bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of a Methodology for Determining ILW Package Monitoring and Inspection Requirements
Authors: Morris Jenny , Hicks Tim , Wickham Stephen , Guida Antonio , McKinney James , Tearle Mark , Morris Darrell , Brownridge Melanie
Published: August 2012

In the UK, monitoring and inspection is required by the regulators to ensure continued safe storage, retrievability and disposability of packages of radioactive waste in interim storage. Monitoring and inspection requirements may also be determined by any requirements for transport, further storage or disposal when the store is emptied. This paper presents an approach to establishing a monitoring and inspection strategy for stored intermediate level waste (ILW) packages that was developed under the UK Nuclear Decommissioning Authority’s Direct Research Portfolio programme. The key aspects to be considered when developing a package monitoring and inspection strategy are identified and used to develop a generic multi-stage process that will enable store operators to identify a justifiable monitoring and inspection strategy for ILW packages in long-term interim storage. Two spreadsheet tools have also been developed to facilitate use of the multi-stage process by store operators (IMPS – Inspection and Monitoring of Packages in Stores and CBAT – Cost / Benefit Analysis Tool). A statistically based methodology has been used to identify the number of waste packages to be inspected and an appropriate frequency of inspection, and aspects such as cost, risk and operational feasibility are taken into account.
A Proposed Figure of Merit for Evaluating the Performance of Radiation Waste Measurement Systems
Authors: Keyser Ronald M. , Twomey Timothy R. , Webster Neil A. , Belbot Michael D.
Published: August 2012

Several different types of detectors, software, and hardware designs are employed in instruments used to monitor radioactive content of freight shipments in order to detect illicit trafficking of nuclear materials. The instruments can be container monitors, hand-held radiation detectors, mobile analysis systems, or fixed radiation portal monitors. However, within the various groupings (e.g., portal monitors), all instruments are expected to solve the same problem, that is, to detect and identify any radioactive material present according to the prescribed investigative procedure or CONOPS. The best way to compare the performance of different instruments is with a numerical score or Figure of Merit (FOM). The FOM must quantify the performance of the instrument with respect to true positives (TP), false positives (FP), and false negatives (FN). The minimization of FN for certain radionuclides (e.g., uranium and plutonium or other Special Nuclear Material (SNM)) is more important than the minimization of FN for non-threat nuclides (e.g., low level NORM). In a similar way, the minimization of FP for SNM is more important than falsely reporting the common NORM nuclides which represent no threat. System performance depends on the measurement detail (e.g., transit speed through a portal, measurement time in general, amount and distribution of NORM), therefore the test conditions should also be included in the statement of the FOM. The FOM is expected to vary significantly with the above measurement details. A FOM has been developed based on the number of true positives (TP), the number of false important positives (FIP), the number of FP, the number of true positives for SNM (TPSNM), and the number of false positives for SNM (FPSNM). This formula rates the overall performance with extra weight given to FP and FN for SNM. Examples will be shown for testing of portal monitors.
Intermediate and Long-Term Radiological Consequences of an Uncontrolled Access of Saline Solution Into the Asse Mine
Authors: Ustohalova Veronika , Ku¨ppers Christian
Published: August 2012

The risk of radioactive contamination in the biosphere surrounding the Asse salt mine has been assessed to determine the possible radioactive exposure to humans if the mine collapses. Geological conditions and anthropogenic activities have made the mine instable and allow salt-saturated ground water to seep in. This uncontrolled brine inflow significantly increases the risk of the mine collapsing. If the mine collapses, the brine will be pressed into groundwater, where the radionuclides can migrate into the biosphere and cause radioactive exposure. The key issue discussed in this paper is estimating the short- and long-term radiation burden for humans under several possible scenarios of radionuclide release. Only a radioecological model able to quantify and estimate processes taking place can generate usable results. This work develops the radioecological model describing both radionuclide migration and the resulting radiological exposure along several exposition pathways. Development of the model took into account the sorption processes, solubility limits and special aspects of decay chain migration. The radiological exposure was estimated under non-equilibrated conditions for the case of short-time expositions. At the end of this paper, the model’s background, the results of the computations and a comparison of several scenarios will be presented.
Radioactive Waste Management Challenges and Progress in Iraq
Authors: Al-Musawi Fouad , Shamsaldin Emad S. , Cochran John R.
Published: August 2012

The government of Iraq, through the Ministry of Science and Technology (MoST) is decommissioning Iraq’s former nuclear facilities. The 18 former facilities at the Al-Tuwaitha Nuclear Research Center near Baghdad include partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. These 18 former facilities contain large numbers of silos and drums of uncharacterized radioactive waste and approximately 30 tanks that contain or did contain uncharacterized liquid radioactive wastes. Other key sites outside of Al Tuwaitha include facilities at Jesira (uranium processing and waste storage facility), Rashdiya (centrifuge facility) and Tarmiya (enrichment plant). The newly created Radioactive Waste Treatment Management Directorate (RWTMD) within MoST is responsible for Iraq’s centralized management of radioactive waste, including safe and secure disposal. In addition to being responsible for the uncharacterized wastes at Al Tuwaitha, the RWTMD will be responsible for future decommissioning wastes, approximately 900 disused sealed radioactive sources, and unknown quantities of NORM wastes from oil production in Iraq. This paper presents the challenges and progress that the RWTMD has made in setting-up a radioactive waste management program. The progress includes the establishment of a staffing structure, staff, participation in international training, rehabilitation of portions of the former Radioactive Waste Treatment Station at Al-Tuwaitha and the acquisition of equipment.
Establishment of Research and Development Priorities Regarding the Geologic Disposal of Nuclear Waste in the United States and Strategies for International Collaboration
Authors: Nutt Mark , Voegele Michael , Birkholzer Jens , Swift Peter , McMahon Kevin , Williams Jeff , Peters Mark
Published: August 2012

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technologies (OFCT) has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel (UNF) and high level radioactive waste (HLW). The U.S. has, in accordance with the U.S. Nuclear Waste Policy Act (as amended), focused efforts for the past twenty-plus years on disposing of UNF and HLW in a geologic repository at Yucca Mountain, Nevada. The recent decision by the U.S. DOE to no longer pursue the development of that repository has necessitated investigating alternative concepts for the disposal of UNF and HLW that exists today and that could be generated under future fuel cycles. The disposal of UNF and HLW in a range of geologic media has been investigated internationally. Considerable progress has been made by in the U.S and other nations, but gaps in knowledge still exist. The U.S. national laboratories have participated in these programs and have conducted R&D related to these issues to a limited extent. However, a comprehensive R&D program investigating a variety of storage, geologic media, and disposal concepts has not been a part of the U.S. waste management program since the mid 1980s because of its focus on the Yucca Mountain site. Such a comprehensive R&D program is being developed and executed in the UFDC using a systematic approach to identify potential R&D opportunities. This paper describes the process used by the UFDC to identify and prioritize R&D opportunities. The U.S. DOE has cooperated and collaborated with other countries in many different “arenas” including the Nuclear Energy Agency (NEA) within the Organisation for Economic Co-operation and Development (OECD), the International Atomic Energy Agency (IAEA), and through bilateral agreements with other countries. These international activities benefited the DOE through the acquisition and exchange of information, database development, and peer reviews by experts from other countries. Recognizing that programs in other countries have made significant advances in understanding a wide range of geologic environments, the UFDC has developed a strategy for continued, and expanded, international collaboration. This paper also describes this strategy.
Technical Know-How for the Investigation and Modelling of Topographic Evolution for Site Characterisation
Authors: Doke Ryosuke , Yasue Ken-ichi , Niizato Tadafumi , Nakayasu Akio
Published: August 2012

Geological hazard assessments are being used to make important decisions relevant to nuclear facilities such as a repository for deep geological disposal of high-level radioactive waste. With respect to such repositories, topographic evolution is a key issue for description of the long-term evolution of a groundwater flow characteristics in time spans of tens to hundreds of thousands of years. The construction of topographic evolution models is complex, involving tacit knowledge and working processes. Therefore, it is important to externalise, that is to explicitly present the tacit knowledge and decision-making processes used by experts in the model building unambiguously, with thorough documentation and to provide key knowledge to support planning and implementation of investigations. In this study, documentation of the technical know-how used for the construction of a topographic evolution model is demonstrated. The process followed in the construction of the model is illustrated using task-flow logic diagrams; the process involves four main tasks with several subtasks. The task-flow followed for an investigation to estimate uplift rates linked to the task-flow for the modelling of topographic evolution is also illustrated. In addition, the decision-making processes in the investigation are expressed in logical IF-THEN format for each task. Based on the documented technical know-how, an IT-based Expert System was constructed. In future work, it is necessary to analyse the knowledge, including the management of uncertainties in the modelling and investigations, and to integrate fundamental ideas for managing uncertainties with expert system.
Applying Fluid Dynamics Simulations to Improve Processing and Remediation of Nuclear Waste
Authors: Knight Kelly J. , Berkoe Jon , Rosendall Brigette , Peltier Joel , Kennedy Chris
Published: August 2012

Transport and processing of nuclear waste for treatment and storage can involve unique and complex thermal and fluid dynamic conditions that pose potential for safety risk and/or design uncertainty and also are likely to be subjected to more precise performance requirements than in other industries. From an engineering analysis perspective, certainty of outcome is essential. Advanced robust methods for engineering analysis and simulation of critical processes can help reduce risk of design uncertainty and help mitigate or reduce the amount of expensive full-scale demonstration testing. This paper will discuss experience gained in applying computational fluid dynamics models to key processes for mixing, transporting, and thermal treatment of nuclear waste as part of designing a massive vitrification process plant that will convert high and low level nuclear waste into glass for permanent storage. Examples from industrial scale simulations will be presented. The computational models have shown promise in replicating several complex physical processes such as solid-liquid flows in suspension, blending of slurries, and cooling of materials at extremely high temperature. Knowledge gained from applying simulation has provided detailed insight into determining the most critical aspects of these complex processes that can ultimately be used to help guide the optimum design of waste handling equipment based on credible calculations while ensuring risk of design uncertainty is minimized.
Generic Design Assessment of Long-Term Spent Fuel Storage for New Reactors in the UK
Authors: Watson D. , Streatfield I. , Grundy C. , Price-Water S. , Glazbrook D. , Fisher C.
Published: August 2012

In the UK the Office for Nuclear Regulation and the Environment Agency developed the Generic Design Assessment process in response to a request from the UK Government. The process allows the regulators to jointly assess new nuclear reactor designs, in advance of any site-specific proposals to build a nuclear power station. Two reactor types are currently being assessed within Generic Design Assessment: • AREVA and Electricite´ de France’s UK EPR®; • Westinghouse Electric Company’s AP1000®. This paper will present the outcome of the assessment of radioactive waste management within the Generic Design Assessment process. One aspect of particular interest is the management of spent fuel from proposed new reactors as the assessment is based on an assumption that it will be sent for disposal. Therefore the paper will specifically consider the management of spent fuel and how this affects the regulatory decisions. The paper will look at four aspects. The first of these is to give a short overview of the Generic Design Assessment process. This will be followed by a summary of the Generic Design Assessment Radioactive Waste Management assessment on the acceptability of: • The types of waste and spent fuel. • The plans for conditioning of the wastes. • The safety issues associated with short-term storage. • The safety issues associated with long-term storage. • The issues associated with the disposal of the wastes. • The safety issues associated with decommissioning the reactors. The third aspect will be to look at the work commissioned by the Office of Nuclear Regulation in support of the Generic Design Assessment of radioactive waste management and how this has affected the regulatory decisions. This work has looked at the long-term stability of spent fuel in storage and the potential faults associated with the storage and handling of the spent fuel. The paper will end with the main conclusions of the radioactive waste management assessment within Generic Design Assessment. Looking at how storage of spent fuel can affect transport, disposal and decommissioning and how work by licensees could alter these conclusions.
Methodology for Identification of Conditionally Released Materials From Decommissioning Using the OMEGA Code
Authors: Zachar Matej , Danisˇka Vladimi´r , Necˇas Vladimi´r
Published: August 2012

The nuclear facility decommissioning process is characterized by production of large volumes of radioactive materials to be managed. The materials with radioactivity slightly above the defined unconditional clearance levels could be released conditionally for a specified industrial application in accordance with developed scenario ensuring that the limits for radiation exposure of workers and public will not be exceeded. The paper presents the methodology for identifying and assessment of the physical and radiological parameters of material items suitable for using in defined scenarios of conditional reuse of materials resulting from the decommissioning process. The calculation of mentioned parameters is done by using the analytical code OMEGA with implemented calculation stream for evaluation the conditional release process. The necessary input parameters for the OMEGA code are maximum concentrations of each radionuclide in the conditionally released materials which do not cause the doses above the limit values for the possible scenarios of further use of materials outside the nuclear site. The examples of such scenarios are assessed and presented in other related papers proposed for the ICEM conference (e.g. papers ICEM2011-59120 or ICEM2011-59128). Based on the inventory database of physical and radiological characteristics of decommissioned facility, the calculation algorithm separates the relevant material items with activity concentrations between unconditional and conditional clearance levels. The amount, material composition and radioactivity of such material items are declared as output parameters of the calculation and are also discussed in the paper. Such data could be used as one of the bases in the benefit analysis of potential application of conditional release within the frame of decommissioning waste management.
Demonstration Test of Underground Cavern-Type Disposal Facilities, Fiscal 2010 Status
Authors: Akiyama Yoshihiro , Terada Kenji , Oda Nobuaki , Yada Tsutomu , Nakajima Takahiro
Published: August 2012

A test to demonstrate practical construction technology for underground cavern-type disposal facilities is currently underway. Cavern-type disposal facilities are a radioactive waste repository excavated to a depth of 50 to 100 m below ground and constructed with an engineered barrier system (EBS) that is a combination of low-permeable bentonite material and low-diffusive cementitious material. The disposed materials are low-level radioactive waste with relatively high radioactivity, mainly generated from power reactor decommissioning, and certain transuranic wastes that are mainly generated from spent fuel reprocessing. The project started in fiscal 2005*, and since fiscal 2007* a full-scale mock-up of a disposal facility has been constructed in an actual sub-surface environment. The main objective of the demonstration test is to establish construction procedures and methods which ensure the required quality of an EBS on-site. Certain component parts of the facility had been constructed in an underground cavern by fiscal 2010, and tests so far have demonstrated both the practicability of the construction and the achievement of the required quality. This paper covers the project outline and the test results obtained by the construction of certain EBS components. * All fiscal years end on March 31 of the following calendar year; for example, fiscal 2005 is the year from April 1, 2005, to March 31, 2006.
Industrial Experience Feedback of a Geostatistical Estimation of Contaminated Soil Volumes
Authors: Faucheux Claire , Jeanne´e Nicolas
Published: August 2012

Geostatistics meets a growing interest for the remediation forecast of potentially contaminated sites, by providing adapted methods to perform both chemical and radiological pollution mapping, to estimate contaminated volumes, potentially integrating auxiliary information, and to set up adaptive sampling strategies. As part of demonstration studies carried out for GeoSiPol (Geostatistics for Polluted Sites), geostatistics has been applied for the detailed diagnosis of a former oil depot in France. The ability within the geostatistical framework to generate pessimistic / probable / optimistic scenarios for the contaminated volumes allows a quantification of the risks associated to the remediation process: e.g. the financial risk to excavate clean soils, the sanitary risk to leave contaminated soils in place. After a first mapping, an iterative approach leads to collect additional samples in areas previously identified as highly uncertain. Estimated volumes are then updated and compared to the volumes actually excavated. This benchmarking therefore provides a practical feedback on the performance of the geostatistical methodology.
Self-Disposal Option for Heat-Generating Waste
Authors: Ojovan Michael I. , Poluektov Pavel P. , Kascheev Vladimir A.
Published: August 2012

Self-descending heat generating capsules can be used for disposal of dangerous radioactive wastes in extremely deep layers of the Earth preventing any release of radionuclides into the biosphere. Self-disposal option for heat-generating radioactive waste such as spent fuel, high level reprocessing waste or spent sealed radioactive sources, known also as rock melting concept, was considered in the 70s as a viable alternative disposal option by both Department of Energy in the USA and Atomic Industry Ministry in the USSR. Self-disposal is currently reconsidered as a potential alternative route to existing options for solving the nuclear waste problem and is associated with the renaissance of nuclear industry. Self-disposal option utilises the heat generated by decaying radionuclides of radioactive waste inside a heavy and durable capsule to melt the rock on its way down. As the heat from radionuclides within the capsule partly melts the enclosing rock, the relatively low viscosity and density of the silicate melt allow the capsule to be displaced upwards past the heavier capsule as it sinks. Eventually the melt cools and solidifies (e.g. vitrifies or crystallizes), sealing the route along which the capsule passed. Descending or self-disposal continues until enough heat is generated by radionuclides to provide partial melting of surrounding rock. Estimates show that extreme depths of several tens and up to hundred km can be reached by capsules which could never be achieved by other techniques. Self-disposal does not require complex and expensive disposal facilities and provides a minimal footprint used only at operational stage. It has also an extremely high non-proliferation character and degree of safety. Utilisation of heat generated by relatively short-lived radionuclides diminishes the environmental uncertainties of self-disposal and increases the safety of this concept. Self-sinking heat-generating capsules could be launched from the bottom of the sea as well as from intermediate-depth or deep boreholes. Self-disposal can also be used with a novel purpose — to penetrate into the very deep Earth’s layers beneath the Moho’s discontinuity and to explore Earth interior.
Post-Iodine Production NORM Repository Technical Design and Reconstruction Experience in Turkmenistan
Authors: Gelbutovskiy A. B. , Cheremisin P. I. , Troshev A. V. , Egorov A. J. , Boriskin M. M.
Published: August 2012

Iodine extraction from the mineral water at the Balkanabat Iodine and Khazar chemical plants in Turkmenistan resulted in their territories contamination by NORM in very significant quantities. The bulk quantity of the accumulated waste is represented by sand, soil and charcoal mix. Charcoal was originally used for the Iodine sorption. It contains the radionuclides which have been presented in the mineral water and then been adsorbed during the Iodine extraction process. Plants territory remediation project results are given in the paper. The main topics are the radioactive waste collection, packaging, transportation and it’s placement to the repository. The given information relates covers the repository commissioning, filling and partial closure. The data on the NORM contaminated waste characteristics and quantities are described. The data on waste transportation containers and repository features are submitted also. One can find the information on the basic technical solutions of the repository construction, filling and closure, including covering coatings, drainage, inspection wells and equipment used. The information is valid as of 01.05.2011.
Application of an Environmental Remediation Methodology: Theory vs. Practice—Reflections and Two Belgian Case Studies
Authors: Blommaert W. , Mannaerts K. , Pepin S. , Dehandschutter B.
Published: August 2012

Like in many countries, polluted industrial sites also exist in Belgium. Although the contamination is purely chemical in most cases, they may also contain a radioactive component. For chemically contaminated sites, extensive regulations and methodologies were already developed and applied by the different regional authorities. However and essentially because radioactivity is a federal competence, there was also a necessity for developing a legal federal framework (including an ER-methodology [1]) for remediation of radioactive contaminated sites. Most of the so-called radioactive contaminated sites are exhibiting a mixed contamination (chemical and radiological), and hence the development of such methodology had to be in line with the existing (regional) ones concerning chemical contamination. Each authority having their own responsibilities with regard to the type of contamination, this makes it more complicated and time-consuming finding the best solution satisfying all involved parties. To overcome these difficulties the legal framework and methodology — including the necessary involvement of the stakeholders and delineation of each party’s responsibilities — has to be transparent, clear and unambiguous. Once the methodology is developed as such and approved, the application of it is expected to be more or less easy, logic and straightforward. But is this really true? The aim of this document is to investigate as well the impact of factors such as the type of radioactive contamination — levels of contamination, related to NORM activity or not, homogeneous or heterogeneous, the differences in licensing procedures,… — on the application of the developed methodology and what could be the consequences in the long run on the remediation process. Two existing case studies in Belgium will be presented ([2]). The first case deals with a historical radium contaminated site, the second one with a phosphate processing facility still in operation, both with (very) low levels of radioactivity but containing very large volumes of contaminated materials. These case studies will demonstrate that, although the applied methodology will be the same in both cases, the impact of e.g. sampling strategy, scenario definitions, modelisations, final destination of the land, presence of chemotoxic components, dose or risk assessments, uncertainties, derivation of clean-up radionuclide guidelines, stakeholder involvement and waste treatment could be important on licensing, cost-estimate, planning and final outcome of the environmental remediation activities to be executed.
Environmental and Radiological Remediation Under Canada’s Global Partnership Program 2004–11
Authors: Washer Michael J.
Published: August 2012

Following the “911” attack on the USA in 2001 the international community under Canada’s G8 leadership established a $20 billion Global Partnership initiative in 2002 to collaboratively address threats to global security posed by the proliferation and potential terrorist use of Weapons and Materials of Mass Destruction (WMMD) and related materials and knowledge. This major international initiative addressed four priority areas: (1) Chemical Weapon Destruction; (2) Nuclear powered submarine eliminations; (3) Nuclear and radiological security; and (4) Employment for former weapon scientists. Additionally the initiative has addressed Biological Non-Proliferation. Canada’s execution of all these program areas has resulted in substantial environmental benefits aside from the eradication and securing of WMMD. This paper reviews the environmental and radiological remediation achievements of the four primary Global Partnership program areas addressed under Canadian funding 2004 through 2011.
Disposition of Transuranic Residues From Plutonium Isentropic Compression Experiment (Pu-ICE) Conducted at Z Machine (LA-UR-10-05649)
Authors: Goyal Kapil K. , Humphrey Betty J. , Gluth Jeffry , French David M.
Published: August 2012

In 1992, the U.S. Congress passed legislation to discontinue above- and below-ground testing of nuclear weapons. Because of this, the U.S. Department of Energy (DOE) must rely on laboratory experiments and computer-based calculations to verify the reliability of the nuclear stockpile. The Sandia National Laboratories/New Mexico (SNL/NM) Z machine was developed to support the science-based approach for mimicking nuclear explosions and stockpile stewardship. Plutonium (Pu) isotopes with greater than ninety-eight percent enrichment were used in the experiments. In May 2006, SNL/NM received authority that the Z Machine Isentropic Compression Experiments could commence. Los Alamos National Laboratory (LANL) provided the plutonium targets and loaded the target assemblies provided by SNL/NM. Three experiments were conducted from May through July 2006. The residues from each experiment, which weighed up to 913 pounds, were metallic and were packaged into a 55-gallon drum each. SNL/NM conducts the experiments and provides temporary storage for the drums until shipment to LANL for final waste certification for disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. This paper presents a comprehensive approach for documenting generator knowledge for characterization of waste in cooperation with scientists at the two laboratories and addresses a variety of essential topics.
Independent Technical Reviews for Groundwater and Soil Remediation Projects at US Department of Energy Sites
Authors: Kaback Dawn S. , Chamberlain Grover , Morse John G. , Petersen Scott W.
Published: August 2012

The US Department of Energy Office of Environmental Management has supported independent technical reviews of soil and groundwater projects at multiple DOE sites over the last 10 years. These reviews have resulted in significant design improvements to remedial plans that have accelerated cleanup and site closure. Many have also resulted in improved understanding of complex subsurface conditions, promoting better approaches to design and implementation of new technologies. Independent technical reviews add value, because they provide another perspective to problem solving and act as a check for especially challenging problems. By bringing in a team of independent experts with a broad experience base, alternative solutions are recommended for consideration and evaluation. In addition, the independence of the panel is significant, because it is able to address politically sensitive issues. The expert panel members typically bring lessons learned from other sites to help solve the DOE problems. In addition, their recommendations at a particular site can often be applied at other sites, making the review even more valuable. The review process can vary, but some common lessons ensure a successful review: • Use a multi-disciplinary broadly experienced team; • Engage the panel early and throughout the project; • Involve regulators and stakeholders in the workshop, if appropriate; • Provide sufficient background information; • Close the workshop with a debriefing followed by a written report. Many groundwater remediation challenges remain at DOE sites. Independent technical reviews have and will ensure that the best capabilities and experience are applied to reduce risks and uncertainties.
Peculiarities of ChNPP Site Remediation During Its Decommissioning: Final State of ChNPP Site
Authors: Mitichkin Sergey V. , Seyda Valeriy A.
Published: August 2012

To review peculiarities of ChNPP site remediation during its decommissioning, final state of ChNPP site and prospects of its development.
Modeling Approaches for Evaluating the Effects of Heterogeneity on Two-Phase Flow Associated With the Migration of Waste-Generated Gas From SF/HLW- and L/ILW Repositories in Low-Permeability Formations
Authors: Papafotiou A. , Senger R. , Alcolea A. , Lanyon G. W. , Ewing J.
Published: August 2012

Different low-permeability formations are considered as potential host rocks for Low / Intermediate Level waste (L/ILW) and spent fuel / high-level / long-lived intermediate-level waste (SF/HLW/ILW) in Switzerland. As part of a generic site evaluation process, emphasis is on the assessment of Mesozoic limestones, marls and claystones in six potential siting areas in Northern and Central Switzerland. An important aspect in the site evaluation process is the characterization of the low-permeability formations in terms of spatial variability of the relevant transport properties (porosity, permeability, clay content), as they may affect the migration of waste-generated gas from L/ILW and SF/HLW/ILW repositories. Numerical modeling studies are presented, aimed at quantifying the impact of spatial variability of rock properties on gas release through the host rock on a deca- to hectometer scale. For this purpose, 2D models of an emplacement tunnel embedded in a low-permeability host rock are developed for both a sequence of limestones and marls with distinct lithological variability (Effingen Beds) and a claystone of moderate spatial variability (Opalinus Clay). For the Effingen Beds, a composite geological model is implemented, comprising stochastic representations of the different facies and the fracture systems. The facies model displays spatial variations in clay content, porosity and permeability within the different facies. The fracture model accounts for the hydraulic effects associated with the faults and fracture systems in the siting area under consideration. For the Opalinus Clay the available geostatistical information (experimental variograms of clay content, porosity, hydraulic conductivity) is used to generate a stochastic facies model based on a log-normal permeability distribution. A separate fracture network model is not established for the Opalinus Clay. A generic gas source-term is assigned to the emplacement tunnel and hydrostatic pressures are initially assumed for the host-rock domain. The comparison of the simulations with different permeability realizations indicates that the heterogeneity of the host rock introduces strong differences in the propagation of the gas pressure perturbation, resulting in significant variations in the lateral propagation of the gas front in the host rock and the gas pressure build-up in the emplacement tunnels. Despite these differences, the calculated peak pressures in the disposal cavern and gas breakthrough along the upper model boundary are similar for multiple realizations, which compare well with the simulated results for a homogeneous model with equivalent averaged properties.
DfD: An Approach to New Reactor Designs
Authors: Devgun Jas S.
Published: August 2012

Features relevant to decommissioning are now becoming a part of the design process for the new nuclear power reactors. This is a significant contrast to when the currently operating reactors were designed several decades ago and when decommissioning was not given much thought during the design process. At a time of nuclear renaissance (even with a slowdown in the post-Fukushima accident era), when many reactors are in various construction stages and many more planned, design considerations for decommissioning will have a significant impact on these projects. The goals are to have fewer components, provide modular designs for construction and eventual deconstruction, address major component removal issues, minimize the generation of radioactive waste, and provide for better radiological safety. This paper lays out a framework of the Designing for Decommissioning (DfD) approach.
Uncertainties on the Extension of a Polluted Zone
Authors: de Fouquet Chantal , Benoit Yves , Carpentier Claire , Fricaudet Bruno
Published: August 2012

Data collected during the sampling of polluted sites are mainly used - through an exploratory and variographic analysis, to characterize to characterize the concentration level and the spatial variability; - at fixed support, to estimate the concentrations in order to map the pollution. Kriging gives also the standard deviation of the estimation error, making it possible to delimit the zones in which the estimation is considered to lack in precision. If a proportional effect is present the map of error standard deviation has to be corrected to take into account the increase of spatial variability with the local concentration mean.
Remote Decommissioning of Blind Cells: APM 214 Facility CEA Marcoule
Authors: Calixte Olivier , Garnier Thomas
Published: August 2012

The APM 214 facility (Marcoule Pilot Workshop) owns blind cells which were used in various chemical processing operations. They are now at the end of their working lives and D&D operations have been planned. Due to the high level of radiation these operations must be partially carried out using remote-controlled equipment. The CEA has entrusted ONECTRA to define the best way to dismantle the blind cells with remote-controlled operations. The first step started with an inventory of the main physical characteristics, constraints, nature of remote operations to perform and imagining the machines necessary to complete the tasks at hand. A further step has seen the designing of these specialized machines.
Decommissioning of the Radio Chemical Hot Laboratory of the European Commission Joint Research Centre of Ispra
Authors: Ugolini Daniele , Rossi Francesco , Basile Francesco
Published: August 2012

The construction of the Radio Chemical Hot Laboratory (RCHL) of the Joint Research Centre (JRC) of Ispra began in the early 1960s while the laboratory activities started in 1964. In 1976 an annex to the main building was built. At this time the RCHL main research activities were in environment and biochemistry by means of radioactive tracers; neutron activation analyses; extraction of actinides from radioactive liquid waste coming from the nuclear fuel reprocessing plants; and analyses of U, Pu, and Th in samples from the nuclear fuel cycle in order to determine the isotopic ratio and the burn-up. In 1978, a new area of laboratories named “Stabularium” was built to study the metabolism of heavy metal on laboratory animals. Complementary to the laboratory three pneumatic transfer systems for irradiated sources connected the RCHL to two research reactors. The decommissioning activities of the 2650 m2 facility started in January 2008 and they were completed at the end of 2010 with the release for unrestricted use of all the buildings of the facility. They consisted in five main tasks; pre-decommissioning, licensing, dismantling, waste management, and final survey. The main pre-decommissioning activities were the physical and radiological characterization of the facility. The principal licensing activity was the preparation of the delicensing documentation to obtain the license termination from the safety authorities. Dismantling consisted in the removal of all the equipments and ancillary systems, of the pneumatic transfer system, and in the decontamination of the structures of the controlled zone. The waste management was limited to the transfer of the waste and of the clearable material to the centralized waste management facility. The final survey consisted in the final radiological characterization to quantify the concentration of any residual radioactivity remained after the completion of the dismantling activities for the release of the RCHL without any radiological constraints. The safety and radioprotection prescriptions adopted were the minimization of the conventional and nuclear risk for the workers (reducing the dose rate), the minimization of the environmental risks (reducing the external liquid, solid and gaseous releases), and the confinement of the contamination where it was generated. This paper describes the pre-decommissioning, dismantling, and final survey activities undertaken to perform the decommissioning of the RCHL.
Sampling and Decommissioning: CETAMA Seminar and International Overview
Authors: Granier Guy , Roudil Danie`le , Dubot Didier
Published: August 2012

In France, the nuclear industry has developed a substantial program for decommissioning plants and nuclear installations, and for remediating and rehabilitating industrial sites. The Commission for the Establishment of Analysis Methods (CETAMA) is a unit of the French Alternative Energies and Atomic Energy Commission (CEA) whose main objective is to improve the quality of analysis and measurement results in the nuclear field. Analysis is the primary tool for monitoring the spread of nuclear material. This document shows that sampling, in close relation with measurement techniques, is a factor of potential gain and risk reduction for site remediation, decommissioning, and rehabilitation projects.
Engaging Local Stakeholders on Technical Issues: Test Case at the La Hague Reprocessing Plant
Authors: Gilli Ludivine
Published: August 2012

In 2009 and 2010, the Institute for Nuclear Safety and Radiation Protection (IRSN) lead a pilot action dealing with the decommissioning of a workshop located on the site of Areva’s La Hague fuel-reprocessing plant site in Northwestern France. The purpose of the pilot program was to test ways for IRSN and a few local stakeholders (Non-Governmental Organizations (NGOs) members, local elected officials, etc.) to engage in technical discussions. The discussions were intended to enable the local stakeholders to review the operator’s decommissioning application and provide input. The pilot program confirmed there is a definite challenge in successfully opening a meaningful dialogue to discuss technical issues. Three factors influence the extent of the challenge: the knowledge gap between experts and local stakeholders, the conflict between transparency and confidentiality which is inherent with technical topics, and the difficulty for an official expertise institute to hold a dialogue with “outsiders” during an ongoing reviewing process in which it is participating. The pilot program, given its mixed results, also provided valuable lessons for further improvement of stakeholders’ involvement.
Thermo-Hydro-Mechanical Simulation of a Heating and Hydration Experimental Study (the China-Mock-Up) in GMZ Bentonite
Authors: Chen L. , Wang J. , Liu Y. M. , Collin F. , Xie J. L.
Published: August 2012

This paper presents a numerical study of the China-Mock-up test, with the purpose of evaluating the performance of Gaomiaozi (GMZ) bentonite under coupled thermo-hydro-mechanical (THM) conditions. In the paper, the basic THM characteristics of GMZ bentonite are presented first. The formulation of coupling heat, moisture (liquid water and water vapour) and air transfer is given. The model of Alonso-Gens (1) is incorporated to reproduce the mechanical behavior of the GMZ bentonite under unsaturated conditions. With the parameters determined from experimental investigations, numerical simulations of the China-Mock-up test are carried out using the code of LAGAMINE. Owing to the lack of experimental data at present stage, a qualitative analysis of the predictive results is realized. The results suggest that the proposed model is able to reproduce the mechanical behavior of GMZ bentonite, and to predict moisture and air motions under thermal solicitations.
Division of Nuclear Liabilities Between Different License Holders and Owners
Authors: Lindskog Staffan , Sjo¨blom Rolf
Published: August 2012

Sweden was one of the first six countries to build and operate a nuclear power reactor. Thus, there exists a corresponding legacy in terms of liabilities for decommissioning and waste management of the historic facilities. Compliance with the Polluter Pays Principle (PPP) and its corollary on equity between generations implies that plans for decommissioning must be made and funds set aside for its execution. The need for precision in the cost estimates often governs the timing of the technical planning. Cost estimates are treacherous since cost raisers may be identified and evaluated only after considerable efforts have been made. Further complications and challenges arise as a result of changes that take place between construction and decommissioning of facilities in terms of the entities involved as owners, operators, license holders, Authorities and financiers. From this perspective, the present paper summarizes the general legislation as well as the legislation that applies particularly to nuclear activities. It also summarizes the relation between the nuclear decommissioning fund system and financial reporting. Three examples are provided that wholly or partially fall under the Studsvik act (that specifically covers old facilities): • The A˚gesta nuclear power plant. • The Ranstad uranium mining and beneficiation facility. • The Neutron Research Laboratory at Studsvik. The findings include the following: • It is important that the legislation be clear as to what is included and not. • The rationale for the legislation should also be clear and well communicated. • Old agreements can be significant for the assessment of liabilities, even in cases where a party may no longer exist. • Support for assessment of when activities are continuing or not (which may have a strong significance for the liability) can be found in court cases on chemically contaminated soil. • Analysis of facilities and the work carried out at different times can be very helpful in determining whether or not a facility is auxiliary. • In order to be essentially correct, annual reporting must be coherent with the declarations of the funding system and in compliance with the IAR/IFRS standards. • Keeping of searchable records is essential. • Research is essential, not only to provide bases for high quality decisions, but also to promote consensus based on agreement on factual circumstances.
Nuclide Separation by Hydrothermal Treatment and Ion Exchange: A Highly Effective Method for Treatment of Liquid Effluents
Authors: Braehler Georg , Rieck Ronald , Avramenko V. A. , Sergienko V. I. , Antonov E. A.
Published: August 2012

Liquid low level radioactive effluents, when solidified in e. g. cement matrix, contribute to a significant extent to the waste amount to be disposed of in final repositories. Accordingly, since many years scientists and engineers investigate processes to remove the radioactive nuclides selectively from the effluents, to split the raw solution into two separate fractions: a large fraction with activity concentrations below the limits for free release; and a small fraction, containing the activity in concentrated form on e. g. ion exchanger materials (ion exchange has proven to be the most promising method for such “nuclide separation”). The challenge to be taken up is: When (and this is most often the case) the effluent contains organic materials and complexing agents, the formation of e. g. the 60-Co-EDTA complex prohibits its fixation to the ion exchangers. Accordingly the complexing agent needs to be removed or destroyed. The Institute for Chemistry of the Russian Academy of Sciences has applied the method of hydrothermal treatment (at elevated temperature and pressure, 200 °C, 200 bar), supported by Hydrogen Peroxide oxidation, to allow virtually complete removal of radioactive nuclides on inorganic ion exchangers. Pilot plants have been operated successfully in Russian power stations, and an operational plant has been designed. The method is being extended for an interesting and promising application: spent organic ion exchange resins, loaded up to the medium activity level, represent a serious disposal problem. With the hydrothermal process, in a process cycle, the activity can be stripped from the resins, the organic content is destroyed, and the activity is fixed on an inorganic absorber, well suited for final disposal.