PUBLICATION

STATUS OF FISSION PRODUCT RECOVERY AT HANFORD

S. J. Beard and P. W. Smith

(TO be presented a t the American Chemical Society Meeting i n At lan t ic City, New Jersey, September 13 through 17, 1965.)

HANFORD ATOMIC PRODUCTS OPERATION RICHLAND. WASHINGTON

N O T I C E !

This report was prepared for use within General Electric Company in the course of work under Atomic Energy Commission Contract AT-(45-11-1350, and any views or opinions expressed in the report are those of the authors only. This report i s subject to revision upon collection of additional data.

LEGAL N O T I C E This report was prepared as an account of Government sponsored work. Neither the United States,

nor the Commission, nor any person acting on behalf o f the Commission:

Makes any warranty or representation, express or implied, with respect to the accuracy, com- pleteness, or usefulness of the information contained in this report, or that the use o f any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or

Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report.

As used in the above, "person acting on behalf of the Commission" includes any employee or contractor of the Commission to the extent that such employee or contractor prepares, handles or distrib- utes, or provides access to, any information pursuant to his employment or contract with the Commission.

A.

B.

50-3000-366 (9-59) AEC-OL RICHLAND. WASH. UNCLASSIFIED

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

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Tk r e

UNCIASSIFIED

STATUS OF FISSION PRODUCT RECOVERY AT HANFORD I

ABSTRACT

RL-SA-40 Page 2

overy of f i s s ion products was i n i t i a t e d a t Hanford i n Augu t,

The proeram has 1960 with the production of strontium-90 and cesium-137.

since included production of custom quant i t ies of cerium-144 and

promethium-147.

ra re ear ths from Purex acid waste by lead c a r r i e r su l fa te prec ip i ta t ion and

The currenC program involves separation of strontium and

subsequent separation of rare ear ths from strontiuni by oxalate precipi ta t ion.

Strontium i s fur ther purif ied by di (2-ethylhexy1)phosphoric acid solvent

extraction. Cesium-137 i s recovered from aged alkal ine wastes by processing

the alkal ine supernatant solut ion through a bed of alumino-silicate res in .

Several other procesGes have been demonstrated with full l eve l process

solutions i n plant equipment.

by zeol i te ion exchange and nickel ferrocyanide prec ip i ta t ion and from acid

wastes by phosphotungstic ac id precipi ta t ion.

and pur i f ied by anion exchange.

ear ths by d i (2 -ethylhexyl) phosphoric qcid solvent extract ion using s i l v e r

Cesium has been recov'kred from alkal ine wastes

Technetium-99 has been recovered

Cerium-144 has been separated from -the rase

catalyzed persulfate oxidation of cerium( 111) t o cerium(IV) . was subsequently separated from the ceriumrfree ra re ear ths by chromatographic

Promethium-147

ion exchange.

Future plans include the recovery of megacurie quawtitieg of Sr, C s , Ce,

and Pm i n a waste processing plant scheduled f o r s t a r tup i n 1967.

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INTRODUCTION '

A major port ion of the f i s s i o n products i n the world has been generated

a t Hanford a s by-products of the production o f plutonium-239 f o r use i n atomic

weapons.

s c i e n t i s t s and engineers f o r the two-fold purpose of confining the f i s s i o n

products i n the Hanford separations wastes f o r radiological sa fe ty and t o

explo i t the po ten t i a l of f i s s i o n products f o r bene f i c i a l use. A s ea r ly as

1955 research and development work was s t a r t ed on the recovery of selected

f i s s i o n products.

improvement program was s t a r t ed f o r which current plans (l) include ex t rac t ion

and high i n t e g r i t y storage of the long-lived heat emitters,strontium-9O and

cesium-137,and the so l id i f i ca t ion o f the short-l ived f i s s i o n products and

bulk sa l t wastes in underground storage t a n k s following 6 t o 8 years ' aging.

Fiss ion product chemistry has been studied extensively by Hanford

Shortly the rea f t e r development of a waste management

The two development programs proved t o be complementary and although the

object ives a r e d i f f e ren t the end r e s u l t of each i s a technology f o r the re-

covery of the long-lived heat emit ters from the separations p lan t wastes.

I n 1967, the waste management program w i l l be implemented and a t t h a t time

the recovery of large quant i t ies of f i s s i o n products will be real ized by

the waste ex t rac t ion processes. Meanwhile, an inter im f i s s i o n product r e -

covery program i s p a r t of the Hanford production e f f o r t .

The current f i s s i o n product recovery program a t Hanford began i n 1960

when a crash program was i n s t i t u t e d t o supply approximately one megacurie

of strontium-90 t o f u e l prototype thermoelectric devices which were being

developed f o r the A E C ' s Division of Reactor Development.

time the shipment of multi-kilocurie quant i t ies of cesium-137 t o Oak Ridge

National Laboratory was i n i t i a t e d .

A t about the same

I n 1963 the program was expanded t o

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include the recovery and s tockpi l ing of strontium-90 crude and a rare-ear th

crude containing cerium-144 and promethium-147. Since then custom quant i t ies

of cerium-144 and promethium-147 have been i so la ted and shipped t o Battelle

Northwest Laboratory and Oak Ridge National Laboratory. I n addition, promising

flowsheets f o r the recovery and pu r i f i ca t ion of these and other po ten t ia l ly

valuable f i s s i o n products have been f i e l d t e s t ed i n f u l l scale p lan t equipment.

This paper w i l l discuss the f i s s i o n product recovery a c t i v i t i e s cur ren t ly

being conducted a t Hanford and describe the processes used.

flowsheet performance w i l l be discussed f o r the individual f i s s i o n product

Production p lan t

recovery processes. F U l l e v e l p lan t tests of f i s s i o n product recovery and

pu r i f i ca t ion flowsheets not current ly used i n the production operation w i l l be

reviewed.

CUPRENT RECOVERY ACTIVITIES

The f i s s i o n product recovery program a t Hanford now includes the recovery

of strontium-90, promethium-147 and cerium-144 from current ly generated Purex

p lan t separations wastes and the recovery of cesium-137 from Purex s tored

a lka l ine waste supernatant solut ions. The strontium and ra re earths, including

cerium and promethium,are removed from the Purex ac id waste by a lead c a r r i e r

sulfate prec ip i ta t ion process.

from the strontium by oxalate prec ip i ta t ion .

The rare earths are subsequently separated

The strontium f r ac t ion i s then

purif ied by d i (2 -ethylhexyl)phosphoric ac id (D2EHPA) solvent ex t rac t ion o r

s tored as strontium crude. Pur i f ied strontium i s s tored o r loaded i n t o f i l t e r

casks f o r shipment t o user si tes. The rare-ear th crude fraction,containing

cerium and promethium,is stockpiled t o provide an inventory of w e l l aged

promethium-147 f o r future AEC programs. Periodically, promethium i s fu r the r

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pur i f ied and shipped t o a user .

a lka l ine waste supernatant solution d i r e c t l y i n t o shipping casks containing

alumino-sil icate ion exchange media a s needed t o meet AEC shipping schedules.

Cesium-137 i s recovered from aged Purex

Special f i s s i o n product loadouts a re made on demand and process flowsheets

are t e s t ed i n the f i s s i o n products recovery equipment during shutdown periods

i n the operating schedule.

batch solvent ex t rac t ion of shor t cooled feed and shipped t o a User s i t e i n a

f i l t e r cask.

Cerium has been separated from other ra re ear ths by

Technetium-99 has beep recovered from Purex a lka l ine waste super-

natant solut ions by anion exchange.

natant solut ions by zeo l i t e ion exchange and by n icke l ferrocyanide prec ip i ta t ion

and from Purex ac id waste by prec ip l ta t ion w i t h phosphotungstic aci.d.

Cesium has been recovered from the super-

REXOVERY EQUIBENT

Fission product recovery is conducted i n large scale separations f a c i l i t i e s .

Prec ip i ta t ion equipment i n the Purex p lan t i s used f o r the i n i t i a l separation of

strontium, and the r a re ear ths from the Purex ac id waste.

ea r th crude i s accumulated i n an underground vau l t f a c i l i t y from where it i s fed

t o prec ip i ta t ion and concentration equipment i n the B-Plant, a par t ia l ly remodeled

former separations p lan t .

B-Plant equipment and the crude f rac t ions a re s tored i n B-Plant storage tanks.

Strontium crude i s t ransfer red v i a the vau l t f a c i l i t y t o the Strontium Semiworks,

a former p i l o t plant f o r the Redox and Purex processes, for pur i f ica t ion and

subsequent storage o r loadout of the pur i f ied strontium.

s t a t i o n i s located near a waste storage tank farm fo r loading cesium shipping

casks.

tank and waste receiver .

The strontium-rare

The strontium i s separated from the r a re ear ths i n the

A cesium loadout

Former waste storage tanks a r e used f o r the cesium loadout s t a t i o n feed

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LEAD NITRATE

SODIUM HYDROXIDE SODIUM CARBONATE I SODIUM SULFATE t

STRONTIUM, CERIUM, 4 RARE EARTHS

O X . L A TE PRECIPITATION OXALIC ACID

NITRIC ACID

I

F I G W I G-122-300

I

UXC IASSIFIED RL-SA -40 Page 6

LEAD CARRDR SULFATE PRJXIPITATION

Strontium-90 and the rare ear ths , which include cerium-144 and pronethim-147,

are recovered from the Purex ac id waste i n Purex plant centr i fugat ion equipment

o r ig ina l ly used f o r dissolver solut ion c l a r i f i c a t i o n and subsequently made a v a i l -

able f o r this purpose through Purex process improvemmts.

stream from the waste concentrator i s centrifuged t o remove so l ids , primarily

The Purex ac id waste

s i l i c a and su l fa tes , which carry f i s s i o n products. The so l ids a re leached w i t h

n i t r i c ac id t o recover the car r ied f i s s i o n products and the leachate i s combined

w i t h the so l ids -free supernatant solut ion f o r processing.

A l ead-car r ie r su l f a t e prec ip i ta t ion process (* 3), shown schematically i n

Figure I, i s used t o recover strontium, cerium and the ra re ear ths .

and r a re ear ths a re prec ip i ta ted a s the sodium su l fa te - ra re earth su l f a t e double

s a l t by adjust ing the su l f a t e concentration t o one molar and the solut ion pH t o

1.0.

avoided by the addi t ion of a chelat ing agent such as glycol ic ac id .

c i p i t a t e i s metathesized with a mixture of sodium hydroxide and sodium carbonate

t o remove lead su l f a t e and t o convert the strontium, cerium and ra re ear ths t o

acid-soluble carbonates.

The cerium

Lead n i t r a t e i s added t o car ry strontium su l f a t e . I ron p rec ip i t a t ion i s

The pre-

The p rec ip i t a t e i s dissolved i n d i l u t e n i t r i c ac id and

t ransfer red t o interim storage p r i o r t o separat ion of strontium from the cerium

and r a re earths.

I n typ ica l p lan t flowsheet performance only about 70 percent of the s t ron-

t i u m and ra re ea r ths i n the ac id waste are made ava i lab le t o the su l f a t e

p rec ip i t a t ion process due t o l i p i t e d leaching of the f i s s i o n products from the

ac id waste so l id s .

sheet, however i s grea te r than 90 percent.

Recovery of the isotopes i n the su l f a t e p rec ip i t a t ion flow-

Decontamination fac tors grea te r

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,

UNC IASSIFIED

than 100 have been measured f o r i ron and nearly quant i ta t ive separation of

aluminum from the product isotopes has been demonstrated i n the p lan t w i t h the

su l f a t e process. Supporting p i l o t p lan t s tudies(4) have shown similar

decontamination from iron, aluminum and other metal ions including manganese

and magnesium. The removal o f manganese and magnesium i s of pa r t i cu la r import-

ance since these two ions are chemically similar t o strontium i n the subsequent

solvent ex t rac t ion pur i f ica t ion process.

OXALATE PRECIPITATION OF THE RARE EARTHS

The r a re ear ths , cerium and lead a re separated from strontium by the addi t ion

of oxalic ac id t o the crude f i s s i o n product feed as i l l u s t r a t e d i n Figure I.

so lu t ion is made about 0.35 molar i n oxal ic acid while holding the n i t r i c ac id

between 0.25 and 0.5 molar.

cerium, r a re

The strontium-bearing so lu t ion i s concentrated and s tored f o r s tockpi l ing o r t o

await further processing. The oxalate p rec ip i t a t e i s s lu r r i ed from the cen t r i -

fuge with four molar n i t r i c acid, concentrated and t ransfer red t o storage tanks

for aging.

leaving soluble rare earth n i t r a t e s i n storage.

The

This ac id side oxalate process p rec ip i t a t e s the

ear ths and lead, leaving the strontium i n the supernatant solut ion.

The oxalate i n the so lu t ion i s destroyed i n a f e w days by radiolysis

Plant experience with t h i s process has shown good separation of strontium

and the rare earths.

rare earths t o the supernatant solut ions a r e typ ica l ly f ive percent and two per-

cent, respectively. Any losses due t o poor separation i n the prec ip i ta t ion s t e p

are poten t ia l ly recoverable by recycle.

The f r ac t ion of strontium l o s t t o the prec ip i ta te and the

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UNC IASS IFIED RL-SA-40 Page 8

STRONTIUM PURIFICATION BY SOLVENT EXTRACT I O N

The strontium fract ion from the B-Plant i s transferred t o the Strontium

Semiworks v ia the storage vault fo r f i n a l purif icat ion and loadout o r storage.

The purif icat ion i s performed i n two Strontium Semiworks pulse columns and

associated tankage by solvent extraction using d i (2-ethylhewl)phosphoric acid

(D~EEIPA) extractant i n a hydrocarbon di luent with t r ibu ty l phosphate a s a

modifier(5). The process i s i l l u s t r a t ed i n Figure 11.

Strontium, calcium and a small amount of cerium are extracted in to the

solvent i n the first column (HA) effect ively separating strontium from the

majority of the metal ion and f i s s ion product contaminants. An aqueous phase

chelating agent, sodium diethylenetriaminepentaacetate (DTPA) , i s used t o re - duce the ex t rac tabi l i ty of the contaminants. The pH during extraction i s

maintained i n the desired range, between 4.5 and 5 .O, by sodium acetate

buffering. The strontium i s mr t i t i oned f r ~ m calcium and the remainder of

the cerium by selective stripping i n the second column (KC) under carefully

controlled pH conditions made possible by the use of c i t r i c acid a s a stripping

agent. The cerium and calcium are then washed from the solvent with n i t r i c acid

i n a batch contactor and the solvent i s prepared fo r recycle by a batch water

wash t o remove excess acid. The c i t r i c acid i n the s$rontium product solution

i s continuously destroyed during the processing by boiling with n i t r i c acid

and hydrogen peroxide t o avoid sol ids and t a r formation associated with radio-

l y t i c degradation of c i t r a t e . Following analyses t o confirm tha t the strontium I

meets customer specifications and the c i t r a t e ion i s adequately removed, the

product solution i s stored i n the Strontium Semiworks i n a vaul t .

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H

H

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The D2EHPA solvent extract ion process has proven highly e f f i c i e n t f o r

the production scale pur i f ica t ion of strontium.

factors f o r metal ion and f i s s ion product contaminants a re shown i n Table 1

f o r several pur i f ica t ion runs.

been adequate t o provide strontium product with l e s s than ten percent by

weight metal impurities, excluding a l k a l i metals, and l e s s than f ive curie

m i c a 1 decontamination

The contaminant removal e f f ic ienc ies have

percent f i s s ion product impurit ies.

TABLE 1

TYPICAL DECONTAMWTION FACTORS I N STRONTIUM PURIFICATION

- 1 - 2 3 Ca 20 10 25

Ea 20 3 95 5

Pb > 1000 7500 800

Fe - 1300 400

Ce 1600 500 7000

Z r -Nb 8000 100 I200

Ru 2 700 35 400

STRONTIUM LOADOUT I N T O FIIEER CASKS

Purif ied strontium product i s per iodical ly loaded out in to f i l t e r casks

f o r shipments t o o f f - s i t e customers.

the addi t ion of potassium bicarbonate following partial neutral izat ion with

sodium hydroxide. The solution i s adjusted t o a final pH of 10 w i t h excess

potassium carbonate and digested a t 90 C t o assure complete prec ip i ta t ion of

strontium.

Strontium carbonate i s precipi ta ted by

?"ne use of carbonate ra ther than sodium hydroxide i n the f inal

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pH adjustment avoids the prec ip i ta t ion of the strontium decay daughter

zirconium-90

storage of strontium-90.

f i l t e r shipping cask.

it i s dr ied a t temperatures up t o 150 C t o s t ab i l i ze the f i l t e r cake f o r ship-

ment.

determine rate of pressure buildup.

it i s released t o the AEC fo r shipment o f f - s i t e .

( 6 ) which builds up t o s ign i f icant quant i t ies during prolonged

After cooling, the s lur ry i s t ransferred in to a

The strontium prec ip i ta te co l l ec t s on the f i l t e r where

The cask i s checked for leaks, purged with helium and monitored t o

When the cask meets shipping requirements

SEPARATION AND PURIFICATION OF CERIUM AND PROl4J3I”IUM

In preparation f o r cerium-promethium separation, metal ions, primarily

lead, a re removed from the ra re ear th f rac t ion by e i t h e r batch o r continuous

solvent extract ion with the D2E”A solvent used f o r strontium purif icat ion.

The aqueous ra re ear th feed i s adjusted t o pH 2.0 t o 3.0 after the addi t ion of

a chelating agent, N-hydroxyethylethylenediaminetriaacetic ac id (HEDTA) , t o

complex the lead and other metal ions and render them l e s s extractable .

ra re ear ths a re stripped from the solvent with n i t r i c ac id providing a re la t ive ly

pure feed f o r cerium-rare earth separation.

Cerium i s separated from promethium and other ra re ear ths by batch solvent

The

extract ion (7’8) a s i l l u s t r a t e d i n Figure I11 using the same D2EIIPA solvent.

The cerium i s oxidized t o cerium(N) by s i l v e r catalyzed sodium persulfate

oxidant.

leaving the remainder of the rare ear ths ( t r i va l en t ) i n the solution.

Cerium(IV) i s extracted in to $he organic from an acid solut ion

The

cerium i s stripped from the solvent w i t h n i t r i c acid containing a t race of

n i t r i t e ion t o accelerate the reduction of cerium(IV) t o cerium(II1). The

ra re ear th f rac t ion i s separated from the s i l v e r and su l fa te reagents used i n

UNCLASSIFIED

D2EHPA SODIUM HYDROXIDE SILVER NITRATE NITRIC TBP CHELATING AGENT POTASSIUM PERSULFATE ACID

RARE EARTHS LEAD BEARING PROMETHIUM TO UNDERGROUND - CEF

PRC UNDERGROUND

STORAGE WASTE TO

c 3

0 2 E l TBP

SILVER, LEAD AND SULFATE BEARING WASTE TO

UNDERGROUND STORAGE

F A

DILUENT

I

I I

WASTE SOLVENT TO MLVENTI

I I I I I I I

S TR€A M LEGEND RADIOACTIVE AQUEOUS = PROCESS CHEMICALS U 1 1 ORGANIC T O 8 - PLANT

STORAGE FOR PROMETHIUM AGING

I I I I

F I G W I11 G-122- 302

UNCLASSIFIED RL-SA-40 Page 11

the oxidation s t ep by e i t h e r batch o r continuous ex t rac t ion w i t h D2EHPA solvent

i n a pH range of two t o four af ter addi t ion of HEDTA t o so lubi l ize r a re ea r th

su l fa tes .

Promethium has been i so la ted on a p i l o t plant scale@) from the r e l a t ive ly

cerium-free r a re ear th f r ac t ion by chromatographic ion exchange (lo) . The rare

ear ths are loaded onto a column of cat ion exchange r e s in . The rare ear ths a re

then eluted w i t h a chelating solut ion (diethylenetriaminepentaacetic acid (DTPA)

i s favored) through subsequent columns where the rare earths are d is t r ibu ted i n

bands i n an order es tabl ished by the equilibrium between the ca t ion r e s i n and

the chelating agent.

i s su f f i c i en t ly i so la ted t o permit the co l lec t ion of a pure promethium f rac t ion .

About 500 ki locur ies of cerium-144 were i so l a t ed from shor t cooled feed

f o r shipment t o Oak Ridge National Laboratory i n an ea r ly cerium-promethium tes t

run. The cerium was loaded onto a s lu r ry f i l t e r cask a s a sodium sulfate-cerium

su l f a t e cake (l1JU) i n a manner s i m i l a r t o a strontium carbonate loadout.

After moving through severa l columns the promethium band

The

cerium cake was removed from the cask with n i t r i c acid a t the user s i t e .

Rare ea r th f rac t ions have been separated from cerium i n several p lan t scale

runs.

I so la ted rare ea r th f rac t ions have been shipped t o Oak Ridge National Laboratory

and Bat te l le Northwest Iaboratary f o r research and development s tud ies pointed

toward the development of promethium as a hea t source.

e s sen t i a l ly f r e e of cerium were provided for the p i l o t p lan t sca le tests of

F u l l l e v e l tests have been conducted t o invest igate the process parameters.

Rare ear th feeds

f i n a l promethium pur i f ica t ion by chromatographic ion exchange conducted by the

Hanford Chemical Processing Department.

The p lan t scale performance of the cerium-promethium separation flowsheet

has exhibited the typ ica l plant s t a r tup problems f o r flowsheets scaled up

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d i rec t ly from laboratory data.

ment i n the presence of high radiat ion f i e lds , which could not be ant ic ipated

from laboratory t e s t s , have been encountered and methods for overcoming them

have been developed a s a coopra t ive e f f o r t between the plant engineers and

laboratory personnel Despite s t a r tup d i f f i c u l t i e s , s ign i f icant separations

have been achieved between cerium and promethium t o meet the requirements of

Problems associated with the use of plant equip-

the users.

CESIUM RECOVERY BY I O N EXCHANGE

Cesium i s recovered from aged Purex alkal ine waste supernatant solution by

ion exchange with beds of alumino-silicate gel , Decalso*, loaded i n shielded

shipping casks (13j14) a s shown i n Figure I V . The alumino-silicate i s highly

select ive f o r cesium which i s present i n only t race quant i t ies i n the con-

centrated sodium n i t r a t e waste.

and a cooler i s therefore used t o reduce the feed temperature a t the i n l e t t o

the ion exchange bed.

maximized by loading the bed t o near 100 percent breakthrough using a recycle

l ine , not shown i n the i l l u s t r a t ion , t o route the e f f luent back t o the feed tank

Cesium se l ec t iv i ty i s bes t a t low temperatures

The quantity of cesium t h a t can be shipped i n a cask i s

a f t e r in i t ia l breakthrough i s observed.

user s i t e w i t h ammonium n i t r a t e .

Cesium i s eluted from the casks a t the

The production scale recovery of cesium by ion exchange has been highly

Cesium losses a re very low due t o the recycle of e f f luent t o the e f f i c i e n t .

feed tank a f t e r in i t ia l breakthrough.

estimated t o be grea te r than 300 during loading. Some incentive e x i s t s however,

t o increase the amount of cesium t h a t may be loaded on the cesium shipping casks

Decontamination fac tors f o r sodium a r e

i

* Trade name of Pfaudler Permutit

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UNCIASSIFLED RL-SA-40 Page 13

i n order t o reduce the number o f r a i l shipments required t o meet AEC schedules.

A recent ly t e s t ed commercial zeo l i te , Linde AW+OO*, has more than double the

capacity of Decalso f o r cesium l ~ a d i n g ( ~ ~ > ~ ~ ) , providing a po ten t i a l f o r cesium

shipping economies through cask r e s in changeout.

TECHNETIUM RECOVERY

Technetium has been recovered from aged Purex a lka l ine supernatant solutions

i n kilogram quant i t ies by an anion exchange process(’7).

exchange r e s i n was connected i n se r i e s with a cesium shipping cask a t the cesium

loadout s t a t i o n as shown i n Figure IV t o recover the t race quant i t ies o f technetium

A cask f i l l e d w i t h anion

present i n the waste.

so lu t ion which was then concentrated and deni t ra ted as shown i n Figure V. Final

pur i f ica t ion p i o r t o preparation of technetium metal was done by a second anion

exchange cycle.

The technetium was e lu ted from the cask .with n i t r i c ac id

Two p lan t t e s t s were conducted with the recovery of about a kilogram of

technetium i n each tes t . The technetium from the f i r s t recovery tes t was t r ans -

fe r red t o Hanford Laboratories, now Bat te l le Northwest Laboratory, f o r f inal

pu r i f i ca t ion and reduction t o m e t a l for use i n a NASA research program.

pur i f i ca t ion of the technetium from the second recovery t e s t w a s conducted by

F i n a l

the Hanford Chemical Processing Department.

the major contaminants was excel lent .

I n each case decontamination from

* Trade name of Linde Division, Union Carbide Corporation

UNC LASS IFIED

,r> ssw WASHING

SODIUM NITRATE DILUTE

NITRIC ACID

FROM-G FARM LOADOUT STATION

STREAM L €@END RADIOACTIVE AQUEOUS

PROCESS CHEMICALS L”_LI

RADIOACTIVE VAPOR EEEl

I I WASTE T O 1 UN DERGROUN 0 7 STORAGE

NITRIC

FL UT/OAf

ACID

STEAM STRIPPEO NITRIC AClQ TO WASTE

t CONCENTRATED TECHNETIUM

PRODUCT I

STORAGE LOADOUT

FIGURF: V G - 122 - 307

1 .

REFE€ENCES

UNCIASSIFIED RL-SA -40 Page 14

1. Beard, S. J* j R. J. Kofoed, J. J. Shefcik and P. W . Smith, Waste Management Program - Chemical Processing Department, HW-81481. March 24, 1964.

2 . Bray, L. A. and H. H. Van Tuy l , Laboratory Development of a Carr ier- Prec ip i ta t ion Process f o r the Recovery of Strontium from the Purex Wastes, HW-69534. May 9, 1961.

3. Van Tuyl, H. H., Improvement of Strontium Sulfate P rec ip i t a t ion Process, HW-74036. June, 1962.

4. Richardson, G. L., Recovery of Strontium from Stored Purex Waste by b a d Sulfate Carr ie r Prec ip i ta t ion - P i l o t P lan t Studies, BNWL-4. January 15, 1965.

5. Members of the Chemical Development Operation, Hot Semiworks Strontium-90 Recovery Program2 HW-72666. July, 1963.

6. Bray, L. A., Zfrconium Removal from Pur i f ied Strontium, BNWC-24. February 5, 1965

7. Bray, L. A. and F. P. Roberts, Separation of Trivalent Rare Earths from Cerium, HW-78987. October 10, 1963.

8. Bray, 1,. A. and F. P. Roberts, Supplemental Studies on the Separation of Cerium Iv from the Trivalent Rare Earths, HW-84101. October 1, 1964.

9. Buckingham, J. W., Ion Exchange Pur i f ica t ion of Promethium i n the 2 2 2 4 Building, 1-F Cubicle, HW-79029. Ju ly 1, 1964.

10. Wheelwright, E. J. and F. P. Roberts, Development and Demonstration of an Ion Exchange Process f o r Kilogram-Scale Production of High Pur i ty Promethium, HW-78651 REV. October, 1963.

11. Van Tuy l , H. H., Laboratory Studies of the Prec ip i ta t ion of Cerium Sulfate for O f f -Site Shipment, HW-71589. November, 1961.

12. DeMier, W . V., Non-Radioactive Testing of the H M O - I , -11 Fission Product Shipping Cask, Kw-71744. February 19, 1962.

Bray, L. A. and H. R . Van Tuyl, Recovery of Cesium from Purex Tank Farm Supernatant Solution by Ferrocyanide P rec ip i t a t ion and Absorption on Decalso, HW-70874. October 18, 1961.

Smith, F. M., A Laboratory Study t o Define Conditions f o r b a d i n g Strontium-90 on Decalso f o r W O - I 1 Cask Shipment, HW-71574.

13.

14. November 30, 1961.

15 Mercer, B. W., Unpublished Data. Ba t t e l l e Northwest Laboratory, Richland, Washington.

UNCIASSIFIED

UNCIASSIFIED RL-SA-40 Page 15

16. Mercer, B. W . and L. L. Ames, The Adsorption of Cesium, Strontium and Cerium on Zeol i tes from Multication Systems, m-78461. August, 1963.

17. Smith, F. M. and F. P. Roberts, Unpublished Data. Ba t t e l l e Northwest Laboratory, Richland, Washington.

UNC LASS IFIED