bluuebunny/crossref_metadata_2025 huggingface dataset:
Visualization of Radioactive Sources Without Gamma-Radiation With UV Imaging Systems
Authors: Ivanov Oleg , Danilovich Alexey , Stepanov Vyacheslav , Smirnov Sergey , Volkovich Anatoly
Published: December 2010

New UV cameras are suitable for imaging of α-contamination by fluorescence of atmospheric air the near ultraviolet (wavelength — 280 – 390 nm) region. Their parameters are: FOV for detecting in UV spectral region is 8° × 6°. The optical FOV is about 48° × 36°. DayCor SUPERB UV camera has sensitivity 3×10−18 W/cm2 enables detection and displaying corona emission as weak as 1.5 pC at distance 8 m, and capture moving targets without smearing the output image. Instruments sensitivity for alpha contamination registration in terms of minimum measurable activities (MMA) have been estimated as, 40–100 Bq/cm2 (measurement time is 3600–600 sec correspondently).
Establishing a Site for a Slovenian LILW Repository
Authors: Virsˇek Sandi , Sˇpiler Janja , Veselicˇ Miran
Published: December 2010

In this paper we will describe the practice and siting process for a LILW repository in Slovenia. Slovenia is a small country, and, consequently, its nuclear programme is small. It does include almost everything that large programmes have, but we still do not have a repository. In 2004, the Slovenian Radwaste management agency (ARAO) initiated a new procedure to find a proper location for a LILW repository. ARAO asked all Slovenian communities to participate. All communities have the option to withdraw from the process until they have approved the site. We received eight responses, and we prepared methodologies with different parameters to evaluate all potential sites. All sites were assessed by Slovenian experts, and determinations were based on using prepared methodologies. On the basis of their expert opinions, we prepared a prefeasibility study and chose the three most suitable sites. With these three we continued the process. After that time, one of the communities withdrew from the procedure, and another one reversed its proposed site and proposed a new one. For the third site we continued with the programme, and we prepared a feasibility study with a Strategic Environmental Impact Assessment and preliminary safety calculations for the comparison of different concepts of disposal units. If everything goes according to the plan for this site and the concept, we expect site approval in the first half of 2009. After that, we will start preparing everything necessary for building permission.
Novel Antimonysilicate Material Quasar-N for the Removal of Radionuclides From Acidic Decontamination Liquids
Authors: Harjula Risto O. , Paajanen Airi
Published: December 2010

Novel antimony silicate material, commercially available from PQ Corporation (previously Ineos Silicas) is highly selective for the removal of several key radionuclides (Co-60, Sr-90, Cs-137, Pu-236 and Am-241) from acidic and neutral nuclear waste effluents. The paper will summarise most the key results that have been obtained in the previous studies of the material. In addition, new test results on the removal Co-60, Sr-90, Cs-137 and Am-241 from acid media are reported. Static batch experiments and column experiments show that Am can be removed efficiently from nitric and oxalic acid, indicating that Quasar is suitable e.g. for the purification of acidic decontamination solutions.
Thermal Safety Analysis of a Dry Storage Cask for the Korean Standard Spent Fuel
Authors: Cha Jeonghun , Kim S. N. , Choi K. W.
Published: December 2010

A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed.
Separation of Rare Earth Precipitates From LiCl-KCl Eutectic Salts by a Distillation at a Reduced Pressure
Authors: Eun Hee-Chul , Yang Hee-Chul , Cho Yung-Zun , Park Hwan-Seo , Lee Han-Soo , Kim In-Tae
Published: December 2010

Distillation and condensation characteristics of LiCl-KCl eutectic salts containing rare earth precipitates were investigated to separate the rare earth precipitates from the salts effectively. The distillation flux of the salts was increased by about 1,000 times by reducing the ambient pressure from 760 Torr to 0.5 Torr. The salt vapors were almost changed into salt lumps during a salt distillation at the ambient pressure of 0.5 Torr and they were collected in the condensed salt storage. However, fine salt particles were formed when the salt distillation was processed at 10 Torr and it is difficult for them to be recovered. Therefore, it is thought that a salt vacuum distillation and condensation should be processed to recover almost all of the vaporized salts at a pressure below 0.5 Torr.
Separation of Fission Products and Actinides From Savannah River Site High-Level Nuclear Wastes
Authors: Peters T. B. , Poirier M. R. , Fondeur F. F. , Nash C. A. , Hobbs D. T. , Fink S. D.
Published: December 2010

Separation methods for the pretreatment of the liquid fraction of high-level nuclear waste (HLW) at the Savannah River Site (SRS) include solvent extraction for the separation of cesium and adsorption/ion exchange for the removal of strontium and alpha-emitting actinides. The solvent extraction process, referred to as Caustic Side Solvent Extraction or CSSX, uses a calixarene extractant in combination with phase modifiers in a hydrocarbon diluent. Monosodium titanate (MST), a hydrous metal oxide, is the baseline material for the removal of strontium and alpha-emitting radionuclides (principally 238Pu, 239Pu, 240Pu and 237Np). Two pretreatment facilities, the Modular Caustic Side Solvent Extraction Unit (MCU) and the Actinide Removal Process (ARP) facility began radioactive operations at SRS in 2008. Together these facilities can treat approximately 4 million liters of waste per year. The same separation processes are also planned for the much larger Salt Waste Processing Facility (SWPF). The SWPF, which has a design throughput of about 27 million liters per year, is scheduled to begin radioactive operations in 2013. This paper presents an overview of the separation processes as well as recent research and development activities aimed at improving separation performance in the pretreatment facilities.
MARSAME: Multi-Agency Radiation Survey and Assessment of Materials and Equipment
Authors: Snead Kathryn , Azzam Nidal , Petullo Colleen , Bhat Ramachandra , Bias Craig , Alberth David , Falo Gerald , Doremus Steven , Williams W. Alexander , Anderson Amanda , Meck Robert , Powers George
Published: December 2010

The MARSAME manual is a technical resource that describes processes and methods for the measurement of radionuclides in or on materials and equipment. These methods and processes are used to make the decision that a radionuclide is present and, if so, how to quantify the amount and its uncertainty. They are also used to make the decision that no radionuclide more than background is present. The MARSAME manual is technically defensible, and endorsed by the U.S. Department of Defense, the U.S. Department of Energy, the U.S. Environmental Protection Agency, and the U.S. Nuclear Regulatory Commission. The MARSAME manual was published in January 2009, and is available for download from the following website: http://www.epa.gov/rpdweb00/marssim/marsame.html. The MARSAME manual is a supplement to the MARSSIM manual, and closely follows the concepts and processes developed therein. The MARSSIM manual describes processes and methods for the measurement of radionuclides on the surfaces of buildings and soils. The MARSAME manual follows the Data Quality Objectives process to describe the phases of: plan, implement, assess, and decide. A strong emphasis on the planning phases ensures that survey data are the right quantity and quality to meet the Data Quality Objectives and Measurement Quality Objectives for the survey. Non-parametric statistics are used as the tool to make decisions, based on hypothesis testing. Measurement uncertainty, measurement detectability, and measurement quantifiability are important Measurement Quality Objectives. Calculation of measurement uncertainty follows International Standardisation of Measurement Guide 98, and thus, uncertainty estimates include method uncertainties as well as counting uncertainties. The MARSAME manual emphasizes that detection is determined by the “critical level” and not the “minimum detectable concentration.” Adjustments to Poisson statistics are given for counting in a low background. The “minimum quantifiable concentration” specifies the relative measurement standard deviation, generally chosen to be 10%. Data Quality Assessment on the survey results ensures that they are of the right type and quality to support the final decision.
Experience in Choices for Decommissioning the Dounreay Site
Authors: Catlow Fred
Published: December 2010

The paper describes the public participation from the viewpoint of a stakeholder and member of the public. The dialogue between various members of the Dounreay Stakeholders Committee vary widely and do not always seem to represent the views of the wider public. Whilst great care has been taken to select various options for the ultimate condition of the Dounreay site and these have been discussed at great length and the preferred option selected by consensus, there still appears to be some conflict within the local community. It is probable that if the local population had to vote on the options for the future of the Dounreay nuclear site the outcome would be vastly different from that of the Stakeholders Committee. Whilst the politicians have been elected by the people, they represent a distinctly anti-nuclear view (even to the extent of decommissioning) whereas many local people (especially the workers on the Dounreay site) would prefer to see a continuation of nuclear activity at Dounreay. The problem is not only with local politicians but at national level in Scotland itself where the Scottish National Party has formed a coalition with the Green Party on condition all nuclear activities are phased out.
Research Reactor Spent Nuclear Fuel Shipment From the Czech Republic to the Russian Federation
Authors: Svitak Frantisek , Svoboda Karel , Podlaha Josef
Published: December 2010

In May 2004, the Global Threat Reduction Initiative agreement was signed by the governments of the United States and the Russian Federation. The goal of this initiative is to minimize, in cooperation with the International Atomic Energy Agency (IAEA) in Vienna, the existing threat of misuse of nuclear and radioactive materials for terrorist purposes, particularly highly enriched uranium (HEU), fresh and spent nuclear fuel (SNF), and plutonium, which have been stored in a number of countries. Within the framework of the initiative, HEU materials and SNF from research reactors of Russian origin will be transported back to the Russian Federation for reprocessing/liquidation. The program is designated as the Russian Research Reactor Fuel Return (RRRFR) Program and is similar to the U.S. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program, which is underway for nuclear materials of United States origin. These RRRFR activities are carried out under the responsibilities of the respective ministries (i.e., U.S. Department of Energy (DOE) and Russian Federation Rosatom). The Czech Republic and the Nuclear Research Institute Rez, plc (NRI) joined Global Threat Reduction Initiative in 2004. During NRI’s more than 50 years of existence, radioactive and nuclear materials had accumulated and had been safely stored on its grounds. In 1995, the Czech regulatory body, State Office for Nuclear Safety (SONS), instructed NRI that all ecological burdens from its past activities must be addressed and that the SNF from the research reactor LVR-15 had to be transported for reprocessing. At the end of November 2007, all these activities culminated with the unique shipment to the Russian Federation of 527 fuel assemblies of SNF type EK-10 (enrichment 10% U235) and IRT-M (enrichment 36% and 80% U235) and 657 irradiated fuel rods of EK-10 fuel, which were used in LVR-15 reactor.
Characteristics of the Spent Fuel Generated in Korea
Authors: Kook Donghak , Choi Jongwon , Choi Heuijoo , Cho Dongkeun
Published: December 2010

Nuclear power has satisfied the national electric power demand for three decades, and there are only two reactor types in Korea. The nuclear fuel species, however, have a large variety of fuel types, dimensions, initial enrichment, and fuel supply vendors. A spent fuel accumulation problem has arisen like any other country that uses nuclear power. The spent fuel wet storage capacity in the reactor pool is getting close to its limit, and so hence, short & long-term solutions are being actively proposed. First the general status for the nuclear industries and spent fuels will be introduced, then spent fuel characteristics will follow, and last the future anticipation of spent fuel management will close this article.
Modelling Radionuclide Mobility of Spent Fuel in a Deep Geological Repository Located in a Clay Bedrock
Authors: Duro Lara , Tamayo Abel , Bruno Jordi , Marti´nez-Esparza Aurora
Published: December 2010

Source term models are widely used to assess the behaviour of spent nuclear fuel after final disposal. However, most models do not take into account some phenomena which are expected to control the transport of radionuclides through the near field. Some uncertainties arise from this fact, thus making it difficult to obtain proper simulations of radionuclide behaviour in the near field. In this work, we have used a compartmental code to build up an integrated source term model in an attempt to overcome the abovementioned drawbacks. The model developed takes into account radiolytically-mediated matrix dissolution, radioactive decay chains, diffusive transport, and retardation by sorption and secondary phase precipitation, among other processes. In addition, this model has been used to estimate radionuclide mobility from spent fuel located in a conceptual clay geological repository.
Design Options for the UK’s ILW Geological Disposal Facility
Authors: Hicks Tim , White Matt , Baldwin Tamara , Chapman Neil , Neall Fiona , McKinley Ian , Hooker Paul , Richardson Phil , King Samantha
Published: December 2010

Over the last few years, a major national programme of public consultation has been under way in the UK resulting, in 2006, in the announcement by government of geological disposal as the most appropriate solution for the long-term management of the UK’s long-lived and higher-activity radioactive waste and the launch, in 2008, of an implementation programme. The approach being pursued is to solicit volunteer communities to host a geological disposal facility, which may contain not only intermediate-level waste (ILW) and some low-level waste (LLW), but also high-level waste (HLW), any spent fuel (SF) declared as waste, and potentially other materials that may be declared as waste. These wastes have different physical, chemical, thermal and radiological characteristics, and different concepts will be required to accommodate their disposal, potentially in a single facility. The volunteer approach means that the geological environment that might eventually emerge as the preferred location is not known at the outset. Indeed, the siting process may require evaluation of several different geological environments because the UK has rich geological variability for such a small landmass. Consequently, the Nuclear Decommissioning Authority (NDA), which is charged with designing, developing and implementing a geological disposal facility, has investigated facility designs that could be appropriate for a wide range of host rocks and geological environments. This paper presents the results of a project carried out on behalf of the NDA to collate and report information on concepts for the geological disposal of ILW/LLW; a separate project carried out a parallel evaluation of options for disposing of HLW and SF. Initially, the range of geological disposal facility design options available worldwide for the disposal of ILW/LLW was evaluated. Nine disposal concepts were identified and reviewed that would cater for any geological environment likely to arise in the UK. These concepts have different engineering and operational aspects. The appropriateness of each option for implementation in five different generic geological environments was assessed using expert judgement, with input from the NDA, consultants and the UK regulatory agencies. The paper presents a set of generic designs derived from the study and discusses the key issues that would need to be addressed should any of these designs be considered for implementation in specific geological environments in the UK. The findings of this work are intended to provide a resource to support comparisons of alternative disposal concepts and the identification of designs suitable for the disposal of UK ILW/LLW in different geological environments.
Approaches for Modelling Transient Unsaturated-Saturated Groundwater Flow During and After Construction
Authors: White Matt , Guimera` Jordi , Ohyama Takuya , Kosaka Hiroshi , Robinson Peter , Saegusa Hiromitsu
Published: December 2010

Construction of underground research laboratories and geological disposal facilities has a significant transient impact on groundwater flow, leading to a drawdown in the water table and groundwater pressures, and groundwater inflow into shafts, access ways and tunnels accompanied by desaturation of the surrounding rock. Modelling the impact of underground facilities on groundwater flow is important throughout the construction and operation of the facilities, e.g. estimating grouting and water treatment facility requirements during construction, and estimating the rate of resaturation of the engineered barrier system and the establishment of steady-state groundwater flow after backfilling and closure. Estimating the impact of these effects requires modelling of transient groundwater flow under unsaturated conditions at large scales, and over long timescales. This is a significant challenge for groundwater flow modelling, in particular because of the non-linearity in groundwater flow equations, which can have a marked effect on suitable timesteps for transient calculations. In addition, numerical grids need to be developed at appropriate scales for capturing the transition between saturated and unsaturated regions of the sub-surface, and to represent the features of complex hydrogeological structures such as heterogeneous fractured rock. The Japan Atomic Energy Agency (JAEA) has been developing modelling techniques to overcome these problems as part of the Mizunami Underground Research Laboratory (MIU) Project in the Tono area of Gifu Prefecture, Japan. An integrated geological and hydrogeological modelling, and visualisation system referred to as GEOMASS has been developed, which allows for transient unsaturated groundwater flow modelling in the presence of dynamic underground excavation models. The flow simulator in GEOMASS, FracAffinity, allows for such modelling by the application of sophisticated gridding techniques, allowing for modification of hydraulic conductivity in key zones, and by suitable modification of water retention models (the relationship between saturation and pressure, and saturation and hydraulic conductivity). The approaches that have been developed in GEOMASS have been tested through a series of models of increasing complexity, and the testing has demonstrated that there is no significant impact on estimates of regional groundwater flows or local estimates of flow into underground excavations. The tools and approaches that are described in this paper are of significance in all geological disposal projects, where a key requirement is to estimate and understand the hydrogeological regime and the transient response of groundwater flow to underground construction. Such understanding is important for construction, operation and post-closure phases of facility development.
Incineration of Contaminated Oil From Sellafield
Authors: Broadbent Craig , Cassidy Helen , Stenmark Anders
Published: December 2010

Studsvik have been incinerating Low Level Waste (LLW) at its licensed facility in Sweden since the mid-1970s. This process not only enables the volume of waste to be significantly reduced but also produces an inert residue suitable for final disposal. The facility has historically incinerated only solid dry LLW, however in 2008 an authorisation was obtained to permit the routine incineration of LLW contaminated oil at the facility. Prior to obtaining the authorisation to incinerate oils and other organic liquids — both from clean-up activities on the Studsvik site and on a commercial basis — a development program was established. The primary aims of this were to identify the optimum process set-up for the incinerator and also to demonstrate to the regulatory authorities that the appropriate environmental and radiological parameters would be maintained throughout the new process. The final phase of the development program was to incinerate a larger campaign of contaminated oil from the nuclear industry. A suitable accumulation of oil was identified on the Sellafield site in Cumbria and a commercial contract was established to incinerate approximately 40 tonnes of oil from the site. The inventory of oil chosen for the trial incineration represented a significant challenge to the incineration facility as it had been generated from various facilities on-site and had degraded significantly following years of storage. In order to transport the contaminated oil from the Sellafield site in the UK to the Studsvik facility in Sweden several challenges had to be overcome. These included characterisation, packaging and international transportation (under a Transfrontier Shipment (TFS) authorisation) for one of the first transports of liquid radioactive wastes outside the UK. The incineration commenced in late 2007 and was successfully completed in early 2008. The total volume reduction achieved was greater than 97%, with the resultant ash packaged and returned to the UK (for ultimate disposal at the UK LLW Repository) in November 2008.
Modeling of Radiation Fields and Estimation of Doses During Dismantling of RBMK-1500 Reactor Emergency Core Cooling System
Authors: Poskas P. , Simonis A.
Published: December 2010

There is only one nuclear power plant in Lithuania — Ignalina NPP. The INPP operated two similar units with installed capacity of 1500 MW (each). RBMK-1500 were commissioned in 12/1983 and 08/1987, and the original design lifetime was projected out to 2010 and 2015 respectively. But the first Unit of Ignalina NPP was shutdown December 31, 2004, and second Unit will be closed down in 2009. In relation with this, implementation of new technologies for treatment and conditioning of radioactive waste, construction of the interim dry storage facility for spent nuclear fuel and preparation of the licensing documents for decommissioning of the Unit 1, and some other projects were implemented or are under implementation at Ignalina NPP. Implementation of dismantling activities requires detailed knowledge of the radiological situation at the Unit 1. General Radiological Characterization Program for Ignalina NPP Unit 1 based on NUREG-1575 was prepared in 2005–2006 and approved by Regulatory Bodies. In 2008 development of the dismantling projects at Ignalina NPP Unit 1 has been started. Dismantling project of the emergency core cooling system (Building 117/1) is the first project in this field. Based on the radiological surveys data, detailed modeling of the radiation fields has been performed using “VISIPLAN 3D ALARA Planning tool” computer program (SCK CEN, Belgium). So, in this paper detailed information on radiation fields modelling, proposals for optimization of the radiation doses to the personnel, and emergency core cooling system equipment dismantling is presented.
OXIPROBE: A Non Destructive Tool for Determining Steam Generator Oxide Characteristics
Authors: Krasznai John P.
Published: December 2010

CANDU Stations are designed with significant amounts of carbon steel piping in the primary circuit. Although the primary coolant chemistry is such that carbon steel corrosion is minimized, nevertheless magnetite transport from the carbon steel surfaces to the steam generators is a significant issue leading to potential reduction in heat transfer efficiency in the steam generator. There are other contributors to the reduction of heat transfer efficiency such as divider plate leakage whereby some of the coolant short circuits the steam generator tubes and secondary side steam generator tube fouling. CANDU station operators have utilized a number of mitigating measures such as primary and secondary side mechanical and chemical tube cleaning, and divider plate refurbishment to counter these problems but these are all expensive and dose intensive, It is therefore very important to establish the relative contribution of each source to the overall heat transfer degradation problem so the most effective results are obtained. Tube removal and laboratory assessment of the oxide loading is possible and has been utilized but at best it provides an incomplete picture since typically only short lengths of tubes are removed — most often from the hot leg and the tube removal process adversely impacts the primary side oxide integrity. Kinectrics Inc. has developed, qualified and deployed Oxiprobe, a highly mobile non destructive technology able to remove and quantify the deposited oxide loading on the primary surfaces of steam generator tubes. The technology is deployed during shutdown and provides valuable, direct information on: • Primary oxide distribution within the steam generator; • Oxide loading (thickness of oxide) on the primary surfaces of steam generator tubes; • Oxide composition and radiochemical characterization. The End Effector probe can reach either side of the straight section of the steam generator U tube but as currently designed it is unable to be deployed in the U-tube region. The current technology is able to visit 4 tubes simultaneously. The technology is Code classified as a Class 6 fitting by the Canadian Nuclear Safety Commission and registered by the Ontario Technical Standards and Safety Authority as a pressure boundary retaining system. Although the application of the technology to date has been applied to steam generator tubes, in principle it can be applied to any heat exchanger tube, vertical or horizontal. This paper will describe the system, the qualification program for its deployment as well as some actual field results. The applicability of the technology for PWR steam generators is also addressed.
The Sorption and Mechanical Properties of the Modified Concrete Matrix Used for Conditioning of Radioactive Waste
Authors: Dogaru Daniela , Niculae Ortenzia , Dogaru Gheorghe , Jinescu Gheorghita , Duliu Octavian G.
Published: December 2010

The paper describes the results obtained in the laboratory investigations on the mechanical and sorption properties of the cement-based radioactive waste form containing two kinds of sludge in different concentrations. One of them simulates the sludge obtained by treatment of liquid radioactive effluents using an anionic polyelectrolyte named PA-type, and the other one simulates the sludge obtained by decontamination of contaminated surfaces using a hydrogel named pNaAc-type. The influence of the concentration of sludge on the compressive as well as on bending strength of the cement-based radioactive waste forms was studied. The sorption properties of two radionuclides were studied.
Current Status of Phase II Investigations: Mizunami Underground Research Laboratory (MIU) Project
Authors: Tsuruta Tadahiko , Uchida Masahiro , Hama Katsuhiro , Matsui Hiroya , Takeuchi Shinji , Amano Kenji , Takeuchi Ryuji , Saegusa Hiromitsu , Matsuoka Toshiyuki , Mizuno Takashi
Published: December 2010

The Mizunami Underground Research Laboratory (MIU) Project, a comprehensive research project investigating the deep underground environment in crystalline rock, is being conducted by the Japan Atomic Energy Agency at Mizunami City, Central Japan. The MIU Project is being carried out in three overlapping phases: Surface-based Investigation (Phase I), Construction (Phase II), and Operation (Phase III), with a total duration of 20 years. The overall project goals of the MIU Project from Phase I through to Phase III are: 1) to establish techniques for investigation, analysis and assessment of the deep geological environment, and 2) to develop a range of engineering techniques for deep underground application. Phase I was completed in March 2004, and Phase II investigations associated with the construction of the underground facilities are currently underway. Phase II investigation goals are to evaluate the geological, hydrogeological, hydrogeochemical and rock mechanical models developed in Phase I and to assess changes in the deep geological environment caused by the construction of underground facilities. Geological mapping, borehole investigations for geological, hydrogeological, hydrochemical and rock mechanical studies are being carried out in shafts and research galleries in order to evaluate the models. Long-term monitoring of changes in groundwater chemistry and pressure associated with the construction of the underground facilities continue in and around the MIU site, using existing boreholes and monitoring systems. This report summarizes the current status of the MIU Project on results of the Phase II investigations to date.
Evaluation of Mechanical Effects of the Fault Movement on the Engineered Barrier System
Authors: Saito Yuya , Nishimura Mayuka , Hirai Takashi , Tanai Kenji , Naito Morimasa
Published: December 2010

Experiments and numerical analyses have been conducted to investigate the consequences of an active fault crosscutting the engineered barrier system (EBS) of a high-level radioactive waste (HLW) repository. Experiments were performed using laboratory simulation test equipment at a scale of 1:20, which can simulate the chosen shear displacements of 80% and 140% of the buffer thickness, and the chosen shear rates of 0.1 m/s, 0.05 m/s and 0.01 m/s. The experiments have so far demonstrated that the metal overpack will be rotated, but not breached, due to the plasticity of the surrounding bentonite buffer. The total pressure on the bentonite buffer increases with an increase in the shear rate. Numerical analyses were carried out using a finite element method assuming the constitutive model of modified Cam-clay, which was used to assess elasto-plastic behavior of the bentonite buffer.
Abrasive Blasting Unit (ABU)
Authors: Braehler Georg , Welbers Philipp , Kelly Mike , Brunetti Gianfranco , van Regenmortel D.
Published: December 2010

NUKEM Technologies was contracted to supply a dry, automated drum belt (tumbling) Abrasive Blasting Unit (ABU) to the Joint Research Centre of the European Commission in Ispra, Italy. The ABU was installed in the centralised radioactive waste management area of the JRC-Ispra site in Italy. The unit is to be employed for the decontamination to clearance levels of slightly contaminated metal components and, where practical, concrete or heavy concrete (density ∼3200 kg/m3) blocks arising from the dismantling of nuclear facilities. The presentation is based on the successful construction and installation of the ABU at the JRC Ispra site. Among the several possibilities of adapting conventional abrasive units to nuclear applications, an automatic tumbling machine was preferred, due to the larger output and (mainly) for the ease of operation, with minimum direct handling of contaminated material by operators, thus satisfying the ALARA principle. Consideration was also given to Belgoprocess’ successful experience with a predecessor, similar unit. After adequate size reduction batches of up to about 800 kg of material to be decontaminated are automatically introduced into the blasting chamber. Pieces between 100 mm and 800 mm long, between 100 mm and 500 mm wide and between 5 mm and 300 mm high can be effectively treated in the unit, the maximum weight of a single piece being limited to 100 kg. Short lengths of pipe may be included; the final dimensions of pipe to be decontaminated will be established during the nuclear commissioning tests. Other components with hard-to-reach surfaces may also be included. The content of the chamber is tumbled by two bladed drums, while sharp steel grit is sprayed onto the contaminated components, thus removing the surface layer including any contamination. From experience, 30 minutes of treatment is sufficient to remove contamination to levels below expected clearance levels for most materials. The decontaminated components are removed from the blasting chamber automatically and collected in skips. Dust and grit are led to a series of separators; the grit gets recycled to the blasting chamber, cleaned off contaminants such as paint are fed to collection bins, and the dust is bagged into waste drums. Airflow through the whole system cleans the decontaminated components, transports the dust to the collecting area, and acts as a dynamic barrier to limit risks of contamination of the surrounding areas. Prior to release back into the room, the air is filtered in a series of automatically cleaned filters, followed by HEPA filters. The whole facility is operated in an automatic mode: the operators are only required to place drums or pallets of contaminated material onto the feeder, and remove skips of decontaminated material and drums of secondary waste such as dust. The presentation will describe the system and potential applications in the nuclear industry in detail.
Improved Practices for Packaging Transuranic Waste at Los Alamos National Laboratory (LA-UR-09-03293)
Authors: Goyal Kapil K. , Carson Peter H.
Published: December 2010

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 that 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 implemented before they are shipped to WIPP. A defensive approach was used for many years to minimize the number of problematic drums. However, based on review of data associated with headspace gas sampling, NDA and RTR results, and enhanced coordination with the entities responsible for waste certification, many changes have been implemented to facilitate packaging of TRU waste drums with higher isotopic loading at the Plutonium Facility at an unprecedented rate while ensuring compliance with waste acceptance criteria. This paper summarizes the details of technical changes and related administrative coordination activities, such as information sharing among the certification entities, generators, waste packagers, and shippers. It discusses the results of all such cumulative changes that have been implemented at the Plutonium Facility and gives readers a preview of what LANL has accomplished to expeditiously certify and dispose of newly generated TRU waste.
The Results of Testing to Evaluate Crystal Formation and Settling in the Cold Crucible Induction Melter
Authors: Marra James , Stefanovsky Sergey , Lebedev Vladimir , Suntsov Dmitriy
Published: December 2010

The Cold Crucible Induction Melter (CCIM) technology offers the potential to increase waste loading for High Level Waste (HLW) glasses leading to significant improvements in waste throughput rates compared to the reference Joule Heated Melter (JHM). Prior to implementation of a CCIM in a production facility it is necessary to better understand processing constraints associated with the CCIM. The glass liquidus temperature requirement and tolerance to crystal formation for processing in the CCIM is an open issue. Testing was conducted to evaluate crystal formation and crystal settling during processing in the CCIM to gain insight into the effects on processing. A high aluminum/high iron content glass composition with known crystal formation tendencies was selected for testing. A continuous melter test was conducted for approximately 51 hours. To evaluate crystal formation, glass samples were obtained from pours and from glass receipt canisters where the glass melt had varying residence time in the melter. Additionally, upon conclusion of the testing, glass samples from the bottom of the melter were obtained to assess the degree of crystal settling. Glass samples were characterized in an attempt to determine quantitative fractions of crystals in the glass matrix. Crystal identity and relative composition were determined using a combination of x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS). Select samples were also analyzed by digesting the glass and determining the composition using inductively coupled atomic emission spectroscopy (ICP-AES). There was evidence of crystal formation (primarily spinels) in the melt and during cooling of the collected glass. There was evidence of crystal settling in the melt over the duration of the melter campaign.
Understanding Potential Release Mechanisms of Volatile Ruthenium During the Vitrification of High Level Waste
Authors: Lawson Yvonne , Brookes Christopher M. , Steele Carl J. , Sarsfield Mark J.
Published: December 2010

In the U.K., High Level Waste from reprocessing operations is vitrified at the Sellafield Waste Vitrification Plant (WVP). A small number of the nuclides present in the waste have the potential to volatilize during vitrification. In order to prevent release of any radionuclides to the environment it is important to understand the mechanisms by which volatilization may occur and to have suitable controls in place. One element of particular concern is ruthenium, formed during the fission of nuclear fuel, which has the potential to form gaseous species such as RuO4 during the vitrification process and whose behavior must therefore be understood in order to underpin the safe operation of WVP.
The Next EDF Tracking System
Authors: Julli Emmanuelle , Lantes Bertrand
Published: December 2010

EDF’s network of nuclear power plants (NPP) comprises 58 pressurized water reactors. Solid waste arising during plant operation (mainly VLLW, LLW and ILW) are conditioned and sent either to interim storage, an off site treatment plant for additional processing (e.g. the Centraco incinerator or the melting facilities of SOCODEI) or directly to one of the two final repositories operated by ANDRA, the French national radioactive waste management agency. The tracking system allows: - the checking of waste package characteristics against acceptance criteria of the final disposal facilities or off site treatment facilities; and - the transmission of the waste package data to ANDRA and SOCODEI. Since 1992, the EDF computer application DRA has been run on networked computers at EDF and ANDRA, and more recently at SOCODEI. DRA is also a key element in the management of radioactive waste. It allows a large range of inter comparisons to be made between the NPPs in operation and is thus the principal tool used optimize technical and economic performance. After 15 years of use, DRA was technically obsolete and could no longer be successfully developed to meet evolving regulatory requirements. It was therefore decided to completely replace the DRA system and in so doing to introduce new functionality.
Global Developments in Multinational Initiatives at the Back End of the Nuclear Fuel Cycle
Authors: McCombie Charles , Chapman Neil , Isaacs Thomas H.
Published: December 2010

Interest in expanding nuclear power globally continues to grow and various studies are underway to examine all issues associated with much expanded nuclear programmes. The most open questions today are related to the security and non-proliferation implications and to the disposal of radioactive wastes. The security and proliferation concerns have been almost entirely focussed on enrichment technology at the front-end of the nuclear fuel cycle and on reprocessing. Although these are the highest risk areas, it is also important that the potential security problems associated with waste management (in particular with the storage and disposal of spent fuel and radioactive wastes) are not neglected. Furthermore, the costs of national geological repositories imply that, for new or small nuclear programmes, such facilities can be implemented only in the far future, if at all. The international community should continue to strengthen its efforts to highlight the risks and to facilitate solutions that reduce the threats of nuclear materials being distributed widely across the globe. In practice, this challenge has been taken up by a number of organisations that are developing initiatives that can alleviate the potential global security and proliferation problems by promoting multinational approaches to the fuel cycle. This paper addresses those initiatives that are concerned with the storage and final disposal of radioactive wastes and spent nuclear fuel.
Decommissioning of the A-1 NPP Long-Term Storage Facility
Authors: Medved Jan , Vargovcik Ladislav
Published: December 2010

The paper deals with experience, techniques and new applied equipment durig undergoing decommissioning process of the A-1 NPP long-term pool storage and the follow-up decommissioning plan. For rad-waste disposal of the long-term pool storage (where most of the contaminants had remained following the removal of spent fuel) special equipment has been developed, designed, constructed and installed. The purpose of this equipment is the restorage, drainage and fragmentation of cartridges (used as a spent fuel case), as well as treatment of sludge (located at the pool bottom) and of the remaining liquid radwaste. The drainage equipment for cartridges is designed for discharging KCr2 solution from cartridges with spent fuel rods into the handling storage tank in the short-term storage facility and adjustment of the cartridges for railway transport, prior to the liquidation of the spent fuel rod. The equipment ensures full remote visual control of the process and exact monitoring of its technical parameters, including that of the internal nitrogen atmosphere concentration value. Cartridges without fuel and liquid filling are transferred to the equipment for their processing which includes fragmentation into smaller parts, decontamination, filling into drums with their sealed closing and measurement of radioactive dose. For the fragmentation, special shearing equipment is used which leaves the pipe fragment open for the following decontamination. For cleaning the cartridge bottom from radioactive sludge water jet system is used combined with slow speed milling used for preparing the opening for water jet nozzle. The sludge from the cartridge bottom is fixed into ceramic matrix. Nuclear Power Plant JE A-1 (since 1980 in decommissioning) is situated in the locality of Jaslovske´ Bohunice. So far the decommissioning of the Long-term storage was carried out within Stage I of A-1NPP decommissioning. This year the Stage I of decommissioning finished, and the performance of Stage II of decommissioning was started. Decommissioning of the long-term storage facility continues within Stage II of the A-1 NPP decommissioning process.
Implementation of a Geological Disposal Facility (GDF) in the UK by the NDA Radioactive Waste Management Directorate (RWMD): Coupled Modelling of Gas Generation and Multiphase Flow Between the Co-Located ILW/LLW and HLW/SF Components of a GDF
Authors: Bond Alex , Towler George , Paulley Alan , Norris Simon
Published: December 2010

In June 2008 the UK government published a ‘White Paper’ as part of the “Managing Radioactive Waste Safety” (MRWS) programme to provide a framework for managing higher activity radioactive wastes in the long-term through geological disposal. The White Paper identifies that there are benefits to disposing all of the UK’s higher activity wastes (Low and Intermediate Level Waste (LLW and ILW), High Level Waste (HLW), Spent Fuel (SF), Uranium (U) and Plutonium (Pu)) at the same site, and this is currently the preferred option. It also notes that research will be required to support the detailed design and safety assessment in relation to any potentially detrimental interactions between the different modules. Different disposal system designs and associated Engineered Barrier Systems (EBS) will be required for these different waste types, i.e. ILW/LLW and HLW/SF. If declared as waste U would be disposed as ILW and Pu as HLW/SF. The Geological Disposal Facility (GDF) would therefore comprise two co-located modules (respectively for ILW/LLW and HLW/SF). A study has recently been undertaken by NDA RWMD to identify the key Thermo-Hydro-Mechanical-Chemical (THMC) interactions which might occur during both the operational and post-closure phases in order to assess the potential implications of co-location in a range of host rocks. This paper presents supporting modelling work used to help understand the potential interactions between the modules. A multi-phase flow and coupled gas generation model was used to help investigate the potential groundwater and gas fluxes between the modules, in particular considering the operational phase and resaturation behaviour of the different modules. These early phases are important because gas generation rates and hydraulic gradients will be at their maximum, and the pressure gradients associated with GDF operations will, at least initially, dominate over the background hydraulic gradient. The gas generation and multi-phase flow study considered a mudstone host rock in which gas pressurisation might significantly influence resaturation behaviour, or drive water from one module to the other. The results of the multiphase flow modelling show that although gas generation affects pressure recovery in the ILW/LLW module, the smaller size of the HLW/SF excavations compared with the ILW/LLW excavations, and the operational timings, mean that in general the groundwater pressure gradient in the GDF is from the HLW/SF module (higher pressure) to the ILW module (lower pressure). Transport of solutes from the HLW/SF module towards the ILW module is not expected to result in any deleterious interactions, indicating that hydraulic interactions during the resaturation period are unlikely to pose a fundamental barrier to co-location.