Radioactive Waste in Geologic Storage - American Chemical Society

5 Leaching of Fully Radioactive High-Level Waste Glass

Radioactive Waste in Geologic Storage Downloaded from pubs.acs.org by UNIV OF CALIFORNIA SAN DIEGO on 04/19/16. For personal use only.

and Waste-Geologic Environment Interaction Studies D . J. B R A D L E Y Pacific Northwest Laboratory, operated by Battelle Memorial Institute for the U . S . Department of Energy, Richland, W A 99352

As part o f c o n t i n u i n g studies i n waste management, the P a cific Northwest Laboratory (PNL), operated by Battelle Memorial Institute for the Department o f Energy (DOE), has been conduct i n g the H i g h - L e v e l Waste Immobilization Program. The purpose of t h i s program i s to develop and demonstrate technology f o r i n c o r p o r a t i n g nuclear wastes i n t o final waste forms. Release r a t e data for r a d i o n u c l i d e s from fully r a d i o a c t i v e waste forms are needed to evaluate the safety o f nuclear waste glass. P r e s e n t l y , contact w i t h water i s considered the most important r e l e a s e path; t h e r e f o r e , the r e l e a s e p r o p e r t i e s o f waste glass in water are o f primary concern. This report describes the p r e p a r a t i o n and leach t e s t i n g o f fully radioactive zinc borosilicate g l a s s , prepared from power r e a c t o r waste. Leach t e s t s were conducted on t h i s fully r a d i o a c t i v e waste glass t o : • determine the r e l e a s e r a t e s o f r a d i o a c t i v e m a t e r i a l from the glass • compare leach r a t e s to those o f n o n r a d i o a c t i v e waste of the same composition • study elements not included in simulated waste glasses • recommend further studies needed in this a r e a .

0-8412-0498-5/79/47-100-075$05.00/0 © 1979 A m e r i c a n C h e m i c a l Society

76

RADIOACTIVE

WASTE

IN GEOLOGIC

STORAGE


To evaluate the s a f e t y o f geologic r e p o s i t o r i e s , PNL i s a l s o conducting the Waste I s o l a t i o n Safety Assessment Program (WISAP) f o r the DOE. Task #2 of the WISAP program i s aimed a t p r o v i d i n g r e l e a s e rate data from p o t e n t i a l waste forms to be used i n geologic storage. Experiments simulating simple to complex waste form r e l e a s e i n geologic media are o u t l i n e d . EXPERIMENTAL PROCEDURE Nuclear waste glasses are complex mixtures of more than 30 elements, u s u a l l y prepared by methods q u i t e d i f f e r e n t than those used commercially to produce p l a t e g l a s s , f o r example. C a n i s t e r s o f g l a s s ( 1 - f t - d i a χ 6 - f t - t a l l ) have been prepared by d i r e c t furnace melting of a mixture of the waste oxides and an appropriate glass f r i t . The present experiment simulated t h i s procedure on the l a b o r a t o r y s c a l e . In J u l y 1975, 2.7 kg of spent i r r a d i a t e d power reactor f u e l was d i s s o l v e d using l a b - s c a l e equipment i n the h o t - c e l l f a c i l i t i e s at PNL. This f u e l had an average burnup of 55,000 MWD/ MTU. A f t e r e x t r a c t i o n o f uranium and plutonium, the r e s u l t i n g l i q u i d waste was b o i l e d down and heated to 500°C to y i e l d 380 g of f i n e granular oxide m a t e r i a l c a l l e d c a l c i n e . We then ana lyzed t h i s c a l c i n e f o r r a d i o i s o t o p i c content, which included a c t i v a t i o n products, f i s s i o n products and a c t i n i d e s . A z i n c b o r o s i l i c a t e glass f r i t was blended with t h i s waste c a l c i n e i n a 3:1 weight r a t i o to make the waste g l a s s . The composition o f the product glass based upon f r i t composition, O R I G E N / - and the c a l c i n e a n a l y s i s i s shown i n Table I. As you can see, the glass has a complex composition. We made the waste glass i n 100-g batches by melting the f r i t and c a l c i n e i n a s t a i n l e s s s t e e l 304L c y l i n d e r which was placed i n s i d e a resistance-heated furnace. The dimensions o f t h i s c y l i n d e r were 3.8 cm d i a χ 12.7 cm height, with a w a l l thickness of 0.165 cm. The waste was brought up to the melting temperature o f 1050°C i n two hours, held there f o r three hours, and then allowed to furnace c o o l at an average rate o f 220°C/hour. A f t e r c o o l i n g we sectioned the s t a i n l e s s s t e e l c a n i s t e r c o n t a i n i n g the waste glass by using a water-cooled diamond saw. Figure 1 depicts t h i s operation and shows the four s e c t i o n s that we used f o r t h i s leaching study. A l l sections appeared homogeneous without large cracks or v o i d s . This same procedure on subsequent glass runs provided samples of f u l l y r a d i o a c t i v e glass f o r metallography and e l e c t r o n microscopy. Information on those analyses, along with glass-making v o l a t i z a t i o n data i s reported i n BNWL-2252 2 and BNWL-2625. 1' Further d e t a i l s on sample preparation, plus a complete l i s t o f the leaching data obtained, are a v a i l a b l e i n PNL-2664. We began the leaching test on the four glass sections i n February 1976, using the apparatus shown i n Figure 2. P r i o r to (

)

(


5.

BRADLEY

STAINLESS STEEL 304 L

Leaching

of Waste

77

Ghss

MINI-CANISTER

GLASS CRUST

SECTIONING

LEACHING STUDIES TOP SECTION UPPER MIDDLE SECTION LOWER MIDDLE SECTION BOTTOM SECTION

ARCHIVE SECTION 3.81 cm OD Figure 1.

Sectioning of high-level waste ghss canister

CANISTER SECTION OF HIGH LEVEL WASTE GLASS Figure 2.

Long-term IAEA

leaching apparatus

RADIOACTIVE

WASTE

TABLE I. F u l l y Radioactive Waste Glass Radioactive Waste in Geologic Storage Downloaded from pubs.acs.org by UNIV OF CALIFORNIA SAN DIEGO on 04/19/16. For personal use only.

Component Si0 ZnO

2

B

2°3 uo Na 0 κ ο SrO Zr0 2

2

2

2

M0O3

MgO CaO BaO Nd 0 Ce0 Cs 0 BaO La 03 Pr 0 PdO Tc O Sm 0 Np0 Y 0 Pu0 Ru0 Eu 0 Am0 Gd 0 Cm 0 SbO CoO MnO 2

3

2

2

2

2

3

2

y

2

3

2

2

3

2

2

2

3

2

2

3

2

3

I N GEOLOGIC

Composition

Weight Perc< 27.8 21.7 11.3 10.3 4.1 4.1 1.9 1.6 1.6 1.5 1.5 1.5 1.5 1.0 0.8 0.5 0.45 0.44 0.43 0.39 0.29 0.22 0.19 0.12 0.09 0.06 0.06 0.05 0.02 0.01 1.1 χ 6.0 χ

(a) The remaining m a t e r i a l i s made up o f chemicals used i n the f u e l d i s s o l u t i o n and uranium/plutonium separation operations.

STORAGE


5.

Leaching

BRADLEY

of Waste

79

Ghss

the s t a r t o f the leach t e s t , a l l sections were washed i n acetone to remove surface p a r t i c l e s , then the s e c t i o n s were d r i e d i n a i r . Deionized water was used as the leachate and the r a t i o of leachate volume (cm3) to sample surface area (cm2) was 26 cm. The leachate was changed according to the I n t e r n a t i o n a l Atomic Energy A s s o c i a t i o n (IAEA) p r o c e d u r e / - ^ which c a l l s f o r sampling d a i l y f o r four days, weekly f o r eight weeks, monthly f o r s i x months, and semi-annually from then on. On leachatechanging day, a 10 mZ a l i q u o t was removed from the container and a c i d i f i e d to a pH o f 1 with concentrated n i t r i c a c i d . This a c i d was added to prevent adherence o f isotopes on the sample container w a l l s . A n a l y s i s o f the sample c o n s i s t e d o f : • gamma spectroscopy • separation o f cesium and s t r o n t i u m ^ • recount by gamma spectroscopy • beta counting o f separated 90sr • alpha energy a n a l y s i s . Bulk glass leach rates based on various isotopes were c a l c u l a t e d by the equation:

where:

°

2 R i · incremental leach r a t e , g/cm -day a · a c t i v i t y o f isotope i n leachate, sec-1 A · s p e c i f i c a c t i v i t y o f isotope i n sample, sec-1 -gram (based on measured a c t i v i t y of the h i g h l e v e l waste c a l c i n e ) , assumed to be homogeneous. A l l a c t i v i t i e s a and A were decay c o r r e c t e d to the same time. 2 S •geometric surface area of sample, cm t · leaching time, days. Q

0

Q

Q

DISCUSSION Figures 3a and 3b show logarithmic p l o t s o f the leach rate i n g/cm2-day based on various isotopes as a f u n c t i o n o f time. You can see that the leach rates decrease r a p i d l y at the beginning o f the t e s t and l e v e l o f f a f t e r about ten days. This behavior i s c o n s i s t e n t with previous studies on simulated waste glasses. P l o t s o f i s o t o p i c leach rates from the other waste glass sections show good agreement with those i n Figure 3a and 3b. This can be seen c l e a r l y i n Figures 4 and 5. Figure 4 shows -^^Cs and 54Mn leach rates representing sections from the top to the bottom o f the waste glass c y l i n d e r . Figure 5 p l o t s the same rates f o r ^ E u and 239+240p . y u can see that the leach r a t e s vary l i t t l e from top to bottom o f the waste c a n i s t e r . Although these four isotopes represent a wide range i n mass and 1 5

u

0


80 RADIOACTIVE WASTE IN GEOLOGIC STORAGE


5. BRADLEY

Leaching

1

Figure 3b.

of Waste

10

Ghss

81

100

Leach rate as a function of time—upper-middle ghss section

1000


82

RADIOACTIVE

WASTE

IN GEOLOGIC

STORAGE

LEACH TIME, Days

Figure 4. Leach rate as a function of time and ghss section for Cs and Mn: ( ), bottom section; (- · -), upper-middle section; ( ), top section. 137

54


BRADLEY

Leaching

of Waste Glass

10

100

1000

LEACH TIME, Days

Figure 5. Eu: ( 154

Leach rate as a function of time and glass section for + Pu and ), bottom section; (-'-), upper-middle section; ( ), top section. 239

240


84

RADIOACTIVE

WASTE

IN

GEOLOGIC

STORAGE

expected v o l a t i l i t y , they i n d i c a t e that there are no s i g n i f i cant inhomogeneities i n the glass column. This conclusion i s supported by recent gamma spectroscopy on the h i g h - l e v e l waste glass sections.^—^ The leach rates seen here can be compared to those from a s i m i l a r test on simulated waste glass of the same composition. The leach rates for cesium and strontium from the f u l l y r a d i o a c t i v e glass were the same as the leach rates of the simulated glass. Although both t e s t s were done on bulk glass samples, they d i f f e r e d i n c o n f i g u r a t i o n . The f u l l y r a d i o a c t i v e samples were disks [^0.3 cm (height) χ 3.5 cm (diameter)], and the simulated glass sample was a c y l i n d e r [0.73 cm (height) χ 0.98 cm(diameter)] . The s i z e of the sample being leached has a large e f f e c t on the apparent leach r a t e . D i f f e r e n c e s up to a f a c t o r of 100 were noted between -42+60 mesh p a r t i c l e s and bulk samples such as used i n t h i s t e s t . — In c o n t r a s t , r a d i a t i o n i s b e l i e v e d to have a very small e f f e c t on leach r a t e s ; t h i s has been shown for the case of alpha r a d i a t i o n . - Further studies on the e f f e c t s of r a d i a t i o n on leach rate are i n progress. Curves showing the cumulative f r a c t i o n leached were a l s o c a l c u l a t e d from the leaching data based on the equation:

cumulative f r a c t i o n = -τ A w ο where: a · a c t i v i t y of isotope i n leachate, sec-1 A ·specific a c t i v i t y of isotope i n sample, w ©sample weight, grams. Q

0

sec-1

Figure 6 shows these cumulative leach f r a c t i o n s from the top waste glass s e c t i o n . A ranking of the elements studied, based on the f r a c t i o n released a f t e r 639 days of s o l u t i o n contact, i s given i n Table I I . This order of leaching can a l s o be seen i n Figures 3a and 3b. TABLE I I . Element Cs Sr Co Sb Mn Pu Eu Cm Ce

Element Release F r a c t i o n s A f t e r 639 Days F r a c t i o n Released at 639 Days^ }

2.8 2.0 1.6 1.1 7.7 2.3 2.6 1.9 4.8

χ χ χ χ χ χ χ χ χ

2

10~ 10"~ 10~ 10"* 10~3 ΙΟ" 10~4 10~4 10~5

2

2

2

3

Source F i s s i o n product F i s s i o n product A c t i v a t i o n product F i s s i o n product A c t i v a t i o n product Actinide F i s s i o n product Actinide F i s s i o n product

(a) Averaged over the four glass s e c t i o n s .


BRADLEY

Figure 6.

Leaching of Waste Ghss

LEACH TIME, days

Cumuhtive fraction leached as a function of time—top ghss section

86

RADIOACTIVE

WASTE

GEOLOGIC

STORAGE

In s e v e r a l respects, t h i s ranking i s not unexpected; low charge, high m o b i l i t y ions l i k e Ca^ and S r should leach more e a s i l y than the rare earth and a c t i n i d e elements. Although plutonium and cerium have s i m i l a r p r o p e r t i e s as pure oxides, Plutonium was found to have a higher r e l e a s e rate than Cerium or Curium. Obviously, much more work i s needed to understand these d i f f e r e n c e s than was p o s s i b l e with t h i s study. F r a c t i o n a l r e l e a s e can be expressed using F i c k ' s Law; 2 +

+


IN

F

r

=

bT*

or l o g F^ = m log Τ + log b = m log Τ + Β where F · Isotope f r a c t i o n released Τ · Time Β · Intercept m · Slope of the l i n e . The slope i s i n d i c a t i v e of the type of r e l e a s e mechanism. A slope of 0.5 indicates a d i f f u s i o n - c o n t r o l l e d release; a slope of 1.0 i n d i c a t e s that a c o r r o s i o n - r e l a t e d mechanism i s operable. - » - » — The d i f f u s i o n release mechanism i s c h a r a c t e r ized by surface adsorption, ion exchange, and migration. Chemical c o r r o s i o n , or a l t e r a t i o n of the s i l i c a t e l a t t i c e , i s c h a r a c t e r i z e d by hydroxyl attack on s i l i c o n or by hydrogen attack on b r i d g i n g oxygens.^—^ The leachate sampling frequency had l i t t l e e f f e c t on leach rate u n t i l the semi-annual frequency was reached. The change i n leach r a t e , as r e f l e c t e d i n Figures 3a, 3b, 4, 5, and 6, was dramatic i n changing from monthly to semi-annual sampling. The apparent mechanism s h i f t e d from l a t t i c e a l t e r a t i o n to d i f f u s i o n r e l e a s e ; t h i s s h i f t i l l u s t r a t e s the important r o l e of d i s s o l v e d species i n lowering the leach r a t e . These r e s u l t s are c o n s i s tent with the work of E l - S h a m y ^ * d P a u l . i ^ From the curves of cumulative f r a c t i o n r e l e a s e , the general trend over the t o t a l t e s t i n g time shows two d i f f e r e n t slopes for each element. A l l of the elements have a slope l e s s than 0.50 at the beginning of the leach t e s t , i n d i c a t i n g a type of diffusion-controlled release. A f t e r a period of time, t h i s slope g r a d u a l l y approaches a value of 1, i n d i c a t i n g a combina t i o n of r e l e a s e mechanisms. This r e s u l t i s c o n s i s t e n t with discussions of release mechanisms i n the literature.^-»—^ Table I I I summarizes these r e s u l t s by element for the top sec t i o n of h i g h - l e v e l waste g l a s s . (

a n

5.

Leaching

BRADLEY

TABLE I I I .


Elements Mn Co Sr Sb Cs Ce Eu Pu Cm

of

Waste

87

Ghss

Slopes of I n i t i a l and Long-Term Release Mechanisms

I n i t i a l Slope/ Time Period, Days 0.31/1-8 0.33/1-8 0.24/1-8 0.47/1-30 0.40/1-8 0.12/1-100 0.31/1-20 0.21/1-8 0.39/1-150

Long-Term Slope/ Time Period, Days 0.84/8-639 0.81/8-639 0.76/8-639 0.67/30-639 0.91/8-639 0.70/100-639 0.66/20-639 0.64/8-639 0.72/150-639

The scope of t h i s work was not intended to include study and a n a l y s i s of leaching mechanisms beyond t h i s p o i n t , but the data do show that the mechanism i s dependent on time, sampling frequency, and type of element. A l s o , over the long t e s t i n g p e r i o d , two d i f f e r e n t mechanisms account for the r e l e a s e of m a t e r i a l . More work i s needed to increase our understanding of these h i g h - l e v e l waste g l a s s - s o l u t i o n i n t e r a c t i o n s . Release rate data from a c t u a l r a d i o a c t i v e waste forms i s needed to evaluate the s a f e t y of emplacing nuclear wastes i n geologic media. However, i n a d d i t i o n to waste form development s t u d i e s , such as the leach t e s t j u s t described, a comprehensive program was s t a r t e d to o b t a i n release data from candidate waste forms f o r geologic d i s p o s a l . This work i s the o b j e c t i v e of Task #2 of the Waste I s o l a t i o n Safety Assessment Program at P N L / — We currently are e v a l u a t i n g the e f f e c t of leach s o l u t i o n composition on r a d i o isotope r e l e a s e . Figure 7 shows the progression of experiments from simple waste form/solution i n t e r a c t i o n s to those combining the waste form with the containment, s o l u t i o n , b a c k f i l l , and rock media under hydrothermal c o n d i t i o n s . Studies using s i t e specific conditions are a l s o planned, along with possible l a r g e - s c a l e h o t - c e l l or i n s i t u t e s t s . The waste forms and experiments c u r r e n t l y being studied and planned are depicted i n Figure 8. CONCLUSIONS The r e l e a s e rate was determined for 10 r a d i o i s o t o p e s from f u l l y r a d i o a c t i v e waste glasses i n deionized water for a p e r i o d of 1.75 years. For cesium and strontium, good agreement e x i s t s between the leach rates f o r simulated and f u l l y radioactive glass of the same composition. The r e l e a s e rate mechanism i s dependent on time, sampling frequency, and type of element. For t h i s study, only sampling i n t e r v a l s greater than one month had s i g n i f i c a n t impact on the leach r a t e . Over the long t e s t i n g p e r i o d two d i f f e r e n t r e l e a s e


RADIOACTIVE

WASTE

IN

GEOLOGIC

FLOW & STATIC LEACH TESTS WASTE FORM + SOLUTIONS + ROCK MEDIA CONTINUOUS FLOW LEACH TEST WASTE FORM I + SOLUTIONS IAEA & STATIC LEACH TESTS WASTE FORM + SOLUTION SI

MECHANISM STUDIES

LEAST COMPLEX

LARGE SCAIE TESTS

IN-SITU TESTS

MOST COMPLEX

GENERIC STUDIES

SITE SPECIFIC STUDIES AUTOCLAVE STUDIES

Figure 7.

WISAP Task #2 leaching program

STORAGE


BRADLEY

Leaching

of Waste

Ghss

a) SOLUTION EFFECTS

a) SOLUTION EFFECTS

a) SOLUTION EFFECTS

b) TEMPERATURE

b) LOW TEMPERATURE

b) TEMPERATURE

(LOW AND HIGH) 0

FLOW EFFECTS

d) REACTION PRODUCT

0 d)

(LOW AND HIGH)

FLOW EFFECTS CONTAINMENT SYSTEM/GEOLOGY

c) FLOW EFFECTS d) REACTION PRODUCT

ANALYSIS

ANALYSIS

e) RELEASE MECHANISMS

e) RELEASE MECHANISMS

f) CONTAINMENT SYSTEM/

f)

GEOLOGY

CONTAINMENT SYSTEM/GEOLOGY

LARGE SCALE TESTS IN-SITU TESTS

Figure 8.

Waste form release experiment outline


90

RADIOACTIVE

WASTE

IN GEOLOGIC

STORAGE

mechanisms occurred. This i s c o n s i s t e n t with p r e v i o u s l y r e ported observations. S p e c i a l areas needing a d d i t i o n a l work a r e : • the e f f e c t of surface area to leachate volume r a t i o on leaching • the nature and importance o f p r o t e c t i v e f i l m s formed during leaching • s o l u t i o n saturation/chemical-back reactions • a c t i n i d e o x i d a t i o n states/bonding • radiation/radiolysis effects. Release rates o f r a d i o i s o t o p e s should be determined from a c t u a l nuclear wastes. The release rates of these isotopes must be measured under conditions of geologic storage. To obtain t h i s data, Task #2 o f the WISAP w i l l study release from waste forms under a v a r i e t y o f conditions to simulate geologic storage o f nuclear waste m a t e r i a l s .

ACKNOWLEDGMENTS The author appreciates the help o f Y. B. Katayama on the glass preparation; J . H. Westsik, J . E. Mendel, and R. P. Turc o t t e on the t e x t ; and J . C. Nelson, C. E. Bigelow, and J . R. Crockett f o r performing the h o t - c e l l work. LITERATURE CITED

1. M. J. B e l l , ORIGEN - The ORNL Isotope Generation and Depletion Code, ORNL-4628, Oak Ridge National Laboratory, Oak Ridge, TN 37830, May 1973. 2. J . E. Mendel, W. A. Ross, F. P. Roberts, Y. B. Katayama, J. H. Westsik, J r . , R. P. Turcotte, J . W. Wald, and D. J . Bradley, Annual Report on the Characteristics of High-Level Waste Glasses, BNWL-2252, Battelle, Pacific Northwest Laboratories, Richland, WA 99352, June 1977. 3. W. A. Ross, D. J . Bradley, R. L . Bunnell, W. J. Gray, Y. B. Katayama, G. B. Mellinger, J . E. Mendel, F. P. Roberts, R. P. Turcotte, J . W. Wald, W. E. Weber, and J . H. Westsik, Jr., Annual Report on the Characteristics of High-Level Waste Glasses, BNWL-2625, Battelle, Pacific Northwest Laboratories, Richland, WA 99352, May 1978. 4. D. J . Bradley, Leaching of Fully Radioactive High-Level Waste Glass, PNL-2664, Battelle, Pacific Northwest Laboratories, Richland, WA 99352, May 1978. 5. E. D. Hespe, ed., "Leach Testing of Immobilized Radioactive Waste Solids, A Proposal for a Standard Mehtod." Atomic Energy Review 9:1, 1971.

5.

BRADLEY

Leaching

of

Waste

91

Glass


6. P. Arthur and O. M. Smith, "Semi-Micro Qaulitative Analy sis," International Chemistry Series, McGraw H i l l , New York, NY, 1942. 7. Y. B. Katayama, Leaching of Irradiated LWR Fuel Pellets in Deionized and Typical Ground Water, BNWL-2057, Battelle, Pacific Northwest Laboratories, Richland, WA 99352, July 1976. 8. J . E. Mendel, W. A. Ross, F. P. Roberts, R. P. Turcotte, Y. Β. Katayama and J . H. Westsik, J r . , "Thermal and Radiation Effects on Borosilicate Waste Glasses." In: IAEA Sympo sium on Management of Radioactive Waste from the Nuclear Fuel Cycle, IAEA-SM-207/100, 2:49, Vienna, 1976. 9. L . L. Hench, "Leaching of Glass," Workshop on Ceramic and Glass Radioactive Waste Forms, Germantown, MD, January 1977. 10. R. W. Douglas and T. M. M. El-Shamy, "Reactions of Glasses with Aqueous Solutions," Journal of American Ceramic Soci ety, 50(1):1-8, January 21, 1967. 11. F. E. Diebold, Discussions of Glass - Water Interactions, ARH-2905, September 15, 1973. 12. T. M. M. El-Shamy, PhD. Sheffield, UK, 1966.

Thesis,

13. A. Paul, Chemical Durability of Approach, University of Sheffield,

University of

Sheffield,

Glasses, A Thermodynamic Sheffield, UK, 1977.

14. H. C. Burkholder, J . Greenborg, J . A. Stottlemeyer, D. J . Bradley, J . R. Raymond, and R. J . Serne, Waste Isolation Safety Assessment Program Summary of FY 1977 Progress, PNL -2451, Battelle, Pacific Northwest Laboratories, Richland, WA 99352, November 1977. RECEIVED January 16, 1979.

Radioactive Waste in Geologic Storage - American Chemical Society

Recommend Documents