OR N Ld3566' UC- 10 - Chemical Separations Processes

for Plutonium and Uranium TID-4500 (28th ed. )

h

RECOVERY OF PLUTONIUM AND OTHER

TRANSURANIUM ELEMENTS FROM IRRADIATED

PLUTONIUM-ALUMINUM ALLOY BY

ION EXCHANGE METHODS

R , E. Brooksbank W. T. McDuffee

UNION CARBIDE. CORPORA1 ION for the

U.S. ATOMIC ENERGY COMMISSION

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. 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.

DISCLAIMER

Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

d. as an acraunt of borarrimnt

any parhon ecfing on behhlf of

C o n t r a c t NO. W-7405-eng-26

CHEMICAL TECHNOLOGY D I V I S I O N

RECOVERY O F PLUTONIUM AND OTRER TRANSURANIUM ELEMENTS

FROM IRRADIATED PLUTONIUM-ALUMINUM ALLOY

BY I O N EXCHANGE METEODS

R. E . B r o o k s b a n k W . T. M c D u f f e e

MAY 1964

OAK RIDGE NATIONAT, LABOMTORY Oak R i d g e , T e n n e s s e e

operated by UNION CARBIDE CORFORATION

fo r the U . S . ATOMIC ENERGY COMMISSION

T H I S PAGE

WAS INTENTIONALLY

LEFT BLANK

CONTENTS

. . . . . . . . . . . . . . . . . . . . . . . Abstract

. . . . . . . . . . . . . . . . . . . 1 . Introduction

2 . S m a ~ y . . . . . . . . . . . . . . . . . . . . . 3 . Flowsheets . . . . . . . . . . . . . . . . . . . .

. . . . . . . . . . . . . 3.1 Chemical Flowsheet

. . . . . . . . . . . . . 3.2 Equipment Flowsheet

4 . Procedure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 Feed Preparation

4.1.1 Description and History of the Fuel . 4.1.2 Shipping Cask Description and Handling

4.1.3 Charging the Fuel t o the Dissolver . . . . . . . . . 4.1.4 Dissolution of t he Fuel

. . . . . . . 4 . 1. 5 Dissolution Rate Studies

4.1.6 Feed Clarif icat ion and Adjustment . . . . . . . . . . . . . . . 4.2 PlutoniumRecovery

4.2.1 Ion Exchange . . . . . . . . . . . . . . . . . . . . . 4.2.2 F i r s t Stage Operation

. . . . . . . . . . . . . 4.2.3 Second Stage

, 4.2.4 Handling of the Plutonium Product . . 4.3 Transuranium Element Recovery . . . . . . . .

. . . . . . . . . . . 4.3.1 Feed Preparation

. . . . . . . . . . . . . 4.3.2 Ion Exchange

4.3.3 Transurantum Element Product Handling

. . . . . . . . . . . . . . . . . . . . . 5 . Appendix

. . . . . . . . . . . . 5.1 Detailed Process T h f . a

. . . . . . . . . 5.2 Examinat ion of Spent Resin

. . . . . . . . . 6 . Conclusions m d Recommendations

. . . . . . . . . . . . . . . . . 7 . Acknowledgement

8 . References . . . . . . . . . . . . . . . . . . . .

RECOVERY OF PLUTONIUM AND OTHER TRANSURANIUPI ELEMENTS FROM

IRRADIATED I?LUTONl3JIVI-A~II~~1 ALLOY BY

I O N EXCHANGE METHODS

ABSTRACT

I n order t o i so la te plutonium and other transuranium elements such

as americium and curium from the aluminum i n highly irradiated Pu-A1 al loy

fue l elements, an ion exchange process was demonstrated. The ainericiuln

and curium w i l l be used in -the nuclear studies related t o the very heavy

elements. The report describes the process i n d e t a i l and gives resu l t s

from the developmental work and the p i lo t plant demonstrat ion. The pro-

2+ cess consisted i n dissolving the fue l rods i n 6 M HNO -- 0.05 M Hg -- - 3 - 0.03 - M F-, adjusting the valence of the plutonium t o four and clar i fying

the solution by f i l t r a t i o n through a sand-bed f i l t e r precoated with f i l - t e r

aid, a f t e r which the plutonium was recovered by sorption on anion exchange

resin, the sorbed plutonium then being scnibbed f ree of aluminum and

f i ss ion products with 7 M HNO -- 0 . O 1 M F-. The par t ly purified plutoni- - 3 -

urn was then eluted from the resin with 0.7 M HlJO Additional purif ica- - 3' t i o n and decontamination was obtained by repeating the ion exchange s tep.

The plutonium product, meeting radiation specifications, was nearly f ree

of aluminum and had been decontaminated from f i s s ion products by a fac tor

5 of 4.6 x 10 . Overall plutonium recovery processing losses of 2.7 and

12.6%, respectively, occurred with fue l elements i r radiated t o 87 and

greater than 97% bumup.

The Lransuranlum elements were then recoverea from the aluminum-

bearing, plutonium loading waste eff luent solution, which had been

collected and stored during the plutonium-recovery phase of the prograin.

The free acid i n the raw wastes was removed by evaporation t o 11IO0c, a f t e r

which the concentrate was di luted with water u n t i l the concentration of

the aluminum was 2.5 t o 2.8 M. - A t t h i s point, the aluminum n i t r a t e was

nearly monobasic. After c l a r i f i ca t ion t o remove the aluminum hydroxide

t h a t was formed, the. rare ear ths and transuranium elements were sorbed on

anion exchange res in at 6 0 ' ~ . Traces of aluminum and l igh t f i ss ion

products t h a t had sorbed on the jresin were sc~ubL~ed J? l s j~u il; w&s 8 1.1 - LiNO 3 '

The purified rare ear th f rac t ion containing the transuranium elements was

then eluted from the res in with 0.7 M mT0 and evaporated 'Lo s~iiall volume. - 3

The conditions developed and demonstrated in the recormnded. flow-

sheet f o r the recovery of plutonium and other transuranium elements from

irradiated Pu-A1 a l loy f i e 1 permitted average dissolution rates of the

-2 a l loy of about 0.8 m g m i d 1 cm and resulted i n plutonium recoveries

greater than 9%. Some reduction i n plutonium recovery may be expected,

however, as the burnup of the fue l elements approaches 99% because of the

destruction o r masking of the exchange s i t e s on the resin. I n the runs

described, res in exposure t o the radiation was suf f l c l e ~ r t i n Lwo instances

t o destroy about 35% of the exchange s i t e s , a ra ther serious defect Of the

process.

Transuranium element recoveries grea ter than 95% were demonstrated

operating according t o the recommended flowsheet developed during t h i s

program. High recovery of the trarlsurqium elements depended on t h e i r

complete recovery from the aluminum hydroxide cake collected during the

c l a r i f i ca t ion s tep. The cake retained about 35% of the transuranium

elements, which could he recovered. by rinsing with d i lu t e n i t r i c acid

and rei;urning t o the evaporator. High sa l t ing strength (8 M NO-) was - 3 necessary in the feed and scrub solutions i n the ion exchange step.

The process described here of fers a useful approach t o the recovery

of plutonium and. other transuraniun elements from Pu-A1 al loy fue l

e lel~~ents , even those of high burnup. The products, plutonium and. t rans-

uranium elements, w i l l be nearly free of aluminum and therefore adaptable

t o subsequent processes f o r isolat ing and recovering them.

1. INTRODUCTION

The recovery of plutonium and transuranium elements from highly

irradiated Pu-A1 al loy was undertaken t o provide suf f ic ien t quant i t ies

f o r the study of t h e i r nuclear properties and f o r the development and

demonstration of processes f o r t h e i r separation and recovery i n a high

s t a t e of purity. This separation takes place i n two steps: one tha t

provides aluminum-free plutonium product, and a second m e which i so la tes

and recovers the transu-ranium elements (plus rare earths) from the

plutonium-free eff luent of the f i r s t s tep. The purpose of the work was

t o provide a process fo r LIE ~ ~ ~ r ~ u v a l of the alminum t o such an extent

tha t it could no longer be regarded as an interference i n subsequent

protessil~g Lo fulluw. T h i s report presents the resu l t s of' a p i l o t plant

demonstration of a process, featuring anion exchange separation and puri-

f ica t ion methods, in the recovery of plutonium and transuranium elements

froin two l o t s of Pu-A1 al loy fue l rods i r radiated t o greater than 8%

burnup.

Specific objectives of the program were to :

1. Recover the plutonium from two l o t s of highly irradiated Pu-A1

a l loy f'uel rods that. would be suf f ic ien t ly decontaminated from

f i s s i o n products and aluminum f o r use i n fur ther development and

i n the fu r the r production of transuranium elements.

2. Recover the transuranium elements f ree of aluminum, i n a condition

su i tab le f o r use as r a w material i n the development of processes

f o r the separation and recovery of highly purified iiidlvidual

c l e ~ n t . s ,

3. Demonstrate on a p i l o t -plant scale a flowsheet f o r the recovery

of purified plutonium and other transuranium elements as aluminum-

f r ee products .

Prominent problems t o be solved consisted i n dissolving the spent fue l

elements, removing the s i l i c a from the solution, and contending with the

r e s in degradation resul t ing from the re la t ive ly high concentrations of

f i s s ion products i n fue l elements of up t o 99% burnup. Other problems

encountered during the recovery of the transuranium elements were: (1) the

formation of aluminum hydroxide during evaporation of the eff luent from

the plutonium recovery step; (2) the removal of aluminum hydroxide from

the evaporated eff luent ; ( 3 ) recovery of the transuranium elements from the

aluminum hyd-ide f i l t e r cake; and (4) res in degradation which resulted

i n increased los s of material. Some of the problems were unusual because

of the high b u m p of the f'uel elements, and others were due t o the high

salt strengths of the loading and scrubbing solutions.

The process includes dissolut ion of the al loy i n 6 M HNO catalyzed - 3

with 0.05 M - H ~ ~ + and 0.03 M - F-, c la r i f i ca t ion and adjustment of the plu-

toniwn valence t o four and the acid concentration t o 7 M HNO and two - 3 stages of anion exchange purif icat ion and recovery. The process also in-

cludes steps t o recover the transuranium elements from the ion exchange

loading waste eff luent from the plutonium recovery s tep by removing the

free n i t r i c acid by evaporation, adjustment of the acid and alwninum con-

centration t o 0.5 t o 1.0 - N acid deficient and 2.5 t o 2.8 M, - respectively,

followed by anion exchange sorption of the rare ear ths and transuranium

elements. Additional purif icat ion was obtained by scrubbing t races of

sorbed aluminum and l igh t f i ss ion products from the resin, a f t e r which

the transuranium elements (and rare earths) were eluted from the res in

with 0.7 M HNO and concentrated to ' smal l volume by evaporation. - 3

Ion exchange was selected because the technology f o r plutonium re-

covery i s well known and it had been shown tha t the transuranium elements

also sorb un anion exchangers from neutral , high-salt-strength solutions.

A s expected, the process worked,though some defects, especially the low

loading capacity of the ion exchanger f o r sorbing the transuranium elements,

were noted. The information in t h i s report i s presented in the following

order: dissolution of the fuel elements, valence adjustment of the plu-

tonium, and i t s sorption and recovery, f ree of aluminum. This is followed

by a section on the evaporation of the ion exchange eff luent from the

plutonium recovery step, c la r i f ica t ion , and ad. justment of the solution

p r io r t o the sorption and. recovery of the transuranium elements and rare

earths, f ree of both aluminum and plutonium.

2. SUMMARY

Two l o t s of Pu-A1 al loy fue l tha t had been irradiated t o greater than

8 f i burnup were processed on a p i l o t plant scale t o recover about 700 g of

plutonium, r ich i n the heavier isotopes and pure enough t o meet radiation

and puri ty specif icat ions. In addition, about 4 g of transuranium elements

w a s recovered as a purified rare ear th fract ion tha t was essent ia l ly

aluminum-free and sui table as feed material in the f'urther development of

separations processes fo r the recovery of individual transuranium elements .

The process was conducted t o demonstrate a flowsheet featuring anion

exchange methods tha t had been developed t o recover plutonium and other

transuranium elements free of the aluminum original ly present. This f i e 1

consisted of two l o t s of 0.94-in-dim by 5-ft-long plutonium-aluminum alloy

rods canned i n aluminum; one l o t (24 rods) had been irradiated t o about

87% burnup and the second (8 rods) t o 99%. These were dissolved in 6 - M

HNO --0 -05 M Hg2+--0 .03 M F ; di.ssolved s i l i c a was coagulated and removed 3 - -

by digesting the solution wi'th 10 ppm gela t in (followed by plutonium

valence adjustment) and f i l t r a t i o n . After acid adjustment t o 7 M HNO - 3' the plutonium w a s sorbed a t 60 t o 7 0 ' ~ on 20 t o 50 mesh P e m t it SK anion

exchange re s in i n the first of two ion exchange purif icat ion stages. Re-

s idual sorbed aluminum and f i s s ion prod.ucts were scrubbed Pro~rl the resin

with 7 M HNO --0.01 M F- and the plutonium eluted with 0.7 M KNU 'Yhe - 3 - - 3' plutonium product solution from the first purif icat ion stage was adjusted

t o convert the plutonium valence t o four and the acid concentration t o 7 M - HNO a f t e r which the plutonium was sorbed on a second bed of anion exchange

3'

r e s in ident ical t o tha t in the f i r s t . Traces of residual aluminum and

f i s s ion products were scrubbed from the resin, f i r s t with 7 M HNO --0.01 M - 3 -

F- then with a small r inse with fluoride-free 7 M HNO The finished plu- - 3'

tonium product w a s recovered from the res in by elut ion with 0.7 M HNO - 3'

The t ransuranium elements were recovered from the aluminum-bearing

ion exchange loading waste eff luent tha t had been collected and stored

during the plutonium-recovery phase of the program. These were f i r s t

evaporated t o 1 4 0 ~ ~ t o remove free acid and t o concentrate the aluminum,

then diluted with acid and water t o ad just t he acid concentration t o 0.5

t o 1.0 - N acid deficient and the aluminum t o 2.5 t o 2.8 M. - After f i l t r a t i o n

through a bed of Ottawa sand precoated with f i l t e r aid t o remove the

aluminum hydroxide, the c la r i f ied solution was passed a t 0.5 t o 1.0 m l

-1 -2 min cm through a 4-in-diam by 4-0-in-long bed (7.5 l i t e r s ) of 50 t o 100

mesh Dowex 1 X10 anion exchange res in held a t 60 t o 70°C, where the rare

ear ths and the transuranium elements were sorbed. Residual aluminum and

l igh t f i ss ion products were scrubbed from the bed wtth 8 M - LiNO metered 3

t o the res in column a t the same ra te and tenlperature as the feed. The

sorbed rare ear ths and transuranium elements were eluted from the res in

with 0.7 M HNO and the product solution evaporated t o approximately half - 3 volume and stored f o r fur ther processing i n the future.

PluLurllwu was recovered as an aqueous product solution tha t met radi-

at ion specifications. That recovered from the 24-rod l o t was deconta,mi.nat,ed

5 from f i ss ion products by a fac tor of 6.9 x 10 a able 2.1) while tha t from

5 the 8-rod l o t averaged 4.6 x 10 . The average composition of the plutonium

product from each l o t of fue l is given i n Table 2.2. Overall recoveries

totaled 93% a able 2.3) . Ion exchange sorption and scrubbing losses f o r

Table 2 .l. Process Losses and Decontamination Factors i n Recovery of Plutonium and Transuranium Elements from I r rad ia ted

Plutonium Aluminum Alloy

Losses ($ of , ~ e e d ) Gross Feed Ion Exchange De cont aminat ion

Prepaqat ion Sorption Scrubbing actor^

24-RO~ l o t

b'irst cyc! p

Second cycle

Overall

8-Rod l o t

F i r s t cycle

Second cycle

Average eve rall (both l o t s )

Plutonium Recovery

Transuranium Element Recovery

Process 9 -0 b 4 .O

b

a Ratio of gross 7 ' s per un i t weight i n f e ed t o thaL 111 product.

Based on raw waste t o the evaporator.

Table 2.2. Average Composition of Plutonium Product

Total - 1 Isotopic Assay (wt $) C3nc. CountingR.zte (cbsmin, HN03 S o-lr ce "eight ((gliter) Mass Mass Mass Mass Mass

(g) Gross a: Gross y (M) 238 2 39 240 241 242

.24-RO~ lo t 615 8.2 1 . 5 x 10 1.6 x 10 2-13 - - 9 29.3 51.2 12.8 6.5 \O

8-ltod l o t 30.4 0 .19 5.0 x 10' 1.1 x 10 2.8 1.09 0.35 15.27 7.8 75.49

a ;-.ieigkted average.

Table 2.3. Material Balance : Plutonium Recovery .

Amount of Plutonium (g) From:

24-~od Lot 8-Rod ~ o t

calculateda content of rods 576

Feed measurements 700

P i f f e r ~ n r e (gain) . ( 124)

Hecovery process

Specification product 655 30.4

Retained product 12 - - Losses t o waste disposal 8 14. 4C

Stored waste b

3

Total accounted fo r i n process

6'78 678

Difference across process 22 1.6

Process balance, recovery/feed, $ 96.9 96.2

a Based on reactor heat balances.

Later discharged t o waste storage a f t e r processing f o r transuranium element recovery.

C Includes 9.8 g i r revers ibly sorbed on f i r s t cycle res in bed.

the program averaged 2.9 and 4.8%, respectively, with the lower losses

( 1.0 and 1. '$I) being observed during processing the lower&urnup material

present i n the 24-rod l o t of fue l rods (plutonium content was 5 times t h a t

in Lhe ,8-rod lot,,which averaged 4.8 and 7.8%).

The transuranium-element product f ract ion of the composition given

i n Table 2.4 was decontaminated from alwninum by a fac tor of 70 and from

l igh t f i ss ion products by a factor of LO able 2.1). Overall recovery of

the transuranium elements from the raw plutoniuni-recovery waste eff luent

solution averaged 63% a able 2.5), taking into account the 37% loss tha t

occurred during feed preparation. Ion exchange sorption and scrubbing

losses averaged 9 and 476, respectively, but under the ideal conditions

specified in the flowsheet ( ~ e c . 3) averaged 0.2 and 0.4% coupled with an

overal l recovery of greater than 95%. High recoveries of the transuranium

elements depended on a thorough washing of these from the aluminum hydroxide

sol ids generated i n the i n i t i a l evaporation s tep and p a r t i a l l y removed upon

f i l t r a t i o n ; often, as mch as 35 t o 40$ of the transuranium elements present

i n the unclarified feed would be retained i n the bulky f i l t e r cake of these

sol ids . Low sorption and scrubbing losses i n the ion exchange s tep de-

pended on the s a l t strength of the feed and scrub solutions but were inde-

pendent of the acid deficiency over the range 0.5 t~ 1. .O - N. A t . st ill highcr

acid deficiencies, col loidal A ~ ( O H ) occurred i n the feed solution i n la rger 3

amounts and continued t o precipi ta te a f t e r f i l t r a t i o n and in ter fer ing with

the mechanical operation of the ion exchange system.

Radiation damage t o the anion exchange res in was apparent i n both

phases of the program. I n the plutonium-recovery phase, where the P e m . ~ t i t

Tab12 2.4. CoqosLt ion 02 Transuranium Element Prcaduct

G r o s s Coun~ing Rate - a1 Li HNO

3 (counts ruin-' KL-l) Pkclide ( d i s rnin-l m l - l )

Source (M_) (M) (y) 0:

6 Y Ce Cs 137 ,106 S r 144 90 Z r 95

7 iu

24-r0d 0.06 6.4 1.1 i . l r 1 e 9 2 . 2 ~ 1 0 ~ ~ 2 . 7 x 1 0 9 l . g x l 0 l 1 4 . 3 ~ 1 0 ~ l . 8 x 1 0 1 0 5 . 8 x 1 0 7 4 . 1 ~ 1 0 l o t

10 8-r0d 0.10 2.2 -- 4 . 1 - l o 9 8 . 8 : d : 1 i 9 2 . 1 x 1 0 9 5 . 5 ~ 1 0 ~ ~ 3 . 4 ~ 1 0 ' 1 . 9 ~ 1 0 - - l o t

Table 2.5. Material Balance : Transuranium Element Recovery

Alpha Counts Aluminum

(counts/min) ($ Raw ~ e e d ) (moles) ($ Raw ~ e e d )

Plutonium Recovery Waste Effluent:

14 Raw !.laste Feed ( in ) 7.3 x 10 100 1550 100

Prepared Feed (out)

(a) Discharged t o waste 1 .6 x 10 22 490 32 14

(b) Sent t o ion exchanger 4.6 x 10 14

6 3 - 1080 70 - - Total Prepared Feed 6.2 x 1014 8 5 1-570 102

Unaccounted f o r l o s s - - (gain)

Ion Exchange System Losses

Loading

Scrubbing

Piping leak

Tot.aJ. losses

Product Recovery

Recovered product 2 .3 x 10 14

131

Re coverable product 0.6 x 10 14 8 -

Total product 2.9 x 10 14

39

. Total accounted f o r i n bS2 i o n exchange cyctcm

l4 57

SK res in w a s used, a steady increase i n plutonium losses, coupled with an

increasing pressure drop across the res in bed, was observed. The res in was 3

8 replaced wi*h a f resh l o t a f t e r receiving an estimated dose of 4 x 10 rads

d u r b g one period of operation with the 24 fue l rods; losses and the pressure

drop across the r e s in bed then returned t o levels observed e a r l i e r . The

spent r e s in was black, soft, and had l o s t almost 35% of i t s "salt sp l i t t i ng"

7 capacity. A second l o t of exposed (4.6 x lQ rads) res in used during the

period of' recovery uf tlle 8-i'oa 1u.t of f u e l rode, xnci 2 ~ ~ 4 tr.) r:r:~~.>.t,ain

sorbed AL, RI, R,,,lo6, m106, m d Pd that r e s i s t e d all e f fo r t s a t e lut ion

even with the strongest reagents, including aqua regia.

The radiat ion damage t o the Dowex 1 X-10 res in used i n transurarrium

element recovery occurred ( qual i ta t ively ) a t a higher rate , requiring re-

newal of the r e s in a f t e r each l o t of feed had been processed. The appear-

ance of the spent r e s in was s imilar t o tha t observed i n the plutonium-

recovery phase of the program i n t h a t the resin was oi ly , tacky, je t black,

and retained t r aces of A l , RU"~, l3h1O6, and Pd. None of these could be

eluted even with aqua regia .

3. FLOWSHJ3ETS

Development and mod i f ica t ion of the recovery process continued through-

out the program, resul t ing i n a recommended chemical flowsheet tha t gave

good recovery of high-purity products. Oaly m l n u r o~odificationc t~ yre=

viously ex is t ing processirlg equipment were made t o adapt it t o the process

used, but the two ion-exchange columns were equipment specif ical ly in-

s t a l l ed f o r t h i s program.

3.1 Chemical Flowsheet

The chemical flowsheet developed during t h i s program and recommended

for future processing of highly irradiated Pu-A1 al loy fue ls t o recover

residual plutonium and other transuranium elements is shown i n Fig. 3.1.

3.2 Equipment Flowsheet

The equipment i n which the work was conducted is located i n c e l l 1,

Building 4507. Many of the components served dual purposes during d i f fe rent

phases of the program. A l l equipment was made of s ta in less s tee l ; process

piping was of an all-welded design t o minimize corrosion and leakage. The

arrangement of the equipment as used i n the plutonium-recovery phase i s .. ,

. ,

shown schematically in Fig. 3.2 and, as used i n transuranium element re- 2

covery, i n Fig. 3.3. . . . . < .

4. PROCEDURE . , , " ! -,

i

The procedure followed i n recovering plutonium and the other t rans- 1

+ ,. uranium elcmcnts included reed preparation, ion exchange sorption of plu-

tonium and i ts recovery, evaporation of the ion exchange eff luent and i ts

adjustment, ion exchange sorption of the transuranium elements (and r a m

earths) and t h e i r recovery, and f i n a l concentration of the transuranium

product f o r storage. Short descriptions of t he f'uel, i t s history, and the

shipping cask are included as an introductory background preceding the

detailed procedures which follow.

4.1 Feed Preparation

The preparation of feed f o r the ion exchange system included charging

the fue l rods ts the dissc~lver, d i ~ ~ o l u t i o n of the pluLunlum-aluminum al loy

CAKE WASH SOLIDS OXIDANT REMOVAL

UNCLASSIFIED HNO3 2 M NO NO^ 2 . 5 M ORNL Dwp64~1110 F-3 0 . 0 5 M HN03 I .OM

(occasional) ADJUSTMENT HN03 13M

I FUEL ROD BATCH DISSOLVER RAW DISSOLVER SOLUTION ION-EXCHANGE A I I . O M S i 0 . 0 0 2 M

Fe I mole 2. Digest -24 hr. Fe 0.01 M HN03 3 .6 N i 0.1 mole

--t 3. Gelatin strike.

- N i 0.001 M P u 0.1-0.

4. Adjust valence. H g . 0 . 0 5 M g/ l i ter N i 0 . 0 0 0 6 5 M F- O . 0 3 M F.P.00.7glliter H g 0.03M

SPENT CAKE

FROM SECOND CYCLE SCRUB WASTE

HN03 7 .3M Pu 0. I-.3g/Iiter

I * EPa -0.6g11iter

Fig. 3.1. Chemical Flowsheet For Recovery of Plutonium and Trans- uranium El-einent s From I r rad ia ted Plutonium-Aluminum Alloy f i e 1 Rods.

Fig. 3.2. Equipment Plowsheet fo r Recovery of Plutonium From 1 rradiated Plutonium-Aluminura Alloy.

Fig. 3.3. EquLpnent Flmsl.leet f o r Transuranium E l e m n t Recovery From I r r a d i a t e d Pldtonium Recovery W~stes.

and the cladding, plutonium valence ad justrnent , clarif icat ion and f ina l

adjustment of the acid concentration.

4.1.1 Description and History of the Fuel

The f i e1 received for processing consisted of two lo t s of 0.94-in-dim

by 5-ft-long aluminum-clad rods. One lo t , comprising 24 rods, had been

irradiated t o about 87% burnup and had decayed for about 205 days; the other

lo t , eight rods, had been irradiated t o greater than gg$ burnup and decayed

for about 100 days.

The 5-ft rods were half-sections of a longer rod that had been made

from two lengths of 6$ Pu-A1 alloy rod co-extruded in an aluminum cladding

such that at each end of the canned co-extruded rod a 6-in-long aluminum

section w a s formed. Two sections of alloy rod were joined (forming an

aluminum spacer section plug) and the assembly irradiated. Following t h i s

and a f t e r the decay period, the 10-ft rods were severed within the aluminum

spacer sect ion and shipped.

4.1.2 Shipping Cask Description and Handliq

Shipments were made i n a 1.0-ton cask ( ~ i g . 4.1) providing 10-in of

lead shielding. The cask, which also served as a dissolver-charging com-

ponent, was equipped with a 24-slot magazine, the s lo t s being i n three

concentric rows, that was mounted i n the cavity and from which the f'uel

rods could be dropped individually into the dissolver charging chute.

The cavity of the cask was equipped for water cooling and could be made

e i ther liquid and gas t igh t o r vented into a f lash chamber v ia a re l ief

valve (se t a t 17 psig) and a glass wool f i l t e r . The temperature of the

UNCLASSIFIED ORNL Dwg 64-1112

MAGAZINE PLUG COVER PLATE INDEX LEVER (Used in transit \Removed durina -INDEX WHEEL J transit

VENT RECEPTACLE- i MAGAZINE

DRAWER-SEALING BOLT (Removed before plocing on oedestol 1

DRAWER HOUSING (Part of pedestal)

CAVITY DRAl RECEPTACLE

GASKET

OPERATING SCREW CARRIER DRAWER '

UlJtHATING SCRl EXTENSION HANDLE

Fig. 4.1. Shipping Cask f o r Irradiated Plutonium-Aluminum Alloy Fuel i n Charging Position on Slug Chute Pedestal.

cask cavity could be measured with the thermocouple mounted i n the shield-

ing. During shipment of the fuel rods by t r a i l e r truck, the cask rested,

drawer-end d m , i n a cradle that kept the longitudind axis of the cask

a t 1 5 O t o the horizontd. Water, which f i l l ed about 65$ of the cavity,

par t ia l ly covered the rods for cooling. Upon arrival here and a f te r in-

spection, the flash chamber and the relief-valve assemblies were removed,

and the cask was moved t o the containment area above the cel ls fo r tempera-

ture observation. After standing 24 hr, the surface temperature of the

loaded cask was 7 and 12*, respectively, above ambient temperature (26 '~ )

when loaded with the 24- and 8-rod shipmnts. The cavity water w a s then

drained (a f te r sampling) t o a ce l l vessel and transferred t o the inter-

mediate-level-waste storage system. The cover plates were removed from

the drawer and magazine-operating plug, a f t e r which the bolts holding the

top of the drawer against the gasket in the cavity were removed. The cask

was then moved t o the slug-chute pedestal ahd positioned t o align the dis-

charge port; of the cask with the charging port in the pedestal. A t ight

seal between the cask and the pedestal was obtained by a soft gasket

mounted in a groove in the pedestal plate that pressed against the smooth

bottom sides of the cask. The drawer-operating screw in the cask was then

engaged with the extension handle, which extends through the housing on the

pedestal that encloses the cask drawer when in the withdrawn position,

a f te r which the discharge of fie1 from the cask could then begin.

4.1.3 Charging the Fuel t o the Dissolver

After the cask had been positioned on the slug-chute pedestal and the

indexing mechanisms attached, the index arm w a s moved t o place the desired

magazine s l o t over the discharge port i n the cask drawer. The pedestal

drawer w a s then withdrawn by turning the operating screw, moving the cask

drawer t o point where the discharge port was digned with the desired

magazine s lo t . The fuel rod dropped by gravity t o the charging chute and

into the dissolver. Both drawers were closed a f t e r the dissolver w a s

charged.

The i n i t i a l charge of f'uel rods t o the dissolver w a s 6 rods. There-

af'ter, one rod was charged per completed dissolving t o maintain a suitable

undissolved-metal inventory i n the dissolver and obtain a practical dis-

solution rate.

4.1.4 Dissolution of the Fuel

Dissolution of the f i e1 rods was conducted in a dissoLwr ( ~ i g . 4.2)

equipped with a dam-draft condenser and two sets of coils; the upper coil

w a s for cooling, while the l o w e r one waa upp plied with e i ther steam or

cold water. A l a t e r revision of the piping, because of a leak in the

lower coil, permitted the use of e i ther set of coils for heating or cooling

( ~ e c . 4.3.3.1). The off-gas leaving the condenser flowed t o a 1-in Raschig-

ring-filled scrubber into the top of which 1 M NaOH was metered, The non- - condensable f'raction of the off-gas passed from the scrubber via a f l a w

arres ter t o the off-gas header serving the fac i l i ty . The off-gas pressure

i n the dissolver was controlled between 0 and 5 in. of water gage below ce l l

pressure. Control of the steam f l o w rate t o the coils w a s maintained by a

temperature controller that monitored the vessel temperature.

Condensate f r o m the condenser could be returned t o the dissolver o r

diverted t o a collection tank. The i n i t i a l dissolving in a program began

Fig. 4 -2. Dissolver-Evaporator System Showing Eke1 Rod Inventory a t Equilibrium.

with the charging of s ix Rzel rods t o the dissolver, which contained 25

l i t e r s of 6 M HNO Upon completion of the charging operat ion, the tem- - 3' perature was raised t o 60°c, at which time the addition of 40 l i t e r s of

6 M HNO containing 0.082 M w a s begun at 6.5 l i t e r s /h r and continued - 3 - while raising the terqperature t o 100°c. Dissolution of the aluminum end

pieces began at 95 t o 100°C and became rapid enough with the six-rod charge

t o require cooling water in the coils t o control the reaction rate. After

a l l of the second lo t of the dissolvent had been added and the i n i t i a l

rapid reaction bad siihsided, thc conteuLts of the dissolver were digested

at 100 t o 102% for 24 t o 32 hl; o r un t i l the acid concentrat ion decreased

t o about 2.5 M HNO a t which concentrat ion the dissolution rate w a s low. - 3' The first lo t of dissolver solution of a new series usually contained an

abnormally high concentration of aluminum along with a l o w concentration

of plutonium; these solutions were transferred from the dissolver ear l ier

than usual because dissolution of the aluminum end pieces consumed the

bulk of the HNO before appreciable exposure an8 dissolution of the Pu-A1 3 alluy coulcl take place. In such cases, a f te r removal of the first dissolver

solution, the second dissolving of a series began with the addition of 65

l i t e r s of 6.0 M WTO -0.05 M Hg2+--0.03 M F- t o the dissolver over a 2-hr - 3 - - period while keeping the temperature as law as possible with cooling water.

The temperature w a s then carefully raised t o 95 t o 100°C t o s t a r t the

dissolution. Afier the vigorous reaction began, cooling was required for

0.5 t o 1 hr, a f t e r which heating was resumed and the temperature kept a t

100 t o 102 for 24 t o 36 hr, or un t i l the metal equivalent t o one rod had

dissolved.

During all subsequent dissolvings, one fue l rod was charged. t o the

dissolver containing the residue of all mds previously added i n the se r i e s .

2+ Sixty-five l i t e r s of 6 M HNO --0.05 M Hg --0.03 M F- was then metered t o - 3 - - the dissolver a t 30 t o 40 l i t e r s / h r while controll ing the temperature with

cooling water a t 95 t o 100°C. After an i n i t i a l reaction surge tha t re-

quired cooling water, the reaction ra te decreased; thereaf ter , the. eontents

were kept at 100 t o 102OC f o r 24 t o 30 h r u n t i l the metal equivalent t o

one rod had dissolved.

After the l a s t rod of a ser ies had been added, dissolvings were con-

tinued with fresh dissolvent as before, u n t i l no s ignif icant amount of

plutonium appeared i n the dissolver solution.

Vapor evolved during dissolution passed t o a downdraft condenser

where the condensable components were condensed and returned t o the d i s -

solver. The noncondensable fract ion of the off-gas passed from a vapor-

l iquid separator below the condenser t o a 6-in-diam by 40-in-long caust ic

scrubber packed with 0.5-in. Raschig rings. The acid components of the

off-gas were sorbed i n 1 M - NaOH metered a t 3 t o 4 l i t e r s / h r t o the caustic

scrubber. Spent caustic l iquor from the scrubber w a s continuously d i s -

charged t o the intermediate-level waste storage system without sampling.

The residual, nonc~ndensahl~, ac id - fme off-gas passeil f r u m the sembber v i a

a flame ar res tor t o the off-gas header.

A controlled part,i.al vacuum ( 3 t o 4 i n ., wa te r &age) was mintained i n

the dissolver a t all times. A continuous' air purge of 1 .5 cfm intmduced

t o the slug chute kept it f ree of vapor and diluted any exploeive

radiolyt ic gases generated i n the dissnl.ver t o safe conccntratiaus.

4.1.5 Dissolution&ate Studies

Only 20 in. of the 5-ft length of the fue l rods was exposed t o the

dissolvent at any time. Previous dissolution studies with inactive

1- in-d im aluminum rods had showed tha t dissolution progressed along the

submerged sect ion of the rod u n t i l ' t h e thinned section collapsed under

i t s own weight, exposing a new section. The collapses were periodic

r a the r than continuous . The attacked section of the rod assumed the shape

of an i a i c l e ,,

The total Ci~ssul~Lluli of a singlc f i e 1 rod Tares cnndi~c ted tg determine

the aluminum-to-plutonium ra t ios and t o obtain specif ic dissolution-rate

data . I n t h i s study, the rod was exposed t o a single l o t of dissolvent

u n t i l about 50% of the aluminum had dissolved, a t which time the spenr

dissolvent was replaced with a f resh l o t and the dissolution resumed and

continued u n t i l no fur ther aluminum o r plutonium was dissolved. The dis-

solver solut ion was sampled periodically and analyzed f o r plutonium and

aluminum, from which data the quantity of dissolved materials were deter-

mined and plot ted as a function of time (Fig. 4.3). Kelatlng thest: dttta

t o the exposed surface of the rod during the t i m e interval gave an average

-1 -2 overal l spec i f ic dissolution r a t e of about 0.45 mg min cm f o r aluminum

-1 -2 and 0.005 mg min cm f o r plutonium. The aluminum-to-plutonium ra t io

over the range of t h i s run was 120, compared with the theore t ica l 64, indi-

cating t h a t a l l of the plutonium probably had not dissolved.

Similar data were obtained during two normal production (equilibrium)

runs, one dissolving material from the 8 f i burnup l o t of fue l and the other

from the 99$ burnup l o t . I n these, a fresh rod was charged t o the dissolver

UNCLASSIFIED ORNL Dwg 6 4 - 2 1 1

DISSOLUTION TIME. (hr),

S ing le - Rod Chorge

D ISSOLUTION T IME ( h r )

S ix -Rod Chorge

Fig. 4.3, Ty-picnl is solution R a t e o .

already containing residual metal from the previous dissolvings of the

ser ies . Data from these two runs when related t o the surface exposed t o

the dissolvent as a function of t i n z e gave specif ic dissolution ra tes f o r

-1 -2 the al loy of 0.8 and 0.7 mg min cm , respectively. The dissolution

r a t e of the fue l dissolved i n t h i s program, although low, agrees well with

t h a t predicted from similar data plotted i n Fig. 4.4.

4.1.6 Feed Clar i f ica t ion and Adjustment

After the desired amount of plutonium had been dissolved, optionally,

suf f ic ien t 1% ge la t in solution was added t o obtain a 3,0-ppm concentration

of ge la t in i n the dissolver solution, a f t e r which the mixture was digested

f o r 1 h r at 80°c t o coagulate s i l i c a . The plutonium valence was reduced

t o three by the addition of a reducer, 50% ferrous sulfamate solution

(2.9 N), a f t e r which the solution was f i l t e r e d through a 12-in-diam by

1-in-deep bed of Ottawa sand tha t had previously been coated with f i l t e r

a id . The valence of the plutonium i n the f i l t r a t e vas adjusted t o four by

oxidizing. with 2.5 M - NaITO, added t o it, followed by digestion a t 50 t o C

60°c for 0.5 h r . On occasions, the f i n a l valence adjustment was completed

p r io r t o f i l t r a t i o n , without noticeable e f fec t on the process. It was

important, however, t o allow suf f ic ien t time fo r the nitrous gases evolved

i n the reduction of the n i t r i t e t o be d.ischarged from the solution pr ior

t o passing it t o the res in beds; t h i s minimized gas binding and the forma-

t i o n of gas-f i l led voids within the bed. Such conditions promoted

channeling of the solution through the bed, with consequently higher

plutonium iosses.

UNCLASSIFIED ,

ORN L-DWG 64-276 R I '

8 ORNL

@ SR

I CR

~u~~~ BURNUP (%)

Fig. 4.4 Plant n i s s o l u t i o n Data. Pd-Al a l l o y , 6 .O M HNO - 0.05 M - Hg2+ - 0.03 M F-. D a t a from Savannah River (1)) Chalk ~ i v e r , &;ld O a k R'idge ~at . ionZ1 Laboratory (2 , 3, 4, 5 , 6) .

The ac id i ty of the feed solution was adjusted t o 7 M HNO p r io r t o - 3 passing the feed t o the anion exchange resin. This was done i n two ways:

I n the f i r s t four runs, the Teed stream was continuously metered t o the

resin-column head pot, where it as mixz6 .:~i-l;h a filetei-ed s t reaa of 13 M -

HNO such t h a t the net acid concentration was 7 M RNO This procedure 3 - 3'

w a s abandoned i n favor of a batch adjustment because of gas evolution

t h a t occurred when the ac id i ty w a s increased and it developed tha t the

holdup time i n the head pot was insuff ic ient f o r the ac id i f ied stream t o

d ischarge i t s e l f of the evolved gases.

2 PI-mtoniufi Re cove ry

Higlily purified plutonium vas recovered from the adjusted feed

solution in a two-stage process by sorption on chloride-free anion ex-

change resin. Important s teps i n each stage inc&i?ded preparation a ~ d

l~andlinp; of tlie resin, charging resin t o the system, loading plutonium

on the resin, scrubbing and e lu t ion of the sorbed plutonil.u?i, and storing

the finished product.

2 1 Ion Exchange - Two stages of ion exchange purif icat ion were requij ed t o obtain a

plutoniwn product t h a t met radiation specifications. The preparation of

chloride-free ion exchange r e s in and i ts charging t o the coimns followed

by conversion t o the n i t r a t e form were necessary preliminary steps i n the

operation. Occasional removal of degraded res in from the columns by up-

f l o w displacement with 2 M - &(NO ) was a technique developed t o perniit 3 3

changing out the res in remotely.

4.2.1.1 Resin Preparation. - The ' res in used i n plutonium recovery

consisted of 20- t o 40-mesh Permutit SK. Received i n the chloride form,

it w a s converted t o the n i t r a t e p r io r t o use by digestion i n 7 M HNO at - 3 room temperature f o r 2 t o 3 hr, a f t e r which the spent acid was drained

away. The t reated res in w a s washed with demineralized water u n t i l f ree of

chlorides, a f t e r which it was stored as a res in s lur ry u n t i l charged' t o

the column.

4.2.1.2 Charging Resin t o the Column. - The chloride-free res in was

charged t o the column as a very d i lu t e slurry, d i lu t ion water being added

t o keep the res in pa r t i c l e s swept along the piping toward the column during

. ' the operation. The s lur ry was added t o a f'unnel f i t t e d t o the resin-charg-

ing l i n e and was continuously s t i r r e d t o keep the res in i n suspension. A

small stream of water was played over the surface of the res in i n the con-

ta iner , sluicing a , l i gh t s lur ry of res in t o the addition funnel, which was

kept half fill of l iquid at all times. If the s lur ry i n the f'unnel became

too heavy, addition of res in t o it was halted. The diluted s lur ry passed

t o the top of the heated standpipe and f e l l through it into , the column.

Excess water flowed from the bottom of the column through a sealing l eg

("Jack leg") and diversion valves t o the waste col lect ion tanks.

After 7 . 5 l i t e r s ( se t t l ed volume) of res in had been charged ro the

column, the addition f'unnel and l ines were flushed thoroughly with water

and the f'unnel assembly removed and the charging l i n e capped.

Before placing the column i n service, pressure-drop measurements across

the r e s in bed were determined as a f'unction of flow ra te , both with water

m d with 7 M HNO These data when compared with s imilar r e su l t s using - 3'

geometrically ident ica l glass columns were usef'ul i n determining whether ,

t h e resin-bed depth w a s sa t i s fac tory and whether the column had been over-

f i l l e d o r not. The 7 M HITO a l so served t o precondition the res in t o re- - 3 ceive feed. I n pract ice it w a s found tha t i n removing spent res in from the

column ( ~ e c . 4.2 .l. 3) 0.5 t o 1 l i t e r of res in remained. When t h i s happened,

the addition of 8 l i t e r s of f resh res in would es tab l i sh a bed tha t com-

p le t e ly f i l l e d the column and a lso established a small-diameter bed extend-

i n g up in to the standpipe.

4.2.1.3 Removing Spent Resin From the Column. - Spent resin, degrad-

ed from chemical and radiation exposure, w a s removed from the column by

pumping 16 l i t e r s of 2 M - A ~ ( N O ) t o the bottom of the column below the 3 3

r e s i n bed v i a the seal-leg assembly a t 15 t o 20 l i t e r s l h r . The r e s in bed

w a s displaced upward (more o r l e s s in tac t ) t o the airlift intake located

about a t t he midpoint of the standpipe. A stream of f lush water w a s con-

tinuously added t o the top of the standpipe a t 30 l i t e r s l h r t o keep any

r e s in flushed down from the upper section and a lso t o d i lu t e the resin-

s lu r ry as it passed t o the air l if t . The a i r l i f i discharged t o a waste-

col lect ion tank, which was continuously sparged with a i r t o keep the resin

suspended. Additional d i lu t ion water w a s a lso continuously added t o the

co l lec t ion tank. The d i lu t e spent -res i n s lur ry w a s continuously transferred

from the col lect ion tank by steam je t t o the intermediate waste system, the

waste header a l so being continuously flushed with a stream of water t o help

keep it swept f ree of resin. Except f o r those instances where sol ids had

collected on top of the res in bed during loading operations, effect ively

reducing t h e throughput ra te and increasing the exposure t o radiation and

consequent degradation, the technique w a s simple and worked well.

Upon completion of the 2 M - A ~ ( N O ) displacement of spent resin, the 3 3

column was flushed with water, and a second displacement with 2 M - A ~ ( N O ) 3 3 was made t o e f f ec t as complete a removal of spent resin as possible.

4.2.1.4 Changing Types of Resin. - In studies with a resin column

mockup, it was found tha t 7 t o 7.5 l i t e r s of an 8 - l i t e r charge was removed

i n two displacements . Additional displacements removed insignificant

quantit ies of spent resin.

To ef fec t the complete removal of one type of res in pr ior t o charging

another type, and t o obtain the minimum cross contaminat ion between the two,

it was necessary t o pa r t i a l ly recharge the column with about 3 t o 4 l i t e r s

of the new type of resin, then f lu id ize the bed with short bursts of water

t o the bottom of the bed t o mix it thoroughly and then conduct a displace-

ment procedure, removing the mixed resin. The p a r t i a l charging-mixing-dis-

placement procedure would be repeated several times u n t i l only insignificant

amounts of the or iginal res in remained, a f t e r which a f u l l 7 .5- l i ter charge

of the new type of resin would be made.

4.2.2 F i r s t Stage Operation

he i n i t i a l stage of purif icat ion began with the sorption of plutonium

from the feed followed by scrubbing the loaded b e d and elution of t h e pm=

duct. Pr ior t o reloading the next batch of feed, the res in was regenerated

with 7 M HNO - 3'

4.2.2.1 Loading. - The adjusted feed solution was metered a t

-1. -2 5 ml min cm (20 ~ i t e r s / h r ) t o the res in column via the headpot. The

ra te was governed by the pressure drop across the resin bed and was s teadi ly

reduced as the program advanced. The solution flowed by gravity t o the

0.75-in-dim by 93-in-long standpipe, which was maintained a t 60°c by c i r -

cu la t ing warm water through the jacket. A supplementary e l e c t r i c a l l y

heated sect ion w a s mounted below, though it was rarely used. The required

pressure head t o drive solution through the bin-diam by 40-in-deep (7.5

l i t e r s ) bed of Permutit SK (20 t o 50 mesh) res in was supplied by the

height of the solut ion i n the standpipe. The res in bed was maintained a t

60°c by c i rcu la t ing hot water i n the jacket of the column. The res in bed

w a s su~gorted. on a 100-mesh screen reinforced with two layers of 18-mesh

screen, while the top of the res in bed was f ree .

Ion exchange loading wastes were collected i n a ve s s e l from which

samples of the contents were removed f o r analyses. Depending on the plu-

tonium content, the loading wastes were discharged t o a 900-gal storage

tank for l a t e r processing t o recover the transuranium elements ( ~ e c . 4.3))

o r t ransferred t o a salvage storage vessel f o r reworking and plutonium

recovery.

4.2.2.2 Scrubbing the Loaded Resin Bed. - Sorbed f i s s ion products

and residual aluminum were scrubbed from the res in bed with 7 M HNO - 3 -2 metered t o it at 2.5 m l m i d 1 cm u n t i l 16 l i t e r s ( 2 bed volumes) of

solution had passed. After three t o f ive l o t s of feed had passed through

the res in bed (with the 16- l i t e r scrub following each), the 7 M HNO scrub - 3

w a s extended t o 80 l i t e r s t o t a l volume f o r more extensive removal of

residual f i s s ion products and aluminum. After sampling the spent scrub

solutions i n the col lect ion tank, these were discharged t o the intermediate-

l eve l waste system o r stored f o r salvage, depending on the plutonium content.

4.2.2.3 Elution. - The plutoniui sorbed on the res in bed was eluted

-1 -2 with 1.6 l i t e r s of 0.7 N HNO metered t o the column a t 0.5 m l min cm , - 3 the temperature of the resin bed being maintained a t 30°C.

4.2.2.4 Resin Regeneration. - Following an elution, the res in Iras

regenerated a t 60°c by metering 16 l i t e r s of 7.0 M HNO - 4 . 0 1 M F---o.OI M i - 3 - -

-1 -2 A 1 t o the column a t 5 rill rnin cm . Spent regenerating solution, usually

low i n plutonium, was discharged. t o the intermediate-level waste system

o r stored fo r salvage, depending on i t s plutonium content.

Whenever the resin bed was t o be idle f o r any length of time, it was

eluted and allowed t o stand i n 0.7 M HNO with which chemical a t tack was - 3' ' . negligible.

4.2.3 Second Stage

4.2.3.1 Feed Adjustment and Loading. - The plutonium product f ract ions

eluted from the f i r s t ~ s t a g e res in bed were accumulated u n t i l a sui table

volume was on harid. The cornposited f i r s t cycle product was adjusted, f i r s t

'

with ferrous sulfamate a t 30 t o bOc, then with NaN02, and digested f o r 1 h r

t o cnoure tha t the plutonium vslenoe was four.

After the digestion a t ambient temperature t o allow the gases evolved

t o pass from the solution, the ac id i ty of the feed was adjusted t o 7 M - HNO and the adjusted solution was metered t o a second ion exchange system

3' a t conditions ident ical t o those i n loading the f i rs t -s tage res in bed. The

second-stage column was ident ical t o the f i r s t except tha t it w a s not

equipped with a head pot.

4.2.3.2 Scrubbing the Loaded Resin Bed. - Residual contaminants

sorbed on the r e s in were removed from the resin with 80 l i t e r s of 7 M HJW - - 3

-1 -2 0 . O 1 M - F- metered t o the system, held a t 60°c, a t 2.5 m l m i l l cm . Scrubbing was continued with 24 l i t e r s more of fluoride-free 7 K HNO under - 3 the same conditions t o displace fluorides from the resin bed. These scrub

solu-t ions (which usually contained plutonium a t concentrat ions justifying

recovery), a f t e r sampling and determining the plutoniuni content, were

stored f o r l a t e r use i n preparing dissolvent.

).1..2.3.3 Elution and Recovery. - Finiahcd plutonium product was re-

covered. from the res in by elut ion with 16 l i t e r s of 0.7 M - HNO, metered t o

-1 -2 the column maintained a t 3 0 ' ~ at 0.5 m l min cm . The f i r s t 2 t o Lc l i t e r s

of e f f luent displaced from the res in bed was routed t o the scrubbing waste

t o obtain highest plutonium concentrat ion i n the product fraction; no

appreciable loss was incurred by t h i s pract ice. The finished product

solution w a s collected in one of the smaller (35 - l i t e r ) c e l l vessels where

it was thoroughly mixed by a i r sparging and samples removed fo r analysis.

4.2.3.4 Regeneration of the Resin. - The res in bed was allowed t o

stand in 0.7 M HNO u n t i l the next loading. Before loading again, the resin - 3 was regenerated a t 60°c by passing 16 l i t e r s of 7.0 M HNO - 4 . 0 1 M F--- - 3 -

-2 0 . O 1 M - A 1 through the bed at 5 ml min-l cm . The spent regenerant solution

was sent t o the dissolver f o r dissolvent preparation o r discharged t o

waste, depending on the plutonium content.

4.2.4 Handling of the Plutonium Product

The finished plutonium product solution, a f t e r preliminary resu l t s of

analyses were known, was withdrawn in to tared polyethylene bo t t l e s which,

a f te r f i l l i n g with 10 l i t e r s of solution and weighing, were bagged and

placed i n shipping containers. The bo t t l e s were 5 in . i n inside diameter,

with a minimum w a l l thickness of 0.125 in; h e nominal capacity w a s 12

l i t e r s . The screwed bo t t l e caps were f i t t e d with a 0.125-in disk gasket

of neoprene rubber backed with a 0.25-in layer of glass wool f i t t e d under

the cap. In storage of the product recovered during the &-rod program,

the cap assembly w a s pierced with a hypodermic needle t o provide a vent

f o r the escape of radiolyt ic gas. When it was found l a t e r t h a t small

quant i t ies- of the solution escaped v i a the puncture-vents, each cap assembly

was f i t t e d with a posit ive vent consisting of a hypodermic needle (pro-

truding through the cap, glass wool, and gasket into the gas space above

the solution) t o which was attached a section of p l a s t i c hose leading t o

a 0.5-in-dim by 3-in-long glass wool f i l t e r mounted on the body of the

bo t t l e .

The shipping container consisted of a 6-in-diam by 60-in-long s tain-

l e s s s t e e l cylinder f i t t e d with a gas-tight flanged cover. Mounted on the

cover was a valved vent l ine f o r attachment t o an off-gas system f o r re-

leasing radiolyt ic gases generated by the plutonium solution during storage

and shipment. The s ta in less s t e e l cylinder was permanently mounted i n a

2 2-ft by 5.5-ft-high plywood box tha t ensured a safe geometry f o r any array

of containers during storage o r shipment.

Immediately p r io r t o shipment, and periodically during storage, the

radiolyt ic gases generated i n the product solutions were released t o main-

t a i n a reasonably low pressure within the s t e e l cylinder.

4.3 Transuranium Element Recovery

The transuranium elements are not sorbed on anion res in at the high

acid concent rat ion charac ter i s t ic of the plutonium recovery process. A t

zero t o 0.2 M HNO i n a medium 7 t o 8 M in t o t a l n i t r a t e , however, the - 3 - ra re ear ths and transuranium elements are sorbed strongly. The t rans-

uranium element recovery process is based on t h i s principle, so tha t the

i n i t i a l s t ep was t o remove the free acid from the plutonium loading waste

eff luent by evaporation while increasing the aluminum concentrat ion.

4.3.1 Feed Preparation

The preparation of feed consisted 5.n evaporating the f i r s t - s tage plu-

tonium-loading waste eff luent t o about 1 4 0 ~ ~ and then adjusting the acid

deficiency and aluminum concent r a t ions, a f t e r which the aluminum hydroxide

formed during the evaporation was removed by f i l t r a t i o n .

4.3.1.1 Evaporation of the Waste Solution. - The plutonium-loading

waste eff luent from the f i r s t plutonium recovery stage tha t had been, accu-

mulated and stored during tha t phase of the program ( ~ e c . 4.2.2 . l ) was

evaporated t o 1 4 0 ~ ~ i n the dissolver vessel ( ~ i g . 4.2), the piping of which

was arranged such t h a t the condensate could be returned t o the vessel or

diverted t o a condensate collection tank. A s or iginal ly instal led, the

upper set of co i l s i n the vessel were piped f o r cooling water only and the

lower set f o r e i t h e r cooling water o r steam. During the runs, however, the

lower co i l began t o leak, making it impossible t o conduct the evaporation

t o 1 4 0 ~ ~ with t h a t c o i l only. As a resu l t , both co i l s were modified so

t h a t each could be used for heating o r cooling, with the added provision

t ha t a 15-psig posit ive a i r pressure would automatically be maintained on

the co i l a t a l l times .when steam o r water t o the co i l s w a s turned o f f . A

second mod i f ica t ion t o the piping. included re routing the coil-discharge

l ine from the proces's drain (normally nonradioactive) t o a waste-collect ion

tank from which samples could be withdrawn. This permitted evaluation of

the leakage ra te of material from the evaporator into the co i l and ensured

tha t no radioactive material found i t s way into the process drain.

The evaporation of the plutonium-loading waste eff luent solution was

conducted by t ransferr ing 65 l i t e r s t o the dissolver and evaporating the

mixture t o small volume a t normal pressure. Afterwards, 10- t o 30-l i ter

increments of f resh solution were transferred t o the dissolver as the

evaporation progressed, u n t i l 340 t o 370 l i t e r s had been added, the volume

of solution i n the evaporator being held between 30 t o 65 l i t e r s ; the mini-

nun volume under evaporation was held t o 30 l i t e r s t o minimize exposure of

the heating co i l s t o the vapor. After the l a s t increment had been added,

the evaporation was continued un t i l the l iquid temperature reached l q O c ,

a t which time the volume of concentrate was about 30 l i te rgoand a port ion

of the c o i l was exposed t o the vapor; No d i f f i cu l ty i n rais ing the f i n a l

tempe.rature t o 1 4 0 ~ ~ was encountered u n t i l the leak occurred i n the lower

coi l , a f t e r which suff ic ient steam 'Leakage into the vessel resulted i n

pressurizing it. After modifying the piping, mentioned above, the use of

steam i n both co i l s was required t o reach l b O c , and during most of the

last stage of the evaporation the upper co i l was exposed t o the vapor.

The marked increase i n the amount of sol ids i n the concentrate observed

a f t e r the modifications t o the co i l s was made w a s a t t r ibuted t o the drying

of so l ids on the heated surfaces exposed t o the vapor during the l a t t e r

stages of the evaporation, coupled with the excessive steam str ipping of

the HNO from the concentrate t h a t w a s in equilibrium with the dissolved 3

After the concentrated solution had been taken t o 140O~, it was

di luted with water t o a ref lux temperature of llg°C (approximately 3.2 t o

3.3 - M A l ) . The acid deficiency was about l ' t o 1.2 - N before the modifi-

cation t o the co i l s were made, but increased afterward t o 1 .5 t o 1.8 N. -

While an acid deficiency of 0.5 t o 1.0 - N was desirable t o prevent cry-

s t a l l i z a t i o n of the A ~ ( N O ) at the higher acid deficiency, a colloidal 3 3'

suspension of A ~ ( O H ) was present tha t was not removed completely by f i l - 3

t rat ion and cont inued t o precipi ta te slowly during subsequent ion exchange

operat ions, resul t ing i n plugging the res in bed. Sorption 'losses, however,

apparently were not increased s ignif icant ly a t the higher acid deficiency.

4.3.1.2 Clar i f icat ion. - The diluted solution was sampled with

d i f f i c u l t y because of the high sol ids content; during about 50% of the

program the evaporator sampler was inoperative because of plugging with

so l ids . The sol ids present i n the diluted solution were only pa r t ly re-

moved by f i l t r a t i o n a t 60 t o 80°c through a 12-in-diam by 1-in-deep bed of

Ottawa sand precoated with f i l t e r a id. F i l t r a t ion was slow and incomplete.

Apparently precipi ta t ion of f ine ly divided A ~ ( O H ) continued a f t e r the 3

solution had been f i l t e r e d . Solutions which were adjusted t o approximately

0 .5 N - acid deficient f i l t e r e d more readily than those of 1 t o 1.5 N acid - deficiency .

After f i l t r a t i o n and sampling, the f i l t r a t e was transferred by steam

j e t t o a feed hold tank. The f i l t e r cake a f t e r draining as thoroughly as

possible was fluidized by backwashing with water and discharged from the

f i l t e r t o the intermediate-level waste storage system.

Material balances completed about the f i l t e r i n ear ly runs showed t h a t

25 t o of the aluminum and transuranium elements were retained i n the

f i l t e r . These materials were recoverable t o greater than 95% by washing

the cake with two 20- l i te r volumes of 0.5 M HNO and returning the spent - 3 wash t o the succeeding l o t of waste t o be evaporated. Owing t o the lack of

intermediate storage vessels fo r holding the spent wash solution f o r l a t e r

salvage and piping l imitat ions of the system, only two demonstrations of

t h i s procedure were completed.

Routinely, f i n a l adjustment of the f i l t r a t e was not necessary since I

the A ~ ( N O ) concentrat ion i n the feed was desired a t the highest leve l 3 3

obtainable (2.5 M - minimum) . It was d i f f i c u l t t o achieve t h i s concentration

without the A~(NO ) crystal l iz ing in both the f i l t e r and col lect ion tanks 3 3

as the solution cooled t o below 6 0 O c (none of the process vessels other

than the evaporator was equipped f o r heating) because of the d i lu t ion with

condensed steam t h a t operated the t r ans fe r j e t s during the two t r ans fe r

operations tha t were required.

4.3.2 Ion Exchange

Ion exchange operations began with converting the res in t o the chloride-

TI-e n l t r a t e form, and charging it t o the column, &er.which the res in was

then conditioned with 8 M - LiNO Recovery of the transuranium elements 3'

from the feed was then begun followed. by scrubbing and elut ion. The spent

reein waa removed at the end of eadl IUL and replaced with a f resh l o t .

4.3.2.1 Resin Preparation. - The res in used i n the recovery of the

transuranium elements was 50- t o 100-mesh Dowex 1 X10. Before use, the

r e s in was freed of chlorides by repeated digestions i n an equal volume of

8 HNO at 25 t o 30°c, u n t i l no chloride was detected i n the acid, washing - 3 with demineralized water following each digestion. Complete removal of

residual chloride from the r e s in was attained by successively digesting it

with equal volumes of 8 M - LiNO a t 90 t o 100°C u n t i l a negative t e s t f o r 3

chlorides was obtained, again washing with demineralized water a f t e r each

digestion. The chloride-free res in was s t o ~ d as a water s lur ry u n t i l

charged t o the column.

4.3.2.2 Charging Resin t o the Column. - After conductin& repeated

resin-removal procedures (Sec. 4.'2 .I) on the f i r s t cycle colurnn t o minimize

cross contaminat ion with spent residual Permutit SK res in used i n the plu-

tonium recovery operation, chloride-free Dowex 1 X10 res in was charged t o

the column according t o the procedure described i n Sec. 4.2.1.2. The

pressure-drop measurements were then conducted, but, i n t h i s case, using

water and 8 M LiNO instead of water and 7 M HNO - 3 - 3'

4.3.2.3 Conditioning the Resin. - The Dowex 1 X10 re s in was pre-

conditioned f o r loading transuranium elements by metering 24 l i t e r s of 8 M - - 2

LiNO a t 5 m i l min-l cm t o the column held a t 60 t o 7 0 ' ~ . 3

4.3.2.4 Loading. - The adgusted feed so lu t io~ l was metered t o the

-1 -2 column held a t 60 t o 70°C a t 0.5 t o 1.0 m l min cm , depending on the

pressure drop., A t low flow ra t e s ( l e s s than 0.5 ml, min-I cm-*~ the resin

t h a t floated i n both the 8 M - LiNO and the feed solution tended t o r i s e 3

into the standpipe, resul t ing i n a deeper bed, a s ignif icant f rac t ion of

which then was of smaller cross section area. Consequently, there was a

higher pressure drop. While the lower flow ra tes were desirable from the

standpoint of s o q t i o n and low losses, prac t ica l considerations dictated

maintaining the ra tes high enough t o hold the res in bed i n the column. A

second harmful resu l t of low flow rate , stemming from the smaller e f fec t ive

cross section of tha t portion of the bed i n the standpipe, was tha t the col-

umn plugged rapid .1~ with aluminum hydroxide tha t continued t'o deposit from

feed solutions a f t e r f i l t r a t i o n .

The loading waste from the ion exchange bed was collected i n a contin-

uously air-sparged col lec t ion tank, samples being withdrawn a t intervals

during the loading operation t o detect losses. The waste was discharged

t o the intermediate-level waste storage system o r reworked depending on

the transuranium element content. Because of the developmental natum of

the program, as well as the lack of adequate salvage storage capacity,

much of the loading waste solution tha t ordinari ly would have be,en re-

worked t o recover the transuranium element content was discharged from

the process t o the intermediate leve l waste storage.

4.3.2.5 Scrubbing the Loaded Resin Bed. - Residual aluminum-bearing

solutfon and sorbed aluminum and i lgh t f l ss lon pmducts were displaced and

scrubbed from the res in with 24 l i t e r s of 8 M - LiNO metered t o the colwnn, 3

-1 -2 held at 60 t o 70°C, a t 0.5 t o 1.0 m l min cm . The spent scrubbing waste

flowed t o a continuously air-sparged col lect ion tank from which samples

were periodically withdrawn t o determine the progress of the operation.

The spent scrub solution was discharged t o the intermediate-level waste

storage system. Because of the incornpatability of lithium and aluminum

n i t r a t e s i n mixtures a t the concentrations necessary in the feed fo r t h i s

process, reworking these solutions was not generally attempted when s igni-

f icant ly high losses occurred.

4.3.2.6 Elution. - After scrubbing the res in almost free of aluminum

and l ight f i s s ion products, the sorbed transuraniwn elements (and rare

ear ths) were eluted from the bed, held a t 60 t o 70°C, with 16 l i t e r s of

-2 0 .7 M HNO metered it a t 0.5 m l min-l cm . A 2- t o 4 - l i t e r displacement - 3 fraction, routed t o the spent scrubbing waste collection tank, was taken.

Then the product st ream was diverted t o a continuously air-sparged product - col lect ion tank from which samples were withdrawn a t intervals. A l l the

product from an individual l o t of fue l rods was collected in a single .

vessel, so tha t the progress of the elut ion of an individual loading of

the res in involved taking in to account the material recovered from a l l

previous loadings, a practice tha t decreased the precision of recovery re-

s u l t s of any individual run.

4.3.3 Transuranium Element Product Handling

The d i lu te transuranium element product solution recovered by elut ion

from the ion exchange resin was concentrated by evaporation and stored f o r

f'uture use.

4.3.3.1 Evaporation. - Upon completing the elut ion of the last l o t

of transuranium element product from the resin, the d i lu te product w a s

transferred t o the evaporator and concentrated t o about 50% of i t s or iginal

volume and transferred t o storage.

4.3.3.2 Storage. - After sampling. the product concentrate i n the

evaporator, it was transferred by steam je t t o water-jacketed storage tanks

located i n a separate c e l l . The evaporator and an intermediate tank i n

the routing t o the storage tanks were flushed twice with three small

( ? - l i t e r ) l o t s of 0.5 M HNO the spent wash solutions being collected i n - 3' the storage tanks along with the product solution. The f i n a l volume of

the stored product including the washes and steam condensate used t o oper-

a te the t ransfer j e t s was about equal t o tha t of the or iginal d i lu t e pro-

duct ' before concent rat ion.

5. APPENDIX

Processing da ta accumulated during the program are tabulated i n t h i s

sect ion. Results of chemical and physical examinations of the spent res in

from the plutonium recovery s teps of the process are given also.

5.1 Detailed Process Data

Detailed chemical and radiochemical data re la t ing t o dissolution of

t h e individual f u e l rods and the feed solutions prepared from these are

given in Tables 5.1 and 5.2. Similar data f o r the spent waste and scrubb-

ing solutions a re given in Tables 5.3 and 5.4, respectively. The composi-

t i o n of the second-cycle feed solution i s given i n Table 5.5; those of the

loading and scrubbing waste from the second ion exchange cycle are given

i n Tables 5.6 and 5.7, respectively.

The composition of the f i n a l .plutonium product is given i n Table 5.8;

t h a t of the stored first stage plutonium recovery wastes is given i n Table

5.9. The compositions of the evaporator condensate and the ion exchange

feed prepared from the plutonium recovery wastes a re given i n Tables 5.10

and 5.11, respectively; those of the loading and scrubbing wastes collected

during transuranium element recovery are given i n Tables 5.12 and 5.13.

The composition of the transuranium element product f ract ions are given i n

Table 5.14; typica l alpha-energy analyses are given i n Table 5.15. Typical

compositions of the condensate collected during the evaporation of the

d i l u t e transuranium element product are given i n Table 5.16. A detailed

material balance based on alpha counting and aluminum analyses is given

i n Table 5 . l7 .

Table 5 .l. Composition of Dissolver Solutions During Plutonium Recovery

D i s ~ o l v l n g HN03 A1 Counting Rate (counts mi"-' rm-l) Dissolution Bii~h (KZ) ( 5 ) ( e l l i t e r ) p,, a C ~ G S a ~ l m e ~ e m r k ~ c r o s s 7 ( h r )

Averwe 61 of typica l runs 2-31

24-Rod Lot

1.0 . loa 6.7 x lo7 5.1 x lo7 1 .4 x 10'

1.4 x 108

2.0 x lo8 1.7 A 10'

1.8 x lo8 -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- -- --

1 .4 * 108

-- -- --

2.8 x lo7 3.2 x lo5

--

1 . 5 x 10'

8-Rod Lot

7.9 x 108 2.2 ,109

1 .6 x 109

1 . 5 x 10'

1 .g 109

1 . 3 x lo9 2.4 x 10'

1.8 x'109

9.2 x 10'

3.8 x 10'

1-md charge

6-md charge

Residual heel

1 addi t ional rod

l a d d i t i o n a l rod

Residual heel + 1 rod

I

Residual metal heel for remain* mds

6-rod charge

Reaidual metnl heel - 1 md

Residual metnl heel + 1 rod

Residual metal heel

a Single-rod charged t o batch 1 - f resh l o t of diesolvent waa added t o continue with batch 1A ofLer f i r s t had been removed . Residual metal heel dissolved v l t h 3.3 M NaN03 - 1.8 M NaOH vhlch was l a t e r ec id i f led t o Z 2 fl m<O3 with 9 M HITO and dlgeated a t 10-105°C. 3

Uigeeted a l M l n e solution 10 h r and the ac id i f ied mixture 16 hr .

Digested alkaline solution 8 h r and the ac id i f ied mixture 34 hr .

' Cuntalns wW6Yk k t e r l a l returned t o disso lver for cleaqout dissolvlnga.

Table 5.2. Composition of F i r s t Cycle Ion Exchange Feed Solution For Plutonium Recovery

Batch HN03 A 1 Counting Rate (counts min-' ml-l) Nuclide (counts min-' ml-l) No. (M) (f?) PU a cross a ~ r o s s 7 nu 7 ce-R Z r - ~ b 7 TRE p

1 3.1a 0.65 3.0 x lo7

I,A 4.7a -- 2.1 lo7 2 2 . 0 ~ 1.1 1 . 5 x lo7

, 2A 3.1a -- 4.6 x lo7

3 l .ga -- 5.3 lo7 4 3 . 0 ~ -- 5.8 x lo7

5 3.5a - 3.4 107 6 l.ga -- 4.2 x lo7

7 2.@ -- 4.3 x lo7

8 4.0a -- 5.7 lo7

9 1 . 4 ~ - 4 . g x 1 0 7

10 7.6 -- 1.8 lo7

11 7.0 -- 4.2 x lo7 i~ b . 3 -- s.1 n LO? 13 '1.U -- 3.8 x lo7

14 7.8 -- 3.7 lo7

15 7.0 -- 4.1 x lo7 i 6 7.0 -- 4.0 x lo7

17 7.2 - - 6.1 lo7 18 6.8 - - 3.9 lo7

19 6.8 - j . u x i b 7 X, 6.7 -- 5.4 lo7

21 7.1 -- 4.2 x lo7 22 6.7 -- 3.0 x lo7

23 6.9 -- 3.4 x lo7 24 6.9 - 4.5 x lo7

25 6.9 -- 3.5 26 7.4 -- 4.6 x lo7 27 7.0 -- 4.1 x lo7 28 6.7 -- 3.4 lo7

Aver- ,, 7.3 -- 3.6 x lo7

1 ,.U U.91 3.7 X lo6 4.9 X 10' 1.7 x lo9 2 7.2 0.65 1.5 x lo7 1.1 x lo9 5.7 x lo9

3 . 0.35 lo7 8.4 4.9 4 7.2 0.63 1.2 x lo7 8.2 x 10' 4.9 x lo9 2.1 x lolo 1.0 x loi0 4.4 x 10' -- 5 6.6 0.66 1.8 x lo7 1 . 3 x lo9 6.8 u 1n9 6 7.6 0.44 1.2 x lo7 1.4 x lo9 8.8 x lo9 7 7.7 0.75 1.9 x lo7 1.2 x lo9 6.3 x lo9 0 8.0 0.75 4.5 x lo7 7.7 n 10' 5.3 x lo9 9 8 .2 0 . 9 2.6 x lo7 6.9 x moo 3.A in9

10 . 7.9 0.54 6.4 x lo7 3.0 x 10' 1 .5 x lo9 10A 7.2 0.12 9 . 1 ~ 1 0 ~ 2 . 6 ~ 1 0 ' 1 . 4 ~ 1 0 ~

Aver- ,, 7.5 0.56 2.2 x lo7 8.5 x 10' 4.6 x 1u9 8.1 x iu1° 1.0 x lolo 4.4 x lo8 -- a In these m s , the feed stream t o the ion exchange beds was contlnuoualy mixed with a sLlrrUu uf 13 M TINO cdculatcd

t o adjust the ANO concentration t o 7 M. In a l l other runs, the adjustment of m w a s conducted on-a baich b a s s . 3 These values are reported i n d i s rnin-l-ml-' as fi106, ce144, zr95-rn95.

3

Runs 10 through 35.

Table 5.3. Composit ion of F i r s t Cycle Ion Excha-nge Loading Waste Solut ion During Plutonium

Re cove r y

Batch 3 Remarks No. (M) Pu (r Gmss cz Cmss 7

Aver- , age

6.8

Aver- age '1.0 -- .. .. .

-

21,-Rod Lot

3 . 2 x 1 0 7 6 . 6 ~ 1 0 '

2.8 x l o 7 6.1 x lo8

2.1 l o 7 3.9 lo8

4 . g X 1 o 7 1 . 0 ~ 1 0 9

5 . ~ ~ 1 0 ~ 1 . 2 ~ 1 0 9

6 . 4 x 1 0 7 1.:,A1~9 6 . 6 x 1 0 7 1 . 4 x 1 0 9

6.1 x lo6 -- 9 .Ox1o7 2 . 1 ~ 1 0 '

1 . 1 x 1 0 8 2 . 7 ~ 1 0 ~ Reworked as 10A, 10B

7.5 x 107 ' 1.4 l o 9 6 . 5 x 1 0 7 1 . 7 x 1 0 9

7.4 l o 5 1.5 109 7 . ~ ~ 1 0 ~ 1 . 8 ~ 1 0 ~

9.0 x l o7 2.1 x lo9 k t luL tu storage for trans- -- -- uranium element recovery

New f irs t stage resin bed

8-Hod Lot

3 . 2 ~ 1 0 ' 1 . 5 x 1 0 9

8.6 x 10' 4.3 x l o 9

6.8 x 10' 3:9 x lo9 6 . 3 ~ 1 0 ' 3 . 6 ~ 1 0 ~

9 . ~ ~ 1 0 ~ 5 . 7 ~ 1 0 9

6 . 9 ~ 1 0 ' 5 . 1 x 1 0 9

1 . 0 ~ 1 0 9 5 ~ ~ 1 0 9

8 . 5 ~ 1 0 ' 4 . 7 x l o 9

7 .Ox1o8 4 . 0 ~ 1 ~ ~

2.0 loa 1.2 109

5 . 3 x 1 0 7 6 . 6 x 1 0 8

6 . 3 ~ 1 0 ' 3 . 6 x l o 9

Table 5.4. Composition o f F i r s t Cycle Ion Exchange Scrubbing Waste Solution During Plutonium Recovery

Batch No.

HNO 3 Counting Rate (counts min-I ml-l) Remarks

( M) PU a Gross a Gross 7

1

1A

2 2A

3 4, 51 6

7, 8 9, 10

10A, 10B

11

12

13 14

15, 16

17 18

19 X)

21

22

23 24

2 5 26, 27, 28

29

30 31

32, 331 35

Average

1

2

3 4

.5 6

7

8, 9 10 10A

Average

'24-~od Lot

- - -- No scrub between 4 and 5 -- - - No scmb between 7 and 8 -- -- No scrub between 9 and 10 - - - -

-- - - 5 . 9 ~ 1 0 ~ 1 . 7 ~ 1 0 ~ 1.5 x 10 6 '1.9 x 10 8

No scrub between 15 and 16 5.6 x lo7 1 .8 x lo9 3.4 x 107 1.1 109 1 . 3 x lo7 3.8 x 10 8

6.5 x lo7 2.1 109

3.5 107 1.1 109 8.9 x 10 2 . 4 ~ 1 0 8

4.3 x lo7 1 .3 x lo9

3.9 x l o 7 1.2 109 1.2 107 3.4 x 10

8

1.5 x lo7 4.4 x 10 8 No scmb between 26,27, and 28

7.4 x lo1 2.5 x 109

7.i x 107 2.) x lu 9

7.8 x 10 4.2 x 10 8

8-Rod Lot

No scrub between 32,33, and 34

Table 5.5. Composition of Second Cycle Ion Exchange Feed Solution During Plutonium Recovery

Batch HNO 3 Counting Rate (counts min-' ml-') .

No. (MI Pua Gross a Gross y

24-~od L o t

1 through 6 8.4 1.5 x 10 8 1 .5 x 10

8 1.2 x 10

6

7 through 19 4.2 x 10 8 4.2 x 10

8 5.3 x 10 6

6.9

20 through 28 2.8 x 10 8 2.9 x 10

8 3.5 x 10

6 7.2

29 through 35 1.8 x 10 8

1 .8 x 10 8

7.3 2.3 x 10 6

Average 2.6 x 10 8 2.6 x 10

8 3.1 x 10 6

7.5

8-Rod Lot

1 through 6 7.4 1.0 x 10 7 1.1 lo7 2.3 x 10 6

Average 7 2 . 0 x 1 0 . 2 .1 x 10 7 4.0 x 10 6

7.5

Table 5.6. Composition of Second Cycle Ion Exchange Loading Waste Solution During Plutonium

Re cove ry

Batch HNO 3 count ing Rate ( counts min-l d-l)

No. (MI m a G r o s s a G r o s s y -- 24-Rod Lot

1 through 6 5.2 x 10 7.8 x 10 4 5 3.9 x 10

6 8.4

7 through 19 4.1 x 10 5 2.3 x 10 6 2.7 x 10

6 7 3

20 through28 7.1 3.1 x 10 5 -- - - 29 through 35 6.5 3.1 x l o 5 1.5 x l o 6 1.6 x l o 6

8-Rod Lot I

1 through 6 7.7 2.3 x 10 5 .1.6 x 10 4 4.7 x LO 6

7 thm* 9 1.1 x 10 6

1.1 x 10 6

2.2 x 10 6

7.7

10 .9.4 x 10 5 1.2 x 10 6 3.4 x 10 6 7 5

~ O A 6.2 x 10 1.1 x 10 4 4.9 x 10 6

7.4 7

kve rage 5.8 x 10 5 2.2 x 10 6 5.3 x 10

6 7.6

Table 5.7. composition of Second Cycle Ion Exchange Scrubbing Waste Solution During Plutonium

R e cove ry

mo -1 *-I Batch 3 Counting Rate (counts min

No. (M_> Pua! Gross, cr: Gross y

24-~od Lot

5 5 1 through 6 7-3 2.5 x 10 3.7 x 10 1.4 x 10 6

7 through 19 7 - 1 1.6 x 10 5 1.7 x 10 5 1.1 x 10 6

20 through 28 7.6 1.5 x 10' -- 29 through 35 7.2 3.7 x 10 5 - -

Average 7.3 2.3 x 10 5 2.7 x 10 5

5 1 through 6 3.7 x 10 1.3 x 10

6 7 -8

7 through 9 7.6 9.3 x 10 5 1.1 x 10 6

10 - 10A 7.11 5.3 x 10 5 5.7 x 10 5

Average 6.1 x 10 5 9.9 x 10 5 1 .4 x 10

6 7.6

Table 5 .e . Composit ion of Plutonium Pro.3uct

Plutonium Counting Dzta un++ 1 m EN03 " - 0

I so topic Assay /-L d\

U V b l r L C 3 31) u \- P! No. . . . . ( collnts min-l d-9 ( E ) 1 i Mass Pl3ss Mass Mass Mass

( g ) ( g l l i t e r ) Gross r Gross y \o/ A * " - A /

--.-- 2 38 2 39 240 2 41 242

a Average 655

24-~od Lot

8-3od Lot

2.3 1.071

- - 1.059

1 .2 1.034

4.8 1.154

3.2 .1.105

a E -1 -1 SF. a c t i v i t y calculated from veighted average, these da ta = 1.843 x LO a counts min Pu.

8 -1 -1 Sp. a c t i v i t y calculated from weighted average, these data = 2.56 x 1 0 a courts min mg Fu.

Table 5.9. Composition of Composited Raw Plutonium Recovery Wastes

24-~od Lot 8-Rod ~ o t - - -

volume ( l i t e r s )

A 1 (MI

m3 (MI Sp. 0. (g xnl- l )

G ~ O S S a (counts rnin-I d-l)

a-pulse ($ ~ r o s s a)

5.5 Mev

5.8 Mev

6.1 Mev

PU ( g t o t a l )

- 1 Pu conc. (mg l i t e r )

Table 5.10. Typical Composition of Evaporator Condensate Evolved D u r i r g Tratisuratiiuni Element

Feed Preparation

Gross CY Gross y HNO Batch Fraction 3 Number (counts mine' ml-l) (counts min'l d-l) (MI Remarks

8-Rod Lot

Evaporation

Evaporation

Evaporation

Digest ion

Digest ion

Evaporation

Evaporation

Evaporation

Digest ion

Digest ion

Evaporat. ion

Evaporation

Evaporation

Evaporation

Evaporat ion

Eviuporat ion

Evaporation

Evaporation

Digest ion

Digest ion

Table 5.11. Composition of Ion Exchange System Feed Solution During Transuranium Element Recovery

Batch A 1 HNO 3 Gross (counts mine' d - l ) Nuclide ( d i s m i d 1 d - l )

eta) (camma) Ce 144. C s 137 Ru 106 S r 90 Z r 95

Aver - a@;'=

2.5 1.23

3.0 x 10 8 - - 7.3 x 10 9

2 .8 x lo8 1.0 x 1~ lo 7.6 x 10 9

2 . 3 ~ 1 ~ ~ 8 . 0 ~ 1 ~ . 6.4 x 10 9 3 . 2 x 1 0 8 7 . 1 ~ 1 0 5.6 x 10 9

2.6 x lo8 8.5 x 10 5.6 x 10 9

. 9 x 1 0 8 1 . 0 ~ 1 0 lo 8.4 x 10 9

Returned t o Run 8

2.5 x lo8 1.0 x lo lo 7.7 x 10 9

2 . 8 x 1 0 8 8 . 9 ~ 1 0 ~ 6 . 9 ~ 1 0 9

8-Rod Lot

a AD = Acid def ic iant .

Acid.

Tsble 5.12. Cornposit ion of Ion Exchange Loading Waste Solution During Transuranium Element Re coveqs

Batch A 1 Gross ( ccunts mi*-' m l - l ) Nuclide (d i s min-l ml-l) No. (g)

( ~ l p h a ) eta) ( G ~ A ) Ce 144 C s 137 Ru 106 S r 90 Z r 9 5

7 Returned t o Run 8

8 - - 6 . 4 x 1 0 5 1 . 3 ~ 1 0 ~ 3 . 8 ~ 1 0 7.0 x 10 7 -- - - - - - - - Aver-

. 2 .O 7 . 0 ~ 1 0 ~ 2 . 3 ~ 1 0 ' k . 2 ~ 1 0 2.4 x 10' 6.8 x 10' 1.8 x 10' 3.6 x lo9 5.5 x 10 7 %e

- 8-Rod Lot

6 0 11. - - 5 . 7 ~ 1 0 l . 8 x 1 0 9 7 . 5 x 1 0 ° - 6.8 x 10 - - - - - - - - 7

- Aver- o - - 9 . 9 ~ 1 0 ~ k . 7 x 1 o 9 5 . 6 ~ 1 0 ' 2.8 x 10 8 -- - - - - - -

Table 5.13. Composition of Ion Exchange Scrubbing Waste Solution During Transuranium

Element Re cover-

Batch Gross ( counts min-l ml-l) N o . Alpha Beta Gamma

24-RO~ Lot

7 Returned t o Run 8

8 1.3 x 10 6 7.7 x 10

8

8 Average' 1.8 x 10 1.2 x 10 9

Table 5.14. Conrpos it ion of Transuranium Element Froduct

*l . L i FDO 3

- 1 Gross ( c o ~ n t s min ml-') Nuclide (d i s mi*-' ml-l) (g) (2) (M) -

( ~ l ~ h a ) eta) (camma> Ce 144 (C s 137 Ru lo6 ~r 90 zrg5

!&-Rod Lot

Dilute 0.035 2.0 0.13 3 . 3 ~ 1 0 ~ 6 . 3 : < l i 9 7 . 6 ~ 1 0 ~ 4 . 4 : < m 10 - - - - - - - - Concen- 0.060 6.k 1.10 1 . 1 ~ 1 0 ~ 2 . 2 ~ 1 0 ' ~ 2 . 7 ~ 1 0 ~ 1 .9x l0 ' l 4 . 8 ~ 1 0 ~ 1 . 8 x l 0 ~ O 5 . 8 ~ 1 0 7 4 . 1 ~ 1 0 7 t r a t e 8

8-Rod Lot

Dilute 0.06 1.1 -- 1 . 9 x lo9 3.5 x lo9 6.7 x lo8 2.7 x 10 3.5 x lo8 5.6 x 10 7 - - < 2 x 1 0 7 10

Concen- 2.2 -- 4 . 1 ~ 1 0 ~ E . 3 ~ 1 0 ~ 2 . 1 ~ 1 0 ~ G 5 . 5 ~ 1 0 ~ ' 3 . 4 ~ 1 0 ~ 1 . 9 ~ 1 0 0.10

10 - - 1.3 x 10 8 t rate

Table 5.15. Alpha Energy Analysis of Transuranium Element Product

Gross Alpha Alpha Pulse ($. gross alpha)

(counts min-l d - l )

Dilute

Concen- t r a t ed

Dilute

Concen- t r a t ed

24-~0d Lot

3.3 x 10 8

28

1.1 x 10 9 24

8-RO~ Lot

Table 5.16. Typical Composition of Condensate Evolved During Transuranium Element Product

Evaporation

cross a Gross y HNO Fraction -A w-l) - 1 3 Remarks

( c b u i t ~ min ( c u w ~ ~ a l l i l ~ r ~ u 1 - I ) (g)

2 4 - ~ o d Lot

Table 5.17. Material. Balance : Transuranium Element Recovery Based on Alpha Counting and Aluminum Analyses

Alpha Total ' Aluminum Stream (qb of Raw

( counts min-l) (qb of Raw Feed) Feed)

24-Rod Lo t

14 Raw Feed: I N 1 . 3 x 10 100 1061 loo Prepared Feed: OUT

Discharged t o waste 3.2 x l 0 l 3 2 5 220 ' 21

Processed t o ion exchange 9..1 1013 75 863 81

Total accounted for : 1 . 3 x 10 l4 1 . 3 x 10 14

(feed preparation) 100 1083 102

Ion exchange losses

Loading

Scrubbing

Total. losses , 4.5 1013 4.5 35

Product recovery

Hecovered 3.8 x 1013 . 29

Recoverable 3.5 x 1012 3

Total product 4.2 x l0 l3 4.2 x l 0 l 3 32

Total accounted for : (ion exchange)

8-Rod Lo t

Raw Feed: I N 6.0 x 10 100 48 5 100 14

Prepared Feed: OUT

Discharged t o waste 1 . 3 x 10 14

Processed t o ion exchange 3.6 x 10 1 4

Total accounted for : 4.9 x 10 l4 4 . 9 ~ 1 0 14

(feed preparation)

Ion exchange losses

Loading 4.2 x l 0 l 3 7

Scrubbing 1.1 1013 2

~eakage t o sumpa 3.3 1013 5

Total losses 0.6 x lo1' 8.6 x 10 13

14

Product recovery

R a c u v a ~ ~ c l

Recoverable

Total product 2.5 x 10 l4 2.5 x 10 14

42

Total accounted for : 3.4 x 10 l4 3.4 x 10 14

(ion exchange )

a Including material l e f t on incompletely eluted resin.

5.2 Examination of Spent Resin

After three months service in the f i r s t stage column during the pro-

cessing of the 24-rod l o t of fue l rods, the charge of Perruutit SK resin

8 had received an estimated dose of 4 x 10 rads . Upon removal and examin-

a t ion of the r e s in it was observed t o be sof t and je t black but remained

essent ia l ly as d iscre te beads; however, there was some evidence of fines

t h a t indicated a small fract ion of the res in beads had shattered.

During the recovery of plutonium from the eight-rod l o t of fuel, the

resin was removed from the column a f t e r a month of service and examined.

The estimated radiat ion dose received during t h i s interval w a s 4.6 x 10 7

rads. This res in also was jet black and, s t icky (it compacted and aggre-

~ a t e d under pressure) . Prior t o removal, the res in had been exhaustively

eluted with 0.7 M - t o 3 M HNO containing from 0.01 t o 0.1 M F- and with - 3 - 0.7 M - t o 3 M HNO containing 0 .1 M Fe (NH SO ) o r NH20H.H2S04. - 3 - 2 3 2

After removal from the column, samples from the top and bottom of the

r e s in bed were examined in the laboratory. The resin from each location

w a s exhaustively eluted with 0.7 t o 7 M HNO containing 0 .O1 t o 0.1 M F-, - 3 - with reducing agents, and f ina l ly , with aqua regia. I n all cases, plu-

tonium a t about 0.1 rng/ml ( i n the eluant) , and aluminum at about 0.1 mg/ml,

was eluted along with f i ss ion products, indicating these t o be t igh t ly

bonded t o the resin. Samples of res in from each location were then ashed

and the residue dissolved as completely as possible in aqua regia - a small

portion of the ash did not dissolve, even on repeated fuming t o dryness.

The solution of the ash and the residual sol ids were examined f o r plutonium,

aluminum, and other ionic material using radiochemical and spectrographic

procedures. Results were as follows :

Pu

A 1

Pd

Rh

Fe, C r

Z r

Nb

Rare Earths

Cs

Component Found ( g / l i t e r of resin)

Top of Bottom of Residual Solids

Resin ~ e d " from Ash

Resin Bed Dissolution

3.2

0.5

0.14

0.03

Trace

< 0.2

< 2.0

<1.0

- -

Moderate

Strong

Weak

Weak

- - - Faint t race

-- - - -

a Estimated t o comprise about 20 t o 3076 of the bed.

6. CONCLUSIONS AND RECOMMENDATIONS

From the r e su l t s and experience obtained i n t h i s program, it is con-

cluded t ha t :

1. Adequate dissolution ra tes of highly radiated Pu-A1 al loy can be

nbtain~cl witah h M HNO, c.~,t,a.l yzed with 0.05 M Hg and 0.83 M F- t o m e the - 3 - - method prac t ica l f o r pi lot . plant-scale operat ion.

2. Highly purified plutonium, decontaminated from f i s s ion products,

can be recovered with low losses by anion exchange methods. However, a

prelimi'nary feed preparation s tep i n which the bulk of the' aluminum was

removed ( f o r example, by dissolution i n NaOH followed by c e n t r i m a t i o n

of the plutonium hydrates and dissolution of these i n the minimum amounts

of HNO ) and the plutonium concentration increased would resu l t in feed 3

solutions b e t t e r suited f o r processing the highest burhup material because

of the lower un i t losses obtained when processing solutions of higher

plutonium concentrat ion.

3. Transuranium elements (and rare ear ths) can be separated and re-

covered f rom the plutonium-loading waste e f fluent i n good yield and nearly

f r ee of aluminum. The low loading capacity of the anion resin, however,

imposes some l imi ta t ion on the general appl icabi l i ty of the method. .

Diff icu l t ies encountered during the program lead t o the recommendat ion

t h a t :

1. Improved methods f o r sol ids removal and subsequent handling should

be developed t h a t w i l l permit good separation with rapid throughput. In

addition, an improved method of discharging the sol ids t o waste i n the

minimum volume of waste solution without pluffaing process lines w~uld in-

crease recoveries and improve the mechanical operation of' the system.

Solids-free feed solutions are required f o r e f f i c i en t operation of the ion

exchange system.

2. The ion exchange columns should be equipped with an e f f i c i en t

res in removal system tha t permits complete removal of spent res in t o a

un i t where t o t a l dissolution of the res in can be conducted, both f o r i t s

ultimate disposal and fo r the recovery of irreversably sorbed valuable . .

material . This requirement becomes more important with extremely high

b u m p material where resin degradation i s severe.

3. An improved evaporator should be designed tha t permits the con-

centrat ion of solution t o about 20% of the working volume of the vessel

without exposing the heat t ransfer surfaces. Such a design would r e su l t

i n the formation of fewer sol ids i n the concentrate, easier , more rapid

c l a r i f i ca t ion of concentrated solutions, lower losses of valuable material,

l e s s material" f o r rework, increased system capacity, and decreased volumes

of waste solution.

4. A resin hold-down device should be included i n the res in column

design. ' Because of the high densi t ies of the feed and scrub solutions

necessary i n the transuranium element recovery process, the ion exchange

res in tends t o f l o a t . In the present program, it was necessary t o operate

a t minimum flowrates such t h a t the downflow ra t e in the standpipe exceeded

the ra te of r i s e of the res in in the solution. F lex ib i l i t y of operation

was sacrificed as a consequence, since a t below these minimum ra tes res in

rose i n the column and standpipe, resul t ing i n an effect ively longer column

of smaller diameter with increased pressure drop. The need i n t h i s case

was also increased when solutions thatt'ga$' ( typica l a t these radiation

levels) are handled. Gas binding of the res in bed was a serious operational

p1.oblern a ~ d often resulted in partially turning the bed over, with .increased

loss of valuable material. For t h i s reason also, sensing devices f o r measur-

ing Lhe pressure drop aeyoss the r e s in 'bed should not be of the pneumatic

type tha t discharge gas into the column.

7. ACKNOWLEDGEMENT

The authors gra te f i l l y acknowledge the assistance and contribution of

t h e following members of the Chemical Technology Division i n the operation

of the plant :

* T. Aochi

W . T. Bostic

E . W . Brown

J. H. Groover

W . A . Lindsey

R. J. Shannon

W . A. Shannon

J. S. Taylor

J . H. Walker

A. V. Wilder

I n addition, the technical assistance of the following Chemical Technology

Division members i n flowsheet modification and development and i n the pro-

cess chemistry ensured the success of the program:

R . D. Baybarz R. E . Leuze

J. E . Bigelow M. H. Lloyd

S . R. Buxton

Also, the authors a re indebted t o J. H. Cooper, C . E. Lamb, W . R . Musick,

and E . I. Wyatt, along with t h e i r co-workers, a l l of the Analytical Diip5.-

sion, f o r conducting the many analyses tha t were required ' to complete the

program described i n t h i s report . B. T. Wa1ter.s and J. A. Westbrook of

the Health Physics Division helped ensure the safe operation of the plant

and provided a major contribution t o the successf i l completion of the

program.

* Guest s c i e n t i s t from Japan.

REFERENCES

1. W . D. Perkins, isso solution of Pu-A1 Alloy, " DP-702, March, 1962.

2. F. L. Culler, e t a l . , "Chemical Technology Division Annual Progress Repart f o r Period Ending August 31, 1959, " ORNL-~~M, Page 36, September 11, 1959.

3. w. H. Lewis and R. E . Brooksbank, "The Recovery of Plutonium from Various Sources at the OWL Metal Recovery Plant," Proceedings of the AEC Plutonium Fuel Technology Symposium, Hartford, Washington, April 22, 1957.

4. R . E . Brooksbank, "Plutonium-Aluminum Alloy Processing, " ORNL Chemical Te chno'logy Division Annual Informat ion Meeting, September, 1962.

5. F. L. Culler, e t a l . , "Chemical Technology Division Annual Progress Report f o r Period Ending June 30, 1962," OWL-3314, Page 156, September 13, 1962.

6. F. L. Culler, e t a l . , "Chemical Technology Division Annual Progress Report f o r Period Ending May 31, 1963," 0 ~ ~ ~ - 3 4 5 2 , Page 134, September 13, 1963.

T H I S PAGE

WAS INTENTIONALLY

. LEFT BLANK

ORNL-3566 UC-10 - Chemical Separations Processes

for Plutonium and Uranium TID-4500 (28th ed.)

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5. Reactor Division Library 6-7. ORNL - Y-12 Technical Library

Document Reference Section 8-42. Laboratory Records Department

43. Laboratory Records, ORNL R.C. 44. H. H. Abee 45. E. D. Arnold 46. Russell Baldock 47. C. J. Barton 48. R. D. Baybarz 49. J. E. Bigelow 50. G. E. Boyd 51. J. C. Bresee 52. R. E. Brooksbank 53. W. D. Burch 54. T. A. Butler 55. W. H. Carr 56. F. L. Culler 57. H. E. Goeller 58. W. R. Grimes

59. D. M. Hewette 60. F. A. Kappelmann 61. R. H. Lafferty, Jr. 62. C. E. Larson 63. R. E. Leuze 64. W. T. McDuffee 65. 3. P. Nichols 66. F. L. Peishel 67. R. H. Rainey 68. J. T. Roberts 69. L. D. Schaffer 70. M. J. Skinner 71. C. M. Smith 72. J. A. Swartout 73. J. W. Ullmann 74. A. M. Weinberg 75. W. R. Whitson 76. 0. 0. Yarbro 77. P. H. Emmett (consultant) 78. J. J. Katz (consultant) 79. T. H. Pigford (consultant) 80. C. E. Winters (consultant)

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83-588. Given dis tr ibut ion as shown i n TID-4500 (28th ed. ) under Chemical Separations Processes for Pl.11 tond.i.im and Uranium category (75 c o p i e ~ - OTS)