Investigation of Groundwater Composition in Relation to Spent Nuclear Fuel Disposal

1983 ◽  
Vol 26 ◽  
Author(s):  
Fred Karlsson
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Carbonate Content ◽  
Field Equipment ◽  
Carbonate Concentration ◽  
Groundwater Environment ◽  
Suitable Site ◽  
High Level ◽  
High Degree ◽  
Sensitive Parameters

ABSTRACTGroundwater from boreholes in granitic rock at six different sites in Sweden has been sampled and analyzed. Sensitive parameters such as redox potential, pH sulphide– and oxygen content have been measured with a field equipment. This is an integrated part of a program of geological, geophysical, geochemical and hydrogeological investigations with the final aim to select a suitable site for a high-level radioactive waste repository. According to present results deep granitic groundwaters are reducing due to the presence of iron(II) ions. The pH is normally ranging from 7 to 9. The total sulphide content is generally less than 0.5 mg/l. The normal alkalinity range is 90-275 mg/l. Copper, which has been suggested as canister material in the present concept for spent nuclear fuel disposal, is stable in this groundwater environment, except for a very slow sulphide corrosion and a limited initial attack of oxygen from the emplacement operations. The dissolution of the spent nuclear fuel matrix, UO2, is ultimately controlled by the total carbonate concentration. Carbonate content, pH and redox conditions will also be decisive for the mobilities of actinides and technetium. The conditions are generally favouring a high degree of retention for these species in the undisturbed deep groundwater rock environment.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

Technology Development to Support OCRWM Transportation Activities

2004 ◽  
Author(s):  
William H. Lake ◽  
Nancy Slater-Thompson ◽  
Ned Larson ◽  
Franchone Oshinowo
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Private Sector ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Yucca Mountain ◽  
Department Of Energy ◽  
High Level ◽  
Transportation Program

Technology development activities are being conducted by the Department of Energy, Office of Civilian Radioactive Waste Management to support spent nuclear fuel and high-level radioactive waste transport to the federal repository at Yucca Mountain, Nevada in 2010. The paper discusses the motivation for pursuing transport technologies for a private sector operated transportation program, and describes some of the current technologies being pursued.

Key Aspects of the H12 Safety Case

2000 ◽  
Vol 663 ◽  
Author(s):  
Hiroyuki Umeki
Keyword(s):  
Nuclear Fuel ◽  
General Public ◽  
Spent Nuclear Fuel ◽  
Atomic Energy Commission ◽  
Management Program ◽  
Geological Disposal ◽  
Safety Case ◽  
Geological Formation ◽  
High Level ◽  
Key Aspects

ABSTRACTIn Japan, as outlined in the overall high-level radioactive waste (HLW) management program defined by the Japanese Atomic Energy Commission (AEC, 1994), HLW from reprocessing of spent nuclear fuel will be immobilized in a glass matrix and stored for a period of 30 to 50 years to allow cooling. It will then be disposed of in a deep geological formation. Pursuant to the overall HLW management program, an organization with responsibility for implementing HLW disposal will be established around the year 2000. This will be followed by site selection and characterization, demonstration of disposal technology, establishment of the necessary legal infrastructure, relevant licensing applications and repository construction, with the objective of starting repository operation by the 2030s and no later than the mid 2040s.The HLW disposal program is currently in the research and development (R&D) phase and the Japan Nuclear Cycle Development Institute (JNC) has been assigned as the leading organization responsible for R&D activities. The aim of the R&D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan and to promote understanding of the safety concept not only in the scientific and technical community but also by the general public. One of the features of the R&D program is that its progress is documented at appropriate intervals, with a view to clearly determining the level of achievement of the program and to promote understanding and acceptance of the geological disposal strategy by the general public. As a major milestone, the Power Reactor and Nuclear Fuel Development Corporation (PNC, now JNC) submitted a first progress report, referred to as H3 (PNC, 1992), in September 1992.

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.

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