Disposal of high-level nuclear wastes: a geological perspective

1985 ◽  
Vol 49 (351) ◽  
pp. 159-176 ◽  
Author(s):  
A. E. Ringwood
Keyword(s):  
Elevated Temperatures ◽  
Great Majority ◽  
Nuclear Industry ◽  
High Level Waste ◽  
Public Confidence ◽  
Wide Range ◽  
Close Relatives ◽  
High Level ◽  
Drill Holes

AbstractMost countries intend to dispose of their high-level radioactive wastes by converting them into a solidified wasteform which is to be buried within the earth. SYNROC is a titanate ceramic wasteform which has been designed for this purpose on the basis of geochemical principles. It comprises essentially rutile TiO2, ‘hollandite’ Ba(Al,Ti)Ti6O16, zirconolite CaZrTi2O7, and perovskite CaTiO3. The latter three phases have the capacity to accept the great majority of radioactive elements occurring in high-level wastes into their crystal lattice sites. These minerals (or their close relatives) also occur in nature, where they have demonstrated their capacity to survive for many millions of years in a wide range of geological environments. The properties of SYNROC and the crystal chemistry of its constituent minerals are reviewed in some detail and current formulations of SYNROC are summarized. A notable property of SYNROC it its extremely high resistance to leaching by groundwaters, particularly above 100°C. In addition, it can be shown that the capacity of SYNROC minerals to immobilize high-level waste elements is not markedly impaired by high levels of radiation damage. Current investigations are focused on developing a satisfactory production technology for SYNROC and progress towards this objective is described. The high leach resistance of SYNROC at elevated temperatures increases the range of geological environments in which the waste may be finally interred; in particular, SYNROC is well adapted for disposal in deep drill-holes, both in continental and marine environments. The fact that SYNROC is comprised of minerals which have demonstrated long-term geological stability is significant in establishing public confidence in the ability of the nuclear industry to immobilize high-level wastes for the very long periods required.

scholarly journals Transporting and Storing High-Level Nuclear Waste in the U.S.—Insights from a Mathematical Model

2019 ◽  
Vol 9 (12) ◽  
pp. 2437 ◽  
Author(s):  
Sebastian Wegel ◽  
Victoria Czempinski ◽  
Pao-Yu Oei ◽  
Ben Wealer
Keyword(s):  
Nuclear Waste ◽  
The United States ◽  
Nuclear Industry ◽  
High Level Waste ◽  
Geological Repository ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Interim Storage ◽  
Storage Facilities

The nuclear industry in the United States of America has accumulated about 70,000 metric tons of high-level nuclear waste over the past decades; at present, this waste is temporarily stored close to the nuclear power plants. The industry and the Department of Energy are now facing two related challenges: (i) will a permanent geological repository, e.g., Yucca Mountain, become available in the future, and if yes, when?; (ii) should the high-level waste be transported to interim storage facilities in the meantime, which may be safer and more cost economic? This paper presents a mathematical transportation model that evaluates the economic challenges and costs associated with different scenarios regarding the opening of a long-term geological repository. The model results suggest that any further delay in opening a long-term storage increases cost and consolidated interim storage facilities should be built now. We show that Yucca Mountain’s capacity is insufficient and additional storage is necessary. A sensitivity analysis for the reprocessing of high-level waste finds this uneconomic in all cases. This paper thus emphasizes the urgency of dealing with the high-level nuclear waste and informs the debate between the nuclear industry and policymakers on the basis of objective data and quantitative analysis.

Issues in the Timing of High Level Waste Disposal: An International Perspective

2008 ◽  
Author(s):  
Stan Gordelier ◽  
Pa´l Kova´cs
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Nuclear Power ◽  
Nuclear Industry ◽  
High Level Waste ◽  
Geological Disposal ◽  
The Public ◽  
Deep Geological Disposal ◽  
High Level

The world is facing energy difficulties for the future, in terms of security of supply and climate change issues. Nuclear power is virtually carbon free and it contributes to energy security, being a quasi-domestic source. Whilst it cannot provide a complete answer to these challenges, it is certainly capable of providing a significant component of the answer. However, nuclear power remains controversial. In order to gain public acceptance, it is widely recognised that a number of key issues need to be addressed, amongst which is resolution of the high-level radioactive waste (HLW) (including spent fuel) disposal issue. This is an important issue for all countries with an existing nuclear programme, whether or not it is intended that nuclear power should be phased out or expanded — the waste already exists and must be managed in any event. It is equally important for countries planning a new nuclear power programme where none has previously existed. Since nuclear power was first developed over fifty years ago, HLW arisings have been stored as an interim measure. It is widely believed by experts (though not by many opponents of the nuclear industry, nor by the public) that deep geological disposal, after a reasonable cooling time in interim storage, is technically feasible and constitutes a safe option [1] at an acceptable cost. The total volume of HLW from nuclear reactors is relatively small. A key issue, however, is the time-scale for developing such a final disposal solution. Considerations of security and inter-generational equity suggest that geological disposal should be implemented as soon as possible irrespective of whether or not new arisings are created. The question of managing HLW is not necessarily related to the issue of building new nuclear power stations. However, many opponents argue that there has been insufficient demonstration of the long-term safety of deep geological disposal. The same opponents also argue that there should be a moratorium on building new nuclear power plants (NPPs) until the issue of long-term management of HLW is resolved. These arguments have a powerful influence on public opinion towards both the construction of a waste repository and the building of new NPPs. The intent of this paper (developed from the current OECD NEA study on “Timing of High Level Waste Disposal”) is to identify and discuss some of the factors influencing the timing of the implementation of a HLW disposal strategy and to demonstrate to decision makers how these factors are affecting country strategies, based on current experience. Determining an optimum timescale of HLW disposal may be affected by a wide range of factors. The study examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on the timing of HLW disposal and can be balanced in a national radioactive waste management strategy taking the social, political and economic environment into account. There is clear evidence that significant fractions of the public still have serious misconceptions with respect to the issues surrounding nuclear waste. The nuclear industry, together with governments in those countries who would like a component of nuclear power in their energy mix, has a responsibility for and a significant challenge in presenting its case to the public.

Immobilization of high-level waste in ceramic waste forms

1986 ◽  
Vol 319 (1545) ◽  
pp. 63-82 ◽  
Keyword(s):  
Borosilicate Glass ◽  
Nuclear Industry ◽  
Waste Stream ◽  
High Level Waste ◽  
Phase Assemblage ◽  
Crystal Lattices ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Wide Range ◽  
High Level

High-level wastes (HLW) can be incorporated in the crystal lattices of coexisting phases in ceramic waste forms. The properties and performances of ceramic waste forms are largely determined by their phase chemistry, phase assemblage and microstructure. Currently, the best categorized advanced ceramic waste form is SYNROC, a titanate ceramic composed of ‘ hollandite ’ Bat 1(Al,Ti)2^Ti|]*"70 16, zirconolite CaZrTi 2 O 7 , perovskite CaTiO 3 , rutile TiO 2 and minor amounts of metal alloys microencapsulated by the titanate matrix. Two factors contribute to the capacity of synroc to accommodate high (e.g. 20% ) loadings of HLW, together with variations in waste-stream composition. Firstly, the constituent phases can accept, as solid solutions in their crystal lattices, a broad spectrum of cationic species of diverse charge and radius, either singly or by complex substitution mechanisms. Secondly, the phase assemblage itself spontaneously adjusts its modal mineralogy in response to waste stream fluctuations. The presence of both rutile and a source of trivalent titanium (from reaction of rutile with added Ti metal) in the synroc phase assemblage is largely responsible for this flexible and accommodating nature. The titanate minerals in synroc also occur in Nature, where they have survived for many millions of years in a wide range of geological environments. Experimental studies show that synroc is vastly more resistant to leaching by groundwater than borosilicate glass; moreover, its high leach resistance is maintained at elevated temperatures. Experimental and analogue studies indicate that the HLW immobilization properties of synroc are not significantly impaired by radiation damage. These properties show that synroc would provide an effective immobilization barrier for HLW when buried in suitable repositories. They also permit the use of a wider range of geological disposal options than are appropriate for borosilicate glass. In particular, synroc is well suited for disposal in deep drill-holes, both in continental and marine environments. The fact that synroc is composed of minerals that have demonstrated long-term geological stability is important in establishing public confidence in the ability of the nuclear industry to immobilize high-level wastes for the very long periods required.

Long-Term Radioactivity Release From Solidified High-Level Waste -Part I: An Approach To Evaluating Experimental Data

1981 ◽  
Vol 11 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Source Term ◽  
Elevated Temperatures ◽  
Relative Sensitivity ◽  
Reaction Rates ◽  
Waste Form ◽  
High Level Waste ◽  
Leach Rate ◽  
Leaching Experiments ◽  
High Level

Safety and risk analyses for the isolation of radioactive waste in a repository must begin with a source term to quantify the amount of radioactivity released from the waste form under a specific set of conditions. The interaction of the waste form with aqueous solutions is the most important mechanism to consider, as any material released may be dissolved and reach the biosphere. In this regard the behaviour of heat generating high-level waste is of particular importance, because reaction rates are higher at elevated temperatures. A long-term leach rate was derived from previous and continuing experimental work. The purpose of this paper is not to describe the “real case” release but rather to provide guidelines for the design of leaching experiments and determine the required precision for the data. This can be derived from the relative sensitivity of extrapolated leach rates for various parameters measured in laboratory experiments.

scholarly journals Effect of Fe2O3 on the Immobilization of High-Level Waste with Magnesium Potassium Phosphate Ceramic

2019 ◽  
Vol 2019 ◽  
pp. 1-10
Author(s):  
Hailin Yang ◽  
Mingjiao Fu ◽  
Bobo Wu ◽  
Ying Zhang ◽  
Ruhua Ma ◽  
...  
Keyword(s):  
Sintering Temperature ◽  
Potassium Phosphate ◽  
High Level Waste ◽  
Radioactive Nuclide ◽  
Hydration Reaction ◽  
Sintering Process ◽  
Positive Role ◽  
Ceramic Structure ◽  
High Level

For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

Flow Barrier System for Long-Term High-Level-Waste Isolation: Experimental Results

1998 ◽  
Vol 124 (1) ◽  
pp. 88-100 ◽  
Author(s):  
James L. Conca ◽  
Michael J. Apted ◽  
Wei Zhou ◽  
Randolph C. Arthur ◽  
John H. Kessler
Keyword(s):  
Experimental Results ◽  
High Level Waste ◽  
High Level

Dose-Dependence of Pb-Ion Implantation Damage in Zirconolite, Hollandite, and Zircon

1981 ◽  
Vol 11 ◽  
Author(s):  
T. J. Headley ◽  
G. W. Arnold ◽  
C. J. M. Northrup
Keyword(s):  
Radiation Damage ◽  
Structural Damage ◽  
Accelerated Aging ◽  
Dose Dependence ◽  
High Level Waste ◽  
Long Term Stability ◽  
Waste Forms ◽  
Recent Approach ◽  
High Level

The long-term stability of nuclear waste forms is an important consideration in their selection for safe disposal of radioactive waste. Stability against long-term radiation damage is particularly difficult to assess by short-term laboratory experiments. Much of the displacement damage in high-level waste forms will be generated by heavy recoil nuclei emitted during the α-decay process of long-lived actinide elements. Hence, an accelerated aging test which reliably simulates the α-recoil damage accumulated during thousands of years of storage is desirable. One recent approach to this simulation is to implant the waste form with heavy Pb-ions.I- 6 If the validity of this approach is to be fully assessed, two important questions which have not yet been investigated must be answered.(1) Is the structural damage, including cumulative effects, similar for irradiation by Pb-ions and a-recoil nuclei in a given material? (2) Is the dose-dependence of the accumulated damage similar? The purpose of this investigation was to assess the extent of these similarities in selected materials. We utilized transmission electron microscopy (TEM) to characterize the radiation damage and measure its dose-dependence.

An Approach to Analysis of High Level waste Container Performance in an Unsaturated Repository Environment

1993 ◽  
Vol 333 ◽  
Author(s):  
John C. Walton ◽  
Narasi Sridhar ◽  
Gustavo Cragnolino ◽  
Tony Torng ◽  
Prasad Nair
Keyword(s):  
Dominant Role ◽  
Near Field ◽  
Temperature Fields ◽  
Localized Corrosion ◽  
High Level Waste ◽  
Quantitative Methodology ◽  
Waste Container ◽  
Wide Range ◽  
Engineered Barrier ◽  
High Level

ABSTRACTOne of the requirements for the performance of waste packages prescribed in 10CFR 60.113 is that the high level waste must be “substantially completely” contained for a minimum period of 300 to 1000 years. During this period, the radiation and thermal conditions in the engineered barrier system and the near-field environment are dominated by fission product decay. In the present U.S design of the engineered barrier system, the outer container plays a dominant role in maintaining radionuclide containment. A quantitative methodology for analyzing the performance of the container is described in this paper. This methodology enables prediction of the evolution of the waste package environment in terms of temperature fields, stability of liquid water on the container surface, and concentration of aggressive ions such as chloride. The initiation and propagation of localized corrosion is determined by the corrosion potential of the container material and critical potentials for localized corrosion. The coiTOsion potential is estimated from the kinetics of the anodic and cathodic reactions including oxygen diffusion through scale layers formed on the container surface. The methodology described is applicable to a wide range of metals, alloys and environmental conditions.

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