Application of Polymers for the Long-Term Storage and Disposal of Low- and Intermediate-Level Radioactive Waste

2004 ◽  
Vol 145 (1) ◽  
pp. 82-101 ◽  
Author(s):  
Hugues W. Bonin ◽  
Michael W. Walker ◽  
Van Tam Bui
Keyword(s):  
Radioactive Waste ◽  
Intermediate Level ◽  
Long Term Storage ◽  
Intermediate Level Radioactive Waste ◽  
Term Storage

Towards Solving the Problems of the Solid Radioactive Wastes From Decommissioned Soviet-Designed Nuclear Submarines

2007 ◽  
Author(s):  
Bjo̸rn Borgaas ◽  
Ingar Amundsen ◽  
Ole Reistadt
Keyword(s):  
Radioactive Waste ◽  
Nuclear Safety ◽  
Long Term Storage ◽  
Bilateral Agreement ◽  
Solid Radioactive Waste ◽  
The Individual ◽  
Waste Handling ◽  
Term Storage ◽  
Shipyard Workers

This paper describes the cooperation and inspection arrangements for verifying the dismantlement of non-strategic Russian submarines, including waste handling, in accordance with the provisions given in the bilateral agreement between the Governments of Russia and Norway and the Norwegian Plan of Action for Nuclear Safety. The main concerns during the project implementation are nuclear safety, the individual safety of workers and the safe and secure handling of radioactive and toxic wastes. Based on data provided by the shipyard on individual exposure, one dismantlement project results in average in an effective dose for shipyard workers between 1,1 and 1,9 mSv, depending on the scope of the work for each worker. The main conclusion is that the present handling of solid radioactive waste is not optimal as such waste now is being put into the reactor compartments without the possibility for adequate control, including the retrieval for repacking or transfer to adequate storage facilities, after the reactor compartments have been placed in the long term storage facility in Saida Bay. Today there is no repository in Northwest-Russia for defence-related radioactive waste.

Decommissioning and Dismantling Solution Development for Volodarsky Civil Nuclear Fleet Support Ship

2009 ◽  
Author(s):  
Konstantin N. Koulikov ◽  
Rinat A. Nizamutdinov ◽  
Andrey N. Abramov ◽  
Anatoly I. Tsubanikov
Keyword(s):  
Radioactive Waste ◽  
Radiation Safety ◽  
Pollution Source ◽  
Long Term Storage ◽  
Murmansk Region ◽  
State Corporation ◽  
Solid Radioactive Waste ◽  
Federal Special Purpose Program ◽  
Term Storage

Having about 200 tons of solid radioactive waste aboard, the Volodarskiy Floating Technical Base (FTB) is a potential radiation pollution source for the Murmansk region and Kola Bay, as her long-term berthing negatively affects the hull structures. Thereby, Atomflot collaborated with ANO Aspect-Konversia and JSC NIPTB Onega within the frameworks of Federal Special-purpose Program “Assurance of Nuclear and Radiation Safety for 2008 and for the period up to 2015” and developed the Volodarskiy FTB dismantling concept. In 2008 in the course of development of the Volodarskiy FTB dismantling concept the following works were carried out: 1) vessel condition survey, including SRW radiological analysis; 2) feasibility study of the Volodarskiy FTB dismantling alternatives. In this regard the following alternatives were analyzed: – formation of the package assembly in the form of vessel’s undivided hull for durable storage in the Saida long-term storage facility (LTSF); - formation of individual SRW package assemblies for durable storage in the Saida LTSF; - comprehensive recycling of all solid radioactive waste by disposal in protective containers. 3) selection and approval of the dismantling alternative. The alternative of formation of individual SRW package assemblies for durable storage in the Saida LTSF was selected by the Rosatom State Corporation. In this case the works will be performed on a step-by-step basis at the Atomflot enterprise and SRE Nerpa. The conceptual dismantling technology was developed for the selected Volodarskiy FTB dismantling option. The proceedings contain description of options, analysis procedure and proposal for further study of mentioned challenge.

Assessment of Package Performance During Long-Term Storage of Intermediate-Level Radioactive Waste

2003 ◽  
Author(s):  
A. V. Chambers ◽  
W. R. Rodwell ◽  
M. Kelly ◽  
A. R. Hoch ◽  
A. J. Baker
Keyword(s):  
Extended Period ◽  
Intermediate Level ◽  
Passive Safety ◽  
Long Term Storage ◽  
Waste Storage ◽  
Disposal Option ◽  
Intermediate Level Radioactive Waste ◽  
The Uk ◽  
The Impact

In the UK, the need for the long-term management of intermediate-level radioactive wastes arises from the absence of an established deep disposal option. As a consequence there is a requirement to demonstrate the ‘passive safety’ of waste packages in stores over a necessarily lengthy period (perhaps 150 years or more). ‘Passive safety’ is taken here to imply that the waste packages themselves would remain safe without intervention; it is assumed however, that the store building and associated infrastructure would need to be actively maintained and that the store may need to be ventilated. Importantly, methods for waste storage also need to ensure that possible future options for disposal are not foreclosed. An extended period of storage suggests that emphasis will need to be placed by waste producers on understanding a number of issues such as: • the impact of corrosion on container integrity; • the assessment of container vents and other features as a pathway for the release of radioactivity; • the extent to which wastes are conditioned for storage and the selection of appropriate methods; • the generation and behaviour of toxic or flammable gases in the storage facility; and the mitigating steps that might be needed to address adverse impacts (e.g. the type of ventilation that the store requires). In this paper, we review the requirements and current state of knowledge relevant to the assessment of operational releases of radioactive, flammable and toxic gases from wastes in long-term interim storage.

Cementitious Materials for Radioactive Waste Management Within IAEA Coordinated Research Project

2011 ◽  
Author(s):  
Zoran Drace ◽  
Michael I. Ojovan
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Cementitious Materials ◽  
Radioactive Waste Management ◽  
Long Term Storage ◽  
Waste Immobilization ◽  
Materials Used ◽  
Cementitious Systems ◽  
Term Storage

The IAEA Coordinated Research Project (CRP) on cementitious materials for radioactive waste management was launched in 2007 [1, 2]. The objective of CRP was to investigate the behaviour and performance of cementitious materials used in radioactive waste management system with various purposes and included waste packages, wasteforms and backfills as well as investigation of interactions and interdependencies of these individual elements during long term storage and disposal. The specific research topics considered were: (i) cementitious materials for radioactive waste packaging: including radioactive waste immobilization into a solid waste form, (ii) waste backfilling and containers; (iii) emerging and alternative cementitious systems; (iv) physical-chemical processes occurring during the hydration and ageing of cement matrices and their influence on the cement matrix quality; (v) methods of production of cementitious materials for: immobilization into wasteform, backfills and containers; (vi) conditions envisaged in the disposal environment for packages (physical and chemical conditions, temperature variations, groundwater, radiation fields); (vii) testing and nondestructive monitoring techniques for quality assurance of cementitious materials; (viii) waste acceptance criteria for waste packages, waste forms and backfills; transport, long term storage and disposal requirements; and finally (ix) modelling or simulation of long term behaviours of cementations materials used for packaging, waste immobilization and backfilling, especially in the post-closure phase. The CRP has gathered overall 26 research organizations from 22 Member States aiming to share their research and practices on the use of cementitious materials [2]. The main research outcomes of the CRP were summarized in a summary report currently under preparation to be published by IAEA. The generic topical sections covered by report are: a) conventional cementitious systems; b) novel cementitious materials and technologies; c) testing and waste acceptance criteria; and d) modelling long term behaviour. These themes as well as all contributions of participating organizations were further developed in the individual reports to be presented in the IAEA publication. The CRP facilitated the exchange of information and research co-operation in resolving similar problems between different institutions and contributed towards improving waste management practices, their efficiency and general enhancement of safety.

The Management of Historic Intermediate Level Radioactive Waste in the United Kingdom

2003 ◽  
Author(s):  
David Horsley ◽  
Bruce McKirdy
Keyword(s):  
United Kingdom ◽  
Radioactive Waste ◽  
Waste Management ◽  
Storage Conditions ◽  
Intermediate Level ◽  
The United Kingdom ◽  
Environmentally Sound ◽  
Intermediate Level Radioactive Waste ◽  
The Uk

Nirex is the organisation responsible for long-term radioactive waste management in the UK. Our Mission is to provide the UK with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. The United Kingdom has a significant legacy of long-lived intermediate level radioactive waste. This has arisen from 50 years of investigation and exploitation of nuclear technology. Some of the waste is stored in old facilities that do not provide the standards of containment that would be incorporated in modern facilities. Also the risk to people and to the environment from the inventory in these facilities will increase with time as the structures age and degrade, increasing the chance of containment failure. There is, therefore, a need to retrieve this raw waste and process and package it to make it demonstrably safe for continued storage, pending a decision on disposal of radioactive waste. This packaging should, as far as is practicable, be compatible with the UK long-term waste management strategy. Nirex has developed its Phased Disposal Concept for intermediate and low-level radioactive waste. Based on that concept, Nirex has developed waste package specifications and carries out assessments of waste packaging proposals. For legacy wastes it may not always practicable to demonstrate full compliance with all Nirex disposal criteria. This paper describes an approach, agreed between Nirex and BNFL, for managing these wastes. The proposed approach takes account of long-term waste management issues whilst recognising the need for timely improvement of storage conditions.

Numerical Assessment of the Long-Term Safety of the Morsleben Repository for Low- and Intermediate-Level Radioactive Waste

2009 ◽  
Author(s):  
Ju¨rgen Wollrath ◽  
Ju¨rgen Preuss ◽  
Dirk-A. Becker ◽  
Jo¨rg Mo¨nig
Keyword(s):  
Radioactive Waste ◽  
Radiation Protection ◽  
Total Activity ◽  
Intermediate Level ◽  
Probabilistic Uncertainty ◽  
Numerical Assessment ◽  
Safety Case ◽  
Calculation Results ◽  
Intermediate Level Radioactive Waste

The Morsleben repository has been in operation since 1971 as a repository for low- and intermediate-level radioactive waste. Until the end of the disposal phase in 1998 a waste volume of about 37,000 m3 with a total activity of 4.5·1014 Bq was disposed of. Currently, the German Federal Office for Radiation Protection (BfS) is applying for the licence to finally close the repository. Concerning the possible release of radionuclides to the biosphere, the repository is subject to German radiation protection regulations. Their fulfilment has to be proven by means of numerical calculations as a part of the safety case. A simplified repository model has been developed by GRS and used for calculating the consequences of different scenarios and variants, as well as for a probabilistic uncertainty and sensitivity analysis. The application for licensing is, among others, based on these results. In this paper the main features of the model and the underlying assumptions, as well as the most important calculation results are presented and explained.

The Hydrogeologic Environment for a Proposed Deep Geologic Repository in Canada for Low and Intermediate Level Radioactive Waste

2011 ◽  
Author(s):  
Jonathan F. Sykes ◽  
Stefano D. Normani ◽  
Yong Yin ◽  
Mark R. Jensen
Keyword(s):  
Radioactive Waste ◽  
Solute Transport ◽  
Regional Scale ◽  
Intermediate Level ◽  
Geologic Repository ◽  
Site Specific ◽  
Ordovician Limestone ◽  
Intermediate Level Radioactive Waste ◽  
The Impact

A Deep Geologic Repository (DGR) for low and intermediate level radioactive waste has been proposed by Ontario Power Generation for the Bruce nuclear site in Ontario, Canada. As proposed the DGR would be constructed at a depth of about 680 m below ground surface within the argillaceous Ordovician limestone of the Cobourg Formation. This paper describes the hydrogeology of the DGR site developed through both site characterization studies and regional-scale numerical modelling analysis. The analysis provides a framework for the assembly and integration of the site-specific geoscientific data and examines the factors that influence the predicted long-term performance of the geosphere barrier. Flow system evolution was accomplished using both the density-dependent FRAC3DVS-OPG flow and transport model and the two-phase gas and water flow computational model TOUGH2-MP. In the geologic framework of the Province of Ontario, the DGR is located on the eastern flank of the Michigan Basin. Borehole logs covering Southern Ontario combined with site-specific data from 6 deep boreholes have been used to define the structural contours and hydrogeologic properties at the regional-scale of the modelled 31 sedimentary strata that may be partially present above the Precambrian crystalline basement rock. The regional-scale domain encompasses an approximately 18500km2 region extending from Lake Huron to Georgian Bay. The groundwater zone below the Devonian includes units containing stagnant water having high concentrations of total dissolved solids that can exceed 300g/L. The Ordovician sediments are significantly under-pressured. The horizontal hydraulic conductivity for the Cobourg limestone is estimated to be 2 × 10−14 m/s based on straddle-packer hydraulic tests. The low advective velocities in the Cobourg and other Ordovician units result in solute transport that is diffusion dominant with Peclet numbers less than 0.003 for a characteristic length of unity. Long-term simulations that consider future glaciation scenarios include the impact of ice thickness and permafrost. Solute transport in the Ordovician limestone and shale was diffusion dominant in all simulations. The Salina formations of the Upper Silurian prevented the deeper penetration of basal meltwater.

Doses Forming Paths of External Irradiation of Personnel on Radioactive Waste Disposal Facilities

2003 ◽  
Author(s):  
T. V. Yakimenco ◽  
T. I. Paramonova ◽  
V. A. Smirnov
Keyword(s):  
Radioactive Waste ◽  
Fuel Cycle ◽  
Pollution Prevention ◽  
Radioactive Waste Disposal ◽  
External Irradiation ◽  
Long Term Storage ◽  
Average Activity ◽  
Term Storage ◽  
Centralized System

The centralized system of radioactive waste management formed beyond of a nuclear fuel cycle exist in Russia. In territory of Russian Federation 16 regional specialized combines (SO) “Radon” are located, the majority of them is entered into operation in the first half 60th years. The specialized combines “Radon” are carrying out centralized collection, RW transportation, processing, localization and long-term storage of RW low and average activity, delivery service of new sources of ionizing radiation to users, take part in liquidation of radioactive contamination of served territory, carry out an nature-conservative measures of pollution prevention of the environment fromradioactive substances and ensuring radiation safety of population. For years of their existence total number of the personnel has made hundreds person. In most cases the personnel of combines are under a radiation control since time of introduction of objects in operation, about 40 years.

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