Pilot Evaluation of Final Uranium Cycle", HW-59837 (Secret),March 30, 1959.

14. Oberg, G. C., "Denitration of Purex Plant IWW", HW-60161 (Secret), May 1,1959.

15. Platt, A. M., "The Retention of High Level Radioactive Wastes", S osiumon the Reprocessing of Irradiated Fuels, Brussels, Belgium, TID-7534,Book 1, pp. 389 - 406, (1957).

16. Ryan, J. L., and E. J. Wheelwright, "The Recovery, Purification, andConcentration of Plutonium by Anion Exchange in Nitric Acid", HW-55893(Confidential), January 2, 1959.

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17. Schmidt, W. C., "Treatment of Gaseous Effluents", Symposium on theReprocessing of Irradiated Fuels, Brussels, Belgium, TID-753 , Book 1,pp. 362 - 377, (1957).

18. "Operation and Maintenance Manual for In-Line Analytical Instruments",HW-41093 (Unclassified), February 1, 1956.

19. "Purex Technical Manual", HW-31000 (Secret), March 25, 1955.

TABLE 1

SOLVENT EXTRACTION COLUMN I TA

Plate or PackedSection Cartridge Details

Column Height, I .D. Plate Plate Holeft. in. Type Mat'1. Size

in.Dia .

FreeArea

PlateSpacing

in.

LouverPlates

No. FreeArea

PulserAmpL., req.in. Range

c/p

Ptia seContin- Pulsed

uous

HA Ex. 13.9 24Scrub 19.2 32

2D Ex. 14.0 24Scrub 13.2 32

HS 18.6 34 Sieve

NozzleSieve

ss 3/16(a) (a)

Nozzle ss 3/16Sieve (b) (b)

(a) (a)

23(a)

23(b)

(a)

21

21

1

44

43

23i6

2316

1.10.6

1.10.6

35-110

35-110

4 16 0.53 35-110

Org . Org .

Org. Org .

Org. Aq.

18.0 34 Nozzle as 1/8 10 4 -- -- 0.53 35-110 Aq-

28.0 27 Sieve as 3/16 33 4 6 16.5 0.84 35-110 Aq.

13.3 8 Sieve ss 1/8 23 2 6 20 0.84 25-110 Aq.

Aq.

Org .

Aq.

2A Ex. 20.9Scrub 9.8

7 (c) (c) -- (c) (c) -- -- 1.1 25-110 Org. Aq.

7 Sieve ss 1/8 23 2

34 Sieve Fluoro-thene

3/16

-- -- 1.1 25-110 Aq.

23 4 0.53 35-110

Aq .

Org. Aq.

(a) Cartridge composed of 21% free area stainless steel (ss) plates and 23% free area linear polyethylene plates,grouped alternately four stainless steel plates and two polyethylene plates. Hole diameters in stainlesssteel plates = 0.085 in.; polyethylene plates = 0.1875 in.

(b) Cartridge composed of alternate pairs 21% free area stainless steel plates and 23% fluorothene plates. Holediameters in stainless steel plates = 0.085 in.; fluorothene plates = 0.1875 ir.

(c) Packed with 1 in. fluorothene Raschig rings.

r~

IC, 2E

IBX

IBS

2B

10, 20

21.0

26.3

13

To Stack

xc backupFa eltty Aed

XDAbsorberXD H2O -Water

H O HN03 2.684-Standard Hanford !!L32 - x1403c2.H26Uranium Slugs Flow Flow 493 Cake Wa Slurry Wath HNO

Flow Variance Sp .ter, 1.08 apGr. 1.00 C haer HSp.Gr. 1.04 H2

0 2Otbaer HOo

Reaction Gaaa 40HN3 0.3 Flow 0.15 3H03

k lSp. Gr. 1.00 FoNaNOj3Addn. NH I.%4 Flow 0.31 Spr. .0

H2O M H2O 05 - TeSp.Gr. 1.01 Sp.Gr.NaNO3 3.63 Air 961rr i

Flow 11,3(aReacton Water Reaction Gas Cake Wash

H2O 202 Flow (does H2O Kof HN43 as NO HN3 03Flow 3 and NO2

NaOH Addn. S-,.r. 1.01' Temnp. 190-21dt Flow 0.19H2O 5S"Gr. 1.01

T BOHe1

Flow 4. Disolver Centrifuge CentrfugateSp.GOr. 1. S3 Dissolver Dissolution Feed M

(Coati-gAcid Dissolution) H2O l U H 196(CsagCoatng Waate H20 M (-i Sop 2) mmom UNH Z. S Centrifuge HN3 0. 308

JrktRas S ep R--) H2O -! HNO3 10.4 Temp. Boiling HNO3 0.4Jackt Rnse1St2 1)3 11 31100C Flow 69

H2O 0 Temp. Boiling Na0H 1 7 Flow 47.7 Flow 54 Sp.r. 1.63Flow 6.8s NaNO3 1 .0 S . Gr, 1.31 Sp. Gr. 1. 75(hot) Sp.r. 1.82Sp Gr. 1.00 N.NO 1.0- Cake SlurryIn3 nreensNa25iO3 0.03

Flow 42.9 H2O - To UndergroundSp. G r . . 22 UNH 0. 02 Stor ageJacket Rinse Dilution Water HNO3 0. 13

- gT Udrrud H2O Flow 0.27FlwLaaeFlow 6. 85 Stor age Flow 8. 9 .G. .0

Sp.Gr. 1.00 sp.Gr. 1.01'

H W..60116

r Add. Water Addn. AdO (

10.4 Flow Variable Na OZ 6.70

Or-r. 1. 02, 7 Sxr . Flew 0.961.31 Sp.Or. 1.03

HA Fake. m

HNO3 06 Column

Flow 75

Sp.GOr. 1.60

HSS

HNOj 2.0

Flow 59Sp.Gr. 1.06Temp, 35 C

KA P

UN 0. 59IHN 0 0. 176

Flaw 415S. Gr. "a. 97S

i6

IBx

HO mHdo

3 0.Z7

Fe(NHj.03)2 0.03

Flow .20

5.Gr. 1.02

HiP - IBXF

U'N 0. 35vpu lH14O3 0. 12%

Flowy 413Sp.l;r. 0.07Temp. 45C

If

0

To IC Column

IBU M IBSFUN 0.346 Hz 'AHN03 0.058 UNH 0.10

HND3 3.39Flow 411 NPg/Sp.Gr. 0.968 FelNH2

503)2 0.023

Flow 25.7p.Gr. 1.13

Recovered8NoD Addn.

H2OHN3 10.4

Flow 4.61S". Gr. 1.31

Ir

:z

0

mu

H .R I, Sr

H120 ]H2O h

UNH 0.c73 UNH 0.12HNO5 2.53 HN0 3 1. S Soloent {('

Flow 60 N Z Flow 24N .G. I. Fe(8

2 , Spr. 0.85

roen . 605C O3z 0.02 'lemp. 400

CFlow 2l.1

From No.Ib Solvent Header

n r.

lBsU mUN 0.103HNO3 0.543Pu gl

Flow 24.7Sp.Gr. 0.90

NaNO2 Addo.

H2O NNaNOz 0.78

Flow 2.2Sp.Gr. 1.03

2*1

H2OPu2/

HNO3 2.70Na 0.063 ToZA

Fe 0.022 ColumnH2504 0. 044

Flow 27.3S .Gr . 1.01i

em

.55DC

Gr:

ISP

H2O -

HNO3

2.93

Pu R11Fe(NH2

SO3)2 0.024

-Gr.l i

NOT ES:

1. All flows are relative to 100 flowns equals747 gallons /ton uranium.

2. All temperatures are ambient except asotherwise noted.

3. Composition of dissolver reaction gas hasbeen averaged. Composition of gas will vary

with time and rr.ode'of dissolver operation.

4. Specific gravities are given for all solutionsat 25

0C. regardless of operating temperature.

S. Solvent refers to 30 t I vol. per cent TBP ina hydrocarbon diluent of spec fic gravityequal to 0.801.

6. Plutonium concentrations are based on uraniumirradiated to g Pu/ton U.

7. Process control allows t 5% deviation on allspecifications except TOP (See note 5).

8. NaNO2 addition made only when plant Pu

product is recycled.

FIGURE IPUREX TWO-CYCLE

FLOWSHEET

IODINE BACKUPFEED PREPARATIONFIRST DECONTAMINATIONAND PARTITION CYCLE

J.

03

V~

Q

HOO HAOSolvent (S) M

UN

0.06

FoevenOHeadeSp-Gr. 1. Pu gl

Te p. 40C Flow Z9.7H2OUNH 1.2S

NN2dd- HNO3 6.36NaO2". 0pGr .06

FlFlow ZSeTep. 50cp.Gr.a 7 a

H2O -HNO N.N.et i"2S04 0.02

HAX ha 0.1:1aalrent (6'1Fr .01

Flour 3r 1 19p. Gr. 0. , .. Jr. 1, 0,

Ten : . 40 L G===

t o As d RecuVenFrain. No. 1 Waste Co-, ent r atan

lvnrt Neater

AT 281

I mww

i

14H W-60116

IDS

Flow 33

U03 Recovered Sp. Gr. 1.0Acid Temp. 350C

620HN03 10.4UNH 0.05

Flow 8. 5Sp.Gr. 1.31 ZDIS

HNOI 2. so

HNO 0.02

Flow 39/Sp.Gr. 1.00Te 'per0%

FromtRx Column

LUMUN .i.34614N03 0.058 m

F Gr. 411S r. 0.968

i-i

Lf

x

1420HN03 0.01 5

Flow 317Sp.Gr. 1.0Temp. 5

0C

VDUM -

-UN o. TsHN03 0.009Flow 347

5n.O. 0.96

3o Baco cycle

Cony enration

I2DWH2OHNO3 0.52UNH 0.051Fe 0.008

Flow 138Sp G. 1.04

0W.

Jt

Solvent (5)

Flow 334Sp.tGr. 0.85

To ZOTTank

To Cavern via

CaCO3 Red

2UD n Jet UTransfer

040 ~zO *60toOU ProductHZQ 1i[ . - Storage Tank

HN03 Trace UNH 2.19

Flow 268 HNO3 05.117Sp.Gr. 1.0 Flow 61.7

Sp. Gr. 1.70

SEU

UNH 0. 408

HNQ3 0.022

Sp Gr. .3 2EU Concentrator

HN 6 03 |

Flow 3.10Sp. Gr. 1.02

FromIBS Column

ZAF

H2OPu gil13 2.70

Na0

0.063Fe+++ 0.022421504 0.044

Flow 27 . 3

'a2

HZO

0HN3 0. 10H2SO4 0.045

Flow 2. 52

Sp.,Gr. 1.00

ZAP

Pu g/lHN3 0.136

Flor. ?SS Gc 086

Tept . 55oG

2^x H2O i

Solvent (4 HNQ3 t. 48

Floc 7.5 ++ .2

Terrip. 350 C H2S04 0.04Sp. Gr . 0. as Flow 31.

5 G.Fr o m

N o . 2 T o B a .r c y le

Organic Systern Concentration

AT 283

5~rf

28W

Solvent (4

Flow 7. 5Sp. Gr. 0.85

To 20FTank

XAS

HNOj 6. 0

SGr. tlTomp. 60

0C

XAF - HNO3

H2O MXDHNO3 1.0 3.0

Flow Z. 28 -Sp. Gr. 1. 37 XAF

1H20 -HNO3 -. 50

H2204 0.02

Flow 5.715p. Gr. 1. 18Temp. 6SoC

_2BP_ XAWH2O H - {Z MPu all HNQ3 C-91

HZSO4 0.044 Pua lFlow Z.57 Flow 5.5

5r,. Gr. 1.03 . Gr . 1. 1 5K1

XCW

XAx

Re! n (6

Flows:

Resir 0.41t51w 0.83l

ReniMrisin

HNO3 1.-0 Flow!.

Flows:.7Resin 0.76in .1

Slipp 0.83

Sp. %r.1.o3n

Te mp. 50 C

XPD X PC

H2OTo XAF ''--"0" S453 0.f411

.f.E Flow 0.772Resin 6_Flory 0.08

Flows: SResin O. 415

Slip 0.83

Product

x~yXCP-

H2O H2p N3 ^ .4H43 ^ ep oln

S ,Gr. 1, .I2Q, gis,

To tWB lank

NOTES:

1. All flows are relative to 100 flows equals747 gallons/toe uranium.

2. All temperatures are ambient except asotherwise noted.

3. Specific gravities are given for all solutionsat 25

0C, regardless of operating temperature.

4. Solvent refers to 30 ; I vol. per cent TSP ina hydrocarbon diluent of specific gravity

equal to 0.801.

5. Plutonium concentrations are based on uraniumirradiated to g Pu/ton U.

6. Anion exchange resin is Permutit SK, 20 - 40mesh. Flow volumes of resin are for net settledresin in the nitrate form.

7. Process control allows 5% deviation on allspec ifications except TSP (See note 4).

FIGURE 2

PUREX TWO-CYCLEFLOWSHE ET

FINAL URANIUM CYCLEFINAL PLUTONIUM CYCLE

PLUTONIUM ION EXCHANGE

IC,Solvent (4)

Flow 058Sp.Gr. 0.8%

To 10F

To Cavern via CaCO3 Bed Flow 11.4

10 Fe(SA)Addn. eSp.3Gr 1.09!V!] al dd-Temp. 35 C

H2O M H2O M

HN03 ca. 0.01 Fa{NH2Flow 32$02DF5. Gr . 1. 0 Flaw 0-.55SPrSp. Gr. .34 H2

UNH 1.6HNO3 0.35To (NH2

SO3)2 0.012

From Na. 2 Flow 88.8

Organic Sytem mp. 300C

IC U

UH14 0.346 bvlo, vi 4)iN03 0.076 U101H 1.61

S Flow 432 HN03 0.353 Flow 304 ...--.

Sp.Gr. 1.11 ICU Concetrator Sp.Gr. .0.8

Sp.Or. 1.53

Lt

xcx

15

Water Addn.

H2OFlow 241

Sp.Gr. 1.0 H A2 Backup Facility AFRH2O0 Z

2WF iWD - AAF Flow 109 H203 . Flow 34HI Sp.Gr. 1.0 HNO3 2.6_ Sp.Gr. 1.0

HO -H0 - Reflux Flow 53 Reflue1130.4 N0

3 0.84 Ratio 0.25 $p.Gr. 1.08

Flow 365 Flow 385

Sp.Gr. 1.03 Vapor Phase

HAWDFFH2O

H2O M H2 HN3

HNO 0.1 HN03 4.7 - .. e- '0 UNH 0. DOS Flow

IMF Flow 438 Flow 121pG

H2OWFl-S.GGr.ro 12.4EHNO3Fe.905 .3Na 0.006Fe 0 004 Concentrator Evaporator T o Cavern va To C

CaC03 Bed

Sp. Gr. 1.03 AAA RcvrdAi()FlowW

H39.. 03

Hl0.1Aci HNO3HN36.2aie T UNH 0.05 UNHa

Flow 0 Waste 09p[ Sp.Gr. 1.18r 2FlowH2O -- Sp. Gr. 1.19 Concentration Sp -G . -3 Sp.Gr.HN3 5..6 Dlute H2O H

Si 0.32at N .

Pu&U Trc0.01dd.1ddon F.88Ni 0.005 K 0.17 CondensateAl 0. 1 Mn 0. 17 HO Acid 9Na. 0.5 C .0 Flow Variable Ta

PO 0.05 Neutralized Al 0.09 5p.r. 1.0

S 0.2 wase So4." 0. ;P u U T race N aO H A ddn.

H 2O

P

F5.00

Vent JetCondensates

ZDW 3WF

ZAW H120

11113 1.12

Fe*"4

0.01 To 4Nat 0.01 Cavern viaH2SO4 0.02 CaC0

3 BedUNH 0.042

Flow 168Gr. 1.03

AFD

0.005 33 WF

98 Decanter

1. 0

Organic toUnderground t;

hem. Sewer Storage a

BackcycleConcentrator

10. 40.011AW

s6. a xSV

1.03

coveredstoragenk s

NOT ES:

1. All (lows are relative747 gallons/ton urtani

2. All temperatures arenoted.

3WD

HO M1120 -

HNO3 ca. 0.01

Flow 147Sp.Gr. 1.0

3WW

HO MHNO3 9.0

H1*54 0.165Fe 0.08Na+ 0.06UNH 0.344Nu g/l

Flow 20.6

Sp.Gr. 1.3

awnHOW

HNO3 8.0112504 0.120Fett 0.06

Na 0.06UNHt 0.254

Flow 28.0Sp.Gr. 1.33Temp. 50 C

___F TBiP-Daluea.t 105

Solvent 4) Make-up 110H2___

GWnFlow 198 Flow ... l 1403 0C33 0.24Sp.Gr. 0.85 Flow 23 Fo .5

S p G r . 1 . 0 2G m p .

____ AdCntcoO_9 Solvent 4) Contactor

H100lo 198 Flow 3.98N2c03 0.24 So.Gr. 1.02 Sp.Gr. 0.8%2EW

Temp. 50 C Temp. 40 C

Flow 23F4lZ0KSp.1Gr. 1a03t1NO

H210 To No. I T0a

KNaO4 0.78 0.23 3 023 r0.02

Kn4 00 Fo 23Foeadew H2OHA X, IBS Na 0. 27

Flow ~2000 - IOR W1 IOD K 0 . 022KMnO4 Add". _m - Sp. Gr . 1. 02 " 109! WH20 H2O M N03 0.1I33

H;30 M Temp. 50 C H2O HN03 ^-0. 32 Na 0 .48 C03 0. 069

K MtnO4 0. 76 Na2CO3 0. 23 Mn02 0. 022

KMn04 0. 05 Flow 23 Flow . 5 Flow 55, 5Flow 1.6 Sp.Gr. 1.02 Sj. Gr. 1. 03 Sp. Gr. 1.03Sri. Or. 1.01 Flow 251-11I

Sp..Gr. 1.02

To Undergroundstorage

ToBo.-ZDilvent

MakeC0 02

Floww1.6

200

CentiueSolvent &, etr ue Flow 341

oUdron Sp.Gr. 08520F Al S Temp. 35

Sotlnt _j4)

Flow 341To No. 2 Solvent HSp. Gr. 0.85 2DX. ZAX

20R (9)-Flow 341

SI.Gr. 1.03Temp. 50 C

Hz0 M 20W

Na2C03 0. 24 H20 04

Flow 18 CO3 0O. 24

Sp. Gr. 1.03 Fo 8

ToUnderground Sp. Gr. 1.03

Header

3. Specific gravities are given for all solutions

at 250C, regardless of operating temperature.

4. Solvent refers to 30 1 I vol. per cent TBP in

a hydrocarbon diluent of specific gravityequal to 0.801.

5. Plutonium concentiaations are based on uranium

irradiated to g Pu/ton U.

6. Vent Jet Condensate flow is fixed at ca. S. 0 gpm.

7. Process control allows t 5% deviation on all

specifications except TOP (See note 4).

8. UO3 plant recovered acid added on a weight

basis of 100% HNO3 at a ratio of 2. 5 parts of

UO3 acid to 10 parts of AAA plus back up facility

acid.

9. Solvent wash solutions - 100, IOR, and ZORchanged 1000 gallons at one time.

FIGURE 3

PUREX TWO-CYCLEFLOWS HEET

WASTE CONCENTRATION

ACID RECOVERY

BACKCYCLE CONCENTRATIONORGANIC RECOVERY

AT 284

HW-60116

To HA Column

to 100 flows equals

um.

ambient except as otherwiseV%%.

-