UNCLASSIFIED AEC RESEARCH AND DEVELOPMENT REPORT COPY No.

CPD - HLO

THE PUREX PROCESS -A SOLVENT EXTRACTION REPROCESSING

METHOD FOR IRRADIATED URANIUM

E. R. IRISH and W. H. REAS

APRIL 8, 1957

HANFORD ATOMIC PRODUCTS OPERATION

RICHLAND, WASHINGTON

GENERAL^ELECTRIC

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AEC GE niCHLAND, WASH.

UNCLASSIFIED HW-49483 A. I Chemis t ry-Separa t ion P r o c e s s e s ' _ for Plutonium and Uranium

(TID-4500, 13th Ed. )

THE PUREX PROCESS - A SOLVENT EXTRACTION

REPROCESSING METHOD FOR IRRADIATED URANIUM *

By

E. R. I r i sh

R e s e a r c h and Engineer ing Operation Chemical P r o c e s s i n g Departnaent

and

W. H. Reas

Chemical R e s e a r c h and Development Operation Hanford Labora to r i e s

Apri l 8, 1957

HANFORD ATOMIC PRODUCTS OPERATION RICHLAND, WASHINGTON

Work per formed under Contract No. W-31-109-Eng-52 between the Atomic Energy Commiss ion and Genera l E lec t r i c Company

Pr in ted by/for the U .S . Atomic Energy Commission

Pr in ted in USA. P r i c e 60 cents . Available from the

Office of Technical Serv ices U. S. Depar tment of C o m m e r c e Washington 25, D. C.

• P r e s e n t e d at the Aqueous Reprocess ing Symposium in B r u s s e l s , Belgium, May, 1957.

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Chemis t ry-Separa t ion P r o c e s s e s for Plutonium and Uranium

(TID-4500, 13th Ed. )

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Chemis t ry-Separa t ion P r o c e s s e s for Plutonium and Uranium

(TID-4500, 13th Ed. )

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Chemis t ry-Separa t ion P r o c e s s e s for Plutonium and Uranium

(TID-4500, 13th Ed. )

EXTERNAL DISTRIBUTION (Contd. )

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ABSTRACT

This paper describes the Purex Process , which enaploys solvent extraction to separate and purify uranium and plutonium from each other and from, fission products contained in irradiated uranium fuel elements. A description of the over-all process, utilizing tri-butyl phosphate solvent (in a kerosene-type diluent) and nitric acid salting agent, is provided along with chemical process flowsheets. The process chemistry of uranium, plutonium, and fission products is discussed as affected by process variables. Methods of recovery of spent solvent and acid are briefly dis­cussed. Alternate process arrangements are suggested.

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THE PUREX PROCESS - A SOLVENT EXTRACTION

REPROCESSING METHOD FOR IRRADIATED URANIUM

I. INTRODUCTION

The Purex Process is another continuous solvent extraction process which has been developed and demonstrated through laboratory and pilot plant work to recover and purify uranium and plutonium from irradiated uranium fuel. This process performs the same functions and has the same products as the Redox Process , but it differs in the use of solvent and salt­ing agent. Whereas the Redox Process utilizes hexone, methyl isobutyl ketone, for a solvent and aluminum nitrate for a salting agent. The Purex Process utilizes tri-butyl phosphate as the solvent and nitric acid as the salting agent. Advantages of the Purex Process stem primarily from the following points: (1) the tri-butyl phosphate solvent system is safer as a result of the higher flash point and lower volatility of the solvent, and (2) the nitric acid salting agent is readily recoverable (by distillation), permit­ting much lower requirements for both waste storage and essential materials. Aside from these two differences, the two processes parallel each other considerably.

The tri-butyl phosphate, solvent extraction, fuel-processing system involves a maximum of six major steps:

1. Extraction of uranium and plutonium from aqueous solutions into the organic TBP-diluent phase.

2. Partitioning of the uranium and plutonium.

3. Decontamination and recovery of the uranium. 4. Decontamination and recovery of the plutonium.

5. Solvent recovery. 6. Nitric acid recovery (including waste concentration).

All six of these steps will be briefly discussed in this paper, along

with an over-all chemical flowsheet, one of several which are feasible.

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Preparation of the feed solution to the Purex Process was discussed in the previous paper. Solvent extraction contacters for carrying out the process will be discussed in the subsequent paper. Auxiliary processes (such as head-end treatnaent of feeds and tail-end treatments on product solutions) will be discussed in another session.

n . BASIC PROCESS PRINCIPLES

It is not within the scope of this paper to discuss comprehensively all the process chemistry of the Purex Process, Rather, it is the intent to present somie highlights which will lend understanding of the process. Detailed comprehensive data must be derived from the development l i tera­ture.

A. Solvent Action of TBP

The solvent action of tri-butyl phosphate depends on its complexing action. For example: uranyl nitrate reacts with TBP according to the reactions

U02''"^(aq)+2 NC53(aq)+2 TBP(org)£=; U02(N03)2 . 2 TBP(org).

If we define K,̂ as equal to the equilibrium constant for this reaction, we can write the useful relationship defining the distribution coefficient (i, e, , the ratio of the concentration of uranium in the organic phase to that in the aqueous phase),

E°/^U = K^ (N03"(aq)) ^ ( T B P (org))^

where the (NOo ) concentration is the aqueous phase concentration and the TBP concentration is the uncomplexed TBP concentration in the organic phase. Since uranyl nitrate does not form a perfect solution in water under process conditions, K is not constant for the equilibrium expressed but varies somewhat with the concentration of the various solution components.

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However, inspection of the equation provides a general ly co r r ec t explana­

tion of the na ture and magnitude of the effects of the concentrat ions of uranyl

n i t r a t e , n i t r i c acid, and TBP on the dis tr ibut ion ra t io of uranyl n i t ra te .

A s i m i l a r analys is to that for uranyl n i t ra te may be applied for the solvent

extract ion of plutonium IV. However, the concentrat ion of the plutonium

itself may usually be neglected in the analys is s ince the plutonium concen­

t r a t ion i s a lmos t invar iably low enough for i ts se l f -sa l t ing and solvent-

sa tu ra t ing effects to be insignificant. Thus , for plutonium IV:

E ° / ^ P u = Kp^ (N03"(aqi)^ (TBP(orgj)2,

Again the T B P concentrat ion in the organic phase is the uncomplexed TBP

concentrat ion, and if u ran ium is a l so p resen t , this fact must a lso be taken

into account. These re la t ionships show the effect of the n i t ra te ion concen­

t ra t ion in the aqueous phase on the distr ibution coefficients. This effect is

defined as the "sa l t ing" effect. For tunate ly , the degree of TBP complexing

of fission products is much l e s s than that of uranium and plutonium; thus,

a high degree of separa t ion is poss ib le .

B, Typical Distr ibution Coefficients

In Table I the re la t ive o rde r and magnitude of TBP extract ion of

uranium,plutonium,ni t r ic acid, and the pr inc ipa l t roublesome fission products

contained in "aged" i r r ad ia t ed uran ium a r e indicated:

TABLE I

Distr ibution Coefficients for Uranium, Plutonium and Fiss ion Produc t s for Extract ion f rom Feed , 25 C,

ION E ° / ^ U(VI) 8,1

Pu(IV) 1. 55

Pu(VI) 0, 62

HNO3 0. 07

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TABLE I (Contd. )

ION E ° / ^

Z r 0. 02

Ru 0. 01

Pu m 0, 008

Nb 0. 005

R a r e E a r t h s 0. 002

Aqueous phase before equi l ibrat ion: 3M HNOo

200 g r a m s U per l i t e r t r a c e r level

Organic phase (30% TBP) after equil ibrat ion: 60% saturat ion

It is evident f rom this tabulation that uranium and plutonium IV

a r e ve ry ex t rac tab le and that the fission products (z i rconium, ruthenium,

niobium, and r a r e ea r ths ) and plutonium lU a r e quite inextractable .

This cha rac t e r i s t i c difference between uranium and plutonium IV and the

f ission products p e r m i t s the decontamination of uranium and plutonium.

The ex t r eme difference between the distr ibution coefficients for uranium

and plutonium IV and that of plutonium III is the proper ty which pe rmi t s

separa t ion of uranium and plutonium from each other , and this will be

d i scussed la te r .

Two t e r m s in Table I need explanation. The TBP is diluted with

a hydrocarbon diluent to 30 volume per cent in the Purex P r o c e s s . This

is done in o rde r to give the solvent phase a low enough specific gravity

(i. e. specific gravity of 0. 841 at 25 C, which i n c r e a s e s to 0. 975 when the

T B P is 85 per cent sa tura ted) so that it can flow by gravity up through the

contacting aqueous phase . (In the Halex P r o c e s s , the TBP is diluted with

carbon t e t r ach lo r ide , and the aqueous phase flows upwards through the

organic phase) . The second t e r m needing definition is "sa tura t ion" as

applied to TBP. The T B P is cons idered to be 100 per cent sa tura ted when

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all of it is complexed with uranium; for a 30 volume per cent TBP solution, a concentration of 124 grams uranium per liter represents complete satura­tion.

C. Variation of Distribution Coefficients with Nitric Acid

In Table U. are tabulated some distribution coefficients for uranium, plutoniuna and fission products at varying nitric acid concentrations in order to illustrate the effect of variations in the nitrate salting strength on the distribution coefficients.

HNO3 M

0 . 5 2 3 4 5 6

Variat

U

1.18 2.84 3, 28 4. 28 4.16 3.61

TABLE II

ions in Distr ibution Coefficients as a of Aqueous

Pu IV 0.10 0.46 0.59 1,11 1.55 2.62

HNOo Concentration

Ru

0.0073 0, 0031 0. 0009 0. 0005 0.0002 0. 0002

Z r

0.0007 0. 0031 0.001 0.024

-0.089

Function

Nb Rare Ea r th s 0, 0001 0.0003 0. 0004 0. 0012 0, 0018 0.0032

0, 0004 0.0005 0, 0004 0. 0004 0. 0002 0,0001

Aqueous phase before equilibration: tracer level 200 grams U per liter

Organic phase (30% TBP) after equilibration: 80% saturation

It is evident from the data presented in this table that the uranium and plutonium distribution coefficients increase markedly as the nitrate salting effect increases, and zirconium and niobium follow in a similar manner, still sufficiently lower, however, to permit decontamination of the uranium and plutonium. However, the ruthenium distribution coefficients follow an inverse pattern. Thus, for optimum separation of the troublesome fission products from uranium and plutonium, a nitric acid concentration in the medium range (e. g,, 2 to 3 M HNOo) must be selected.

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D, Variat ion of Distr ibution Coefficients with Saturation

In Table III a r e tabulated some distribution coefficients for

uranium, plutonium, and fission products at varying uranium satura t ions of

the solvent in o rde r to i l lus t r a t e the effect of the availabil i ty of uncomplexed

T B P on the dis tr ibut ion coefficients,

TABLE III

E o / a at 25 C

u 16.7 12.1 7 . 9 5 . 4 3 . 6

P u I V 4 . 0 2 . 3 1.6 L I 0, 79

Ru

0,067 0.028 0.0096 0,0037 0. 0016

Z r

0.041 0.025 0.020 0,012 0.009

R a r e Ea r th s

0, 0096 0, 0048 0, 0021 0, 0011 0. 0004

Uranium Saturation of TBP , P e r Cent

28, 0 45. 6 61, 7 72. 0 82,4

Aqueous phase before equil ibrat ion: 3 MHNOo

t r a c e r level

Organic (30% TBP)- to-aqueous r a t io : 2 to 1

As would be expected, the uran ium distr ibution coefficient dec rease s

a s 100 p e r cent sa tura t ion is approached because of the lower concentration

of uncomplexed TBP. The distr ibution coefficients for the other ions

d e c r e a s e in a s imi l a r manner ,

E, Chemis t ry of Plutonium in Pu rex P r o c e s s

Three chemical reac t ions of plutonium valence adjustment a r e impor ­

tant to pe rmi t quantitative plutonium recove ry in the p r o c e s s and a high

degree of separa t ion of u ran ium and plutonium. The f i rs t react ion r e p r e s e n t s

the reduction of hexavalent plutonium to te t ravalent plutonium, the most

ex t rac tab le valence s t a t e :

P u O ^ ^ +NO^' + 2H +^- I^Pu"*"^ + NO, + H2O.

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The n i t r i t e ion can be added to the solution ei ther as a salt (e, g. , NaNO^)

or as a gas (e. g. , NO^), and the react ion is rapid, being essent ia l ly

complete in a few minutes at 50 C.

The second reac t ion indicating the reduction of plutonium IV to

plutonium III by f e r rous ion is l ikewise a lmos t ins tantaneous:

Pu"*"̂ + Fe"*"^ + (NH-^SOg")^ Pu"*"̂ + Fe"*"^ + (NH^SOg"),

F e r r o u s ion (added as f e r rous sulfamate) i s used as the reductant,

with the sulfamate ion act ing as a n i t r i t e supp re s so r . If the sulfamate ion

were not p r e sen t , low concentra t ions of n i t r i t e ion (always found in n i t ra te

s y s t e m s in the absence of sulfamate or some s imi l a r ion) would init iate

the autocatalyt ic oxidation of the f e r rous ion, thus prevent ing plutonium

reduction, A severa l - fo ld excess of f e r rous sulfamate is used to a s s u r e

a reducing solution. When reoxidation of the plutonium is again des i red ,

the reac t ion indicated by the th i rd equation is init iated, again using the

n i t r i t e ion but this t ime as an oxidant:

2NH2SO3" + 2N02~ 7 2N2 + 2SO^°^ + 2H2O followed by

6Pu"^^ + 2NO2" + 8H''" yr 6 Pu^^ + ^ 2 + 4H2O.

The n i t r i t e ion is an ideal oxidizing agent because of i ts ability

to oxidize plutonium only to the te t ravalent s ta te , the most ext ractable

form for fur ther p rocess ing . You will note that sulfate ion is formed, and

if excess ive amounts of sulfate ion a r e formed, plutonium sulfate is formed

which is inext rac tab le during subsequent solvent extract ion process ing .

Thus , the chemical flowsheet must be defined to minimize formation of

this complex.

One other c h a r a c t e r i s t i c of te t ravalent plutonium should be mentioned.

In low ac id i t i es , dependent upon the plutonium concentrat ion and total n i t ra te

concent ra t ions , an inext rac tab le po lymer ic spec ies of plutonium may form.

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A, Co-Decontamination Cycle Flowsheet, HA Column

In Figure 1 the HA Column accomplishes the primary separation of the highly radioactive feed solution into an organic product stream and a first-cycle aqueous waste stream containing greater than 99, 9 per cent of the fission products. The feed solution (HAF) from the dissolvers and feed preparation equipment is fed continuously to the midpoint of the HA Column and flows downward into the extraction section countercurrent to an upward flow of organic extractant (HAX). Conditions of flow and salting strength are regulated in such a manner that the uranium VI and the plutonium IV are coextracted almost quantitatively into the organic phase, leaving most of the fission products and other impurities in the aqueous phase. Some fission products are extracted, however, by the solvent at the feed point and in the lower portion of the column, but these are partially backwashed or "scrubbed' into an aqueous nitric acid scrub s tream (HAS) flowing countercurrent to the product-bearing solvent s tream in the upper (scrub) section of the compound column. The bulk of the impurities and nitric acid leave the bottom of the column in the aqueous effluent s tream (HAW) which is sent to acid recovery. The acid recovery flowsheet will be discussed later. The product bearing stream (HAP) is cascaded from the top of the HA Column to the bottom of the HC Column.

HC Column

The HAP enters the bottom of the HC Column and is contacted by a countercurrent flow of aqueous strip solution (HCX). The uranium and plutonium are stripped back into the aqueous phase, A trace of nitric acid is added to the HCX stream to decrease the susceptibility of the HC system to emulsification. The product-free organic (HCW) is sent to solvent treatment (also to be described later), and the aqueous-effluent (HCP) containing the uranium and plutonium is steam-stripped of residual dissolved and entrained organic phase, then concentrated to meet Partition Cycle feed specifications. After one complete cycle the products are generally decon-

3 4 taminated from the gross fission products by a factor of 10 to 10 , and

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additional decontamination may be needed to meet the product-purity

requirements. Product recoveries may approximate 99. 9 per cent,

B. Partition Cycle (Figure 2) lA Column

The product-bearing s t ream overflows from the Co-Decontamination Cycle concentrator into the lA Column feed preparation section where any plutonium (VI) formed within the concentrator is stabilized as plutonium (IV) by reduction with sodium nitrite. Ruthenium and zirconium-niobium are the principal remaining fission products in the lA Column feed (1 AF), The 1 AF is pumped continuously to the mid-point of the lA Column. The lA Column is operated as a decontamination column in a manner analogous to that described previously for the HA Column. The organic-product stream (lAP) overflows the lA Column and is combined with the organic effluent of the IB Scrub Column (to be described later) to form IBXF.

IBX Column

The combined organic effluent of the lA and IBS Columns (IBXF) is pumped to the bottom of the IB Extraction (IBX) Column, As the organic r i ses through the IBX Column, it is contacted with a countercurrent flow of slightly acidified ferrous sulfamate solution (IBX), Ferrous iron in the aqueous phase reduces the plutonium from valence IV to valence III, which is only weakly complexed by the solvent, thus permitting the plutonium (III) to be extracted into the aqueous phase. Some uranium also tends to strip out of the solvent into the aqueous phase, but the majority is held in the organic by the high salting strength of nitric acid refluxed within the column. The aqueous effluent from the IBX Column (IBXP) containing the plutonium and approximately one per cent of the uranium is fed to the top of the IBS Column, The organic effluent (IBU containing the uranium cascades to the IC Column).

IBS Column

The uranium present in the IBXP is stripped by contacting with fresh 30 per cent TBP (IBS) in the IBS Column. The uranium-bearing organic

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stream (IBSU) overflows the IBS Column and is routed to combine with the lAP to form the IBXF stream, previously described. The plutonium bearing s t ream (IBP) leaves the bottona of the IBS Column and passes to the Final Plutonium Cycle,

After two complete cycles the plutonium is decontaminated from 5

gross fission products by a factor up to 5 x 10 , and additional decontamina­tion may be generally required to meet product-purity requirements. Partition Cycle plutonium decontamination factors are typically 20 to 50. Product recoveries in the range of 99, 9 per cent may be accomplished,

IC Column

The uranium bearing s t ream (IBU) overflows the IBX Column and flows into the bottom of the IC Column. The uranium is transferred into the aqueous phase in the IC Column by countercurrent extraction in a manner comparable to that of the HC Column. The product-free solvent (ICW) is sent to solvent recovery. The aqueous effluent (ICU) is s team-stripped of residual organic phase, and concentrated in the ICU concentrator to meet Final Uranium Cycle feed specifications.

After two complete cycles the uranium is decontaminated from gross fission products by a factor up to 1 x 10 , and additional decontamination may be required to meet product purity requirements. Partition Cycle uranium decontamination factors are typically 50 to 100. Uranium recoveries may approximate 99. 9 per cent.

C. Final Uraniuna Cycle (Figure 3) Feed Preparation

The Final Uranium Cycle completes the removal of fission products and plutonium from the uranium to permit direct handling of the uranium product. Of the fission products originally present, on the order of 0. 01 per cent remain in the feed, along with a small amount of plutonium. Zirconium, niobium, and ruthenium remain as the principal fission product contaminants, although the ratio of ruthenium to zirconium-niobium will vary considerably in different flowsheet variations.

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of hydrolysis products of TBP, namely dibutyl phosphate (DBP), monobutyl phosphate (MBP), and phosphoric acid, primarily DBP, Both the DBP and MBP have a marked effect on plutonium behavior in the Purex Process, the DBP for its strong complexing of plutonium IV and the MBP for its tendency to form a precipitate with plutonium IV. Both DBP and MBP form weak complexes with plutonium III and neither affects the plutonium HI distribution ratio appreciably. However, owing to the complexing action of DBP on plutonium IV, plutonium IV losses during stripping become excessive unless the DBP concentration of the solvent is kept below 0. 001 per cent. Similar behavior exists with uranium also. Because both MBP and DBP are acidic in nature, they are readily removed from the used solvent with the dilute sodium carbonate wash. Dilute caustic is also satisfactory for DBP and MBP removal, but it will precipitate uranium and plutonium whereas sodiuna carbonate forms soluble complexes of these metallic ions.

The carbonate wash also is used to remove residual uranium, plutonium, cind fission products (primarily ruthenium, zirconium, and niobium) from the solvent before recycle. With this simple washing procedure, solvent quality is maintained at a satisfactory quality for continued recycle.

Process degradation of the diluent, a kerosene-type hydrocarbon, under normal conditions is not a serious problem. However, the hydro­carbon selected should be one of high purity, low in unsaturated compounds and preferably non-cyclic in nature for maximum stability to radiation and chemical degradation. A diluent having a high flash point is desirable to minimize safety hazards. The sodium carbonate wash has essentially no beneficial effect on diluent quality and continued development work is in progress to determine the most satisfactory commercial diluents and to devise chemical means for overcoming process degradation of diluents.

F. Nitric Acid Recovery (Figure 6)

The nitric acid recovery flowsheet illustrated (Figure 9) consists of a two-stage distillation and absorption and fractionation. The economic incentive for acid recovery from extraction wastes is based primarily on

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savings of waste storage space, with relatively minor savings resulting from the decrease in nitric acid and sodium hydroxide (for neutralization) consumption. In properly sized equipment, the degree of acid recovery is limited only by the maximum allowable concentration factor which is per­mitted before the formation of excessive solids (from sodium nitrite and ferrous sulfamate). You will note in the flowsheet that the first-cycle waste acid (HAW) is recovered by a double distillation (to permit adequate decon­tamination of the recovered acid) whereas the other acid waste is only dis­tilled once prior to absorption and fractionation. The degree of acid recovery is controlled by adjusting the feed and concentrate acidities so that the feed acidity is approximately equal to the vapor concentration in equilibrium with the concentrate. In this manner the majority (85-95 per cent) of the acid is distilled and then can be fractionated to 50 to 60 weight per cent nitric acid by conventional means. The high-activity acid concentrate (6 to 10 M HNOo) is neutralized and sent to underground storage where further self-concentration (from the heat of fission product decay) can be achieved. This concept will be discussed in more detail in a later session.

Decontamination factors across a double-distillation system (gross R 7

gamma DF on the order of 1 x 10 to 10 ) are sufficient to allow use of the recovered nitric acid in all process s t reams except the final-cycle scrub s t reams. Ruthenium volatilization may become excessive at high (8 to 10 M) nitric acid concentrations, but this volatilization may be suppressed by the presence of low concentrations (ca 0. 03 M) of nitrite ion, either added directly or formed by the reaction of entrained diluent with nitric acid.

An alternate to the flowsheet shown would be to route the 2WW acid concentrate to the HAF rather than to the IWF for disposal. By utilizing this alternate, the small quantities of uranium and plutonium contained therein can be recovered instead of discarded. If this alternate is used, TBP must be prevented from entering the waste system, since excessive hydrolysis products of TBP (DBP and MBP) could be recycled and result in high plutonium and uranium losses from the HC Column in addition to lowering the decon­tamination factor obtained in the HA Column,

UNCLASSIFIED

UNCLASSIFIED -22- HW-49483 A

A second reason for preventing the accumulation of TBP in the acid concentrator is related to chemical safety. TBP will form complexes with nitric acid (and also uranyl nitrate) which, if concentrated to a high degree, can resul t in an exothermic decomposition reaction of explosive violence. Purex Process conditions do not approach those required for such reaction. However, for process safety the product and acid concentrators should be protected to prevent such conditions from arising accidentally. Protection is readily possible since the reactions will not take place at atmospheric p re s su re until the mixture reaches 135 C for uranyl ni t ra te-TBP complexes and 150 C for nitric acid-TBP complexes. Limitingthe maximum temperature of the sytem provides an adequate safeguard.

IV. ALTERNATE PROCESS ARRANGEMENTS

As mentioned ear l ier the flowsheet and flow patterns just discussed (Alternate A in Table IV) are only one variation of several satisfactory arrangements of the Purex Process . Other satisfactory chemical flow­sheets with alternate A (i. e. , three-cycle decontamination) can also be demonstrated, utilizing variations in the acidities of different feed solutions, different flow rat ios , concentrations of TBP, etc. However, even more versati l i ty of the Purex Process can be achieved by utilizing various head-en

TABLE IV

Alternate Purex Process Arrangements

Head-End Solvent Extraction Treatments Cycles

U Pu

A B X

C D E X

3 2 2 2 1

3 2 2 2 1

Tail-End Treatments

U Pu

X X

X X

X X

UNCLASSIFIED

I 1 > I I

^ o

WASTE

ACID RECOVERY

NEUTRALIZED WASTE

FIGURE 6

UNCLASSIFIED 31

CO DECONTAMLNATION

H A S

H 2 O

H N O 3

S p . G r

F l o w

M

2 . 0

1 . 0 7

5 6 . 2

H A F

H 2 O U

P u

H N O 3

S p G r

T e m p F l o w

M

I 8 0

0 9 6

1 6 1

5 0 ' ' C 7 5

H C X

H 2 O H N O 3

S p . G r . T e m p .

F l o w

M

0 0 1

1 0

3 0 ° C 4 8 8

HAP M

J 0 352

HNO3 0 2 TBP Di)

Sp Gr

Flow

0 96 383

T B P H N O 3

Sp G F l o w

%U % P u

H C W

Dil

T

r 0

U

race

8 4 3 5 5

< 0

'•Q

2 J H P D

H ^ O H N O 3

S p G r

F l o *

c a 0 0 1

1 0

4 2 3 4

HCP CONCENTRATOR

HCP M H2O

UNH 0 27

Pu

HNO3 0 15

Flow 497

N a N O ,

H 2 O N a N 0 2

S p G r F l o w

A d d

i

1

1

M

6

07

2 5

H F C

U N H

P u

H N O 3 F l o w

M

1 8 3

0 97

73 7

I A S H ^ O

H N O 3

Sp O r

F l o w

l A F

H 2 O U

P u H N O 3

N a N O ^

Sp G r

T e m p F l o w

l A X

T B P D i l

S p G r T e m p F l o w

M

2 0

1 07 56 2

U

1 8 0

0 9 5 0 0 3

1 6 1

5 0 ° C 7 5

M

0 8 4 6

4 5 0 c 3 5 5

•

I

A

C 0

L

U W N

To Acid Recovery

2 E X

HS^OJ

S p G r

T e m p

F l o w

M

0 0 1

1 0

3 0 ° C

4 3 9

2 D U M

U N 0 36 H N O 3 0 01 T B P Di l S p G r 0 9 7 F l o w 3 7 3

2 D W M

H N O 3 0 76

. e ( N H 2 S 0 3 ) 2

0 0 0 4

Sp G r 1 0 2

F l o w 150

2 E W M

T B P Di l

H N Q , T r a c e

S p G r 0 8 5

F l o w 3 4 5

1 l o S o l v e n t T r e a t m e n t

2 U D M H O

H R O 3 c a 0 F l o w 3 7 8

_

01

l A P M

U N 0 351

P u H N O 3 0 16

T B P Di l

S p G r 0 9 5

F l o w 3 8 3

l A W

H N O 3

N a N O , N a N 0 3

S p G r F l o w

M

0 9 3

0 0 1 7

T r a c e

1 0 3

120

2EU CONCLNTRATOR

2 E U

H ^ O

U N H

H N O 3

F l o w

M

0 2 9 8

0 02

4 4 8

U P r o d u

H 2 O

U N H H N O 3

Sp G r

F l o w

c l M

2 1

0 0 8

1 6 2

6 3 8

HNO- ,

H 2 O

H N O 3 Sp G r F l o w

M

10 4

1 31

4 8 2 A X

T B P Dil

Sp G r

T e m p F l o w

M

0 8 4 6

4 0 ' ' C

10 9

FINAL URANIUM

AEC GE RICHLAND WASH

I

S2^ •31- HW-49483 A

PARTITION

I B X

H 2 O H N O ,

M

0

F e { N H 2 S 0 3 J 0

Sp G r F l o w

1

id

57

0 3

03

I B X F U

P u

H N O 3 T B P D i l

T e m p

S p G r

F l o w

• M

0 3 3

0 2

45< 'C

0 9 5 3

4 2 0

I B S L U N

H N O 3 T B P D i

F l o w

o i l 0 16

37

I B S T B P D i l

T e m p

S p G r F l o w

M

4 5 ° C

8 4 6

36

ICX M

H2O

HNO3 0 01

TBP Dil

Sp Gr Temp Flow

30°C 529

I B X P

H 2 O

U

Pi.

H N O ,

M

0 0 3 5

1 8 4

F e ( N H 2 S 0 3 ) 2 0 0 2 8

Sp G r F l o w

1 0 8

32 *.

ICW M

I B P D 1

HNO3 Trace

l U D M

H 2 O

H N O 3 c a 0

S p G r 1 0 F l o w 4 5 0

0 1

r-I C U

CONCENTRATOR

F e ( S A ) H 2 O

F e ( N H 2

S p G r F l o w

A d d

SO3I

2

1

0

M

2 0 0 3 4

6 0

NaN02 Add

Alternate

To solvent t r ea tmen t

2 B X

H j G

H N O 3

S p G r

T e m p F l o w

M

0 01

1 0

3 0 ^ 0 7 0

HNO3 0 13

T B P Di!

Flow 10 9

2 A W H 2 O

H N O 3

Na"^

N a N 0 2 Sp G r F l o w

M

2 49 0 0 6 4

T t a c e

1 09 4 2 3

2 B W M

T B P DU H N O 3 T r a c e F l o w 10 9

2 B P M

H 2 O

P u

H N O 3 0 2 2

S p G r 1 0

F l o w 7 0

i S D

H N C j

Sp G r

T e m p F l o w

M

T r a c e

1 0

^ S O ^ C

' 8

I P O M

H 2 O

H N O j 0 7

S p G r I 02

T e m p < 5 0 ° C H o w 4 2

P U P P O n U C T M 1

" 2 ^ H N O 3

p G r

F l o w

6 0 1 4 0

0 18

FIGURE 7 PUREX

SOLVENT EXTRACTION FLOW SHEET

100 FLOWS EQUALS 747

GALLONS OF SOLUTION PER

TON OF URANIUM PROCESSED

ALL TEMPERATURES ARE

AMBIENT UNLESS OTHERWISE

SPECIFIED

SPECIFIC GRAVITIES OF

ORGANIC SOLUTIONS ARE

BASED ON A HYDROCARBON

DILUENT WITH A SP GR

OF 0 785

FINAL PLUTONIUM

UNCLASSIFIED

UNCLASSIFIED •32- HW-49483 A

3 T~

Waste Solvent

TBP-Diluent Make-Up Flow 2

lOF , ^

Waste Solvent Feed-30% TBP

Flow 350 Sp. Gr. 0.85

lOS

F r e s h Sodium Carbo

Flow NaCO Sp. G

nate

6 3 0. 5M r. 1.05

w

1 1 0

C

o L U M N

lOR X

Sodium Carbonate Recycle

Flow 35.0 Sp. Gr. 1.05

/

1 / \

\

CENTRIFUGE

S o

<

u u

1—*

W to

X) • iH 1—I

0 in

^

100

Recovered Solvent-3 0%TBP

Flow 352 Sp. Gr. 0.85

Solvent for Recycle

r

Carbonate to Waste

F I G U R E 8

P U R E X SOLVENT T R E A T M E N T F L O W S H E E T

(Based on Figure 7 Flowsheet)

AEC-GE RICHLAND. WASH. UNCLASSIFIED

UNCLASSIFIED

3S J5 O •33- HW-49483 A

HAW M HpO HNO3 Sp. Gr Flow

0.93 1.03 120

IWF M H2O HNO3 Na+ Sp. Gr . Flow

1.47 0.034 1.03 135

High -Activity Concentrator

NaOH NaOH Sp. Gr. Flow

M 19 1.53 5

lAW M H2O HNO3 Na+ Sp. Gr . Flow

0.93 0.017 1.03 120

IWW M H2O HNO3 NaN03 Sp, Gr. Flow

5.7 0.31 1.19 15

Neut. Waste M H2O NaNOj NaOH Sp. Gr. Flow

4 .5 0 .5 1.27 20

2AW H2O HNO3 Na+ Sp. Gr . Flow

M

2.49 0.064 1.09 42 .3

Water Add H2O Sp. Gr. 1.0 Flow 64

I 2WF M

H 2 0 HN03 NaNOj Sp. Gr . Flow

0 . 9 0.009 1.03 496

AFF H2O HNO3 Vapor Phase Flow 481

Low-Activity Concentrator

Acid Fract ionator

Crib

Acid for Recycle

• ^ To Waste Storage FIGURE 9

PUREX ACID RECOVERY FLOWSHEET

(Based on Figure 7 Flowsheet)

AEC-GE-RtCHLAND. WASH. UNCLASSIFIED