Safety and Sustainability Implications of Long Term Storage of Radioactive Waste

Author(s):  
John Rowat

Storage and disposal of radioactive waste are complementary rather than competing activities, and both are required for the safe management of wastes. Storage has been carried out safely within the past few decades, and there is a high degree of confidence that it can be continued safely for limited periods of time. However, as the amounts of radioactive waste in surface storage have increased, concern has grown over the sustainability of storage in the long term and the associated safety and security implications. In response to these concerns, the IAEA has prepared a position paper [1] that is intended for general readership. This presentation will provide a summary of the position paper, and a discussion of some safety issues for further consideration. A key theme is the contrast of the safety and sustainability implications of long term storage with those of early disposal. A number of factors are examined from different points of view, factors such as safety and security, need of maintenance, institutional control and information transfer, community attitudes and availability of funding. The timing and duration of the process of moving from storage to disposal, which are influenced by factors such as the long timeframes required to implement disposal and changing public attitudes, will also be discussed. The position paper focuses on the storage of three main types of waste: high level waste from the reprocessing of nuclear fuel, spent nuclear fuel that is regarded as waste and long-lived intermediate level radioactive waste. Long term storage of mining and milling waste, and other large volumes of waste from processes involving the use of naturally occurring radioactive materials are not discussed. Specialist meetings were held last year by the IAEA on the sustainability and safety of long-term storage to establish and discuss the issues where a broad consensus exists, and to investigate areas where issues remain unresolved. Within the technical community, it is widely agreed that perpetual storage is not considered to be either feasible or acceptable because of the impossibility of assuring active control over the time periods for which these wastes remain potentially hazardous. For high-level and long-lived radioactive waste, the consensus of the waste management experts is that disposal in deep underground engineered facilities — geological disposal — is the best option that is currently available, or likely to be available in the foreseeable future.

Author(s):  
Zenghu Han ◽  
Ralph Fabian ◽  
Ron Pope ◽  
Yung Liu ◽  
James Shuler

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation, Office of Environmental Management, has sponsored a suite of training courses that are conducted annually by Argonne National Laboratory (Argonne) in support of safety and security of nuclear and other radioactive material packages. One of these courses conducted by Argonne since 2000 is the Application of the ASME Code to Radioactive Material Transportation Packaging, which was expanded significantly in 2014 to include dry storage casks, resulting in a change in course title to the Application of the ASME Code to Radioactive Material Packaging/Cask. The purpose of the course is to provide guidance for the application of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code (“ASME Code”) to transportation packaging and storage cask of radioactive materials, including used (or spent) nuclear fuel and high-level waste, and to facilitate the design, fabrication, examination, and testing of packagings and casks. Both regulatory requirements in 10 CFR Parts 71 and 72 and the ASME Code requirements for transportation and storage containments are addressed, with emphasis on the Code Section III, Division 3, “Containments for Transportation and Storage of Spent Nuclear Fuel and High Level Radioactive Material and Waste.” Among the specific topics covered are the application of the ASME Code requirements to structural materials, containments, loading and design; the design of containment internal support structures and buckling analysis; fabrication, welding, examination, and test requirements; quality assurance; physical testing, structural and thermal modeling and analysis considerations; and containment, shielding, and criticality analysis considerations. Special topics covered include non-Code materials, hydrogen gas generation, and aging management for extended long-term storage of used fuel and subsequent transportation. The expanded training course was offered in June 2014 at Argonne with 27 participants representing mainly industry and government agencies. On the basis of the feedback and course evaluation by the participants, the course may be expanded from 3 to 4.5 days in the future to allow more time for in-class discussion and exercises, as well as to include additional topics related to aging management for extended long-term storage of used fuel and its post-storage transportation. The course provides insight into the DOE and the U.S. Nuclear Regulatory Commission (NRC) transportation and storage cask certification processes. The target audience is DOE, DOE contractors, other agency personnel, and commercial transportation packaging and storage cask engineering employees. Those responsible for designing, fabricating, testing, or packaging and casks, as well as preparing or reviewing the associated Safety Analysis Reports, will also benefit from the course.


2017 ◽  
Vol 153 ◽  
pp. 07035 ◽  
Author(s):  
Mikhail Ternovykh ◽  
Georgy Tikhomirov ◽  
Ivan Saldikov ◽  
Alexander Gerasimov

Energy ◽  
2019 ◽  
Vol 170 ◽  
pp. 978-985 ◽  
Author(s):  
R. Poškas ◽  
V. Šimonis ◽  
H. Jouhara ◽  
P. Poškas

2015 ◽  
Vol 14 (3) ◽  
pp. 252-257 ◽  
Author(s):  
Rodney C. Ewing

Author(s):  
A. Meleshyn ◽  
U. Noseck

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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