bluuebunny/crossref_metadata_2025 huggingface dataset:
A Nationwide Modelling Approach to Decommissioning
Authors: Kelly Bernard , Mort Paul , Lowe Andy
Published: December 2010

In this paper we describe a proposed UK national approach to modelling decommissioning. For the first time, we shall have an insight into optimizing the safety and efficiency of a national decommissioning strategy. To do this we use the General Case Integrated Waste Algorithm (GIA), a universal model of decommissioning nuclear plant, power plant, waste arisings and the associated knowledge capture. The model scales from individual items of plant through cells, groups of cells, buildings, whole sites and then on up to a national scale. We describe the national vision for GIA which can be broken down into three levels: 1) the capture of the chronological order of activities that an experienced decommissioner would use to decommission any nuclear facility anywhere in the world — this is Level 1 of GIA; 2) the construction of an Operational Research (OR) model based on Level 1 to allow rapid what if scenarios to be tested quickly (Level 2); 3) the construction of a state of the art knowledge capture capability that allows future generations to learn from our current decommissioning experience (Level 3). We show the progress to date in developing GIA in levels 1 & 2. As part of level 1, GIA has assisted in the development of an IMechE professional decommissioning qualification. Furthermore, we describe GIA as the basis of a UK-Owned database of decommissioning norms for such things as costs, productivity, durations etc. From level 2, we report on a pilot study that has successfully tested the basic principles for the OR numerical simulation of the algorithm. We then highlight the advantages of applying the OR modelling approach nationally. In essence, a series of “what if…” scenarios can be tested that will improve the safety and efficiency of decommissioning.
Delivering Step Change Improvements to UK Low Level Waste Strategy
Authors: Dean Jason , Rossiter David
Published: December 2010

The UK Nuclear Industry continues to produce significant quantities of Low Level Waste (LLW) as decommissioning projects generating waste become more prevalent. Current infrastructure and projected increasing waste volumes will deliver a volumetric shortfall of storage capacity in the near future. Recently established as a stand alone site licence company, the Low Level Waste Repository (LLWR) near Drigg, in West Cumbria (formerly operated and owned by British Nuclear Group) is tasked with managing the safe treatment and disposal of LLW in the UK, on behalf of the Nuclear Decommissioning Authority (NDA). The problem is complex involving many stakeholders with potentially different priorities. Previously, most nuclear waste generators operated independently with limited integration with other similar organisations. However, the current financial, programme and technical pressures require collaborative working to facilitate a step-change improvement in LLW management. Achieving this quickly is as much of a challenge as delivering robust cost effective technical solutions. NDA is working in partnership with LLWR to develop a LLW Strategy for the Nuclear Industry and has in parallel commissioned a number of studies by the National Nuclear Laboratory (NNL), looking at opportunities to share best practice. A National Strategy Group has been established to develop a working partnership between the Nuclear Decommissioning Authority, LLW Repository Ltd, Regulators, Stakeholders and LLW Consignors, promoting innovation, value for money, and robust implementation of the waste hierarchy (avoid-reduce-re-use-recycle). Additionally the LLWR supported by the NNL have undertaken a comprehensive strategic review of the UK’s LLW management activities. Initial collaborative work has provided for the first time a detailed picture of the existing strategic baseline and identified significant national benefits from improving the way LLW is forecasted, characterised, segregated, and treated in line with the waste hierarchy. Implementation of volume reduction technologies, such as incineration and metal treatment, is critical to mitigate the LLWR capacity gap and reduce NDA’s liabilities. The cumulative effect of these solutions has the potential to reduce lifetime costs by several £billion and extend the life of the existing LLWR site to 2070 and possibly beyond. This work has informed the NDA’s UK Nuclear Industry LLW Strategy, published for consultation in June 2009 and the Draft UK LLW Management Plan which sets out how the strategy will be implemented. Technical and infrastructure solutions have been found to exist via the supply chain supporting deliver of the necessary step changes in the near future. Work continues to reduce the LLW inventory forecast uncertainties and evaluate strategic implementation options in more detail, e.g. benefits of national vs. local treatment and disposal solutions, plus on gaining the corresponding stakeholder acceptance and operational authorisations.
WIPP: A Perspective From Ten Years of Operating Success
Authors: Gregory Phillip C.
Published: December 2010

The Waste Isolation Pilot Plant (WIPP), located 35 miles east of Carlsbad, New Mexico, USA is the first and, to the author’s knowledge, only facility in the world for the permanent disposal of defense related transuranic (TRU) waste. Soon after plutonium was first synthesized in 1940 by a team of scientists at the University of California Berkley Laboratory, the need to find a permanent repository for plutonium contaminated waste was recognized due to the more than 24,000 year half-life of Plutonium-239 (239Pu). In 1957 the National Academy of Sciences published a report recommending deep geological burial in bedded salt as a possible solution. However, more than 50 years passed before the solution was achieved when in 1999 WIPP received the first shipment of TRU waste from Los Alamos National Laboratory. Ten years later, more than 7,600 shipments of TRU waste have been disposed of in rooms mined in an ancient salt bed more than 2,000 feet underground. This paper provides a brief history of WIPP with an overview of the technical, regulatory, and political hurdles that had to be overcome before the idea of a permanent disposal facility became reality. The paper focuses primarily on the safe, uneventful transportation program that has moved 100,000-plus containers of TRU waste from various U.S. Department of Energy (DOE) generator and/or storage sites across the Unites States to southeastern New Mexico.
Analysis of Management and Disposal Alternatives for Low Activity Radioactive Waste
Authors: Anderson Keith D.
Published: December 2010

Government and commercial decommissioning and remediation activities at nuclear sites and facilities often generate in large volumes and mass radioactive waste that is of low activity. In large part, the low activity radioactive waste generated from decommissioning and remediation activities are below regulatory limits for low level radioactive waste, while not meeting site specific dose and risk-based cleanup criteria. Policies and strategies for the remediation, management, and disposal of low activity radioactive waste in these circumstances often explore alternative options to fully licensed disposal. Primary goals in exploring alternative disposal options are protection of the public and the environment, while achieve a cost-effective solution. Vague regulations and arbitrary interpretation by federal and state regulators may enhance alternative disposal options or quash efforts by those conducting the remediation and decommissioning. Fundamental to the issues is the overriding concern of a lack of clarity and increased liability in the regulatory structure of the grey area that is low activity radioactive wastes. This paper explores the current efforts by the United States of America and International regulatory community to better define low activity radioactive waste and to provide effective and protective waste management and disposal policies and strategies.
The Effect of Waste Loading on the Characteristics of Borosilicate SRS SB4 Waste Glasses
Authors: Stefanovsky Sergey , Kobelev Alexander , Lebedev Vladimir , Polkanov Michael , Suntsov Dmitriy , Marra James
Published: December 2010

The glasses containing surrogate of Sludge Batch 4 (SB4) waste with high concentrations of aluminum and ferrous oxides within the waste loading (WL) range of ∼50 to 70 wt.% produced in alumina crucibles in a resistive furnace and in a 236 mm inner diameter cold crucible were examined by X-ray diffraction, electron microscopy and infra-red spectroscopy. At relatively low WLs (50 to 60 w.%) the products remain predominantly vitreous with minor crystalline spinel structure phase corresponding in chemical composition to trevorite-magnetite solid solution. At higher WLs (>60 wt.%) amount of crystalline phases increased significantly and a nepheline phase was observed. The products with WL of 65 to 70 wt.% were spinel/nepheline glass-ceramics. The glass with 60 wt.% WL slowly cooled in accordance with the canister centerline cooling (CCC) conditions contained minor nepheline. Occurrence of nepheline reduces the chemical durability of nuclear waste glasses.
Cold Crucible Vitrification of SRS SB4 Waste at High Waste Loadings
Authors: Stefanovsky Sergey , Kobelev Alexander , Lebedev Vladimir , Knyazev Oleg , Polkanov Michael , Marra James
Published: December 2010

The test on determination of maximized waste loading was performed at the Radon bench-scale facility equipped with a 236 mm inner diameter cold crucible. Waste surrogate was vitrified using a commercially available Frit 503-R4 (in wt.%: 8 Li2O, 16 B2O3, 76 SiO2). Waste loading ranged between ∼45 and 70 wt.%. Viscosity of the melt with 50 wt.% waste loading remained lower that 100 Poise (maximum value for glass melting) even at temperatures below 1300 °C. Electric resistivity value for this glass allowed electric melting at temperatures of 1100–1150 °C and higher. Melt viscosity increased with the increase of waste loading and glass with 60 wt.% waste loading had appropriate viscosity for melting at temperatures over 1350 °C. This glass was much “shorter” than the glass with 50 wt.% waste loading. Its electric resistivity at the same temperature was higher than that of glass with 50 wt.% waste loading and allowed this glass to be produced at temperatures 1250 °C and lower, but the high viscosity was a restricting factor and this glass required temperatures ∼ 1400 °C and over to be produced. The melt with 70 wt.% waste loading was too inhomogeneous and viscous and its viscosity and electric resistivity could not be measured precisely. As a result the glasses with 70 and 60 wt.% waste loadings required melting temperature as high as 1350–1400 °C whereas the glass with 50 wt.% waste loading may be produced at 1150–1250 °C. Wet slurry (∼50 and ∼70 wt.% water content) in the amount of 625.1 kg was processed and glass in the amount of 186.6 kg was produced and poured into 12 canisters for 91 hrs. Average slurry feed rate, glass productivity and specific glass productivity were 6.86 kg/hr, 2.16 kg/hr, and 68.8 kg/(m2xhr) respectively.
Algorithmisation of Dismantling Techniques in Standardised Decommissioning Costing Using the Standardised Cost List
Authors: Beza´k Peter , Danisˇka Vladimi´r , Reha´k Ivan , Necˇas Vladimi´r
Published: December 2010

Recently developed computer code OMEGA for evaluation and optimisation of decommissioning options implements the standardised cost structure (IAEA, OECD/NEA, EC, 1999) as the universal structure [PSL] for calculation and optimisation of decommissioning scenarios. One of the groups of decommissioning activities is the dismantling of systems and structures. These activities variously depend on the complexity of dismantled systems, e.g. reactors, equipment of the primary systems or standard components like pipes, valves, motors, tanks etc. Type and extent of decommissioning activities depends also on local conditions for dismantling like dose rate, decommissioning equipment category, local working conditions, etc. These factors determine the selection of techniques for dismantling, depending on material and radiological status of the equipment (type of technique and their manual or remote application). This approach enables proper planning and performing of individual decommissioning phases.
Modelling Long-Term Corrosion of Cemented Waste Forms in Salt Brines
Authors: Kienzler Bernhard , Metz Volker
Published: December 2010

Since 1979, leaching and corrosion experiments have been performed at the Asse II salt mine using simulated full-scale cemented waste forms. The cement blocks, doped with 137Cs and uranium, have been exposed to saturated salt brines. The brines have been sampled regularly and analysed with respect to the mobilized radionuclides, pH, and major solution components. In 2006, four of the cement blocks samples were recovered and solid samples were obtained by drilling vertically into the corroded cement forms. The drill cores as well as abraded drill dust were analysed with respect to chemical and mineralogical compositions, distributions of radionuclides and major waste components, thermogravimetric and mechanical properties. Results of these methods show consistently that cement forms, having an initial W/C = 0.5, were homogeneously corroded in MgCl2 rich brine. No vertical or radial variations were found with respect to major chemical components and mineral phases. In the cement blocks corroded in NaCl saturated brine, spatial heterogeneities in the distribution of major anions (chloride, nitrate) and the trace cation Cs+ show that these cement / NaCl systems have not been completely homogenized within 22 years. Concentration profiles of Cl−, NO3− and Cs+ were compared with calculated profiles to quantify diffusion coefficients and to get information on the kinetics of relevant processes. Simulations of chemical alteration of the cement / brine system indicate that both MgCl2-rich and NaCl systems are close to equilibrium with respect to stabilities of secondary mineral phases.
The Collaborative EC Project ReCosy
Authors: Kienzler Bernhard , Buckau Gunnar
Published: December 2010

The Collaborative Project (CP) ReCosy (Redox Controlling Systems) was established in 2008. It falls within the EURATOM program and is implemented within the European Commission’s 7th Framework Program. Main objectives of CP ReCosy are the sound understanding of redox phenomena controlling the long-term release/retention of radionuclides in nuclear waste disposal and providing tools to apply the results to Performance Assessment/Safety Case. Although redox is not a new geochemical problem, different questions are still not resolved and thus raised by implementers and scientists. Radionuclide redox transformations on minerals are considered to be decisive scenarios in various FEP lists. 32 institutions from 13 European countries contribute to the 4-years CP by i) development of advanced analytical tools, ii) investigations of processes responsible for redox control (thermodynamically and kinetically controlled processes, surface reactions and microbial processes), iii) provision of required data on redox controlling processes, and iv) response of the disposal systems to external/internal disturbances. Systems which are investigated comprise simple very well defined ones under controlled laboratory conditions, complex laboratory systems, near-natural systems in the laboratory, real systems, and near-field systems controlled spent nuclear fuel. The scientific results of the CP ReCosy are reported during annual workshops (AW). The first AW took place in February 10–12, 2009. The scientific contributions underwent a peerreview and are published in the form of workshop proceedings together with scientific-technical overview papers of the different project workpackages. The present paper summarizes the main results obtained after the first year of ReCosy. The status during the project preparation phase of the CP ReCosy is published elsewhere [1].
Remediation of Uranium- and CHC-Contaminated Groundwater on a Former Nuclear Fuel-Element Production Site
Authors: Wo¨rner Jo¨rg , Margraf Sonja
Published: December 2010

Since July 2002 a pump & treat remedial action has been in operation for the extraction of Uranium- and “CHC-compounds” contaminated ground-water flowing below an industrial area used before as production site for nuclear fuel elements. So far about 17 kg Uranium and 22 kg CHC-compounds have been eliminated. As described earlier [1] two ground-water plumes have transported the Uranium downstream to locations where two main remediation wells are placed. They show characteristic differences with respect to their Uranium-nuclide vector. More than 220.000 m3 ground-water of the whole 350.000 m3 have been pumped therefrom. Preferred ground-water paths cause blending the two plumes’ waters. The first plume’s water characterized by decreasing Uranium-235 ratios is meanwhile dominated by the second plume’s water of up to 85% in case of well W10. The overall activity of 122 MBq Uranium-235 has been extracted from the saturated zone with decreasing quantities. Only a small area remains where the first plume’s pure ground water will be extracted in future. Both target values of 20 μg/l and 10 μg/l for dissolved Uranium and CHC-compounds, respectively, could not yet be reached for all remediation wells. In accordance to the 10μSv-concept of radiation protection the improvement of specific parameters will still be necessary in order to release the complete site from the German Atomic Law.
Comparison of Actinides Separation by Coprecipitation and Chromatographic Resin (Dipex®) for Gross Alpha Determination
Authors: Lara-Robustillo Esperanza , Rodri´guez Alcala´ Marina
Published: December 2010

The determination of gross alpha in wastes that have alpha, beta and gamma emitters is important from a regulatory point of view. The level of alpha radioactivity allowed is limited in a low and medium level radioactive location for safety reasons, so it is necessary to measure it with high precision. In order to quantify the concentration of alpha emitters it is necessary to perform an adequate radiochemical separation before measurement. In this paper has been studied and compared the results obtained by two separation methods: one by coprecipitation using BaSO4, Fe(OH)3 and/or CaC2O4 as precipitant agents and the other one by extraction chromatography using Actinide Resin (Dipex®). The separation procedures have been applied to spent ion exchange resins from Spanish Nuclear Power Plants. Once the sample was dissolved by acid digestion in a closed vessel microwave equipment, and the radionuclides of interest were isolated of the rest by the techniques before mentioned, the gross alpha was measured by Liquid Scintillation Counting (LSC), using the α/β discrimination if necessary (when the separation was not completed), since this technique is today one of the most used and accurate for measuring radioactivity because LS spectrometry can detect practically all types of radiation with high efficiency.
Expectations for Managing Contaminated Ground and Groundwater: Developing a Common View of NDA and Regulators
Authors: Clark Anna
Published: December 2010

The management of contaminated ground and groundwater is a notable contributor to dealing with the challenge we face in cleaning up the legacy of the UK’s civil nuclear industry in a safe, cost-effective and environmentally responsible manner. To facilitate this mission, the Nuclear Decommissioning Authority, Environmental Regulators and Safety Regulators are working together to develop common expectations for the management of contaminated ground and groundwater arising on and extending off nuclear licensed sites in the UK. The aims of this work are to: • set out shared expectations for land quality management, explaining any differing expectations where consensus is difficult; • interpret expectations to ensure they are clear and implementable, facilitating planning of programmes and deliverables; • provide a framework for dialogue against which progress in land quality management can be mapped; • promote positive action to manage land quality in a proportionate and sustainable manner to achieve consistent standards; and • identify whether areas of the regulatory framework or NDA contractual requirements warrant review and propose improvements for consideration, as appropriate. This paper outlines the process currently ongoing to identify the best way of achieving these aims in a manner that avoids compromising the respective statutory obligations, duties and functions of each party.
The Use of Hot-Isostatic Pressing to Process Nuclear Waste Forms
Authors: Stewart Martin W. A. , Moricca Sam A. , Eddowes Tina , Zhang Yingjie , Vance Eric R. , Lumpkin Gregory R. , Carter Melody L. , Dowson Mark , James Michael
Published: December 2010

ANSTO has developed a combination of tailored nuclear waste form chemistries coupled with the use of flexible hot-isostatic pressing processing technology to enable the successful incorporation of problematic nuclear wastes into dense, durable monoliths. This combined package also enables the design of waste forms with waste loadings well in excess of those achievable via baseline melting routes using borosilicate glass, as hot-isostatic pressing is not constrained by factors such as glass viscosity, crystallisation and electrical conductivity. In this paper we will discuss some of our experiences with problematic wastes, namely plutonium wastes, sludges and HLW such as the Idaho calcines.
Field Examples of Waste Assay Solutions for Curium-Contaminated Wastes
Authors: Chard Patrick M. J. , Croft Stephen , Hutchinson Ian G. , Turner Tom W. , Ross Ann , Greenhalgh Barrie
Published: December 2010

Passive neutron coincidence counting is a mature technique for the assay of Pu in nuclear material. It is widely deployed in safeguards and waste inventory verification applications. The presence of 242Cm and 244Cm in spent fuel wastes, often poses a severe challenge owing to the relatively short spontaneous fission half-life for these isotopes and the subsequent prolific spontaneous fission neutron emission. This is a well documented problem, compounded by the fact that for most waste assay applications, neutron assay techniques are not capable of distinguishing between these Cm isotopes and the even isotopes of Pu, which are normally of interest in waste assay applications. Therefore the presence of even small quantities of these isotopes can result in gross over-estimation of the Pu inventory, if an appropriate correction is not made. Previous theoretical studies carried out recently have illustrated the potential magnitude of the problem, with reference to the fundamental nuclear data and typical isotopic compositions of wastes. Neutron multiplicity counting can, in principle, differentiate between isotopes that undergo spontaneous fission, however in practice the uncertainties in waste assay are such that this is rarely beneficial. More practical “compensation” techniques use combinations of different assay techniques (for example passive and active neutron counting) and knowledge of the actinide ratios in the waste stream fingerprint. In this paper we describe various waste assay applications as case studies. For each example we describe the nature of the challenge and show how solutions have been developed for applications where the presence of curium has caused problems. We describe the technical solutions, showing the limitations and assumptions of each. We also emphasise the role of robust Quality Assurance procedures, to ensure that the techniques are implemented reliably and with predictable outcomes. Finally, we describe the benefits that have been realised for the plant operations teams, with regard to improved measurement accuracy, avoidance of false over-estimation of the Pu inventory and subsequent improvement in plant throughput.
Hydrogeologic Modelling in Support of a Proposed Deep Geologic Repository in Canada for Low and Intermediate Level Radioactive Waste
Authors: Sykes Jonathan F. , Normani Stefano D. , Yin Yong , Sykes Eric A. , Jensen Mark R.
Published: December 2010

A Deep Geologic Repository (DGR) for Low and Intermediate Level radioactive waste has been proposed by Ontario Power Generation for the Bruce Nuclear Power Development site in Ontario, Canada. The DGR is to be constructed at a depth of about 680 m below ground surface within the argillaceous Ordovician limestone of the Cobourg Formation. This paper describes a regional-scale geologic conceptual model for the DGR site and analyzes flow system evolution using the FRAC3DVS-OPG flow and transport model. This provides a framework for the assembly and integration of site-specific geoscientific data that explains and illustrates the factors that influence the predicted long-term performance of the geosphere barrier. In the geologic framework of the Province of Ontario, the Bruce DGR is located at the eastern edge of the Michigan Basin. Borehole logs covering Southern Ontario combined with site specific data have been used to define the structural contours at the regional and site scale of the 31 sedimentary strata that may be present above the Precambrian crystalline basement rock. The regional-scale domain encompasses an 18.500km2 region extending from Lake Huron to Georgian Bay. The groundwater zone below the Devonian is characterized by units containing stagnant water having high concentrations of total dissolved solids that can exceed 300g/l. The computational sequence involves the calculation of steady-state density independent flow that is used as the initial condition for the determination of pseudo-equilibrium for a density dependent flow system that has an initial TDS distribution developed from observed data. Long-term simulations that consider future glaciation scenarios include the impact of ice thickness and permafrost. The selection of the performance measure used to evaluate a groundwater system is important. The traditional metric of average water particle travel time is inappropriate for geologic units such as the Ordovician where solute transport is diffusion dominant. The use of life expectancy and groundwater age is a more appropriate metric for such a system. The mean life expectancy for the DGR and base case parameters has been estimated to be in excess of 8 million years.
Further Development of Iodine Immobilization Technique by Low Temperature Vitrification With BiPbO2I
Authors: Mukunoki Atsushi , Chiba Tamotsu , Suzuki Yasuhiro , Yamaguchi Kenji , Sakuragi Tomofumi , Nanba Tokuro
Published: December 2010

The authors describe progress in the development of low temperature vitrification with BiPbO2I (BPI) as a promising immobilization technique by which Iodine-129 is recovered by BiPbO2NO3 to form BPI, and then solidified into a lead-boron-zinc glass matrix (PbO-B2O3-ZnO) using a low temperature vitrification process. The microscopic structure of BPI glass was analyzed by various analytical techniques, such as XRD (X-ray diffraction), NMR (nuclear magnetic resonance analysis), and XPS (X-ray photoelectron spectroscopy), using several types of glass samples. The results obtained provide structural information on key elements in BPI glass and can be applied for modeling the structure of the BPI glass, simulated by molecular dynamics. The previous work suggested that the leaching behavior of iodine from BPI glass depended upon the chemical conditions of the solution. Further leaching tests using solutions under varying conditions were carried out in order to predict mechanisms of iodine leaching. Normalized elemental mass loss values of iodine in simulated seawater and bentonite pore water are almost the same as those of boron, showing that iodine dissolves congruently with BPI glass, whereas iodine dissolves incongruently in Ca(OH)2 solutions of pH 9 and 11. To demonstrate the feasibility of the BPI vitrification process, recovery tests of iodine from spent iodine filters were conducted and a prototype melting furnace was developed for scale-up tests of glass sample. It was found that more than 95% of iodine can be recovered from the spent iodine filter and that the prototype furnace can produce approximately 0.5 liters of homogeneous glass.
Experience in Chemical Decontamination of PWR Systems and Components
Authors: Steinkuhler Claude , Lenie Koen , Coomans Reginald
Published: December 2010

Tecnubel has recently performed various chemical decontamination of French and Belgian Pressurized Water Reactors (PWR) systems and components. The purpose of this paper is to present and compare these experiences. The objectives of these operation were the reduction of the general surface contamination together with the elimination of hot spots in Residual Heat Removal Systems (RHRS), Chemical and Volume Control Systems (CVCS) and Reactor Coolant Pumps (RCP). This reduction of contamination leads to the reduction of dosimetry to the maintenance personnel and allows the works on critical equipment. An additional challenge for three of these projects lay in the execution of a complicated operation on the critical path of a reactor refueling shutdown. The chemical decontamination were performed by circulating an adequate fluid in the systems or around the components. Since the contamination was generated at hot conditions during power operation, a redox attack on the surface was necessary. The EDF systems and components were decontaminated using a qualified EDF process of the EMMAC family. The Reactor Coolant Pump from the Belgian PWR was treated with the NITROX process, qualified by Westinghouse. The functions required by the decontamination system were very diverse and therefore an existing decontamination loop, which was previously developed for the decontamination of small circuits, was re-developed and adapted for bigger volumes by DDR Consult and Tecnubel. The results of five decontamination are presented and detailed in terms of efficiency and waste production. These projects were: the chemical decontamination of the RHRS of Flamanville 1 NPP, of the CVCS non regenerative heat exchanger at St Laurent des Eaux NPP, of the RHRS and CVCS of Bugey 2 NPP and of two RCP at the Westinghouse Belgian Service Center.
Experience With Technical Advisory Groups in the Japanese HLW Disposal Programme
Authors: Tsuchi Hiroyuki , Kitayama Kazumi , Deguchi Akira , Takahashi Yoshiaki , Ohe Toshiaki , McCombie Charles , McKinley Ian
Published: December 2010

Recognising the benefits to be gained by integrating national and international experience into a rapidly growing programme, NUMO already in June 2001 set up both domestic and international technical advisory committees. These groups supported NUMO during the time when many of the basic, often innovative, concepts that characterise the Japanese deep geological disposal programme were developed. Overall experience with those technical advisory committees was very positive: most of the essential programme elements have been put in place and, in some technical areas, NUMO is now up at the front along with other leading programmes. On the other hand, after a decade, NUMO recognized the necessity to revisit their roles and rethink the overall NUMO policy on technical support, advice and review. This paper provides a summary of NUMO’s experience with advisory committees and an outlook on the future use of such bodies.
The Development and Use of T2GGM: A Gas Modelling Code for the Postclosure Safety Assessment of OPG’s Proposed L&ILW Deep Geologic Repository, Canada
Authors: Suckling Paul , Calder Nicola , Humphreys Paul , King Fraser , Leung Helen
Published: December 2010

As part of the postclosure safety assessment of Ontario Power Generation’s (OPG’s) proposed Deep Geologic Repository (DGR) for Low and Intermediate Level Waste (L&ILW) at the Bruce site, Ontario, a Gas Generation Model (GGM) has been developed and used to model the detailed generation of gas within the DGR due to corrosion and microbial degradation of the organics and metals present. The GGM is based on a kinetic description of the various microbial and corrosion processes that lead to the generation and consumption of various gases. It takes into account the mass-balance equations for each of the species included in the model, including three forms of organic waste (cellulose, ionexchange resins, and plastics and rubbers), four metallic waste forms and container materials (carbon and galvanised steel, passivated carbon steel, stainless steel and nickel-based alloys, and zirconium alloys), six gases (CO2, N2, O2, H2, H2S, and CH4), five terminal electron acceptors (O2, NO3−, Fe(III), SO42−, and CO2), five forms of biomass (aerobes, denitrifiers, iron reducers, sulphate reducers, and methanogens), four types of corrosion product (FeOOH, FeCO3, Fe3O4, and FeS), and water. The code includes the possibility of the limitation of both microbial and corrosion reactions by the availability of water. The GGM has been coupled with TOUGH2 to produce T2GGM; a code that models the generation of gas in the repository and its subsequent transport through the geosphere. T2GGM estimates the peak repository pressure, long time repository saturation and the total flux of gases from the geosphere. The present paper describes the development of T2GGM and the numerical modelling work undertaken to calculate the generation and build-up of gas in the repository, the two-phase exchange of gas and groundwater between the repository and the surrounding rock, and between the rock and the surface environment. The results have been used to inform the safety assessment modelling.
Thermal Cutting Technologies for Decommissioning of Nuclear Facilities
Authors: Bienia Harald
Published: December 2010

Remote disassembly of radiologically burdened large components is among the most sophisticated and complex activities in the dismantling of nuclear installations. High local dose rates and contamination levels, combined with complicated designs and geometries of the object to be dismantled, plus insufficient accessibility, imply major challenges in the dismantling of nuclear facilities. Usually the shielding effect of water is used during the dismantling period. Other dismantling activities require dry ambiences.
Radiological Characterization of a Copper/Cobalt Mining and Milling Site
Authors: Arno Matthew G. , Arno Janine Katanic , Halter Donald A. , Berry Robert O. , Hamilton Ian S.
Published: December 2010

Extensive copper and cobalt ore deposits can be found in the Katanga Province of the Democratic Republic of the Congo near the city of Kolwezi. These deposits have been mined via open pit and underground mines since the 19th century with many changes in control of the mines including colonial industrial control and Congolese government control. With the recent re-establishment of a relatively stable democratic government in the DRC, foreign investors returned to the area to restart mining activities that were abruptly terminated in the 1990’s due to political turmoil. Some of these new projects are being performed in accordance with World Bank and International Finance Corporation Social & Environmental Sustainability standards. As part of these standards, radiological characterization of the mines, processing facilities, and surrounding environment was conducted to establish current conditions, evaluate human health and ecological risks, and provide a basis for establishment of radiation safety and environmental remediation programs. In addition to naturally occurring radioactive materials associated with the copper/cobalt ore, the site was reputedly historically used to store ore from the Shinkolobwe uranium mine, the source of the uranium ore for the World War II Manhattan project. The radiological characterization was conducted via extensive gamma radiation surveys using vehicle-mounted sodium-iodide detectors, random grid composite soil sampling, biased soil sampling of areas with elevated gamma radiation levels, and sampling of surface water features. The characterization revealed broad areas of elevated gamma radiation levels of up to 160 μGy/hr in two distinct areas believed to be the Shinkolobwe uranium mine ore storage locations. Other areas, with gamma radiation levels of up to 80 μGy/hr, were detected associated with copper/cobalt ore refinery tailings and waste rock (overburden) sediments. The gamma radiation surveys revealed that elevated radiation levels were largely confined to areas previously disturbed by mechanized mining activities. Radiological contaminants in local surface water sources were within drinking water standards with the exception of one river heavily polluted with both uranium and other metals by waste streams from an ore processing and refining facility. Surrounding areas that appeared to be undisturbed by mining, including agricultural areas, native villages, and urban colonial-architecture cities, exhibited soil concentration and gamma radiation levels consistent with expected background levels.
Implications of Microbial Redox Catalysis in Analogue Systems for Repository Safety Cases
Authors: West Julia M. , McKinley Ian G. , Stroes-Gascoyne Simcha
Published: December 2010

A detailed assessment of studies of oxidising redox fronts around fractures at depth in otherwise “reducing” environments suggests that the usual explanation, in terms of past disturbances that have resulted in deep penetration of oxidising water, are incompatible with hydrogeological and/or geochemical observations. An alternative hypothesis, microbial catalysis of kinetically slow or hindered reactions involving oxyanions such as sulphate or carbonate, appears potentially more credible. Although still not always taken into account by the geochemical community, the role of microbial metabolism in low temperature geochemistry is supported by the rapidly expanding database on subsurface microbial populations. These populations are demonstrated to be viable and, therefore, could potentially be active at levels close to or below current detection limits in deep geological systems. Indeed, inspection of information available from several analogue studies or repository site characterisation programmes suggests that such activity may explain some of the geochemical anomalies encountered. This paper examines the current (indirect) evidence for microbial redox catalysis in relevant subsurface rock matrix environments and considers the implications that this would have for the development of site understanding — and in particular the identification of factors that may distinguish between different locations during site selection. Further, it examines the wider implications of more extensive roles of microbes in repository systems on the overall post-closure safety case and the need for further focused analogue studies to develop answers to these open questions.
The Ruprechtov Natural Analogue Site (CZ) Study: Mobile Natural Organic Matter Identification, Characterisation and Link to PA Relevant Processes
Authors: Havlova´ Va´clava , Cˇervinka Radek , Noseck Ulrich , Brasser Thomas , Havel Josef
Published: December 2010

The Ruprechtov Natural Analogue (CZ) Programme has been focused on studying real system processes, relevant to performance assessment (PA) of sediment formations that can form the overburden of geological repository host rocks. The site has been extensively studied due to its geological constitution (granite – kaolin – clay – U mineralisation – organic matter). The presented study used Ruprechtov unique but well-described geological conditions in order to identify and characterise mobile organic matter (MOM) that can be easily released into groundwater and can influence PA relevant specie migration due to complexation/sorption reaction. The modern analytical method MALDI-TOF MS was used for characterisation. It was found that only a small fraction of sedimentary natural organic matter (NOM) from the site was easily releasable (max. 5%) as MOM, resulting in low organic substance concentration in natural groundwater. MOM amount released was decreasing with increasing NOM content. MALDI-TOF MS proved to be a useful tool to characterize organic substances, either natural ones or artificially released from natural organic matter samples. A noticeable fingerprint for all the MOM compounds analysed was found at MALDI-TOF MS spectra. This showed that MOM from the Ruprechtov site was in all cases composed of molecules with low molecular weight (under 1000 Da). As determined by the consequent geochemical analyses, despite groundwater reducing conditions MOM compounds would be mainly interacting with U(VT) in the groundwater, being present as more abundant U specie. Good correspondence of results enabled to consider the extracted humic acid HA 12/3 as a mobile organic matter fraction representative.
WASAN: A Methodology for Analysing Waste Minimisation
Authors: Shaw Duncan , Blundell Neil
Published: December 2010

This paper presents a new methodology called Waste And Source-matter ANalyses (WASAN) which helps a group to analyse waste-production and build actions to minimise avoidable waste. Designed for Licencees of nuclear facilities as Health and Safety Executive (HSE) Guidance on what constitutes good practice, WASAN uses lessons from Systems Thinking, Waste Management Hierarchy, Hazard Operability (HAZOP) Studies, As Low As Reasonably Practicable (ALARP), simulation modelling and sensitivity analysis. These lessons are used to support groups when developing plans to reduce avoidable waste production. We report on the development of WASAN and its use in three workshops. This paper reports progress on the development of waste management guidance for the HSE and does not currently represent its final regulatory view.
An Overview on the National Radwastes Management Strategy in the Context of the Nuclear Program Development in Romania
Authors: Adrian Panait , Radu Maria , Barariu Gheorghe , Negut Gheorghe , Litescu Cristian
Published: December 2010

At present the main radwaste generator is Cernavoda NPP with 2 × 700 MW units in operation. For the year 2016, the Plant is predicted to have 4 units in operation. The National Strategy for radioactive waste management was drafted by National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for coordinating spent nuclear fuel and radioactive waste management in Romania. Last year, the Government of Romania established that a new nuclear power plant would be build, so, with this future operational NPP, the quantity of waste will increase from the today evaluations. Therefore the national strategy for radioactive waste management should be reviewed accordingly. The paper presents some preliminary results obtained upon the R&D Programs related to the radioactive waste disposal facilities.
Some Aspects of Radioecological Monitoring of High Aquatic Plants From Water-Bodies Within the Chernobyl Accident Exclusion Zone
Authors: Shevtsova Natalie L. , Gudkov Dmitriy I. , Yavnyuk Andrian A.
Published: December 2010

The impact analysis of low doses of ionizing radiation on the breathers in natural populations is an important part of radiobiological studies of non-human biota. The main aim of our studies was to investigate some cytogenetic, morphological and reproductive rates of the common reed (Phragmites australis (Cav.) Trin. ex. Steud.) from different water bodies within the Chernobyl accident exclusion zone. The absorbed dose rate for littoral emergent plants in sampling water bodies was varied from 1.3E−02 to 1.6E−01 Gy/h. The rate and main types of chromosome aberrations in roots meristems, morphological damages in seed germs, as well as rates of germinating ability and power were analyzed. There were registered rather low rate of germinating ability (14–48%) and germinating power (<1) of seeds from all sampling water bodies with high levels of radioactive contamination in comparison to control ones. Against the general suppressed background the effect of relative stimulation of more affected seeds was observed. With increase of absorbed dose in range of 1.3E−02–1.6E−01 Gy/h the number of germinated seeds was increased. At the same time the number of morphological damages of seeds was increased as well. There was determined the positive correlation between absorbed dose rate and chromosome aberration rate in roots of the common reed from sampling water bodies. The highest rate of chromosome aberrations (up to 17%) were registered in plants with high level of morphological deviations in seeds germs. The data obtained from the complex analysis of natural populations of the common reed from the radioactive contaminated water bodies testify about rather high level of genetic efficiency of low doses of long-term exposure.
Sustainable Covers for Uranium Mill Tailings, USA: Alternative Design, Performance, and Renovation
Authors: Waugh William J. , Benson Craig H. , Albright William H.
Published: December 2010

The U.S. Department of Energy Office of Legacy Management is investigating alternatives to conventional cover designs for uranium mill tailings. A cover constructed in 2000 near Monticello, Utah, USA, was a redundant design with a conventional low-conductivity composite cover overlain with an alternative cover designed to mimic the natural soil water balance as measured in nearby undisturbed native soils and vegetation. To limit percolation, the alternative cover design relies on a 160-cm layer of sandy clay loam soil overlying a 40-cm sand capillary barrier for water storage, and a planting of native sagebrush steppe vegetation to seasonally release soil water through evapotranspiration (ET). Water balance monitoring within a 3.0-ha drainage lysimeter, embedded in the cover during construction, provided convincing evidence that the cover has performed well over a 9-year period (2000–2009). The total cumulative percolation, 4.8 mm (approximately 0.5 mm yr−1), satisfied a regulatory goal of <3.0 mm yr−1. Most percolation can be attributed to the very wet winter and spring of 2004–2005, when soil water content exceeded the storage capacity of the cover. Diversity, percent cover, and leaf area of vegetation increased over the monitoring period. Field and laboratory evaluations several years after construction show that soil structural development, changes in soil hydraulic properties, and development of vegetation patterns have not adversely impacted cover performance. A new test facility was constructed in 2008 near Grand Junction, Colorado, USA, to evaluate low-cost methods for renovating or transforming conventional covers into more sustainable ET covers.
Strategic Environmental Assessment for UK LLW Management
Authors: Craze Andrew , Davis Pete , Clark Matthew
Published: December 2010

NDA is delivering a Strategic Environmental Assessment (SEA) to underpin the UK Nuclear Industry Low Level Waste Strategy. The purpose of this assessment is embed sustainability issues into our decision making and to fulfil our requirements under the European Union’s Strategic Environmental Assessment (SEA) Directive (2004/42/EU) and transposing UK Regulations, and to underpin the development of the strategy. The outputs of the SEA have provided input into particular aspects of the strategy, leading to a more robust and better informed result. Development of options to be assessed under the SEA has looked at a number of factors, including: • what the strategy is aiming to achieve; • expectation from stakeholders as to what should be addressed; • consideration of tactical approaches to implementation of the strategy in addition to high level strategic issues; • links to other projects and programmes (for example the Environmental Safety Case for the Low Level Waste Repository. The SEA aims to provide a robust assessment of the environmental and sustainability impacts of alternative strategies for providing continued capability and capacity for the management and disposal of LLW in the UK. The assessment also considers other, more tactical, issues around implementation of the strategy, for example: issues around the location of LLW management facilities; the environmental impacts of alternative waste treatment options (metal recycling etc); considerations of alternative approaches to the classification of radioactive waste and opportunities that would result. Critical to the development of the SEA has been the involvement of statutory and non-statutory stakeholders, who have informed both the output and the approach taken.