bluuebunny/crossref_metadata_2025 huggingface dataset:
Implications of Patent Law Changes for Medical Device Manufacturers
Authors: Stone Gregory M.
Published: June 2009

2007 saw significant changes in patent law, and 2008 is promising additional significant developments. Those changes have significant effects on the value of patents and on how medical device developers and manufacturers should manage their patent strategy. The paper and discussion will detail those changes and provide practical guidance on what steps a medical device company can take to properly address those changes.
Tools for Nanosurgery: Applying Carbon Nanopipettes to Cell Physiology
Authors: Schrlau Michael G. , Bau Haim H.
Published: June 2009

There is a need in medical and biological research to probe cells with high spatial resolution. We have developed a method for fabricating integrated carbon nanopipes with diameters ranging from 10’s to 100’s of nanometers to form carbon nanopipettes (CNPs) without the need for cumbersome assembly. CNPs are easily adapted to standard cell physiology equipment and simultaneously provide an accessible nanoscale channel for mass transport and a conductive pathway for electrical measurements. We present the fabrication method, the characterization of the device, and their ability to inject fluids into cells, specifically, calcium mobilizing second messengers.
Experimental Study and CFD Analysis of the Volumatic® Spacer
Authors: Abreu Sofia , Silva Lui´s , Teixeira Senhorinha F. C. F. , Antunes Henedina , Marques Helena , Ramalhete Nuno
Published: June 2009

The aim of the present study was to evaluate the performance of Ventilan® coupled to the Volumatic® spacer. An experimental study was conducted using the four stage Multistage Liquid Impinger (MSLI). This enabled the observation that half of the pharmacological dose is retained within the body of the spacer. Therefore, the air flow inside the spacer was studied, along the respiratory cycle, using Fluent™. Several recirculation regions were observed inside the spacer that may influence the aerosol flow and also the deposition of the drug.
Wireless Central Apnea Response System for Neonatal Intensive Care
Authors: Hill Jonathan M. , Adrezin Ronald S. , Eisenfeld Leonard
Published: June 2009

An apnea event is defined as the cessation of breathing for 10 to 20 seconds, generally accompanied by bradycardia (decrease in heart rate), cyanosis, or both. Unlike apnea in adults, which is usually caused by an obstruction in the airway, central apnea events appear in premature babies because the autonomic nervous system is not yet fully developed.
Curing Time Effect on the Fraction of 137Cs From Cement-Ion Exchange Resins-Bentonite Clay Composition
Authors: Plecas Ilija , Dimovic Slavko
Published: June 2009

To assess the safety of disposal of radioactive waste material in cement, curing conditions and time of leaching radionuclides 137Cs have been studied. Leaching tests in cement-ion exchange resins-bentonite matrix, were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Incremental leaching rates Rn (cm/d) of 137Cs from cement-ion exchange resins-bentonite matrix after 180 days were measured. The results presented in this paper are examples of results obtained in a 20-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.
Bioremediation of Organic Pollutants in a Radioactive Wastewater
Authors: Oboirien Bilainu O. , Molokwane P. E. , Chirwa Evans M. N.
Published: June 2009

Bioremediation holds the promise as a cost effective treatment technology for a wide variety of hazardous pollutants. In this study, the biodegradation of organic compounds discharged together with radioactive wastes is investigated. Nuclear process wastewater was simulated by a mixture of phenol and strontium, which is a major radionuclide found in radioactive wastewater. Phenol was used in the study as a model compound due to its simplicity of molecular structure. Moreover, the biodegradation pathway of phenol is well known. Biodegradation studies were conducted using pure cultures of Pseudomonas aeruginosa and Pseudomonas putida. The rate of phenol degradation by both species was found to be higher in the test without strontium. This suggests some degree of inhibition in the degradation of phenol by strontium. There was no phenol degradation in the sterile controls. The results indicate the feasibility of the biodegradation of organic pollutants discharged in radioactive effluents by specialised microbial cultures.
Contaminated Land and Groundwater Management at Sellafield: A Large Operational Site With Significant Legacy and Contaminated Land Challenges
Authors: Reeve Phil , Eilbeck Katherine
Published: June 2009

Sellafield is a former Royal Ordnance Factory used since the 1940’s for the production and reprocessing of fissile materials. Leaks and spills from these plants and their associated waste facilities has led to radioactive contaminated ground legacy of up to 20 million m3. Consideration of land contamination at Sellafield began in 1976, following discovery of a major leak from a waste storage silo. Over the past three decades there has been a programme of environmental monitoring and several phases of characterization. The latest phase of characterization is a £10million contract to develop second generation conceptual and numeric models. The Site Licence Company that operates the site has been subject to structural changes due to reorganizations within the British nuclear industry. There has also been a change in emphasis to place an increased importance on accelerated decommissioning. To address these challenges a new contaminated land team and contaminated land and groundwater management plan have been established. Setting and measuring performance against challenging objectives is important. The management plan has to be cognizant of the long timescales (ca. 80 years) for final remediation. Data review, collation, acquisition, analysis, and storage is critical for success. It is equally important to seize opportunities for early environmental gains. It is possible to accelerate the development and delivery of a contaminated land and groundwater management plan by using international experts.
Environmental Assessment of Decommissioning Radioisotope Thermoelectric Generators (RTG) in Northwest Russia
Authors: Hosseini A. , Standring W. J. F. , Brown J. E. , Dowdall M. , Amundsen I. B.
Published: June 2009

This article presents some results from assessment work conducted as part of a joint Norwegian-Russian project to decommission radioisotope thermoelectric generators (RTG) in Northwest Russia. Potential worst case accident scenarios, based on the decommissioning procedures for RTGs, were assessed to study possible radiation effects to the environment. Close contact with exposed RTG sources will result in detrimental health effects. However, doses to marine biota from ingestion of radioactivity under the worst-case marine scenario studied were lower than threshold limits given in IAEA literature.
Current Status of the Japanese Geological Repository Programme
Authors: Kitayama Kazumi
Published: June 2009

The programme for disposal of radioactive waste in Japan is now moving ahead on a number of fronts. On the regulatory side, responsibility for TRU waste disposal has been assigned to NUMO and guidelines for the safety goals for disposal of LLW have been published. NUMO, as the implementer for the deep geological disposal programme, has been developing the special tools for project management that are needed as a result of the decision to adopt a volunteering approach to siting. NUMO is also building up the technical infrastructure for flexible tailoring of site characterisation, repository design and the associated safety assessment to the conditions found in any volunteer site. This work requires openness and transparency in decision-making but, as several sites may need to be investigated in parallel, particular emphasis is placed on operational practicality.
Removal: An Alternative to Clearance
Authors: Feinhals J. , Kelch A. , Kunze V.
Published: June 2009

This presentation shows the differences between the application of clearance and removal, both being procedures for materials leaving radiation protection areas permanently. The differentiation will be done on the basis of the German legislation but may be also applicable for other national legislation. For clearance in Germany two basic requirements must be given, i.e. that the materials are activated or contaminated and that they result from the licensed use or can be assigned to the scope of the license. Clearance needs not to be applied to objects in Germany which are to be removed only temporarily from controlled areas with the purpose of repair or reuse in other controlled areas. In these cases only the requirements of contamination control apply. In the case of removal it must either be proved by measurements that the relevant materials are neither activated nor contaminated or that the materials result from areas where activation or contamination is impossible due to the operational history considering operational procedures and events. If the material is considered neither activated nor contaminated there is no need for a clearance procedure. Therefore, these materials can be removed from radiation protection areas and the removal is in the responsibility of the licencee. Nevertheless, the removal procedure and the measuring techniques to be applied for the different types of materials need an agreement from the competent authority. In Germany a maximum value of 10% of the clearance values has been established in different licenses as a criterion for the application of removal. As approximately 2/3 of the total mass of a nuclear power plant is not expected to be contaminated or activated there is a need for such a procedure of removal for this non contaminated material without any regulatory control especially in the case of decommissioning. A remarkable example is NPP Stade where in the last three years more than 8600 Mg were disposed of by removal and only 315 Mg were released by clearance, even before the decommissioning licensing procedure was finished.
The Carbon Matrices Made of Pyrolised Phtalocyanines as a Base for Encapsulation of the Long-Lived Nuclides of Iodine, Technetium and Minor Actinides
Authors: Tikhonov V. I. , Moskalev P. N. , Kapustin V. K.
Published: June 2009

The creation and careful investigation of suitable materials and forms for transmutation of the long-lived radioactive waste (RW) is mainly in the starting stage. A new carbon material formed as a result of pyrolisis of bisphtalocyanine, Pc2Me, gives a chance to solve this goal successfully. The pyrolysis takes place under an argon (Ar) atmosphere at temperature of 700 – 800°C. The release of atoms encapsulated inside this carbon matrix occurs only at temperatures above 1200°C, and a correlation between the efficiency of the atoms’ release and their atomic radius has been revealed. It is caused with creation of closed microcavities in the carbon skeleton during pyrolysis of MeC2. Due to inert features and high thermostability of carbon, an inculcation of the long-lived radionuclides in these microcavities by means of their phtalocyanines pyrolysis gives unique opportunities for both their transmutation and storage. The first results on encapsulation within matrixes of radionuclides of europium (Eu), technetium (Tc), iodine (I) and “minor actinides” are presented. The efficiency of encapsulation is close to 100% for all studied elements excluding iodine, for the last one, it is near 85–90%. The results on thermochemical stability, leaching and other tests of these matrixes are presented.
Planning for the Recreational End Use of a Future LLR Waste Mound in Canada: Leaving an Honourable Legacy
Authors: Kleb H. R. , Zelmer R. L.
Published: June 2009

The Low-Level Radioactive Waste Management Office was established in 1982 to carry out the federal government’s responsibilities for low-level radioactive (LLR) waste management in Canada. In this capacity, the Office operates programs to characterize, delineate, decontaminate and consolidate historic LLR waste for interim and long-term storage. The Office is currently the proponent of the Port Hope Area Initiative; a program directed at the development and implementation of a safe, local long-term management solution for historic LLR waste in the Port Hope area. A legal agreement between the Government of Canada and the host community provides the framework for the implementation of the Port Hope Project. Specifically, the agreement requires that the surface of the long-term LLR waste management facility be “conducive to passive and active recreational uses such as soccer fields and baseball diamonds.” However, there are currently no examples of licensed LLR waste management facilities in Canada that permit recreational use. Such an end use presents challenges with respect to engineering and design, health and safety and landscape planning. This paper presents the cover system design, the environmental effects assessment and the landscape planning processes that were undertaken in support of the recreational end use of the Port Hope long-term LLR waste management facility.
Remediating While Preserving Wetland Habitat at an LLR Waste Site in Canada
Authors: Kleb H. R. , Zelmer R. L.
Published: June 2009

The Low-Level Radioactive Waste Management Office was established in 1982 to carry out the federal government’s responsibilities for low-level radioactive (LLR) waste management in Canada. The Office operates programs to characterize, delineate, decontaminate and consolidate historic LLR waste for interim and long-term storage. In this capacity, the Office is currently considering the remediation of 9,000 cubic metres of contaminated sediment in a coastal marsh in the context of a major remediation project involving multiple urban sites. The marsh is situated between the Lake Ontario shoreline and the urban fringe of the Town of Port Hope. The marsh is designated a Cattail Mineral Shallow Marsh under the Ecological Land Classification system for Southern Ontario and was recently named the A.K. Sculthorpe Marsh in memory of a local community member. The marsh remediation will therefore require trade offs between the disruption of a sensitive wetland and the removal of contaminated sediment. This paper discusses the issues and trade-offs relating to the waste characterization, environmental assessment and regulatory findings and thus the remediation objectives for the marsh. Considerations include the spatial distribution of contaminated sediment, the bioavailability of contaminants, the current condition of the wetland and the predicted effects of remediation. Also considered is the significance of the wetland from provincial and municipal regulatory perspectives and the resulting directives for marsh remediation.
The UK Government’s Global Partnership Programme: Its Achievements Over the Past Five Years and Challenges Ahead
Authors: Heyes Alan
Published: June 2009

Through the Global Partnership the UK continues to make a significant contribution to improve national and global security. Over the past year the UK has continued to implement a wide range of projects across the breadth of its Global Partnership Programme. As well as ensuring the Programme is robust and capable of dealing with new challenges, the UK has cooperated with other donor countries to help them progress projects associated with submarine dismantling, scientist redirection, enhancing nuclear security and Chemical Weapons Destruction. The Global Partnership, although only five years old, has already achieved a great deal. Some 23 states, plus the European Union, are now working closer together under the Global Partnership, and collectively have enhanced global regional and national security by reducing the availability of Weapons of Mass Destruction (WMD) materials and expertise to both states of concern and terrorists. Considerable progress has already been made in, for example: • Improving the security of fissile materials, dangerous biological agents and chemical weapons stocks; • Reducing the number of sites containing radioactive materials; • Working towards closure of reactors still producing weapon-grade plutonium; • Improving nuclear safety to reduce the risks of further, Chernobyl style accidents; • Constructing facilities for destroying Chemical Weapons stocks, and starting actual destruction; • Providing sustainable employment for former WMD scientists to reduce the risk that their expertise will be misused by states or terrorists. By contributing to many of these activities, the UK has helped to make the world safer. This paper reports on the UK’s practical and sustainable contribution to the Global Partnership and identifies a number of challenges that remain if it is to have a wider impact on reducing the threats from WMD material.
Operational Programs for National Radioactive Waste and Spent Fuel Management Programme in Slovenia
Authors: Zˇeleznik Nadja , Kralj Metka , Mele Irena
Published: June 2009

The first separate National Radioactive Waste and Spent Fuel Management Programme (National Programme) was prepared in Slovenia in 2005 as a supplementary part of the National Environmental Action Programme and was adopted in February 2006 by the Slovenian Parliament. The new National Programme includes all topics being relevant for the management of the radioactive waste and spent fuel which are produced in Slovenia, from the legislation and identification of different waste streams, to the management of radioactive waste and spent fuel, the decommissioning of nuclear facilities and management of (TE)NORM in the near future from 2006 up to the 2015. The National Programme identified the existing and possible future problems and proposed the technical solutions and action plans for two distinctive periods: 2006–2009 and 2010–2015. According to the requirement of Act on Protection against Ionising Radiation and Nuclear Safety the national Agency for Radwaste Management (ARAO) prepared the operational programmes for the four year period with technical details on implementation of the National programme. ARAO gained the detailed plans of different involved holders and proposed 9 operational programmes with aims, measures, individual organizations in charge, expenses and resources for each of the programmes. The Operational programmes were already reviewed by the Ministry of Environment and Physical Planning and are under acceptance. The orientation of the radioactive waste management according to the National Programme and operational activities within additional limitations based on the strategical decisions of Slovenian Government is presented in the paper.
Switching From Deferred Dismantling to Immediate Dismantling: The Example of Chooz A, A French PWR
Authors: Grenouillet Jean-Jacques
Published: June 2009

Located in the north of France, close to Belgian border, Chooz A is the first PWR that was built in France from 1962 to 1967. When it was shutdown in 1991, a deferred dismantling strategy was selected. Further to an evolution of EDF decommissioning strategy in 2001, the decommisioning of the plant was accelerated by reducing the safe enclosure period to only a few years. Thus Chooz A will be the first PWR to be fully dismantled in France and it gives a good insight of what is needed to reactivate a plant for final dismantling after a safe enclosure period.
Evaluation of Fourier Transform Profilometry Performance: Quantitative Waste Volume Determination Under Simulated Hanford Waste Tank Conditions
Authors: Jang Ping-Rey , Leone Teresa , Long Zhiling , Mott Melissa A. , Norton O. Perry , Okhuysen Walter P. , Monts David L.
Published: June 2009

The Hanford Site is currently in the process of an extensive effort to empty and close its radioactive single-shell and double-shell waste storage tanks. Before this can be accomplished, it is necessary to know how much residual material is left in a given waste tank and the chemical makeup of the residue. The objective of Mississippi State University’s Institute for Clean Energy Technology’s (ICET) efforts is to develop, fabricate, and deploy inspection tools for the Hanford waste tanks that will (1) be remotely operable; (2) provide quantitative information on the amount of wastes remaining; and (3) provide information on the spatial distribution of chemical and radioactive species of interest. A collaborative arrangement has been established with the Hanford Site to develop probe-based inspection systems for deployment in the waste tanks. ICET is currently developing an in-tank inspection system based on Fourier Transform Profilometry, FTP. FTP is a non-contact, 3-D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, FTP is capable of determining the height (depth) distribution (and hence volume distribution) of the target surface, thus reproducing the profile of the target accurately under a wide variety of conditions. Hence FTP has the potential to be utilized for quantitative determination of residual wastes within Hanford waste tanks. We have completed a preliminary performance evaluation of FTP in order to document the accuracy, precision, and operator dependence (minimal) of FTP under conditions similar to those that can be expected to pertain within Hanford waste tanks. Based on a Hanford C-200 series tank with camera access through a riser with significant offset relative to the centerline, we devised a testing methodology that encompassed a range of obstacles likely to be encountered “in-tank.” These test objects were inspected by use of FTP and the volume of the test objects determined. The volumes of nondescript test objects were independently determined and were not known to the FTP operators. Several stages of testing are ongoing with successive stages imposing aspects that present increasing difficulty and increasingly more accurate approximations of in-tank environments. We report the Stage 1 results of this multi-stage evaluation of FTP performance.
The Korean Final Repository for Low- and Intermediate-Level Radioactive Waste
Authors: Kim Young-Ki , Koh Kwang-Hoon , Lee Sang-Sun , Choi Gi-Won , Lee Byung-Sik , Kwon Heon-Woo , Lee Hyeong-Nam , Yoo Jun-Sang , Park Bok-Ok
Published: June 2009

The underground waste repository site is located at Gyeongju and is selected for the disposal of all the Low- and Intermediate-Level Radioactive Waste (LILW). Its operation is scheduled in the beginning of 2009. The repository, with a disposal capacity of 800,000 drums, will be constructed in granite bedrock near the seashore at the Gyeongju site. The repository will be constructed in phases to reach its final capacity of 800,000 drums. In the first phase of construction, the repository will have a capacity to dispose of 100,000 drums. The first phase of the repository design consists of an access shaft, a construction tunnel, an operating tunnel, an unloading tunnel, and six (6) silos. The silos are located at 80 to 130 meters below Mean Sea Level (MSL), in bedrock. Each silo is 24.8m in diameter and 52.4m in height. The silo will be reinforced with shotcrete, rockbolts and concrete lining for rock supports, and the lining will also act as an engineered barrier to limit radioactive nuclide release after closure. After serving its intended function the repository will be backfilled and sealed. The primary objective of backfilling and sealing is to prevent ground-water flow into the silos through the tunnel system and to prevent inadvertent intrusion into the repository after closure.
Possible Approach to Cleaning “Problematic” LRW With Large Contents of Suspended Particles, Oils and Other Organic Substances
Authors: Ilin V. , Karlin Yu. , Laurson A. , Volkov Eu. , Dmitriev S.
Published: June 2009

A general structural scheme for cleaning “problematic” liquid radioactive wastes (LRW) containing a large amount of suspended particles, oils and other organic substances has been proposed. The technological scheme includes two main stages: 1) separation of suspended particles, oil product emulsions and the larger part of colloidal particles from LRW by filtration, 2) purification of radioactive waters from radionuclides by membrane-sorption to the levels of radiation safety norms applied. The filtration stage is considered as a three-step process of “problematic” LRW treatment including: 1) “problematic” LRW extraction from storage tanks with a robot type device intended for washing out the bottom sediment (slurry), 2) separation of suspended particles, oil product emulsions and larger part of colloidal particles from LRW by filtration through porous or gauze diaphragms of 0.1 to 10 μm pores (cells) in size, 3) concentration of separated slurry up to 100–200 g/l. Two main options of the membrane-sorption technologies, AQUA-EXPRESS and Reverse Osmosis, for LRW purification have been considered. Two possible options of porous or gauze diaphragms productivity and lifetime increase between their surface regenerations have been shown: 1) possibility of an oxidizer introduction into initial LRW, 2) possibility to rotate a filtering element (disk or cylinder type).
Windscale Pile Reactors: Decommissioning Progress on a Fifty Year Legacy
Authors: Sexton Richard J.
Published: June 2009

The decommissioning of the Windscale Pile 1 reactor, fifty years after the 1957 fire, is one of the most technically challenging decommissioning projects in the UK, if not the world. This paper presents a summary of the 1957 Windscale Pile 1 accident, its unique challenges and a new technical approach developed to safely and efficiently decommission the two Windscale Pile Reactors. The reactors will be decommissioned using a top down approach that employs an array of light weight, carbon fiber, high payload robotic arms to remove the damaged fuel, the graphite core, activated metals and concrete. This relatively conventional decommissioning approach has been made possible by a recently completed technical assessment of reactor core fire and criticality risk which concluded that these types of events are not credible if relatively simple controls are applied. This paper presents an overview of the design, manufacture and testing of equipment to remove the estimated 15 tons of fire damaged fuel and isotopes from the Pile 1 reactor. The paper also discusses recently conducted characterization activities which have allowed for a refined waste estimate and conditioning strategy. These data and an innovative approach have resulted in a significant reduction in the estimated project cost and schedule.
Analysis on Uranium Isotope in the Facilities of Nuclear Fuel Materials Using Depleted Uranium
Authors: Jeon Jong Seon , Jung Ki Chul , Park Sang Gyu , Kim Tae Hyun , Lee Jae Min
Published: June 2009

This study checked the degree of contamination of depleted uranium used as a chemical catalyst in the manufacturing process within the facilities of nuclear fuel materials to analyze the environmental sample for abandoning their industrial factory sites and investigated how many times of contamination were made compared to (natural) abundance of isotopes if contamination was made within the facilities. In order to analyze the degree of uranium contamination, the researcher of this study divided the upper and lower parts of 20 points from the surface of the earth within the factory site made of concrete and extracted 40 samples from the surface of the earth and 15 samples for checking air and surface water contamination. The study checked the concentration of uranium existing in small quantity in the samples by liquefying a large amount of samples using pre-treatable acid percolation.
Use of a Shielded Low Resolution Gamma Spectrometer for Segregation of Free Release and Low Level Waste
Authors: Wilkins C. G. , Alvarez E. , Cocks J. , Davison L. , Mattinson A.
Published: June 2009

In the UK, low level radioactive waste (LLW) is sent to the national Low Level Waste Repository (LLWR) at Drigg in Cumbria. Strict rules limit the specific activity of waste that is sent to the LLW Repository and waste producers and consignors have to demonstrate that the waste they send to the repository meets its conditions for acceptance. However, the limited capacity of the Low Level Waste Repository means that it is just as important for waste consigners to ensure that inactive ‘free release’ or ‘exempt’ waste is not inadvertently sent to the repository. Incorrect segregation of waste in a decommissioning activity can mean that large amounts of the waste produced is below the exemption limit and could therefore be disposed of in conventional landfill. Sellafield Ltd. is using a pair of Canberra WM2750 Clearance Monitors to assay 100 litre packages of soft waste produced in some of their decommissioning activities at Sellafield. The WM2750 uses low resolution gamma spectrometry (LRGS) to determine the radionuclide content of packages or drums of LLW up to a maximum of 140 litre capacity. It uses a lead shielded measurement chamber to reduce the local radiation background along with high efficiency sodium iodide (NaI) detectors in order to obtain the measurement sensitivity required to be able to distinguish between LLW and exempt waste in a measurement time of less than 1 minute per package. This paper describes the waste monitoring process and the design of the clearance monitor — in particular how it was calibrated and the performance testing that was carried out to ensure that waste items identified by the monitors as being exempt waste are suitable for disposal to a conventional landfill site.
Development of a Carbon-14 Bioseperation Technique for Cleanup of Nuclear Graphite
Authors: Molokwane Pulane E. , Chirwa Evans M. N.
Published: June 2009

A preliminary study on the bioseperation of Carbon-14 was carried out using a mixed-culture of microorganisms obtained from the Pebble Bed Modular Reactor Company (PBMR). The culture demonstrated the ability to bioaccumulate radiocarbon-14 (C-14) from solution. The experiment consisting of a growth vessel and a biofilter connected in a closed loop. The biofilter was not installed for the purpose of treatment but rather as a method of isolation of microorganisms for further processing. Significant amounts of C-14 were detected in the trapped cells in the biofilter, significantly higher than in controls taken before adding carbon sources containing C-14. The microorganisms were grown under micro-aerobic conditions with graphite carbon and commercially purchased powdered carbon as the predominant supplied carbon sources. Small amounts of sucrose (500 mg/L) were added at 48 hour intervals to promote the growth of heterotrophic microorganisms. Additional work is required to determine the amount of C-14 escaping through gases produced as a product of metabolism (CO2 and CO) and the total carbon metabolized by the microorganisms in order to report with accuracy the degree of separation of C-14 from the C-14/C-12 mixture. A proof of concept study is underway to determine the C-14 mass balance, characterize the microorganisms in the reactor, and establish the presence or absence of processes that might have affected the preliminary observations.
Cost Calculations at Early Stages of Nuclear Research Facilities in the Nordic Countries
Authors: Iversen Klaus , Salmenhaara Seppo , Backe Steinar , Cato Anna , Lindskog Staffan , Callander Clas , Efraimsson Henrik , Andersson Inga , Sjo¨blom Rolf
Published: June 2009

The Nordic countries Denmark, Norway and Sweden, and to some extent also Finland, had very large nuclear research and development programs for a few decades starting in the nineteen fifties. Today, only some of the facilities are in use. Some have been decommissioned and dismantled while others are at various stages of planning for shutdown. The perspective ranges from imminent to several decades. It eventually became realized that considerable planning for the future decommissioning is warranted and that an integral part of this planning is financial, including how financial funds should be acquired, used and allocated over time. This necessitates that accurate and reliable cost estimates be obtained at all stages. However, this is associated with fundamental difficulties and treacherous complexities, especially for the early ones. Eventually, Denmark and Norway decided not to build any nuclear power plants while Finland and Sweden did. This is reflected in the financing where the latter countries have established systems with special funds in which money is being collected now to cover the future costs for the decommissioning of the research facilities. Nonetheless, the needs for planning for the decommissioning of nuclear research facilities are very similar. However, they differ considerably from those of nuclear power reactors, especially with regard to cost calculations. It has become apparent in the course of work that summation types of cost estimation methodologies give rise to large systematic errors if applied at early stages, in which case comparison based assessments are less biased and may be more reliable. Therefore, in order to achieve the required quality of the cost calculations, it is necessary that data and experience from authentic cases be utilized in models for cost calculations. It also implies that this calculation process should include a well adopted learning process. Thus, a Nordic co-operation has been established for the exchange and evaluation of cost-related information on nuclear research facilities. The aim is to identify good practices, accumulate experience, compile data from actual plants and projects, and to derive methodology for cost calculations, especially for early stages. The work includes the following tasks which constitutes the bulk of the present paper: • identification of good practice with regard to the following: - strategy and planning, - methodology selection, - radiological surveying, - uncertainty analysis; • descriptions of relevant plants, features and projects: - decommissioning of reactor DR 1 in Denmark, - decommissioning of reactor R 1 in Sweden, - decommissioning of the pilot scale uranium fuel reprocessing plant in Norway, - planning for the future decommissioning of the TRIGA reactor in Finland; • techniques for assessments of costs introduction.
Destruction of Nuclear Organic Waste by Supercritical Water Oxidation: Scale-Up of the Process
Authors: Moussiere S. , Roubaud A. , Joussot-Dubien Ch. , Turc H.-A. , Fournel B.
Published: June 2009

In order to design and then define appropriate dimensions for a supercritical oxidation reactor, a 2D and 3D simulation of the fluid dynamics and heat transfer during the oxidation process has been performed. The solver used is a commercial code, Fluent 6.2. The turbulent flow field in the reactor, created by the stirrer is taken into account with a k-omega model and a swirl imposed to the fluid. In the 3D case the rotation of the stirrer can be modeled thanks to the sliding mesh model. The reactivity of the system is taken into account with a classical combustion model EDC. Comparisons with experimental temperature measurements validate the ability of the CFD modeling to simulate the supercritical water oxidation process. Simulation results provide us a view inside the reactor on the flow, temperature fields and the oxidation localization and development. Results indicate that the flow can be considered as piston-like, heat transfers are strongly enhanced by the stirring. Hence the scaling up of the reactor volume, to reach a treatment capacity of 1 Kg/h of pure organics, can be done regarding the necessary residence times and temperature distribution needed for a complete destruction of the organic matter.
Coincidence Gate Utilization Factors for Neutron Correlation Counters With Up to Three Components in the Die-Away Profile
Authors: Croft S. , McElroy R. D. , Kane S. C.
Published: June 2009

We present analytical expressions for the gate utilization factors (GUFs), up to fourth order, for both signal triggered and random triggered histograms based on a three-component capture time profile. These are useful for refined design performance calculations of passive neutron multiplicity counters using shift register correlation analysis. To our knowledge, these expressions are new.
Relative Performance of a TGS for the Assay of Drummed Waste as Function of Collimator Opening
Authors: Kane S. C. , Croft S. , McClay P. , Venkataraman R. , Villani M. F.
Published: June 2009

Improving the safety, accuracy and overall cost effectiveness of the processes and methods used to characterize and handle radioactive waste is an on-going mission for the nuclear industry. An important contributor to this goal is the development of superior non-destructive assay instruments. The Tomographic Gamma Scanner (TGS) is a case in point. The TGS applies low spatial resolution experimental computed tomograghy (CT) linear attenuation coefficient maps with three-dimensional high-energy resolution single photon emission reconstructions. The results are presented as quantitative matrix attenuation corrected images and assay values for gamma-emitting radionuclides. Depending on a number of operational factors, this extends the diversity of waste forms that can be assayed, to a given accuracy, to items containing more heterogeneous matrix distributions and less uniform emission activity distributions. Recent advances have significantly extended the capability to a broader range of matrix density and to a wider dynamic range of surface dose rate. Automated systems sense the operational conditions, including the container type, and configure themselves accordingly. The TGS also provides a flexible data acquisition platform and can be used to perform far-field style measurements, classical segmented gamma scanner measurements, or to implement hybrid methods, such as reconstructions that use a priori knowledge to constrain the image reconstruction or the underlying energy dependence of the attenuation. A single, yet flexible, general purpose instrument of this kind adds several tiers of strategic and tactical value to facilities challenged by a diverse and difficult to assay waste streams. The TGS is still in the early phase of industrial uptake. There are only a small number of general purpose TGS systems operating worldwide, most being configured to automatically select between a few configurations appropriate for routine operations. For special investigations, one may wish to widen the repertoire but there is currently little guidance as to the trade-offs involved. In this work, we address this weakness by studying the performance of a typical TGS arrangement as a function of collimator opening, scan pattern and scan time for a representative selection of simulated waste forms. Our focus is on assessing the impact on the precision and accuracy of the quantitative assay result but we also report the utility of the imaging information in confirming acceptable knowledge about the packages.
Packaging and Transportaion of Radioactively Contaminated Lead
Authors: Gleason Eugene , Holden Gerard
Published: June 2009

Under the management of the Nuclear Decommissioning Authority (NDA) the government of the United Kingdom has launched an ambitious program to remediate the nation’s nuclear waste legacy. Over a twenty-five year period NDA plans to decommission several first generation nuclear power plants and other radioactive facilities. The use innovative, safe “fit for purpose” technologies will be a major part of this complex program. This paper will present a case study of a recently completed project undertaken in support of the nuclear decommissioning activities at the Sellafield site in the United Kingdom. The focus is on an innovative application of new packaging technology developed for the safe transportation of radioactively contaminated lead objects. Several companies collaborated on the project and contributed to its safe and successful conclusion. These companies include British Nuclear Group, Gravatom Engineering, W. F. Bowker Transport, Atlantic Container Lines, MHF Logistical Solutions and Energy Solutions. New containers and a new innovative intermodal packaging system to transport the radioactive lead were developed and demonstrated during the project. The project also demonstrated the potential contribution of international nuclear recycling activities as a safe, economic and feasible technical option for nuclear decommissioning in the United Kingdom. Eugene J. Gleason is Vice President for Government Affairs and International Initiatives at MHF Logistical Solutions. Gerard Holden is Director for Waste packaging at Gravatom Engineering Limited.
Final Covering of the Ronneburg Uranium Mining Site
Authors: Hoepfner Uwe
Published: June 2009

The rehabilitation of WISMUT’s former Ronneburg uranium mining site involves backfilling of waste rock to the Lichtenberg open pit. The relocation project comprises about 110 million m3 of sulphide-bearing and AMD-generating waste rock which makes it the most important and most cost-intensive single surface restoration project conducted by WISMUT at the Ronneburg site. The backfilled waste rock has to be covered on an area of about 220 ha to control water infiltration and gas diffusion. Design planning for the final cover placement which began in 2004 had to be based on a comprehensive cost-benefit analysis as well on field tests of alternative cover options which are in compliance with legal requirements. An intensive testing program concerning the vadose zone of soil covers has therefore been started in 2000. The paper presents an overview of the monitoring program and the results of the vadose zone measurements. The water of soil covers have to be predicted for extended evaluation periods. Therefore water balance simulations of single layer covers (storage and evaporation concept) taking current and future soil and climate conditions into account are performed with the HYDRUS_2D code.
Progress and Experiences From the Decommissioning of the Eurochemic Reprocessing Plant
Authors: Gilis R. , Lewandowski P. , Ooms B. , Reusen N. , Van Laer W. , Walthe´ry R.
Published: June 2009

Belgoprocess started the industrial decommissioning of the main process building of the former EUROCHEMIC reprocessing plant in 1990, after completion of a pilot project in which two buildings were emptied and decontaminated to background levels. The remaining structures were demolished and the concrete debris was disposed of as industrial waste and green field conditions restored. The Eurochemic reprocessing plant operated from 1966 to 1974 to process fuel from power reactors and research reactors. The main building is a large concrete structure, comprising a surface area of 55,000 m2, concrete volume 12,500 m3, and 1,500 Mg of metal components. The building is divided into multiple cells. About 106 individual cell structures have to be dismantled, involving the removal and decontamination of equipment from each cell, the decontamination of the cell walls, ceilings and floors, the dismantling of the ventilation system. Most of the work involves hands-on operations under protective clothing tailored to each specific task. Tool automation and automatic positioning systems are successfully applied. In view of the final demolition of the main process building, the main process building is divided into three parts — each part is isolated from the others. In the middle of 2008, after the removal of the NDA-IPAN/GEA installation, the eastern part will be demolished. The paper presents a status overview of the decommissioning and decontamination activities at the main process building of the former Eurochemic reprocessing plant on the nuclear site of Dessel in Belgium. The specific BELGOPROCESS approach will be highlighted, in which the decommissioning activities are carried out on an industrial scale with special emphasis on cost minimisation, the use of technology on an industrial representative scale and the specific alpha contamination of equipment and building surfaces, requiring that the decommissioning work is done with adequate protective clothing. Also specific breathing and cooling air systems have been provided to allow the operators to carry out the decommissioning tasks in acceptable working conditions.
Radioactive Waste Disposal in Germany: No Site Decision - Keeping Competence
Authors: Kienzler Bernhard , Geckeis Horst , Gompper Klaus , Klenze Reinhardt
Published: June 2009

The research programme of the Institut fu¨r Nukleare Entsorgung (INE) at the Forschungszentrum Karlsruhe is dedicated to elaborate the fundamental understanding of radionuclide/actinide interactions with various components of the groundwater and with the relevant hostrock materials at disposal relevant trace concentrations. INE’s research programme was not biased after enactment of the Gorleben moratorium in 2001. This paper presents current R&D with respect to application in performance assessment/safety case of nuclear waste disposal. Focus is given to the leading role of the institute in various projects within EU framework programmes.
Study on Creating Hydraulic Tomography for Crystalline Rock Using Frequency Dependent Elastic Wave Velocity
Authors: Yoshimura K. , Sakashita S. , Ando K. , Bruines P. , Blechschmidt I. , Kickmaier W. , Onishi Y. , Nishiyama S.
Published: June 2009

The objective of this study is to establish a technique to obtain hydraulic conductivity distribution in granite rock masses using seismic tomography. We apply the characteristic that elastic wave velocity disperses in fully saturated porous media on frequency and this velocity dispersion is governed by the hydraulic conductivity — this characteristic has been confirmed in laboratory experiments. The feasibility and design of the field experiment was demonstrated in a first step with numerical simulations. In a second step we applied the technique to the fractured granite at the Grimsel Test Site in Switzerland. The emphasis of the field campaign was on the evaluation of the range of applicability of this technique. The field campaign was structured in three steps, each one corresponding to a larger spatial scale. First, the seismic tomography was applied to a small area — the two boreholes were located at a distance of 1.5m. In the following step, we selected a larger area, in which the distance of the boreholes amounts to 10 m and the field corresponds to a more complex geology. Finally we applied the testing to a field where the borehole distance was of the order of 75 m. We also drilled a borehole to confirm hydraulic characteristic and reviewed hydraulic model in the 1.5m cross-hole location area. The results from the field campaign are presented and their application to the various fields are discussed and evaluated.