bluuebunny/crossref_metadata_2025 huggingface dataset:
Site Survey of Former Naval Base in Andreyev Bay, Northwest-Russia: Radiation Levels and Radionuclide Concentrations On and Below the Surface Level
Authors: Reistad Ole , Steenhuisen Frits , Dick O̸ystein B. , Grepstad Gisle , Hustveit Styrkaar , Amundsen Ingar
Published: June 2009

This paper presents the main results of the program to examine radiation levels and radionuclide concentrations on and below the surface level at the former Russian naval base in Andreyev Bay, Murmansk County. Presently, this base represents an exceptional case regarding future remediation and cleanup operations due to the accident risk (∼ max. fuel inventory of 100 submarine cores) and degree of contamination (over 25 years with continuous release — still ongoing — of radionuclides into the terrestrial and marine environment). The first part of the survey consists of about 1030 measurement points established as a grid with 10 m and 5 m mesh size for the measurement of dose rate in two heights above the ground level (0.1 m, 1 m), radionuclide concentrations, drilling of 50 boreholes for further examination of the radionuclide releases on site and the establishment of a 1:500 map of the area. These surveys were completed 2002–04. The results for dose rate measurements taken 1 m above the ground level varies between background levels and 3 mSv/h. Additional measurements were completed around the main building structures at the site and as part of a geological survey of the site. The activity concentration levels for Cs-137 and Sr-90 were measured in 250 points being part of the same measurement grid as above. The results for both isotopes range from normal fallout levels from atmospheric nuclear weapons testing to above 1 MBq/kg. The main conclusion is that continuous releases of fission products from spent nuclear fuel and fuel residues in defect storage pools have, together with inadequate storage facilities for large amounts of solid radioactive waste, led to heavy contamination of fission products in large areas. The 1:500 map is not public accessible. Thus, the second part of the survey was to analyse and document the results in adequate maps. These maps, geo-referenced to the UTM WGS84 system, have been established on the basis of commercial available satellite images from 2004 and show the measurement grid and results for radiation and activity levels in addition to coastline contour and selected infrastructure elements.
Concept and Development of Information Flow Diagram for Literature Survey
Authors: Noda Masaru , Sato Shoko , Tsuchi Hiroyuki , Koike Akihisa , Ueda Tadashi
Published: June 2009

The Information Flow Diagram for Literature Survey (IFDLS) has been developed to manage information and procedure in the literature survey phase of the PIAs selection process. It is a tool utilizing information technology, which can organize, analyze, and evaluate information from literature survey and manage their process systematically. IFDLS is able to show the flow of information and data, and the history of information management processing. Information coverage and quality is not homogenous throughout the country and, in some areas, there may not even be sufficient data available to be able to reach a judgment on conformity with the site-specific evaluation factors (SSEF). Literature surveys could only be conducted on a volunteer area before it is nominated as a PIA. However, the absence of information on any factor mentioned will not constitute disqualification of the area. On the contrary, an attempt to compare these sites with analogous areas in Japan will be done to assemble sufficient data and consequently make a decision on whether to proceed further. The application of IFDLS to literature survey phase of the PIA selection process is being proposed. The concept, construction, application and evolution of IFDLS towards application phase on a trial basis are discussed.
Melting Characteristics of the Stainless Steel Generated From the Uranium Conversion Plant
Authors: Choi W. K. , Song P. S. , Min B. Y. , Oh W. Z. , Jung C. H.
Published: June 2009

The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO2), calcium oxide (CaO) and aluminum oxide (Al2O3). Also, Calcium fluoride (CaF2), nickel oxide (NiO), and ferric oxide (Fe2O3) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO2) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF2) and a high amount of silica were more effective for a melt decontamination of stainless steel wastes contaminated with uranium. During the melting tests with stainless steel wastes from the uranium conversion plant (UCP) in KAERI, we found that the results of the uranium decontamination were very similar to those of the uranium oxide from the melting of stimulated metal wastes.
Enlargement of the Baldone Near-Surface Radioactive Waste Repository
Authors: Dreimanis A.
Published: June 2009

A unified analysis of the enlargement of the Baldone near-surface radioactive waste (RW) repository RADONS considers the interplay of the existing engineering, safety and infrastructure premises, with the foreseen newly socio-technical features. This enlargement consists in construction of two additional RW disposal vaults and in building a long-term storage facility for spent sealed sources at the RADONS territory. Our approach is based on consecutive analysis of following basic elements: - the origin of enlargement – the RADONS safety analysis and a set of optimal socio-technical solutions of Salaspils research reactor decommissioning waste management; - the enlargement – a keystone of the national RW management concept, including the long-term approach; - the enlargement concept – the result of international co-operation and obligations; - arrangement optimization of new disposal and storage space; - environmental impact assessment for the repository enlargement – the update of socio-technical studies. The study of the public opinion revealed: negative attitude to repository enlargement is caused mainly due to missing information on radiation level and on the RADONS previous operations. These results indicate: basic measures to improve the public attitude to repository enlargement: the safety upgrade, public education and compensation mechanisms. A detailed stakeholders engagement and public education plan is elaborated.
Selection and Evaluation of Inner Material Canididates for Spanish Highlevel Radioactive Waste Canisters
Authors: Puig Francesc , Dies Javier , Sevilla Manuel , de Pablo Joan , Pueyo Juan Jose´ , Miralles Lourdes , Marti´nez-Esparza Aurora
Published: June 2009

This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose.
Factors Affecting Public and Political Acceptance for the Implementation of Geological Disposal
Authors: Neerdael Bernard
Published: June 2009

The main objective of this paper is to identify conditions which affect public concern (either increase or decrease) and political acceptance for developing and implementing programmes for geologic disposal of long-lived radioactive waste. It also looks how citizens and relevant actors can be associated in the decision making process in such a way that their input is enriching the outcome towards a more socially robust and sustainable solution. Finally, it aims at learning from the interaction how to optimise risk management addressing needs and expectations of the public and of other relevant stakeholders. In order to meet these objectives, factors of relevance for societal acceptance conditions are identified, described and analysed. Subsequently these factors are looked for in the real world of nuclear waste management through cases in several countries. The analysis is conducted for six stages of a repository programme and implementation process, from policy development to the realisation of the repository itself. The diversity of characteristics of such contexts increases insight in the way society and values of reference are influencing technological decision making. These interrelated factors need to be integrated in step by step decision making processes as emerging the last years in HLW disposal management. In the conclusions, the effect of each factor on acceptance is derived from the empirical record. In the course of carrying out this analysis, it became clear that acceptance had a different meaning in the first three stages of the process, more generic and therefore mainly discussed at policy level and the other stages, by nature more site-specific, and therefore requesting both public and political acceptance. Experience as clearly addressed in this report has shown that a feasible solution has its technical dimension but that “an acceptable solution” always will have a combined technical and social dimension. If the paper provides tentative answers to the central question how factors affect public and political acceptance, it also aims at illustrating the added value of broadening the technical dimension with social dialogue and insight in value judgements.
Development and Performance Assessment of a Soil Washing Equipment for Soil Contaminated With Radionuclide
Authors: Kim Gye-Nam , Moon Jei-Kwon , Jung Chong-Hun
Published: June 2009

The purpose of this study is to develop a soil washing system and to define the most suitable experimental conditions for the individual elemental equipment in a soil washing system for decontaminating the radioactive soil from around a TRIGA (Training, Research, Isotope, General Atomic) reactor in Korea. Analysis results have shown that the main radionuclides were Cs137 and Co60, the soil particle size ranges from 0.063 mm to 1.0 mm and the radioactive concentration was the strongest in a soil particle smaller than 0.063 mm as predicted. Meanwhile, an oxalic acid was found to be the most efficient chemical agent for washing, especially of cobalt. The scrubbing time of four hours was an optimum time to obtain a removal efficiency of more than 75% for 137Cs and 60Co. A mixing ratio of the soil weight to the volume of the oxalic acid solution, 1:10, was observed to be the best for a washing and it was estimated to be reasonable for 2 cycles of a scrubbing with 1.0M of oxalic acid to avoid a generation of an excessive waste-solution.
Innovative In-Situ Determination of Unsaturated Hydraulic Properties in Deep Loess Sediments in North-West Bulgaria
Authors: Mallants Dirk , Karastanev Doncho , Antonov Dimitar , Perko Janez
Published: June 2009

In the framework of selecting a suitable site for final disposal of low- and intermediate level short-lived radioactive waste (LILW-SL) in Bulgaria, site characterization is ongoing at the Marichin Valog site, North-West Bulgaria. The site is characterized by a complex sequence of loess, clayey gravel, and clay layers, of which the first 30–40 m are unsaturated. Proper knowledge about unsaturated water flow and concomittant radionuclide transport is key input to safety assessment calculations. Constant-head infiltrometer tests were carried out at several meters below ground surface to determine the unsaturated hydraulic properties of silty loess, clayey loess, and clayey gravel layers. Individual infiltrometers were equipped with 0.5-m-long filter sections; the shallowest filter was from 2 to 2.5 m depth, whereas the deepest was from 9.5 to 10 m depth. Infiltration tests provided data on cumulative infiltration and progression of the wetting front in the initially unsaturated sediments surrounding the infiltrometer. A cylindrical time-domain reflectometry TRIME probe was used to measure water content variations with time during progression of the wetting front. Access tubes for the TRIME probe were installed at 0.3 to 0.5 m from the infiltrometer tubes. By means of an inverse optimization routine implemented in the finite element code HYDRUS-2D, field-scale soil hydraulic parameters were derived for all layers. Results show a great consistency in the optimized parameter values, although the test sites were several meters apart. Apparently the size of the affected volume of soil was large enough to reduce the effect of spatial variability and to produce average field-scale hydraulic parameters that are relevant for large-scale predictions of flow patterns and radionuclide migration pathways.
Interim Storage Technology of Spent Fuel and High-Level Waste in Germany
Authors: Geiser H. , Schro¨der J.
Published: June 2009

The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 countries. Spent fuel assemblies of PWR, BWR, VVER, RBMK, MTR and THTR as well as vitrified high level waste containers are stored in full metal casks of the CASTOR® type. Also MOX fuel of PWR and BWR has been stored. More than two decades of storage have shown that the basic requirements (safe confinement, criticality safety, sufficient shielding and appropriate heat transfer) have been fulfilled in any case — during normal operation and in case of severe accidents, including aircraft crash. There is no indication of problems arising in the future. Of course, the experience of more than 20 years has resulted in improvements of the cask design. The CASTOR® casks have been thoroughly investigated by many experiments. There have been approx. 50 full and half scale drop tests and a significant number of fire tests, simulations of aircraft crash, investigations with anti tank weapons, and an explosion of a railway tank with liquid gas neighbouring a loaded CASTOR® cask. According to customer and site specific demands, different types of storage facilities are realized in Germany. Firstly, there are facilities for long-term storage, such as large ventilated central storage buildings away from reactor or ventilated storage buildings at the reactor site, ventilated underground tunnels or concrete platforms outside a building. Secondly, there are facilities for temporary storage, where casks have been positioned in horizontal orientation under a ventilated shielding cover outside a building.
Waste Management Regulations and Approaches in the EU: Potential Areas for Enhancement or Harmonization
Authors: Salzer Peter , Butler Gregg , Chapman Neil , McCombie Charles , McGlynn Grace
Published: June 2009

This paper describes the results of a survey and study performed for the European Commission on “Regulations Governing Radioactive Waste Disposal in EU Countries”. Its main purposes were to provide a survey of the regulations governing the disposal of all forms of radioactive waste in all EU Members States and, based on this study, to consider the potential for harmonization in different regulatory areas. Three key parts of the study are presented and the results discussed: collection and assessment of national data, including its verification by national stakeholders, application of multiattribute analysis methodology to identify optimal waste classification scheme and a workshop of national authorities regulating disposal of radioactive waste. For five determined regulatory issues, the workshop carried out a “strengths, weaknesses, opportunities and threats” (SWOT) analysis of the impacts of harmonization. Opinions expressed in the paper are those of the authors and do not represent the EC’s official position.
Radiation Safety Requirements for Radioactive Waste Management in the Framework of a Quality Management System
Authors: Salgado M. M. , Beni´tez J. C. , Pernas R. , Gonza´lez N.
Published: June 2009

The Center for Radiation Protection and Hygiene (CPHR) is the institution responsible for the management of radioactive wastes generated from nuclear applications in medicine, industry and research in Cuba. Radioactive Waste Management Service is provided at a national level and it includes the collection and transportation of radioactive wastes to the Centralized Waste Management Facilities, where they are characterized, segregated, treated, conditioned and stored. A Quality Management System, according to the ISO 9001 Standard has been implemented for the RWM Service at CPHR. The Management System includes the radiation safety requirements established for RWM in national regulations and in the License’s conditions. The role of the Regulatory Body and the Radiation Protection Officer in the Quality Management System, the authorization of practices, training and personal qualification, record keeping, inspections of the Regulatory Body and internal inspection of the Radiation Protection Officer, among other aspects, are described in this paper. The Quality Management System has shown to be an efficient tool to demonstrate that adequate measures are in place to ensure the safety in radioactive waste management activities and their continual improvement.
Radiological Characterisation of Waste in Interim Storage Building of COVRA
Authors: van Velzen L. P. M. , Welbergen J.
Published: June 2009

At COVRA spatial dose rate distribution measurements were performed in December 2004 and December 2006 in the interim L/ILW storage building (LOG). This storage facility consists out of four large storage halls (height * width * depth; 7m * 40m * 70m) each with a volume of about 20000 m3 (see photo 1). The scope of this study is to investigate the benefits of the waste storage strategy (see figure 1) and procedures for minimization of the dose to the workers and the public. The main aim of the measurements in 2004 was: - to validate the applied L/ILW storage strategy; - to examine, if spatial collected data can be used to detect unforeseen differences in radiation level. The results of these measurements of spatial dose showed a number of unforeseen hotspots at different locations, so that it could be concluded that the applied storage strategy and procedures has to be improved. Further the dose rate at the height of 6 m, mainly responsible for the sky-shine dose rate, being an important part of the dose rate to the public at the site boundary, has to be reduced by more shielding (1). In December 2006 a second serial of spatial radiological and non-radiological data have been collected. The applied nondestructive INDSS-R (INDoor Survey System-Radiation) method has been improved, so that the following 3-dimensional data could be collected between 0.5 m and 5.5 m: - dose rate (by pressurized ionisation chamber). - nuclide depended gamma photon flux (3 * 3 NaI). - temperature and relative humidity. These last two non-radiological parameters were measured to verify the storage conditions of the waste. The main aim of these 3 dimensional collection was to verify the second stated aim of 2004.
Proficiency Test for Non-Destructive Assay of 220 Litre Radioactive Waste Drums by Gamma Assay Systems
Authors: van Velzen L. P. M. , Bruggeman M. , Botte J.
Published: June 2009

The European Network of Testing Facilities for the Quality Checking of Radioactive Waste Packages (ENTRAP) initiated a feasibility study on how to organize in the most cost effective way an international proficiency tests for non-destructive, gamma-ray based, assay of 220 liter radioactive waste drums in the European Union at a regular time interval of 2 or 3 years. This feasibility study addresses all aspects of proficiency testing on radioactive waste packages including the design of a commonly accepted reference 220 liter drum. This design, based on the international response on a send out questionnaire, includes matrixes, radioactive sources; a solution to overcome the tedious and expensive international transport costs of real or even simulated waste packages, general cost estimation for the organization of, and the participation in the proficiency test. The proposed concept for the proficiency testing and the estimated costs are presented. The participation costs of the first proficiency test are mainly determined by the manufacturing of the non-radioactive 220 liter drum (± 55%). Applied reference sources, transport of the drum and reference sources and participation costs in the proficiency test contribute each about ± 15%.
Concrete Crushing and Sampling: A Methodology and Technology for the Unconditional Release of Concrete Material From Decommissioning
Authors: Gilis R. , Lewandowski P. , Ooms B. , Reusen N. , Van Laer W. , Walthe´ry R.
Published: June 2009

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1990, after completion of a pilot project. Two small storage buildings for final products from reprocessing were dismantled to verify the assumptions made in a previous paper study on decommissioning, to demonstrate and develop dismantling techniques and to train personnel. Both buildings were emptied and decontaminated to background levels. They were demolished and the remaining concrete debris was disposed of as industrial waste and green field conditions restored. Currently, the decommissioning operations carried out at the main building have made substantial progress. They are executed on an industrial scale. In view of the final demolition of the building, foreseen to start in the middle of 2008, a clearance methodology for the concrete from the cells into the Eurochemic building has been developed. It considers at least one complete measurement of all concrete structures and the removal of all detected residual radionuclides. This monitoring sequence is followed by a controlled demolition of the concrete structures and crushing of the resulting concrete parts to smaller particles. During the crushing operations, metal parts are separated from the concrete and representative concrete samples are taken. The frequency of sampling meets the prevailing standards. In a further step, the concrete samples are milled, homogenised, and a smaller fraction is sent to the laboratory for analyses. The paper describes the developed concrete crushing and sampling methodology.
Grimsel Test Site — Phase VI: Review of Accomplishments and Next Steps
Authors: Vomvoris Stratis , Kickmaier Wolfgang
Published: June 2009

The Grimsel Test Site owned and operated by Nagra is located in the Swiss Alps (www.grimsel.com). The sixth Phase of investigations was started in 2003 with a ten-year planning horizon. With the investigations and projects of Phase VI the focus is shifted more towards projects assessing perturbation effects of repository implementation and projects evaluating and demonstrating engineering and operational aspects of the repository system. More than 17 international partners participate in the various projects, which form the basic organisational ‘elements’ of Phase VI, each one further structured in field-testing, laboratory studies, design and modelling tasks, as appropriate. Each project phase is planned with a duration of 3 to 5 years, to facilitate all practical and administrative aspects, ensuring flexibility for updating the overall plan with the recent findings. Scientific and engineering interaction among the different projects is ensured via the annual international meeting and ad-hoc meetings, as appropriate. As Phase VI approaches its mid-term point, a review of the accomplishments to date is performed to provide a sound basis for the detailed planning of the next steps. The accomplishments to date are described and assessed below; the opportunities with the on-going projects as well as new projects – currently under discussion – are also presented and discussed. The on-going projects include: studies of the long-term diffusion with emphasis on the processes in the rock matrix (LTD); colloid studies under in-situ generation conditions and migration velocities closer to velocities expected in an actual repository site (CFM); studies of the long-term cement interactions with natural systems (LCS); evaluation of full-scale engineered systems under simulated heat production and long-term natural saturation (NF-Pro/FEBEX); gas migration through engineered barrier systems (GMT); emplacement of shotcrete low-pH plug (ESDRED/Module IV); test and evaluation of monitoring systems (TEM). In addition, various shorter term projects assessing, for example, new geophysical investigation tools, wireless transmission, testing new tools and training for in-situ tracer transport studies have been performed and/or are planned for the near future.
Application of Spent Fuel Characterization and Leaching Studies for Validating Alteration Models
Authors: Quin˜ones Javier , Sabate Joaquin Cobos , Iglesias Eduardo , Rodriguez Nieves , Marti´nez-Esparza Aurora
Published: June 2009

From the Spanish point of view, one of the key issues related to the HLW performance assessment is knowing and predicting, or modelling, the behaviour of spent fuel under geological repository conditions. Taking into account this objective, several experiments have been performed in order to split and determine the influence of different variables on the final stability of the spent fuel matrix in the geological repository. This paper presents some of the leaching results obtained with spent fuel and chemical analogues (UO2, alpha doped – UO2, SIMFUEL,) their application to extrapolate the corrosion behaviour for a long period of time and compare with corresponding data obtained using models. This procedure allows pointing out some of the uncertainties whose minimization is necessary to improve the models useful for performance assessment studies.
Environmental Safety Aspects of the New Spent Nuclear Fuel Management and Storage System at Ignalina NPP
Authors: Poskas P. , Ragaisis V. , Adomaitis J. E.
Published: June 2009

In the framework of the preparation for the decommissioning of the Ignalina Nuclear Power Plant (INPP) a new Interim Spent Nuclear Fuel Storage Facility (ISFSF) will be built in the existing sanitary protection zone (SPZ) of INPP. In addition to the ISFSF, the new spent nuclear fuel management activity will include all necessary spent nuclear fuel retrieval and packaging operations at the Reactor Units, transfer of storage casks to the ISFSF, and other activities appropriate to the chosen design solution and required for the safe removal of the existing spent nuclear fuel from storage pools and insertion into the new ISFSF. The Republic of Lithuania regulations require that the average annual dose to the critical group members of population due to operation of nuclear facility shall not exceed dose constraint. If several nuclear facilities are located in the same SPZ, the same dose constraint shall envelope radiological impacts from all operating and planned nuclear facilities. The paper discusses radiological safety assessment aspects as relevant for the new nuclear activity to be implemented in the SPZ of INPP considering specificity of Lithuanian regulatory requirements. The safety assessment methodology aspects, results and conclusions as concern public exposure are outlined and discussed.
Radiological Characterization of the Nuclear Waste Streams of the Belgian Nuclear Research Centre SCK•CEN
Authors: Maris Patrick , Cornelissen Rene´ , Bruggeman Michel
Published: June 2009

The radiological characterization of nuclear wastes of a research centre is difficult seen the many different processes that generate waste. Since these wastes may contain radionuclides relevant for the disposal option, the nuclide content and activity have to be known. Considering the fact that some wastes are generated only in minor quantities, complex approaches, involving sampling and successive analysis are not justified. Basic physical models can generally be applied to estimate activity ratios, from which the radionuclide inventory can be determined by non-destructive assay on waste-packages. This article discusses waste streams at the Belgian Nuclear Research Centre SCK•CEN and explains how nuclide inventories and activity are determined. The physical models, used to derive activity ratios, and other simple approaches are discussed.
The Prototype Fast Reactor at Dounreay, Scotland: Process and Engineering Development for Sodium Removal
Authors: Mann A. , Herrick R. , Smith M. , Fletcher B. , Gunn J. , Husband W.
Published: June 2009

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal of the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120°C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage and is expected to be deployed within the next few years.
Setup and Organisation of a NDA-System Procurement Project
Authors: Botte John , Gielen Paul
Published: June 2009

Belgoprocess is momentarily in the process of purchasing its fifth NDA-system. Measurement systems are, although based on general designs, not from the shelf items but tailormade sophisticated and highly automated devices. It is obvious that such a project cannot be carried out by solely a NDA team, but needs a multifunctional team. This team combines NDA expertise with experts in civil works, electrical and mechanical engineering, procurement, IT, safety and legal administration. From less positive experiences in the past, Belgoprocess learned a lot and has now a systematic in place. This systematic structures the project from definition of requirements to operation, a two to three year process. This paper defines the phases of a NDA project and gives for each phase some do’s and don’ts. A second subject is the writing and handling of the vast but needed and required documentation. It gives a brief overview of the over thirty documents and files needed. The described, integrated and formal approach reduces the risk of failing projects, systems not meeting the expectations or denied qualification. It puts clear agreements in place, smoothening the relationship between company, supplier and authorities.
Closure of the Fast Flux Test Facility: Current Status and Future Plans
Authors: Lesperance C. P. , Doebler S. V. , Burke T. M.
Published: June 2009

The Fast Flux Test Facility (FFTF) was a 400 MWt sodium-cooled fast reactor situated on the U.S. Department of Energy’s (DOE) Hanford Site in the southeastern portion of Washington State. DOE issued the final order to shut down the facility in 2001, when it was concluded that there was no longer a need for FFTF. Deactivation activities are in progress to remove or stabilize major hazards and deactivate systems to achieve end points documented in the project baseline. The reactor has been defueled, and approximately 97% of the fuel has been removed from the facility. Approximately 97% of the sodium has been drained from the plant’s systems and placed into an on-site Sodium Storage Facility. The residual sodium will be kept frozen under a blanket of inert gas until it is removed later as part of the facility’s decontamination and decommissioning (D&D). Plant systems have been shut down and placed in a low-risk state to minimize requirements for surveillance and maintenance. D&D work cannot begin until an Environmental Impact Statement has been prepared to evaluate various end state options and to provide a basis for selecting one of the options. The Environmental Impact Statement is expected to be issued in 2009.
The Belgian Approach Towards the Study of the Compatibility of Eurobitum With the Geological Disposal Environment
Authors: Valcke Elie , Gens Robert
Published: June 2009

In Belgium, EUROBITUM bituminized radioactive waste containing large amount of soluble salts (NaNO3) is to be disposed of in a final repository in a clay formation. Since the emplacement of the waste will induce many interdependent processes that could negatively affect the interesting radionuclide retarding properties of the clay, the study of the compatibility of EUROBITUM is very complex. To better structure the research and to identify possible knowledge gaps, NIRAS/ONDRAF, the Belgian Radioactive Waste Management Agency, developed the safety functions and safety statements approach. In this paper, we present the application of this approach for the case of EUROBITUM. The approach is illustrated with new and old results on water uptake, swelling, swelling pressure build-up, and ageing, obtained from tests performed in the laboratories of SCK•CEN, the Belgian Nuclear Research Centre.
Determination of 125I and 131I in Radioisotope Wastes
Authors: Kang Sang Hoon , Choi Ke Chon , Lee Heung N. , Han Sun Ho , Jee Kwang Yong
Published: June 2009

In order to measure a low activity of 125I and 131I in radioisotope wastes, we took into consideration various sample preparation and separation methods, such as an acid decomposition, an acid leaching and a combustion method. In a previous study, the maximum chemical yield of iodine by an acid leaching was found to be 78.0%. However, in this study, the maximum chemical yield of the acid decomposition method and the combustion method with a radioiodine reference solution was found to be 99.1% and 84.5%, respectively. We selected the acid decomposition method for the analysis of radioisotope waste samples due to its high chemical yield and short preparation and separation time. The chemical yield of the acid decomposition method depends on the reaction time at each experimental stage, added amount of H3PO3 and H2O2, and the pH of the condensed solution and the condition of the AgI precipitation. The important point for the highest recovery rate from a acid decomposition method is to maintain enough reaction time and pour 10 mL of 30% H3PO3 before a distillation, and drop 1 mL of H2O2 when the condensed solution is trapped in the Florence flask. Through a study of the acid decomposition method we found an optimal preparation and separation method of 125I and 131I in radioisotope wastes due to the merits of a short reaction time and high recovery rate, and a counting system was applied to LEPS for the 125I and HP Ge gamma-ray spectrometer for 131I.
Determination of 35S in Radioisotope Wastes by a Wet Oxidation
Authors: Lee Heung N. , Kang Sang-Hoon , Ahn Hong Joo , Sohn Wook Hyun , Jee Kwang Yong
Published: June 2009

The oxidation studies of a sulfur to a sulfate ion by various oxyhalide oxidants in organic (thiourea, methionine) and inorganic (sulfate, thiophosphate) compounds were carried out in an acidic solution. The optimized result of the oxidation reaction was obtained when a bromate compound (BrO3−) as an oxidant and a 3 M HNO3 solvent. The chemical yield for the oxidation of the organic and inorganic sulfur compounds to a sulfate ion was monitored as 80% for thiophosphate, 87% for methionine, and 100% for thiourea and sulfate within 5% RSD. The oxidation of thiourea required at least 1.6 equivalents of the bromate in an acidic solution. In the case of the oxidation of methionine and thiophosphate, the oxidation yield was above 80% if the bromate was used at 20 times that of the substrates. The chemical yield in the paper sample (WypAll) exceeded 100% because of its background sulfur contents (910 ppm). The sulfate ion was quantitatively measured by using GPC and/or LSC counting of 35S followed by precipitates of BaSO4. The interfering nuclides (14C, 32P) were removed in an acidic condition. The minimum detectable activity (MDA) of 35S was found to be 0.1 Bq/g.
Development of Technology for Cementation of Oil Containing LRW Using Porous Concrete
Authors: Dmitriev S. A. , Varlakov A. P. , Germanov A. V. , Gorbunova O. A. , Barinov A. S. , Vasendin D. R.
Published: June 2009

A new technology of oil containing liquid radioactive waste conditioning has been developed at SIA “Radon”. A porous concrete matrix is placed into special containers and impregnated with oil containing liquid radioactive waste. The waste is effectively fixed in the porous cement matrix. The final product has all the regulated properties. The content of oils in the cement compound can be up to 40% wt. The technology excludes negative influence of oils on hydration of cement which usually occurs at co-cementation of oils with salt liquid radioactive waste.
Challenges Dealing With Depleted Uranium in Germany: Reuse or Disposal
Authors: Mo¨ller Kai D.
Published: June 2009

During enrichment large amounts of depleted Uranium are produced. In Germany every year 2.800 tons of depleted uranium are generated. In Germany depleted uranium is not classified as radioactive waste but a resource for further enrichment. Therefore since 1996 depleted Uranium is sent to ROSATOM in Russia. However it still has to be dealt with the second generation of depleted Uranium. To evaluate the alternative actions in case a solution has to be found in Germany, several studies have been initiated by the Federal Ministry of the Environment. The work that has been carried out evaluated various possibilities to deal with depleted uranium. The international studies on this field and the situation in Germany have been analyzed. In case no further enrichment is planned the depleted uranium has to be stored. In the enrichment process UF6 is generated. It is an international consensus that for storage it should be converted to U3O8. The necessary technique is well established. If the depleted Uranium would have to be characterized as radioactive waste, a final disposal would become necessary. For the planned Konrad repository — a repository for non heat generating radioactive waste — the amount of Uranium is limited by the licensing authority. The existing license wouldn’t allow the final disposal of large amounts of depleted Uranium in the Konrad repository. The potential effect on the safety case has not been roughly analyzed. As a result it may be necessary to think about alternatives. Several possibilities for the use of depleted uranium in the industry have been identified. Studies indicate that the properties of Uranium would make it useful in some industrial fields. Nevertheless many practical and legal questions are open. One further option may be the use as shielding e.g. in casks for transport or disposal. Possible techniques for using depleted Uranium as shielding are the use of the metallic Uranium as well as the inclusion in concrete. Another possibility could be the use of depleted uranium for the blending of High enriched Uranium (HEU) or with Plutonium to MOX-elements.
Characteristics of Solidified Products Containing Radioactive Molten Salt Waste
Authors: Kim In-Tae , Park Hwan-Seo , Cho Yong-Zun , Kim Kwang-Wook , Park Seong-Won , Kim Eung-Ho
Published: June 2009

For a treatment of molten salt wastes generated from a pyroprocessing of oxide spent fuel, we had suggested a stable chemical route, named GRSS (Gel-Route Stabilization & Solidification), and a subsequent consolidation method. By using this method, a series of monolithic wasteforms with different conditions were fabricated, and then their physicochemical properties were investigated. A simulated salt containing 90wt% LiCl, 6.8wt% CsCl, and 3.2wt% SrCl2 was treated with a gel-forming material system, Si/Al/P = 0.4/0.4/0.2 and 0.35/0.35/0.3, and the gel-products were treated at 1100C° after mixing with borosilicate glass powder, where the salt loadings were about 16∼20wt%. The solidified products had a density of 2.3∼2.35g/cm3, a micro-hardness of 4.69∼4.72GPa, a glass transition temperature of 528∼537C°, and a thermal expansion coefficient of 1.65×10−7∼3.38×10−5/C°. Leaching results by the PCT-A method revealed leached rates, 10−3∼10−2g/m2day and 10−4∼10−3g/m2day for Cs and Sr, respectively. From the long-term ISO leaching test, the 900day-leached fraction of Cs and Sr predicted by a semi-empirical model were 0.89% and 0.39%. The leaching behaviors indicated that Cs would be immobilized into a Si-rich phase while Sr would be in a P-rich phase. The experimental results revealed that the GRSS method could be an alternative method for a solidification of radioactive molten salt wastes.
Radioactive Spent Resins Conditioning by the Hot Super-Compaction Process
Authors: Roth Andreas , Centner Baudouin , Lemmens Alain
Published: June 2009

Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of resins) is achievable using hot compaction techniques. This paper will describe the application of the Resin Hot Compaction Process, at Tihange NPP.
Current Trends for Packaging Transuranic Waste at Los Alamos National Laboratory (LA-UR-07-4785)
Authors: Goyal Kapil K. , Carson Peter H. , Enriquez Alejandro E.
Published: June 2009

Transuranic (TRU) waste leaving the Plutonium Facility at Los Alamos National Laboratory (LANL) is packaged using LANL’s waste acceptance criteria for onsite storage. Before shipment to the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, each payload container is subject to rigorous characterization to ensure compliance with WIPP waste acceptance criteria and Department of Transportation regulations. Techniques used for waste characterization include nondestructive examination by WIPP-certified real-time radiography (RTR) and nondestructive assay (NDA) of containers, as well as headspace gas sampling to ensure hydrogen and other flammable gases remain at safe levels during transport. These techniques are performed under a rigorous quality assurance program to confirm that results are accurate and reproducible. If containers are deemed problematic, corrective action is taken before shipment to WIPP. Currently this activity is possible only at the Laboratory’s Waste Characterization, Reduction, and Repackaging Facility. To minimize additional waste requiring remediation, WIPP waste acceptance criteria must be applied at the point of waste generation. Additional criteria stem from limitations of RTR or NDA instruments or lack of appropriate sampling and analysis. This paper presents the changes that have been implemented at the Plutonium Facility and gives readers a preview of what LANL expects to accomplish to expeditiously certify and dispose of newly generated TRU waste.