bluuebunny/crossref_metadata_2025 huggingface dataset:
Designing a Modular Rapid Manufacturing Process
Authors: Nagel Jacquelyn K. S. , Liou Frank W.
Published: December 2010

Freeform fabrication and additive fabrication technologies have been combined with subtractive processes to achieve a variety of fully integrated rapid manufacturing systems. The combination of separate fabrication techniques into one rapid manufacturing system results in unit manufacturing process integration, sometimes known as a hybrid system. However, the design methods or approaches required to construct these integrated systems are vaguely described or not mentioned at all. The final product from any integrated system is affected not only by the unit manufacturing processes themselves, but also by the extent the individual units are assimilated into an integrated process. A wide variety of integrated and hybrid manufacturing systems and current manufacturing design methodologies are described in this paper, along with their similarities and differences. Through our extensive review, it was discovered that there are five key elements to a reliable integrated rapid manufacturing system: process planning software, motion system, control system, unit manufacturing process, and a finishing process. By studying the manner in which all other systems have been integrated, a table of successful integrated manufacturing system element combinations has been complied, documenting each of the key element choices, resulting in a variety of modular designs. This paper further discusses the importance of the five elements in manufacturing system integration, and how an integrated system is the way to move forward in the manufacturing domain. To that end, a rapid manufacturing system design methodology was developed that explores designs via process analysis to discover integration potential. Cost-benefit analysis is then used to assess the performance of the new system. This analysis determines if all needs have been met, while staying within the constraints of time and resources. Additionally, a table of common issues and obstacles encountered during manufacturing system development has been compiled to assist in the design and development of future rapid manufacturing systems. To illustrate the design methodology, our modular design experience with a laser aided manufacturing process is presented. Unit manufacturing process integration has increased the productivity and capabilities of our system, which reduced resource volume and increased productivity.
Optimization of Stirring Parameters Through Numerical Simulation for the Preparation of Aluminum Matrix Composite by Stir Casting Process
Authors: Su Hai , Gao Wenli , Zhang Hui , Liu Hongbo , Lu Jian , Lu Zheng
Published: December 2010

The flow behavior of the fluid has a significant effect on the particle distribution in the solid-liquid mixing vessel. The stir casting process is generally conducted in a closed crucible, in which the flow pattern is invisible. Therefore, numerical simulation is a forceful tool to guide the experimental research. In the present study, the fluid flow in the stirred crucible during stir casting has been simulated using finite element method. The effects of some important stirring process parameters, such as the blade angle, rotating speed, the diameter of the impeller, and the stirrer geometry, on the flowing characteristics of the molten matrix have been investigated in order to achieve the effective flow pattern to uniformly disperse the ceramic particles in the molten matrix. The simulation results show that the process parameters have significant effects on the flow behavior of the fluid in the stirred crucible. The various combinations of these parameters are beneficial to generate a suitable condition for the composite casting. Further experimental investigation reveals that the present work can provide a guide for the industrial preparation of aluminum matrix composite with a uniform particle reinforcement distribution by stir casting process.
Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code, CUPID
Authors: Cho Hyoung Kyu , Yun Byong Jo , Park Ik Kyu , Jeong Jae Jun
Published: December 2010

For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID. It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model.
Transfer Function Modeling of Zero-Power Dynamics of Circulating Fuel Reactors
Authors: Cammi A. , Di Marcello V. , Guerrieri C. , Luzzi L.
Published: December 2010

In this paper, the zero-power behavior of circulating fuel reactors (CFRs) has been investigated by means of a zero-dimensional neutron kinetics model that provides a simplified but useful approach to the simulation of the dynamics of this class of nuclear reactors. Among CFRs, the most promising is the molten salt reactor (MSR), which is one of the six innovative concepts of reactor proposed by the “Generation IV International Forum” for future nuclear energy supply. One of the key features of CFRs is represented by the fission material, which is dissolved in a liquid mixture that serves both as fuel and coolant. This causes a relevant coupling between neutronics and thermo-hydrodynamics, so that fuel velocity plays a relevant role in determining the dynamic performance of such systems. In the present study, a preliminary model has been developed that is based on the zero-power kinetics equations (i.e., reactivity feedbacks due to temperature change are neglected), modified in order to take into account the effects of the molten salt circulation on the drift of delayed neutron precursors. The system dynamic behavior has been analyzed using the theory of linear systems, and the transfer functions of the neutron density with respect to both reactivity and fuel velocity have been calculated. The developed model has been assessed on the basis of the available experimental data from the molten salt reactor experiment (MSRE) provided by the Oak Ridge National Laboratory. The results of the present work show that the developed simplified theoretical model is well descriptive of the MSRE zero-power dynamics, allowing a preliminary evaluation of the effects due to the circulation of the fuel salt on the neutronics of the system. Moreover, the model is of general validity for any kind of CFRs, and hence is applicable to study other MSR concepts in order to have some indications on the control strategy to be adopted in the MSR development envisaged by Generation IV.
Three-Dimensional Thermohydrodynamic Analysis of a Wet Clutch With Consideration of Grooved Friction Surfaces
Authors: Jang J. Y. , Khonsari M. M. , Maki Rikard
Published: December 2010

A model is developed to investigate the effect of radial grooves and waffle-shape grooves on the performance of a wet clutch. Three-dimensional formulation of the governing equations, boundary conditions, and numerical solution scheme are presented for modeling the thermal aspects of the engagement process in a wet clutch. The thermal model includes full consideration of the viscous heat dissipation in the fluid as well as heat transfer into the separator, the friction material, and the core disk. The convective terms in the energy equations for the oil as well as the heat conduction equations in the bounding solids are properly formulated to determine the temperature fields corresponding to the domains between grooves. Roughness, centrifugal force, deformability, and permeability of the friction material with grooves are taken into account. The effects of groove geometry such as groove depth, grooved area, and number of grooves on the engagement characteristic of a wet clutch are investigated. It is also shown that the thermal effects in a wet clutch influence the engagement time and the torque response and should be included in the analytical studies.
Performance of an Absorber With Hydrophobic Membrane Contactor at Aqueous Solution-Water Vapor Interface
Authors: Ali Ahmed Hamza H.
Published: December 2010

In this study, a detailed modeling of the heat and mass transfer processes inside a plate-and-frame absorber with hydrophobic microporous membrane contactor at aqueous solution-water vapor interface as a part of a chiller model is developed. The absorber is a component of a 5 kW cooling capacity single effect lithium bromide-water absorption chiller with a hot water thermally driven generator, a water-cooled absorber, and a condenser. The model is used to investigate the performance of the absorber in case the chiller operates at different values of the inlet driving hot water and cooling water (coolant) temperatures. The results clearly indicate that for the same cooling capacity of the chiller and compared with the performance at the design point value, increasing the inlet driving hot water temperature results in an increase in the required absorber size and consequently a decrease in the absorber performance, while decreasing the cooling water (coolant) inlet temperature leads to slight decreases in the required absorber size and consequently an increase in the absorber performance. The effect is prominent and can be used to decrease the absorber size for chillers work in places where the option of lower inlet coolant temperature is available with normal driving hot water temperature.
Electrogalvanized Low Carbon Steel Adhesion Tendency in Friction Processes Under Mixed Lubrication Regime
Authors: Miguel V. , Coello J. , Manjabacas M. C. , Calatayud A. , Ferrer C. , Martínez A.
Published: December 2010

Although some authors work at times with large flat dies and evaluate friction under more realistic conditions than usual, pressure is not totally controlled. In any case, cohesive friction does not appear to have been well studied in literature, but pressure and sliding velocity may provide useful information about preventing the cohesive phenomenon in sheet stamping processes. In this work, the coefficient of friction (COF) for DC-05 electrogalvanized steel is experimentally evaluated under lubrication regime by flat face dies. These tests are also considered to reproduce friction conditions in the die-sheet-blankholder system at some stages of the deep drawing process. High pressure condition in a flat friction system can also be considered for studying the friction behavior in the die radius. This work investigates the influence of contact pressure and sliding velocity of the sheet on the COF value. Adhesion tendency during sliding is also evaluated. Sheets were lubricated with a prelube type mineral oil and different lubricant film thicknesses are present on the sheet as a result of the draining off time effect, an aspect that will be evaluated later. Although sliding velocity has almost no influence on the COF value, pressure has an influence that may be expressed by a potential mathematical function. The COF value tends to be constant for high enough pressure values. This behavior may be explained, in part, from the viewpoint of zinc acting as a typical soft metallic lubricant. Sliding velocity is the most important variable from the adhesion phenomenon point of view, which appears more frequently for low velocity values. The draining off time, which some research works consider fixes the initial lubrication conditions in friction tests, has no significant effect when a mineral oil, typically used as a prelube, is selected as a lubricant. The authors found that pressure is the most important variable for the COF value. Velocity is the determining factor for the adhesion phenomenon in friction processes under mixed lubrication.
Enhancement of Adaptability of Parallel Kinematic Machines With an Adjustable Platform
Authors: Bi Z. M. , Kang B.
Published: December 2010

The design of an industrial robot involves a number of conflicting objectives in general. A parallel kinematic machine (PKM) is known to achieve high precision and heavy load capacity but with the sacrifice of large workspace and high dexterity. Therefore, existing PKMs are mostly dedicated to a specific task with relatively poor adaptability to task variations. In this paper, the concept of an adjustable platform is proposed to enhance the adaptability of a PKM for various tasks. It is demonstrated that the adjustment of the dimensions of a base platform or end-effector platform has a significant impact on the performance of a PKM including its workspace and overall stiffness distribution. Both offline and online adjustment modes are presented. The offline adjustment brings a new dimension for reconfigurability and thus increases the versatility of a parallel robot for different tasks. The online adjustment turns a parallel robot into a redundant PKM, and therefore its overall performance against task requirements can be improved. A planar PKM and a Stewart robot with an adjustable platform are used as case studies in order to demonstrate the advantages of the proposed concept.
A Practical Approach to Characterise and Assess a Site Drainage System in Support of Site Restoration
Authors: Bartlett Angela , Coppins Gavin , Burgess Peter
Published: December 2010

Part of the nuclear site restoration and delicensing process involves the characterisation and assessment of below-ground drainage systems. Site restoration is currently underway at Harwell in Oxfordshire where there is a complex drainage system that has developed over more than 70 years. Drainage decommissioning involves visual inspections, jet-washing and radiological surveys prior to final grouting. Prior to decommissioning, the structural and radiological condition of the network was unknown and now requires characterisation and assessment against defined clean-up criteria before the land can be released for future use. This paper outlines the application of an innovative Geographic Information System (GIS) and data management methodology to assess the structural and radiological condition of the drainage network at Harwell. The approach demonstrates the importance of efficient data collection and storage, implemented using UKAEA’s IMAGES land quality data management system. It also details several GIS techniques that can be utilised to accurately position below-ground surveys and record pipe material and diameter for surveyed drain sections. The aim of the monitoring process is to identify any activity which exceeds the delicensing criteria set by the Nuclear Installations Inspectorate (NII). The relevant criteria are the residual activity and the risk to a future site occupier. The maximum missable activity for each pipe class was determined by considering instrumentation characteristics and natural background radiation levels. The equipment used (a caesium iodide scintillation probe connected to a ratemeter), was limited by the small diameter of some drains. The hypothesis chosen was that residual activity was most likely to lie along the base of the pipe, given that the drains generally ran with only a limited depth. The response was measured for 137Cs gamma radiation and then used to calculate the net count rate for a variety of pipe configurations. Variability of the data was investigated in relation to static probe response and variations in counts per second along a pipe length. Analysis of the data using GIS showed clear spatial patterns, indicating systematic variability within each pipe material and diameter class. Clearance criteria were derived from this analysis and will be used to establish whether the specific sections of drain can remain in situ or require removal. This approach improves safety and cost performance of the decommissioning process and ensures that a transparent and auditable record of the status of the drainage system remains for the future re-use of the land as the Harwell Science and Innovation Campus.
The Use of Radiological Characterisation in Support of the Design and Build of a New Facility in an Area of Elevated Dose Rate
Authors: Hughes Karl , Thornley David , Pienkowski Czeslaw
Published: December 2010

A new building, a maintenance facility, is to be constructed in the ‘separation area’ of the Sellafield Site. Sellafield is a complex and busy nuclear facility covering about two square miles in the north-west of the United Kingdom. The facility, which is to provide necessary storage and maintenance functionality to support bulk waste retrievals from legacy silos, is being built in close proximity to spent fuel storage ponds, intermediate level legacy waste stores and a pipebridge carrying active materials. In addition, the site is adjacent to a main pedestrian and vehicle thoroughfare and a railway line used for the regular transfer of nuclear materials. The facility itself is to be built on the site of a recently demolished active facility. The ambient gamma dose rates in the construction area have been measured to be typically 5 to 50 μSv/hr. Although there are known specific sources of dose rate close to the construction site, the relative contributions of each within the envelope of the new facility are unknown. This paper describes how a series of gamma dose rate measurements, gamma spectroscopy measurements and gamma-ray imaging surveys, using RadScan® 800, have been used to better understand the origins of the dose rates at a number of key locations within the area. The results of the survey are being used to assess the requirements for shielding within the structure of the facility in order to reduce the dose to personnel that will work there when it is operational. The results are also being used to identify whether any localised shielding would prove beneficial for reduction of the dose both during construction and occupancy. The unique mix of facilities surrounding the site have meant that the contributions to the dose rate come not only from adjacent facilities in the form of unscattered, higher energy, penetrating radiation, but also in the form of lower energy scattered radiation, both from radiation that has passed through shielding and also in the form of skyshine. Skyshine is the term for the radiation that has escaped a facility and ‘bounced’ off the sky to return to earth. The radiometric measurements, which were collected over a period of a few days, were able to provide a useful understanding of each of these sources of dose, and to quantify the relative contributions of each.
A Systematic Planning Tool for Environmental Characterisation
Authors: Wilcox Steven , Wilkins Richard , Lyons Martin
Published: December 2010

Many organisations are currently dealing with long standing legacy issues in clean up, decommissioning and demolition projects. Industry is required to ensure that all bulk articles, substances and waste arisings are adequately characterised and assigned to the correct disposal routes in compliance with UK legislation and best practice. It is essential that data used to support waste sentencing is of the correct type, quality and quantity, and that it is appropriately assessed in order to support defensible, confident decisions that account for inherent uncertainties. AMEC has adopted the Data Quality Objectives (DQO) based methodology and the software package Visual Sample Plan (VSP) to provide a better, faster, and more cost effective approach to meeting regulatory and client requirements, whilst minimising the time spent gathering data and assessing the information. The DQO methodology is based on a scientific approach that requires clear objectives to be established from the outset of a project and that there is a demonstration of acceptability of the results. Through systematic planning, the team develops acceptance or performance criteria for the quality of the data collected and for the confidence in the final decision. The systematic planning process promotes communication between all departments and individuals involved in the decision-making process thus the planning phase gives an open and unambiguous method to support the decisions and enables the decision-makers (technical authorities on the materials of concern) to document all assumptions. The DQO process allows better planning, control and understanding of all the issues. All types of waste can be sentenced under one controllable system providing a more defensible position. This paper will explain that the DQO process consists of seven main steps that lead to a detailed Sampling and Analysis Plan (SAP). The process gives transparency to any assumptions made about the site or material being characterised and identifies individuals involved. The associated calculation effort is reduced using the statistically based sampling models produced with VSP. The first part of this paper explains the DQO based methodology and Visual Sample Plan and the second part shows how the DQO process has been applied in practice.
Transport of the Reactor Pressure Vessels in the Greifswald Nuclear Power Plant
Authors: Borchardt Ralf
Published: December 2010

Five WWER-440 reactors are being dismantled on the Greifswald Nuclear Power Plant (KGR) site. The strategy for the dismantling of the reactor units 1 to 4 (operation time 12–17 years) was to cut and pack the components remotely. For this purpose dry and wet cutting areas were installed. The remote cutting and packing of the reactor pressure vessel and its internals was successfully tested with non-activated original reactor components of units 7 and 8 from October 1999 until July 2003. From August 2004 until July 2007 the internals from reactor units 1 and 2 were cut, packed and transported to the on-site Interim Store North (ISN). For the reactor 5 it was planned to transport the RPV in one piece and the reactor internals in shielding and transport containers to the interim store for decay storage. In December 2003 the RPV of unit 5 was lifted and transported to the interim store. From April 2006 up to July 2006 the reactor internals of unit 5 were packed and transported to the interim store. After the evaluation of the experience made during the transport and the radiological measurements and samplings taken from the RPV unit 1, the strategy for the dismantling of the reactors was changed. The reactor pressure vessels of the units 1 to 4 and the reactor internals of the units 3 and 4 should be removed as complete parts and stored as shielded large components in the ISN. In summer 2005 EWN applied for the new strategy at the responsible licensing authority and in August 2007 this license was granted. In November 2007 the reactor pressure vessels of the units 1 and 2 were transported into the ISN. The transport of the reactor pressure vessels and the internals from units 3 and 4 is planned in the period from March till September 2009. These transports of the reactor pressure vessels and internals show that the dismantling of the reactors with dismantling and interim storage of large components could not only be an alternative for cutting but could also be favored from the economical and radiological point of view.
Decommissioning Strategies Worldwide: A Re-Visited Overview of Relevant Factors
Authors: Laraia Michele
Published: December 2010

This paper highlights current trends and developments in selecting decommissioning strategies worldwide. Radiological conditions, spent fuel and radioactive waste management, funding, economics and the development of suitable technology are some common factors for taking decisions on the timing and circumstances of decommissioning. Although safe enclosure is the default option for many shut down facilities typically due to lack of ready cash, delay in dismantling may have serious disadvantages such as loss of expertise and long term uncertainties. Currently, of the many large nuclear installations permanently shut down, only a fraction have been or will be in the near term totally dismantled and decommissioned to unrestricted release state. A trend towards immediate dismantling seems to emerge in some countries, and is supported by IAEA positions, but this appears to be due to country-, site- or plant-specific conditions of limited generic applicability. In recent years, and often as the result of international efforts, the situation is evolving, and provisions and infrastructures including funding are being established to cope with decommissioning challenges. This factor seems in principle to encourage immediate, total dismantling. However, the worldwide overview of decommissioning strategies does not offer a clear pattern. New factors have come into being, such as stakeholder opinions, in particular those of local communities, and now play a significant role in decision-making. The conditions of the nuclear industry at large (e.g. the “nuclear renaissance”) considerably changed over the last few years and are going to affect decommissioning in the near future. Strategies such as restricted release (brownfields), incremental decommissioning or entombment seem to offer new prospects. The author reviewed first the worldwide situation around the year 2000, and offers in this paper some reflections about changed world’s conditions and how these affect the decommissioning scenarios.
Progress and Experiences From the Decommissioning of the Eurochemic Reprocessing Plant
Authors: Walthe´ry Robert , Lewandowski Patrick , Ooms Bart , Reusen Nancy , Van Laer Wim
Published: December 2010

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 has been divided into three parts — each part is isolated from the others. In September 2008 the eastern part of the building has been 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 strategy 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.
Developing Design Premises for a KBS-3V Repository Based on Results From the Safety Assessment
Authors: Andersson Johan , Hedin Allan
Published: December 2010

As a part of the planned license application for a final repository for spent nuclear fuel the Swedish Nuclear Fuel and Waste Management Co. (SKB), has developed design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel. The purpose is to provide requirements from a long term safety aspect, to form the basis for the development of the reference design of the repository and to justify that design. Design premises typically concern specification on what mechanical loads the barriers must withstand, restrictions on the composition of barrier materials or acceptance criteria for the various underground excavations. These design constraints, if all fulfilled by the actual design, should form a good basis for demonstrating repository safety. The justification for these design premises is derived from SKB’s most recent safety assessment SR-Can complemented by a few additional analyses. Some of the design premises may be modified in future stages of SKB’s program, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety.
Lessons Learned From the Operation of a LILW National Disposal Centre: The Cabril and the Spanish Case
Authors: Navarro Mariano , Go´mez Fernando , Garci´a Emilio
Published: December 2010

Spain occupies a relatively important position in experience in the field of LILW management. The management of LILW in Spain may be defined as an integrated system encompassing the entire spectrum; from production controls, removal and transport, to disposal. In this system, a clear definition of the responsibilities of each of the people involved plays a fundamental role. ENRESA, the organization in charge of radioactive waste management in Spain, has been operating the El Cabril LILW disposal facility since 1992, this installation being a key component in the national LILW management programme. Over the years ENRESA has acquired significant operating experience from a multi-disciplinary point of view, including technical, economic and social aspects. To date, since the design phase of the facility and over more than fifteen years, ENRESA has adopted a series of decisions and has undertaken programmes and activities that have allowed the installation to evolve into the reality that it now is. The aim of this paper is to present the lessons learned from a strategic point of view and in relation to the most relevant and significant aspects that have facilitated the normal operation of the facility, and the development of specific solutions to the challenges posed by the performance of activities, and the emerging needs of the Spanish programme.
Stakeholder Participation for the Legacy Ponds and Legacy Silos (LP&LS) Facility at Sellafield, Cumbria, UK: The Nature and Effectiveness of the Dialogue
Authors: Whitton John
Published: December 2010

The Legacy Ponds and Silos (LP&S) facilities are part of the UK nuclear legacy located at the Sellafield Site, Cumbria. There are four individual facilities containing nuclear wastes that have accumulated over a period of approximately 50 years. Waste retrieval and conditioning, in preparation for decommissioning, is currently being carried out by the site operator. LP&S have recently proposed a re-engagement with stakeholders following the initial engagement in December 2005. This paper reviews this earlier engagement in terms of the nature of dialogue that was carried out when compared against definitions of deliberation provided in the literature. The aim of this paper is to provide those planning future engagement with a better understanding of how the nature of dialogue can vary and uses participation and deliberation as indicators of effective engagement. A concern of those working towards a programme of effective stakeholder participation in 2005 was how to ensure genuine dialogue and stakeholder representation in such a strictly controlled and regulated environment with a technical complexity that challenges technical specialist and layperson alike. LP&S recognised that effective dialogue with stakeholders on the available technical options and their associated societal impacts would form a significant part of this process if options were to prove resilient. However, the challenge presented to LP&S was how to engage stakeholders on a variety of projects, whilst ensuring the output could be used by the projects as part of their technical decision making. Initial contact was made with stakeholders in December 2005, as part of a ‘Baseline’ Best Practicable Environmental Option (BPEO) study. A BPEO study leads to decisions on waste management based on an integrated assessment of alternative options, on the basis of factors such as the occupational and environmental risks, the environmental impacts, costs and social implications of the options. An engagement framework was also developed by LP&S to understand the engagement within the context of that taking place nationally and internationally. LP&S plan to reconvene stakeholders as part of the commitment made to them in 2006 to review progress on decommissioning projects. Currently, the strategy by which this will be achieved is unclear, however it is proposed that rather than full engagement, LP&S will provide a series of information updates to the West Cumbria Site Stakeholder Group (WCSSG). Representing local interests only and potentially reducing the opportunity for deliberation, further research is recommended to assess the nature of dialogue and the role of deliberation in this type of forum.
Regulatory Review of Preliminary Safety Assessment for the Baˇit¸a-Bihor Repository, Romania
Authors: Biurrun Enrique , Haverkamp Bernt , Ro¨hlig Klaus-Ju¨rgen
Published: December 2010

In the framework of a PHARE project DBE TECHNOLOGY GmbH carried out a regulatory review of the Preliminary Safety Assessment Report (PSAR) for the near surface LILW repository Baˇit¸a-Bihor, Romania. During the review process several shortcomings of the PSAR from the regulatory point of view were identified. Main findings concerned shortcomings in regard to the operational safety, especially in the field of mining safety. Other points of criticism were connected with the post-closure part of the PSAR. From the regulatory point of view, the long-term calculations appeared to lack a sufficient level of conservatism and the necessary clarity and traceability that would be required to allow a proper evaluation of the concept used for calculating the long-term radionuclide migration and the particular results of the PSAR. Respective recommendations were drafted for the Romanian regulator about the changes that should be implemented in the next version of the Safety Assessment Report. Another focus of this project was an investigation on the potential implementation of the Hydraulic Cage Concept for Baˇit¸a-Bihor. This concept, which originally was developed by DBE TECHNOLOGY GmbH for implementation at the Richard Repository, Czech Republic, helped to decrease the potential future radiological impact for the Richard repository. As the geological situation and the condition of the waste packages are similar for the two repositories, the Hydraulic Cage Concept or some adaptation of it seemed to be an obvious choice to improve the safety of the Baˇit¸a-Bihor repository should that turn out to be necessary. During execution of the review, it became clear that without prejudging the outcome of the next version of the Safety Assessment Report it is likely that some kind of measure will be needed to reduce the potential radiological impact. According to the outcome of the project, an adapted simplified Hydraulic Cage System would be a significant improvement to limit the potential future radiological impact.
Waste Reduction by Re-Use of Low Activated Material
Authors: Ehrlicher Ulrich , Pauli Heinz
Published: December 2010

A multidisciplinary institute, equipped with research reactors and accelerator-driven research installations produces and, in the case of PSI, collects radioactive waste on one hand and requires material, especially for shielding purpose, on the other hand. The legislative framework for radiation protection, financial reasons and limited storage capacity strongly force Paul Scherrer Institute and comparable facilities to minimize radioactive waste. Besides free release of inactive components, recycling and re-use of low-level radioactive material in controlled areas are the best means for waste minimization. The re-use of slightly activated steel plates as a shielding material and the recycling of irradiated reactor graphite as a filling material embedded in mortar may give examples and encouragement for similar activities. Besides the advantages for radiation protection, the financial benefit can be measured in millions of dollars.
Implementation and Ongoing Development of a Comprehensive Program to Deal With Canada’s Nuclear Legacy Liabilities
Authors: Metcalfe Douglas , Yuen Pui Wai , McCauley Dave , Brooks Sheila , Miller Joan , Stephens Michael
Published: December 2010

Nuclear legacy liabilities have resulted from 60 years of nuclear research and development carried out on behalf of Canada by the National Research Council (1944 to 1952) and Atomic Energy of Canada Limited (AECL, 1952 to present). These liabilities are located at AECL research and prototype reactor sites, and consist of shutdown reactors, research facilities and associated infrastructure, a wide variety of buried and stored waste, and contaminated lands. In 2006, the Government of Canada adopted a new long-term strategy to deal with the nuclear legacy liabilities and initiated a five-year, $520 million (Canadian dollars) start-up phase, thereby creating the Nuclear Legacy Liabilities Program (NLLP). The objective of the long-term strategy is to safely and cost-effectively reduce risks and liabilities based on sound waste management and environmental principles in the best interests of Canadians. The five-year plan is directed at addressing health, safety and environmental priorities, accelerating the decontamination and demolition of shutdown buildings, and laying the groundwork for future phases of the strategy. It also includes public consultation to inform the further development of the strategy and provides for continued care and maintenance activities at the sites. The NLLP is being implemented through a Memorandum of Understanding between Natural Resources Canada (NRCan) and AECL whereby NRCan is responsible for policy direction and oversight, including control of funding, and AECL is responsible for carrying out the work and holding and administering all licences, facilities and lands. The paper summarizes achievements during the first three years of program implementation in the areas of decommissioning and dismantling; waste recovery and environmental restoration; the construction of enabling facilities to analyze, handle and store the legacy waste; and, planning for the long-term management of the radioactive waste.
Treatment of Core Components From Nuclear Power Plants With PWR and BWR Reactors
Authors: Viermann Jo¨rg , Friske Andreas , Radzuweit Jo¨rg
Published: December 2010

During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK® cask. The MOSAIK® family of casks is subject of a separate presentation at the ICEM 09 conference.
Building 18: Operating Feedback From Cleaning and Dismantling of Glove Boxes and Shielding Lines
Authors: Jeanjacques Michel , Bremond Marie Pierre , Gautier Laurent , Vieillard Guy , Estivie´ David , Pichereau Eric
Published: December 2010

This summary presents the clean-up and dismantling works on glove boxes and shielded lines in building 18. The Atomic Energy Commission’s centre at Fontenay Aux Roses (CEA-FAR) was created in 1946. A radiochemical laboratory devoted to R&D programs focused on studies of irradiated fuel reprocessing processes, waste treatment processes and studies and production of transuranic elements was built in the early 1960s. The R&D program was finally halted on 30 June 1995. The purpose of the building 18 clean-up program is to minimise the nuclear and traditional hazards and to reduce as much as possible the production of high- and medium-level waste during subsequent dismantling work. The paper will deal with a short description of the operations as well as the lessons learned and the feedback experience which could be of any help in on going or planned DD&R projects.
Cleansing and Dismantling of CEA-Saclay Nuclear Licensed Facilities
Authors: Jeanjacques Michel , Gle´varec Rebecca , Delaire Isabelle
Published: December 2010

This summary presents the cleansing and dismantling operations currently realized on the CEA center of Saclay (CEA-Saclay). It was initiated at the beginning of the 2000 years a cleansing and dismantling program of the old Nuclear Licensed Facilities (NLF). Currently this program relates to the Hot Laboratories (Laboratoires de Haute Activite´: LHA) and the old workshops of the Liquid Waste Treatment Plant (Station des Effluents Liquides: STEL).
Developing Ceramic Based Technology for the Immobilisation of Waste on the Sellafield Site
Authors: Scales C. R. , Maddrell E. R. , Dowson Mark
Published: December 2010

National Nuclear Laboratory, in collaboration with the Australian Nuclear Science and Technology Organisation, is developing hot isostatic press (HIP) based ceramic technology for the immobilisation of a diverse range of wastes arising from nuclear fuel processing activities on the Sellafield site. Wasteform compositions have been identified and validated for the immobilisation of these plutonium containing wastes and residues in glass-ceramic and ceramic forms. A full scale inactive facility has been constructed at NNL’s Workington Laboratory to support the demonstration of the technology. Validation of the inactive wasteform development using plutonium has been carried out at ANSTO’s Lucas Heights facility. A feasibility study has been conducted to evaluate the construction and operation of a plutonium active pilot facility which would demonstrate the immobilisation of actual residues in the NNL Central Lab. This could form the basis of a facility to treat the plutonium wastes and residues in their entirety. The technology is being explored for the immobilisation of additional wastes arising on the Sellafield site taking advantage of the investment already made in skills and facilities.
Decommissioning of Buildings 105X and 122X at Belgoprocess
Authors: Ooms Bart , Lievens Bert , Van Laer Wim , Van Campfort Etienne , Walthe´ry Robert
Published: December 2010

The Eurochemic reprocessing plant was built between 1960 and 1966 and operated from 1966 until the end of 1974. During these eight years of active operation, Eurochemic reprocessed 181.5 t of natural and slightly enriched uranium fuels (less than 4.5% initial 235U enrichment) from various experimental and power reactors, and 30.6 t high enriched uranium fuels from testing reactors, generating approximately 50 m3 of high-level liquid waste from power reactor fuels (LEWC, low enriched waste concentrate ) and 850 m3 from research reactor fuels (HEWC, high enriched waste concentrate). As a result of reprocessing and cleaning operations (1975–1981), generated intermediate and high level wastes were put into temporary storage, pending the availability of appropriate treatment, conditioning and storage facilities. Immediately after LEWC and HEWC vitrification, the corresponding storage vessels were rinsed and decontaminated. The rinsing and decontamination program started in April 1986 and was interrupted between September 1987 and July 1989 in view of possibly reusing the vessels for storage of similar HLLW (high level liquid waste) solutions. Because the storage building itself was not aircraft crash resistant, it was decided not to use the storage vessels anymore and to proceed the decontamination with more aggressive chemicals. Due to this time gap however, and especially because vitrification came to an end in September 1991, a considerable volume of decontamination liquids was produced after this time and stored, pending the availability of the bituminization installation. In 2005 and 2006 a research program was performed. For both buildings and vessels images and samples were collected and dose rate measurements were executed. The paper presents an overview of the different studies that were indispensable in order to be able to select the most appropriate decommissioning strategy.
On Fluidization of Borosilicate Glasses in Intense Radiation Fields
Authors: Ojovan Michael , Mo¨bus Guenter , Tsai Jim , Cook Stuart , Yang Guang
Published: December 2010

The viscosity is rate-limiting for many processes in glassy materials such as homogenisation and crystallisation. Changes in the viscous flow behaviour in conditions of long-term irradiation are of particular interest for glassy materials used in nuclear installations as well as for nuclear waste immobilising glasses. We analyse the viscous flow behaviour of oxide amorphous materials in conditions of electron-irradiation using the congruent bond lattice model of oxide materials accounting for the flow-mediating role of broken bonds termed configurons. An explicit equation of viscosity was obtained which is in agreement with experimental data for non-irradiated glasses and shows for irradiated glasses, first, a significant decrease of viscosity, and, second, a stepwise reduction of the activation energy of flow. An equation for glass-transition temperature was derived which shows that irradiated glasses have lower glass transition temperatures. Intensive electron irradiation of glasses causes their fluidisation due to non-thermal bond breaking and can occur below the glass transition temperature. Due to surface tension forces fluidisation of glasses at enough high electron flux densities can result in modification of nano-size volumes and particles such as those experimentally observed under TEM electron beams.
Tailoring Engineering Activities to D&D Projects
Authors: Negin Charles A. , Urland Charles S. , Szilagyi Andrew P. , Collazo Yvette T. , Santos Joseph K. , Gladden John B.
Published: December 2010

Engineering is an important element of Deactivation and Decommissioning (D&D) project technical planning, scheduling, estimating, and execution. Understanding the scope of engineering and related design, deciding when in a project’s schedule these activities should be conducted, and specifying the products to be generated from each engineering task are important management functions. These subjects are addressed in a guidance report developed by the U.S. Department of Energy (DOE) described in this paper.