Composites of cement and radioactive waste are being used for disposal of low level waste and may be used for high level waste. It is important to be able to predict their behavior in case of intrusion of leaching waters. The purpose of this paper is to determine the mechanisms of leaching of the cement and how the presence of radioactive waste components affects these mechanisms in composites.

A review of the leaching behavior of 76–68 glass shows that it cannot be explained in terms of diffusion in the glass, which has been the basis for several leach models. Instead, we present two models based on a dissolution rate impeded by surface processes: the accumulation of a protective layer of insoluble reaction products, and adsorption of reaction products on the surface. The resulting predicted time dependences are identical and predict a change from linear to parabolic rate laws for soluble species, which is found to agree with the data over a range of temperatures. Incongruent release is attributed primarily to solubility effects. The relative merits of the models are discussed on the basis of the effect of surface area-to-volume ratio in static leach tests. Their relevance to modeling repository behavior is discussed.

Cement waste forms prepared by three processes, casting, cold pressing, and FUETAP (Formed Under Elevated Temperatures and Pressure) have been compared for their leachability by using the MCC-1 leach test. The results indicate that releases of plutonium are not controlled by the waste form matrix and that there is no significant overall advantage to any of the three cement processes from a leachability viewpoint.

The effects on the leachability of SRL 131 glass as a result of slight changes in the glass composition were monitored by leachate and surface analyses. The compositional changes had marked effects on solution and layer composition and on the precipitated alteration products that formed after 546 days of static leaching. The leaching behavior of long-lived actinide elements were measured, with Pu and Am showing little tendency to become solubilized and Np and U continuously accumulating in solution.

Specimens of borosilicate glass were irradiated in a material testing reactor at a neutron fluence of 1019–1020 nvt and a temperature of below 250°c. Tritium released from the specimen was determined by means of a specially designed sampling system and liquid scintillation counter.

Temperature and time dependence of the releasing behavior of tritium from borosilicate glass was obtained in the temperature range of 200 – 700°C. Very small amount of tritium as compared with the amount contained in the specimen was released in several hours at temperatures below 700°C. The greater part of tritium released from the specimen was tritium oxide, which is controlled by the diffusion step of tritium in the specimen. The diffusion coefficient of tritium in borosilicate glass was also obtained, which agreed with that of water in glasses.

Shallow land burial of sodium-bearing low-level wastes requires an assessment of safetyconsequences relating to (1) groundwater interactions with the residual sodium in the event of a breach of containment and (2) release of tritium dissolved in sodium along with hydrogen (both of which will permeate through adjacent soil). This paper describes the tests and evaluations being performed to establish a data base on safety consequences of burying low-level radioactive sodium wastes (RSW). This data base can be used in establishing the amounts of sodium that may be tolerated in burying RSW and in establishing related criteria.

The effects of the surface area-to-solution volume ratio on waste glass leach rates are investigated from a theoretical point of view. Simple leach models are discussed first. Correlation variables to interpret the results of similar leaching experiments performed at different values of the surface area-to-solution volume ratio are obtained for static leach testing. For dynamic leaching conditions, the source term required for risk assessment is derived and its dependence on the leachant flow rate and leach specimen surface area is discussed. The findings are upheld by a more complex leach model, the mathematical formulation of which has been implemented in a computer code named LIX. When tested against actual PNL 76–68 glass leaching data, LIX shows excellent capabilities in reproducing the experimental evidence, in particular the effects of the surface area-to-solution volume ratio.

Water samples from the disposal trenches of two low-level radioactive waste burial sites were analyzed for their inorganic, organic, and radionuclide contents. Since oxidation of the trench waters can occur during their movement along the groundwater flow path,experiments were performed to measure the chemical and physical changes that occur in these waters upon oxidation. In addition to the highly acidic conditions of the oxidized Maxey Flats trench water, low concentrations of chelating agents, shown to exist in trench waters, may be responsible for keeping radionuclides in solution following oxidation of anoxic trench leachates from Maxey Flats and West Valley burial sites.

A test method based on partial leachant exchange at constant intervals until leachate concentrations become stabilized has been used to obtain a data-base in evaluating long-term leaching patterns. Applied to basalt and granite specimens and compared with ground-water analyses, correlations which can be used in long-term leachability extrapolations have been established. They are based on the following principles: (i) Control of solubilities of major components by combined aluminosilicate species; (ii) Continued increase in levels of soluble species, e.g. Na+ and Cl−, with time; (iii) Initial rapid leaching of SO42−, along with Ca2+ and K+, followed by slowing down, probably due to the presence of selectively leachable phases; (iv) Initial retardation of Mg and Fe leaching followed by regrowth as sorption sites become saturated.

This paper summarizes a series of experiments to determine the strontium leachability of cement-based sludge-slurry hydrofracture grouts.

The hydrofracture process has been used at ORNL since 1966 for the routine disposal of intermediate-level waste (ILW) solutions. In this process, cement and other additives are mixed with a waste stream to form grout, which is then injected into a shale bed at a pressure sufficient to cause fracturing along the horizontal bedding planes. This injected grout soon hardens, fixing the radionuclides between layers of the massive Conasauga shale formation.

Modified IAEA dynamic leach tests on hydrofracture grout specimens showed improved leach resistance (by a factor of 3 to 5) as the curing time increased from 28 d to 91 d. A weak trend toward lower leachability with increased amounts of dry solids was observed.

The moving boundary-plus-diffusion leach model fits the dynamic leach data successfully in most cases. An apparent diffusion coefficient of 5 × 10−12 cm2/s and a moving boundary coefficent of 1 × 10−7 s−1 were obtained when one of the best grout compositions was leached in distilled water.

During the past year we have conducted a modified MCC-1 leach test on a 145 kg block of a cast cement waste form. The leach vessel was a 200 liter Teflon®-lined drum and contained 97.5 liters of deionized water. The results of this large-scale leach test were compared with the results of standard MCC-1 tests (40 ml) on smaller samples of the same waste form. The ratio of leachate volumes between the large and small scale tests was 2500 and the ratio of sample masses was 150,000. The cast cement samples for both tests contained plutonium-doped incinerator ash.

The leachates from these tests were analyzed for both plutonium and the matrix elements. Evaluation of plutonium plateout in the large-scale test indicated that the majority of the plutonium leached from the samples deposits onto vessel walls and little (<3 × 10−12M) remains in solution. Comparison of elemental concentrations in the leachates indicates some differences up to 5X in the concentration in the large- and small-scale tests. The differences are attributed to differences in the solubilities of Ca, Si, and Fe at pH ˜11.5 and at pH ˜12.5. The higher pH observed for the large-scale test is a result of the larger quantities of sodium in the large block of cement.

Measurements of the electrical conductivity of leachate solutions as a function of time can be used as an efficient, informative means of evaluation and comparison in the development of nuclear waste forms and in the preliminary analysis of their corrosion resistance in distilled water. Three separate applications of this technique are described in this work. These are: (1) its use in the optimization of the corrosion resistance of a crystalline wasteform (monazite); (2) a study of the protective ability of the surface layer (“gel layer”) which forms on the nuclear waste glass Frit 21 + 20 wt. % SRW in distilled water, and (3) making comparisons of the overall corrosion resistance of three different nuclear wasteforms (i.e., monazite, SYNROC, and borosilicate glass). A complete solution analysis of the borosilicate glass leachate and a straightforward analysis of the conductivity results agree to within ±20%. In the absence of a complete, time consuming solution analysis, conductivity measurements can be used to estimate reliably the total ionic concentration in the leachate to within a factor of 2.

A study correlating the leachability of 137 Cs from small-scale to large-scale cement forms was performed. The waste forms consisted of (a) organic ion exchange resins incorporated in Portland I cement, with a waste-to-cement ratio of 0.6 and a water-to-cement ratio of 0.4 (as free water) and (b) boric acid waste (12% solution) incorporated in Portland III cement, with a waste-to-cement ratio of 0.7. 137Cs was added to both waste types prior to solidification. The sample dimensions varied from 1 in. ×1 in. to 22 in. × 22 in. (diameter × height). Leach data extending over a period of 260 days were obtained using a modified IAEA leach test. A method based on semi-infinite plane source diffusion model was applied to interpret the leach data, which allows prediction of the amount of 137Cs leached from the forms as a function of leaching time and waste form dimensions. A reasonably good agreement between the experimental and calculated data was obtained for both types of cement waste forms.

The computer code QTERM has been used to calculate the total released activity from a single glass block when in contact with brine in a salt dome repository as a function of: (1) waste form properties, (2) leaching mechanisms, (3) retention or precipitation of specific radionuclides in surface layers, (4) thermal history of the repository and (5) decreasing activity as a function of time.

Selection of the best rock types for radwaste disposal will depend on their having minimal permeability, maximal flow dispersion, minimal chance of forming new wide aperture fractures, maximal ion retention, and minimal thermal and mining disturbance. While no rock is perfect, thinly bedded complex sedimentary sequences may have good properties, either as repository rocks, or as cover to a repository.

Long time prediction of such favorable properties of a rock at a given site may be best modelled from studies of in situ rock properties. Fracture flow, dispersion history, and geological stability can be derived from direct observations of rocks themselves, and can provide the parameters needed for convincing demonstration of repository security for appropriate times.

Evaluation of the threat of natural dissolution of host evaporites to the integrity of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico has taken into consideration (1) the volume of “missing” rock salt, (2) the occurrence (or not) of characteristic dissolution brines, (3) geomorphic features, some of which are unrelated to dissolution, and (4) the time intervals over which dissolution may have been active. Even under the assumption that all “missing” halite was originally present and has been removed by dissolution, there is no evidence of active preferential removal of the lower Salado Formation halite by any geologically reasonable process. The geologic record contains evidence of dissolution in the Triassic and Jurassic; to constrain all removal of basinal halite to the late Cenozoic yields an unrealistically high rate of removal. Application to the lower Salado of a stratabound mechanism known to be active in Nash Draw, a near-surface feature within the Basin, allows a minimum survival time of 2,500,000 years to be predicted for the subsurface facility for storage of radioactive waste at WIPP. This calculation is based on an analysis of all known dissolution features in the Delaware Basin, and takes into account the wetter (pluvial) climate during the past 600,000 years.

This project encompasses investigations of selected natural analogues of minerals and rocks containing radioelements and fission-product elements, to ascertain how radionuclides and their daughter elements may migrate from sites of origin in response to long-term natural processes such as heating due to an igneous intrusion, circulating hydrothermal systems, diffusion through the rock matrix, weathering and erosion. Comparison of natural occurrences with conceptual models based on laboratory data will furnish a test of such models over time periods comparable to or longer than those expected for a repository.

At Oak Ridge National Laboratory the Pumpkin Valley Shale is used as a host formation for hydraulic fracturing waste disposal. Determination of the relationships between the distribution of different lithologies and porosity-permeability trends within this host formation allows these properties, important to hydraulic fracturing operations, to be related to measurable and mappable geological and petrological parameters. It also permits extrapolation of such patterns to little-studied portions of the Pumpkin Valley Shale. Such knowledge better allows for the satisfactory operation and assessment of the hydraulic fracturing at Oak Ridge National Laboratory.

Pertechnetate oxyanion (99TcO4−), a potentially mobile species in leachate from a breached radioactive waste repository, was removed from a brine solution by precipitation with sulfide, iron, and ferrous sulfide at environmental pH's. Maghemite (γ-Fe23 ) and goethite (α-FeOOH) were the dominant minerals in the precipitate obtained from the TcO4−)-ferrous iron reaction. The observation of small particle size and poor crystallinity of the minerals formed in the presence of Tc suggested that the Tc was incorporated into the mineral structure after reduction to a lower valence state. Amorphous ferrous sulfide, an initial phase precipitating in the TcOÝ-ferrous iron-sulfide reaction, was transformed to goethite and hematite (α-Fe2O3 ) on aging. The black precipitate obtained from the TcOÝ-sulfide reaction was poorly crystallized technetium sulfide (Tc2S7 ) which was insoluble in both acid and alkaline solution in the absence of strong oxidants. The results suggested that ferrous- and/or sulfide-bearing groundwaters and minerals in host rocks or backfill barriers could reduce the mobility of Tc through the formation of less-soluble Tc-bearing iron and/or sulfide minerals.

No exact analogue to a radioactive waste disposal system exists but by studying natural analogues of the most important components or subsystems of a waste repository, confidence in long-term predictions can be increased. The U.S. Nuclear Regulatory Commission (NRC) is supporting research on igneous intrusions into proposed repository-type host rock (basalt, tuff, crystalline hard-rock, and salt), on uranium ore body analogues, on actinide solubility as a function of natural complexing agents in closed basin lakes and on field migration of radionuclide species at low-level waste and uranium mill tailings disposal sites. The NRC is studying analogues to ensure that important parameters which might be overlooked in laboratory tests and in modeling because of problems with scaling time, space and geologic complexity can be identified and their importance bounded.