Estimates of potential radionuclide releases from HLW storage facilities in geologic formations are an integral part of the technical basis for the regulation of nuclear waste disposal. We have attempted to assess the importance of the variation of several variables and model assumptions to the calculation of radionuclide discharge from hypothetical repositories. The conceptual models of the repository sites are consistent with our current understanding of the characteristics of the geologic environments being studied by the Department of Energy. It must be stressed that we have not attempted to accurately model any real sites; at the present time, the available data are not sufficient for this purpose. Large uncertainties exist in the characterization of the solubilities and sorption of radionuclides, in the description of the regional and local hydrogeology and in the mathematical treatment of contaminant transport in the presence of fracture flow and matrix diffusion. The fraction of waste that could be accessed by breaching the repository, the effects of mixing processes within the facility and the probability of human intrusion are not known. We have completed preliminary calculations of potential radionuclide discharges from hypothetical repositories in tuff and bedded salt [1,2]. Our theoretical approach and computational methods are illustrated in this paper with examples for a repository in volcanic tuff. We also briefly discuss the importance of radionuclide source term assumptions to the calculations for a bedded salt repository.

The Fuel Cycle Risk Analysis Division of Sandia National Laboratories has been funded by the U.S. Nuclear Regulatory Commission to develop a methodology for use in assessing the long-term risk from the disposal of radioactive wastes in deep geologic formations. As part of this program, the Dynamic Network (DNET) model was developed to investigate waste/near field interactions associated with the disposal of radioactive wastes in bedded salt formations. The model is a quasi-multi-dimensional network model with capabilities for simulating processes such as fluid flow, heat transport, salt dissolution, salt creep, and the effects of thermal expansion and subsidence on the rock units surrounding the repository. The use of DNET has been demonstrated in the analysis of a hypothetical disposal site containing a bedded salt formation as the host medium for the repository [1]. An example of this demonstration analysis is discussed. Furthermore, the outcome of sensitivity analyses performed on the DNET model are presented.

It has been recognized that coupled thermomechanical, thermohydrologic, and hydromechanical processes may play an important role in the behavior of the geologic formation around a nuclear waste repository. A numerical code, ROCMAS, was recently developed at Lawrence Berkeley Laboratory to calculate coupled heat transfer, fluid flow, and mechanical deformation in a fractured porous medium. Three scoping studies using this model are described in this paper, with the following preliminary conclusions. The thermomechanical simulations of a fractured rock mass show that the fractures can absorb the thermally induced rock deformations, thus possibly explaining the observed behavior of the granite rocks in the experiment at Stripa, Sweden. The variations in fracture aperture profiles in the hydromechanical simulations indicate that the rigid fracture assumptions used in most uncoupled studies may not be a good representation of fracture behavior. The thermohydromechanical model study reveals that perturbations induced by a heater may close a fracture near the heater borehole and stop the water inflow to the borehole.

The Waste Package Performance Assessment (WAPPA) code has recently been completed, verified, and documented within the SCEPTER program for ONWI. This is a systems model of the gradual degradation of physical and retardation barriers comprising a nuclear waste package, under repository isolation conditions, in various geologic media. The model describes and predicts the temporal and spatial extent of the loss of containment capability for each barrier and for the entire waste package, on timescales of hundreds of thousands of years. The output quantifies the residual state of integrity of waste package components as a function of time and calculates heat, and radionuclide fluxes to the repository.

The model includes five distinct degradation process models (Radiation, Thermal, Mechanical, Corrosion, and Leaching), which are driven both internally by waste decay and externally by repository stress and fluids. The process models are coupled at system level via state and coupling variables. The approach to degradation modeling adopted is barrier-integrated and process-sequential, allowing modification of the entire waste package sequentially by the five process models within each time step.

The code is intended to serve as a tool for evaluating both the relative and the absolute performance of waste package design concepts, for guiding or selecting a preliminary design among conceptual designs, and for support of licensing activities.

Surface uplift and buoyancy flow are two of the major long-term, far-field perturbations to the geologic formations around a nuclear waste repository. The allowable surface uplift has been accepted in the literature as a criterion limiting the repository waste loading density. It has also been recognized in generic modeling studies that the buoyancy distortion of the ambient groundwater flow around a repository is a major mechanism for radionuclide transport. However, the buoyancy considerations have yet to be quantified for the purpose of optimal design of a radioactive waste repository. We suggest the possibility of using the buoyancy gradient when compared with the ambient regional (horizontal) gradient as a measurable thermohydrologic factor that controls the design of waste loading. This study indicates that the buoyancy gradient could in some cases become the main controlling factor. To decrease the buoyancy effects, much lower values of waste loading density should be considered in the design of the waste repository.

The Environmental Protection Agency (EPA) has issued a draft standard (40CFR191) [1] which specifies permissible limits for radionuclide releases from a high-level waste repository to the accessible environment. The U. S. Nuclear Regulatory Commission (NRC) has published a proposed rule (10CFR60) [2] which contains technical criteria for geologic disposal of high-level waste to facilitate compliance with the EPA Draft Standard. The three main numerical criteria specified in 10CFR60 are:

The inability of empirical single value elemental Rd values to model radionuclide retardation in natural groundwater systems has demonstrated the need for developing a comprehensive integrated phenomenological model of the physicochemical and hydrologic mechanisms involved in retardation. The model must account for: (1) radioactive decay, (2) chemical precipitation/dissolution of bulk phases, (3) chemical substitution reactions, (4) isotopic exchange reactions, (5) cation and anion exchange, (6) specific adsorption, (7) diffusion into adjacent flow paths, (8) diffusion into fluid not involved in flow, (9) diffusion into the solid matrix and (10) ultrafiltration. The formulation of such a model is proposed, based on a reaction-path-simulation code with modifications for adsorption processes (by a surface-complexation site-binding model) and diffusional processes. The coupling of the physicochemical portion of this model to a ground-water transport code can be achieved by alternately iterating through the chemical and physical portions of the code during each time step. Sensitivity analysis used to determine dominant species, dominant retardation mechanisms and temporal and spatial homogeneities in the flow system will allow the integrated model to be reduced to a manageable form.

The hydrofracture process is currently being used at Oak Ridge National Laboratory (ORNL) for the permanent disposal of locally generated radioactive waste solutions. In this process, the waste solution is mixed with a blend of cement and other solid additives; the resulting grout is then injected into an impermeable shale formation at a depth of 200 to 300 m (700 to 1000 ft). The grout sets a few hours after completion of the injection, fixing the radioactive waste in the shale formation.

Between 1966 and 1979, ORNL has used hydrofracture to dispose of more than 8 million L (2 million gal) of waste grout containing more than 600,000 Ci of radionuclide activity. Calculations show that operational costs are on the order of $0.16/L ($0.60/gal). Construction of a new hydrofracture disposal facility has been completed, and operational injections were initiated in June 1982.

In support of the responsibility of the Nuclear Regulatory Commission (NRC) to promulgate a regulation for disposal of high level waste in a geologic repository, a technical support program was conducted at Brookhaven National Laboratory (BNL). The principal objectives of the program were (1) to identify and provide an understanding of those factors and processes which affect repository performance, and (2) to provide the information needed to develop and support proposed regulations, criteria, and performance objectives that will ultimately be issued as the final rule for disposal of high level radioactive wastes in geologic repositories, known as 10 CFR 60 [1]. As part of the technical support program in the development of nuclear waste form criteria, three main areas of transuranic (TRU) waste management have been evaluated: immobilization processes and waste forms proposed for ultimate geologic disposal of TRU waste; management of TRU waste by decontamination; and potential problems associated with gas generation by certain TRU wastes [2]. The evaluation results from a review of, primarily, relevant DOE research and development efforts reported up to the end of CY 80. Also, only the definitions, criteria, and performance objectives detailed in the Advance Notice of Proposed Rulemaking (10 CFR 60) [3] were considered in this evaluation. This is important to note since significant differences in the waste package definition and in the statement of performance objectives as well as in other areas exist between the Advance Notice [31 and the Proposed Rule [1].

It is widely known that certain areas of the earth are characterized by such high concentrations of toxic elements that their use by plants or animals is precluded. A water hole whose water contains potentially toxic quantities of As is a common example of such a naturally poisoned area. The dry salt basins of the Basin and Range Province in Nevada and western Utah are largely relict inland salt seas and, because many of the salt basins have total interior drainage of both surface and ground water, these basins represent unique geochemical environments for the entrapment and concentration of mobile elements. Through leaching, transport, and ultimate deposition from throughout the water shed, the closed salt basins are hydrologic sinks in which toxic materials such as As, Sb, Hg, and other heavy metals become concentrated in a relatively confined area.

A new method to ceramize inorganic ion exchangers loaded with nuclear waste has been developed. It is simpler and cheaper than methods used previously, e.g. hot pressing. The inorganic ion exchangers, sodium titanate and ZrO2, were turned into final ceramic waste form by mixing them with a Finnish red clay in weight ratio 1:4 at maximum. The tiles moulded frKm the wet, bakeable mixture were ceramized at 1020–1060°C. The leach rates of Sr, Cs and Co from the tiles determind by a 7 dynam c ISO-test were after six months of leaching 10−6−10−7 g/cm2/d, in decreasing order. Mechanically the tiles are very durable: flexural strengths were in the range of 20–45 meganewtons per square meter.

Geophysical measurements can, depending on the techniques employed, be used at low-level nuclear waste disposal sites to map the subsurface stratigraphy, identify interfaces, locate objects and targets, and aid in selecting locations for obtaining sample cores and placing monitoring wells. The geophysical techniques of ground penetrating radar and electrical resistivity have been applied in a complementary manner at nuclear waste disposal sites. The data obtained from the field surveys have been integrated in developing geophysical models of these sites.

A two-layer electromagnetic geologic model has been developed that allows the prediction of radar attenuation and depth of penetration. Field tests conducted at a lowlevel nuclear waste disposal site show values consistent with the model.

Simulations of waste forms that might be produced by slagging pyrolysis incineration of low-level transuranic (TRU) wastes stored at the Idaho National Engineering Laboratory (INEL) have been fabricated containing the transuranic isotopes 237Np, 239pu, 24lAm, 244Cm at levels of approximately 1 μCi per gram of each.

Leach tests were performed using frit and vitrified monolithic specimens of average INEL TRU waste, portland cement monoliths made with frit as aggregate, and vitrified monoliths of INEL soil and simulated Rocky Flats sludge. Static leach tests were performed at 90, 70, 40, and 25°C in deionized water for up to 364 days. Leachates were analyzed for the TRU elements by alpha spectrometry. The following generalizations can be made:

1. Cemented frit and vitrified sludge waste forms produce leachates with the highest pHs (>11) and have the lowest TRU leach rates, 10−4 g/m2.d at 90°C.

2. Neptunium has a higher leach rate than the other three TRU elements by as much as two orders of magnitude for all waste forms tested except cemented frit.

3. Only the vitrified soil samples display a marked temperature dependence for leach rates of all four TRU elements.

At Oak Ridge National Laboratory, we have developed cementitious grout formulations which can incorporate up to 50 wt % of a transuranic (TRU) waste generated during startup operations at the Rocky Flats Plant Building 374, Golden, Colorado. Trial grout mixes were tested in order to minimize the water content and maximize the waste loading. The effects of water reducers and set regulators on these mixes were also examined.

Durability and physical property measurements were performed on formulations containing 40 wt % waste, which are compatible with standard processing equipment. The densities, porosities, and compressive strengths of these solids are reported. Compressive strengths of the samples were found to increase by as much as 69% after the sample had been exposed to leachants for 28 d at 900 C.

Ferrite treatment was found to be effective in removing actinides from Rocky Flats process waste solutions. With one ferrite treatment, plutonium concentrations were consistently reduced from around 10−4 g/l to between 10−7 and 10−8 g/l. Three to four grams of solids were produced per liter of solution treated as compared to 15.7 g/l observed for the flocculant precipitation technique now in use at Rocky Flats. Less cement was required for the immobilization of ferrite solids resulting in further reductions in waste generation. The success of ferrite treatment in laboratory studies has prompted serious consideration of this technique for implementation in plant operations.

Radioactive defense waste currently stored at the Savannah River Plant in liquid form is to be immobilized by incorporation into a borosilicate glass. The glass melter for this process will consist of a refractory lined tank enclosed in a water-cooled, steel vessel and fed by an aqueous slurry of glass frit plus radioactive waste. As an aid to understanding the melting process and scaling data from small melters, conservation principles were used to develop equations for predicting the melt rate for a feedpile of arbitrary radius plus additional information regarding feedpile behavior and energy consumption.

Operation of joule-heated, continuous slurry-fed melters has demonstrated that off-gas aerosols are generated by entrainment of feed slurry and vaporization of volatile species from the melt. Effective off-gas stream decontamination for these aerosols can be obtained by utilizing a suitably designed and operated wet scrubber system. Results are presented for performance tests conducted with an air aspirating - type venturi scrubber processing a simulated melter off-gas aerosol. Mass overall removal efficiencies ranged from 99.5 to 99.8%. Details of the testing program and applications for melter off-gas system design are discussed.

As part of an interagency agreement between the Bureau of Mines, U.S. Department of the Interior, and the U.S. Department of Energy through its contractor, EG&G Idaho, Inc., the Bureau conducted six melting tests at its Albany Research Center to assess the feasibility of melting transuraniccontaining wastes. The tests were conducted with simulated wastes in a conventional refractory-lined electric arc furnace. Charge materials included concrete, soil, metal, wood, CaO- and Na2O-containing chemical waste sludges, cement, and polyethylene mixed in various proportions in both unburned and partially incinerated forms. The investigation showed that it is possible to melt these materials in a 1-metric-ton conventional electric arc furnace and separate the slag and metal provided that suitable fluxes are added to condition the silicious slags. However, the electric arc furnace cannot be considered an efficient incinerator. The molten slags were poured into 210-liter steel drums having a 0.64-cm-thick steel chill plate on the bottom. All slags were tapped from the furnace satisfactorily. The concrete and sludge materials required the most energy for melting (2.2 to 2.6 kwhr/kg). The highest electrode consumption occurred when the sludges were melted (0.04 kg/kg product). A high alumina-chrome refractory is satisfactory for use as a furnace lining in melting these wastes. Offgases and particulates from all of the tests were sampled and analyzed. The greatest amounts of particulate matter in the offgas streams were obtained from melting sludges and incinerated wastes.

It is recommended that if a conventional electric arc furnace is used to melt transuranic waste that it (1) be fed slowly to prevent excessive fumes and flames and (2) be operated continuously to minimize startup problems and improve efficiency.

Studies of inactive uranium tailings piles have shown that tailings sands containing Ra-226 and other radionuclides may be dispersed by wind and water erosion, causing contamination of adjacent areas. To conduct an effective cleanup operation, it is necessary that boundaries of contamination be well defined. To accomplish this, data from surface gamma-ray surveys made under the Measurement Monitoring Program of the U.S. DOE's Uranium Mill Tailings Remedial Action Project (UMTRAP) are first used to delineate a general outline of the contaminated area. Then, data fram portable scintillometer surveys and fran sealed-can gammaray analyses of soil samples are used to more precisely define the perimeter of Ra-226 contamination. These field measurements are supported by radiochemical analyses of randomly selected samples. Because of its adaptability to the widely varying chemical composition of the material in these samples, the camplexing agent EDTA is used in a complexometric leaching procedure to analyze Ra-226. By this procedure, natural concentrations of Ra-226 in soil (∼1 pCi/g) can be measured routinely. The potential limit of detection is in the 0.1–0.5 pCi/g range. Details of the method, which includes leaching of radium followed by radon de-eranation, are described. Comparative data for various soil and tailings samples are presented.

Brookhaven National Laboratory, in order to provide technical assistance to the NRC, has measured a number of physical and chemical characteristics of soils from two commercial radioactive waste disposal sites: one at Barnwell, SC, and the other near Richland, WA. Materials commonly used for the containment and the solidification of radioactive waste are subject to underground corrosion. The data reported can be used to assess the corrosivity of each burial environment on a waste form or high integrity container. Samples of trench backfill soil, believed to be representative of the soil that will contact the buried waste forms were collected and analyzed. Earth resistivities were measured at each site. Laboratory measurement of the resistivities of water saturated backfill soils from both sites show values greater than 104 ohm-cm. This is indicative of soils having relatively small amounts of soluble ions and of soils likely to be very mildly corrosive to steel. The backfill soil from the Barnwell site is acidic having a pH ∼5. Measurements on Hanford backfill soil indicate a neutral environment; pH ∼7. The exchange acidity of the backfill soils from both sites was <2 meq per 100 g of soil and in most cases an alkaline reaction was observed. The pHs and exchange acidities of these soils are consistent with those of mildly corrosive soils. Aqueous extracts of saturated pastes were used to determine the concentrations of the following ions: Ca2+, Mg2+, K+, Na+, HCO3−, SO4=, and Cl−. The sulfide content of each of the soils was measured in a mixture of soil and an antioxidant buffer solution.