CORROSION OF SIMULATED NUCLEAR WASTE GLASS

Journal o f Radioanaly tical and Nuclear Chemistry, Articles, Vol. 122, No. 1 (1988) 91-102

CORROSION OF SIMULATED NUCLEAR WASTE GLASS

S. MUSIC, M. RISTI(~, M. GOTI(~, J. FORI~

Rud]er Bo~kovi~ Institute, P.O.Box 1016, 41001 Zagreb [YugoslaviaJ

(Received November 13, 1987)

Glasses are suitable host matrices for the immobilization of high-level radioactive wastes. The corrosion behaviour of nuclear waste glass in water is of considerable importance, since a potential route for returning of radionuclides to the biosphere is their leaching from the waste form into groundwater and subsequent transport by the groundwater to the surface. In this study, the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilieate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates (g'cm-2"day-~), as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na20, 10.39% B~O3, 45.31% SiO2, 13.42% ZnO, 6.61% TiO2 and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed.

Introduction

Borosilicate glass is one of the leading candidate materials for the immobil izat ion

of high-level radioactive wastes) -4 The concept of using borosilicate glass as a host

matr ix for radioactive wastes is based on the formation of solid solutions, which are

characterized by good inertness to chemicals and high thermal stability. Borosilicate

glasses containing actual radioactive wastes are not very sensitive to a/ / t /7 radiation.

Finally, French findings have proven that commercial vitrification o f high-level radio-

active wastes is possible with the use of alkali-borosilicate glass.

The current concept of a final repository for high-level radioactive wastes involves

a multi-barrier approach. In accordance with this concept , the nuclear waste glass

will be cast into a metal container and buried in a stable geologic formation. Clay

minerals (for example, bentonite) will be used as backfill material in order to provide

an additional barrier for radionuclides.

Waste radionuclides may enter into the biosphere, if groundwater penetrates a

metal container and comes in contact with the surface of nuclear waste glass. For

this reason, safety evaluation for the isolation of high-level radioactive wastes in a

Elsevier Sequoia S. A., Lausanne Akaddmiai Kiad6, Budapest

S. MUSIC et al.: CORROSION OF SIMULATED NUCLEAR WASTE

final repository must take into account the corrosion behawour of a glass waste form during a long period of time. Long-term corrosion (up to 10 4 years) cannot be measured directly, but it can be estimated on the basis of short-term laboratory experiments. Generally, in these experiments the corrosion of nuclear waste glasses can be accelerated by chemically aggressive solutions, high temperature and pressure, intensive alpha radiation, etc.

In a narrow sense, the term "corrosion of nuclear waste glass" is often used to describe ability of this glass to resist corrosion in aqueous medium (leaching pro-

perties). The experimental measurement of leach rates has been reviewed by MENDEL: In this work, the preparation and characterization of borosilicate glasses of dif-

ferent composition were investigated. The vitrification of simulated nuclear waste oxides with borosilicate matrix was also investigated. The chemical corrosion in water of these glasses was measured, because the reactions between water and surface of the nuclear waste glass are responsible for possible release of radionuclides into the biosphere. The experimental results are illustrated by means of selected examples.

Experimental

All chemicals were of A. R. grade. Chemicals in the form of oxides and carbonates were used for the preparation of glass samples. Simulated nuclear waste glasses were prepared as follows. The mixture of hydrous oxides ("waste oxides") and the glass- forming oxides was melted in air atmosphere. The melt was cast in a preheated graphite mould, and then the glass was cooled at room temperature. Chemical compositions (wt. %) of the glass samples, prepared for the investigation in this work, are given in Tables 1 and 2.

Table 1 Composition of samples G-1 to G-6 in weight %

Waste Sample Na 2 0 B~ O~ SiO 2 Li 2 0 CaO ZnO TiO~ oxides

(wo-1)

G-1 27.95 8.84 46.45 3.52 4.43 8.80 99.99 G-2 17.10 11.38 49.60 14.49 7.24 I00.01 G-3 14.93 14.92 55.26 7.31 2.99 4.58 99.99 G-4 18.53 12.24 46.95 14.69 7.78 100.19 G-5 18.49 I0.01 52.55 3.98 5.01 9.95 99.99 G-6 15.61 10,39 45.31 13.42 6,61 8.66 I00

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S. MUSI(~ et al.: CORROSION OF SIMULATED NUCLEAR WASTE

Table 2 Composition of samples G-7 toG-18 in weight %

Sample Na 20 B2 03 SiO2 Li20 CaO SrO ZnO TiO2 Waste oxides

WO-1 WO-2 WO-3

G-7 G-8 G-9 G-10 G-11 G-12 G-13 G-14 G-15 G-16 G-17 G-18

27.0 18.5 17.0 27.0 27.0 27.0 18.5 18.5 18.5 17.0 17.0 17.0

9.0 11.5 12.0 9.0 9.0 9.0

11.5 11.5 11.5 12.0 12.0 12.0

47.0 3.5 2.5 2.0 9.0 52-0 3.5 2.5 2.0 10.0 49.5 2.0 12.5 7.0 47.0 3.5 2.5 2.0 9.0 47.0 3.5 2.5 2.0 9.0 47.0 3.5 2.5 2.0 9.0 52.0 3.5 2.5 2.0 10.0 52.0 3.5 2.5 2.0 10.0 52.0 3.5 2.5 2.0 10.0 49.5 2.0 12.5 7.0 49.5 2.0 12.5 7.0 49.5 2.0 12.5 7.0

+9.0

+9.0

+9.0

+9.0

+9.0

+9.0

+9.0

+9.0

+9.0

Table 3 Composition of simulated waste oxides in weight %

Waste oxides Component

WO-1 WO-2 WO-3

Fe203 61.75 Cr2 03 2.06 NiO 0.24 MnO 0.47 CoO 0.41 La 2 03 11.80 Ce203 11.95 Prz 03 1.98 Sm~ 03 2.06 Gd203 7.28

61.49 95.09 2-05 3.17 0.24 0.38 0.47 0.73 0.41 0.63

35.33

Hydrous oxides ("waste oxides") were coprecipitated from nitrate salts solution

at pH ~ 9.5. The coprecipitation was performed using NaOH solution. The copreci.

pitate was washed with doubly distilled water containing a small amount of NH4 OH.

The cleaned precipitate was dried at 105 ~ Chemical compositions (wt. %) of

simulated waste oxides (WO-1, WO-2 and WO-3) are given in Table 3.

The leach rates (g �9 cm -2 �9 day-1 ) were determined using the standard leach test

proposed by IAEA (see Figure 4 in Reference 5). In our experiments this method was

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slightly modified, in such a manner tb_at the glass surface was completely exposed to water attack. In all leach tests, the ratio of solution volume to sample surface area was restricted at 10 cm. Doubly distilled water (not free from atmospheric CO2) with an initial pH ~ 5.9 was used. Glass samples were thermostated at 25---0.1 ~

Analytical chemistry of solutions was performed using inductively coupled plasma atomic emission spectroscopy (IPC-AES). Determination of trace concentrations in leachates was limited with the detection capabilities of this spectroscopic technique.

R e s u l t s a n d d i s c u s s i o n

Generally, borosilicate glass can be defined as a supercooled liquid (solid solution) with a different B2 O3/SIO2 molar ratio. The modifying oxides added to the borosilicate matrix affect the glass properties, such as the melting temperature, chemical durability, colour, thermal expansion, etc. Borosilicate glasses with high SiO2 content are chemically very resistant. On the other hand, they are characterized by high melting temperatures. Borosilicate glasses with lower melting temperatures are more suitable for the immobilization of radioactive wastes, because with these glasses during the vitrification process there are no major problems with volatile radionuclides (for example with radionuclides of Cs and Ru).

Nuclear waste borosilicate glass may contain up to 35 wt. % of waste oxides. 6 In Sweden, for instance, the upper limit of nuclear waste loading is defined with 9 wt. % of radioactive fission products. 4 In addition to the glass-forming oxides, the nuclear waste glasses contain more than 30 different oxides, which include the nuclear waste itself plus various additives whose purpose is to modify the chemical and physical properties of the solidification matrix.

The corrosion behaviour in aqueous medium of nuclear waste glass depends on many factors. 7 An incomplete list includes: - glass composition, - chemical composition of aqueous medium (pH, buffer capacity, silicic acid concentration, etc.), - water supply (surface area/volume ratio, different flow conditions), - temperature, - effects of backfill material and container corrosion products, - and radiation effects.

In the present work, borosilicate matrices of different composition (samples G-1 to G-5 and G-7 to G-9) were prepared. X-ray diffraction analysis has shown that these samples are amorphous materials for X-rays. Samp,es G-7 to G-18 contained 2 wt. % of SrO, which is added as the glass network modifier and also to simulate the radioactive strontium. The chemical compositions of glass matrices were determined taking into consideration the general properties of the borosilicate glass. For instance, large additions of Na § ions reduce the corrosion resistance of borosilicate glasses.

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S. MUSIC et aL: CORROSION OF SIMULATED NUCLEAR WASTE

However, if the glass contains different alkali ions (for example, Li § and Na § the corrosion resistance of alkali borosilicate glass improves (synergetic effect). The incorporation of ZnO and TiO2 improves significantly the chemical durability of the borosilicate glass. Also, the addition of alkaline earths, as the glass network modifiers, improves selective alkali leaching. Nomlally, CaO is added, but MgO, SrO and BaO act similarly.

The leach tests were performed with doubly distilled water, which is a chemically more aggressive medium than brine. Experiments were performed at 25-+O.1 ~ and this temperature is lower than the temperatures (40-90 ~ usually used by other researchers. This difference in the experimental temperatures is not critical, since the surface temperature of nuclear waste glass will be 40 ~ or less, if and when it finally comes in contact with water. Only for high loadings (~ 35 wt. %) the surface temperature of nuclear waste glass can be expected in critical situations to be

90 ~ or less. Figures 1 and 2 show typical examples for the time dependence of leach rates of

the glass forming elements. The slope of lines changes as a function of time. At the

Time, days

_50 20 /-,0 60 80 100 120 1/.,0 I I I I I I I

~

-~-8

Fig. 1. Leach rates of Na, Si, Ti and Zn, measured for sample G-2, as a function of time

beginning of the corrosion process (within two weeks), the leach rates decrease �9 abruptly. After two weeks, the corrosion process is significantly retarded. The experimental results obtained for samples G-l, G-2, G-3, G-4 and G-5, indicate the following order of leaching from the borosilicate matrix: Li > Na > Si > Ti > Zn.

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S. MUSI(~ et a].: CORROSION OF SIMULATED NUCLEAR WASTE

Time, doys 0 20 /-,0 60 80 I00 120 T40

Ti

Fig. 2. Leach rates of Li, Na, Si and Ti, measured for sample G-3, as a function of time

_30

-4

Time,doys 20 L.O 60 80 100 120

I I I ]

-5

Na

SJ

0

Fig. 3. Leach rates of Na, Si, B, Ti and Zn, measured for sample (-:-6, as a function of time

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Time, doys 20 40 60 80 100 120 140 160 180

-3

-5

-6

4"

"0 -7 E

t~ 0 - 8

Co

Mn

Fe

l.'ig. 4. Leach rates of Ni, Cr, ('o, Mn and t,'e, measured for sample G-6, as a function of time

Time, days

_2 ( 20 40 60 80 100 120 [ t I I I I

-3

-4

-5

Fig. 5. Leach rates of Ce, Pr and Sin, measured for sample (;-6, as a function of time

Leach rates o f Ca measured a f te r six m o n t h s o f cor ros ion in doub ly dist i l led wate r

are given in Table 4.

Sample G-6 is doped with 8 .66 wt. % of s imula t ed waste oxides (WO-1). The

mix tu re o f s imula ted waste oxides (WO-I ) is p r epa red to s imula te the co r ros ion

p roduc t s , the n e u t r o n ac t iva t ion p roduc t s and the fission p roduc t s .

Table 4 l.each rate of Ca measured after six months of

corrosion in doubly distilled water at 25 ~

Sample R, g �9 cm -2 �9 day- '

G-1 1.70- 10 -7 G-3 3 .90. 10 -7 G-5 3 .43. 10 -7

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S. MUSI~ et al.: CORROSION OF SIMULATED NUCLEAR WASTE

The leach rates measured for sample G-6 are shown in Figures 3, 4 and 3. Figure 3

indicates that the leach rates of the glass forming elements, measured after 112 days, are in the range of values R ~ 1 �9 10 -6 to 3 �9 1 0 - S g . cm -2 �9 day -1. Figure 4

shows the leach rates of Fe, Cr, Ni, Mn and Co, measured for sample G-6, as a func-

tion of time. The following order of leaching of these elements is observed (after

six months): Ni > Cr > Co > Mn > Fe. Leaching of Ni, Cr and Co is pronounced at the beginning of the corrosion process. The leach rate of Ni changes from R = 5.8 �9 10 -3 to 2.8 �9 10 -6 g �9 cm -2 �9 day "x , after 6 months. On the other

hand, the leaching of Fe from borosilicate matrix is very small. The leach rate of Fe changes from R = 1.3 �9 10 -s to 1.3 �9 10 -8 g �9 cm -2 �9 day -1 for the same time

interval (1 day to 6 months).

Figure 5 shows the leaching of Ce, Pr and Sm from sample G-6, as a function of time. The leach rates of Ce, Pr and Sm decrease rapidiy with the increase o f time.

For instance, at the beginning of the corrosion process (after 1 day) the leach rates of these elements are in the range of values R = 4.7 �9 10 -3 (for Sm) to 1.1 �9 10 -3

g �9 cm-2 . day-1 (for Pr). After 112 days, the leach rates are in the range of values R = 1.0 �9 10 -8 (for Sin) to 6.0 �9 10 -7 g �9 cm -2 �9 day -1 (for Ce).

Cerium is often used by researchers to simulate the actinides. Recently, APTED,

MCVAY and WALD a investigated the leaching of Np, U and Pu from defense waste

glass under simulated repository conditions. The authors found that leaching behaviour of uranium does not show any consistent relationship to either neptunium or plutonium

release behaviour. Also, they concluded that the use of uranium as a "master" indicator

of actinide leaching is questionable. In spite of this finding, the simulation of high-

level radioactive wastes is very important at the laboratory level to avoid difficulties

in work with "ho t" materials, as well as to eliminate the radiation protection.

The leach rates measured for sample G-9, as a function of time, are shown in

Fig. 6. The following order of leaching is obtained: Na > B > Si > Ti > Zn.

The results of our study showed that leaching behaviour of Ca and Sr is similar.

Generally, the leach rate of Ca is greater than that o f Sr. At the beginning of corrosion

process the leaching of Ca and Sr is significant. However, with the increase of time,

as well as the pH of leachant, the leaching of Ca and Sr decreases abruptly. At the

first stage, leaching of Ca and Sr is controlled by diffusion/adsorption processes on

the glass surface. This process is followed by the surface precipitation of CaCO3 and SrCO3 or the structural incorporation of Ca and Sr in the passive layer.

Figure 7 shows the leach rate of La, measured for sample G-17, as a function of time. Leach rate of La changes from R = 7.0 �9 10 -s to 2.4 �9 10 -6 g �9 cm , 2 �9 day "1

for a given time interval (I to 90 days). La 3+ cations precipitate as hydroxide at

higher pH values in relation to the other lanthanide cations. Generally, lanthanide(III) cations undergo intensive hydrolysis at pH ~ 6 to 8 (the direction Lu 3§ - La3+). Ex-

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S. MUSIC et al.: CORROSION OF SIMULATED NUCLEAR WASTE

-5 -~ 3~

-6

A

Cn

0 _

Time, days 20 40 60 80 100 120 140 160~,.

I 1 I I I I I !

o Na

aB ~Si

Ti

1 Zn

Fig. 6. Leach rates of Na, B, Si, Ti and Zn, measured for sample G-9, as a function of time

T i m e , d a y s

. 0 20 4 0 60 80 100 120 -41 I I I i I I

"~E

%~ La

Fig. 7, Leach rate of La, measured for sample G-17, as a function of time

perimental conditions for the precipitation of La(OH)3 are investigated by MUSIC

and ~ IPALO-~ULJEVI(~. 9 La 3+ cations are sensitive to the presence of carbonate ions

and this leads to the formation of lanthanum carbonate. The reactions of hydrolysis of lanthanide(III)cations and their affinity for carbonate ions play an important role during the formation of a passive surface layer on the nuclear waste glass.

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S. MUSI(~ et al.: CORROSION OF SIMULATED NucLEAR WASTE

In accordance with CLARK and coworkers, 10,1 t the mechanism of glass corrosion in aqueous medium involves the following processes: ion exchange of alkali and alkaline earth cations present in the glass with H § or Ha O § ions from the corrosion solution, network dissolution, in which the structural bonds in the glass are attacked by OH- ions, pitting of the glass surface due to the preferential leaching of highly stressed regions, partially phase-separated or crystallized glasses, or the presence of impurities and heterogeneities, - precipitation of ionic species from solution onto the glass surface, - development of surface protective (passive) layers, and clouding or

staining of the glass surface through weathering. The corrosion behaviour of simulated nuclear waste glass (or actual nuclear waste

glass) is very dependent on the chemical and structural properties of the passive layers formed on the glass surface. The chemical and structural properties of passive layers have been investigated by many researchers.

Tile presence of Zn 2§ in the glass or in the solution induces the formation of a thin protective layer on a glass under aqueous attack. 12 This layer is mostly amorphous. The crystalline structures discovered were ascribed to Zn4(OH)2 Si20T (hemimorphite) and Zn2 SiO4 (Zn-orthosilicate).

Van ISEGHEM and coworkers 13 investigated the corrosion behaviour of five simulated European reference high-level nuclear waste glasses (borosilicate matrices). The increase of leachant temperature to 150 ~ and higher leads to pronounced crystallization in the surface layer. Cubic analcime (NaAISi2 06 �9 H2 O) and a tetragonal zeolite (NasA13 Sis O16 " 6H2 O) were observed at 150 ~ Three crystal types were detected at 190 ~ an alcime, orthorhombic eucriptite (LiA1SiO4 �9 2H2 O) and a needle-shaped Ca, Si based crystal (structure not determined). Protective surface layers were thick up to 60/am, or more.

CELESTANI and coworkers 14 investigated the effects of Cs2 O, SrO, U30s and ZnO on the durability (leaching behaviour) of a borosilicate glass. They found that Cs20 depresses the durability, whereas the strontium and uranium oxides leave practi- cally unchanged the leaching properties. The influence of ZnO is completely different; it increases the glass durability.

MURAKAMI and BANBA 1 s used the Soxhlet technique to investigate the leaching properties of borosilicate glass containing 14 wt. % of simulated high-level radioactive waste. Four surface layers were detected on the glass and their concentration profiles were determined. The presence in surface layers of the crystalline phase with a general formula of (Ca, Ba, La, Ce, Nd)x(Mn, Fe, Zr, Mg, Ni, A1)y(Si, Al)z(O, OH)m is suggested.

MEANS and SPINOSA 16 found that the reprecipitation of released cations (radionuclides) onto the glass surface affects the morphology and composition of protective layers.

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The adsorption of radionuclides on the glass surface, during the corrosion process,

plays an important role in decreasing the radioactivity of leachate. Hydrous oxides

(particularly of iron ions) formed on the glass surface are strong adsorbents for a wide

variety of radionuclides. The adsorption of trace elements (or radionuclides) on hydrous

iron oxides was discussed by MUSIC and RISTIC. 1 7

Conclusions

The experimental results of our study indicate that borosilicate glass is a promis-

ing material for the immobilization of high-level radioactive wastes. This study showed

that a simulated nuclear waste glass with the chemical composit ion (weight %), 15.61%

Na20, 10.39%, B2Oa, 45.31% SiO2, 13.42% ZnO, 6.61% TiO2 and 8.66% waste

oxides, is characterized by good corrosion resistance in water medium. Also, this glass

is characterized by low melting temperature. These findings are very important for

the technology of vitrification of high-level radioactive wastes. Incorporation of ZnO

into the glass structure improves significantly the chemical durability of borosilicate

glass. For this reason Zn-borosilicate glasses can be considered as the next generation

of materials for the immobilization of high-level radioactive wastes. 1 s - 2 0

We wish to thank Prof. Stanko POPOVI~, Prof. Marko HERAK and Dr. Goran VUJICI~ for their help during the experimental performance of this work.

References

1. B. C. SALES, L. A. BOATNER, H. NARAMOTO, C. W. WHITE, J. Non-Cryst. Solids, 53 (1982) 201.

2. L. L. HENCH, D. E. CLARK, J. CAMPBELL, Nucl. Chem. Waste Manage., 5 (1984) 149. 3. D. M. STRACHAN, K. M. KRUPKA, B. GRAMBOW, Nucl. Chem. Waste Manag., 5 (1984) 87. 4. H.-P. HERMANSSON, H. CHRISTENSEN, I.-K. BRONER, L. WERME, D. E. CLARK, Nucl.

Chem. Waste Manag., 5 (1985) 315. 5. J. E. MENDEL, Nucl. Chem. Waste M anag., 3 (1982) 117. 6. N. E. BIBLER, P. KENT SMITH, Characterization of borosilicate glass containing Savannah

River Plant radioactive waste, in: Advances in Materials Characterization, D. R. ROSSINGTON, R. A. CONDRATE, R. L. SNYDER (Eds.), Materials Sci. Res., VoL 15, Plenum Press, New York and London, p. 591.

7. R. GRAUER, Synthesis of Recent Investigations on Corrosion Bchaviour of Radioactive Waste Glasses, EIR-Bericht Nr. 538, March 1985, CH-5303 Wiirenlingen, Switzerland.

8. M. J. APTED, G. L. MCVAY, J. W. WALD, Nucl. TechnoL, 73 (1986) 165. 9. S. MUSIC, J. ~IPALO-~ULJEVIC, Colloid Polymer Sci., 256 (1978) 970.

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10. D. E. CLARK, L. URWONGSE, C. MAURER, Nucl. Teehnol., 56 (1982) 212. 11. D. E. CLARK, C. G. PANTANO, Jr., L. L. HENCH, Glass Corrosion, Books for Industry, New

York 1979; citation via Reference 10. 12. R. A. LEWIS, S. MYHRA, R. L. SEGALL, R. St. C. SMART, P. S. TURNER, J. Non-Cryst.

Solids, 53 (1982) 299. 13. P. Ph. Van ISEGHEM, W. TIMMERMANS, R. DE BATIST, Mat. Res. Soe. Syrup. Proc., 44

(1985) 55. 14. G. CALESTANI, A. MONTENERO, E. FERRAGUTI, G. INGLETTO, M. BETrINELLI, J.

Non-Cryst. Solids, 84 (1986) 452. 15. T. MURAKAMI, T. BANBA, Nuel. TechnoL, 67 (1984) 419. 16. J. L. MEANS, E. D. SPINOSA, Ceramic BulL, 65 (1986) 780. 17. S. MUSIC, M. RISTI~, L Radioanal. Nuel. Chem., 120 (1988) 289. 18. R. P. SCHUMAN, N. D. COX, G. W. GIBSON, P. V. KELSEY, Jr., Evaluation of forms for the

immobilization of high-level and transuramc wastes, Report EGC~FM-6045, EG&G Idaho Inc., August 1982, Idaho Falls, Id. 83415.

19. R. P. TURCOTTE, J. m. WALD, Devitrification behavior in a zinc borosilite nuclear waste glass, Report PNL-2247, Battelle Pacific Northwest Laboratories, March 1978, Riehland, Wa. 99352.

20. S. MUSIC, M. GOTI~, S. POPOVI~, B. GR~.ETA, J. RadioanaL Nucl. Chem., 116 (1987) 141.

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CORROSION OF SIMULATED NUCLEAR WASTE GLASS

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