Surface layer formation on corroded nuclear waste glasses

Surface layer formation on corroded nuclear waste glasses

Journal of Non-Crystalhne Solids 67 (1984) 245-264 North-Holland, Amsterdam 245 S U R F A C E LAYER F O R M A T I O N ON C O R R O D E D NUCLEAR W A...

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Journal of Non-Crystalhne Solids 67 (1984) 245-264 North-Holland, Amsterdam

245

S U R F A C E LAYER F O R M A T I O N ON C O R R O D E D NUCLEAR W A S T E GLASSES *

B.C. SALES, C.W. W H I T E , G . M . B E G U N ** a n d L.A. B O A T N E R Solid State Dwtston, Oak Ridge National Laboratory, Oak Ridge. TN 37830, USA

The corrosion of a borosillcate nuclear waste glass Js accompanied by the formation of an altered layer on the glass surface This layer has a chermcal composition that is radically different from that of the bulk glass and can be many rmcrons duck In ttus study, the combined techniques of Rutherford backscatterlng depth profile analysis, Raman scattenng, quanutative solution analysis, and solution conductivity measurements have been used to investigate the elemental distribution, chermcal stability, chenucal composition, and protective ability of the altered surface layers formed on several simulated nuclear waste glasses that were corroded in aqueous media The properties of altered glass surface layers formed by exposure to distilled H20, acidic, basic, and bnne solutions were investigated. Exposure to the different solutions resulted in significant variations in the characteristics of the corroded surface layer. In particular, the surface layers formed in one soluUon were frequently unstable when the glass was re-exposed to different corrosion conditions. Raman spectroscopy was employed in estabhslung that for glasses corroded in dlstdled H20, strontium IS present in the altered surface layer in the form of SrCO3

I. Introduction Borosilicate glass is one of the leading c a n d i d a t e materials for the p r i m a r y c o n t a i n m e n t of high-level radioactive waste. A l t h o u g h the c o m p o s i t i o n of the glass that might actually be used is currently unspecified, the present general p l a n for its application as a p r i m a r y waste form calls for radioactwe waste to be dissolved in the m o l t e n glass at relatively high temperatures. The m o l t e n borosilicate g l a s s - w a s t e mixture would be p o u r e d into metal cannisters which w o u l d be sealed a n d then transported to a n u n d e r g r o u n d storage repository for final disposal. O n c e in place, the most likely m e c h a n i s m b y which radioactive material could be removed from the glass a n d r e i n t r o d u c e d into the biosphere is through the i n t r u s i o n of g r o u n d water into the repository. Accordingly, achieving a n u n d e r s t a n d i n g of the m e c h a n i s m s of nuclear waste glass corrosion in aqueous m e d i a is critical to assessing the l o n g - t e r m safety of this particular solution to the p r o b l e m of n u c l e a r waste disposal. * Research sponsored by the Division of Materials Sciences, US Department of Energy under contract W-7405-eng-26with Union Carbide Corporation ** Chermstry Division, ORNL 0022-3093/84/$03.00 © Elsevier Science Pubhshers B V (North-Holland Physics Pubhshlng Division)

246

B C Sales et al / Surface layer formation on corroded nuclear wastes

The corrosion of a borosilicate nuclear waste glass is accompanied by the formation of an altered layer on the glass surface. This layer, which has a chermcal composition radically different from the bulk glass, can be many microns thick and is composed primarily of metal compounds that are relatively insoluble in certain pH ranges. As the borosllicate glass matrix dissolves, various metal cations (including waste cations) dispersed m the glass are free to go either into solution or to combine with the various anions in solution to form chemical compounds that make up the altered surface layer. In a static corrosion experiment large concentration gradients will exist adjacent to the glass surface and, hence, any relatively stable compounds that form will tend to precipitate back onto the surface of the corroding glass. Understanding the physical and chemical properties of the altered glass surface layer is important for two major reasons: First, the mechanism(s) of nuclear waste glass corrosion are reflected in the composition and spatial structure of this layer (as well as in the concentration of corrosion products present in the corroding solution). Second, this layer may, in some cases, provide a barrier that slows the attack of the underlying glass by the corroding solution. In this study, the combined techniques of Rutherford backscattering depth profile analysis (RBSDPA), quantitaUve solution analysis, Raman scattering, and solution conductivity measurements have been used to investigate the elemental distribution, chemical stability, chemical composition, and protective ability of the altered surface layers formed on several borosilicate simulated nuclear waste glasses that were corroded in aqueous media.

2. Experimental 2.1. Glass preparatton

Three "reference" nuclear waste glass compositions were investigated. These were: Frlt 21 containing 20 wt% simulated Savannah River defense waste (SRW), Frlt 131 containing 29 wt% SRW, and Frit 21A [1-3]. The compositions of these three glasses are given in table 1. The composition of the Frit 21A glass is identical to the Frit 21 composition except that CaO was not added. Frequently an additional 3 wt% of either UO 2, Cs20, or SrO was included to facilitate detection of these important nuclear waste elements in the altered surface layer as well as in the leachate. The appropriate amounts of all oxides, carbonates, or sulfates were thoroughly mixed and then melted in a platinum crucible at 1160°C for 3 h. The molten glass was poured into a heated mold of spectroscopically pure carbon, annealed at 550 °C for 2 h, and then slowly cooled to room temperature. Samples measuring 1 × 1 × 0.2 cm were cut from the resulting glass billet with a diamond saw which used ethylene glycol for cooling instead of water. For the ion scattering and some Raman measurements one side of each sample was polished to a 1 micron finish with A1203.

B C Sales et al / Surface layer formanon on corroded nuclear wastes

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Table 1 Glass compos~tlons used tn mvesttgatmg surface layer formation on corroded nuclear waste glasses .Compound

Frtt 21 + 20 wt% SRW

F n t 131 + 29% SRW

F n t 21A

StO 2 B203 Na20 TIO 2 CaO Ll 2° Fe203 AI203 MnO 2 NiO NaESO 4 La203 ZrO 2 MgO

42 0 80 16.35 80 4.8 3.2 10.9 2 15 30 1.35 0 25 -

41 1 10.4 14.8 07 12 4.0 15 8 31 4.4 19 0.4 04 04 14

55.2 10 5 19.5 10 5 4.3

2.2. Glass corrosion The glass samples were corroded in a static solution at 90 o C in carefully cleaned Teflon containers for times ranging from 0-36 days. The ratio of the surface area of the glass sample to the volume of corroding liquid was fixed at 0.1 c m - 1. The effects of four different corroding liquids on the glass samples were investigated. These were: triply distilled water, a 0.01 N HCI solution (pH = 2), 0.01 N N a O H solution (pH = 12), and brine (a nearly saturated NaCI solution). -

2.3. Rutherford backscattering depth profile analysts (RBSDPA) The general theory of Rutherford backscattering is described in several excellent review articles [4,5]. The application of the technique of RBSDPA to the study of nuclear waste glass corrosion has been discussed in previous articles [6-8] so that only a brief description will be given here. After allowing the corroded glass sample to dry in air, a 10 nm thick A1 film was evaporated onto the polished surface to prevent charging during the subsequent backscattering measurements. The Al-coated glass was then placed in a vacuum chamber and bombarded by a 2 MeV beam of He + ions. By analyzing the energy distribution of the scattered He + ions which arrive at a solid state detector placed at a scattering angle of - 1 6 0 °, quantitaUve concentration versus depth profiles were determined for many of the elements in the glass. Concentration profiles could be probed to a maximum depth of 1-2 # m below the surface with a depth resolutxon of 0.01-0.02 btm.

248

B C Sales et a l /

Surface layer formatton on corroded nuclear wastes

2.4. Solutton analyses

At the conclusion of a given corrosion period each solution was analyzed for the following elements: Si, B, Na, Fe, Mn, Ni, Ca, A1, Li, Sr, U, TI, and in some cases Cs. With the exception of the U concentraUon which was measured using fluorimetry and the Cs concentration which was measured w~th atomic absorption spectroscopy, the concentrations of the remainder of the elements in solution were determined using inductively coupled plasma (ICP) emission analysis. To facilitate comparisons with solution data from other research groups, the concentration of each element found in solution was converted to a normalized corrosion rate with units of g / m 2 d. The corrosion rate R was determined using the following equation: R = C,V M, St '

(1)

where C, is the concentration (g/ml) of element t found in solution, V the volume of solution (ml), S the geometric surface area (m 2) of the glass, t the time (days) the sample was exposed to the solution, and M, the mass fraction of element t present in the uncorroded glass. 2.5. Raman scattermg

The Raman spectrometer and ancillary equipment have been described in detail elsewhere [9,10]. Briefly, a Ramanor HG-2S spectrometer (Instruments SA.) equipped with a cooled photomultiplier tube and pulse counting electronics was used to obtain the Raman spectra. A Nicolet 1170 signal averager was used to control the spectrometer. Multiple scans were made when necessary and the spectra were stored on magnetic tape. Raman spectra were excited by means of the 514.5 nm line of a Spectra-Physics model 164 argon-ion laser. The glass samples (1 × 1 × 0.2 cm) or small portions of the powder produced by the corrosion were placed at the focus of the laser beam and Raman spectra were observed at 90 o. Polarization measurements were made by rotating the plane of polarization of the exciting laser beam by 90 o. 2.6. Solutton conductivtty analysis

As a nuclear waste glass (or similar material) corrodes in distilled water, mobile ions are released into solution resulting in an increase in the solution conductivity. For total ionic concentrations lower than - 1 0 -3 mol./1, the effects of ion-ion interactions on the solution conductivity can be neglected and the activity coefficients of all ions taken to be unity. In this dilute limit, the solution conductivity, o, is then: o = Y'.lq, la, N,,

(2)

B C Sales et al / Surface layer formatton on corroded nuclear wastes

249

where q, is the charge of the anion or cation in solution, N, is the concentration in m o l . / c m 3 of the t th type of ion in solution, and o, is the ionic conductance per charge of the ith ion m the dilute limit. With the exception of the H ÷ and O H - ions, the values for o, at 25 °C of almost all cations and anions m solution are between 40 and 80 cm 2 m h o / m o l . [11]. The concentration of the H ÷ and O H - ions, however, can be determined directly by measuring the solution pH. When the total ionic concentration m solution is greater than about 10 -2 mol./1, the effects of ion-ion interactions m solution must be considered as has been done, for example, in the Debye-HiJckel theory [12]. If the average charge of the ions in solution is known, then eq. (2) can be used to estimate the total ionic concentration. For example, when a glass waste form is corroded, a reasonable charge to assume for the ions in solution is 2. Thus, if the solution conductivity rises by 100 ~tmho/cm at room temperature, an average value for o, of 60 cm 2 m h o / m o l , can be assumed, which lmphes a total concentration of ions m solution of 0.33 × 10 -6 m o l . / c m 3 (i.e., 0.33 x 10 -3 mol./l). Because of the weak variation of o, from ion to 1on, this estimate should be accurate to within a factor of 2. The solution conductivity measurements were made using a standard Teflon corrosion container that was modified slightly so that two 0.55 m m diameter Pt wires could be inserted in the solution. The conductance was measured with a YSI Model 32 conductance meter, and the cell constant for each Teflon container was determined using a solution of known conductivity (10 -2 N, KCI at 25 o C). A change in the solution conductivity of 0 . 1 / t m h o / c m could be easily measured. This corresponds roughly to a 50 parts per billion change in the total ionic concentration. 3. Results and discussion 3.1. Elemental composttton of surface layer 3.1.1. Frtt 21 + 20 wt% S R W +

3 wt% UO 2

Four samples of a uranium-doped simulated nuclear waste glass (Frit 21 containing 20 wt% SRW plus an additional 3 wt% of U O 2) were exposed for 6 h at 90 o C to either distilled water, a 0.01 N HCI (pH 2) solution, a 0.01 N N a O H (pH 12) solution, or a brine solution. The high energy portion of the backscattering spectra for the four glass samples corroded in these four solutions is shown in fig. 1 (closed circles) along with the backscattering spectrum for an uncorroded glass (open circles). When this glass composition was corroded in distilled water (see fig. la), the surface concentrations of U, Fe, Ti, and most transition metals were two to four times larger than their concentrations in the bulk glass. The concentrations of N a and Si in the surface region, however, are much lower than in the bulk glass (note that the backscattering data for N a and Si are not shown in fig. 1). These findings are in agreement with our previous RBS data on similar glasses which were corroded in distilled water [6-8]. The normahzed concentrations of U, Ti, Fe,

B C Sales et al / Surface layer formatwn on corroded nuclear wastes

250 3 a)

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Fig 1 Energy spectra of 2.0 MeV 4He+ ions backscattered from uncorroded (©) and corroded (o) sxmulated borosdicate nuclear waste glass (Fnt 21 + 20 wt% SRW + 3 wt% UO2). Each glass sample was corroded for 6 h at 90 o C m either (a) dLstdledwater, (b) 0.01 N HCI (pH = 2), (c) 0.01 N NaOH (pH = 12), or (d) brine. The arrows denote the onset energy for scattenng from U, Fe, Mn, Tn, and Ca. A comparison of the backscattenng results shown m (a) and (d) shows the quahtatlve stmdantnesof the altered surface layer for glass samples leached for this time period m &stdled water or brine. In both cases, enhanced concentrations of U, Fe, Mn. and T] are present m the surface layer. The results shown m (b) for leaching m a pH 2 solution clearly illustrate the selectnve removal (Le., leaching m the tradmonal sense) of uranium from the nuclear waste glass. M n , a n d Ni in solutions as d e t e r m i n e d b y q u a n t i t a t i v e analysis (table 2) were f o u n d to be m u c h lower than those of Si, Na, a n d B. These results are in good qualitative agreement with the backscattering data. W h e n glass with this composition is corroded in a n acidic solution, the c o m p o s i t i o n of the surface layer changes dramatically relative to that of the surface layer formed b y corrosion in distilled water. I n particular, in the acidic solution, u r a n i u m is selectively leached from the glass to such an extent that n o u r a n i u m was detected within 1 # m of the surface (see fig. lb). This effect is shown more clearly in fig. 2 where the U c o n c e n t r a t i o n versus depth profiles are directly c o m p a r e d for an u n c o r r o d e d glass a n d a glass corroded in either distilled water or a p H 2 (0.01 HC1) solution. After corroding the glass, q u a n t i t a t i v e analysis of the p H 2 solution indicates that the U, Ti, Fe, a n d M n

B C Sales et al. / Surface layer format:on on corroded nuclear wastes

251

Table 2 Normahzed corrosion rates (g/m 2 d) of vanous elements present m borosihcate glass containing SRW. The glass specimens were all corroded for 6 h at 90 o C m either dtstdled water, pH 2, or pH 12 solunon Element a)

Water

(0.01 N HCI)

Sl B Na Tl L1 Ca Fe Mn Nl A1 Sr U

4.0 45 51 < 0.01 51 25 < 0.05 < 0 04 < 0.04 3.1 2.4 0 15

4.4 9.9 10.7 28 11 0 9.5 46 8.7 8.5 8.1 8.5 11.0

(0 01 N NaOH) 12.4 13.7 < 0.0009 15.1 1.9 0.33 < 0 02 < 0 16 15.1 4.1 73

a~ The values shown for all elements except U and Sr represent average corrosion rates determined using the three compositmns: Fnt 21 + 20 wt% SRW + 3 wt% of either UO2, SrO, or Cs20. The values shown for U and Sr, however, were obtamed usmg only the glasses doped with 3 wt% UO 2 and SrO, respectively.

c o m p o u n d s , which p r e c i p i t a t e d on the surface of the glass c o r r o d e d in distilled w a t e r are, in fact, soluble in the p H 2 solution. It is interesting to note, however, that the b a c k s c a t t e r i n g results (fig. l b ) show that there is still a thin o u t e r layer ( - 5 0 n m thick) which c o n t a i n s increased (relative to the b u l k c o m p o s i t i o n ) c o n c e n t r a t i o n s of Fe, M n , a n d Ti. In the p H 2 solution, s o d i u m is largely d e p l e t e d f r o m the surface layer b u t the silicon c o n c e n t r a t i o n r e m a i n s close to that of the b u l k value for this glass. I n a b a s i c solution ( p H 12) the b a c k s c a t t e r i n g results shown in fig. l c i n d i c a t e that a surface layer is f o r m e d whose c o m p o s i t i o n is similar to that of the b u l k glass. T h e solution analysis d a t a (table 2), however, i m p l y that the surface layer s h o u l d be rich in M n , Fe, a n d Ti since the c o n c e n t r a t i o n s of these e l e m e n t s in the leachate solution is very small. A l t h o u g h it is p o s s i b l e that these t r a n s i t i o n m e t a l s are p r e s e n t in solution in the f o r m of colloids, a m o r e likely e x p l a n a t i o n is that p a r t o f surface layer s i m p l y flaked off the surface d u r i n g the d r y i n g process since this effect has b e e n o b s e r v e d b y us for thicker surface layers. This conclusion is also s u p p o r t e d b y the b a c k s c a t t e r i n g d a t a for the S r - d o p e d glass (see fig. 3c) which shows that when the glass is c o r r o d e d in a p H 12 solution, the surface layer b e c o m e s highly e n r i c h e d in Fe, M n , a n d Ti as expected f r o m solution analysis. A s shown in fig. l d , c o r r o d i n g the glass in a b r i n e solution results in an altered glass surface layer whose c o m p o s i t i o n is similar to that f o r m e d after c o r r o d i n g the glass in distilled water. I n general, elements which have c o r r o s i o n rates larger t h a n Si (the m a j o r c o n s t i t u e n t of the glass) are d e p l e t e d f r o m the surface layer (e.g., U in a p H 2 solution) while elements which have c o r r o s i o n -

252

B C Sales et aL / Surface layer formation on corroded nuclear wastes

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Fig. 2. U r a m u m concentration versus depth profdes from a simulated nuclear waste glass ( F n t 2 1 + 2 0 wt% S R W + 3 wt% UO2) corroded for 6 h at 9 0 ° C m dlstdled water (e) or a pH = 2 solution (I). For comparison, the U concentration versus depth profile from an uncorroded glass ( O ) is also shown. The exact depth to wluch the u r a m u m has been removed cannot be deterrmned due to the onset of scattering from Fe and M n m the layer (see fig. lb).

rates much smaller than Si (e.g., Fe in H 2 0 ) are concentrated in the surface layer. As noted above, the apparent disagreement between the p H 12 backscattering results and the pH 12 solution data is most likely due to physical loss of the surface layer as a result of flaking of the leached glass specimen during drying. An important conclusion resulting from the solution analyses is that the uranium compound which precipitates on a glass corroded in distilled water does not form on a glass corroded in either a basic or acidic solution. Hence, at least for short corrosion periods, U is only selectively retained on the glass surface when the material is corroded by a solution with a pH in the neutral region. 3.1.2. Frit 21 + 2 0 wt% S R W + 3 wt% SrO

Four samples of a strontium doped simulated nuclear waste glass (Frit 21 + 20 wt% S R W + 3 wt% SrO) were exposed for 6 h at 9 0 ° C to either distilled water, a 0.01 N HC1 (pH 2) solution, a 0.01 N N a O H (pH 12) solution, or a brine solution. The high energy portion of the backscattering

B C Sales et al

/

Surface layer formation on corroded nuclear wastes

253

spectra for each of the four corroded glasses is shown m fig. 3 (closed circles) along with the backscattering spectrum for an uncorroded glass (open circles). After corroding this glass composition in distilled water (fig. 2a), some depleUon of Sr, Na, and Si is apparent within 0.1 /~m of the surface. The Fe, Mn, and Ti concentrations, however, are 2-4 times larger in the surface layer than m the bulk glass. It is interesting to compare these results with some of the earlier data given in ref. [6]. The results shown in fig. 4 of ref. [6] were obtained when a glass consisting of Frit 21 plus 5 wt% SrO was corroded in distilled H20 for various time periods. This glass did not contain the 20 wt% simulated SRW that is present m the material used in the current work, and the behawor of strontium in the surface region is apparently completely different in the two different glass compositions. A comparison of the RBS results shown in fig. 4 of ref. [6] after 2.5 h of leaching with the RBS data

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Fig. 3 Energy spectra of 2 0 MeV 4He+ ions backscattered from uncorroded (©) and corroded (O) simulated borosdlcate nuclear waste glass (Fnt 21 + 20 wt% S R W + 3 wt% SrO) Each glass sample was corroded for 6 h at 90 o C in either (a) dlstdled water, (b) 0.01 N HCI (pH = 2), 0 01 N N a O H (pH = 12), or (d) brine The arrows denote the onset energy for scattering from Sr, Fe, Mn, TI, and Ca. Strontmm ts obviously depleted m the surface layer following leaclung in either distilled H 2 0 or b n n e and is essentially totally removed from the surface by ieaclung m a pH 2 solution (b). Large surface concentrations of Fe, Mn, TI, and Ca are produced when the glass is leached in a pH 12 solution as shown in (c)

B C Sales et al.

254

/

Surface layer formation on corroded nuclear wastes

presented here in fig. 3a shows that the strontium concentration at the surface is increased for the Frit 21 + 5 wt% SrO glass, while strontium leaches from the surface of the Frit 21 glass containing 20 wt% SRW plus 3.0 wt% SrO. This difference in the corrosion behavior of the two glasses is attributed to the more rapid increase in the solution pH that occurs as the Frit 21 + 5 wt% SrO glass corrodes. As was the case for U in the U-doped glass, when the Sr-doped glass is corroded in an acidic solution (pH 2) for 6 h no Sr was detected within 0.5-0.75/~m of the glass surface - i.e., in this surface region all of the Sr has been selectively leached from the glass in the acidic environment. No measurable increase of the Fe and Mn concentrations was found in the surface layer but about a twofold increase in the Ti concentration was observed within 50 nm of the surface. In fig. 4, the Sr concentration versus depth profiles are directly compared in detail for an uncorroded glass, glass corroded in distilled water, and a glass corroded m a pH 2 solution. After exposing the Sr-doped glass to a basic corroding solution (pH 12) for 6 h (fig. 3c), the altered surface layer was enriched in Fe, Mn, and Ti but was depleted in both Si and Na. The Sr and Ca concentrations in the surface layer were roughly the same as those characteristic of the bulk glass. The composition of the surface layer formed after corroding the glass in a brine solution (fig. 3d) is qualitatively simdar to that found after corroding the

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Fig. 4. Strontium concentration versus depth profiles for a simulated nuclear waste glass ( F n t 21 + 20 w t ~ SRW + 3 wt% S t • ) corroded for 6 h at 90 o C m distilled water (O) or a pH = 2 soluuon (m). The Sr concentraUon versus depth profile for an uncorroded glass ( 0 ) is also shown for comparison. A comparison of the Sr and U (see fig. 2) profiles following a 6 h exposure to distilled H 2 0 shows the opposite behavior of these two waste glass constituents under these cond~tlons.

B C Sales et al / Surface layer formatton on corroded nuclear wastes

255

glass in distilled water. One major difference associated with corroding a borosilicate glass in a concentrated brine solution versus distilled water, however, is that in brine the p H of the solution is essentially fixed while m distilled water the p H steadily increases to a value between 9 and 10 as alkali ions are released from the corroding glass. This difference is particularly important with regard to Sr which tends to precipitate as SrCO 3 (see section 3.3) at the higher p H ' s but is readily soluble (and hence very mobile) at the p H characteristic of brine. 3.1.3. Frtt 21 + 20 wt% S R W + 3 wt% Cs20 A sample of simulated S R W glass doped with an additional 3 wt% of Cs was exposed to distilled water at 90 o C for 6 h. The total backscattering spectra for the corroded (closed circles) and uncorroded (open circles) Cs-doped glass are shown in fig. 5. The surface layer of the corroded glass is obviously enriched in T,, Mn, and Fe, but is depleted of Na, Cs, and Si. The narrow peak in the backscattering spectra at about 1.1 MeV is due to the 100 ,~ AI film which was evaporated on the glass to prevent charging during the RBS measurement. Detailed concentration versus depth profiles for Cs are shown in fig. 6 for an uncorroded and corroded Cs-doped glass. 3.2. Chemwal stabdity o f the altered surface layer

In order to assess the chemical stabihty of the various surface layers which form on borosilicate glass under different corrosion conditions, several two-

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256

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stage corrosion experiments were performed. Samples of Frlt 21 + 20 wt% SRW containing 3 wt% of either UO2, SrO, or Cs20 were initially exposed for 6 h at 90 o C to either distilled water, a 0.01 N HC1 solution, or a 0.01 N N a O H solution. During this initial corrosion step, an altered surface layer was produced on the glass. The corroded glass was then removed from the first solution and placed in a second solution with a different chemical composition for 1 h. Three two-stage corrosion experiments were performed as follows: In one set of experiments, doped borosilicate-SRW glass samples were first corroded for 6 h in distilled H 2 0 and were then corroded m an acidic (pH 2) solution for 1 h. In the second set of experiments, glass specimens were initially corroded in a pH 12 solution for 6 h and then exposed to a pH 2 solution for 1 h. The third set of experiments was essentially a reversal of the second set in that the samples were first corroded in a pH 2 solution for 6 h and then in a pH 12 solution for 1 h. After all three types of two-stage corrosion, the solutions from the 1 h corrosion experiments were analyzed quantitatively for the dements shown in table 3. By comparing the corrosion rates from these experiments with those obtained when an altered layer was not "pre-grown" on the glass (table 2), two general conclusions can be drawn. First, the transition-metal-rich layer which forms on the glasses corroded in either

B C Sales et al / Surface layerformanon on corroded nuclear wastes

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Table 3 Solution analyses results from three two-stage corrosion experiments using borosdtcate glass contamlng SRW. Each glass was exposed for 6 h at 90 o C to etther a dtstdled water, pH 2, or pH 12 solution. The glass was then removed from the first soluuon and placed m a second solution with a different chenucal composmon for 1 h. The normalized corrosion rates (g/m2/d) determined from the elements present m the 1 h leachate are also given below Element a)

St B Na Tt L1 Ca Fe Mn N1 AI Sr U

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8 85 16.0 18.0 20 22 0 78.0 95 48 0 82 0 16 3 67.0 80.0

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a) The values shown for all elements except U and Sr represent average corrosion rates determined using the three composlUons:Frlt 21 + 20 wt% SRW + 3 wt% of etther UO2, SrO, or Cs20. The values shown for U and Sr, however, were obtained using only the glasses doped with 3 wt% UO2 and SrO, respectwely.

distilled water or a basic solution (pH 12) is rapidly removed when the glass is exposed to a n acidic solution ( p H 2). Second, the silicon-rich layer wbach forms o n glasses corroded in a n acidic solution ( p H 2) is rapidly dissolved when the glass is placed in a basic solution (pH 12). 3.3. Chemtcal composmon o f the surface layer As n o t e d m the i n t r o d u c t i o n , the c o n c e n t r a t i o n of certain elements in the corroded surface layer is a p p a r e n t l y a result of the f o r m a t i o n of insoluble metal c o m p o u n d s o n the glass surface. C h a n g i n g the p H of the corroding solution can change the solubility limit for various metal c o m p o u n d s b y several orders of m a g n i t u d e . C o n s i d e r i n g the solubility of the most c o m m o n metal c o m p o u n d s (carbonates, hydroxides, hydrated oxides, borates, nitrates, etc.) relative to the solubility of silica at a particular p H (temperature fixed), if the solubility limit of the c o m p o u n d is greater t h a n that of silica, the c o m p o u n d will go into aqueous solution rather t h a n precipitate o n the glass surface. F o r each p H , t h e solubility limit c a n be calculated for each potential c o m p o u n d and, when the solubility limit for a particular c o m p o u n d falls below that of silica, this c o m p o u n d will precipitate o n the surface of the glass leading to the f o r m a t i o n of a n altered surface layer.

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Surface layer formatzon on corroded nuclear wastes

When borosilicate nuclear waste glass is corroded m a fixed volume of distilled water, the solution p H steadily increases to a final value of between 9 and 10. Using the available thermochemical data and following the analysis of G r a m b o w [13], the most likely insoluble compounds formed on a borosilicate glass corroded in distilled water are: Fe(OH)3, Mn(OH)z or MnCO3, SrCO3, C a C O 3, Ti(OH)4, and U O 3 or U O 2 (OH)2. The technique of Raman spectroscopy was used in a search for specific chemical compounds which could form on the surface of corroded borosilicate glasses. The glasses investigated were Frit 21A (see table 1) with and without 5 wt% SrO. These simpler glasses were studied because of the &fficulty associated with interpreting the R a m a n spectra from the simulated nuclear waste glass (i.e., Frit 21 + 20 wt% SRW) which contains 20 to 30 different elements. The Raman spectrum for an uncorroded slice of Frit 21A is shown in fig. 7. Many of the details of this spectrum are not understood, but the large peak at about 870 cm-1 is associated with a T i - O vibration and the shoulder at about 1046 cm -1 is probably due to nonbridging oxygen bonds [14]. The R a m a n spectrum of an uncorroded slice of Frit 21A + 5 wt% SrO is essentially the same as the R a m a n spectrum of Frit 21A. Specimens of Frit 21A and Frlt 21A + 5 wt% SrO were corroded in distilled water at 90 ° C for 36 days. After removing the specimens from the water, a thick white, powder-like layer was evident on the surface. This layer was very weakly attached to the surface and could be removed with a stream of air. I

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Surface layerformanon on corroded nuclear wastes

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During the corrosion period the pH of the aqueous solutions for each sample increased to values between 9.9 and 10.3, and the final weight of each specimen was about 9% less than the initial weight. The Raman spectrum from the surface layer of each glass is shown in fig. 8. The Raman spectrum from the corroded Sr-doped glass exhibits a sharp peak at approximately 1090 cm -], but a similar peak is not observed in the spectrum for pure Fnt 21A. This peak is shown more clearly in fig. 9 where I

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only the portion of the Raman spectrum near 1100 cm -a is shown for each glass. The position of the peak observed at 1090 cm-~ for the altered surface layer of the Sr-doped glass is the same as the frequency of the intense symmetric vibration mode for the CO 3 ion. These data confirm that Sr is present in the altered layer in the form of SrCO 3 - as expected from the solubihty analysis of Grambow [13]. The Raman spectrum of the corroded Frit 21A specimen in the spectral region below 800 cm-a exhibits much sharper peaks than would be expected for a glass. These sharper features probably indicate the mitxal formation of crystalline material in the surface layer. The formation and identification of crystalline phases on nuclear waste glasses corroded in aqueous solutions at higher temperatures (150-300 °C) have been well documented [15,16].

3.4. Protective abthty of the surface layer The normalized corrosion rates of elements from most borosllicate nuclear waste glasses corroded in distilled water decrease with time [17,18]. For many of the elements in the glass, most of the decrease is due to the approach of the

B C Sales et al

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Surface layerformatton on corroded nuclear wastes

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solubility limit for a particular compound [e.g., Fe(OH)3]. For other elements in the glass such as Na, Ll, B, and to a lesser extent Si, a decrease in the release rate is not due to a solubihty limit and may be related to a diffusion barrter provided by the transttion-metal-rich surface layer which forms on the corroded glass. In order to determine how effective this surface layer is in protecting a prototype nuclear waste glass, several samples of Frit 131 + 29 wt% SRW (table 1) were precorroded in distilled water at 90°C for various lengths of time. During this precorrosion period, an altered layer was produced on the surface of the glass. The precorroded glass with its altered surface layer intact was then re-exposed to fresh distilled water, and the total release of ions into solution was momtored using the solution conductance techmque. The solution conductivity vs time curves for the precorroded glasses were then compared to the conductivity curve of an uncorroded piece of the same glass. The variation of the solution conductivity with time is shown in fig. 10 for uncorroded and precorroded glass samples. As is evident from fig. 10, for glass specimens that were precorroded for longer times, the subsequent total release of ions into solution is lower. After precorroding the glass for 30 days, the subsequent corrosion rate (as determined from the total release of ions into solution after 2 days of corrosion in fresh distilled water) is about four times smaller than the imtial corrosion rate. These results are m good agreement with the findings of Wallace and Wicks [18] who concluded that, for the same glass (Frit 131 + 29% SRW) corroded under the same conditions, the decrease in the release rates of B and other highly soluble ions is due to the diffusion barrier provided by the surface layer. It should be emphasized, however, that for other nuclear waste I

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Surface layer formatton on corroded nuclear wastes

glasses with very similar compositions (i.e., P N L 76-68 glass and Frit 21 + 20% SRW [19]) the altered layer is not nearly as effective in reducing the corrosion rate, and other effects such as those included in the sorbed reaction product model [20] may be dominant. At present it is not possible to predict a p r t o r t how effective an altered layer will be in shielding the underlying glass from a corroding solution.

4. Summary The techniques of Rutherford backscattering depth profile analysis, quantitative solution analysis, R a m a n scattering, and solution conductivity analysis have been combined to investigate the elemental composition, chemical stability, chemical composition, and protective ability of the altered surface layer that formed on several simulated nuclear waste glasses (table 1) corroded in aqueous media. The major findings of this study are: (i) Glasses corroded in distilled water or in a basic 0.01 N N a O H (pH = 12) solution develop surface layers that are rich in relatively stable transition metal compounds [e.g., Fe(OH)3, Mn(OH)2 , and Ti(OH)4 ]. (il) Glasses corroded in an acidic 0.01 N HC1 (pH = 2) solution develop surface layers that are depleted in B and all metal elements (except perhaps Ti and Fe) but are rich in Si. (iii) The transition-metal-rich layer, which forms on glasses corroded in distilled water or a basic solution, is rapidly removed when the glass is exposed to an acidic solution (pH 2). (iv) The silicon-rich layer which forms on glasses corroded in an acidic solution (pH 2) is rapidly dissolved when the glass is placed in a basic solution (pH 12). (v) For simulated nuclear-waste-containing glasses corroded in distilled water, Sr and Cs are selectively leached from the glass while U is selectively retained in the surface layer as a precipitate of a relatively insoluble U compound. (vi) For glasses corroded in a basic solution, Cs and U go into solution at about the same rate as the borosilicate matrix elements (Si, B), but Sr will eventually precipitate on the glass surface (or in the container) as SrCO 3. (vii) The elements U, Sr, and Cs are all selectively leached from glasses corroded in a 0.01 N HC1 solution (pH = 2). For example, corroding a U-doped glass in a p H 2 solution for 6 h resulted in a surface layer in which no uranium was detectable within 1/~m of the surface (see fig. lb). (viii) The uranium compound, which precipitates on a glass corroded in distilled water, does not form on a glass corroded in either an acidic or a basic solution. Hence, at least for short corrosion periods, uranium is only selectively retained by the glass when the glass is corroded by a solution with a p H in the neutral region.

B C Sales et al / Surface layer formatwn on corroded nuclear wastes

(ix)

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263

The Raman spectra from simpler borosilicate glasses corroded in distilled water clearly show the presence of the CO 3 ion in conjunction with the presence of Sr in the glass. These data confirm the presence of S r C O 3 in the surface layer as expected from solubility calculations. The Raman spectrum of the corroded Frit 21A specimen in the spectral region below 800 cm -1 exhibits much sharper peaks than would be expected from a glass. These relauvely sharp features are probably due to the initial formation of crystalline material in the surface layer. Solution conductivity measurements on specimens of Frit 131 + 29 wt% SRW precorroded in distilled water indicate that the drop in the normalized corrosion rates with time of elements such as B is due to the presence of the transition-metal-rich surface layer. For other nuclear waste glasses with smular compositions, however, (i.e., PNL 76-68 glass and Frit 21 + 20 wt% SRW) the altered layer ts not so effective in reducing the corrosion rate of the underlying glass.

It is a pleasure to acknowledge Joe Stewart and Dave Heine of the O R N L Analytical Chenustry Division for their interest and assistance with the work, J.O. Ramey for her help with the illustrations, and LazeUe Tyler for her efforts in preparing the manuscript.

References [1] M J. Plodmec, in: Scientific Basis for Nuclear Waste Management, Vol 1, ed., G J. McCarthy (Plenum, New York, 1979) p. 31. [2] M.J. Plodmec, in: Scientific Basis for Nuclear Waste Management, Vol. 2, ed., C.J M. Northrup Jr (Plenum, New York, 1980) p. 223. [3] G . G Wicks, W C. Mosely, P.G Wlutkop and K.A. Saturday, J. Non-Crystalline Sohds 49 (1982) 413. [4] L C. Feldman, C R C Reviews in Solid State and Materials Science (May 1981) p 142 [5] W K Chu, J.W Mayer, M.-A Nicolet, T.M. Buck, G. Ansei and F. Elsen, Tlun Sohd F d m s 17 (1973) 1 [6] B.C Sales, L A Boatner, H Naramoto and C W. Wlute, J Non-Crystalhne Sohds 53 (1982) 201 [7] B C Sales, L A Boatner, H. Naramoto and C.W. Wlute m. SclenUflc Basic for Nuclear Waste Management, Vol. 4, e d , S.V Topp (Elsevier-North Holland, New York, 1982) p. 83. [8] B.C Sales, C.W. Wlute and L.A. Boatner, Mater. Lett., m press [9] G M. Begun, G W Beal, L.A. Boatner and W.J. Gregor, J. R a m a n Spectrosc 11 (1981) 273. [10] G M Begun and C.E. Bamberger, J. R a m a n Spectrosc. 13 (1982) 284 [11] G Milazzo, Electrochelmstry (Elsevier, New York, 1963) p. 60. [12] J O ' M Bockns and A K.N. Reddy, Modern Electrochemistry, Vol. 1 (Plenum, New York, 1977). [13] B. Grambow, in: Scientific Basis for Nuclear Waste Management V, Vol. 11, ed., W Lutze (Elsevier-North-Holland, New York, 1982) p. 93 [14] T Furakawa, K.E. Fox and W.B. Wbate, J. Chem. Phys 75 (1981) 3226. [15] J.K. Bates and M.J. Stemdler, m: SclenUflc Basis for Nuclear Waste Management VI, Voi 15, ed., D.G. Brooions (Elsevier-North-Holland, New York, 1983) p. 83. [16] W Lutze, G.Malow and H. Rabe, in. SclenUflc Basis for Nuclear Waste Management VI, Vol 15, e d , D G. Brookms (Elsevier-North-Holland, New York, 1983) p. 37.

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[17] D.E. Clark, C A Mauer, A.R Jurgensen and L Urwongse m SclenUfic Basis for Nuclear Waste Management V, Vol. 11, ed., W. Lutze (Elsevier-North-Holland, New York, 1983) p 1. [18] R.M Wallace and G.G Wicks, in: Scientific Basis for Nuclear Waste Management VI, Vol 15, ed, D G. Brookans (Elsevier-North-Holland, New York, 1983) p 23 [19] B C. Sales, M. Petek and L A Boatner, in: Scientific Basis for Nuclear Waste Management VI, Vol. 15, ed, D.G. Broolons (Elsevier-North-Holland, New York, 1983) p 251 [20] W.L. Kuhn and R.D. Peters in: SclenUfic Basis for Nuclear Waste Management VI, Vol 15, ed., D.G Broolons (Elsevier-North-Holland, New York, 1983) p 167