bluuebunny/crossref_metadata_2025 huggingface dataset:
Coupling Time-Dependent Sorption Values of Degrading Concrete With a Radionuclide Migration Model
Authors: Perko Janez , Mallants Dirk , Jacques Diederik , Wang Lian
Published: December 2010

Safety assessment of radioactive waste disposal facilities is usually carried out by means of simplified models. Abstraction of the numerical model from the real physical environment is done in several steps. One of the most challenging issues in safety assessment concerns the long time scales involved and the evolution of engineered barriers over thousands of years. For some processes occurring in specific engineered barriers the uncertainties related to long time scales are addressed by implementing conservative assumptions in the radionuclide migration models. Other processes such as chemical concrete degradation, however, can be estimated for long time periods by the use of coupled geochemical transport models. For many near-surface disposal facilities, concrete is a very important engineered barrier because it is used in the construction of disposal modules or vaults, in production of high-integrity monoliths and their backfilling and for waste conditioning. Knowledge on the durability of such concrete components and its relation to radionuclide sorption is important for a defensible safety assessment. Chemical degradation typically occurs as the result of decalcification, dissolution and leaching of cement components and carbonation. These reactions induce a gradual change in the solid phase composition and the concrete pore-water composition, from “fresh” concrete porewater with a pH above 13 to a pH lower than 10 for “evolved” porewater associated with fully degraded concrete. The focus of this work is to analyse the behaviour of the disposal facility in terms of radionuclides sorption values depending on the geochemical evolution of engineered barriers. The time-dependency of the concrete mineralogy and porewater is coupled with sorption values that are characteristic for the four concrete degradation states: (i) State I with a pH larger than 12.5, controlled by the dissolution of alkali-oxides, (ii) State II with a pH at 12.5 controlled by the dissolution of portlandite, (iii) State III with a pH between 12.5 and 10 when all portlandite is dissolved and the pore water composition is determined by different cement phases including calcium-silicate hydrates (C-S-H phases), and (iv) State IV with a pH lower than 10 with calcite and aggregate minerals present. Above mentioned pH values are valid for a system with a temperature of 25°C. Sorption values are obtained from a literature review. The time-dependency of the sorption values Rd is implemented in a one-dimensional radionuclide migration model used for release calculations from the planned near-surface disposal facility at Dessel, Belgium. Calculated releases will be discussed for radionuclides typical of low- and intermediate level short-lived (LILW-SL) waste.
Management of Metal Waste With High Concentration of Natural Radionuclides: Its Problems and Experience
Authors: Gelbutovskiy Alexander B. , Cheremisin Peter I. , Troshev Alexander V.
Published: December 2010

The given work describes the problem and the offered solutions for the NORM contaminated metal waste generated in the oil & gas and other non-atomic industries. The Russian Federation regulatory base, the fundamental metal LLW treatment scheme in accordance with Sanitary Rules and standards are given. It is shown, that one of the treatment ways is the processing at a specialized enterprise. Such an only enterprise in Russian Federation is the Joint-Stock Company (JSC) “ECOMET-S”. Its main activities are radwaste management, processing and disposal of the metal LLW. The operations’ main goal is to reduce the waste volume to its minimum before further burial and to return the clean metal to the unrestricted commercial turnover. The worked out technology and the enterprise industrial capacity are represented. JSC “ECOMET-S” activities results and processing and disposal schedule are given.
Automated Vehicle and Waste Package Survey System
Authors: Desrosiers Arthur E. , Mann Philip H.
Published: December 2010

SEC provides radiological control services for the D&D of the K-25 Gaseous Diffusion Plant in Oak Ridge TN. The demolition of the structure of the building has been initiated. Building debris will be disposed in a waste management facility on the Oak Ridge Reservation. This will require an estimated 70,000 truck shipments and each truck shipment must be inspected for radiological contamination. The inspection process includes measurements with hand held survey instruments and preparation of a “clearance ticket.” We developed a system for automatically generating the clearance ticket in order to reduce costs. The system automates data collection and report preparation by using a programmable handheld data terminal (HDT) that acquires the instrument readings and prints a preformatted clearance ticket. We estimate that the number of labor hours required to perform these truck inspections will be reduced by 33% compared to previous methods. The key to success for this project was developing a simple software interface that provides a graphic diagram of the truck and the survey points, provides step-by-step directions and status during the inspection, and enters data onto a preformatted report that can be printed directly from the HDT. Symbols differentiate the contact and smear measurements. The inspector simply positions the instrument probe and presses a button on the HDT to acquire data. The HDT software evaluates each measurement and instructs the inspector to continue taking measurements or respond to elevated levels of contamination. When the survey is completed, the HDT provides a clearance message on the LCD screen and the inspector may proceed to the next truck survey. At the end of the work turn, the stored results are printed for review and signature. Although the time required to complete the measurements is not reduced, the time required to prepare documentation is drastically condensed. Although the initial implementation is limited to surveying large trucks, the process may be extended to any repetitive waste management survey, such as inspections of shipping packages. Future versions of the software may also be expanded to minimize the time required for each measurement.
Study on Performance Assessment for HLW Repository in China
Authors: Chen Weiming , Wang Ju , Su Rui , Li Yunfeng
Published: December 2010

This paper presents the latest achievements of performance assessment (PA) for high level radioactive waste (HLW) repository in China. The first PA model, taking Beishan granite site as an example, is established with GoldSim and is verified by comparison with Japanese PA model. Then the behaviors of granite repository in Beishan area are simulated. The results from these simulations show that the engineered barrier is the most important part inside the repository, especially its bentonite plays a key role in the retardation of repository after the nuclide is released from the vitrified waste. Five sensitive parameters are identified and two design parameters are optimized. As a result, it has been proved that performance assessment model is a necessary tool to understand the behaviors of repository, to identify sensitive parameters, and to optimize design parameters.
Integration of the H2 Inhibition Effect of UO2 Matrix Dissolution Into Radiolytic Models
Authors: Duro Lara , Tamayo Abel , Bruno Jordi , Marti´nez-Esparza Aurora
Published: December 2010

Different models describing the dissolution mechanism of spent nuclear fuel under repository conditions have been developed in the last years. One of the most evolved ones is the Matrix Alteration Model (MAM), which is an Alteration/Dissolution source term model based on the oxidative dissolution of spent fuel. Oxidant and reducing species can be naturally or radiolytically-generated. The experimentally-observed inhibition of matrix dissolution by H2, was integrated into MAM by considering that H2 is able to consume the oxidant species responsible for UO2 oxidation, e.g. H2O2. As a consequence, MAM predicts lower H2O2 concentrations for systems containing larger amounts of dissolved H2. Radiolysis experiments carried out by Pastina and coworkers have shown that under specific conditions such as high linear energy transfer (LET) radiation and absence of solid phase, dissolved H2 has a negligible effect on the H2O2 concentration, thus suggesting that the H2 inhibition effect catalyzed by the matrix surface has not been properly implemented in MAM. Modelling exercises performed in this work confirm such point and reveal the necessity of considering H2-activation when modelling this kind of systems. In addition, it has been demonstrated that, for high LET radiation, a clear dependence exists between the extent of the H2 activation and the integral LET of the radiation. The integration of the function describing such dependence allows improving the implementation of the H2 inhibition effect in MAM.
MGAv10: The Latest Evolution in the Multi-Group Analysis Code
Authors: Croft Stephen , Bosko Andrey , Gunnink Ray , Philips Sasha , Lamontagne Joe , Koskelo Markku , McElroy Robert
Published: December 2010

At many points in the safe and transparent handling of plutonium materials the relative isotopic composition of the principle isotopes needs to be known. Sometimes this information may be of primary interest — such as in the verification of safeguard declarations or in the confirmation of the reactivity of mixed oxide fuel. At other times, e.g., for radioactive waste characterization, the isotopic composition may be needed to calculate specific thermal power or specific spontaneous fission rates for the item under study, which can subsequently be combined with calorimetric and correlated neutron counting measurements, respectively, in order to make quantitative assessments of the mass of Pu and associated nuclides that are present in an item. The Multi-Group Analysis code MGA is a highly regarded and widely used computer code for the analysis of high resolution gamma ray spectra in order to extract the relative isotopic composition of plutonium for a diversity of items with minimal prior information. It has been honed over many years to give reliable results for a broad range of measurement scenarios commonly encountered in the fuel cycle. The nuclear industry is not dormant however and the demands on such codes continue to shift as a combination of technology and necessity open up new application areas. For example, while MGA had its origins in the analysis of clean spectra on product material principally for nuclear safeguards applications taken with germanium detectors having good low-energy resolution, it is now widely applied to the characterization of drummed waste forms and the complex spectra from such items acquired with much larger volume and poorer resolution detectors often used in such applications for the dual use of quantitative assay of the many gamma-emitters. This new domain of operational experience resulted in the need to enhance MGA to deal with spectra of poor statistical quality and also to cope with some of the complications that arise in the analysis of unusual spectra. Together with some additional changes made to incorporate feedback since the release of version 9.63 (which had minor revisions denoted by the letters A through H) of the code this has resulted in the creation of MGA v10. In this paper we shall outline the main changes to the code explaining why they were conceived and implemented. We illustrate what kinds of measurement problems can now be addressed over and above the previous capabilities which have been preserved and verified by the same set of regression tests that have been applied to previous generation of the code.
Radioactive Oil Decontamination Development: An Overview
Authors: Krasznai John P.
Published: December 2010

Insulating, hydraulic and vacuum pump oils are used extensively in CANDU plants. It is inevitable that at some point in the equipment life cycle that these oils will become waste products and their disposal needs to be properly managed. The presence of radioactivity and conventional contamination in the oils (mixed waste) pose special challenges to the waste manager. This paper provides an overview of waste oil streams that have been problematic in the CANDU nuclear industry and the decontamination processes that were developed for each to effectively remove a variety of radioactive species including tritium as well as conventional hazardous materials such as PCB, lead and cadmium.
Dose Validation in the Dutch Interim Waste Storage Facility
Authors: Welbergen Jeroen , van Velzen Leo
Published: December 2010

All radioactive waste in the Netherlands is collected by COVRA (acronym for Central Organisation for Radioactive Waste) that operates a facility for treatment of waste including interim storage buildings for HLW, LILW and TENORM (Calcinate and U3O8). Like many other waste management organisations, COVRA developed and adopted different waste storage strategies for different types of waste. The basis of all strategies is Isolation, Control and Surveillance (ICS), a principle aimed at minimization of dose to operators and the public alike. The stacking of waste in the storage buildings directly follows from this principle. To minimize radiation exposure of employees, waste is stacked in blocks. In these block the packages with low dose rates are placed on the outside and are shielding packages with higher dose rates and neutron sources inside. The packages with the lowest dose rate are stacked against the outer walls to minimize radiation into the environment. In 2004, a novel Non-Destructive Assay (NDA) method was used to validate the applied waste storage strategies in terms of spatial dose rate distribution. With this method measurements were performed in one of the interim storage modules for LILW. The dose rate at a height of 6m, mainly responsible for the sky-shine at the site boundary, was somewhat higher then expected. Based on the experience and feedback, the NDA method was developed further, into the present INDSS-R (acronym for INDoor Survey System – Radiation) method. This new method was put to the test in 2006 with a second series of spatial radiological measurements. The main aim was now to verify the reproducibility of the method. In 2008, a third series of measurements was carried out and the following data was collected: • dose rate (using a pressurized ionisation chamber) • nuclide depended gamma photon flux (by means of a 3*3″ NaI detector ) • thermal neutron flux measurements (by means of LiI(Eu) detector) The results of the third serial of spatial radiological and thermal neutron flux measurements will be presented and compared to the 2004 and 2006 data on the operational dose for operators and for the public.
Separation of Lanthanoid Phosphates From the Spent Electrolyte of Pyroprocessing
Authors: Amamoto Ippei , Kofuji Hirohide , Myochin Munetaka , Tsuzuki Tatsuya , Takasaki Yasushi , Yano Tetsuji , Terai Takayuki
Published: December 2010

This study is carried out to make the pyroprocessing hold a competitive advantage from the viewpoint of environmental load reduction and economical improvement. As one of the measures is to reduce the volume of the high-level radioactive waste, the phosphate conversion method is applied for removal of fission products from the melt as spent electrolyte in this paper. Though the removing target elements in the medium are alkali metals, alkaline earth metals and lanthanoid elements, only lanthanoid elements and lithium form the insoluble phosphates by reaction with Li3PO4 or K3PO4. Therefore, as the first step, the precipitation experiment was carried out to observe the behaviours of elements which form the insoluble precipitates as double salts other than simple salts. Then the filtration was experimented to remove lanthanoid precipitates in the spent electrolyte using Fe2O3-P2O5 glass system as a filtlation medium which is compatible material with the glassification. The result of separation of lanthanoid precipitates by filtration was effective and attained almost 100%.
Dismantling the Reactor Containment of Germany’s First NPP
Authors: Eickelpasch Ludger
Published: December 2010

Decommissioning and dismantling of a Nuclear Power Plant (NPP) is a challenging task with very long duration. Coming to the very end of a dismantling project it is still worth to think about optimization possibilities such as parallel work. Some additional constrains generated by parallel activities may raise but it can still be effective due to reduction of the overall project schedule.
Technical Performance Characterization of Fourier Transform Profilometry for Quantitative Waste Volume Determination Under Hanford Waste Tank Conditions
Authors: Monts David L. , Jang Ping-Rey , Long Zhiling , Norton Olin P. , Gresham Lawrence L. , Su Yi , Lindner Jeffrey S.
Published: December 2010

The Hanford Site in western Washington state is currently in the process of an extensive effort to empty and close its radioactive single-shell and double-shell waste storage tanks. Before this can be accomplished, it is necessary to know how much residual material is left in a given waste tank and the chemical makeup of the residue. The Institute for Clean Energy Technology (ICET) at Mississippi State University is currently developing an quantitative in-tank inspection system based on Fourier Transform Profilometry, FTP. FTP is a non-contact, 3-D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, FTP is capable of determining the height (depth) distribution (and hence volume distribution) of the target surface, thus reproducing the profile of the target accurately under a wide variety of conditions. Hence FTP has the potential to be utilized for quantitative determination of residual wastes within Hanford waste tanks. We report the results of a technical feasibility study to document the accuracy and precision of quantitative volume determination using the Fourier transform profilometry technique under simulated Hanford waste tank conditions.
Factors Influencing the Performance and Lifetime of Fibrous Glass and Metal Media HEPA Filters
Authors: Waggoner Charles A. , Parsons Michael S.
Published: December 2010

High efficiency particulate air (HEPA) filters are used in a variety of nuclear applications as final air treatment units. The design of air filtration systems in nuclear facilities that will function well requires a significant amount of knowledge about the challenge conditions that these filters will be exposed to. Additionally, risk assessments conducted as a review of these systems need also to be based on knowledge of filter challenge parameters during upset conditions that may be used as design basis conditions. This paper presents a summary of findings of factors that influence the performance lifetime of conventional fibrous glass media filters. These factors include aerosol challenge (particle size distributions and mass concentrations), media velocities, wetting conditions, and changes in gas density. These data are correlated to design considerations for new systems and to process upset conditions used in risk assessments, particularly those involving fires. Data from filter testing activities are also compared to filter loading models and to literature information regarding aerosol emission rates from combustion of various materials. Additional data are provided relating the performance of metal media HEPA filters under conditions that exceed maximum ranges for fibrous glass filters. These discussions focus on selection factors between these two categories of filter units.
A Preliminary Postclosure Safety Assessment of OPG’s Proposed L&ILW Deep Geologic Repository, Canada
Authors: Little Richard , Avis John , Calder Nicola , Garisto Nava , Gierszewski Paul , Leung Helen , Limer Laura , Penfold James , Towler George , Walke Russell , Walsh Robert
Published: December 2010

Ontario Power Generation (OPG) is proposing to build a Deep Geologic Respository (DGR) for Low and Intermediate Level Waste (L&ILW) near the existing Western Waste Management Facility at the Bruce site in the Municipality of Kincardine, Ontario. The Nuclear Waste Management Organization (NWMO), on behalf of OPG, is currently preparing an Environmental Impact Statement (EIS) and Preliminary Safety Report (PSR) for the proposed repository. This involves investigation of the site’s geological and surface environmental characteristics, conceptual design of the DGR, and technical studies to demonstrate the operational and long-term safety of the proposed facility. A preliminary postclosure safety assessment (SA) was undertaken in 2008 and 2009. Consistent with the guidelines for the preparation of the EIS for the DGR and the regulatory guide on assessing the long-term safety of radioactive waste management, the SA evaluated the DGR’s performance and its potential impact on human health and the environment through pathway analysis of contaminant releases, contaminant transport, receptor exposure and potential effects. Consideration was given to the expected long-term evolution of the repository and site following closure (the Normal Evolution Scenario) and four disruptive (“what if”) scenarios (Human Intrusion, Severe Shaft Seal Failure, Open Borehole, and Extreme Earthquake), which considered events with uncertain or low probability that could disrupt the repository system. Conceptual and mathematical models were developed and then implemented in a range of software tools including AMBER, to provide estimates of impacts such as dose, FRAC3DVS, for detailed 2D and 3D groundwater flow and transport calculations, and T2GGM, a code that couples the Gas Generation Model (GGM) and TOUGH2 and models the generation of gas in the repository and its subsequent 2D transport through the geosphere. Calculations have been undertaken to assess the impact of radionuclides on human and non-human biota and the impact of non-radioactive species on humans and the environment. The results indicate that the DGR system provides a high level of postclosure safety.
I-Graphite Waste Management in France
Authors: Ozanam Odile , Ouzounian Gerald
Published: December 2010

In France about 23,000 tons of irradiated graphite waste were generated mainly from 9 nuclear gas-cooled reactors (UNGG type). All these reactors are now shut down. In France, a final disposal route has been decided for i-graphite waste management by the French Parliament in 2006. The planning act of June 28, 2006 has provisioned that a research and investigation programme shall be established with a view to developing disposal options for graphite waste. The low specific activity of these wastes means shallow disposal facilities located in a geological layer with a low permeability. The main radionuclides that need specific attention for the long term safety are Chlorine 36 and Carbon 14. This establishes the main requirements for the graphite disposal. According to these requirements, underground disposal in a clay layer, at shallow depth, offers good performance in terms of the delay and the reduction of the radionuclide release; it also offers guarantees for the cover’s hydrodynamic and transport properties. The progress of the programme makes it now necessary for a site to be found. Based on a bibliographical study, a wide-spread call for applications took place during the second semester of 2008.
NUMO-RMS: A Practical Requirements Management System for the Long-Term Management of the Deep Geological Disposal Project
Authors: Ueda Hiroyoshi , Suzuki Satoru , Ishiguro Katsuhiko , Oyamada Kiyoshi , Yashio Shoko , White Matt , Wilmot Roger
Published: December 2010

NUMO (Nuclear Waste Management Organization of Japan) has the responsibility for implementing deep geological disposal of high-level (HLW) and transuranic (TRU) radioactive waste from the Japanese nuclear programme. A formal Requirements Management System (RMS) is planned to efficiently and effectively support the computerised implementation of the management strategy and the methodology required to drive the step-wise siting processes, and the following repository operational phase. The RMS will help in the comprehensive management of the decision-making processes in the geological disposal project, in change management as the disposal system is optimised, in driving projects such as the R&D programme efficiently, and in maintaining structured records regarding past decisions, all of which lead to soundness of the project in terms of long-term continuity. The system is planned to have information handling and management functions using a database that includes the decisions/requirements in the programme under consideration, the way in which these are structured in terms of the decision-making process and other associated information. A two-year development programme is underway to develop and enhance an existing trial RMS to a practical system. Functions for change management, history management and association with the external timeline management system are being implemented in the system development work. The database format is being improved to accommodate the requirements management data relating to the facility design and to safety assessment of the deep geological repository. This paper will present an outline of the development work with examples to demonstrate the system’s practicality. In parallel with the system/database developments, a case research of the use of requirements management in radioactive waste disposal projects was undertaken to identify key issues in the development of an RMS for radioactive waste disposal and specify a number of use cases to guide the overall development of the system. The findings of the case research will also be shown in the paper to provide general information on the application of an RMS in a radioactive waste disposal programme, the difficulties of successful implementation and suggestions on how these difficulties can be overcome.
Implementation of a Geological Disposal Facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): The Potential for Interaction Between the Co-Located ILW/LLW and HLW/SF Components of a GDF
Authors: Towler George , Hicks Tim , Watson Sarah , Norris Simon
Published: December 2010

In June 2008 the UK government published a ‘White Paper’ as part of the “Managing Radioactive Waste Safety” (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK’s higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). This paper presents an overview of a study undertaken to assess the implications of co-location by identifying the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions that might occur during both the operational and post-closure phases, and their consequences for GDF design, performance and safety. The MRWS programme is currently seeking expressions of interest from communities to host a GDF. Therefore, the study was required to consider a wide range of potential GDF host rocks and consistent, conceptual disposal system designs. Two example disposal concepts (i.e. combinations of host rock, GDF design including wasteform and layout, etc.) were carried forward for detailed assessment and a third for qualitative analysis. Dimensional and 1D analyses were used to identify the key interactions, and 3D models were used to investigate selected interactions in more detail. The results of this study show that it is possible for ILW/LLW and HLW/SF modules to be co-located without compromising key safety functions of different barrier components, and this reflects international precedents, e.g. the Andra and Nagra repository designs. There are two key technical issues that need to be managed in designing the geometry of the co-located GDF: (i) the heat flux from the HLW/SF module interacting with the ILW/LLW module, and (ii) the potential for development of an alkaline plume from the ILW/LLW module interacting with the HLW/SF module; particularly within fractured host rocks.
Radioactive Waste: Show Time?
Authors: Code´e Hans , Verhoef Ewoud
Published: December 2010

Time will render radioactive waste harmless. How can we manage the time radioactive substances remain harmful? Just ‘wait and see’ or ‘marking time’ is not an option. We need to isolate the waste from our living environment and control it as long as necessary. For the situation in the Netherlands, it is obvious that a period of long term storage is needed. Both the small volume of waste and the limited financial possibilities are determining factors. Time is needed to let the volume of waste grow and to let the money, needed for disposal, grow in a capital growth fund. An organisation such as COVRA — the radioactive waste organisation in the Netherlands — can only function when it has good, open and transparent relationship with the public and particularly with the local population. If we tell people that we safely store radioactive waste for 100 years, they often ask: “That long?” How can we explain the long-term aspect of radioactive waste management in a way people can relate to? In this paper, an overview is given of the activities of COVRA on the communication of radioactive waste management.
Shared, Regional Repositories: Developing a Practical Implementation Strategy
Authors: Verhoef Ewoud , McCombie Charles , Chapman Neil
Published: December 2010

The basic concept within both EC funded SAPIERR I and SAPIERR II projects (FP6) is that of one or more geological repositories developed in collaboration by two or more European countries to accept spent nuclear fuel, vitrified high-level waste and other long-lived radioactive waste from those partner countries. The SAPIERR II project (Strategic Action Plan for Implementation of Regional European Repositories) examines in detail issues that directly influence the practicability and acceptability of such facilities. This paper describes the work in the SAPIERR II project (2006–2008) on the development of a possible practical implementation strategy for shared, regional repositories in Europe and lays out the first steps in implementing that strategy.
Decommissioning of AECL Whiteshell Laboratories
Authors: Koroll Grant W. , Bilinsky Dennis M. , Swartz Randall S. , Harding Jeff W. , Rhodes Michael J. , Ridgway Randall W.
Published: December 2010

Whiteshell Laboratories (WL) is a Nuclear Research and Test Establishment near Winnipeg, Canada, operated by AECL since the early 1960s and now under decommissioning. WL occupies approximately 4400 hectares of land and employed more than 1000 staff up to the late-1990s, when the closure decision was made. Nuclear facilities at WL included a research reactor, hot cell facilities and radiochemical laboratories. Programs carried out at the WL site included high level nuclear fuel waste management research, reactor safety research, nuclear materials research, accelerator technology, biophysics, and industrial radiation applications. In preparation for decommissioning, a comprehensive environmental assessment was successfully completed [1] and the Canadian Nuclear Safety Commission issued a six-year decommissioning licence for WL starting in 2003 — the first decommissioning licence issued for a Nuclear Research and Test Establishment in Canada. This paper describes the progress in this first six-year licence period. A significant development in 2006 was the establishment of the Nuclear Legacy Liabilities Program (NLLP), by the Government of Canada, to safely and cost effectively reduce, and eventually eliminate the nuclear legacy liabilities and associated risks, using sound waste management and environmental principles. The NLLP endorsed an accelerated approach to WL Decommissioning, which meant advancing the full decommissioning of buildings and facilities that had originally been planned to be decontaminated and prepared for storage-with-surveillance. As well the NLLP endorsed the construction of enabling facilities — facilities that employ modern waste handling and storage technology on a scale needed for full decommissioning of the large radiochemical laboratories and other nuclear facilities. The decommissioning work and the design and construction of enabling facilities are fully underway. Several redundant non-nuclear buildings have been removed and redundant nuclear facilities are being decontaminated and prepared for demolition. Along with decommissioning of redundant structures, site utilities are being decommissioned and reconfigured to reduce site operating costs. New waste handling and waste clearance facilities have been commissioned and a large shielded modular above ground storage (SMAGS) structure is in final design in preparation for construction in 2010. The eventual goal is full decommissioning of all facilities and infrastructure and removal of stored wastes from the site.
An Integrated System for Conducting Radiological Surveys of Contaminated Sites
Authors: McCown Jay P. , Rogers Donna M. , Waggoner Charles A.
Published: December 2010

This paper describes an integrated detection system that has been developed to conduct radiological surveys of sites suspected of contamination of materials such as depleted uranium. This system utilizes cerium activated lanthanum bromide and thallium activated sodium iodide gamma detectors and can be easily adapted to include units for detecting neutrons. The detection system includes software controlling the collection of radiological spectra and GPS data. Two different platforms are described for conducting surveys, a modified zero turn radius (ZTR) mower and a three-wheeled cart that is manually pushed. The detection system software controlling data collection has components that facilitate completing a gridless survey on user specified spacings. Another package confirms that all data quality activities (calibrations, etc.) are conducted prior to beginning the survey and also reviews data to identify areas that have been missed for which data quality falls below user designated parameters. Advanced digital signal processing algorithms are used to enhance the interpretation of spectra for conducting background subtractions and for mapping. Data from radiation detection instruments and GPS antennae are merged and made compatible with mapping using Geosoft Oasis montaj software. A summary of system performance during field-testing is included in the paper. This includes survey rate, detection limits, duty cycle, supporting ancillary equipment/material, and manpower requirements. The rate of false positives and false negatives is discussed with the benefits of surveys conducted using synergetic detection systems such as electromagnetic induction imaging.
Involvement of ANDRAD in Endorsement of Decommissioning Documentation of Nuclear Facilities in Romania
Authors: Dinca Marin
Published: December 2010

National Agency for Radioactive Waste — ANDRAD — is in Romania, by law, the competent authority for the disposal administration of spent nuclear fuel and radioactive waste and for the coordination of the predisposal management of spent nuclear fuel and radioactive waste, inclusive decommissioning of nuclear facilities. Government Ordinance (GO) No. 11/January 30, 2003 and Government Decision (GD) No. 1601/December 23, 2003 established the ANDRAD’s foundation and organization. In accordance with GO No. 11/2003, republished, on the safe management of the radioactive waste, ANDRAD has the responsibility to endorse the decommissioning documentation issued by the main radioactive waste generators (nuclear installations and other major radiological installations: radioactive waste treatment plants, radioactive waste storage facilities, post irradiation examination laboratories, centres for radioisotopes production etc.). ANDRAD receives for endorsement some of the documentation for decommissioning that is provided by enforced norms for each type of nuclear facility. There are presented the nuclear facilities that must have decommissioning documentation endorsed by ANDRAD, the type of documents submitted by license holder to ANDRAD and the procedure of endorsement in relation with the regulatory body (CNCAN) approval of the decommissioning documents.
Radiological Assessment of Petroleum Pipe Scale Waste Streams From Dry-Rattling Operations
Authors: Hamilton Ian S. , Berry Robert O. , Arno Matthew G. , Fruchtnicht Erich H.
Published: December 2010

Petroleum pipe scale consists of inorganic solids, such as barium sulfate. These solids can precipitate out of brine solutions that are pumped out of oil wells as part of normal oilfield operations. The precipitates can nucleate on down hole pipe walls, causing the buildup of hard scales in some tubulars in a pipe string, while leaving others virtually untouched. Once the scale buildup is sufficient to restrict flow in the string significantly, the tubulars are removed from service. Once removed, tubulars are transported to storage yards for storage, subsequent inspection, and possible recycling. Many of the tubulars are never returned to service, either because the threads were too damaged, pipe walls too thin, or the scale buildup too thick. Historically, the tubular refurbishment industry used primarily one of two processes, either a high-pressure water lance or a dry, abrasive “rattling” process to ream pipes free of scale buildup. The dry rattling process was primarily for touching up new pipes that have rusted slightly during storage; however, pipes with varying levels of scale were reamed, leaving only a thin coating of scale on the inner diameter, and then returned to service. Chemically, radium is an analog for barium, and radium is present in parts-per-million quantities in the brines produced from downhole pumping operations. Thus, some of the scales contain radium salts. When the radium-bearing scales are reamed with a dry process there is the possibility of generating radioactive aerosols, as well as bulk waste materials. At Texas A&M University, and under the university’s radioactive materials broad scope license, an outdoor laboratory was constructed and operated with dry rattling equipment restored to the “as was” condition typical of historical pipe cleaning yards. A battery of measurements were obtained to determine the radiological and aerodynamic properties of scale-waste products liberated from the inner surfaces of a variety of tubulars acquired from working oilfields. Four dose pathways were analyzed: inhalation of scale dust released during dry rattling, ingestion of same, external exposure from uncleaned pipes, and external exposure from pipe scales dispersed on the ground in the natural discharge pattern of the rattling machine for given, recorded meteorological conditions. For the aerosol dispersion research, dust (mass) loading, and the respirable size-fraction component of generated dusts was determined. Additional research revealed the pulmonary solubility class of respirable scale aerosols, as well as the radon emanation rate of bulk scale materials.
Curing Time Effect on the Fraction of 137CS From Immobilized Radioactive Evaporator Sludge by Cement
Authors: Plecas Ilija , Dimovic Slavko
Published: December 2010

Traditional methods of processing evaporator concentrates from NPP are evaporation and cementation. These methods allow to transform a liquid radioactive waste into a more inert form, suitable for a final disposal. To assess the safety for disposal of radioactive mortar-waste composition, the leaching of 137Cs from immobilized radioactive evaporator concentrate into a surrounding fluid has been studied. Leaching tests were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Incremental leaching rates Rn(cm/d) of 137Cs from evaporator concentrates after 180 days were measured. The results presented in this paper are examples of results obtained in a 20-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.
Contingency Options for the Dry Storage of Magnox Spent Fuel in the UK
Authors: Morris Jenny , Wickham Stephen , Richardson Phil , Rhodes Colin , Newland Mike
Published: December 2010

The UK Nuclear Decommissioning Authority (NDA) is responsible for safe and secure management of spent nuclear fuel. Magnox fuel is held at some Magnox reactor sites and at Sellafield where it is reprocessed using a number of facilities. It is intended that all Magnox fuel will be reprocessed as described in the published Magnox Operating Programme (MOP) [1]. In the event, however, that a failure occurs within the reprocessing plant, the NDA has initiated a programme of activities to explore alternative contingency options for the management of wetted Magnox spent fuel. Magnox fuel comprises metallic uranium bar clad in a magnesium alloy, both of which corrode if exposed to oxygen or water. Consequently, contingency options are required to consider how best to manage the issues associated with the reactivity of the metals. Questions such as whether Magnox spent fuel needs to be dried, how it might be conditioned, how it might be packaged and held in temporary storage until a disposal facility becomes available, all require attention. During storage in the presence of water, the corrosion of Magnox fuel produces hydrogen (H2) gas, which requires careful management. When uranium reacts with hydrogen in a reducing environment, the formation of uranium hydride (UH3) may occur, which under some circumstances can be pyrophoric, and might create hazards which may affect subsequent retrieval and/or repackaging (e.g. for disposal). Other factors that may affect the choice of a viable contingency option include criticality safety, environmental impacts, security and Safeguards and economic considerations. Magnox fuel has been successfully dry-stored as intact fuel elements in CO2- and air-filled primary and secondary cells at Wylfa Power Station, UK. Storage of some fuel elements in the Wylfa secondary cells has been carried out successfully for over 25 years. Other relevant experience includes the French UNGG (Uranium Naturel Graphite Gaz) and U.S. Hanford N-Reactor spent fuels, both of which have been retrieved and dried after decades of wet storage. The dried fuels are respectively stored in sealed canisters in modular vault stores at Cadarache (CASCAD) and Hanford (Canister Storage Building). The applicability of these and other potential store designs, such as concrete and metal casks and silos, to the storage of Magnox spent fuel is discussed.
Strategy and Practice in Spent Sealed Sources Management in Belgium
Authors: De pooter Vincent , Vanleeuw David , Van Zaelen Gunter , Braeckeveldt Marnix
Published: December 2010

Radioactive sources are used for a variety of purposes, e.g. in medical treatment and diagnosis, research applications, measurement, testing, detection and calibration in industry, educational activities in colleges and universities etc. As part of its mission, ONDRAF/NIRAS, the Belgian Radioactive Waste Management Agency, draws up an inventory of all radioactive substances and nuclear installations on the Belgian territory. In recent years this inventory has been used to launch specific campaigns for the collection of different types of radioactive sources. In addition to this, the Royal Decree of 23 May 2006 concerning the transposition into Belgian law of the Spent High Activity Sealed Sources and the Management of Orphan Sources Directive of the EU (2003/122/EURATOM) has led to an increase in the number of requests addressed to ONDRAF/NIRAS for the collection of these types of radioactive waste and to an intensified collaboration between ONDRAF/NIRAS and the Belgian Safety Authority FANC/AFCN towards an effective management of orphan sources. Specific properties of these spent sources such as their activity, external dose rate, weight, size and/or their invalid special form certificate may complicate the transport and final treatment and conditioning of this type of waste and that is why these operations require careful attention. An overview of the radioactive sources already collected as radioactive waste or still present in the nuclear installations, different cases and problems encountered are presented in this paper, as well as the waste management options adopted by ONDRAF/NIRAS to deal with this type of waste.
Geochemical Behaviour of Uranium in Sedimentary Formations: Insights From a Natural Analogue Study
Authors: Noseck Ulrich , Havlova Vaclava , Suksi Juhani , Brasser Thomas , Cervinka Radek
Published: December 2010

Groundwater data from the natural analogue site Ruprechtov have been evaluated with special emphasis on the uranium behaviour in the so-called uranium-rich clay/lignite horizon. In this horizon in-situ Eh-values in the range of −160 to −280 mV seem to be determined by the SO42−/HS− couple. Under these conditions U(IV) is expected to be the preferential redox state in solution. However, on-site measurements in groundwater from the clay/lignite horizon show only a fraction of about 20% occurring in the reduced state U(IV). Thermodynamic calculations reveal that the high CO2 partial pressure in the clay/lignite horizon can stabilise hexavalent uranium, which explains the occurrence of U(VI). The calculations also indicate that the low uranium concentrations in the range between 0.2 and 2.1μg/l are controlled by amorphous UO2 and/or the U(IV) phosphate mineral ningyoite. This confirms the findings from previous work that the uranium (IV) mineral phases are long-term stable under the reducing conditions in the clay/lignite horizon without any signatures for uranium mobilisation. It supports the current knowledge of the geological development of the site and is also another important indication for the long-term stability of the sedimentary system itself, namely of the reducing geochemical conditions in the near-surface (30m to 60 m deep) clay/lignite horizon. Further work with respect to the impact of changes in redox conditions on the uranium speciation is on the way.