Transplutonium Elements-Production and Recovery - ACS Publications

6

Downloaded by UNIV OF MASSACHUSETTS AMHERST on February 27, 2016 | http://pubs.acs.org Publication Date: July 20, 1981 | doi: 10.1021/bk-1981-0161.ch006

Recovery of Americium-241 From Aged Plutonium Metal L. W. GRAY, G. A. BURNEY, T.A. REILLY, T. W. WILSON, and J. M. McKIBBEN Ε. I. du Pont de Nemours & Company, Savannah River Laboratory, Aiken, SC 29808

The Savannah River P l a n t (SRP) was requested to separate approximately 5 kg of Am from about 850 kg of plutonium metal containing, nominally, 11.5% Pu. A f t e r separation and p u r i f i c a t i o n , both a c t i n i d e s were p r e c i p i t a t e d as oxalates and c a l c i n e d to t h e i r r e s p e c t i v e oxides. The PuO was shipped to the U.S. Department of Energy Hanford S i t e for use as f u e l in the Fast F l u x Test Facility (FFTF); the AmO was shipped to the Oak Ridge N a t i o n a l Laboratory Isotope Sales Pool for use as a neutron source in many f i e l d s , predominantly petroleum well-logging. 241

240

2

241

2

A l a r g e - s c a l e process was developed s p e c i f i c a l l y for SRP a p p l i c a t i o n using e s t a b l i s h e d d i s s o l u t i o n , separation, purifica tion, p r e c i p i t a t i o n , and c a l c i n a t i o n technology. However, adaptation of the process to e x i s t i n g plant facilities required a s u b s t a n t i a l development e f f o r t to c o n t r o l c o r r o s i o n , to avoid product contamination, to keep the volume of process and waste s o l u t i o n s manageable, and to d e n i t r a t e s o l u t i o n s with formic a c i d . The Multipurpose Processing Facility (MPPF), designed f o r recovery of transplutonium isotopes, was used for the first time for the p r e c i p i t a t i o n and c a l c i n a t i o n of americium. Also, for the first time, l a r g e - s c a l e formic a c i d d e n i t r a t i o n was performed in a canyon v e s s e l at SRP. Conceptual Process Because it was necessary to use a process that would work in e x i s t i n g equipment, a process was designed (diagrammed in Figure 1) i n v o l v i n g the f o l l o w i n g operations: •

D i s s o l u t i o n . Plutonium metal was d i s s o l v e d in 1.67M sulfamic a c i d at about 25°C to 60 +_10 g Pu/L. The Pu0 c o a t i n g on the surface of the metal plus the PuH (where χ = 2.0 to 2.7) produced from the r e a c t i o n of ^ ( g ) with plutonium metal formed a sludge which was c o l l e c t e d and subsequently d i s s o l v e d separately using hot 14M HNO3 c o n t a i n i n g 0.2M KF. 2

x

0097-615 6/ 81 /0161 -0093$05.00/ 0 © 1981 American Chemical Society

In Transplutonium Elements—Production and Recovery; Navratil, James D., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 1981.

T R A N S P L U T O N I U M

221-F

AGED PLUTONIUM METAL


B-LINE

E L E M E N T S

HNO3 HF

H NS0 H 2

SLUDGE DISSOLVER

METAL DISSOLVER

3

H2C2O4 PLUTONIUM FINISHING:

IL

PRECIPITATOR |—»CQ . H

I

HNO3-

Pu0

CATION EXCHANGE COLUMN

I STORAGE I

CANYON

2

H 0 2

2

ΗΝ0

3

SOLVENT EXTRACTION:

AQUEOUS SCRUB HNO-,

3 0 % ΤΒΡ-η (η -paraffin) j

Χ

Λ2 MIXE R - S E T Τ L E R

MIXER-SETTLER (EXTRACTION) SECOND

(STRIPPING)

PLUTONIUM

-SOLVENT

CYCLE

RECOVERY

H 0 2

LOW

ACTIVITY

HNO3

WASTE

EVAPORATION: H

2

JEVAPORATORS 0 (2 STAGES)

H

2

C

2

0

4

H

FEED ADJUSTMENT Π-F

FRAMES:

C 0 RERUN : • •

EVAPORATION: DENITRATION :

1.

, N 0

2

2

C

2

H S 0 2

0

4

-

4

-

H N O 3 CATION EXCHANGE COLUMN

2

|H O 2

EVAPORATOR (DENITR ATOR]

Π

,

Να, F e , P u , N i , S O 4 (TO

5

WASTE)

"H 0 2

MPPF EVAPORATOR

H

2

C

2

0

4

-

PRECIPITATOR

co

2

H 0 2

A m 0 PRODUCT

2

Figure 1.

0

2

PRODUCT

Y—ΝαΝ0

FEED ADJUSTMENT

2

4

1

241

Process for recovery of A m from aged plutonium metal


2


6.

GRAY

ET AL.

241

Am

Recovery

from Aged

Pu

95

•

Feed Adjustment f o r E x t r a c t i o n . D i s s o l v e r s o l u t i o n was accumul a t e d and d i l u t e d to 95% 2^Am02. By s e l e c t i v e blending, impurities in the shipped product, predominantly lead and n i c k e l , were kept below 2%. Experimental Procedures All experiments were conducted using normal laboratory g l a s s ware. Chemicals used were t e c h n i c a l grade chemicals removed d i r e c t l y from process chemical hold tanks where p o s s i b l e ; r e s i n s were from the same production l o t s as would be placed in the process column equipment. Cation exchange feed rates were the same as obtainable in plant equipment. Laboratory Results D i s s o l u t i o n of Plutonium Metal Plutonium metal d i s s o l v e s r e a d i l y in sulfamic acid at ambient temperatures according to the r e a c t i o n Pu° + 3 H - ^ P u +

+ 3

(NH2SO3H)

+ 3 / 2 H (g) 2

(1)

The d i s s o l u t i o n rate at about 25°C depends upon acid c o n c e n t r a t i o n and surface area of the metal. T y p i c a l l y , i n i t i a l batches of s o l u t i o n from the d i s s o l v e r average 50 +5 g Pu/L; the concentrat i o n increases to 60 +10 g Pu/L when using a cycle of 1 hour d i s s o l v i n g time followed by displacement of two- t h i r d s of the s o l u t i o n . A more complete treatment of both ambient temperature and elevated temperature d i s s o l v i n g experiments is given e l s e where U _ , 2, 3).


6.

GRAY E TA L .


Storage

2 4 1

Am

Recovery

from Aged

of Dissolved Plutonium

Pu

97

Solution

Simulated storage experiments showed (Figure 2) that r a d i o l y s i s would be inadequate for valence adjustment of P u ( l l l ) to Pu(lV) w i t h i n the a v a i l a b l e time frame. It was also necessary to assure that plutonium s u l f a t e s would not p r e c i p i t a t e during storage. The s o l u b i l i t y of plutonium vs. n i t r i c acid concentra t i o n at various concentrations of s u l f a t e is shown in Figure 3. Because the plutonium concentration in canyon tanks is kept at because the s o l u b i l i t y of N a H S 0 4 is exceeded during the first stage of evaporation. However, downstream processing through c a t i o n exchange d i c t a t e s that o x i d a t i o n , not h y d r o l y s i s , must be the mode of d e s t r u c t i o n of the sulfamate ion. Sodium americyl s u l f a t e is also r e l a t i v e l y i n s o l u b l e in n i t r i c acid (Table I ) . The s o l u b i l i t y of this s a l t is also ex ceeded during the first stage of evaporation. However, subse quent a c i d s t r i p p i n g of the s o l u t i o n s reduces the n i t r i c a c i d c o n c e n t r a t i o n and the s a l t s r e d i s s o l v e .


TRANSPLUTONIUM

ELEMENTS


7.5M HNO-

Time, days

Figure 2.

Radiolytic oxidation of Pu(IH) to Pu(IV) in sulfamic acid-nitric acid solutions

1 ι ι ι—ι—ι—ι—ι—ι—ι—ι—Γ

0

1 2

4

6

8

10

12

14

Nitric Acid C o n e , M

Figure 3. Calculated equilibrium concentrations of Pu(IV) in nitric acid solutions containing sulfate ions: (O) 0.25M H SO , (A 0.40M H SO (Φ) 0.2M H SO 2

k

2

if


2

h


GRAY E TA L .

241

Am

Recovery

from Aged

Pu

Nitric Acid C o n e , M

Figure 4.

Solubility of sulfate salts in nitric acid solutions: Ο NHf salts

TABLE I.

ΗΝ0 , M 3

S o l u b i l i t y of Sodium Americyl

+

Na , M

S0

4>

M

Να* salts, (Ο)

Sulfate

Am in s o l u t i o n , g/L (in equilibrium with s o l i d )

0.8

2.1

1.0

0.15

1.4

2.3

1.1

0.25

1.4

3.3

1.5

0.15

4.2

2.1

1.0

0.4

4.2

3.1

1.5

0.2

a. All s o l u t i o n s contained Fe (1.5 g/L), Cr (0.4 g/L), and N i (0.2 g/L).


100

TRANSPLUTONIUM

ELEMENTS


The p a r t i a l decomposition of d i s s o l v e d TBP in the evapora t i o n step leads to the p r e c i p i t a t i o n of a white organophosphate solid. Complete decomposition to an a c i d soluble compound r e quired extended b o i l i n g o f MOM n i t r i c acid s o l u t i o n s as would be achieved i f the first evaporation step were a f a c t o r of 50. Use of this evaporation f a c t o r in the first stage, however, led to excessive c o r r o s i o n of the s t a i n l e s s s t e e l process equipment. The first evaporation step was l i m i t e d to a f a c t o r o f 25 to l i m i t the HNO3 concentration to 98% of the trace Zr, Nb, and Pu ions. O x a l i c a c i d in the feed does not a f f e c t the s o r p t i o n be h a v i o r of C r and N i ^ ions on the c a t i o n r e s i n . However, about 75% of the N i ( l l ) ions were in the s o r p t i o n and wash e f f l u e n t s because r e s i n a f f i n i t y for N i ^ ion is lower than that for C r and Am"*" i o n s . 3 +

+

+

3 +

3

More than 99% of the sodium was separated when two a c i d washes were made a f t e r the loading c y c l e . The first wash was about 15 bed volumes of 0.2M H SO4-0.05M H 2 C 2 O 4 ; the second wash was about 5 bed volumes of 0.25M H N O 3 , which removed the remaining sodium and also flushed s u l f a t e and oxalate from the r e s i n bed. 2

E l u t i o n with 5M HNO3 at 0.5 mL/Cmin-cm^) removed about 87% of the americium in four bed-volumes and >99% in eight bed-volumes.


6.

GRAY E TA L .

2 4 1

Am

Recovery


Formic A c i d D e n i t r a t i o n

from Aged

of Product

101

Pu

Solution

Results of formic acid d e n i t r a t i o n simulation experiments (Figure 5) showed that at >90°C, d e n i t r a t i o n began when s u f f i c i e n t 23.5M formic acid had been added to bring the s o l u t i o n concentration to about 0.06M formic a c i d . The lowest f r e e - a c i d concentrations f o r the laboratory s o l u t i o n s were obtained when a formic a c i d - t o - f r e e n i t r i c acid mole r a t i o of about 1.6 to 1.9 was used. This r a t i o y i e l d e d a f i n a l f r e e - a c i d concentration of 0.7 to 0.8M. In the region where free n i t r i c a c i d is 96% o f the N a , S O 4 " , F e , Pu, and f i s s i o n product ions. However, the process r e j e c t e d only 88.8% of the n i c k e l and

Transplutonium Elements-Production and Recovery - ACS Publications

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