bluuebunny/crossref_metadata_2025 huggingface dataset:
Radioactive Waste Management in the Czech Republic
Authors: Duda Vitezslav
Published: March 2009

Radioactive waste and spent nuclear fuel are generated in the Czech Republic as a consequence of the peaceful use of nuclear energy and ionising radiation in many industries, particularly in the generation of nuclear energy, health care (therapy, diagnostics), research, and agriculture. The current extent of utilisation of nuclear energy and ionising radiation in the Czech Republic is comparable with that of other developed countries. The Concept of Radioactive Waste and Spent Nuclear Fuel Management is a fundamental document formulating government and state authority strategy for the period up to approximately 2025 (affecting policy up to the end of the 21st century), concerning the organizations which generate radioactive waste and spent nuclear fuel. The Concept puts forward solutions to provide for the disposal of waste in compliance with requirements for the protection of human health and the environment without excessively transferring any of the current impacts of nuclear energy and ionising radiation utilisation to future generations. The Concept was approved by the government of the Czech Republic in 2002. According to the Concept high level waste and spent nuclear fuel generated at the Dukovany and Temeli´n nuclear power plants will eventually be disposed of in a deep geological repository. Such a repository should commence operation in 2065. Work aimed at selecting potentially suitable sites began in 1992, but the final site has not yet been determined. In compliance with the aforementioned Concept, the Radioactive Waste Repository Authority (RAWRA) is responsible for finding two suitable sites till 2015. The current stage of evaluation covers the whole territory of the Czech Republic and involves complex criteria and requirements. On the basis of current findings RAWRA suggested six potential sites for further investigation at the beginning of the year 2003.
Analysis of the Excavated Archaeological Iron Using Xray-CT
Authors: Yoshikawa Hideki , Ueno Kenichi , Honda Takashi , Yamaguchi Shingo , Yui Mikazu
Published: March 2009

In order to evaluate the long-term corrosion behavior of carbon steel, we investigated the rust of archaeological iron buried in soil. It is difficult to obtain experimental data of the long-term corrosion in the laboratory. However, it is possible to obtain corrosion data over several hundred years by using archaeological iron and to develop a reliable model for the long-term corrosion behavior by using such natural analogue data. Since these archaeological samples are very rare and important, we can not get agreement to destroy it for analysis. The rust of the sample has been analyzed no-destructively and quantitatively using high-power X-ray computed tomography. The X-ray strength was developed in the two-demensional image. We observed a rust layer distinct from the inner iron metal as a main body by using X-ray map element concentration. A mass-balance quantity of rust calculation was performed from the amount of corroded layer. A sample of axe which was excavated in the Izumo-Taisha-ruin (Shimane prefecture) was analyzed by using the method. The region in Izumo is famous as the production area of the iron from ancient times in Japan. The axe is traditional Japanese type, made of iron, and probably used for a foundation ceremony of the building. The sample has been buried under the column of the shrine and enveloped by clay. It is assumed that the axe remained under reducing conditions until its discovery in 2001 for about 750 years. We have also investigated the corrosion of the gas pipe buried in the soil in several decades as natural analogue study. By the comparison of these data with the corrosion data of water pipe (cast iron) buried in clay soil at most for 100 year, the results of this study do not exceed the extrapolated pitting corrosion depth based on the corrosion depth of the cast iron pipe.
Combined HRGS/PNCC Systems for the Assay of Plutonium Contaminated Waste Arising From the Drigg Retrieval Project
Authors: Wilkins C. G. , Croft S. , Daniels B. , Wardle S.
Published: March 2009

Between 1959 and 1967, Plutonium Contaminated Material (PCM) was stored in the magazines at the Low Level Waste Facility at Drigg. However, PCM is now classified as Intermediate Level Waste — rather than Low Level Waste — and so a programme to remove the PCM is underway. In support of this programme a suite of safety related non-destructive assay instrumentation has been supplied to the Drigg Site in order to characterise the PCM prior to transportation. Three identical assay systems employing High Resolution Gamma Spectrometry (HRGS) and Passive Neutron Coincidence Counting (PNCC) have been installed in the new waste retrieval facilities on three of the PCM magazines. These Retrieval Module Monitors allow the fissile content and radionuclide inventory of individual items of PCM to be determined before they are placed with other items into 200 litre drums and transported across the Drigg site to a temporary drum storage facility. In addition, two assay systems of a different design have also been installed in a new PCM Drum export facility. The purpose of these Drum Monitors is to determine the fissile content and radionuclide inventory of drums of PCM that have been retrieved from the magazines before they are transported to Sellafield. Like the Retrieval Module Monitors, the Drum Monitors also uses the HRGS and PNCC techniques to measure drums of up to 400 litre capacity but in this case a scanning detector is used instead of three fixed gamma detectors. This paper describes the characteristics of both the Retrieval and Drum Monitors, the particular requirements for these systems and the means by which they were designed, built and tested in order to ensure that they delivered their primary safety function, which was to not underestimate the fissile content of the items of PCM they measure.
Preliminary Calculations of Radionuclide Release From Near-Field to Biosphere Using Computer Code Goldsim
Authors: Prva´kova´ S. , Dˇu´ran Juraj
Published: March 2009

This paper presents preliminary calculations for the performance assessment of the hypothetical disposal facility for spent nuclear fuel in the Slovak Republic. In the concept of geological disposal, two host rock formations are under consideration: granite and clay. Slovakia is now in site selection phase therefore the lack of input data for calculations is compensated from the international reports [1, 2]. The calculations of radionuclide release were made for the spent nuclear fuel disposed in a hypothetical underground repository situated in clay host formation. One-dimensional mathematical model of the radionuclide release through the engineered and natural barrier system up to one million years was simulated using a flexible compartment model tool Goldsim. Both deterministic and probabilistic calculations have been applied. The input parameters for probabilistic calculations are selected from the probability density functions (PDF) using Monte Carlo method. Radionuclide release to the biosphere is considered via contaminated groundwater supply taken from a deep well. Three different exposure pathways were considered: consumption of the root vegetables, consumption of meat and drinking of water from well. The results of preliminary assessment of radiological consequences are expressed in terms of release rates on the aquifer-well interface together with the corresponding individual annual doses for average member of the critical group of adults living in the vicinity of the repository. The regulatory limits regarding the annual doses for the concept of geological disposal in Slovakia are under preparation.
Characterization and Treatment Experience of TRU Sludges During Decommissioning NPP A-1 in Slovak Republic
Authors: Majersky D. , Sekely S. , Katrlik J.
Published: March 2009

Many special tasks have to be solved during the decommissioning of the NPP A-1 in Jaslovske Bohunice, Slovak Republic. One of them is the safe treatment of TRU sludges from the past. After the removal of fuel elements from steel casings, deposits of radioactive sludges at the bottoms of the casings were observed. High gamma radioactivity and contents of transuranic radionuclides in these sludges were the main reason for the selection and preparation of the sludge immobilization directly in the original steel casings. After TV and radiation monitoring, sampling of the sludge and laboratory verification, the in-situ immobilization of the sludge was realized. The applied SIAL matrix (mainly Si and Al compounds) has been developed and optimized to encapsulate the sludge giving high compressive strength and high index of leachability of the resulting product. Realisation of the sludge immobilization in 45 steel casings was performed.
Sellafield Decommissioning Programme: Progress Update
Authors: Challinor S. F.
Published: March 2009

In 1999 the UK government announced a step change in the strategy for the delivery of the UK civil nuclear clean-up programme. BNFL has responded to the Governments announcement by changing the strategic direction and increasing the priority on remediation activities across the Company. BNFL has extensive experience in decommissioning nuclear facilities having undertaken remediation and decommissioning operations on BNFL sites for many years, encompassing a wide range of projects including reactors, fuel cycle plants and Research and Development facilities. This paper describes the challenges posed by, and the progress made, on some of the range of decommissioning projects undertaken on the Sellafield site as part of its decommissioning and remediation portfolio. These decommissioning operations cover a variety of redundant fuel cycle facilities ranging in size and complexity in both beta gamma and alpha contamination environments utilising manual and remote decommissioning techniques to systematically and progressively reduce the hazard on the site.
The Disposal of HLLL Waste in a Granitic Formation: The Status of Research in France
Authors: Mouroux B. , Merceron T. , Vilogeux D. , de Me`redieu J.
Published: March 2009

In accordance with the French Law of 1991 December 30, Andra is preparing a report on the possibilities of geological disposal to be submitted in 2005. For that purpose, a research programme has been launched to assess the suitability of granite geological formations for the implementation of a high-level long-lived radioactive waste repository. The programme relies mainly on Andra’s participation in experiments in Swedish, Swiss and Canadian underground research laboratories. The main issues under investigation include granite fracturing and the behaviour of engineered barriers. Furthermore, a typological analysis of French granites provides an opportunity to apply results to the French geological context. Advances in knowledge afford a better understanding of the phenomena relating to granite and preliminary repository designs. They constitute the basis for design studies and safety analyses to be conducted before 2005.
Development of Foam Floating Decontamination System for Radioactive Solid Wastes From MOX-Fuel Fabrication
Authors: Okada T. , Uematsu S. , Tobita N. , Kato J.
Published: March 2009

The foam floating decontamination (FFD) method consists of floating, carrying and removing processes of minute particle contaminatns, such as MOX powder, by foam. In this study the zirconia powder was used to simulate MOX powder. This study describes the behavior of floating zirconia powder within the FFD method. It has shown that the foam is able to float and to carry the zioconia powders, which have grain size distributions of φ 5–50 μ m, from the bottom to the top of a column. It is expected that the FFD would be very useful as a decontamination method of MOX fuel fabrication installations.
Legitimacy as the Key: The Long-Term Management of Radioactive Waste in the UK
Authors: Murray Chris , Wild David , McCall Ann , Mathieson John , Russell Ben
Published: March 2009

This paper provides an overview of the current status of radioactive waste management in the UK from the point of view of Nirex, the organisation responsible for providing safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. Essentially, it argues that: • the waste exists and must be dealt with in an ethical manner; • legitimacy is the key to public acceptance of any attempt to solve the waste issue; and • credible options and a new political will allow, and indeed, compel this generation to deal with it. In doing this, the paper takes account of a number of recent announcements and ongoing developments in the UK nuclear industry, in particular: • the recent announcement that Nirex is to be made independent of industry; • the Department of Environment, Food and Rural Affairs and Devolved Administrations’ Managing Radioactive Waste Safely consultation exercise; • the creation of the Committee on Radioactive Waste Management to oversee the consultation; • the creation of the Nuclear Decommissioning Authority to manage the civil nuclear site clean-up programme; • proposals for improved regulation of Intermediate Level Waste conditioning and packaging; and • proposals by the European Commission for a new radioactive waste Directive. These institutional and policy changes amount to an evolution of the back-end of the fuel cycle that represents the most radical transformation in the UK nuclear industry for many years. In a large part, this is a transformation made necessary by past failures in trying to impose a solution on the general public. Therefore, in order for these changes to result in a successful long-term radioactive waste management programme, it is necessary to pay as much attention to political and social concerns as scientific and technical ones. Primarily it is crucial that all parties involved act in an open and transparent manner so that the decisions made achieve a high degree of legitimacy and thus public acceptance. Crucially too, the problem must be framed in the correct term — that the waste exists irrespective of the future of nuclear power and that this is an issue that must be addressed now. Thus there is a legitimacy of purpose and scope in moving forward that addresses the ethical imperative of this generation dealing with the waste. Put together with the action the government is taking to create the necessary institutional framework, Nirex believes that for the first time in a generation the UK has the building blocks in place to find a publicly acceptable, long-term solution for radioactive waste.
Decommissioning and Decontamination and Radioactive Waste Management Within the Framework of Remediation of Environmental Liabilities in the Nuclear Research Institute Rez plc
Authors: Podlaha Josef
Published: March 2009

After more than 45 years of activities in the nuclear field, there are many environmental liabilities that shall be remedied in the Nuclear Research Institute Rez plc, Czech Republic. The Nuclear Research Institute Rez plc (NRI) is a leading institution in all areas of nuclear R&D in the Czech Republic. NRI operates two research nuclear reactors and many facilities such as a hot cell facility, research laboratories, technology for radioactive waste (RAW) management, radionuclide irradiators, electron accelerators, etc. Except of facilities that will be decommissioned in the future, there are many retired facilities that will be decommissioned soon. There are three main areas of remediation: [1] decommissioning of old obsolete facilities (e.g. decay tanks, liquid RAW storage tanks, old RAW treatment technology, special sewage system), [2] processing of RAW resulting from operation and dismantling of nuclear facilities (e.g. RAW from reconstruction of the VVR-S research reactor), and [3] elimination of spent fuel from research reactors. The goal is to remedy the environmental liabilities and eliminate the potential negative impact on the environment. Based on this postulate, optimal remedial actions have been selected and recommended for the environmental remediation. Remediation of the environmental liabilities with the large potential impact on the environment has already been started. The most significant items of environmental liabilities are discussed. The main goal is to describe the current state, the progress and the future activities in remediation of environmental liabilities.
The Nirex Coherent Approach to Achieving Criticality Safety
Authors: Wood P. , Askarieh M. , Lock P.
Published: March 2009

The role of Nirex is to provide the UK with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. This will include all intermediate-level waste and some low-level waste (ILW and LLW). Nirex has developed a Phased Disposal Concept based on conditioning the wastes then isolating them deep underground. Based on this phased disposal concept, Nirex advises waste producers on the conditioning and packaging of radioactive waste. The ILW and LLW which comprise the ‘reference volume’ of the Phased Disposal Concept contain approximately 5 t Pu-239 and 28 t U-235. Nirex’s work is based on a number of standards for safety and environmental protection set by the Government. Three different Regulators ensure application of these standards in the transport, operation and post-closure phases. Therefore a coherent approach to criticality safety is required. The approach to criticality safety begins by avoiding criticality resulting from the way in which the waste is packaged, by controlling the package design including the level of fissile material. The concept safety assessments will support a screening level of approximately 5 g Pu-239 equivalent. In specific cases where waste producers indicate that their proposed packages will not comply with the screening level, package-specific criticality safety assessments can be incorporated into the concept through the change control process. Engineering measures are available to prevent criticality for such time as the waste packaging affords a high level of containment. In the long term, however, after deterioration of the physical containment provided by the waste packages, there would be the possibility of movement of fissile material out of the waste packages and subsequent accumulation into new configurations which could in principle lead to a criticality. It is conceivable that a criticality could adversely impact on the performance of a repository after closure because, for example, of the heat that would be produced affecting the engineered barriers. It is therefore necessary to consider the post-closure criticality safety of the repository concept. Nirex is currently undertaking a programme of work on ‘understanding criticality under repository conditions’. The aim of the programme is to obtain a better understanding of the processes that would control the nature and magnitude of a criticality under the particular conditions of the concept repository. An essential component of demonstrating criticality safety is to ensure waste packagers develop operating arrangements, and provide objective evidence in the form of criticality compliance assurance documentation, to demonstrate how fissile material will be controlled to meet levels defined in concept criticality safety assessments. This paper will describe the coherent approach and the processes by which it is applied.
Deployment of the RASP System for Cutting a Contaminated Power Reactor Turbine Shaft
Authors: Sanders M. J. , Jeanes H. , Fray S. , Arnold H.-U.
Published: March 2009

A need was identified for the in-situ size reduction of large components such as glove boxes, tanks and other components. A development programme resulted in a new cutting method that could be remotely deployed for the in-situ segmentation of large components of this type. RWE NUKEM has selected RASP as acronym for this technology Remotely-Operated Advanced Segmentation Process or RASP. A review of the standard manual cutting methods highlighted a number of significant shortcomings including operative safety; high energy cutting tools generating high temperatures; spread of contamination over wide areas; risk from sharp edges and operative fatigue. The development of new technology was therefore focused on improving safety for operatives and developing an inherently safe and flexible technology, which can be used for in-situ segmentation of any large components of various sizes and configurations. Important criteria included: minimal space requirements; fast set up times for use in confined spaces or high dose areas; the ability to cut in any direction; the ability to cut virtually any material and geometry; the minimization of secondary waste and low cutting temperature so flammable materials can be cut. It was concluded that a dry low energy mechanical cutting principle, which did not require cooling, would meet these criteria. The development program produced a low energy tool that uses the principles of a reciprocating saw together with a novel operating system that applies the principles of fuzzy logic. RASP technology can be used for virtually any cutting task and is among the safest and most versatile segmentation technologies for the in-situ processing of nuclear components. After completing a test programme in which a number of metal artefacts were segmented, RASP was been set up to cut a redundant SGHWR turbine shaft into pieces to facilitate packaging. The complete turbine assembly is a large heavy item with a steel shaft of 430mm diameter. The modular RASP framework was set up to enclose the unit and the system put to work and a hoist was set up to support the cut portion of the shaft on completion of the cut. The work was completed safely and without incident.
Spent Resin Treatment Test in the ATR Fugen Nuclear Power Station Using Low Pressure Oxygen ICP
Authors: Sano Kazuya , Higashiura Norikazu , Katagiri Genichi , Fujisawa Morio , Shimamura Takeo
Published: March 2009

Low-pressure oxygen plasma processing using ICP (Inductively coupled plasma) is the technology that takes volume reduction and stabilizing treatment of an organic material such as ion exchange resin mildly. We evaluated, for example, treatment performance such as the volume reduction, the weight reduction, and γ specter, H-3,C-14 for the spent ion exchange resin which has actually been produced in the ATR Fugen power station. As a result, a volume reduction rate was achieved not less than 90% (1/10), and a weight reduction rate not less than 95% (1/20). Co −60 was held not less than 0.998 in a processing chamber. In addition, distribution to an effluent gas was not more than 10−6.
Development of an Online Performance Assessment System
Authors: Takase Hiroyasu , Inagaki Manabu , Noguchi Toshihide
Published: March 2009

Performance assessment (PA), as a tool to support decisions associated with geological disposal, requires the facilitation of various views of stakeholders. A straightforward approach to this is for variety of people to actually try carrying out PA calculations to stimulate, test, and strengthen their arguments. In the past, however, they have been carried out only by a limited number of experts who share a similar discipline. This is due to a number of reasons, e.g., methodology of PA is difficult for a non-expert to follow, availability of resources necessary for PA is limited, and a single expert in a specific field cannot conduct an overall PA calculation because of multidisciplinary nature of PA. To tackle these issues, we developed a system on the World Wide Web that allows a wider spectrum of people to run PA codes and access associated databases on demand. Moreover they can build their own PA models in an object-oriented-online environment. With this system, independent and voluntary network communities can construct their own safety cases so that, collectively, a variety of views can be presented to support stakeholders in forming their opinion. In this paper, the methodology of the online performance assessment and configuration of the software system is presented with example applications.
The Benefits of Cementitious Encapsulation Matrices for the Conditioning of Intermediate Level Waste
Authors: Guppy R. M. , Vines S. P. , Wisbey S. J.
Published: March 2009

The UK has significant quantities of radioactive waste, which have arisen over the past fifty years or so, largely as a result of nuclear power, reprocessing and defence programmes. The intermediate level wastes arising as a result of these activities, exhibit a high level of physical and chemical diversity, and must be managed safely in a way that protects existing and future generations and the environment. Development work has been conducted since the early 1980s to identify suitable conditioning materials and techniques that are compatible with the needs of safe long-term management, including interim storage, transport and future deep geological disposal. From these studies cementation emerged as the one medium which could satisfy all the key waste management criteria. Other materials were not ruled out and may offer benefits in specific applications. The advantages of conditioning ILW with cement include: • the extensive experience of its use in a wide variety of contexts; • the raw materials are relatively cheap and have a long shelf life; • cement is processed in relatively simple plant at room temperature, with safety and cost benefits for plant operators; • the product is fire resistant and of relatively low toxicity; • cement is capable of immobilising a wide range of wastes ranging from solids to aqueous slurries; • cement provides desirable product properties. Desirable properties include: • suitable strength, • chemical control of radionuclide leading to enhanced retention, • good corrosion protection for steels, • low permeability, • tolerance to radiation, • durability over extended timescales, and • good radiation self-shielding properties. Several waste packaging plants are now operational in the UK using cement-based encapsulants. These are currently conditioning ILW for interim storage, in a manner suitable for future transport and compatible with the Nirex phased deep disposal concept. This paper will describe the development of cement-based encapsulants to meet the needs of UK radioactive wastes, and will provide examples of the supporting product quality data.
The Potential Impact of Oil and Other Non-Aqueous Phase Liquids (NAPLs) on the Long-Term Management of Radioactive Wastes
Authors: Askarieh M. M. , Harris A. W. , Wisbey S. J.
Published: March 2009

The presence of non-aqueous phase liquids (NAPLs) in radioactive wastes has the potential to adversely affect their long-term management. The buoyancy of light NAPLs can represent a separate pathway for their migration from a waste management facility, such as a deep repository, to the accessible environment. Their inherent toxicity and potential burden of radionuclides need to be addressed. Nirex is currently developing an understanding of the behaviour of NAPLs and exploring the means of mitigating any adverse impact. NAPLs such as oils and solvents are present in existing wastes, but NAPLs can also be generated by degradation of some solid organic materials, such as plastics. Wastes arising in the United Kingdom that may contain NAPLs include liquids and sludges contaminated with oils, and waste items containing trapped oil, for example gearboxes and pumps. The reference inventory being assessed by Nirex also contains significant quantities of organic materials which can be considered to be potential precursors to the generation of NAPLs. A programme of work has been instigated by Nirex to develop a better understanding of the behaviour of NAPLs. The programme includes the following aspects: understanding the mechanisms by which NAPLs might be produced and existing NAPLs degraded and destroyed: • consideration of the containment that could be offered by packaging of wastes containing NAPLs; • investigating the extent to which radionuclides may be entrained in NAPLs; • understanding the migration of NAPLs in the near-field and in geological systems; • the impact of NAPLs on the surface properties of repository backfill and the geosphere; • development of assessment tools to quantify the potential risk due to NAPLs. This paper will describe the scope of this programme of work, and will provide examples from the ongoing programme to demonstrate that suitable long-term waste management solutions can be developed for NAPL containing wastes.
A Radio Toxicity Index for Fusion Waste
Authors: Di Pace Luigi , Natalizio Antonio
Published: March 2009

With current designs of future fusion power plants, it is evident that a significant quantity of operational radioactive waste will be produced over the lifetime of the plant. This waste will be mostly due to the replacement of in-vessel components (IVCs) on a regular basis, currently assumed to be every five years. This potentially large quantity of waste raises issues about its ultimate disposal, particularly the nature of the disposal facility required to accommodate it. The term invessel component includes the divertor and the breeding blanket, the “fuel” in a fusion reactor. In this perspective only, the waste resulting from IVC replacement is analogous to the fuel waste arising from fission power plants, and this comparison, whether justified, or not, could prejudice the fusion waste disposal solution. As fusion in-vessel component waste is significantly and fundamentally different from fission reactor fuel waste, it is essential that the fusion disposal solution be based solely on its needs. To highlight this fundamental difference between the fusion and fission operational waste, a radio toxicity index has been defined, which may prove to be of value in defining appropriate requirements for the disposal of fusion operational waste. Uranium has been the basis of the fission power industry and it is found in nature in concentrations varying, typically, from 0.1 to 1%, and in some cases ore bodies with concentrations up to 25% have been found. Because uranium is a radioactive element, and is quite common in the earth’s crust, it offers an opportunity to be used as a benchmark for comparing potential fusion and fission power reactor radioactive waste. As U-238 is the most abundant isotope of uranium found in nature (>99%), it is proposed that the radio toxicity of U-238 be used to assess the relative radio toxicity of relevant fusion and fission waste. The ratio of the radio toxicity of a given material to that of U-238 is referred to as the radio toxicity index. Therefore, a waste material with a radio toxicity index equal to one would be considered acceptable for disposal in the earth’s crust, in the same manner that uranium tailings are disposed of in the mining industry. The results of studies performed for typical fusion breeder material indicate that there is no compelling economic reason for reprocessing. Furthermore, the radio toxicity index for such materials indicates that there are no technical reasons — i.e., there does not appear to be a need for deep, geological disposal of spent fusion breeder material. On the contrary, the application of the radio toxicity index to spent fission fuel has demonstrated, from a waste disposal perspective, that there are compelling reasons for reprocessing to separate low radio toxicity fission products from the high radio toxicity actinides, which can be reused. This conclusion augurs well for a future fusion power industry and goes a significant distance in demonstrating the potential environmental advantages of fusion power.
Decommissioning Liabilities and Cost Considerations
Authors: Cloutier William A.
Published: March 2009

The premature cessation of operations at several large commercial reactors in the United States has created for the owners of these facilities an accelerated liability for decommissioning. A majority of the owners of these facilities, however, still elected to proceed with immediate dismantling, even though, in many instances, the adequacy of the available funding had not been established. With limited financial resources, project success often depends upon the ability of the owner to address and resolve project encumbrances, regulatory constraints and the growth of the work scope in an expeditious and cost-effective manner. The common lesson-learned recognized in the performance of these major projects over the last 15 years, is that planning for decommissioning needs to be started earlier and include a comprehensive cost analysis so as to facilitate post-shutdown activities. This paper will summarize the processes used to identify and quantify decommissioning liabilities in the US, as well as in other countries. In particular, the objectives in developing a cost estimate will be explored, the types of estimates discussed, and the need to integrate the estimate within the ongoing planning for decommissioning. Strategic cost considerations will be identified, as well as their impact on the financial resources required. Case studies will be presented, identifying both similarities and differences in purpose and in scope. The paper will discuss the key planning tools, for example, facility characterization assessments for radiological, hazardous and toxic contaminates. Program management is the single largest expense incurred in plant decommissioning since it is a highly regulated and controlled process. However, in most instances, the oversight of decommissioning operations requires only a fraction of the original operating organization. This paper will explore the need to streamline and transition the operating staff to one that can effectively support decommissioning activities while minimizing the overall expense. Waste conditioning and disposal is a major technical, as well as financial element, in the facility decommissioning. The availability and cost of regional and national disposal facilities is a key consideration in the options selected for decommissioning, including timing, approach and methods selected. The formation and integration of a waste management strategy will be discussed along with the sensitivity of the decommissioning cost and schedule to the strategy selected. The paper will conclude with several observations relating to the need to include financial planning in any decommissioning evaluation, a discussion of lessons-learned from ongoing decontamination and dismantling projects, and common misconceptions. Recommendations will be offered for owners of those facilities currently considering decommissioning, as well as those in the early planning stage.
Strategic Environmental Assessment and Environmental Impact Assessment as Part of a Stepwise Decision Making Process
Authors: Atherton Elizabeth , McCall Ann
Published: March 2009

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. One of the lessons that Nirex has learned from previous experience in the UK and internationally is the importance of developing due process for finding a long-term solution for radioactive waste management. We have been investigating best practice in this area and incorporating the findings into the work that we undertake. Projects which will have an impact on the environment are subject to EC Directives on Strategic Environmental Assessment (SEA) and Environmental Impact Assessment (EIA). The EIA Directive has already been implemented into law within Member States while the SEA Directive has to be implemented by 2004. Nirex believes that radioactive waste management programmes will have to adhere to the principles outlined in the Directives. Nirex has been investigating how the frameworks set out in the Directives could be used to: • Develop a stepwise approach to decision making in the UK; • Engage stakeholders during the stepwise decision making; • Enable stakeholders’ issues and concerns to be addressed. This paper will outline how Nirex has been developing its work in these areas including reference to the Nirex Involvement Programme, which uses different consultation and dialogue techniques to enable people to engage with Nirex’s work programme.
Vitrification of DOE Simulated Radioactive Waste by Induction-Heated Cold-Crucible Melter Technology
Authors: Aloy A. S. , Soshnikov R. A. , Lopukh D. B. , Bickford D. F. , Herman C. C. , Holtzsheiter E. W. , Goles R. W.
Published: March 2009

Certain waste streams of the US DOE contain radioactive refractory oxides and other components like aluminum zirconium and chromium, which present difficulties during their processing and immobilization. The vitrification of such waste in joule-heated melters at high waste loading is possible only at a temperature exceeding 1150°C. The Khlopin Radium Institute (St.-Petersburg, Russia) jointly with the US Department of Energy has performed a feasibility study on the suitability of the Cold-Crucible Induction Heated Melter (CCIM) technology for the single-stage solidification of a surrogate sludge (C-106/AY-102 HLW Simulant), similar in composition to the High Level Waste (HLW) found at DOE’s Hanford Site (Richland, USA). During the experiments, slurry of simulated sludge and glass formers was metered directly to the CCIM, melted, and the glass product was poured from the melter. The melts were conducted at a mean melt temperature of 1350°C. The experiments produced borosilicate glass wasteforms with a waste oxide loading of 70 weight percent. According to the X-Ray diffraction analysis, the final product had a glass-crystalline structure. The crystalline phase was represented by spinel, (Fe,Mn)Fe2O4, uniformly distributed over the wasteform. The chemical durability of the samples was tested by the Product Consistency Test (PCT), and was considered durable according to the DOE specifications for HLW. In the course of the experiments, data were accumulated on the specific electric power consumption and the throughput of the facility.
Practical Considerations of Site Clearance
Authors: LaGuardia Thomas S.
Published: March 2009

The US Nuclear Regulatory Commission (NRC) established criteria for acceptable residual radioactivity related to decommissioning nuclear power plants in the US [1]. A level of 25 mRem per year to the maximum exposed individual by site-specific pathways analysis, with ALARA is acceptable to the NRC. Systems and structures containing very low levels of radioactivity that meet this criteria are deemed acceptable to abandon in place as part of the decommissioning process and termination of the license. Upon license termination by the NRC, the owner may then demolish and remove remaining structures. In practice, site-specific criteria imposed by local state mandates, company management decisions, real estate value, and long-term liability potential have driven nuclear plant licensees to adopt an alternative disposition for these materials. Although the reasons are different at each site, the NRC’s criteria of 25 mRem per year are not the controlling factor. This paper will describe the regulatory process for termination of the license, and the other factors that drive the decision to remove radioactive and non-radioactive material for decommissioning. Several case histories are presented to illustrate that the NRC’s criteria for license termination are not the only consideration.
The Risk to Health From Exposure to Low Doses of Ionising Radiation
Authors: Wakeford Richard , Tawn E. Janet
Published: March 2009

Controversy continues over the shape of the dose-response curve describing the risk of stochastic health effects (cancer and hereditary disorders) following exposure to low doses of ionising radiation. Radiological protection is currently based upon the assumption that the dose-response curve has no threshold and is linear in the low dose region. This position is challenged by groups suggesting either that this approach seriously underestimates the true risk at low doses or that low-level exposure results in no risk (a threshold dose exists) or even a beneficial effect (“radiation hormesis”). In this paper, we examine the epidemiological and radiobiological bases of the linear no-threshold model and some of the alternatives that have been proposed. We conclude that the evidence for a material deviation from a linear no-threshold dose-response relationship at low doses is not persuasive and that the standard model provides the most parsimonious description of the available scientific evidence.
The Difficulties of Undertaking Site Characterisation on Operational Nuclear Licensed Sites
Authors: Booth Peter , Gordon Robert
Published: March 2009

Undertaking site investigation on nuclear licensed sites is very different to site investigation on other sites, even those which are heavily contaminated. This paper sets out to highlight some of the issues which need to be considered when formulating a fit for purpose, yet defensible site characterisation programme on such a site. The first and most important aspect of the work is to set out clearly your objectives. There may be a number of reasons why a site investigation is being undertaken. These could include purely fulfilling your site licence conditions as an operator or they could be more specific like supporting a defined de-licensing or decommissioning project, installing a monitoring network, or determining the extent of ground contamination. Ensuring that a conceptual model exists is the next step, even if only at a preliminary stage, as this coupled with the desk study will help formulate the site characterisation programme. Logistical issues as well as technical requirements need to be factored in, but in order to maintain transparency it is important to declare the latter first. Like other sites with ground contamination, issues like sampling and analysis need to be considered. Clearance procedures on nuclear licensed sites are extremely stringent and can lead to delays. These need to be considered, especially if sending samples offsite for organics analysis. The laboratories themselves need to be licensed to handle radioactive samples and the transport regulations also need to be adhered to. Other logistical issues requiring consideration include safety cases, plant modification proposals and waste disposal. The technical side itself sets its own challenges in that decisions need to be closely linked into the logistics. Will the samples and data be collected primarily through intrusive techniques or is there a requirement to utilise non-intrusive methodology? How do you defend the proposed site sampling strategy when you have access restrictions? Do you need to have permanent monitoring facilities? These are just some of the questions which need to be answered if a site operator is to have a transparent and defensive site investigation programme on a nuclear licensed site.
Rationale for Independent Site Clean-Up and Radiological Clearance Standards
Authors: Genoa Paul H.
Published: March 2009

Over the past few years, the U.S. nuclear power industry has gained substantial experience and appreciation of the technical complexity and rigor required to meet a performance-based site clean-up standard. Five large power reactors and several smaller ones are now well along the path to license termination. They have not been on this journey alone. There has been a steep learning curve for all stakeholders involved in the process including state and federal radiation regulators, legislators, and the public. We have all learned that the translation of results from a post remediation survey interpreted through pathway modeling for comparison with a dose-based clean-up standard is for many a leap of faith. Our regulator has an understandable desire to address this uncertainty by demanding conservative analysis at each turn. As a result, it is extremely demanding to demonstrate that a clean-up standard in the 0.15–0.25 mSv/a range has been met. It is not likely that a standard in the 10 μSv/a level, typically associated with radiological clearance standards, can be practically demonstrated while still meeting the current expectations of U.S. Nuclear Regulatory Commission for technical rigor.
Discharge Reductions: Value for Money?
Authors: Morley R. G. , Waring J. , Coates R. , Taylor R. H.
Published: March 2009

The paper uses the historical record of radioactive discharges from BNFL’s Sellafield reprocessing site in the UK and seeks to identify what have been the key drivers for change, particularly over the past 20 years of significant discharge reductions. The paper examines the current context for ongoing and future discharges from the site, including intergovernmental commitments such as the OSPAR Sintra statement and the developing UK policy framework, together with BNFL’s work with a wide range of ‘green’ stakeholders. The paper outlines the principal components of BNFL’s decision-making processes for discharge control and abatement, and how these interact with the relevant external pressures. It then analyses whether the overall drivers and outcomes align with the declared desire of the UK government to ensure that the taxpayer receives value for money in the new national arrangements for managing legacy wastes from the nuclear industry.
DFR Decommissioning: The Breeder Fuel Removal Project — Retrieval and Processing Facilities
Authors: Bonnet C. , Potier P. , Ashton B. Morris
Published: March 2009

The Dounreay site, in North Scotland, was opened in 1955 and a wide range of nuclear facilities have been built and operated there by UKAEA (The United Kingdom Atomic Energy Authority) for the development of atomic energy research. The Dounreay Fast Reactor (DFR) was built between 1955 and 1957, and operated until 1977 for demonstration purposes and for producing electricity. Today, its decommissioning is a key part of the whole Dounreay Site Restoration Plan that integrates the major decommissioning activities such as the fuel treatment and the waste management. The paper presents the contract strategy and provides an overview of the BFR project which consists in the removal of the breeder elements from the reactor and their further treatment. It mainly provides particular details of the Retrieval and Processing Facilities design.
The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World’s First Radioactive Fusion Research Reactor for Burial
Authors: Rule Keith , Perry Erik , Chrzanowski Jim , Viola Mike , Strykowsky Ron
Published: March 2009

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor (Fig. 1) will approach 2 Million Kilograms with an associated volume of 2500 cubic meters. The materials were activated by 14 Mev neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project.