High-level Waste Storage Research Articles

Radiation damage ininsulators at high doses

Abstract Radiation damage at high doses (approaching or exceeding one displacement per atom) has been studied widely in metals and alloys in view of its importance in fission and fusion reactor technology. Insulators have received less attention, but their use-actual or potential-in reactors and as host materials for radioactive waste has stimulated recent studies. Insulators present conceptual problems compared with metals: electronic as well as atomic processes may contribute to radiation damage and different sub-lattices may not be affected by radiation in the same way. Ionization damage (radiolysis) in NaCl at high doses and temperatures provides a model system where the production of metal and halogen-based defect clusters can be studied experimentally and theoretically in some detail. It also has direct practical relevance to the storage of high-level radioactive waste in salt repositories. Fast neutron damage to high doses will be sustained by insulators used in any future fusion reactors. Basic st...

Thermal-diffusion aging in solidified radioactive wastes

There is particular importance in the storage of high-level wastes (for example, vitrified ones) in the stability of the solid matrix (glass) containing radionuclides. This paper examines the most rapid process in glass component redistribution near inhomogeneities (inclusions, pores, cracks, etc.), as these inevitably arise during vitrification. The authors consider two model problems, whose solutions enable one to estimate the changes in component concentrations near inhomogeneities.

Application of chemsimul for groundwater radiolysis

The application of the radiation chemical computer program chemsimul, for the calculation of radiolysis in connection with the storage of high level waste has been studied. Methods have been developed for the diffusion of gases out of the irradiated system, for the continuous addition of Fe 2+ to and removal of Fe(OH), from the solution, and for reactions with a solid UO 2-wall. The validity of the program has been confirmed by comparing calculated results with results of well-documented experiments described in the literature. In three of these experiments, pure water containing only oxygen and hydrogen was irradiated. In another experiment, the production of Fe(III) and H` by the irradiation of aqueous ferrous sulphate solutions at natural pH values was determined. The agreement between the calculated and experimental results was good in all cases, except for H + in the last case, where a deviation of 20% was found. Even more complex systems, such as bentonite/water mixtures irradiated with γ- or β-radiaton, were simulated by computer calculations. Also in these cases the agreement between experiment and calculation was satisfactory.

The Transport of Uranium and Technetium Through The unsaturated Tuffs, Yucca Mountain, Nevada

AbstractAn 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.

Radiation Dose Deposition and Colloid Formation in a Rock Salt Waste Repository

For a series configurations of high-level waste (HLW) storage in a salt repository, gamma transport and deposition have been calculated together with the heating of the salt around waste containers. These time-dependent data were used to calculate colloid growth due to irradiation using a theory by Jain and Lidiard. The results show that by a proper choice of storage parameters the colloid fraction can be limited to a few percent. Overpacking by a few centimeters of steel will reduce the amount to < 1%. With the methods described a safe and economic design of HLW containers for final disposal will be possible.

WP-CAVE for dry underground storage of high-level nuclear waste : Akeson, B A; Bergman, S G A; Sagefors, K I In: Rock Mechanics: Caverns and Pressure Shafts (papers to the ISRM Symposium, Aachen, 26–28 May 1982)V2, P861–872. Publ Rotterdam: A. A. Balkema, 1982

WP-CAVE for dry underground storage of high-level nuclear waste : Akeson, B A; Bergman, S G A; Sagefors, K I In: Rock Mechanics: Caverns and Pressure Shafts (papers to the ISRM Symposium, Aachen, 26–28 May 1982)V2, P861–872. Publ Rotterdam: A. A. Balkema, 1982

Numerical Simulation of Flow and Transport in Fractured Tuff

ABSTRACTThe unsaturated, fractured tuff of Yucca Mountain in the Nevada Test Site is one of the potential sites for geologic storage of high-level radioactive waste. A modeling study of flow and transport in this geologically complex site is presented. Numerical models of mass and heat flow in conjunction with analytical solutions are being used for sensitivity and pathway analysis studies and to aid in design and interpretation of laboratory and field flow and transport tests in tuff.

Heat transfer in high-level waste management

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed.

Thermomechanical In-Situ Experiments and Finite Element Computations

In the ultimate storage of high level wastes in rock salt thermomechanical phenomena caused by the temperature rises play an important role for elaborating the storage concept. Important factors are both large space effects concerning the repository and the geological formation and small space phenomena influencing to a considerable extent the interaction between waste blocks and the host rock. While the large space aspects are largely inaccessible to an experimental investigation, the small space effects, especially the convergence behavior of individual heated boreholes, are eligible for a systematic experimental and computational study allowing verification of available computational methods and thermomechanical material laws for rock salt.

Migration Rates of Brine Inclusions in Single Crystals of Nacl

Rock-salt deposits have been considered as a possible medium for the permanent storage of high-level radioactive wastes and spent fuel. Brine inclusions present in natural salt can migrate toward the waste if the temperature and the temperature gradients in the vicinity of the radioactive waste are large enough. This migration is due to the dissolution of salt at the hot side of the salt-brine interface, ion diffusion through the brine droplet, and the precipitation of salt at the cold side of the salt brine interface.

Transport of Actinides Through a Bentonite Backfill

Compacted bentonite has been proposed as a suitable backfill material in the Swedish concept for underground storage of high-level waste. The backfill barrier will serve both as a physical barrier, preventing convective water flow and allowing radionuclide penetration only by diffusion, and as a chemical barrier capable of chemically interacting with the radionuclides.

Dedicated-site, interim storage of high-level nuclear waste as part of the management system

Dedicated-site interim storage of high-level reprocessed nuclear waste and of spent fuel rods is proposed as a long-term integral part of the systems approach of the national nuclear waste isolation program. Separation of interim sites for retrievable storage from permanent-disposal repositories should enhance ensurance of the performance of the latter; maintenance of retrievability at separate sites also has many advantages in both safety and possible use of waste as resources. Interim storage sites probably will not be needed beyond about 100 years from now, so the institutional and technical considerations involved in their choice should be much less stringent than those for the selection of permanent sites. Development of interim sites must be concurrent with unabated effort to identify and to develop permanent repositories.

A new irradiation effect and its implications for the disposal of high-level radioactive waste.

Materials containing alkali metals or alkaline earths are sensitized by bombardment with either ions, electrons, or photons to chemical attack by atmospheric moisture. The implications of this effect on the proposed immobilization and long-term storage of high-level nuclear waste in glass or similar materials is discussed.

Physical properties of glass for immobilization of high level radioactive waste

The physical properties of borosilicate glass make it suitable for the storage and disposal of high-level radioactive wastes. Waste glass has excellent radiation stability. As the result of over 10 years of testing it is known that the physical properties are relatively unaffected by high doses of contained radioactivity. Similar detailed studies have shown that the effects of self-heating in canisters of properly-formulated waste glass are also relatively minor. Glass is only one of a number of potentially satisfactory immobilization forms. It differs from the other forms mainly in that its processing has been developed to the point that a waste glass production plant is now operating in France. Using the known properties of waste glass as a baseline for comparison, important factors in the evaluation of alternate waste forms can be delineated. Among the factors that enter into the evaluation are: ability to deal with changing waste compositions, process complexity, waste loading, canister interactions, thermal stability, radiation stability, transmutation effects, and leachability.

Laboratory Simulation of High-Level Liquid Waste Evaporation and Storage

Simulated high-level liquid waste (HLLW) slurries were prepared by combining a series of soluble stock solutions composed of commonly available chemical reagents. The properties simulated included ...

High-Level Waste Ceramics: Materials Considerations, Process Simulation, and Product Characterization

A ceramic is one of the alternatives for solidification and storage of high-level wastes (HLW). The procedure for developing a tailor-made ceramic with HLW ions fixed in mutually compatible, refractory and leach-resistant crystalline phases has been developed. Cold engineering-scale evaluation of an early ceramic formulation at Pacific Northwest Laboratories (PNL) has resulted in a product that is easily crystallized and has more than adequate thermal stability and leaching resistance. When combined with the continuous pelletizing and coatings processes developed at PNL, the results to date demonstrate that the tailor-made ceramic and the multibarrier waste forms are very promising advanced alternatives to glass as an HLW solid. 5 figures, 3 tables.

Geohydrology of Buried Triassic Basin at Savannah River Plant, South Carolina

A buried Triassic basin with possible potential for storage of high-level radioactive wastes was discovered beneath the southern third of the Savannah River Plant site near Aiken, South Carolina. Investigation into the characteristics of the basin was aimed at understanding the geohydrology to determine its compatibility with the safe storage of waste. Before it was completed, the investigation was deferred pending more detailed evaluation of alternative concepts of waste storage and management. However, some of the geohydrologic results of the investigation are reported herein. Seismic surveys, gravity and magnetic surveys, and the drilling of several exploratory wells indicate that the Triassic basin is about 30 mi long, 6 or more mi wide, and perhaps 5,300 ft thick. One well penetrated the Triassic border, a second was in the center of the basin, and a third investigated an inferred intrabasin fault. The rock is predominantly mudstone of very low permeability with a few lenses of poorly sorted gritty sand. The water yield of all the exploratory wells is extremely low, and water-transmitting fractures are virtually nonexistent. In two wells within the basin, heads above land surface have been measured that cannot be explained by connection with a recharge area. Possible explanations are tectonic compression, temperature increase, and osmotic-membrane phenomena. Whether the high head is general over the entire basin or only in segments of it is unknown.

Deformation of rock salt in openings mined for the disposal of radioactive wastes

Deformation of Rock Salt in Openings Mined for the Disposal of Radioactive Wastes With the storage of high-level radioactive waste in salt structures, unique mine stability problems will occur as a result of the elevated temperatures. To predict flow in rock salt, scale models of salt pillars and their surrounding rooms were fabricated from cores taken in the Carey salt mine, Lyons, Kansas. Tests were conducted at temperatures of 22.5°, 60°, 100°, and 200°C for axial loads of 2000, 4000, 6000, 8000, and 10000 psi at each temperature. These tests showed that marked increases occur in the rates of deformation of salt pillars at high loads and especially at elevated temperatures. For all combinations of axial loads and temperatures, it was observed that there is initially a high rate of deformation that diminishes with time. Creep rates were found to continue to decrease even after more than 3 years of testing. An empirical relationship between pillar deformation, stress, temperature, and time was developed from the tests and is expressed as $$\dot \varepsilon = 0.39 \cdot 10^{ - 37} T^{9.5} \sigma ^{3.0} t^{ - 0.70} ,$$ $$\varepsilon = 1.30 \cdot 10^{ - 37} T^{9.5\sigma3.0} t^{0.30} ,$$ where\(\dot \varepsilon\) = strain rate (in./in./hr),e = cumulative deformation (in./in.),T = absolute temperature (°K),σ = average pillar stress (psi), andt = time (hr). For comparative purposes, model pillar tests were conducted on samples of bedded salt, as well as dome salt from six different mines in the United States and from the anticlinal structure of the Asse II salt mine in Northeast Germany. In general, the deformational behavior for the various types of salt was similar at room temperature as well as at elevated temperature even though some variations in the rates of deformation were observed. From model tests it was also observed that greatly accelerated rates of deformation will occur in excavated cavities where thin shale beds occur in the pillars at the roof and floor interfaces. Since the shalcs serve as friction reducers, effective confining stresses in the roof and floor are not transmitted into the pillars; thus the pillars under these conditions are weaker than where shale partings are not present at the tops and bottoms of the pillars.

Geologic, Hydrologic, and Safety Considerations in the Storage of Radioactive Wastes in a Vault Excavated in Crystalline Rock

A recent investigation established the technical feasibility and indicated the high degree of safety that could be afforded by the storage of high-level radioactive wastes in unlined vaults excavated in crystalline rock 1500 ft beneath the surface of the Savannah River Plant near Aiken, S. C.The crystalline rock at the proposed site is covered by 1000 ft of unconsolidated sediments conSisting predominantly of sand and clay. A virtually impermeable layer of clay separates the rock from the overlying sediments in which several prolific water-bearing zones occur. The separation of the waters above and below this clay layer is confirmed by their different chemical composition and by the presence of dissolved helium-bearing gas only in the water in the rocks beneath the clay. Based on geologic and hydrologic information obtained in an intensive drilling and testing program, upper limits on the rates of water movement through the crystalline rock are calculated to be 1.5 to 7 ft/year, depending upon the degree of fracturing of the rock. Comparable data on the unconsolidated sediments lead to a calculated maximum rate of water movement of 350 ft/year.The most significant driving force for the migration of radionuclides from the storage site is derived from the natural water movement, coupled with effects due to dispersion and ion exchange. Characteristics of the waste, heat generation, and radiolysis have, by contrast, only small effects on migration. Three barriers prevent migration of the radionuclides: the very low permeability of the rock in which the storage vault is located, the virtually impermeable clay layer separating the rock and sediments, and the ion exchange properties of the sediments. Anyone of these barriers is capable of confining the radionuclides well within the plant boundaries for a time much greater than the 600-year period required to render the wastes innocuous.

Treatment and Storage of High-Level Radioactive Wastes

"Treatment and Storage of High-Level Radioactive Wastes." Nuclear Science and Engineering, 18(4), pp. 534–535