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

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.

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.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

Long-Term Release from High Level Waste Glass - Part IV: The Effect of Leaching Mechanism

1984 ◽  
Vol 44 ◽  
Author(s):  
Eberhard Freude ◽  
Bernd Grambow ◽  
Werner Lutze ◽  
Harald Rabe ◽  
Rodney C. Ewing
Keyword(s):  
Linear Time ◽  
High Surface ◽  
High Level Waste ◽  
Corrosion Mechanism ◽  
Surface Areas ◽  
One Year ◽  
The Stability ◽  
High Level ◽  
Long Term Behavior

During the past ten years extensive data have been determined for the corrosion of nuclear waste forms in short-term laboratory experiments (usually less than one year). The long-term behavior of glass has been inferred by: (1) the acceleration of corrosion rates at high temperatures [1]; (2) the use of high surface areas of the glass to small volumes of solution [1]; and the analysis of natural glasses altered over long periods of geologic time [2, 3]. The most recent efforts have concentrated on understanding the mechanisms of corrosion [1, 4, 5]. The corrosion mechanism may be used to make long-term extrapolations of the “stability” of the waste form. In this paper, we consider a linear time dependence for the corrosion under near saturation conditions and use a rate equation in the QTERM code [6, 7, 8] to model the long-term behavior of the German glass, C-31−3EC [9], JSS A [10, 11] and SRL TDS 131 [1]. The data base for C-31−3EC has been published elsewhere [9, 12, 13, 14], and we include experimental work completed by Rabe for boron and silica, at 200°C.

Long-Term Radioactivity Release from Solidified High-Level Waste - Part II Parametric Study of Waste form Properties, Temperature and Time

1982 ◽  
Vol 15 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Parametric Study ◽  
Computer Code ◽  
Salt Dome ◽  
Waste Form ◽  
Glass Block ◽  
High Level Waste ◽  
Surface Layers ◽  
History Of ◽  
High Level

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.

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