Standardized test methods that are currently in use or under development appear to adequately cover most of the testing that may be required to demonstrate compliance with the WAPS. It may be important to complete standardization of those tests that are under development, and to develop additional standardized tests for those specifications for which no standardized tests exist. A significant amount of work would be involved in this effort. Therefore, before this effort is undertaken, DOE must decide whether there is a need for a set of standardized test methods that would receive an “official approval” for use in waste compliance testing. If a set of approved tests is needed, DOE would need to determine what types of tests to include in the “approved list”, who should develop these tests, and which organizations should review and approve the test methods. Test method review and approval might be accomplished through the use of the Materials Review Board (MRB), an organization that was created by DOE for the purpose of reviewing and approving key test methods and data related to the repository licensing process. Alternatively, other means of standardizing these tests might be considered, such as processing the tests through the American Society for Testing and Materials for publication as ASTM standards.

If a set of approved tests is adopted, producers would not be required to use these tests; however, if a producer wished to propose other tests for compliance testing, it would be reasonable to require that such test procedures undergo a review/approval process similar to that which the officially approved tests had undergone. Once approved, these alternate procedures could be used for compliance testing.

Nuclear waste glass leaching has been studied extensively during the past ten years. Although much has been learned concerning the kinetics and mechanisms of glass dissolution, it does not appear that accurate predictions can yet be made concerning the release kinetics for specific elements from a given glass as a function of environmental conditions. In order to reliably predict elemental release rates, one needs to know: 1) how a given element is incorporated into the glass structure, 2) how specific sites in the glass react with water, 3) how the composition and reactivity of the leachate influence glass reactivity, 4) how the structure and reactivity of the glass changes in surface alteration layers, and 5) how glass dissolution modifies the chemistry of the leachate. At our current level of understanding, we are only able to make qualitative predictions concerning each of the above factors which allow us to make “order of magnitude” or “upper limit” predictions for radionuclide release rates.

A Large-Scale Leach Test Facility (LSLTF) has been constructed at the U.S. Department of Energy's Savannah River Laboratory (SRL) to perform static leach tests on 24-inch (610 mm)-diameter canister sections cut from simulated (nonradioactive) waste forms cast under reference conditions. The equipment and test procedures are designed to closely correspond to MCC-l leach test criteria.

Less than a factor of 3 increase in leachability results from combined scale-up, glass-cracking, leached surface area estimation, and surface roughness effects. This factor is dominated by surface roughness of saw cut surfaces. The factor is negligible when compared to the 200,000/1 ratio of glass sample masses.

The MCC-l, and other small-scale leach tests have been valuable in determining the relative merits of alternative waste form compositions. However, the actual waste glass to be stored in repositories is subject to fracture, devitrification, and container/glass interactions, which are difficult to simulate on a laboratory scale. Large-scale leach tests integrate these and other possible waste form characteristics that are not represented in typical small samples.

The facility, equipment, test method and results of one year leach test- ing are discussed. These results substantiate the applicability of small scale test data which precede this work, and the use of small scale tests to simulate leaching of the Defense Waste Processing Facility's borosilicate glass product. Exceptionally good sampling statistics make the large scale data particularly well suited for verification of models of leachability rates.

A systematic matrix of leach tests on a Rest Valley waste glass composition is described. These tests are intended to provide quantitative systematic information on leachant composition effects. and accordingly elucidate the controlling orocesses in hiuh-solute media. e.g. silicate ground water. brine. etc. In order to separate the effects of the nature and concentration of solutes from pH effects. leach tests were first carried out on the glass in de-ionized water under highly interactive conditions (high surface to volume ratio) in a configuration which pre- vented lowering the pH due to the intrusion of CO:. After determination of the resulting pH. the leach behavior of the glass was characterized in a series of high-dilution tests in various leachants buffered at this pH value in order to quantify the effects of leachant composition and concen- trations on the leaching orocess. The high-dilution conditions were used at this stage to ensure that the leachant composition mould not be signifi- cantly affected by accumulation of leached glass components. It was concluded that increased ionic strength appears to depress the leach rates in near-neutral media but enhances the leach rates under moderately basic conditions.

Repository-oriented leaching experiments involving Savannah River Laboratory (SRL) 165 type glass under a γ-radiation field (1= 0.2 × 104 R/h) have been performed by the Nevada Nuclear Waste Storage Investigations (NNWSI) project. In this communication, we discuss glass surface analyses obtained by SEM, nuclear resonance profiling, and SIMS together with leachate solution data in relation to a mechanism that couples diffusion, hydrolysis (etching and gelation), and precipitation to qualitatively describe the release of different glass components to the leachant solutions. The release of mobile (e.g., Li) and partly mobile (e.g., B) species is controlled primarily by interdiffusion with water species across the interdiffusion zone. Glass components that are immobile in the interdiffusion zone are released to the solution by etching. For prediction of long-term steady-state concentrations of glass components with low solubility, the relative rates of release from the glass and secondary mineral precipitation must be taken into account.

The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 & 165, and PNL 76-68), which were leached at 90°C (all glasses) or hydrated in a vapor-saturated atmosphere at 202°C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. In addition, molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 micrometer layer on SRL-131 glass formed by leaching at 90°C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl.

The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali-hydronium ion (H30+) interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H+ interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups according to:

20H = H20 molecular + 00

where 00 refers to a bridging oxygen, and OH refers to a hydroxyl group attached to a silicate polymer. The hydrated layer on the nuclear waste glasses appears to be of relatively low water content (4 to 7% by weight) and is not substantially hydroxylated. Thus, these layers do not have many of the properties associated with “gel” layers.

A Monte-Carlo code has been developed to infer dissolution rates of silicate glasses from their chemical composition. The glasses are considered as solid solutions of oxides and silicates, being divided into elementary cells of network formers, in which network modifiers can be incorporated. The different types of cells are randomly distributed and a resistance capacity to dissolution is assigned to each cell. We have tested the case where each cell is removed in a constant time as soon as it is in contact with the solution and a second one where each cell is dissolved in a time depending on the number of already dissolved neighbor cells. The first hypothesis leads to a percolation threshold when the proportion of fast etching cells reaches 25 %. Above, the dissolution front progresses via contiguous channels of fast etching cells. In the second hypothesis the variations of the etch rate with the proportions of each component are much closer to a linear combination of individual etch rates. These predictions have been successfully compared to experimental data on different series of silicate glasses of increasing complexity. Such a treatment could be straightforwardly extended to nuclear glasses.

Structural changes occurring at the surface of leached borosilicate glasses containing 3%wt of uranium have been investigated using fluores- cence EXAFS spectroscopy at grazing incidence. X-ray penetration depths between 30 Å and above 1000 Å are possible by varying the incidence angle of the X-ray monochromatic radiation in the vicinity of the critical angle for total external reflection, øc. This enables the surface and bulk structure to be compared. It is found that a local increase in uranium concentration takes place near the surface of the glass during the first 30 minutes of leaching time. Uranium is coordinated with oxygen and other uranium atoms. The local distribution can be modelled by a square planar island structure comprised of uranyl-type units. This structure changes with corrosion time. After 30 minutes of leaching there is evidence of partial removal of the surface layer.

The U. S. Department of Energy is currently evaluating the feasibility of using copper-base materials for the manufacture of nuclear waste con- tainers. One site under consideration for geologic disposal of nuclear waste is at Yucca Mountain, Nevada. One feature of this waste repository will be the initial presence of ionizing gamma radiation at high dose rates, which may alter the corrosive medium. To evaluate such effects, three copper-base materials (pure copper, 7% aluminum-copper and 30% nickel-copper) have been exposed (presgntly up to 14 months) to a gamma radiation field of approxi- mately 1 × 104 roentgens/hr. The exposure environments have been: 1) both groundwater (regional to the repository site, although taken from a lower elevation) at 95°C; 2) the water-vapor saturated air phase above it; and 3) air/water vapor at 150°C. In addition to uniform corrosion, both pitting and crevice corrosion have been observed. Characterization of the corrosion layers by X-ray diffraction has shown the presence of mixed copper(I) and copper(II) oxides. Studies by Auger Electron Spectroscopy (AES) have also been conducted in order to further characterize the compositions and structures of these corrosion products.

The anodic dissolution of copper has been examined in deaerated, 0.1 M HCl aqueous solution in the presence of H202. Concentrations of H202 up to 0.2 M were studied at a rotating copper disk-platinum ring electrode. The open circuit potential (OCP) of copper was found to depend on both peroxide concentration and rotation rate. The OCP shifts towards more positive values with increasing H202 concentration (C) and decreasing rotation rate (ω). The dependence of OCP on (C/ω1/2) was the same s for oxygenated solutions reported earlier [1], at small values of (C/ω1/2). At higher values of (C/ω1/2), departure from the expected behavior was observed. The current- voltage curves for anodic dissolution of copper were also influenced by the presence of peroxide. The curves recorded with the potential scanned in the positive direction showed the expected 60 mV slope, but the reverse scans showed significant departures. At a given potential scan rate, hysteresis was observed which was larger for higher H202concentrations, lower rotation rates, and more positive anodic potential limits. Monitoring the cuprous ions at the outer Pt ring revealed that there was a complex set of events taking place at the copper surface, including film formation and the appear- ance of cupric ions.

The autoradiolytic oxidation of Pu(IV) and Am(III) has been investigated in saline solution of different NaCl concentrations, pH and specific α-activities. The high linear energy transfer (LET) property of α-radiation in solution enhances the oxidation of the Cl ion and thus increases the redox potential (Eh) of the NaCI solution system to a value of more than +1000 mV vs. NHE. Radiolytic reaction products are found as being comparable with those from the Cl2 saturated NaCl solution 'and create a metastable system consisting of C12/HCl0/Cl0−/Cl−. The oxidation of Pu(IV) and Am(III) in such solutions is dependent on the NaCl concentration, pH and specific a-activity involved. The quantitative oxidation to Pu(VI) and Am(V) is ob- served in the concentrated NaCl solution at pH > 7 with a moderately high specific activity (> 1 Ci/L).

The Nevada Nuclear Waste Storage Investigations project has completed a series of experiments that provide insight into groundwater chemistry and glass waste form performance in the presence of a gamma radiation field at 90°C. Results from experiments done at 1 × 103 and 0 R/hr are presented and compared to similar experiments done at 2 × 105 and 1 × 104R/hr. The major effect of radiation is to lower the groundwater pH to a value near 6.4. The addition of glass to the system results in slightly more basic final pH, both in the presence and absence of radiation. However, there is essentially no difference in the extent of glass reaction, as measured by elemental release, as a function of dose rate or total dose, for reaction periods up to 278 days.

Preliminary short-term experiments to study the effect of gamma radiation on the waste package materials interactions in the environment of a nuclear waste repository in basalt were completed. Experimental parameters were similar to those expected during the saturated post- closure period of the repository, i.e., a temperature of 200°C and a hydrostatic pressure of ≤70 bar. The test components included various combinations of synthetic groundwater (GR4), low carbon steel coupons, CH4, basalt, and basalt/bentonite mixtures (packing). The tests were run in low carbon steel pressure vessels.

It was found that gamma radiation at a dose rate of 104rad/hr for a duration of one and two months increased the H2 yield in all tests. Increases in the organic carbon yield, the sulfate ion concentration, and the corrosion rate of the coupon were observed in some of the tests. These latter results varied with the combination of waste package components included in the tests. Evidence for the quench effect was observed in tests which included basalt.

Minerals of the thorite (ThSiO4) mineral group have been leached as fine powders in bicarbonate-carbonate solutions for time periods up to 288 days.The radioactivities of the solutions were measured for 238U, 234U, 232Th, 230Th and 228Th by alpha spectrometry, preceded by chemical separation and purification of U and Th. The results confirm the preservation of disordered regions of higher chemical reactivity in an already fully-damaged aperiodic structure. The damage is most evident for the short-lived nuclide 228Th and to a much lesser extent for the long-lived nuclide 234U because the alpha-recoil tracks anneal with time. The calculated mean life of an alpha-recoil track in these metamict thorium silicates is approximately 1,500 to 6,000 years. This study demonstrates that alpha-decay products are selectively leached from damaged, aperiodic materials, but the damage is subject to low-temperature annealing over relatively short periods of time.

X-ray absorption spectroscopy has been used to investigate the Ta LIII and LI absorption edges of metamict and annealed pyrochlore minerals A1-2B2O6Y0-1. The LIII edge absorption spectra exhibit a pronounced doublet in primary absorption resonance for the annealed samples which have the pyrochlore structure. The doublet in the LIII edge resonance becomes less pronounced and for some samples even merges Into a singlet for the samples in the unannealed, fully damaged (metamict) state. The LIII edge resonance of the damaged samples closely resembles Ta205 or disordered manganotantalite (MnTaNbO6 ), while the LIII edge resonance for the annealed samples most closely resembles the synthetic pyrochlore KTaWO6·H20. These results are supported by the Ta LIIIedge x-ray absorption data. Fourier transforms of the Ta LIII edge x-ray absorption fine structure (EXAFS) show first neighbor Ta-0 bond lengths of 1.96 ± 0.05 A for the samples in the annealed and metamict state.

The complex structure of Ta205 is used as a model for the Ta-containing oxygen polyhedra in metamict pyrochlores, as it embodies some of the topologic features consistent with the reduction of long-range periodicity with increasing alpha-decay dose: 1) greater distortion of the primary coordination shell around Ta, 2) slightly increased mean second nearest-neighbor <Ta…X> distances, and 3) a wider distribution of second nearest-neighbor distances. These modifications are incorporated in a large unit cell with a b translation of 40 A.

Annealing of radiation damage in artificial sodium silica glass has been studied by absorption spectroscopy for different radiation sources (fission fragments produced by n,f reactions, alpha decays produced by n, α reactions, and 2 MeV electrons) as a function of annealing time and temperature. The annealing observed can be described best by a single exponential function

A(t) = a exp (− αt)

with an Arrhenius type temperature dependence of the annealing coefficient α

α(T) = α0 exp (− EA/kT)

For the three different radiation sources similar activitation energies EA were deduced from the data. Possible annealing processes are discussed.

A groundwater radiolysis model has been developed at Battelle-Columbus which is applicable to groundwaters containing bicarbonate species. The model consists of a chemical mechanism which describes interactions between groundwater species and radiolytic species. Due to the chemical kinetics nature of the model, elementary reactions can be added to extend its range of applicability to other groundwaters. This paper describes the chemical kinetics and influences of bicarbonate species in the model.

Near-field modelling is concerned with the description of the migration, chemical and degradation processes that may occur within an engineered radioactive waste repository and its immediate environs. The object is to gain understanding of such processes in order to predict the long-time evolution of the repository and to assess the degree of containment provided by the proposed engineered construction. The conditions of primary interest to our programme concern the waste contained within a steel canister and buried in a concrete environment within a clay geology. The chemistry of the near-field is controlled in that it is the consequence of the choice of near-field components, but it may be extremely complex. Intrusion of external groundwater and degradation of the chosen materials will lead to variations in the chemistry in both space and time. It is vitally important to understand these changing chemical conditions since they determine the solubility and sorption of any released radionuclides. In this paper, we describe the computer program CHEQMATE (CHemical EQuilibrium with Migration And Transport Equations), which has many applications in modelling various changes in chemistry in the near-field. The program combines an ionic migration code with the geochemical program PHREEQE [1]. The program maintains local chemical equilibrium in the system as the transport processes evolve. The program includes automatic mineral accounting; solid phases are added or removed from the equilibrium as precipitation or dissolution occurs. We illustrate the use of the CHEQMATE program with an example of a coupled chemical and transport problem, particularly relevant to the near-field of a waste repository.

Experiments have been completed in the basalt-Tc-doped groundwater sys- tem to quantify the mobility of Tc in the waste package and near field envion- ment in a repository located in basalt. The experiments were completed using a batch sorption technique in which 8 rushed basalt and Tc-doped groundwater were reacted at 85, 100, 125 and 150 °C for periods of time up to 118 days. The water to rock ratio was 10:1 ml/g. In another set of experiments, basalt crushed under anoxic conditions and Tc-doped groundwater were reacted at 85°C at different water to rock ratios (10:3, 20:3, 40:3 and 80:3 ml/g).

The rate and percentage of Tc removed from solution increased with increasing temperature and decreasing water to rock ratio. Also, basalt crushed under anoxic conditions removed >87% of Tc from solution at 85°C in 7 days while no Tc was removed from solution with basalt crushed in air at the same temperature after 118 days. These results are consistent with the hypothesis that (1) ferrous iron in basalt phases preferentially consumes oxygen and will not reduce Tcuntil all free oxygen in the system is consumed. Once oxygen is consumed, reduction of Tc is quite rapid. These data indicate that the potential exists in a basalt hydrothermal environment to satisfactorily immobilize Tc. Reliable application of these data will depend on experimental and theoretical estimates of oxygen consumption with time in the waste package environment following repository closure, subsequent saturation and long term hydrothermal reactions.

An area containing unsaturated fractured tuffs at Yucca Mountain, Nevada, is one of the potential sites for geologic storage of high-level radioactive waste. The Environmental Protection Agency (EPA) 40 CFR 191 Regulation limits the cumulative releases of many radionuclides from the repository to the accessible environment for 10,000 years after disposal [I]. Numerical models can be used to determine if the EPA containment requirement is met.

In this paper a preliminary set of transport calculations for uranium and technetium is discussed. First, a foundation for the calculations, a comprehensive, referenced geochemical/geophysical model containing the current stratigraphic, petrologic, hydrogeologic, geochemical, and material data for the Yucca Mountain site was compiled. Second, the integrated transport of uranium and technetium from the repository to the water table was modeled. An expected-case flow scenario and a worse-case flow scenario were used in the calculations. The sorption of technetium was neglected in the worse-case scenario. Results show that in the case of uranium, the estimated transport was only moderately sensitive to the magnitude of the flow because sorption had a significant effect on the retardation. In the case of technetium, when sorption was neglected, the flow dominated the transport. The EPA contain- ment requirements were met for all cases. These preliminary calculations will be used as a basis to investigate the effects of physical and geochemi- cal processes on the long-term transport of radionuclides at Yucca Mountain.