bluuebunny/crossref_metadata_2025 huggingface dataset:
The Regulation of Legacy Situations Where Reduction in Long Term Radiation Hazard Requires an Increase in Short Term Detriments
Authors: Fisher Steve , Turton David
Published: March 2009

Legacy situations on nuclear sites usually occur at the end of the life of a facility or site. These situations, such as decommissioning or historic waste treatment, are often responsible for long term hazards to workers and the public. Reducing the magnitude of these long-term hazards will generate waste disposals which may impact on the environment and may result in workers incurring additional doses. The increase in these detriments seems at first glance to be in conflict with the usual aims of the regulators. However by putting the reduction of long term radiation hazard from legacy situations in the context of Government policy and guidance, the approach of current regulatory practice adopted for such situations is considered. The regulatory framework is presented from the high level policy, then through the application of the principles of ALARA, reducing risks — protecting people, Best Practicable Environmental Option (BPEO) and Best Practicable Means (BPM). Issues that the regulators expect the nuclear site operator to address are discussed from both the protection of the public and the environment and examples of the practical implications of the regulatory approach are described. ‘Softer’, but important essentials for the operator to adopt in the handling of legacy situations are also raised. These essentials such as openness and stakeholder dialogue, have in the past been poorly performed leading to a lack of trust and understanding by the public.
Instrumental Photon Activation and Applications in a Nuclear-Waste Inspection Purpose
Authors: Jeanneau F. , Gmar M. , Huot N. , Laine´ F. , Lyoussi A. , Payan E. , Pillot Ph. , Roux L. , Saurel N.
Published: March 2009

The development of non-destructive methods to inspect nuclear-waste containers is important for radioactive-waste management and non-proliferation purposes. This paper will present studies and results carried out by a method based on photon interrogation (photofission) which allows the determination of the actinide quantity contained in the waste. High-energy photons (produced by an electron accelerator associated with a Bremsstrahlung tungsten target) will induce photofission reactions on the actinides. Then the flux of delayed neutrons, which is directly proportional to the amount of actinides, is measured with 3He detectors. Since the beginning of 1990’s, our team in CEA has been working on the development of this method and the improvement of the existing simulation code. The two main tools will be introduced: OPERA (tool for the simulation of photonuclear reactions) which includes photonuclear cross sections in a Monte-Carlo code based on MCNP4C, and SAPHIR (Irradiation and Photon-Activation System), a device allowing experimentations for research and development programs. The applications of these tools will be illustrated mainly with two examples: 1) The feasibility study of an inspection device for old concrete containers will be reported. Two campaigns of measurements have been performed in order to determine the sensitivity and the detection limits in the case of four different types of concrete containers, in terms of nature and geometry. 2) Nuclear-waste producers and managers have been interested by the active photon interrogation possibilities to measure actinide quantity in wastes of high activity, vitrified or compacted, with constraints like a dose rate around 400 Gy/h at 27 cm from the container. The simulation-code improvement has allowed some calculations, based on the SAPHIR facility, which have shown a good linearity between the actinide mass and the number of detected neutrons, in spite of a very high passive noise and the presence of a lead protection. Several R&D programs will be also presented. On one hand, measurements are performed on real wastes, chosen for parameter which could define a limitation of the measurements, in order to improve the method and to evaluate the detection limits. For instance, tomography can be performed with this experimental device: quantity and position of actinides in the waste can be calculated. On the other hand, a new method is studied, using the delayed-gamma flux in order to quantify and to identify the different actinide isotopes contained in the waste. These methods and device offer a large panel of results in terms of measurements and simulations. Our team is now involved in several prospecting and R&D programs in order to improve the current method and to find some new applications for nuclear-waste management.
Results of the Full Scale Testing of the Remote Dismantling in Greifswald NPP
Authors: Borchardt Ralf , Raasch Ju¨rgen
Published: March 2009

For the dismantling of the pressure vessels and internals of the Greifwald WWER units 1–4, remote dismantling procedures are necessary. In order to realise this, 3 different cutting areas are used, i.e.: a wet cutting pool for highly activated parts, a dry cutting area for less activated parts and a specific cutting area for the annular water tank. Techniques and procedures were tested in a mock-up installation, with original components from the not operated reactor units 7 and 8. Special emphasis was given to evaluate the expected radiation exposure of the personnel, the maximum application time of the tools and the minimization of the secondary waste. The complete testing program was executed from October 1999 to September 2002. The authorised expert was involved in all steps of the testing. Thus a firm basis for the licensing and the execution of the remote dismantling could be established. Based on the results of the model dismantling and the selection of tools and procedures, the times necessary for the different operation steps could be calculated. With these figures, the manpower and the ambient dose rate it was possible to calculate the time need for the dismantling and also the expected dose commitment. This information was mandatory for the project planning and the preparation of the licensing documents. Presently the implantation of components and systems are going on in unit 2 and the start of cutting of the first activated pressure vessel is scheduled for end of 2003.
Stabilization of Underground Solvent Storage Tanks
Authors: Smail Timothy R. , Herb Annamarie M. , Hall Monica C.
Published: March 2009

The Old Solvent Tanks (OST), located at the Savannah River Site (SRS) Old Radioactive Waster Burial Ground (ORWBG), are comprised of 22 underground storage tanks that were used to store spent radioactive solvent and aqueous wastes generated from the plutonium-uranium extraction (PUREX) process. The OSTs were installed at various dates between 1955 and 1968 and used to store the spent solvents until 1974. The spent solvents stored in the OSTs were transferred out from 1976 through 1981 leaving only residual liquids and sludges that could not be pumped out. Final remediation goals for the ORWBG include an overlying infiltration control system. If the tanks were to structurally fail, they would collapse causing potential for onsite worker exposure and release of tank contents to the environment. Therefore, as an interim action, methods for stabilizing the tanks were evaluated. The preferred remedial action was “Grouting of the Tank Wastes In-situ.” The primary function of the grout is to provide structural stability of the tanks by filling void space with material that prevents tank collapse. Incidental to any mixing that may occur, residual material in the tanks will be incorporated into the grouting mixture. The incidental grouting will ultimately improve environmental protection by rendering the residual material immobile. To accomplish this task, the SRS Environmental Restoration Division (ERD) teamed with the Savannah River Technology Center (SRTC) to determine a remedial design strategy and to translate this strategy into a construction specification and drawings for implementation. The OST remedial design strategy contained the following key aspects for performance requirements and acceptance criteria: • Grout mix; • Tank atmosphere testing; • Grout delivery system and camera monitoring system; • Off-Gas HEPA filter system and environmental monitoring; • OST Sealing and labeling. From November 2001 through February 2003 all 22 Old Solvent Tanks were successfully stabilized. This paper will discuss the systems designed to perform and monitor the grouting operation, the grouting process, and the radiological controls and wastes associated with grouting the Old Solvent Tanks.
Non-Oxidative Destruction of TNT, RDX, and HMX on Contaminated Soil Using Subcritical (Hot/Liquid) Water
Authors: Hawthorne Steven B. , Lagadec Arnaud J. M. , Miller David J. , Hammond Peter J.
Published: March 2009

Subcritical (hot/liquid) water was used in a simple static (non-flowing) vessel to treat three soils from former defense sites which were contaminated with the explosives TNT (12 wt.%), or RDX (0.62 wt.%) and HMX (0.16 wt. %). Significant degradation of RDX began at 100 C, and at 125 C for TNT and HMX, with the bulk of the undergraded explosives remaining in the soil rather than in the water phase. Based on HPLC/UV analysis, intermediate degradation products formed, but quickly degraded at < 250 C. Remediations performed using a generator-powered mobile pilot-scale unit (4 to 6 kg soil) with 4-L of water at 275 C for 1 h of real soils resulted in > 99.9% destrcution of TNT and HMX, and > 99.5% desstruction of RDX. None of the mutagenic nitroso derivatives of RDX and HMX were formed. “Microtox” acute toxicity tests with Vibrio fischeri showed no significant (compared to background) residual toxicity in either the process wastewaters or leachates from the treated soils. The operation is closed-loop (no air or water emissions), and process water can be recycled without treatment. Initial cost analysis indicates that the process should be competitive with other approaches such as bioremediation.
DIAMEX Process Development to Separate Trivalent Actinides From High Active Concentrates
Authors: Modolo G. , Seekamp S. , Vijgen H.
Published: March 2009

The DIAMEX process is one of the most promising processes for partitioning the trivalent actinides from the highly active raffinates (HARs) coming from the PUREX process. In the present work, experiements for the definition of a DIAMEX process for treating high-active concentrates (HACs), the product after the concentration and denitration of HAR, have been performed. Within the DIAMEX process, oxalic acid has to be added to the DIAMEX feed in order to prevent third-phase formation due to co-extraction of some fission products, e.g. Zr and Mo. However, the experimental results indicate that after adding oxalic acid to the HAC a precipitate is formed which, in addition to Zr and Mo as the main constituents, also contains considerable amounts of lanthanides, and thus will lead perhaps to a loss of trivalent actinides. This unusual behaviour of Ln(III) coprecipitation was further investigated. Thus, other Zr, Mo complexing agents were tested and with D-mannitol promising results were achieved. With D-mannitol we did not observe any precipitates and the extraction experiments could be performed using the DIAMEX solvent composed of 1.0 mol/L DMDBTDMA in TPH. Extraction results as the basis for the development of a continuous counter-current test will also be presented.
Lessons Learned Siting and Successfully Operating Two Large L/ILW Disposal Facilities in the U.S.
Authors: Prince Robert E. , Magnus Victor , Latham James W.
Published: March 2009

This paper addresses the experience, knowledge, and expertise that Duratek has acquired while performing environmental remediation at two large low-level radioactive waste (LLRW) disposal facilities in the United States. Environmental remediation and related waste disposal has been the company’s primary line of business line since it was founded in 1969. It has disposed of more than half of the low-level radioactive waste generated in the U.S. over the past thirty years, working with almost every radioactive waste generator in the country. That experience has allowed the company to develop a unique understanding of safe, efficient, and cost-effective LLRW disposal methods. The paper also tracks the history of waste disposal technology at the Barnwell Disposal Site in South Carolina and the U.S. Department of Energy Environmental Restoration Disposal Facility (ERDF) at Oak Ridge, Tennessee. In particular, it describes the evolution of trench design, operations, and disposal procedures for these facilities. It also discusses the licensing of one the most active waste disposal sites in the U.S., the success of which has been assured to customers and stake-holders because of: • Well trained personnel who are dedicated to the design, construction and operation of safe and efficient disposal facilities; • Commitment to strong community relations; • Comprehensive knowledge of proven disposal strategies, technologies, and management practices; • Capability and readiness to respond rapidly to routine and emergency situations; • Established record of comprehensive and responsive communications with regulatory authorities; • Commitment to quality, compliance and personnel health, and safety; and • Financial systems that ensure long-term facilities management.
Third Update on Environmental Remediation of Historic LLR Waste Sites in Canada (1997-2003)
Authors: Zelmer R. L. , Case G. G.
Published: March 2009

Canada’s Low-Level Radioactive Waste Management Office (LLRWMO) continues to make significant progress toward the remediation of orphan sites contaminated with historic Low Level Radioactive (LLR) waste. Since its establishment in 1982, the LLRWMO, which is operated by Atomic Energy of Canada Limited, has acted as the agent of the federal government in this area, taking policy and priority direction from the federal department of Natural Resources Canada. The LLRWMO has investigated and decontaminated structures and properties at many sites across the country. It has removed contaminated soil, debris and radioactive artifacts to interim storage or interim, in situ containment. It has worked with communities and regulatory agencies to develop locally acceptable waste management solutions for the short- and long-term. This paper provides an update on the progress of environmental remediation programs and projects of the LLRWMO made since the last reporting at the Sixth ICEM Conference in Singapore in 1997. Emphasis is placed upon the areas of sustained interim waste management and community problem solving in this period. In addition, comment is provided on the future of the program. On behalf of the federal government, the LLRWMO was appointed in 2000 July to act as the proponent for the Port Hope Area Initiative (PHAI), a ten-year $260 M undertaking that will see historic LLR wastes currently found in various community locations consolidated into safe, long-term management facilities, yielding environmental benefits for present and future generations. This activity is breaking new ground in the implementation of community recommended solutions and signals the way forward in Canada’s historic waste program.
Remote Handled Removal and Repackaging of Wastes at the VKTA Site Rossendorf
Authors: Kiolbassa A. , Helwig U. , Gu¨ldner E.
Published: March 2009

The paper will present the potential applications for EMSM-Elektro Master Slave Manipulators and other remote controlled equipment with the intention of reducing radiological exposure to personal.
Development and Evolution of a Site Survey System: Groundhog
Authors: Davies Mike , Murley Robert , Adsley Ian
Published: March 2009

Traditional techniques for the assessment of pollutants in contaminated land, notably brown-field sites, may not yield the speed and accuracy now required for estimates of risk and remediation cost. Detailed site investigation is often limited by the time and cost of laboratory-based analysis techniques and time-consuming data collation phases. Thus, relatively straightforward technical issues, such as the mapping of priority areas of a site, can be unnecessarily delayed and expensive. The GROUNDHOG system was developed to address these problems and to provide a platform for the development of a range of techniques for the radiological survey of potentially contaminated land. The system brings together the best of well-established and recent technologies. Visualisation of the survey results is improved by the use of Geographical Information Systems and Database systems allow an audit trail to be maintained as part of a Quality Assurance programme. Development of the Groundhog system has continued, increasing the sensitivity of the system for some applications, using gamma radiation spectrometry systems to provide qualitative measurements and constructing ruggedised systems for surveys of areas where the risks associated with manual surveys are deemed unacceptable. In recent years, ‘conventional’ Groundhog surveys have been performed on many nuclear and non-nuclear sites, for a wide range of reasons: de-licensing nuclear facilities; pre- and post-remediation surveys of contaminated land; during the remediation of contaminated land, to reduce waste volume. Specialised versions of the system have been developed and used for the location of discrete nuclear fuel ‘particles’ on beaches, sub-surface measurements have been made for estimating waste volume and a submarine survey has been conducted. This paper describes some of the projects completed and the technologies used to perform the work.
Doses Forming Paths of External Irradiation of Personnel on Radioactive Waste Disposal Facilities
Authors: Yakimenco T. V. , Paramonova T. I. , Smirnov V. A.
Published: March 2009

The centralized system of radioactive waste management formed beyond of a nuclear fuel cycle exist in Russia. In territory of Russian Federation 16 regional specialized combines (SO) “Radon” are located, the majority of them is entered into operation in the first half 60th years. The specialized combines “Radon” are carrying out centralized collection, RW transportation, processing, localization and long-term storage of RW low and average activity, delivery service of new sources of ionizing radiation to users, take part in liquidation of radioactive contamination of served territory, carry out an nature-conservative measures of pollution prevention of the environment fromradioactive substances and ensuring radiation safety of population. For years of their existence total number of the personnel has made hundreds person. In most cases the personnel of combines are under a radiation control since time of introduction of objects in operation, about 40 years.
The Radioactive Waste Management Programme in Spain
Authors: Beceiro Alvaro R. , Vico Elena , Neri Emilio. G.
Published: March 2009

The paper will start with an overview of the organisational and financing framework established in Spain for the safe and efficient management of radioactive waste and decommissioning of nuclear installations. Since its creation by Royal Decree in 1984, ENRESA, the Spanish Radioactive Waste Management Agency, is in charge of both activities. ENRESA is a state owned company whose shareholders are CIEMAT (Centro de Investigaciones Energe´ticas, Medioambientales y Tecnolo´gicas) and the State Industrial Holding (SEPI), both governmental institutions. In Spain the Directorate General for Energy Policy and Mines of the Ministry of Economy (MINECO) plays the leading role in controlling nuclear activities, since it is the body responsible for awarding licenses and permits for installations and activities within the framework of the existing nuclear legislation. The Nuclear Safety Council (CSN) was set up in 1980 as the only competent body in matters of nuclear safety and radiological protection, and is generally responsible for the regulation and supervision of nuclear installations. Any license granted by MINECO is subjected to the mandatory and binding report of the CSN. The paper will review the steps undertaken for solving the national problems associated with the management of radioactive waste and decommissioning of nuclear installations, including uranium mining and milling facilities, and will address the lessons learnt from the activities developed by ENRESA and the future goals to be met. Regarding the L/ILW (Low and Intermediate Level Radioactive Waste) programme, the main milestones of El Cabril L/ILW disposal facility will be described highlighting the most relevant events as well as the foreseen activities, most of them focus on optimizing the capacity of the already operating installation. The elaboration and signature of a Protocol, at the end of 1999, for collaboration on the radiological Surveillance of Metallic Materials in order to detect the possible presence of radioactive materials is worth to be mentioned because of the involvement and agreement of several public and private organisations as well as the administration. Concerning the SF and HLW (Spent Nuclear Fuel and High Level radioactive Waste) programme, the solutions adopted in order to solve the insufficient capacity of the storage pools at NPPs will be described as well as the evolution of the final disposal programme since its beginning and the foreseen goals to be achieved before the year 2010. The last activities will deal with the experience gained during the decommissioning of Vandello´s I NPP and the future decommissioning projects. The decision taken in 2002 by the Spanish authorities to close down Jose´ Cabrera NPP in April 2006, before its 40 years lifetime, has had an impact on ENRESA’s activities.
Free Release Waste Characterisation During the Decommissioning of Windscale Pile 2 Chimney
Authors: Frith A. J. , Dickson E. D. , Hewetson J. A. , McCormick K.
Published: March 2009

The decommissioning of Windscale Pile 2 Chimney resulted in the removal of one of most historically significant and prominent features of the Sellafield site. The project was the first large scale concrete “free release” operation to be undertaken on the Sellafield nuclear site, producing 4500 tons of concrete of which 3000 tons was demonstrated to be “free release”. The paper describes the radiometric techniques employed in the characterization and segregation of the concrete into Low Level Waste (LLW), Very Low Level Waste (VLLW) and Free Release categories. It examines the robust solutions that were developed to meet the technical and regulatory challenges of the project, which included the definition of free release, the selection of averaging volumes, the testing and validation of the monitoring systems employed for bulk quantities of concrete and the identification and removal of small numbers of fuel particles from the free release waste stream. As a result of interest shown by other BNFL and UKAEA decommissioning projects at Sellafield, the paper finishes by discussing ways in which the system may be developed to assay other waste streams.
Independent Monitoring of Solid Low Level Radioactive Waste Disposals in the UK
Authors: Daish S. R. , Leech N. A.
Published: March 2009

The independent monitoring of solid low level radioactive waste (LLW) in the United Kingdom is undertaken by NNC Limited on behalf of The Environment Agency to ensure that disposals are within the authorised limits. Waste consignments are seized by the Agency prior to disposal and are transported to the Waste Quality Checking Laboratory (WQCL) at Winfrith, where the contents are analysed and assessed by destructive and non-destructive testing. All work performed at the laboratory is qualified by a Quality Assurance System which has been accredited by the United Kingdom Accreditation Service (UKAS). This paper outlines the regulatory framework for control of LLW disposals in the UK and describes the techniques used at WQCL for radioactive waste assessment.
Proceeding Toward a License Application for U.S. Nuclear Waste Repository: Total System Performance Assessment Approach
Authors: McNeish Jerry , Swift Peter , Howard Rob , Sevougian David , Kalinich Donald , MacKinnon Robert
Published: March 2009

The development of a deep geologic repository system in the United States has progressed to the preparation of an application for a license from the U.S. Nuclear Regulatory Commission. The project received site recommendation approval from the U.S. President in early 2002. The next phase of the project involves development of the license application (LA) utilizing the vast body of information accumulated in study of the site at Yucca Mountain, Nevada. Development of the license application involves analyses of the total system performance assessment (TSPA) of the repository, the TSPA-LA. The TSPA includes the available relevant information and model analyses from the various components of the system (e.g., unsaturated geologic zone, engineered system (waste packaging and drift design), and saturated geologic zone) (see Fig. 1 for nominal condition components), and unites that information into a single computer model used for evaluating the potential future performance or degradation of the repository system. The primary regulatory guidance for the repository system is found in 10 CFR 63, which indicates the acceptable risk to future populations from the repository system. The performance analysis must be traceable and transparent, with a defensible basis. The TSPA-LA is being developed utilizing state-of-the-art modeling software and visualization techniques, building on a decade of experience with such analyses. The documentation of the model and the analyses will be developed with transparency and traceability concepts to provide an integrated package for reviewers. The analysis relies on 1000’s of pages of supporting information, and multiple software and process model analyses. The computational environment represents the significant advances in the last 10 years in computer workstations. The overall approach will provide a thorough, transparent compliance analysis for consideration by the U.S. Nuclear Regulatory Commission in evaluating the Yucca Mountain repository.
The Stepwise Approach to Decision Making for Long-Term Radioactive Waste Management: Activities of the OECD/NEA Forum on Stakeholder Confidence
Authors: Pescatore C. , Va´ri A.
Published: March 2009

The context of long-term radioactive waste management is being shaped by changes in modern society. Values such as health, environmental protection and safety are increasingly important in our society, demanding new forms of risk governance in dealing with hazardous activities. These changes necessitate, in turn, new forms of dialogue and decision-making processes that include a large number of stakeholders. The development and implementation of radioactive waste management schemes take place over a long time scale, on the order of decades. In this complex context, a “decision” no longer means opting, in one go and for all time, for a complete package solution. Instead, a decision is one step in an overall, cautious process of examining and making choices that preserve the safety and well-being of the present generation and the coming ones while not needlessly depriving the latter of their right of choice. Consideration is thus increasingly being given to concepts such as “stepwise decision making” and “adaptive staging” in which the public, and especially the most affected local public, are meaningfully involved in the planning process. This paper is based on the work of the Forum on Stakeholder Confidence. It reviews the current developments regarding the stepwise approach with the aim to pinpoint where it stands, to highlight its societal dimensions, to analyse its roots in social sciences, and to identify guiding principles and issues in implementation.
The Hanau Land Remediation and Soil Segregation Project
Authors: Simon Gerold G. , Sokcic-Kostic Marina , Christ Bernhard G.
Published: March 2009

Radioactively contaminated land and its remediation is a growing sector of the Nuclear Business. The contamination is often associated with both naturally occurring and man-made radionuclides. A significant proportion of the waste contains only small amounts or low concentrations of radionuclides, and therefore much of the remediated material can be considered to be for “free release”. It has often not been possible to get adequate specific information on these materials, so a monitoring system is needed for their classification and characterisation. At the former Fuel Fabrication Facility NUKEM-Old in Hanau, Germany, all process equipment and installations were dismantled, building structure surfaces have been decontaminated. After final measurements of building contamination, mainly applying in-situ gamma-spectrometry, and approval for “free release”, or for dumping the material, all buildings were demolished. Since then, an Automated Soil Segregation System is being successfully used for the segregation and measuring of the remaining 20.000 tonnes of radioactively contaminated, or at least suspicious soil material, building foundations and sewerage system components.
Remediation of One Million Tons of Low-Level Radioactive Waste at the Department of Energy Fernald Closure Project
Authors: Murphy Cornelius
Published: March 2009

The Fernald Waste Pits Remedial Action Project (WPRAP) is located within the Department of Energy (DOE) Fernald Closure Project (FCP) Site located 32 km (20 miles) northeast of Cincinnati, Ohio. The FCP covers 424 ha (1,050 acres) of land in a rural, agricultural community. Fluor Fernald, Inc., is the Prime Contractor to the DOE for management of the FCP remediation. The WPRAP is removing approximately one million tons of low-level radioactive waste from eight storage pits which cover 15 ha (38 acres). This waste was generated during the FCP uranium metal production years of 1952 to 1989. Radioactive leachate from these wastes contributed to the contamination of an 80 ha (200 acres) portion of the Great Miami Aquifer. This aquifer is a drinking water source for the greater Cincinnati area. This unique project is one of the largest in the history of the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA)/ Superfund program. The objective of the project is the removal of all of the uranium and thorium contaminated wastes, soils and sludges from the waste pits area of the FCP. The facility in which these wastes are processed was financed and constructed by the Shaw Group (Shaw) and is operated jointly by Shaw and Fluor Fernald. Wet soils and sludges from the waste pits are excavated and thermally dried, then blended and analyzed. Once the waste has been determined to meet criteria for transportation and disposal, it is loaded into specialized railcars and transported by exclusive-use train to the Envirocare Waste Disposal Facility 3,200 km (2,000 miles) away in Clive, Utah. This project is presently about 72% complete. More than 600,000 tons of waste material have been safely transported off site by 95 exclusive-use trains. Waste shipments are projected to be completed by late next year (2004). The progress of the WPRAP to date demonstrates that a major DOE facility remediation project can be safely and successfully executed in partnership with private industry and local stakeholders utilizing proven commercial best practices and existing site labor resources. This paper details project performance to date, challenges encountered, and the cooperation of the DOE, the Environmental Protection Agency (EPA); Fluor Fernald, Inc.; Shaw, local labor unions, and the local community in planning and successfully executing the WPRAP. The cost of the WPRAP to the U.S. Government is projected to be about four hundred million dollars ($400,000,000.00).
Design and Construction of Multi-Layered Permeable Reactive Barrier for Removing Radionuclides, Nitrate, and Perchlorate at Los Alamos National Laboratory
Authors: den Baars P. Scott , Kaszuba John P. , Cota Ted , Myers Jonathan , Longmire Patrick , Strietelmeier Betty A. , Taylor Tammy P.
Published: March 2009

Los Alamos National Laboratory (LANL) and Shaw Environmental, Inc. (Shaw) designed and constructed a multilayered permeable reactive barrier (PRB) to remove contaminants from shallow alluvial groundwater within Mortandad Canyon at LANL. This project was developed as a pilot project for LANL to conduct research and development and proof of concept and as such does not meet all identified target contaminant concentrations, but provides LANL data for future applications of the technology. Shaw worked jointly with LANL scientists in selecting the site, conducting a geotechnical and hydrogeologic investigation with contaminant characterization for waste disposal, preparing a design basis report, conducting geochemical and groundwater flow modeling, and preparing both conceptual and final detailed engineered designs. Geochemical modeling of the PRB multibarrier processes was conducted to predict influent and effluent contaminant concentrations and evaluate the potential for mineral precipitation and reduction of effective porosity in the barrier. A numerical model of groundwater flow was constructed to simulate hydrogeologic conditions in Mortandad Canyon and then used to simulate flow with the PRB in place. The Mortandad Canyon PRB is designed to remove radionuclides (americium-241, plutonium-238 and 239/240, and strontium-90), nitrate, and perchlorate from alluvial groundwater. The PRB consists of a funnel and gate constructed of sealable sheet piling driven through the alluvium and into the underlying volcanic tuff. The gate is designed as a braced cofferdam. The gate contains four sequential media cells consisting of lava rock gravel, mineral apatite (a calcium phosphate), biobarrier, and limestone gravel. The lava rock gravel will sorb colloids (sorbed with americium, plutonium, and strontium) from the alluvial groundwater. The apatite will remove soluble metals and radionuclides through sorption processes. The biobarrier serves as a host microorganisms that biodegrade nitrates and perchlorate. The limestone gravel functions to buffer the biobarrier effluent. In addition, there will also be sorption of soluble plutonium, americium, and metals within the biobarrier and limestone layer. A series of sampling ports and monitoring wells were installed within the reactive media cells. The purpose of the funnel is to direct shallow alluvial groundwater through the gate. This project was a joint effort between LANL and Shaw. The initial feasibility studies and bench scale treatability were conducted at LANL. The LANL laboratory data was used as the basis for design criteria. The hydrogeologic and geochemical modeling, engineering design, and construction were performed by Shaw with LANL guidance and input.
Suggestion to Waste Classification for Scaling Factor Method
Authors: Masui Hideki , Kashiwagi Makoto , Mu¨ller Wolfgang , Lante`s Bertrand
Published: March 2009

Scaling factors (SF) are widely used to determine the nuclide specific radioactivity concentration of a waste package. In this paper, an appropriate waste classification for applying the same SF value is considered through a study of differences in physicochemical behavior of nuclides and a comparison of nuclide data obtained by a radiochemical analysis of actual wastes from several waste streams. Corrosion product (CP) nuclides show only minor differences in production/transportation behavior through all waste streams because they are generated by the activation of reactor materials and have low solubility in common. Therefore a unified SF for all waste streams is considered applicable, though the SF can at best be determined for each individual waste stream. Fission product (FP) nuclides and alpha-emitters are generated by neutron capture and nuclear fission and their solubility varies. If Cs-137 is selected as the key nuclide, distinct differences in nuclides ratios are recognized between homogeneous waste (e.g. resins, concentrates) and heterogeneous waste (e.g. filter cartridges, dry active waste). This is mainly because the release behavior and the solubility of alpha emitters and FP nuclides differ from those of Cs-137. Our study suggests that all waste streams can be divided into those two categories. On the other hand, some countries selected Co-60 as key nuclide for alpha-emitters and some FP nuclides. If Co-60 is selected as the key nuclide, it may be helpful to categorize power plants according to their fuel failure history. This is because the generation mechanism of the key nuclide differs from that of the difficult to measure (DTM) nuclides. Within each categorization, insignificant differences are recognized in terms of ratios of DTM nuclides to Co-60, for both nuclides have rather low solubility. Therefore a unified SF can be applicable, though further categorization of SF is possible for more accurate estimate.
Challenges in Electronic Cooling: Opportunities for Enhanced Thermal Management Techniques — Microprocessor Liquid Cooled Minichannel Heat Sink
Authors: Schmidt Roger
Published: March 2009

The volumetric heat dissipated by computer equipment at each level of the package from the chip to the chassis is having a tremendous impact on the thermal management of computer equipment. Because of the consumer’s insatiable desire for increased performance the competitive pressures are driving the computer manufacturer to pack as much processor/memory performance within the smallest volume possible. The consumer views high performance in a compact package as a benefit. These market pressures seem to be in direct conflict with the desire to continue to provide air cooling solutions for the foreseeable future. Because of these trends in power and package design other cooling technologies beside air are now becoming viable techniques but each must be weighed with many other factors that influence the cooling technology selected. These factors will discussed along with two specific IBM server packages and their associated cooling technology employed. Finally a microprocessor liquid cooled minichannel heat sink will be described and its performance presented as applied to a current microprocessor (IBM Power4) chip.
Convective Boiling in Parallel Micro-Channels
Authors: Hetsroni G. , Klein D. , Mosyak A. , Segal Z. , Pogrebnyak E.
Published: March 2009

Experiments were performed with clear water and with surfactant flowing in parallel triangular micro-channels. The study is based on systematic measurements of temperature and flow pattern by infrared radiometry and high-speed digital video imaging. Different flow patterns were observed simultaneously in various micro-channels at a fixed value of water or surfactant flow rates. Depending on flow and heat flux, pressure and temperature instabilities in the heated micro-channels were studied. This work develops a practical modeling approach for two-phase micro-channel heat sinks and considers also effect of surfactant on convective boiling in micro-channels.
Critical Heat Flux in Microchannels: Experimental Issues and Guidelines for Measurement
Authors: Bergles A. E. , Kandlikar S. G.
Published: March 2009

Critical heat flux limit is an important consideration in the design of any flow boiling unit. Before the use of microchannels under flow boiling conditions becomes widely accepted in critical applications, such as electronics cooling and laser lenses, it is essential to develop CHF data for microchannels. The experiments required to obtain this information pose unique challenges as the channel dimensions become smaller. The issues of parallel channel instability, experimental control, experimental uncertainty, and conjugate effects need to be carefully addressed. These issues are addressed in the present paper, and guidelines helpful in the design of CHF experiments are outlined.
Single-Phase Flow and Heat Transport in Microchannel Heat Sinks
Authors: Garimella Suresh V. , Singhal Vishal
Published: March 2009

Microchannel heat sinks are widely regarded as being amongst the most effective heat removal techniques from space-constrained electronic devices. However, the fluid flow and heat transfer in microchannels is not fully understood. The pumping requirements for flow through microchannels are also very high and none of the micropumps in the literature are truly suitable for this application. A wide-ranging research program on microchannel heat sinks and micropumps is underway in the Electronics Cooling Laboratory at Purdue University. This article provides an overview of the research being conducted to understand fluid flow and heat transfer in microchannels and to identify pumping requirements and suitable mechanisms for pumping in microchannels.
Fabrication Technologies for Advanced Heat Transfer Applications
Authors: Grande William J.
Published: March 2009

Microchannel devices with channel widths in the range from one micron to several hundred microns have become increasingly important structures for heat transfer applications. This paper will examine several classes of fabrication technologies that are employed in the field. Established technologies, such as bulk and surface micromachining, high aspect ratio machining, and conventional machining have been perfected over the last several decades and are the principal methods of creating microchannel structures. Synthesizing technologies, such as wafer bonding and micromolding techniques, allow these primary structures to be assembled together into working devices or enable high volume, low cost manufacturing using microstructured masters. Established and synthesizing technologies have reached a high level of performance and are generally the subject of refinement efforts rather than innovative investigations. Active research into new materials, processes, and fabrication strategies are found in the class of emerging technologies. This paper will briefly review the dimensional spectrum of microchannels and survey the established and synthesizing technologies. Then an exploration of emerging technologies will be made. The topic of rapid prototyping will be given particular emphasis.
Characterization of the Hydrodynamically Developing Flow in a Microtube Using Molecular Tagging Velocimetry
Authors: Thompson B. R. , Maynes D. , Webb B. W.
Published: March 2009

There is a need for increased understanding of the momentum transport phenomena in micro-fluidic geometries to aid in the design and optimization of such devices. Micro-molecular tagging velocimetry (μMTV) has been used to characterize the hydrodynamic developing flow in a microtube with an inner diameter of 180 μm. μMTV is a non-intrusive laser-based technique for obtaining detailed measurements of velocity profiles in flows dominated by a single velocity component. μMTV measurements are made by directing an ultra-violet laser beam into a flow containing phosphorescent tracer molecules. The laser beam excites a line of phosphorescence in the flow. Subsequently, two digital images, separated by a short time delay, of the line are captured by a CCD camera. The displacement of the tracer molecules between the images can be determined from the two images and the velocity of the flow is thus calculated. Velocity profile data at ten axial locations within the hydrodynamic developing region of a 180 μm diameter tube were acquired using the μMTV approach. The uncertainty for these measurements ranged from 1.5% to 5.5% of the center line velocity. Data were taken at Reynolds numbers, Re, of 60, 140, 290, and 340. It was observed that a vena-contracta existed in the first few tube diameters for all Re. The velocity profiles obtained very close to the tube entrance exhibited a uniform velocity core flow surrounded by regions of relatively stagnant fluid in the near wall regions. The profiles evolved in the downstream direction until the classical parabolic distribution was observed. The total hydrodynamic entry length agrees well with values published in the literature for macroscale flows, obtained from numerical simulation.
A Study on Stirring Characteristics in a Microchannel With Various Arrangement of Blocks
Authors: Heo Hyeung Seok , Suh Yong Kweon
Published: March 2009

In this study a newly designed microchannel is proposed. This design comprises periodically arranged simple blocks. In this configuration, the stirring is greatly enhanced at a certain parameter set. To characterize the flow field and the stirring effect both the numerical and experimental methods were employed. To obtain the velocity field, three-dimensional numerical computation to the Navier Stokes equations are performed by using a commercial code, FLUENT 6.0. The fluid-flow solutions are then cast into studying the characteristics of stirring with the aid of Lyapunov exponent. In this study the Lyapunov exponents are computed manually because the commercial code does not provide the corresponding option. In the experiment, flow visualization for the stirring effect is performed by using pure glycerin in one tank and glycerin mixed with a fluorescent dye in the other. The numerical results show that the particles’ trajectories in the microchannel heavily depend on the block arrangement. It was shown that the stirring is significantly enhanced at larger block-height and it reaches maximum when the height is 0.8 times the channel width. We also studied the effect of the block stagger angle, and it turns out that the stirring performance is the best at the block angle 45°.
Pressure Losses in a Network of Triangular Microchannels
Authors: Niklas Mariusz , Favre-Marinet M.
Published: March 2009

The flow characteristics of a network of parallel microchannels (hydraulic diameter: 110 μm) were investigated both experimentally and numerically in the present work. The cross-section of the micro-channels was triangular for further application to micro heat pipes. Measurements of the pressure drop across the microchannels network showed a dramatic increase of the pressure losses and a departure from the law of fully developed flow as soon as the Reynolds number of the flow exceeded about 10. Numerical computations of the flow were performed by using the classical laws of hydrodynamics in order to explain this surprizing result. They showed a good agreement with the experimental results, which suggests that there are no size effects at the length scale used in the experiments. Moreover, the mechanisms responsible of the large pressure drop in the high-range of Reynolds number are identified by the numerical analysis. They correspond to extra head losses due to separation in several parts of the test cell.
Time-Dependent Electrokinetic Flow in Infinitely Extended Rectangular Microchannels With Slip
Authors: Yang Jun , Kwok Daniel Y.
Published: March 2009

Rectangular microchannel is the most popular shape to be widely used in Microelectromechanical devices. However, analytical solutions of flow in this shape of microchannels are seldom reported because of its two dimensional problem. Furthermore, microflows in microchannels are more difficult to describe. In this paper, we addressed analytically the problem of laminar electrokinetic slip flow in microchannels with rectangular cross-section subjected to a time-dependent pressure gradient and a time-dependent electric field. The analytical solution has been determined based on the Debye-Hu¨ckel approximation and its solution is useful to precisely control microflow with slip in rectangular cross-section microchannels.
Gas Flow Study in MEMS Using Lattice Boltzmann Method
Authors: Tang G. H. , Tao W. Q. , He Y. L.
Published: March 2009

Isothermal gas flows in two-dimensional microchannels are investigated with the lattice Boltzmann method. The slip velocity on the solid boundaries can be obtained reasonably when bounce–back reflection is combined with specular reflection in a certain proportion. Behaviors in the microchannel flow including velocity distribution, nonlinear pressure drop, and average friction factor are examined. The pressure distribution, the average friction factors and the mass flow rates are compared with those predicted by Arkilic’s model and experimental data and the agreement is reasonably good. Furthermore, the effects of bounce-back proportion rb on the slip velocity are investigated and its value is chosen to be 0.7 to best match the data from Arkilic’s model and available experimental data.