bluuebunny/crossref_metadata_2025 huggingface dataset:
The Emergence of Sustainable Practice Within Decommissioning
Authors: Adamson David William , Francis Jonathan
Published: December 2010

Despite the advance of sustainable practice and energy efficient techniques outside of the nuclear industry, at the start of the 21st Century there was a lack of published guidance aimed at their adoption at specifically nuclear facilities. Even with the establishment of the Nuclear Decommissioning Authority, there is very little guidance published on how to adopt sustainable practices during decommissioning. There have been instances where energy efficiency had affected design and operations decisions. Projects aimed at responsible housekeeping, switching off lights, and changes to the nuclear ventilation design philosophy illustrate a desire for action, but these activities were championed by interested and motivated employees. Sustainable practice had not at that time received a strategic lead that resulted in a management structure to enable a coordinated and concerted effort in sustainable practice. This paper traces the progress during the 20th and early 21st Centuries, whereby sustainable practice is now established within a much firmer foundation of case study, guidance and organisational structure; to embed sustainable practice within the United Kingdom’s current decommissioning programme. It looks at the development of relevant literature and, through interviews with key managers and external stakeholders, demonstrates (i) the degree to which two essential guidance documents (the NiCOP and CIRIA SD:SPUR) are permeating the industry, (ii) how the current work of the Characterisation and Clearance Group has evolved to influence the decontamination and dismantling planning procedures and (iii) the transition from identifying ‘free-release’ materials to actually releasing them for re-use in the community.
Implementation of the Best in Class Project Management and Contract Management Initiative at the Department of Energy’s Office of Environmental Management
Authors: Van Camp Scott G. , Deiters Michael G. , Stevenson Jeremy S. , Jamison Timothy P.
Published: December 2010

Since its creation in 1989, the Department of Energy (DOE), Office of Environmental Management (EM) has struggled with a legacy of inadequate project management and contract management. This has been manifested in recurring scope changes, cost overruns and schedule delays, and has been documented in multiple internal and external reviews. EM has committed itself to improving project performance and undertaken a number of proactive management initiatives including the development of a “Best in Class” Project Management and Contract Management organization (i.e., the BICPM Initiative). During 2007, EM assessed the status of project management and contract management at 15 EM sites. These assessments evaluated strengths and weaknesses in 12 key project management capabilities and three contract management benchmarks. The January 2008 Compilation Assessment Report showed that EM faces significant challenges in its mission execution due to staffing shortages, project and contract management integration, insufficient project-oriented culture, and lack of a clear role for Headquarters in BICPM. EM then formulated a strategy to meet their objectives in the March 2008 Corporate Implementation Plan. It summarizes BICPM efforts, introduces the vision for BICPM, identifies the strategy for achieving BICPM, and describes a process for implementing BICPM. That is, it acts as a roadmap to address EM’s challenges. It also documents 18 Recommended Priority Actions (RPAs) that are the key to correcting these challenges. These RPAs provide a clear path forward that can be communicated to the entire EM organization and provide the foundation upon which a BICPM culture can be built. EM has since gained considerable momentum and progress towards institutionalizing BICPM. This paper provides a discussion of the BICPM Initiative and its implementation.
A Review of Vortex Amplifier Design in the Context of Sellafield Nuclear Operations
Authors: Birch M. J. , Doig R. , Francis J. , Parker D. , Zhang G.
Published: December 2010

Vortex amplifiers have for over 30 years been used to ensure containment of glove-box ventilation in the event of a barrier breach, the most likely such breach being damage to the glove itself. Containment is achieved using fluidic principles to control the glove-box depression and ventilation rate under both normal and emergency conditions; in the event of such a breach vortex amplifiers can switch quickly between these two states without recourse to electrical, pneumatic or manual intervention. This paper begins by summarising the developments in vortex amplifier design used at the Sellafield site by successive companies engaged in fuel technology, reprocessing and decommissioning (British Nuclear Fuels PLC (BNFL), BNFL Engineering Limited, British Nuclear Group and Sellafield Limited). The main reasons for design changes have been practical issues of set-up, cleaning, filter and waste minimisation, and space limitations. The development culminates in the use of a smaller version of the vortex amplifier (VXA) which is a nearly exact geometrical scaling of its predecessor and which has been standard design for over a decade. Initial use of this device, the mini–VXA, led to a substantial increase in the amount of inert gas needed to maintain the required oxygen-depletion conditions within the glove-box, implying some escape of oxygen into the glove-box. The use of the mini–VXA introduced practical issues relating to (i) its control characteristics and (ii) the reverse flow of air in the supply port. Comparison with the published design specification demonstrates that the geometrical scaling process has led to a slightly hysteric characteristic. Tests conducted by the authors indicate (i) that the origin of the escaping oxygen is the control air feeding back through the supply ports and (ii) that a prototype chamber and orifice plate arrangement between the glove-box and mini–VXA significantly reduces the inert gas demand in normal usage. This prototype arrangement introduced problems in maintaining a clean environment in the chamber, so the chamber and orifice was substituted by a detachable cowl that enabled the mini–VXA to be located within the glove-box and provided access for cleaning.
Education and Industry Partnership: A Case Study of Co-Delivery
Authors: Mercer Tim , Francis Jonathan
Published: December 2010

One of the essential elements for safe operation of a nuclear licensed site is the availability to the licensee in sufficient numbers of suitably qualified and experienced people to carry out and manage the operations and associated design work. In the last few years, there have been a number of reports to illustrate the recent and current problems of recruiting such people to work in the traditional locations for nuclear personnel in the North-West of England. Concern for the immediate future is exacerbated by a peculiar demographic of the people currently employed in positions demanding higher level skills. In response to the growing realization that there is an impending skills gap that needs to be filled, Sellafield Limited’s Talent Management team (and latterly with support of the NDA) have been working with a number of education and training providers to put in place bespoke courses aimed at overcoming this shortage. In the absence of a steady stream of willing graduates from technical and management courses, the primary strategy has been to encourage life-long learning and up-skilling amongst its employees, targeting those who, for whatever reason upon leaving school, missed their opportunity to study and progress to train at a high level, but who possess that potential and have now developed a keenness to proceed with that study in later life. One Foundation Degree has been selected for development of a unique approach to higher education. The work of University of Central Lancashire and its West-Cumbrian education and training partners has featured as a case study in other media, but this paper reports on a fresh development within that work: co-delivery. Co-delivery relates to a partnership of educationalists and industrialists, with an emphasis on industrial numbers on the course development steering group. The means by which a significant proportion of the course is strongly workplace related are presented and the benefits and problems that this introduces are discussed. The course uses the industry as a vehicle to communicate concepts and develop problem-solving skills. Rather than the major vocational aspects being confined to just a few ‘workplace’ modules, the industry permeates all modules and co-delivery is part of a good many. A report is also made on the areas of provision where the major capability and expertise is located in just a few industrialists; and how that aspect is learned within a co-delivery course.
Surface Decontamination by Photocatalysis
Authors: Wilbraham R. J. , Boxall C. , Taylor R. J.
Published: December 2010

Currently in the nuclear industry, surface contamination in the form of radioactive metal or metal oxide deposits is most commonly removed by chemical decontamination, electrochemical decontamination or physical attrition. Physical attrition techniques are generally used on structural materials (concrete, plaster), with (electro)chemical methods being used to decontaminate metallic or painted surfaces. The most common types of (electro)chemical decontamination are the use of simple mineral acids such as nitric acid or cerium (IV) oxidation (MEDOC). Use of both of these reagents frequently results in the dissolution of a layer of the substrate surface increasing the percentage of secondary waste which leads to burdens on downstream effluent treatment and waste management plants. In this context, both mineral acids and MEDOC can be indiscriminate in the surfaces attacked during deployment, e.g. attacking in transit through a pipe system to the site of contamination resulting in both diminished effect of the decontaminating reagent upon arrival at its target site and an increased secondary waste management requirement. This provides two main requirements for a more ideal decontamination reagent: Improved area specificity and a dissolution power equal to or greater than the previously mentioned current decontaminants. Photochemically promoted processes may provide such a decontamination technique. Photochemical reduction of metal ion valence states to aid in heavy metal deposition has already been extensively studied [1], with reductive manipulation also being achieved with uranium and plutonium simulants (Ce) [2]. Importantly photooxidation of a variety of metals, including neptunium [3], has also been achieved. Here we report on the potential application of this technology to metal dissolution.
Decommissioning of a Uranium Conversion Plant and a Low Level Radioactive Waste for a Long Term Disposal
Authors: Choi Yun D. , Hwang D. S. , Chung U. S.
Published: December 2010

A decommissioning project for a uranium conversion plant was conducted to restore it to a safe environmental condition and minimal low level radioactive wastes which were converted to stable chemical forms for a long term disposal. In the middle of 2004, a decommissioning program for a conversion plant, which was constructed in 1982, and treated about 300 tons of natural uranium until it was shut down in 1992, obtained its approval from the regulatory body. Actual dismantling and decontaminating activities have been performed since July 2004 and will be finished by December 2009. The decommissioning works were mainly divided into two parts: the inside of the building containing the process equipments; the lagoon sludge generated during the plant operation. The decommissioning of the inside of the building was carried out by dismantling the process equipment, which were firstly segmented and decontaminated by polishing and washing with steam and chemicals or melting, and then decontamination for the surfaces inside the building by excavating or grinding the concrete walls. The decontamination goals were below 0.2Bq/g for the metallic segments and below 0.4Bq/cm2 for the concrete walls. Decontamination methods were selected according to the degree of contamination and a minimization of the low level radioactive wastes was conducted throughout the decommissioning work. The lagoon sludge waste had two types, one was an various inorganic nitrate salt mixture containing a very low concentration of uranium, about 200∼300ppm, in Lagoon-II and the other was an inorganic nitrate salt mixture containing a few percent of uranium in Lagoon-I. To treat these sludge wastes a thermal decomposition facility was constructed and operated to produce stable sludge wastes containing uranium oxides which are stable in the air. The final sludge wastes after a thermal treating for the sludge waste of Lagoon-I could be reused. The final residual radioactivity for the inside of the building will be measured to confirm a complete decontamination of the uranium to back ground level and then the building will be considered for another use.
Critical Evaluation on the Use of Non-Destructive Assay of Nuclear Packages Through Destructive Breakdown and Inventory Recovery
Authors: Hepworth Stephen , Way Rob , Sharpe Jonathan
Published: December 2010

Two historic waste packages on the Sellafield Site with potentially high fissile content were destined for re-packaging. Prior to relocation and subsequent breakdown, each item underwent a campaign of non-destructive assay. The aim of the assay was to gather information that would assist with the production of a safety case. The assay work consisted of: conventional x-ray radiography to determine the identity of the contents; gamma imaging and three-dimensional tomographic re-construction to determine the location of the gamma emitting material; and neutron coincidence counting coupled with gamma spectrometry to assign a fissile mass. Most items on the Sellafield Site that undergo non-destructive assay normally remain intact or are re-packaged with minimal interference of the content. However, in this instance each item was dismantled and the fissile material recovered. This paper provides a comparison of the measured results with the actual results for each technique. The x-ray radiographic information was used to construct a three dimensional representation of the contents of each item. This information was useful in identifying the plant items contained within. The results were discussed with plant operators who were familiar with the historic plant. The operators were able to identify areas of likely accumulations of fissile material. The two-dimensional gamma survey and subsequent threedimensional re-construction revealed the location of the gamma emitting materials within the packages. It was assumed that areas of increased gamma activity indicated areas of increased fissile mass. The neutron coincidence counting, gamma measurements and subsequent modelling provided an estimate of fissile mass for each item. The fissile mass estimate is an essential component of the safety case planning. However, it does not provide all of the information required to plan an ALARP breakdown strategy. The combination of the information gathered through assay was essential to construct a safety case that considered: transport requirements; criticality risk; dose to operator and breakdown methodology. Subsequent studies suggest that improved combination of the data acquired from the assay programme and more post-assay discussion of the specific issues and implications between the various stakeholder groups (e.g. assay engineers and safety case authors) may have led to a fissile mass result with lower uncertainties, which given a more favourable example would have resulted in a significant cost savings through reduced criticality assessment burden and mitigation actions and reduced security control.
Repository Site Characterization: Comparing International Experience
Authors: Goldsworthy Martin , Popp Till , Seidel Knut , Bruns Johannes
Published: December 2010

An important part of the work described here was a study of existing international experience in investigating deep geological repository sites. The objective of this study was to derive a basis for planning the content and extent of investigations which might be carried out in Germany in the future. Such investigations would be required in the course of a site selection process for a repository for HLW (high level radioactive waste). For this purpose information on suitable sites was gathered, mainly from literature sources. Suitable in this context meant two things. Firstly, the investigated site should be in rock similar to four being considered in Germany (salt, clay, crystalline and other hard rock under a clay cover). Secondly, the investigations carried out could reasonably be considered as being intended to lead to the use of the site as a repository. The investigation processes were presented, analysed and compared. The comparison was based on the quality and the intensity of the methods employed to obtain the information necessary for deciding between candidate repository sites in terms of safety and the feasibility of construction. In the final stage of the work the analysis and presentation method developed for the international sites was applied to the investigations already carried out at three German sites (Gorleben — a prospective HLW repository, Morsleben — an existing but now not operational repository for radioactive waste and Konrad — a repository currently under construction). The reported investigatory work was compared with the ideal investigations developed on the basis of the existing international experience.
Options for the Removal of Contaminated Concrete From the Bore of the Windscale Pile Chimney
Authors: Campbell Colin , Hepworth Stephen , Hunt Jeremy
Published: December 2010

A legacy of the 1957 Windscale Pile reactor fire is the penetration of radioactive contamination into the internal surface concrete of the chimney bore. Gamma imaging of Cs-137 has shown that the contamination is widespread throughout the chimney, and core samples have shown that the contamination has penetrated to depths of around 5–25mm. The Pile chimney is 100m tall and has an internal bore diameter of 15m. It is constructed of a hard concrete comprised of Whinstone aggregate. The baseline decommissioning scheme is to remove approximately 5–25mm of the surface concrete from the entire bore of the chimney. The technology baseline in 2006 was to remove layers of contaminated concrete by mechanical means using shavers or scabblers. However, risks associated with mechanical technologies that may preclude their use include: the ability of mechanical devices to remove the hard concrete; clogging of the devices due to wet concrete; and deployment of the delivery systems. This paper discusses the options under consideration to reduce the risks associated with the removal of the contaminated concrete through application of alternative techniques. The present baseline technology is high pressure water jetting technique. Demonstrations have shown that this technology can successfully remove concrete without significant reaction forces. However, an inherent problem with this technology is the production of secondary liquid effluent waste, which would need to be treated by an appropriate conditioning process. To address the secondary effluent waste issue, technologies that produce little or no secondary waste have been considered. The technologies that have been considered are laser scabbling, microwave scabbling and nitrogen jet blasting. The paper discusses each technique in turn, highlighting their advantages and disadvantages. The results of an in-active laser scabbling and high pressure liquid nitrogen jetting trial are presented. The paper concludes with a discussion of the merits of each technology in support of the future strategy for concrete removal.
Measurement of Solid-Liquid Mixtures Using Electrical Tomography Measurement Techniques
Authors: Bolton Gary , Stanley Stephen
Published: December 2010

Electrical Impedance Tomography measurement techniques have been applied to a variety of solid-liquid processes in the laboratory and on industrial plant. This paper reviews the advances in the measurement techniques to determine key process information in solid-liquid systems such as concentration mapping, mixture homogeneity, interface detection and suspension velocity. A number of applications to solid-liquid flow applications are presented. The use of the technology for improved design and operation is highlighted, as are the opportunities for on-line sensing for flow measurement, fault detection and process control. A recent development in high-speed electrical imaging has allowed velocity maps to be calculated for fast flowing suspensions (up to 10 ms−1). The methodology for determining mixture homogeneity in both pipeline flows and agitated tanks are summarized. Finally the application of a linear ERT electrode array to identify interfaces during the settling of solid-liquid mixtures is presented.
Fluorescence Spectral Imaging as a Tool for Locating Uranium Deposited on Surfaces
Authors: Monts David L. , Wang Guangjun , Su Yi , Jang Ping-Rey , Waggoner Charles A.
Published: December 2010

In the environment, metallic uranium readily oxidizes to form uranium compounds that contain the uranyl (UO2+2) moiety. For more than a hundred and fifty years, it has been known that when illuminated with ultraviolet (UV) light, uranyl compounds exhibit characteristic fluorescence in the visible region (450–650 nm). We report our efforts to develop a transportable, quantitative Fluorescence Spectral Imaging (FSI) system as a tool for locating and quantifying uranyl compounds dispersed in soils and on other surfaces. A project is underway to develop a set of sensors to locate expended depleted uranium (DU) rounds and to process soil and debris to recover the material from domestic firing ranges. The FSI system can also be utilized to monitor excavation of DU munitions and separation of uranyl compounds from soils. FSI images are acquired by illuminating a surface with a UV light and using a narrow bandpass filter on a camera, recording an image of the resulting fluorescence. The FSI image provides both spatial and spectral information. The FSI system is described and its performance characterized using field samples.
Development of Joint Regulatory Guidance on the Management of Higher Activity Radioactive Wastes on Nuclear Licensed Sites
Authors: Bacon Mick , Ilett Doug , Whittall Andy
Published: December 2010

In 2006 the UK Governments response to recommendations by its Committee on Radioactive Waste Management (CoRWM) established, in England and Wales, that geological disposal, supported by safe and secure interim storage, is the preferred route for the long-term management of higher-activity radioactive waste (i.e. that which is not suitable for near-surface disposal). It also gave the responsibility for delivering the programme for a deep geological repository to the Nuclear Decommissioning Authority (NDA). The Scottish Government has a policy of long term, near site, near surface safe and secure interim storage. To support the open and transparent approach promised by Government, the Health and Safety Executive (HSE), the Environment Agency and the Scottish Environment Protection Agency (SEPA) are developing joint guidance on the management of higher-activity radioactive waste to explain regulatory objectives in securing safe and secure interim storage and the associated management of radioactive wastes. The guidance comes in two parts: • Guidance on the regulatory process; • Technical guidance modules. The guidance promotes a cradle to grave approach to radioactive waste management and by aligning the regulatory interests of environmental and safety regulators it delivers one of the Government’s “Better Regulation” objectives. This paper describes the process by which the joint guidance was produced with particular emphasis on stakeholder engagement. It describes the key features of the guidance, including the concept of the radioactive waste management case (RWMC). Finally the problems encountered with dissemination and implementation are discussed together with measures taken by the regulators to improve these aspects.
Increasing Operational Efficiency in a Radioactive Waste Processing Plant
Authors: Turner T. W. , Watson S. N.
Published: December 2010

The solid waste plant at Harwell in Oxfordshire, contains a purpose built facility to input, assay, visually inspect and sort remote handled intermediate level radioactive waste (RHILW). The facility includes a suite of remote handling cells, known as the head-end cells (HEC), which waste must pass through in order to be repackaged. Some newly created waste from decommissioning works on site passes through the cells, but the vast majority of waste for processing is historical waste, stored in below ground tube stores. Existing containers are not suitable for long term storage, many are already badly corroded, so the waste must be efficiently processed and repackaged in order to achieve passive safety. The Harwell site is currently being decommissioned and the land is being restored. The site is being progressively delicensed, and redeveloped as a business park, which can only be completed when all the nuclear liabilities have been removed. The recovery and processing of old waste in the solid waste plant is a key project linked to delicensing of a section of the site. Increasing the operational efficiency of the waste processing plant could shorten the time needed to clear the site and has the potential to save money for the Nuclear Decommissioning Authority (NDA). The waste processing facility was constructed in the mid 1990s, and commissioned in 1999. Since operations began, the yearly throughput of the cells has increased significantly every year. To achieve targets set out in the lifetime plan (LTP) for the site, throughput must continue to increase. The operations department has measured the overall equipment effectiveness (OEE) of the process for the last few years, and has used continuous improvement techniques to decrease the average cycle time. Philosophies from operational management practices such as ‘lean’ and ‘kaizen’ have been employed successfully to drive out losses and increase plant efficiency. This paper will describe how the solid waste plant at Harwell has continuously increased the throughput of RHILW, which should lead to significant programme savings.
An Overview of US EPA’s Current Radioactive Waste Management and General Radiation Protection Efforts
Authors: Peake R. Thomas , Schultheisz Daniel , Setlow Loren W. , Littleton Brian , Rosnick Reid , Czyscinski Ken
Published: December 2010

The United States Environmental Protection Agency’s (EPA) Radiation Protection Division is the portion of EPA (or the Agency) that develops environmental standards for radioactive waste disposal in the United States. One current issue of concern is the disposal of low activity radioactive waste (LAW), including wastes that would be produced by a radiological dispersal device (RDD), for which current disposal options may be either inconsistent with the hazard presented by the material or logistically problematic. Another major issue is related to the resurgence in uranium mining. Over the past several years, demand for uranium for nuclear power plant fuel has increased as has the price. The increase in price has made uranium mining potentially profitable in the US. EPA is reviewing its relevant regulations, developed primarily in the 1980s, for potential revisions. For example, in-situ leaching (also known as in-situ recovery) is now the technology of choice where applicable, yet our current environmental standards are focused on conventional uranium milling. EPA has two actions in process, one related to the Clean Air Act, the other related to revising the environmental standards that implement the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). Separately, but related, EPA has developed over the last several years uranium mining documents that address technologically enhanced natural occurring radioactive materials (TENORM) from abandoned uranium mines, and wastes generated by active uranium extraction facilities. Lastly, in 1977 EPA developed environmental standards that address nuclear energy, fuel fabrication, reprocessing, and other aspects of the uranium fuel cycle. In light of the increased interest in nuclear power and the potential implementation of advanced fuel cycle technologies, the Agency is now reviewing the standards to determine their continued applicability for the twenty-first century.
Technology Development for Decommissioning in FUGEN and Current Status
Authors: Kitamura Koichi , Sano Kazuya , Nakamura Yasuyuki , Matsushima Akira , Matsuo Hidehiko , Ishiyama Masahiro , Tezuka Masashi , Haneda Takahiro
Published: December 2010

The decommissioning program of proto-type Advanced Thermal Reactor (ATR) FUGEN has started in 2008 as first decommissioning of the commercial-scale water reactor. It consists of four periods, considering the transportation of spent fuels and the radioactive decrease of highly activated materials. It is expected that the whole program of decommissioning will be completed until 2028. Now, the decommissioning is under the first period, spent fuels and heavy water has been carrying out from FUGEN, and a part of the turbine system with relatively low radioactive contamination has been dismantled. FUGEN has a complicated core structure consisting 224 fuel channels with pressure tubes and calandria tank, etc. and used heavy water as moderator, unlike other light water reactor (LWR). The materials of the core structure were highly activated due to a long term operation, tritium and C-14 were generated, and the facilities were contaminated by them. Thus, it is important to study the dismantling technology of the reactor core and the decontamination technology, considering characteristics of FUGEN such as core structure and radioactive inventory in advance. In this presentation, the contents of the decommissioning program and its current status such as dismantling work of a part of the turbine system, the studying situation of dismantling technology of reactor core using Abrasive Water Jet (AWJ) which is a candidate of cutting technologies, the examination of tritium decontamination in heavy water system, the study of decontamination technology for C-14 will be presented mainly.
Verification Test of Clearance Automatic Laser Inspection System for Surface Contamination Measurement
Authors: Sasaki Michiya , Ogino Haruyuki , Ichiji Takeshi , Hattori Takatoshi
Published: December 2010

Recently, a clearance automatic laser inspection system (CLALIS) has been developed for clearance measurement of scrap metals and concrete debris. It utilizes three-dimensional laser scanning, gamma-ray measurement and Monte Carlo calculation, and its outstanding detection ability has been verified. In Japan, when an object is removed from a radiation-controlled area, the activity level must be lower than the surface contamination density standard of 4 Bq/cm2 for beta and gamma emitters, which is one-tenth of the surface contamination density limit. According to the clearance inspection report published by the Nuclear Safety Commission of Japan, the activity level of waste must be compared with not only the clearance level but also the surface contamination density standard for clearance inspection. To demonstrate that CLALIS can also be used for the measurement of surface contamination, a verification test was carried out using actual metal waste samples of various shapes, sizes and activity levels at Kashiwazaki-Kariwa Nuclear Power Station of Tokyo Electric Power Company. As a result, it was clarified that CLALIS gives a conservative value for surface contamination compared with the conventional GM survey meter measurement. This because the activities of metal waste samples were estimated using the total count rate, a fixed average surface area of 100 cm2 and the conservative source position assumed in the Monte Carlo calculation for the calibration factor. In a nuclear power plant, the actual judgment of whether an object can removed from a radiation-controlled area is based on whether the result of surface contamination measurement is lower than the detection limit, which is significantly lower than 4 Bq/cm2. According to this criterion, CLALIS provides an almost identical judgment to the GM survey meter, which means that CLALIS can be used as a rational clearance monitor to carry out clearance level and surface contamination inspections in a single radiation measurement. The prospective detection limit for CLALIS at nuclear power plants is also discussed and compared with that for the conventional surface contamination monitors.
The Rock Creep Evaluation in the Analysis System for the Long-Term Behavior of TRU Waste Disposal System
Authors: Ohno Shintaro , Morikawa Seiji , Mihara Morihiro
Published: December 2010

MACBECE (Mechanical Analysis system considering Chemical transitions of BEntonite-based and CEment-based materials) is an analysis system to evaluate the long term mechanical behavior in the TRU (TRans Uranium) waste disposal system. TRU wastes are low level radioactive wastes that include long-lived nuclides. MACBECE is the system to calculate mechanical and chemical behavior in the near field including engineered barriers in the TRU waste disposal system, and subsequently to evaluate the hydraulic property in various components. MACBECE can evaluate the mechanical/ hydraulic alterations of two types of barrier materials, cement-based and bentonite-based materials. These materials, in the long-term, may be altered internally by chemical reaction. To evaluate the long-term mechanical behavior due to the chemical transitions, the nonlinear elastic model for the cement-based materials and the elasto-viscoplastic model for the bentonite-based materials were applied, based on the data from various laboratory tests. Also, proposed models based on the data from permeability tests were applied to evaluate the alteration of hydraulic conductivity. So as to realize the high reliable evaluation, the variable-compliance-type constitutive model proposed by Okubo (1992) was newly adopted for the rock creep evaluation in MACBECE. This creep model can express analytically nonlinear visco-elastic behavior and the over-peak-strength state behavior. This paper focuses on the adoptability of rock creep model in MACBECE. Applicability of the rock creep model is verified by conducting the numerical simulations with the long-term mechanical interaction between the TRU waste disposal system and the host rock.
Interdigitated Electrode Array Based Sensors for Monitoring of Caesium
Authors: Nickson Ian D. , Boxall Colin , Port Simon N.
Published: December 2010

The requirement for on-line and in-situ monitoring of analytes in process and effluent streams and in ground waters has become increasingly more important in recent years. We therefore describe the development of the transduction element for a fully automated online instrument for the detection of caesium. The sensor layer for this instrument employs an Ion Selective Conductimetric microsensor (ISCOM) as the detector. This is based upon a plasticized polymeric membrane incorporating a selective ionophore, overlaying two interdigitated microelectrode arrays. A direct relationship has been observed between the bulk conductance (as determined by the microelectrodes) of the ionophore loaded membrane and the concentration of the primary ions in solution. Caesium selective ISCOMs were prepared using a ion selective membrane containing the commercially available ionophore Calix[6]arene-hexaacetic acid hexaethyl ester, polyvinylchloride (PVC) and plasticiser Nitrophenylether (NPOE). Methods of ion selective membrane deposition have been investigated in order to obtain a maximal response. The relative levels of membrane components have also been varied in order to further enhance the ISCOM reponse. We also present preliminary data concerning the caesium selectivity with respect to a range of possible interferents, including rubidium.
The Development of a Method for the Simultaneous Measurement of Cerium (IV) and Chromium (VI) Species in Nitric Acid Media
Authors: Nickson Ian D. , Boxall Colin , Jackson Angela , Whillock Guy O. H.
Published: December 2010

The corrosion of stainless steel in nitric acid media is a major concern for the nuclear industry. Several reprocessing schemes such as PUREX (Plutonium Uranium Reduction Extraction) and UREX (Uranium Reduction Extraction) utilise nitric acid media, and an understanding of the behaviour of key chemical species in these process streams is vital if their effect on associated corrosion reactions and their rates is to be accurately assessed and quantified. This will allow for more accurate prediction of the working lifetime of any stainless steel surface in contact with the process stream in question. Two such key species that are found in nuclear process streams are cerium as Ce(IV) and chromium as Cr(VI), both of which may act as corrosion accelerants. The redox chemistry of cerium and chromium in highly active liquor (HAL) will depend on nitrous acid concentration, temperature, acidity, total nitrate and possibly the influence of other dissolved species and hence an analytical technique for the on-line measurement of these quantities would be useful for lifetime prediction and corrosion prevention. As a result of this, a strategy for the simultaneous measurement of both Ce(IV) and Cr(VI) species in the presence of other ions typically found in process streams (such as Iron, Magnesium Neodymium and Aluminium) has been developed. The work presented will discuss the design and implementation of the electrochemical techniques that we have used in the development of this strategy and in the measurement of the species in question.
Threats and Benefits: Updated Information on Local Opinions Regarding the Spent Nuclear Fuel Repository in Finland
Authors: Kojo Matti , Kari Mika , Litmanen Tapio
Published: December 2010

The aim of the paper is to provide updated information on local opinion regarding the siting of a spent nuclear fuel repository in Finland. The main question is how the residents of the municipality perceive the threats and benefits of the repository. In accordance with the Decision in Principle by the Council of State passed in 2000, the Olkiluoto area in Municipality of Eurajoki was chosen as the location for the repository to accommodate spent nuclear fuel produced in Finland. Updated information on local opinions is needed as the siting process is approaching the next phase, the application for a construction license by 2012. The nuclear waste management company Posiva, owned by the utilities Teollisuuden Voima and Fortum Power and Heat, has also applied for a new Decision in Principle (DiP) for expansion of the repository. The data provided in this paper is based on a survey carried out in June 2008. The respondents were selected from the residents of the municipality of Eurajoki and the neighbouring municipalities using stratified random sampling (N = 3000). The response rate of the survey was 20% (N = 606). The paper is part of a joint research project between the University of Jyva¨skyla¨ and the University of Tampere. The research project “Follow-up research regarding socio-economic effects and communication of final disposal facility of spent nuclear fuel in Eurajoki and its neighbouring municipalities” is funded by the Finnish Research Programme on Nuclear Waste Management (KYT2010).
A Study on the Once-Through Back-End Fuel Cycle Scenario
Authors: Lee Yoon Hee , Song Jongsoon , Lee Jongkuk , Lee Kunjai
Published: December 2010

There are three options for spent fuel management, recycle, once-through and wait and see. The national policy for spent fuel in Korea is “wait and see” and it has to be clearly decided for spent fuel management. The final disposal is the last stage of the fuel cycle and it is essential even though the recycling option will be chosen for spent fuel management policy. And the long-term strategy for spent fuel management considering safety and retrievability is needed. In this study, once-through fuel cycle was focused on for back-end fuel cycle. The international trend for SF management policy and the Korean situation has been investigated. The once-through back-end fuel cycle scenarios has been developed and screened in point of technical and economical aspect. The optimal scenario has been derived by relative comparison and the long-term SF management strategy has been proposed which satisfies both domestic conditions and international trends.
Training Activities and Perspectives in the Radioactive Waste Management Area of Moscow SIA “Radon”
Authors: Batyukhnova O. G. , Arustamov A. E. , Ojovan M. , Dmitriev S. A. , Drace Z. , Arustamova N. A.
Published: December 2010

The education service for specialists dealing with radioactive waste was established in Russia (former USSR) in 1983 and was based on the capabilities of two organisations: Moscow Scientific and Industrial Association «Radon» (SIA “Radon”) and Lomonosov’s Moscow State University. These two organizations are able jointly to offer training programs in the science fundamentals, applied research and in practical operational areas of the all pre-disposal activities of the radioactive waste management (RWM). Since 1997 this system was upgraded to the international level and now acts as the International Education Training Centre (IETC) at SIA “Radon” under the guidance of the IAEA. During last 12 years more than 350 specialists from 33 European and Asian countries enhanced their knowledge and skills in RWM. The IAEA supported many specialized regional training courses and workshops, fellowships, on-the-job training, and scientific visits which are additional means to assure development of personnel capabilities. Efficiency of training was analysed at IETC using the structural adaptation of educational process as well as factors, which have influence on education quality. In addition social-psychological aspects were also taken into account in assessing the overall efficiency. The analysis of the effect of individual factors and the efficiency of education activity were carried out based on appraisal results and post-course questioning of attendees.
Characterization of a Fe-Based Alloy System for an AFCI Metallic Waste Form
Authors: Williamson Mark J. , Sindelar Robert L.
Published: December 2010

The AFCI waste management program aims to provide a minimum volume stable waste form for high level radioactive waste from the various process streams. The AFCI Integrated Waste Management Strategy document has identified a Fe-Zr metallic waste form (MWF) as the baseline alloy for disposal of Tc metal, undissolved solids, and TRUEX fission product wastes. Several candidate alloys have been fabricated using vacuum induction melting to investigate the limits of waste loading as a function of Fe and Zr content. Additional melts have been produced to investigate source material composition. These alloys have been characterized using SEM/EDS and XRD. Phase assemblage and specie partitioning of Re metal (surrogate for Tc) and noble metal FP elements into the phases is reported.
Characteristics on the SAP-Based Wasteform Containing Radioactive Molten Salt Waste
Authors: Park Hwan-Seo , Kim In-Tae , Kim Hwan-Young , Ahn Byeung Gil , Lee Han-Soo
Published: December 2010

This study investigated a unique wasteform containing molten salt wastes which are generated from the pyro-process for the spent fuel treatment. Using a conventional sol-gel process, SiO2-Al2O3-P2O5 (SAP) inorganic material reactive to metal chlorides were prepared. By using this inorganic composite, a monolithic wasteform were sucessfully fabricated via a simple process, reaction at 650°C and sintering at 1100°C. This unique wasteform should be qualified if it meets the requirements for final disposal. For this reasons, this paper characterized its chemical durability, physical properties, morphology and etc. In the SAP, there are three kinds of chains, Si-O-Si as a main chain, Si-O-Al as a side chain and Al-O-P/P-O-P as a reactive chain. Alkali metal chlorides were converted into metal aluminosilicate (LixAlxSi1−xO2−x) and metal phosphate (Li3PO4 and Cs2AlP3O10) while alkali earth and rare earth chlorides were changed into only metal phosphates (Sr5(PO4)3Cl and CePO4). These reaction products were compatible to borosilicate glasses which were functioned as a chemical binder for metal aluminosilicate and a physical binder for metal phosphates. By these phenomena, the wasteform was formed homogenously above μm scale. This would affect the leaching behaviors of each radionuclides or component of binder. The leach rates of Cs and Sr under the PCT-A test condition were about 10−3g/m2day. The physical properties (Cp, k, ρ, Hv, and etc) were very reasonable. Other leaching tests (ISO, MCC-1P) are on-going. From these results, it could be concluded that SAP can be considered as an effective stabilizer on metal chlorides and the method using SAP will give a chance to minimize the waste volume for the final disposal of salt wastes through further researches.
Study of the Radioactive Silt Sediment Cementation Techniques
Authors: Varlakov Andrey P. , Karlin Sergey V. , Barinov Alexandr S. , Zaharova Elena V. , Ermolaev Viatcheslav M.
Published: December 2010

Radioactive silt deposits (RSD) of the spent fuel rod cooling pools were used as an object of these studies. In this paper the following techniques were considered for the silt processing: the cementation of silt without a prior treatment, cementation of the dried silt product, cementation of the calcined silt product, and the impregnation of the calcined silt with high penetrating cement grouts. This paper reports the results of the following studies: the chemical and isotope silt composition, physicochemical silt properties, properties of the cement grouts and solidified compounds, obtained as a result of various silt processing techniques. For every processing technique the following cement compound quality index stated in the Russian standard GOST R 51883-2002 was determined: the mechanical strength, freeze-thaw resistance and leach rate of 137Cs. Effects of variations in the thermal silt treatment mode, in the water/cement ratio, and in the binder material type have been determined. Based upon the results of the studies a comparative assessment of the silt cementation techniques has been performed.
Cementation of Problem LRW Using Porous Concrete
Authors: Varlakov Andrey P. , Germanov Alexandr V. , Dmitriev Sergey A. , Barinov Alexandr S. , Arustamov Artur E.
Published: December 2010

Problem wastes are defined as those wastes which are solidified with difficulty, have poor volumetric efficiencies or produce an unsatisfactory product, demands much of radiation safety of technological process. These wastes are organic liquid (oil, solvents, extractants, scintillation cocktail and others), high containing surface-active substances, subacid aqueous saline and high containing alpha nuclides LRW. The technology consists in as follows. First of all, porous concrete is prepared by a special mixer and placed into a container. The container is the standard drum 100 or 200 l. At the age-hardening of concrete not less than 28 days, when main hydration processes are finished, problem LRW is pumped into the pore space of the porous concrete matrix through previously positioned the special feed tube. The hardened porous concrete matrix is impregnated with problem liquid radioactive wastes which penetrate into the pore space. Depending on kind of LRW contents of waste in the final product has made 50–65% vol. The final conditioning product has meet requirements. LRW is fixed very reliable inside pores of the cement matrix.