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.

The IAEA Coordinated Research Programme on the Performance of High-Level Waste Forms and Packages Under Repository Conditions

1992 ◽  
Vol 294 ◽  
Author(s):  
Vladimir S. Tsyplenkov
Keyword(s):  
Near Field ◽  
Research Programme ◽  
High Level Waste ◽  
Future Research ◽  
Spent Fuel ◽  
Waste Forms ◽  
Safety Assessments ◽  
High Level ◽  
Coordination Meeting

ABSTRACTThe IAEA initiated, in 1991, a Coordinated Research Programme (CRP), with the aim of promoting the exchange of information on the results obtained by different countries in the performance of high-level waste forms and waste packages under conditions relevant to final repository. These studies are being undertaken to obtain reliable data as input to safety assessments and environmental impact analyses, for final disposal purposes. The CRP includes studies on waste forms that are presently of interest worldwide: borosilicate glass, Synroc and spent fuel.Ten laboratories leading in investigation of high-level waste form performance have already joined the programme. The results of their studies and plans for future research were presented at the first Research Coordination Meeting, held in Karlsruhe, Germany, in November 1991. The technical contributions concentrated on effecting an understanding of dissolution mechanisms of waste forms under simulated repository conditions. A quantitative interpretation of the chemical processes in the near field is considered a prerequisite for long-term predictions and for the formulation of a "source term" for performance assessment studies.

Degradation of phosphatic waste forms incorporating long-lived radioactive isotopes

2001 ◽  
Vol 65 (5) ◽  
pp. 589-601 ◽  
Author(s):  
D. Read ◽  
C. T. Williams
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Radioactive Waste Disposal ◽  
Radioactive Isotopes ◽  
Long Term Stability ◽  
Waste Forms ◽  
Trivalent Actinides ◽  
High Level ◽  
Hlw Repository

AbstractThis paper provides a brief perspective on synthetic, phosphate-based waste forms for high level radioactive waste (HLW). Evidence in support of their long-term stability is then discussed by reference to the degradation of natural monazites with emphasis on the fate of released uranium, thorium and the rare earths (REE). It is apparent that the REE can be mobilized and fractionated at temperatures anticipated in a HLW repository (∼200°C). This provides an indication of the likely fate of the trivalent actinides (Am(III), Cm(III)) if incorporated in similar matrices. Thorium, though released on alteration of monazite, tends to re-concentrate locally in secondary, microcrystalline phases. In relative terms, U is readily removed from monazites. Although it can be re-concentrated in alteration products, the potential exists for substantial loss of U to groundwater. The findings of this research have important implications for the performance of radioactive waste disposal systems where there is a clear need for improved chemical data to describe the precipitation-dissolution of phosphate phases. It is concluded that monazite-like ceramics designed for the containment of HLW will retain tetravalent actinides but may release uranium in response to natural degradative processes.

Low Alkaline Cement Used in the Construction of a Gallery in the Horonobe Underground Research Laboratory

2010 ◽  
Author(s):  
Masashi Nakayama ◽  
Haruo Sato ◽  
Yutaka Sugita ◽  
Seiji Ito ◽  
Masashi Minamide ◽  
...  
Keyword(s):  
Fly Ash ◽  
Research Laboratory ◽  
Concrete Lining ◽  
Energy Agency ◽  
Tunnel Support ◽  
Long Term Stability ◽  
Alkaline Conditions ◽  
High Level ◽  
Underground Research Laboratory

In Japan, any high level radioactive waste (HLW) repository is to be constructed at over 300 m depth below surface. Tunnel support is used for safety during the construction and operation, and shotcrete and concrete lining are used as the tunnel support. Concrete is a composite material comprised of aggregate, cement and various admixtures. Low alkaline cement has been developed for the long term stability of the barrier systems whose performance could be negatively affected by highly alkaline conditions arising due to cement used in a repository. Japan Atomic Energy Agency (JAEA) has developed a low alkaline cement, named as HFSC (Highly Fly-ash Contained Silicafume Cement), containing over 60 wt% of silica-fume (SF) and fly-ash (FA). HFSC was used experimentally as the shotcrete material in construction of part of the 140m deep gallery in the Horonobe Underground Research Laboratory (URL). The objective of this experiment was to assess the performance of HFSC shotcrete in terms of mechanics, workability, durability, and so on. HFSC used in this experiment is composed of 40 wt% OPC (Ordinary Portland Cement), 20 wt% SF, and 40 wt% FA. This composition was determined based on mechanical testing of various mixes of the above components. Because of the low OPC content, the strength of HFSC tends to be lower than that of OPC. The total length of tunnel using HFSC shotcrete is about 73 m and about 500 m3 of HFSC was used. The workability of HFSC shotcrete was confirmed in this experimental construction.

Aging Studies of Pu-238 and -239 Containing Calcium Phosphate Ceramic Waste-forms

2013 ◽  
Vol 1518 ◽  
pp. 73-78 ◽  
Author(s):  
Shirley K. Fong ◽  
Brian L. Metcalfe ◽  
Randall D. Scheele ◽  
Denis M. Strachan
Keyword(s):  
Calcium Phosphate ◽  
Accelerated Aging ◽  
Waste Form ◽  
Calcium Phosphate Ceramic ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Aging Studies ◽  
Flow Through ◽  
Pyrochemical Reprocessing

ABSTRACTA calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.

Titanate Waste Forms for High Level Waste — An Evaluation of Materials and Processes

1981 ◽  
pp. 123-130 ◽  
Author(s):  
R. G. Dosch ◽  
A. W. Lynch ◽  
T. J. Headley ◽  
P. F. Hlava
Keyword(s):  
High Level Waste ◽  
Waste Forms ◽  
High Level

Overview of the U.S. Department of Energy Advanced Waste Forms Development

2018 ◽  
Author(s):  
Kimberly Gray ◽  
John Vienna ◽  
Patricia Paviet
Keyword(s):  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
The United States ◽  
Department Of Energy ◽  
High Level Waste ◽  
Fuel Cycles ◽  
Waste Forms ◽  
The U.S

In order to maintain the U.S. domestic nuclear capability, its scientific technical leadership, and to keep our options open for closing the nuclear fuel cycle, the Department of Energy, Office of Nuclear Energy (DOE-NE) invests in various R&D programs to identify and resolve technical challenges related to the sustainability of the nuclear fuel cycle. Sustainable fuel cycles are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety and limit proliferation risk. DOE-NE chartered a Study on the evaluation and screening of nuclear fuel cycle options, to provide information about the potential benefits and challenges of nuclear fuel cycle options and to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States. The identification of these promising fuel cycles helps in focusing and strengthening the U.S. R&D investment needed to support the set of promising fuel cycle system options and nuclear material management approaches. DOE-NE is developing and evaluating advanced technologies for the immobilization of waste issued from aqueous and electrochemical recycling activities including off-gas treatment and advanced fuel fabrication. The long-term scope of waste form development and performance activities includes not only the development, demonstration, and technical maturation of advanced waste management concepts but also the development and parameterization of defensible models to predict the long-term performance of waste forms in geologic disposal. Along with the finding of the Evaluation and Screening Study will be presented the major research efforts that are underway for the development and demonstration of waste forms and processes including glass ceramic for high-level waste raffinate, alloy waste forms and glass ceramics composites for HLW from the electrochemical processing of fast reactor fuels, and high durability waste forms for radioiodine.

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.

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