bluuebunny/crossref_metadata_2025 huggingface dataset:
Demonstration of Gas-Permeable Seals for Radioactive Waste Repositories: Laboratory and In-Situ Experiments
Authors: Rueedi Joerg , Marschall Paul
Published: August 2012

In low/intermediate-level waste (L/ILW) repositories, anaerobic corrosion of metals and degradation of organic materials produce hydrogen, methane, and carbon dioxide. Gas migration in a L/ILW repository is one of the processes evaluated in the safety assessment of deep geological disposal in low-permeability formations, in particular with respect to the development of gas pressures in the repository caverns which could negatively affect the host rock or the engineered barrier system (EBS). In order to restrict build-up of gas overpressures in the emplacement caverns, Nagra (National Cooperative for the Disposal of Radioactive Waste, Switzerland) has proposed design options aimed at increasing the gas transport capacity of the backfilled underground structures, compromising neither the low hydraulic conductivity nor the radionuclide retention capacity of the EBS (Nagra, 2008). They involve specially designed backfill and sealing materials such as high porosity mortars as backfill materials for the emplacement caverns and sand/bentonite (S/B) mixtures with a bentonite content of 20% to 30% for the seals themselves and for backfilling other underground structures. These increased gas permeability materials can supplement the gas flow that is expected to occur through the excavation damaged zone (EDZ) and avoid the creation of overpressures. Preliminary experimental studies have confirmed the gas transport capacity of the S/B mixtures and demonstrated the ability to design mixtures with specific target permeabilities for water and gas flow (Nagra, 2008). Two-phase flow modelling studies have shown that the gas transport capacity of seals is largely dependent on their permeability and length. More detailed models of sealing elements show a rather complex history of seal saturation during the early saturation phase and the later gas escape phase (Gaus et al., 2010). Note, however, that current modelling approaches are based on parameters and conceptual understanding of small-scale laboratory experiments. Two large(r) scale experiments which aim at validating and, if necessary, improving current conceptual models for the resaturation and gas invasion processes into S/B seals and the determination of up-scaled gas / water permeabilities of S/B seals (i.e. two-phase flow parameters for large-scale models) have been initiated and will be highlighted in the paper. The first one, a mock-up experiment, was set up in 2010 as part of the EU 7th FP project FORGE, aiming at demonstrating seal performance on an intermediate (decimetre scale). The second one is a large-scale experiment (metre-scale), the Gas-Permeable Seal Test (GAST), which was also initiated in 2010 at the Grimsel Test Site (GTS). For GAST, a seal will be emplaced at the GTS to demonstrate the effective functioning of gas-permeable seals on a realistic scale and with realistic boundary conditions (‘proof of concept’).
Decommissioning of the A-1 NPP Heavy Water Evaporator Facility
Authors: Medved Jan , Vargovcik Ladislav
Published: August 2012

The paper deals with experience and techniques in the application of remotely controlled robotic devices for the dismantling of the A-1 NPP technological equipment during undergoing decommissioning process of the A-1 NPP, which is characterized by high level of radioactivity and contamination. For liquidation of the heavy water evaporator has been applied a mobile robotic system MT 80, which had been developed, designed and constructed as a general-purpose decommissioning equipment. The heavy water evaporator as a part of the NPP heavy water system is located inside the main production unit building in Room No. 220 where the inner surface contamination is from 101 Bq/cm2 to the level of 103 Bq/cm2, dose rate up to 1.5 mGy/h and the feeding pipeline contained LRAW with high tritium content. The first step was the development of a work procedure with special focus on the elimination of activity and aerosols leaking into the environment. Special tooling was developed for application with the robot, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Then, civil engineering modifications were made to the workplace and new technology was installed, including an efficient exhaust system. After draining-off the remains of LRAW and rinsing the pipeline, fragmentation of the pieces of equipment was started. The fragments are being deposited into drums which are transported in shielded containers to the decontamination facility, where their activity is reduced prior to their storage or further use. All operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. This experience will be exploited in the continuation of work in Room No. 219, where the second identical heavy water evaporator is located. Nuclear Power Plant A-1 is situated in the locality of Jaslovske´ Bohunice. The A-1 NPP decommissioning has been organized according to distinctively licensed phases, which allows collecting input data for next stage during the running decommissioning stage. Since 2009 the Stage II of the A-1 NPP decommissioning has been in process.
Centre de la Manche Institutional Control Period: After the First Safety Review
Authors: Dutzer Michel , Vervialle Jean Pierre , Andre´ Alain , Marchiol Albert
Published: August 2012

Centre de la Manche disposal facility is the first French surface disposal facility dedicated to low and intermediate level short lived radioactive waste. It started up in 1969. After a continuous improvement, in the design of disposal vaults, in operational modes, in the whole process of waste management, in the safety approach, the last packages were received in 1994. 527,000 m3 of waste packages have been disposed during the 25 years of operation. The facility was licensed for the institutional control period in 2003. The disposal vaults are covered with a multilayer capping system that includes a bituminous membrane to provide protection against rainwater infiltration. Water that might infiltrate through the membrane is collected by the bottom slab of the vaults to a pipe network implemented in an underground gallery. Measurements show an overall infiltration rate of about 3 l/m2/year that complies with the objective of Andra of a few liters per square meter and per year. Investigations are performed in order to assess the behavior of the membrane in the long term. For this purpose periodically samples of the bituminous membrane are taken and measurements are performed. As at the beginning of the operational period waste packages were not conditioned in accordance with the specifications that are presently prescribed to waste generators, some settlements can be observed on the ancient part or the facilities. At the end of 2009 some excavation works were performed in an area where a settlement of few tens of centimeters was observed. The integrity of the membrane could be observed and the adequacy of the selection of this option for the watertightness of the capping system was so confirmed. Environmental monitoring includes radiological and chemical measurements for discharge, underground water and surface water. In the particular framework of Centre de la Manche, a contamination of groundwater by tritiated wastes occurred in 1976. Theses wastes were removed in 1977 and 1978 but tritium can still be detected under the facility. However the impact to a hypothetical critical group, using only local water, is presently about 0,4 μSv per year (2009). A safety review of the facility was performed in 2009 by the regulatory body. It was noticed that the behavior of the facility was consistent with predictions. The progressive strategy proposed by Andra to improve the capping system, in particular to achieve long term stability of the slopes, was assessed. The process implemented by Andra to maintain memory of the facility was discussed. In this way the periodical safety reviews are part of the dynamic process through which the monitoring of the facility will become progressively more and more passive.
Industrial Complementarities Between Interim Storage and Reversible Geological Repository
Authors: Hoorelbeke Jean-Michel
Published: August 2012

The French Act voted in 2006 made the choice of deep geological disposal as the reference option for the long term management of high level (HLW) and intermediate level long-lived waste. The Cige´o repository project aims at avoiding or limiting burden to future generations, which could not be achieved by the extension in time of interim storage. The reversibility as provided by the Act will maintain a liberty of choice for waste management on a duration which is comparable to new storage facility. Interim storage is required to accommodate waste as long as the repository is not available. The commissioning of the repository in 2025 will not suppress needs for interim storage. The paper describes the complementarities between existing and future interim storage facilities and the repository project: repository operational issues and planning, HLW thermal decay, support for the reversibility, etc. It shows opportunities to prepare a global optimization of waste management including the utilization at best of storage capacities and the planning of waste emplacement in the repository in such a way to facilitate operational conditions and to limit cost. Preliminary simulations of storage-disposal scenarios are presented. Thanks to an optimal use of the waste management system, provision can be made for a progressive increase of waste emplacement flow during the first operation phase of the repository. It is then possible to stabilize the industrial activity level of the repository site. An optimal utilization of interim storage can also limit the diversity of waste packages emplaced simultaneously, which facilitates the operation of the repository. 60 years minimum interim storage duration is generally required with respect to HLW thermal output. Extending this interim storage period may reduce the underground footprint of the repository. Regarding reversibility, the capability to manage waste packages potentially retrieved from the repository should be analyzed. The 2006 French Act provides for a research program on interim storage to be carried out along with the repository project development. This program has been guided by the complementarities between interim storage and the repository project. The main research issues address the longevity of storage facilities, up to 100 years, their versatility with regard to waste package types and their modularity to match future needs progressively. In parallel the dialogue between Andra and waste producers will continue to propose optimized waste management scenarios.
Uncertainty Analysis of Activity Measurement of New Monitoring System for Free-Release for NPP A-1 Decommissioning, Slovakia
Authors: Slaninka Alojz , Sla´vik Ondrej , Necˇas Vladimi´r
Published: August 2012

New free release monitoring post with a large volume 600 L container counting geometry was designed and developed. The monitoring system is able to monitor a material also in standard counting geometry of 200 L drum. Using counting geometry of 600 L rectangular container that is equipped with self-discharger is able to increase the total monitoring capacity. The monitoring system is based on a pair of electrically cooled semiconductor HPGe detectors that are placed into a modifiable vertical or horizontal pair of lead collimators. The monitoring system is integrated with an industrial scale for determination of massic activities of measured materials and in addition by a rotating table in the case of 200 L drums monitoring. Monitoring system is integrated into transportable ISO container with constant environmental conditions that are ensured by air-condition unit. Full-energy peak detection efficiency (FPE) polynomial curves for various densities of measured material were in both cases determined by ISOCS calibration code based on designed counting geometry and delivered ISOCS/LabSOCS detector characterization. Uncertainty analysis of massic activity measurement by container and drum monitoring system in designed counting geometry is introduced below in more detail.
Dealing With a Pa-231 Contaminated Ventilation Duct
Authors: Irving Clare , Burgess Pete , Stevens Keith
Published: August 2012

Cambridge University had worked in the 1960s and 1970s with Pa-231, a decay product of U-235. The fume cupboards discharged into ventilation ducting made from asbestos cement. The university wished to refurbish the laboratory and the RPA had negotiated over many years with the Environment Agency to set up a project to remove the ducting both to reduce the radiological hazards and as part of a programme to remove unwanted circuits and upgrade the ventilation system to modern standards. Contamination levels were significant and low dose rates were measurable on the external surface. The aim was to be able to remove the ducting and treat it as asbestos waste, rather than to have to treat the debris as asbestos contaminated radioactive waste. The age of the contaminant was such that a large fraction of the decay chain had grown in, giving a mixture of alpha, beta and gamma emissions. The most useful nuclides for surface monitoring were Pb-211 and Tl-207, both of which are energetic beta emitters. A wide energy range beta detector was used, but it was fitted with a filter to absorb any alpha radiation which otherwise would have contributed to the signal for good surfaces but not for dusty, damp or rough surfaces and would have contributed to the uncertainty in the activity assessment. Samples were checked using gamma spectrometry to confirm that only Pa-231 and its progeny were present in significant quantities. The gamma spectrum is complicated and this paper describes the difficulties in confirming that the spectrum only contained the Pa-231 decay chain. The vast majority of the contaminated ducting was successfully consigned as asbestos, rather than radioactive, waste. The other problem was dealing with the soft waste produced during the dismantling process. This was monitored using simple equipment and it was possible to demonstrate that it could be disposed of with the rest of the waste under the relevant UK legislation.
The Use of Standard Alpha and Beta Surface Scintillation Contamination Monitors to Confirm the Contamination Fingerprint and to Check on Source Quality
Authors: Burgess Pete
Published: August 2012

Normally, beta and alpha surface contamination monitors are used with a simple counting threshold, i.e. any pulse over a predetermined amplitude is counted. This is very different from gamma monitoring, where the use of counting windows is very popular and the use of full multi-channel analysis is common. Many current surface contamination ratemeters have the capacity to drive dual phosphor detectors and can be set up to provide beta and alpha channels. Effectively, the beta channel is a counting window, i.e. all pulses which are bigger than the threshold and smaller than the alpha threshold are counted. Larger pulses go into the alpha channel. This paper addresses how this can be used with beta only and alpha only detectors to provide information on the source. The detector is set up conventionally to a defined point for the lowest beta energy anticipated. The instrument is then switched to alpha + beta mode and the alpha threshold set to 3 times the beta threshold. With this set up, the alpha to beta channel count rate ratio varies smoothly by a factor of 14 between Y-90 (Emax 2.27 MeV) and C-14 (Emax 0.16 MeV). Hence the instrument can be used to estimate the energy of an unknown beta contaminant or to confirm that a mixed beta fingerprint has essentially the same mix. The same approach can be used with alpha probes to confirm the source quality. The main worry with alpha monitoring is the surface condition. A poor surface condition will lead to a low count rate. Using the channel ratio method will identify grubby sources. The resulting ratio can be used either as a go/no trigger, i.e. any surface with a low ratio will be treated as untrustworthy, or alternatively the ratio can be used to correct the reading to give a better estimate of surface activity.
Detection of Shielded Sealed Radioactive Sources in Radioactive Waste by Non-Destructive Assay Techniques
Authors: van Velzen Leo , van der Marck Steven
Published: August 2012

Experiences dealing are seldom reported on the detection of sealed radioactive sources (SRS) or its shielding in waste by non-destructive assay (NDA) methods in literature. Further the knowledge and experience in this field of waste characterization will vary from specialist till basic. The main aim of this paper is to give some guidance with the aid of an overview or scheme with which the possibility can be assessed of the detection and of a shielded SRS in (historic) waste packages. This aim could be reached by simulating the gamma flux emitted by a SRS at the outline of a standard 220 litres drum. The simulations have been performed with the Monte Carlo Neutron Photon transport code (MCNP). The results, visualized by means of iso-plots, are then be used for assessing the detection probability of the SRS by available NDA techniques. The following conclusions could be made: • “Heavy and dens” objects (e.g. shielded sealed radioactive source) can be detected and an indication of its location can be obtained by a sudden significant decrease of the gamma photon flux (cold-spot) at the outline of the drum. This cold-spot is caused by the “heavy or dens” object itself. A cold-spot can be detected the best by analyzing and visualizing the gamma photon flux from the backscatter area (e.g. range 100 keV up to 200 keV) and second best by analyzing the total gamma flux. • With relative simple radiological equipment (e.g. dose rate meter) the possible presence and an indication of the location of a shielded SRS can be estimated. • The detection possibility of detecting a shielded SRS depends not only on the applied NDA technique, but also on the applied effort (e.g. number of measurements, measuring time, applied measurement grid), the properties of the waste packages (e.g. density of the waste matrix and uniformity of the inside dispersed radioactivity), the location and the properties of the shielding of the SRS (e.g. geometry). • For confirmation and the determining of the exact location advanced NDA techniques (e.g. transmission computer tomography or other NDA that techniques gives three dimensional results) have to be applied.
Delivering Value for Money: Trust and Verify?
Authors: Laird Alastair
Published: August 2012

Accurate estimates for national Environmental Management remediation work programs are an essential ingredient of ensuring that plans can be adequately funded. They also form the basis of value measurement as the work is executed on an annual or program basis. However, the inherent uncertainties of many of the Environmental Management (EM) and decommissioning tasks, both in terms of the technical challenges faced, options available, end states to be achieved; and the general risks and uncertainties associated with the hazard and its characterisation means that many estimates were always going to have very high levels of uncertainty. In 2002 the United Kingdom Nuclear Liabilities Estimate was quoted as £48Bn when the government restructured the UK civil nuclear industry and set out the basis for forming what was to become the Nuclear Decommissioning Authority (NDA). By 2005 the NDA had assessed the costs as £56Bn but by 2008 the costs had significantly increased to £73Bn and continue to rise. How does this relate to the more immediate challenges of ‘working off’ the plan and demonstrating Value for Money can be achieved in the near term? In parallel the US Department of Energy Environmental Management Office introduced its ‘Best in Class’ initiative in 2007 — the intention being to tackle underperformance and drive improvements in the baselines and the contractor delivery programs. This paper compares and contrasts UK and US EM program performance issues and covers several interdependent topic areas including: a) Government funding impacts, b) Contractor program estimates, c) Program Controls requirements, and d) Independent assurance requirements. This paper attempts to answer the question “how can governments demonstrate Value for Money in EM”.
Use of Source Term and Air Dispersion Modeling in Planning Demolition of Highly Alpha-Contaminated Buildings
Authors: Droppo James G. , Napier Bruce A. , Rishel Jeremy P. , Bloom Richard W.
Published: August 2012

The current cleanup of structures related to cold-war production of nuclear materials includes the need to demolish a number of highly alpha-contaminated structures. The process of planning for the demolition of such structures includes unique challenges related to ensuring the protection of both workers and the public. Pre-demolition modeling analyses were conducted to evaluate potential exposures resulting from the proposed demolition of a number of these structures. Estimated emission rates of transuranic materials during demolition are used as input to an air-dispersion model. The climatological frequencies of occurrence of peak air and surface exposures at locations of interest are estimated based on years of hourly meteorological records. The modeling results indicate that downwind deposition is the main operational limitation for demolition of a highly alpha-contaminated building. The pre-demolition modeling directed the need for better contamination characterization and/or different demolition methods—and in the end, provided a basis for proceeding with the planned demolition activities. Post-demolition modeling was also conducted for several contaminated structures, based on the actual demolition schedule and conditions. Comparisons of modeled and monitoring results are shown. Recent monitoring data from the demolition of a UO3 plant shows increments in concentrations that were previously identified in the pre-demolition modeling predictions; these comparisons confirm the validity and value of the pre-demolition source-term and air dispersion computations for planning demolition activities for other buildings with high levels of radioactive contamination.
Adsorption of Cesium Radionuclides by the Composite Sorbents Carbon Fiber/Transition Metals Ferrocyanides
Authors: Sheveleva Irina V. , Zheleznov Veniamin V. , Bratskaya Svetlana Yu. , Kuryavyi Valery G. , Avramenko Valentin A.
Published: August 2012

Among various methods of cesium removal from aqueous solutions, sorption using transition metals ferrocyanides is the most efficient method due to extremely high affinity of cesium ions to ferrocyanides. The efficiency of transition metals ferrocyanides application is known to depend on the crystal size being the highest for nanocrystals. Although nanocrystals are difficult to handle in direct application, they can be used in composite materials. In this case two main problems arise: how to control the crystal size of transition metals ferrocyanides and fix them reliably in the supporting matrix. Here we suggest a new route to preparation of composite materials selective to cesium ions using transition metals ferrocyanides stabilized by siloxane-acrylate latexes. The size of transition metals ferrocyanides is controlled by the size of latex particles and their stability is determined by ionization of polyacrylic acid carboxylic groups on the functionalized latex surface. These functionalized particles can be used as precursors in preparation of composite materials by sedimentation and polymerization of latexes on the solid surface of porous matrix, e.g. carbon fibers. Several routes of preparation of carbon fiber based composite materials using functionalized latexes and sorption properties of the obtained materials are discussed. The effect of preparation conditions (method used, carbon fiber polarization potential, concentration of latexes functionalized with transition metals ferrocyanides) on cesium uptake by composite sorbents from solutions of various salinity is reported.
Pilot Test of Precipitation Setup for Dust Supressor and Transuranic Elements Removal From Wastewaters of Chernobyl Nuclear Power Plant
Authors: Avramenko Valentin , Bratskaya Svetlana , Marinin Dmitry , Terzi Anatoliy , Yarmolyuk Mariya
Published: August 2012

In this paper we report the results of pilot tests of flocculation/precipitation setup for dust-suppressor (DS) and transuranic elements (TUE) removal from wastewaters of Chernobyl Nuclear Power Plant (ChNPP), including those of the Object “Shelter”. Tests were performed on the pilot unit (PU), which included service tank, precipitation tank, and accumulation tank, 300 dm3 each, connected with pipelines with dosing and pumping equipment, and throttle valves providing controlled dosing, mixing, precipitation and mechanical filtration of radioactive wastewaters under different conditions. The reagent compositions used in pilot tests were based on coagulant POLYPACS-30 LF (aluminum polyoxychloride), synthetic cationic flocculants Besfloc K6634, K510CA, K6732 («Kolon Life Science, Inc», South Korea) varying in molecular weight and charge density, and natural cationic flocculant «Chitofloc» (Institute of Chemistry FEBRAS, Russia). The following wastewater parameters were controlled during the pilot tests: pH, dry residue, oxygen consumption, total α- and β-activity, isotope composition, optical density and DS content. The precipitation setup demonstrated lower efficiency DS removal from evaporator concentrates due to high ionic strength suppressing the electrostatic interactions between coagulants/flocculants and oppositely charged colloids of DS and TUE. The residual DS concentration was below 1 mg/L that corresponds to decontamination factor above 300 for the drainage water samples tested. The chitosan-based “Chitofloc” flocculant appeared to be the reagent which was the least sensitive to negative effect of ionic strength; however, the decontamination factor in DS removal was not higher than 5 due to suppressing of electrostatic interactions in high salinity media. Analysis of α-activity of water samples after flocculation/mechanical filtration revealed that TUE were not detected in the drainage water samples with DS content reduced to 2 mg/L that corresponds to TRU decontamination factor above 10000 and confirms immobilization of TUE in DS precipitate.
Development of Comprehensive Techniques for Coastal Site Characterisation: Integrated Palaeohydrogeological Approach for Development of Site Evolution Models
Authors: Amano Kenji , Niizato Tadafumi , Ota Kunio , Lanyon Bill , Alexander W. Russell
Published: August 2012

Radioactive waste repository designs consist of multiple safety barriers which include the waste form, the canister, the engineered barriers and the geosphere. It is widely considered that the three most important safety features provided by the geosphere are mechanical stability, favourable geochemical conditions and low groundwater flux. To guarantee that a repository site will provide such conditions for timescales of relevance to the safety assessment, any repository site characterisation has to not only define whether these features will function appropriately today, but also to assess if they will remain adequate up to several thousand to hundreds of thousand years into the future, depending on the repository type. The case study described here is focussed on the palaeohydrogeology of the coastal area around Horonobe in northern Hokkaido, Japan. Data from JAEA’s ongoing underground research laboratory project is being synthesised in a Site Descriptive Model (SDM) with new information from the collaborating research institutes to develop a Site Evolution Model (SEM), with the focus very much on changes in the Sea of Japan seaboard over the last few million years. This new conceptual model will then be used to assess the palaeohydrological evolution of the deep geosphere of coastal sites of Japan.
The Removal, Transportation and Final Treatment and Conditioning of the THETIS Research Reactor Spent Fuel of the University of Ghent (Belgium) Achieved in 2010
Authors: Thierens Hubert , Monsieurs Myriam , De pooter Vincent , Noynaert Luc , Maris Patrick , Ooms Luc , Lewandowski Patrick , Braeckeveldt Marnix , Strijckmans Karel
Published: August 2012

The THETIS research reactor on the site of the Nuclear Sciences Institute of the Ghent University has been in operation from 1967 until December 2003. This light-water moderated graphite-reflected low-enriched uranium pool-type reactor has been used for various purposes e.g. the production of radio-isotopes and activation analyses. During the first years its core power was 15 kW. In the early ’70, a core enlargement allowed for operation at typically 150 kW, while the maximum was allowed to be 250 kW. The fuel was 5% enriched uranium cladded with AISI304L stainless steel, with graphite plugs at both ends of the tubes. In order to decommission the reactor, the spent fuel and other nuclear materials present had to be removed from the reactor site. Ghent University entrusted SCK·CEN, the Belgian Nuclear Research Centre, with the study of the further management of the spent fuel. Various options such as reprocessing, intermediate storage awaiting final disposal were investigated. However the characteristics and the small amount of spent fuel (84.64 kg of UO2) made these solutions very expensive. In the meantime ONDRAF/NIRAS, the Belgian radioactive waste management agency, was developing together with Belgoprocess, a solution for final conditioning in 400 liter drums and further intermediate storage of the spent fuel in its nuclear facilities at the BELGOPROCESS site in Dessel. This conditioned waste is foreseen to enter the future geological disposal site after the intermediate storage period only after 2050. Finally SCK·CEN recommended this solution for the back-end of the THETIS spent fuel and Ghent University declared this spent fuel as radioactive waste. Once the feasibility for conditioning and storage was demonstrated, further actions were taken in order to unload the spent fuel out of the reactor and to transport it to the PAMELA-installation at the Belgoprocess site in Dessel. Finally after receiving all necessary licensing authorisations from the FANC/AFCN, the Belgian nuclear safety authority, the operations started at the reactor site beginning of 2010 and the spent fuel was placed into the intermediate storage building after conditioning at the Belgoprocess site at the end of 2010. The paper will focus on: - the inventarisation and characterization of the spent fuel and other nuclear materials; - the operations at Ghent University and Belgoprocess sites; - the conclusions drawn from the operations.
Perspectives on Reversibility and Retrievability of a Geological Repository in France
Authors: Dumont Jean-Noe¨l , Hoorelbeke Jean-Michel
Published: August 2012

In France it is required by law that the future geological repository for high-level waste (HLW) and intermediate-level long-lived waste (ILW-LL) be “reversible.” but the conditions for reversibility will be defined after the license application for construction. This situation requires scientists to establish dialogue with the stakeholders and issue specific proposals before a new law is promulgated as a preliminary to obtaining authorization to build a waste repository. Such dialogue can identify the various expectations of stakeholders. These expectations call for answers of various natures: better mutual understanding through clarifications on specific subjects, technical features to be incorporated in the design, and proposals for governance features of the disposal process. These answers are being developed in the framework of the program conducted by Andra. This dialogue is a continuous process. While a major decision point will be met around 2016, concerning the license for construction, discussion and revisions will continue as part of the disposal program, until the final closure of the repository, or even later.
35 Years of Incineration in Studsvik: Lessons Learned and Recent Modifications and Improvements
Authors: Lindberg Maria , Lo¨vstrand Joakim , von Kronhelm Karin
Published: August 2012

Since the incinerator in Studsvik was taken into operation in 1976 it has been operating at a level of 350–500 tonnes per year. The incinerator treats waste from both the nuclear industry and from other sectors generating radioactive waste such as hospitals, research companies/facilities and academic institutions. The incineration facility has been upgraded several times during its lifetime. The upgrades includes, change of off gas treatment as well as new control systems and currently the commissioning of a sister pyrolysis plant. Several new waste streams have also been approved for treatment in the last few years.
State of the Art for Fabricating and Emplacing Concrete Containers Into Large Horizontal Disposal Caverns in the French Geological Repository
Authors: Bosgiraud Jean-Michel , Guariso Maurice , Pineau Franc¸ois
Published: August 2012

The research and development work presented in this paper was initialized by Andra in 2007. The work necessary for manufacturing and testing a full scale demonstrator is presently implemented. The case story is twofold. The first part is related to the initial development of a high performance concrete formulation used for fabricating concrete storage containers (containing Intermediate Level and Long Lived Waste primary canisters) to be stacked and emplaced into 400-m long concrete lined horizontal disposal vaults (also called cavern), excavated in the Callovo-Oxfordian clay host formation at a 550 to 600-m depth, with an inside diameter of approximately 8-m. The fabrication of the concrete boxes is illustrated. The second part presents the outcome at the end of the detailed design phase, for a system which is now being manufactured (for further test and assembly), for the emplacement of the concrete containers inside the vault. The application was engineered for remote emplacing a pile of 2 concrete containers (the containers are preliminarily stacked in a pile of 2, inside a hot cell, thanks to a ground travelling gantry crane). The emplacement process is justified and the related emplacement synoptic is illustrated. The test campaign is scheduled in 2011–2012. The successful completion of the technical trials is mandatory to confirm the mechanical feasibility of remotely emplacing concrete containers into large horizontal disposal caverns over long distances. The later display of the machinery at work in Andra’s showroom will be instrumental for the confidence building process involving the various stakeholders concerned by the public enquiry period (mid-2013) preceding the deep geological repository license application (2014–2015).
Environmental Remediation of an ALSTOM Grid Industrial Site (France)
Authors: Rome´ro Ste´phanie
Published: August 2012

ALSTOM Grid is the project owner of the remediation of a former industrial site, located in Saint-Ouen, north of Paris. The industrial activity (power transformer production) started in 1921 and stopped in 2006. The type of pollution is linked with the former activity. It’s an organic pollution: hydrocarbon, PCB and chlorinated volatile organic compounds. The clean-up concerns soil and groundwater. The main specificity of the project is that the remediation is operated inside the existing industrial buildings which must be kept in place and restituted to the owner after the works. The treatment of soil requires excavating soil up to 9 m deep (1 m under the level of the groundwater) inside the buildings. As a consequence, some impressive devices were set up to ensure the stability of the buildings during the clean-up, like support structures of the foundations and strengthening of the building fronts. In the same time, it has to be pointed out that great diversity of clean-up actions is performed on the site: the soil is excavated to be treated on site (bioremediation or chemical treatment) or off site. The treatment of groundwater consists of pumping the oil staying on the surface and oxidizing the dissolved pollution. This project is probably the first experience of this scale in France with multi-contaminated soil and groundwater decontamination in keeping and reinforcing the existing buildings.
Magnox Swarf Storage Silo Liquor Effluent Management: Sellafield Site, Cumbria, UK
Authors: Le Clere Stephen
Published: August 2012

The Sellafield Magnox Swarf Storage Silo (MSSS) was constructed to provide an underwater storage facility for irradiated magnox cladding metal swarf, as well as miscellaneous beta-gamma waste from several sources. Liquid effluent arisings from hazard reduction activities at this facility represent the toughest effluent treatment challenge within the company’s Legacy Ponds & Silos portfolio. The key requirement for hazard reduction has generated many substantial challenges as the facility is readied for decommissioning. This has demanded the production of carefully thought out strategies for managing, and overcoming, the key difficulties to be encountered as hazard reduction progresses. The complexity associated with preparing for waste retrievals from the Magnox Swarf Storage Silo, has also generated the demand for a mix of creativity and perseverance to meet the challenges and make progress.
Development of a Computer System for Support and Documentation of Clearance of Buildings of Nuclear Installations
Authors: Winkler Christoph , Wo¨rlen Stefan , Thierfeldt Stefan , Schartmann Frank
Published: August 2012

Clearance of buildings of nuclear installations in Germany has to comply with general requirements laid down in Section 29 of the German Radiation Protection Ordinance (RPO; Strahlenschutzverordnung), clearance levels as stipulated in Table 1 of Annex III RPO and technical guidance given in the technical standard DIN 25457. Clearance procedures are usually proposed by the licensee and are made part of a decommissioning licence after appropriate review by the authorities. The clearance procedures consist of a radiological characterisation of the various parts of the buildings, a decontamination step if necessary, followed by measurements for checking the success of decontamination, and the actual decision measurements for demonstrating compliance with clearance levels. These three steps rely on various measurement methods, selected depending on the relevant radionuclides to be measured, like sampling with subsequent gamma spectrometry (and if necessary separate evaluation of beta and alpha emitting nuclides), measurements with surface contamination monitory and in situ gamma spectrometry. For a nuclear power plant, the number of samples taken during this procedure can easily reach a few 10,000, while the number of single measurements can reach several 100,000 single surface measurements and several 10,000 measurements with in situ gamma spectrometry. This large number of data together with the correct interpretation according to the valid nuclide vector, the penetration depth of the contamination, the correction for radioactive decay etc have to be managed, which is a error-prone process if carried out manually or with insufficient support by customised software. For this reason, a versatile software tool has been developed by Brenk Systemplanung GmbH that supports the operator in all aspects of clearance of buildings, based on the extensive experience with various clearance procedures in nuclear power plants that Brenk Systemplanung has carried out in the past. Many steps of the data evaluation and report generation during the clearance process outlined above can now be carried out automatically. This includes the time-consuming steps of: • support during the acquisition of the geometry data for a room, • generation of the lists for sampling and measurements, • automatic acquisition of the measurement and sampling results from the instruments, • evaluation of compliance with clearance levels, both for measurements on the entire surface and for statistically distributed measurements or samples, • generation of the documentation on clearance, • storage of all data, generation of nuclide vectors, correction for radioactive decay, connection with photo, CAD drawings etc.
Uranium Metal Oxidation, Grinding, and Encapsulation in BoroBond®: TRU Waste Management
Authors: Cook Kevin S. , Addington Larry A. , Utley Beth
Published: August 2012

Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond®, pre-oxidation of uranium metal with Fenton’s reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation in BoroBond® resulted in pressure increase rates at 60 °C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fenton’s reagent system consisting of a simple reagent mix of FeSO4·7H2O, H2O2 and HCl effectively oxidized 1/4″ cubes of uranium metal in under four days at room temperature. Increased peroxide addition rate, increased FeSO4·7H2O concentration and low pH all increase the corrosion rate. Densalloy SD170 with an average particle size of 581 μm with 7.63% of particles less than 90 μm was milled so that over 90% of the Densalloy mass measured less than 90 μm in 6 hours of milling. Acceptable wear rates were seen on wear components that were from standard materials (Nitronic SS and 440SS).
Clearance of Buildings of NPP Wu¨rgassen (Germany) Using an Innovative Clearance Concept
Authors: Wo¨rlen Stefan , Steinbusch Marco , Kummer Alexander , Thierfeldt Stefan , Schartmann Frank
Published: August 2012

The nuclear power plant Wu¨rgassen in Germany entered into decommissioning in 1995. The project has currently advanced to the clearance of the buildings. Clearance of buildings has commenced in 2007 with establishing the clearance procedure in pilot rooms. It has entered the “productive” phase in 2010 and will last until about 2014. In 2007, Brenk Systemplanung has been commissioned by E.ON / NPP Wu¨rgassen to perform clearance of the pilot rooms in the turbine building. It has been one of the goals of this work to establish an innovative clearance procedure which would not rely on 100% surface measurements but which should be based on the use of in situ gamma spectrometry and the application of approaches to minimise the number of clearance measurements. In 2010 and 2011 Brenk Systemplanung has been commissioned by E.ON / NPP Wu¨rgassen to perform clearance of rooms with partly considerably higher contamination levels in other buildings. Central points of the concept for clearance of buildings which is presented in this paper were the establishment of averaging areas of up to 10 m2, which is an ideal size for the use of in situ gamma spectrometry, and the application of statistical methods, which allow an efficient performance of the clearance measurements. As a first step the measurement instrumentation necessary for the performance of clearance measurements had to be qualified. The innovative concept was first tested and implemented in the pilot rooms and has now been used also in various other parts of the NPP Wu¨rgassen. The procedure which has been established by Brenk Systemplanung for NPP Wu¨rgassen allows rapid measurements making best use of the averaging area and measurement area of collimated in situ gamma spectrometry and minimising the number of single measurements in comparison to other clearance procedures for nuclear installations of similar size. The paper highlights the relevant approaches and the conceptual differences and innovations.
Use of Highly Pressurized Liquid Nitrogen Technology for Concrete Scabbling Application at SICN Nuclear Facility
Authors: Moggia Fabrice , Vaudey Claire-Emilie , Varet Thierry , Toulemonde Vale´rie , Richard Fre´de´ric , Anderson Gary , Damerval Fre´de´rique
Published: August 2012

The decommissioning process is a quite long and complicated stage who may take few years or decades to be achieved. Generally, this process involves the implementation of a large number of technologies dedicated to cutting and decontamination operations. Based on this finding, the Clean-Up Business Unit of AREVA with Air liquide decided to start the development of a new technology based on the use of liquid nitrogen (−140°C / 3500 bar).
Mobility of Ra-226 and Heavy Metals (U, Th and Pb) From Pyritic Uranium Mine Tailings Under Sub-Aqueous Disposal Conditions
Authors: Dave´ Nand K.
Published: August 2012

Laboratory lysimeter studies were undertaken to evaluate the leaching characteristics and mobility of Ra-226 and other heavy metals (U, Th and Pb) from pyritic uranium mine tailings under sub-aqueous disposal conditions for assessing the long-term radiological stability of such waste repositories. The experiments were conducted using three types of un-oxidized tailings: fine, coarse and gypsum depleted mill total tailings. The results showed that Ra-226 was leached from surface of the submerged tailings and released to both surface water and shallow zone pore water during initial low sulphate ion concentration of the surface water cover in all three cases. The release of Ra-226 was further enhanced with the onset of weak acidic conditions in the surface water covers of both coarse and gypsum depleted mill total tailings. With additional acid generation and increasing sulphate and iron concentrations, the dissolved Ra-226 concentrations in the water covers of these tailings gradually decreased back to low levels. Pb was also leached and mobilized with the development of moderate acidic conditions at the surface of the submerged coarse and gypsum deplete tailings. No leaching of U and Th was observed.