bluuebunny/crossref_metadata_2025 huggingface dataset:
The Hydrogeologic Environment for a Proposed Deep Geologic Repository in Canada for Low and Intermediate Level Radioactive Waste
Authors: Sykes Jonathan F. , Normani Stefano D. , Yin Yong , Jensen Mark R.
Published: August 2012

A Deep Geologic Repository (DGR) for low and intermediate level radioactive waste has been proposed by Ontario Power Generation for the Bruce nuclear site in Ontario, Canada. As proposed the DGR would be constructed at a depth of about 680 m below ground surface within the argillaceous Ordovician limestone of the Cobourg Formation. This paper describes the hydrogeology of the DGR site developed through both site characterization studies and regional-scale numerical modelling analysis. The analysis provides a framework for the assembly and integration of the site-specific geoscientific data and examines the factors that influence the predicted long-term performance of the geosphere barrier. Flow system evolution was accomplished using both the density-dependent FRAC3DVS-OPG flow and transport model and the two-phase gas and water flow computational model TOUGH2-MP. In the geologic framework of the Province of Ontario, the DGR is located on the eastern flank of the Michigan Basin. Borehole logs covering Southern Ontario combined with site-specific data from 6 deep boreholes have been used to define the structural contours and hydrogeologic properties at the regional-scale of the modelled 31 sedimentary strata that may be partially present above the Precambrian crystalline basement rock. The regional-scale domain encompasses an approximately 18500km2 region extending from Lake Huron to Georgian Bay. The groundwater zone below the Devonian includes units containing stagnant water having high concentrations of total dissolved solids that can exceed 300g/L. The Ordovician sediments are significantly under-pressured. The horizontal hydraulic conductivity for the Cobourg limestone is estimated to be 2 × 10−14 m/s based on straddle-packer hydraulic tests. The low advective velocities in the Cobourg and other Ordovician units result in solute transport that is diffusion dominant with Peclet numbers less than 0.003 for a characteristic length of unity. Long-term simulations that consider future glaciation scenarios include the impact of ice thickness and permafrost. Solute transport in the Ordovician limestone and shale was diffusion dominant in all simulations. The Salina formations of the Upper Silurian prevented the deeper penetration of basal meltwater.
Design and Installation of a Permeable Treatment Wall to Passively Remove Strontium-90 From Groundwater at a Former Commercial Nuclear Fuel Reprocessing Facility in New York
Authors: Bablitch Douglas , Malhotra Rupeet , Frappa Rick , Warner Scott
Published: August 2012

A permeable treatment wall (PTW) was designed and installed at the West Valley Development Project (WVDP), a former commercial nuclear fuel reprocessing facility in New York, to passively remove and contain the expansion of strontium-90 (Sr-90) in the site groundwater. AMEC engineers and geologists have collaborated with researchers at the State University of New York at Buffalo (UB) and West Valley Environmental Services LLC to design and install an approximately 260 meter (860-foot) long by 0.9 meter (3-foot) thick zone of granular zeolite (a natural aggregate composed of approximately 85 percent of the mineral clinoptilolite) that will remove Sr-90 in situ from groundwater through ion-exchange reactions. The PTW was designed to meet the functional objectives for up to 20 years; performance monitoring will be conducted regularly and will be used to assess the lifetime efficacy of the PTW. The zeolite-filled PTW was the selected remedial alternative due to its hydraulically passive operations and lower life-cycle cost over other more traditional active treatment alternatives such as pump and treat. The design relied heavily on detailed site characterization of site soils and groundwater conditions and bench and pilot scale evaluations of various zeolite materials. The design specified the use of a one-pass trencher to simultaneously remove unconsolidated aquifer material composed of glaciofluvial-derived silt, sand, and gravel from ground surface to depths up to approximately 9 meters (30 feet) and replace the excavated zone with zeolite along the entire alignment while keying the PTW at least 0.9 meter (3 feet) into the underlying low-permeability glacial till. Several technology demonstrations were conducted to assess implementability using the one-pass trencher prior to completing the final design. During full-scale implementation, excavated sediment was conveyed directly into a prefabricated containment structure pending final disposition. The passive design provides a cost effective and sustainable alternative for treatment of Sr-90 and potentially other exchangeable radioactive ions in groundwater where these constituents migrate in unconsolidated materials.
Evolution of Sorption Properties in Large-Scale Concrete Structures Accounting for Long-Term Physical-Chemical Concrete Degradation
Authors: Perko Janez , Jacques Diederik , Mallants Dirk , Seetharam Suresh
Published: August 2012

Long-term safety of radioactive waste disposal facilities relies on the longevity of natural or engineered barriers designed to minimize the migration of contaminants from the facility into the environment. Especially near surface disposal facilities, such as planned by ONDRAF/NIRAS for the Dessel site in Belgium, long-term safety relies almost exclusively on the containment ability of the engineered barriers (EB) with concrete being the most important EB material used. Concrete is preferred over other materials mainly due to its favourable chemical properties resulting in a high chemical retention capacity, and owing to its good hydraulic isolation properties. However, due to the long time frames typically involved in safety assessment, the chemical, physical and mechanical properties of concrete evolve in time. The alterations in concrete mineralogy also cause changes in pH and sorption behaviour for many radionuclides during chemical degradation processes. Application of dynamic sorption of concrete requires an adequate knowledge of long-term concrete degradation processes, knowledge of the effect of changing mineralogy to sorption of radionuclides and knowledge of large-scale system behaviour over time. Moreover, when applied to safety assessment models, special attention is required to assure robustness and transparency of the implementation. The discussion in this paper focuses on the sorption properties of concrete; selection of data, rescaling issues and on the hypotheses used to build a robust and yet transparent dynamic model for large-scale concrete structures for assessing the long-term performance.
Remediation of Canada’s Historic Haul Route for Radium and Uranium Ores: The Northern Transportation Route
Authors: Geddes Brian , Wenzel Chris , Owen Michael , Gardiner Mark , Brown Julie
Published: August 2012

Established in the 1930s, the Northern Transportation Route (NTR) served to transport pitchblende ore 2,200 km from the Port Radium Mine in Canada’s Northwest Territories to Fort McMurray in Alberta. From there, the ore was shipped 3,000 km by rail to the Town of Port Hope, Ontario, where it was refined for its radium content and used for medical purposes. Later, transport and refinement focussed on uranium. The corridor of lakes, rivers, portages and roads that made up the NTR included a number of transfer points, where ore was unloaded and transferred to other barges or trucks. Ore was occasionally spilled during these transfer operations and, in some cases, subsequently distributed over larger areas as properties were re-developed or modified. In addition, relatively small volumes of ore were sometimes transported by air to the south. Since 1991, the Low-Level Radioactive Waste Management Office (LLRWMO), working with communities and its consulting contractors, has conducted surveys to identify and characterize spill sites along the NTR where soils exhibit elevated concentrations of uranium, radium and/or arsenic. In addition to significant areas of impact in Fort McMurray, contamination along the NTR was centred in the Sahtu region near Great Bear Lake and along the southern part of the Slave River. Early radiological investigations found contaminated buildings and soil and occasionally discrete pieces of pitchblende ore at many transfer points and storage areas along the NTR. Where possible, survey work was undertaken in conjunction with property redevelopment activity requiring the relocation of impacted soils (e.g., at Tulita, Fort Smith, Hay River, and Fort McMurray). When feasible to consolidate contaminated material locally, it was placed into Long Term Management Facilities developed to manage and monitor the materials over extended timelines. Radiological activity generated by these engineered facilities are generally below thresholds established by Canadian regulators, meaning they are straightforward to maintain, with minor environmental and community impacts. Securing community acceptance for these facilities is critical, and represents the predominant development component of plans for managing ore-impacted soils. In those circumstances where local consolidation is not achievable, materials have been relocated to disposal facilities outside of the region. The LLRWMO is continuing a program of public consultation, technical evaluation and environmental assessment to develop management plans for the remaining ore-impacted sites on the NTR. This paper will highlight current activities and approaches applied for the responsible management of uranium and radium mining legacies.
Industrial Scale HL-LL Waste Canister Transfer Demonstrator
Authors: Champ Pascal , Bardet Jacques , Attias Benjamin
Published: August 2012

In 2007, Cegelec has been attributed a contract by ANDRA under the ESDRED program (Engineering Studies and Demonstrations of Repository Designs) financed by 13 organisations across Europe, for the design, manufacturing and trial test of a full scale industrial demonstrator of HL-LL waste canister transfer system. Completed in 2009, this project gives a good synthesis of solutions the industry can offer in order to manage HLW transportation with regard to deep geological disposal. This paper highlights solutions developed and results achieved with numerous pictures of the actual system.
Performance Assessment Modelling for the Geological Siting Regions for the L/ILW and HLW Repositories in the Context of the Swiss Site Selection Plan
Authors: Poller Andreas , Schneider Ju¨rg W. , Zuidema Piet , Holocher Johannes , Mayer Gerhard
Published: August 2012

In Switzerland, the Nuclear Energy Law requires the disposal of all radioactive waste in deep geological repositories. The procedure for selecting the repository sites is defined in the Sectoral Plan for Deep Geological Repositories and consists of three stages. In Stage 1, the National Cooperative for the Disposal of Radioactive Waste (Nagra) proposed geological siting regions based on criteria relating to safety and engineering feasibility. As part of Stage 2, Nagra has to select at least one site within each siting region, to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. In order to achieve these objectives, the state of knowledge of the geological conditions in the siting regions has to be sufficient to perform the provisional safety analyses. In October 2010, Nagra submitted a report which documents Nagra’s technical-scientific assessment of this precondition, based on a comprehensive list of processes and parameters relevant for safety and engineering feasibility. A part of this assessment consists of test calculations for the provisional safety analyses. This paper summarizes how the numerous test calculations have been identified, how the concepts of radionuclide release from the repository are implemented into numerical codes and how input data and results are organized in order to ensure transparency and traceability.
Preparing for the Construction Phase of the Port Hope Area Initiative: Canada’s Largest Low-Level Radioactive Waste Clean-Up Project
Authors: Fahey Christine A. , Case G. Glenn , Denby Andrea T. , Palmeter Tim
Published: August 2012

The Port Hope Area Initiative (PHAI) is a federally-sponsored project to clean-up the historic low-level radioactive waste (LLRW) arising from the operations of Eldorado Nuclear, a former federal crown corporation, and its private sector predecessors. The waste is mostly in the form of contaminated soil and was generated between 1932 and 1988 when Eldorado was dissolved. This paper provides a summary of the origin of the waste, the initial attempt to address the waste issue, and the current initiative which began in 2001. More specifically, this paper provides an introduction to the project and the work accomplished in Phase 1, specific details regarding the ongoing Transition Phase 1A and a sneak peak at the plan for Phase 2 when the construction and remediation activities will be performed and the wastes safely emplaced in new engineered containment mounds.
Design and Operation of the Combined Technology Automated Waste Characterization System
Authors: Mason J. A. , Looman M. R. , Price R. A.
Published: August 2012

This paper describes the design and operation of the Combined Technology Automated Waste Characterisation System (CTA-WCS) at JRC Ispra. The WCS was designed for the measurement of fission products and uranium and plutonium containing waste arising from nuclear fuel and nuclear materials processing and reactor operations. The WCS covers a range of activity including Low and Intermediate Level Waste (LLW and ILW). The system is designed to measure the waste in 200 and 400 (440) litre drums with a maximum drum weight of 1500 kg. Gamma-ray measurements of radio-nuclide content are performed by a gamma ray measurement station which functions as either a Segmented Gamma Scanner (SGS) or Tomographic Segmented Gamma-ray Scanner (TSGS). Either of these two techniques may be employed to perform the functions of drum screening, non-destructive examination (NDE) and, where appropriate, final drum assay. Coupled to the gamma ray station is a surface dose-rate measurement station, which employs 6-shielded Geiger-Muller detectors. Active and passive neutron measurements are performed by an advanced, graphite lined Differential Die-away (DDA) system, which comprises the neutron measurement station. The DDA performs conventional passive neutron totals, coincidence, and multiplicity counting and active DDA total neutron counting. Data analysis is based on the use of a range of matrix calibrations, some determined by Monte Carlo analysis. Linking the gamma ray and neutron measurement stations is an automated roller conveyor with a 20 drum buffer capability and a weight measurement station. Drums are identified by bar code reading technology. Once loaded, the system performs automatic assay of up to 20 drums and then returns the drums to the buffer position on the conveyor. The first WCS of this type was supplied to the European Commission at the Joint Research Centre (JRC) Ispra in northern Italy and it was commissioned at the end of 2007. It is now in a phase of pilot operation. Results will be presented from the first drum measurement campaign.
Decommissioning of Obsolete Nuclear Facilities in the Nuclear Research Institute Rez
Authors: Podlaha Josef , Svoboda Karel , Hansli´k Eduard
Published: August 2012

After more than 55 years of activities of the Nuclear Research Institute Rez (NRI) in the nuclear field, there are some obsolete nuclear facilities that shall be decommissioned. NRI is a leading institution in all areas of nuclear R&D in the Czech Republic. NRI has had a dominant position in the nuclear programme since it was established in 1955 as a state-owned research organization and it has developed to its current status. In December 1992, NRI has been transformed into a joint-stock company. The Institute’s activity encompasses nuclear physics, chemistry, nuclear power, experiments at the research reactor and many other topics. Main issues addressed in NRI in the past decades were concentrated on research, development and services provided to the nuclear power plants operating VVER reactors, development of chemical technologies for fuel cycle and irradiation services to research and development in the industrial sector, agriculture, food processing and medicine. The NRI operates two research nuclear reactors, many facilities as a hot cell facility, research laboratories, technology for radioactive waste (RAW) management, radionuclide irradiators, an electron accelerator, etc. The obsolete facilities to be decommissioned comprise various research facilities and facilities for RAW management. Decommissioning of nuclear facilities NRI is the only ongoing decommissioning project in the Czech Republic. Decommissioning started in 2003 and will be finished in 2014. Some facilities have already been successfully decommissioned.
Implementation of the International Structure for Decommissioning Costing: Examples and Related IAEA Projects
Authors: Daniska Vladimir , Laraia Michele , O’Sullivan Patrick
Published: August 2012

In 1999, IAEA, the OECD Nuclear Energy Agency and the European Commission jointly proposed the standardised listing of decommissioning activities [1] to serve as a general basis for presentation of decommissioning costs and for promoting the harmonisation in decommissioning costing. The standardised listing of activities [1] was developed in three hierarchical levels based on analysis of typical decommissioning activities identified in various decommissioning projects. The structure [1] has been currently updated by the same organisations as the International Structure for Decommissioning Costing (ISDC) based on the experience gained over ten years of use of the original standardised listing [2]. First part of the paper presents the revised ISDC. The principle of the three-level original hierarchical structure has been preserved. Re-definition of the content and re-structuring was done to avoid ambiguity and to ensure comprehensiveness. Paper presents two basic approaches for implementation of the ISDC structure in costing — converting the cost data available in specific cost structures, mostly according the work breakdown structures of decommissioning projects into ISDC and implementation of the ISDC as the cost calculation structure. Examples of the second approach are given to show that this approach is feasible and may have several advantages. An ORACLE based costing model with implemented of the extended ISDC for detailed costing and an Excel based costing model for preliminary costing at IAEA for research reactors are given.
Laboratory Experiments on the Alteration of Highly-Compacted Bentonite by Alkaline Solution and the Effects on Physical Properties
Authors: Miyoshi Satoru , Morikami Shinya , Kimura Yukinobu , Jinno Tomoko , Yamamoto Shuichi
Published: August 2012

The laboratory experiment was done that 1.0mol/L sodium hydroxide solution was injected to the compacted bentonite whose density is the same as the prospected value in the concept of the intermediate-level disposal in Japan in the circumstance of 70°C temperature. After the injection of the alkali solution for approximately 600 days, the bentonite was taken out of the apparatus and some sorts of analysis were done. The accompanying minerals in the bentonite, calcedony and quartz, were dissolved and disappeared in XRD charts. Then analcime was precipitated as a secondary mineral. Although montmorillonite was dissolved, the mass fraction of it was kept approximately. The hydraulic conductivity of the bentonite calculated using the flow rate at the end of the injection of alkali solution was smaller than the prospected value based on a widely-used empirical model of the hydraulic conductivity of compacted bentonite as a function of the equivalent concentration of pore solution, montmorillonite partial void ratio, and the ratio of sodium ion equivalent to the exchangeable cation equivalent. The reasons for the difference were supposed to be the decrease of pore size brought by mineral dissolution and the large viscosity of pore solution involving high concentration aqueous silicon.
Spanish Approach to Large Decommissioning Projects
Authors: Santiago Juan Luis , Rodri´guez Alejandro
Published: August 2012

The Spanish experience related to the decommissioning of nuclear facilities includes the decommissioning of the Vandello´s I Nuclear Power Plant, the decommissioning of the CIEMAT Nuclear Research Centre and the decommissioning of the Jose´ Cabrera Nuclear Power Plant. This paper reviews the key aspects of these projects and describes the lessons learned related to preparatory activities, auxiliary facilities, decommissioning technologies, material management and site remediation and release.
The Decommissioning of the CIEMAT Nuclear Research Center
Authors: Vidaechea Sergio , Ondaro Manuel
Published: August 2012

CIEMAT, formerly the Nuclear Energy Board (JEN), is the Spanish Centre for Energy-Related, Environmental and Technological Research. Located in Madrid, it used to have more than 60 facilities in operation that allowed a wide range of activities in the nuclear field and in the application of ionising radiations. Particularly significant among these facilities were the research reactors, particle accelerators, hot cells and nuclear fuel manufacturing and processing plants. At present the centre, which is authorised as a single nuclear facility, includes various installations, some of them are now obsolete, shut down and in dismantling phases. In 2000 CIEMAT started the “Integrated plan for the improvement of CIEMAT installations (PIMIC)”, which includes activities for the decontamination, dismantling and rehabilitation of obsolete facilities. This paper will describe the decommissioning process carried out at CIEMAT Research Centre, and will review the key aspects of the project, including site remediation and release.
Radium Diagnosis Campaign
Authors: Gabillaud-Poillion Florence
Published: August 2012

In line with the approaches already adopted in France during the 90s on various sites where research and/or radium-extraction activities were mostly conducted in the past, the French public authorities wish from now on to pursue their prevention and site-rehabilitation approach inherited from the French craftsman and medical sectors that used that radioelement. As a matter of fact, radium has been in use in several medical activities, notably in the initial methods of cancer therapy. Similarly, it was also used in some craftsman activities, such as the clock industry, for its radioluminescent properties, the fabrication of lightning conductors or cosmetics until the 60s. Those activities have generated various traces of pollution that have remained today. On the basis of the different inventories of industrial sites where radium may have been held or used, and notably the inventory updated by the French Institute for Radiation Protection and Nuclear Safety (Institut de radioprotection et de suˆrete´ nucle´aire – IRSN) in 2007 at the request of the French Nuclear Safety Authority (Autorite´ de suˆrete´ nucle´aire – ASN), French State services have potentially identified 134 sites that hosted radium-related activities in France. The radiological status of those sites is either unknown or very partially known by State services. Sites include both dwellings or commercial premises and derelict lands. The “Radium Diagnosis Campaign” (Operation Diagnostic Radium), consists of a radiological survey carried out by the IRSN. In cases where traces of radium are detected, plans call for the implementation of precautionary measures and of a medical follow-up of the relevant populations. Lastly, radium-contaminated sites are rehabilitated by the French National Radioactive Waste Management Agency (Agence nationale pour la gestion des de´chets radioactifs – Andra). That voluntary and positive approach on the part of public authorities is fully financed by public funds, and consequently, at no cost for the tenants of the premises involved, whether it involves the diagnosis, the individual health follow-up or the rehabilitation. The first diagnosis phase, which focused on the Iˆle-de-France Region (Paris and suburbs), was launched in September 2010. At the end of the year, six sites grouping a total of 40 premises or dwellings had already been fully surveyed. Traces of pollution were detected on 10 of those premises where rehabilitation activities are already under way. So far, that approach has been received favourably by the populations concerned.
Laser Removal of Contaminants on the Metal Surface
Authors: Won Hui-Jun , Jung Sun-Hee , Jung Chong-Hun , Choi Byung-Seon , Lee Kune-Woo , Moon Jei-Kwon
Published: August 2012

A Q-switched Nd:YAG laser with a 532 nm and 450 mJ/pulse was employed to study the decontamination characteristics of Type 304 stainless steel specimens and aluminum specimens artificially contaminated with CsNO3, Co(NH4)2(SO4)2, Eu2O3 and CeO2, respectively. The relative atomic molar percent of the contaminants on metal surface before and after laser irradiation was determined by EPMA. The morphology of specimen surface was examined by SEM. The ablation behavior of contaminants on Type 304 stainless steel specimens was investigated for the decontamination variables such as a number of laser shots, laser fluence and an irradiation angle. Their optimum values were found to be 8, 13.3 J/cm2 and 30°, respectively. The removal of contaminants on aluminum surface, however, was found to be more difficult by laser.
Geostatistical Methodology for Waste Optimization of Contaminated Premises
Authors: Desnoyers Yvon , Dubot Didier
Published: August 2012

The presented methodological study illustrates a geostatistical approach suitable for radiological evaluation in nuclear premises. The waste characterization is mainly focused on floor concrete surfaces. By modeling the spatial continuity of activities, geostatistics provide sound methods to estimate and map radiological activities, together with their uncertainty. The multivariate approach allows the integration of numerous surface radiation measurements in order to improve the estimation of activity levels from concrete samples. This way, a sequential and iterative investigation strategy proves to be relevant to fulfill the different evaluation objectives. Waste characterization is performed on risk maps rather than on direct interpolation maps (due to bias of the selection on kriging results). The use of several estimation supports (punctual, 1 m2, room) allows a relevant radiological waste categorization thanks to cost-benefit analysis according to the risk of exceeding a given activity threshold. Global results, mainly total activity, are similarly quantified to precociously lead the waste management for the dismantling and decommissioning project.
Self Cleaning High Efficiency Particulate Air (HEPA) Filtration Without Interrupting Process Flow
Authors: Chadwick Chris
Published: August 2012

The strategy of protecting the traditional glass fibre HEPA filtration train from it’s blinding contamination and the recovery of dust by the means of self cleaning, pre-filtration is a proven means in the reduction of ultimate disposal volumes and has been used within the Fuel Production Industry. However, there is an increasing demand in nuclear applications requiring elevated operating temperatures, fire resistance, moisture resistance and chemical composition that the existing glass fibre HEPA filtration cannot accommodate, which can be remedied by the use of a metallic HEPA filter media. Previous research suggests that the then costs to the Department of Energy (DOE), based on a five year life cycle, was $29.5 million for the installation, testing, removal and disposal of glass fibre HEPA filtration trains. Within these costs, $300 was the value given to the filter and $4,450 was given to the peripheral activity. Development of a low cost, cleanable, metallic, direct replacement of the traditional filter train will the clear solution. The Bergman et al work has suggested that a 1000 ft^3/min, cleanable, stainless HEPA could be commercially available for $5,000 each, whereas the industry has determined that the truer cost of such an item in isolation would be closer to $15,000. This results in a conflict within the requirement between ‘low cost’ and ‘stainless HEPA’. By proposing a system that combines metallic HEPA filtration with the ability to self clean without interrupting the process flow, the need for a tradition HEPA filtration train will be eliminated and this dramatically reduces the resources required for cleaning or disposal, thus presenting a route to reducing ultimate costs. The paper will examine the performance characteristics, filtration efficiency, flow verses differential pressure and cleanability of a self cleaning HEPA grade sintered metal filter element, together with data to prove the contention.
Sampling vs. Taking Some
Authors: Francois-Bongarcon D.
Published: August 2012

Collecting a sample is a delicate task that is NOT naively equivalent to simply “taking some of the material”. The question examined is: “What is it exactly?” The problem of sampling in general, and for nuclear decontamination in particular, is properly defined. A theory is presented (Gy’s Theory of Sampling, a.k.a. TOS) that brings all the answers and allows us to put them to work. The author draws form his lifelong experience in research, teaching and practical applications in this domain to emphasize the critical odds (i.e. risks) of not taking sampling explicitly into account when assessing grades and concentrations. The evolution of the acceptance of this theory in the nuclear industry is finally illustrated, and a hopeful glimpse into the future concludes the presentation.
Regulation of Ageing Reprocessing Facilities in the UK
Authors: Maitland R. P.
Published: August 2012

The UK’s strategy for spent Magnox reactor fuel demands continued operation of the Magnox Reprocessing facility at Sellafield (located in the North West of England) to reprocess the remaining spent fuel in the shutdown Magnox reactor stations and from the two remaining operational Magnox reactor stations, Wylfa and Oldbury. Safety, security, environmental, transport, energy and economic issues provide the initiative to continue reprocessing in ageing facilities that are prone to chronic operational and nuclear safety challenges. One of the responsibilities of the UK’s Office for Nuclear Regulation is to regulate the safety of continuing Magnox Reprocessing Operations against relevant health and safety legislation; this largely non-prescriptive framework requires dutyholders to demonstrably reduce risk so far as is reasonably practicable. This paper articulates the often complex balances that have to be made to demonstrate compliance with safety law to sustain continued operation of ageing reprocessing facilities. This paper details how the UK’s regulatory framework facilitates a flexible, proportionate and goal-setting approach to regulating operational facilities where it is difficult to satisfy relevant good practice or standards that would be expected of a modern facility. The challenges presented by regulation of ageing, operational facilities is analogous to those from legacy waste retrieval and decommissioning; this paper reflects the versatility of the UK’s regulatory approach to these two different areas of the fuel cycle.
Experience of Application of New Remote Controlled Instruments for Scanning of Distribution of Radioactive Contamination in Rooms With High Dose Rate
Authors: Stepanov V. E. , Danilovich A. S. , Ivanov O. P. , Potapov V. N. , Smirnov S. V. , Volkovich A. G.
Published: August 2012

When carrying out the decommissioning of nuclear facilities is necessary to measure the distribution of radioactive contamination in the rooms and the equipment at high levels of background radiation. In the decommissioning of the reactor in the MR NRC “Kurchatov Institute” for such problems is developed and applied a few special systems with remote control. For a survey of high-level objects used mounted on the robot radiometric system. To determine the composition of pollutants and use a portable collimated spectrometric system. To obtain a detailed distribution of contamination of used remote-controlled gamma camera.
An Assessment of the Radiological Impact of Human Intrusion at the UK Low Level Waste Repository (LLWR)
Authors: Hicks Tim. , Baldwin Tamara , Cummings Richard , Sumerling Trevor
Published: August 2012

The UK Low Level Waste Repository Ltd submitted an Environmental Safety Case for the disposal of low-level waste (LLW) to the Environment Agency on the 1st of May 2011. The Environmental Safety Case (ESC) presents a complete case for the environmental safety of the Low Level Waste Repository (LLWR) both during operations and in the long term (Cummings et al, in these proceedings). This includes an assessment of the long-term radiological safety of the facility, including an assessment of the potential consequences of human intrusion at the site. The human intrusion assessment is based on a cautiously realistic approach in defining intrusion cases and parameter values. A range of possible human intrusion events was considered based on present-day technologies and credible future uses of the site. This process resulted in the identification of geotechnical investigations, a housing development and a smallholding as requiring quantitative assessment. A particular feature of the site is that, because of its proximity to the coast and in view of expected global sea-level rise, it is vulnerable to coastal erosion. During such erosion, wastes and engineered barrier materials will be exposed, and could become targets for investigation or recovery. Therefore, human intrusion events have been included that are associated with such activities. A radiological assessment model has been developed to analyse the impacts of potential human intrusion at the site. A key feature of the model is the representation of the spatial layout of the disposal site, including the engineered cap design and the large-scale spatial heterogeneity of radionuclide concentrations within the repository. The model has been used to calculate the radiation dose to intruders and to others following intrusion at different times and at different locations across the site, for the each of the selected intrusion events, considering all relevant exposure modes. Potential doses due to radon and its daughters in buildings constructed on excavated spoil from the repository are a particular concern. Options for managing the emplacement of the radium-bearing waste packages with regard to human intrusion have been assessed. These calculations show that a managed waste emplacement strategy can ensure that calculated doses are consistent with regulatory guidance levels.
The Global Threat Reduction Initiative’s Radiological Security Cooperation With Russia
Authors: Blanchard Tiffany A. , Abramson William J. , Russell James W. J. R. , Roberts Catherine K.
Published: August 2012

The United States (U.S.) Department of Energy (DOE) / National Nuclear Security Administration’s (NNSA) Global Threat Reduction Initiative (GTRI) supports both U.S. and international threat reduction goals by securing vulnerable nuclear and radiological material located at civilian sites throughout the world. GTRI’s approach to reducing the threat posed by vulnerable, high-activity radioactive sources includes removing and disposing of orphan or disused radioactive sources; implementing physical security upgrades at civilian sites containing radioactive sources; and establishing a cooperative sustainability program at sites to ensure that upgrades are maintained. For many years GTRI has collaborated successfully with the Russian Federation and international partners to improve radiological security in Russia. This paper provides a synopsis of GTRI’s accomplishments and cooperation with Russia in the following areas: 1.) recovering and disposing of orphan and disused radioactive sources, 2.) recovering and disposing of radioisotope thermoelectric generators (RTGs), and 3.) providing physical security upgrades at civilian sites that contain vulnerable radiological material.
Overview of Radium Legacies in Belgium
Authors: Dehandschutter B. , Jadoul L. , Mannaerts K. , Pepin S. , Poffijn A. , Blommaert W. , Sonck M.
Published: August 2012

The Belgian metallurgical company, Union Minie`re, has been a key-player in the sector of radium production between 1922 and 1969. The factory based in Olen has extracted radium from minerals and produced radium sources during that period. The radium production facilities have been dismantled in the 70s but legacies of the former production have still to be remediated. An overview of these legacies and of their radiological characteristics will be given. Next to the sites related to radium production, other radium legacies are related to NORM industries, essentially from the phosphate sector (phosphogypsum and CaF2 stacks).
Methodological and Practical Bases of Providing Information Support to Activities on Environmental Remediation of the Spent Nuclear Fuel and Radioactive Waste Temporary Storage Facility in Gremikha
Authors: Stepennov D. B. , Varnavin A. P. , Zakharchev A. A. , Pillette-Cousin L.
Published: August 2012

Remediation of a spent nuclear fuel (SNF) and radioactive waste (RW) temporary storage facility is a multifaceted process that includes a number of stages, such as development of a remediation programme, performance of comprehensive engineering and radiological survey, development of a remediation design, removal of SNF and RW up to the site cleanup. At any stage of the remediation, making of justified decisions is ensured by availability and completeness of associated information. Huge amount of information has to be managed. Therefore an information analysis system (IAS) was developed by the National Research Centre «Kurchatov Institute» within the framework of the project for environmental remediation of the SNF and RW temporary storage facility in Gremikha with financial and technical support provided by France (CEA) and the Russian Federation (Rosatom). The IAS accumulates all information about the project: technical and radiological characteristics of objects/facilities, cartographic information, documentation, data on the project participants, technologies and equipment involved. The IAS architecture includes the following functional subsystems: data management, data analytical processing, project management, geoinformation, 3D modeling, and public information. The IAS allows developers and performers of environmental remediation of the SNF and RW temporary storage facility in Gremikha to fulfill tasks arising at all stages of the work. The IAS operating experience can be transferred for use during surveys and remediation of any radiation hazardous facilities.
Human Reliability-Based MC&A Methods for Evaluating the Effectiveness of Protecting Nuclear Material
Authors: Dura´n Felicia A. , Wyss Gregory D.
Published: August 2012

Material control and accountability (MC&A) operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. MC&A activities, from monitoring to inventory measurements, provide critical information about target materials and define security elements that are useful against insider threats. However, these activities have been difficult to characterize in ways that are compatible with the path analysis methods that are used to systematically evaluate the effectiveness of a site’s protection system. The path analysis methodology focuses on a systematic, quantitative evaluation of the physical protection component of the system for potential external threats, and often calculates the probability that the physical protection system (PPS) is effective (PE) in defeating an adversary who uses that attack pathway. In previous work, Dawson and Hester observed that many MC&A activities can be considered a type of sensor system with alarm and assessment capabilities that provide reccurring opportunities for “detecting” the status of critical items. This work has extended that characterization of MC&A activities as probabilistic sensors that are interwoven within each protection layer of the PPS. In addition, MC&A activities have similar characteristics to operator tasks performed in a nuclear power plant (NPP) in that the reliability of these activities depends significantly on human performance. Many of the procedures involve human performance in checking for anomalous conditions. Further characterization of MC&A activities as operational procedures that check the status of critical assets provides a basis for applying human reliability analysis (HRA) models and methods to determine probabilities of detection for MC&A protection elements. This paper will discuss the application of HRA methods used in nuclear power plant probabilistic risk assessments to define detection probabilities and to formulate “timely detection” for MC&A operations. This work has enabled the development of an integrated path analysis methodology in which MC&A operations can be combined with traditional sensor data in the calculation of PPS effectiveness. Explicitly incorporating MC&A operations into the existing evaluation methodology provides the basis for an effectiveness measure for insider threats, and the resulting PE calculations will provide an integrated effectiveness measure that addresses both external and insider threats. The extended path analysis methodology is being further investigated as the basis for including the PPS and MC&A activities in an integrated safeguards and security system for advanced fuel cycle facilities.
Environmentally-Friendly Recovery of Uranium From Nuclear Fuel Scraps Generated at Nuclear Fuel Fabrication Process
Authors: Kim Kwang-Wook , Chung Dong-Youg , Lee Eil-Hee , Lee Keun-Young , Lee Kune-Woo , Moon Jei-Kwon
Published: August 2012

A way to recover pure uranium as UO4 from UO2 and (U,Gd)O2 nuclear fuel scraps by using carbonate solution with hydrogen peroxide was studied and the nuclear fuel fabrication characteristics of the UO2 powder prepared from the UO4 was evaluated.
High Waste Loading Glass Formulations for Hanford High-Aluminum High-Level Waste Streams
Authors: Kruger Albert A.
Published: August 2012

The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP’s overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33–50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.
Kola Peninsula Radwaste Management in the Framework of the Italian-Russian Cooperation Agreement for Global Partnership
Authors: Nobile Massimiliano , Izmailov Dmitry , Spadoni Antonino , Uryvskiy Oleg
Published: August 2012

In 2003 a Cooperation Agreement, envisaging a grant of 360 Meuro in 10 years, was signed between Italy and Russia in the framework of the “so-called” Global Partnership Program. So far, in five years of concrete work, a total of 146 Meuro have been engaged through 33 contracts and 121 Meuro have been already paid. Together with other significant lines of activities the management of waste, presently located at Andreeva Bay site, is surely the most important one, not only for the amount of allocated funds (about the half of available ones) but also for the urgency to solve serious environmental problems linked to the actual unsafe conditions of these waste, both as regards possible incidents or malicious attacks from outside. Volume of Low Level and Intermediate Level (LL/IL) waste at Andreeva Bay site are 17000 tons and 3000 tons, respectively for solid and liquid; moreover is expected, during the next 15 years, the production of additional 33000 tons of liquid and 8000 tons of solid, due to the need of demolishing existing facilities and final closing of the site. As regards radwaste management facilities, the solution defined in OBIN (justification of investment) has been analyzed and critically reviewed. After the completion of a conceptual study, the construction of three facilities (solid and liquid treatment ones and a protective structure for storing on site temporarily conditioned waste) was decided and a contract was assigned in 2010 to Ansaldo Nucleare/AtomStroyExport for the detailed design of the above mentioned facilities, which is expected to be completed in 2012. Some important modifications to the initial time schedule have been taken into account and now the operation of the facilities is expected by 2015, two years before the previous estimate.
New Generation of Diamond Toolings for Facilitating Decommissioning Operations
Authors: Besson Claude , Mariotti Erico , Mouysset Alexandre , zur Nedden Lorenz , Delannay Bernard
Published: August 2012

Diamond tools are well proven cutting, drilling and grinding technologies in many applications but need to be specifically optimized and adapted for the complex and varied structures of nuclear power plant in view of decontamination and decommissioning. The proper development and use of diamond tools in these extreme and complex conditions can only be achieved thanks to the combined talent of experienced nuclear plant contractors, engineers, technicians, operators of diamond tools, and the use of specialized equipment. This present paper is an overview of the possible applications of diamond tools in the different operations of Nuclear Decommissioning and Decontamination.