bluuebunny/crossref_metadata_2025 huggingface dataset:
Solidification of Spent Ion Exchange Resins Into the SIAL® Matrix at the Dukovany NPP, Czech Republic
Authors: Tatranský Peter , Pražská Milena , Harvan Dávid
Published: February 2014

Based on the decision of the State Office for Nuclear Safety, the Dukovany NPP has been obliged to secure the efficient capacities for the disposal of spent ion exchange resins. Therefore, in September 2010, based on the contract with supplier company AMEC Nuclear Slovakia s.r.o. has begun with pumping and treatment of ion exchange resins from the storage tank 0TW30B02, situated in the auxiliary building.

The SIAL® technology, developed in AMEC Nuclear Slovakia, has been used for the solidification purposes. This technology allows an on-site treatment of various special radioactive waste streams (resins, sludge, sludge/resins and borates) at the room temperature.

The SIAL® matrix and technology were licensed by the Czech State Office for Nuclear Safety in 2007.

On-site treatment and solidification of spent ion exchange resins at Dukovany NPP involves process of resin removal from tank using remotely operated manipulator, resin transportation, resin separation from free water, resin filling into 200 dm3 drums and solidification into SIAL® matrix in 200 dm3 drums using the FIZA S 200 facility.

The final product is observed for compressive strength, leachability, radionuclide composition, dose rate, solids and total weight. After meeting the requirements for final disposal and consolidation, the drums are being transported for the final disposal to the Repository at Dukovany site.

During the 3 month’s trial operation in 2010, and the normal operation in 2011 and 2012, 189 tons of dewatered resins have been treated into 1960 drums, with total activity higher than 920 GBq. At the end of trial run (2010), 22 tons of dewatered resins were treated into 235 drums.

During standard operation approximately 91 tons in 960 drums (2011) and 76 tons in 765 drums (2012) were treated. The weights of resins in the drum ware in the range from 89–106 kg and compressive strength limit (10 MPa) has already been achieved 24 hours after fixation. The final measured strength values ranged from 19.0–34.7 MPa and real leachability values ranged from 0.03–0.65%, far below the 4% limit value. Collective effective dose of all workers in 2012 was 7.7 mSv (12.6 mSv in 2011, 6.2 mSv in 2010). Average individual effective dose in 2012 was 0.55 mSv (14 workers), and maximal individual effective dose was 2.25 mSv. This approach allows fast, safe and cost effective immobilization and transformation of dangerous radioactive waste such as sludge and resins into the solid form, which is suitable for long term storage or disposal.
The Development of Non Destructive Remote Measurement Method of Concrete Contamination
Authors: Stepanov V. E. , Potapov V. N. , Ivanov O. P.
Published: February 2014

In the frame of the decommissioning of nuclear power plants or laboratories, the penetration depth of the contamination in concrete surfaces: walls, floor, is unknown. Its knowledge requires sample analysis, that is time consuming and expensive. The main goal of the work is to define and evaluate a non-destructive measurement technique for the evaluation of the contamination depth in concrete. Estimation of accuracy of measurements for different contamination levels, time of measurement, value of natural radionuclides (NRN) concentration in concrete and background radiation dose were carried out. The type of relevant detector depends of selected limiting sensitivity, the weight and sizes of the device.
Graphite Waste Pit Cleanup
Authors: Thiebaut Valérie , Gerber Jean-Luc
Published: February 2014

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments.

In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pit had to be retrieved and transferred into a newer and safer pit. So, this project consists in the full retrieval and transfer of 15 m3 of water mixed with graphite dust located in the decladding facility, as well as the complete cleanup and decontamination of the pit. The equipment and process necessary for retrieval operations were designed, built and tested.

The process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. In the pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit.

As irradiant fragments have been discovered in the pit, specific studies are in progress in order to remove them to the laboratory for dissolution. This work is expected to 2014.
Nuclear Fuel Assemblies and Fuel Debris Monitor for Characterization of Fuel Assemblies for Final Storage
Authors: Sokcic-Kostic M. , Langer F. , Schultheis R. , Klein Ch.
Published: February 2014

NUKEM monitor FAMOS is designed to measure the burn up of fuel assemblies (FA’s) and allows the calculation of the content of fissile material and other actinides. The knowledge of the isotopic composition and activities of FA’s is required for a safe handling and storage of the elements.

The monitor analyses the measurement results on the basis of burn up calculations which improve the quality of the results, with respect to previous systems, which did not use those.
XRD Studies of Cement Embedding Matrices Containing Complexing Reagents From Decontamination Practices
Authors: Ionascu L. , Nicu M. , Turcanu C. , Dragolici F. , Rotarescu Ghe.
Published: February 2014

The study of embedding matrix of radioactive waste is an important issue concerning the need to assure the radiological safety requirements for final disposal. The conditioning of the radioactive wastes by cementation process involves also a structural investigation by X-ray diffraction (XRD) of the samples prepared with cement and complexing agents at different concentration. This paper gives useful information about the influence of complexing agents related to damages produced in concrete microstructures. The experimental data obtained on samples kept in real repository environment for a period up to 15 years, offer structural information regarding the interaction between the different decontamination agents with the major components of the matrix.
D&D Knowledge Management Information Tool: A Web Based System Developed to Share D&D Knowledge Worldwide
Authors: Lagos L. , Upadhyay H. , Shoffner P.
Published: February 2014

Deactivation and decommissioning (D&D) work is a high-risk and technically challenging enterprise within the U.S. Department of Energy complex. During the past three decades, the DOE’s Office of Environmental Management has been in charge of carrying out one of the largest environmental restoration efforts in the world: the cleanup of the Manhattan Project legacy.

In today’s corporate world, worker experiences and knowledge that have developed over time represent a valuable corporate asset. The ever-dynamic workplace, coupled with an aging workforce, presents corporations with the ongoing challenge of preserving work-related experiences and knowledge for cross-generational knowledge transfer to the future workforce [5]. To prevent the D&D knowledge base and expertise from being lost over time, the DOE and the Applied Research Center at Florida International University (FIU) have developed the web-based Knowledge Management Information Tool (KM-IT) to capture and maintain this valuable information in a universally available and easily accessible and usable system. The D&D KM-IT was developed in collaboration with DOE Headquarters (HQ), the Energy Facility Contractors Group (EFCOG), and the ALARA [as low as reasonably achievable] Centers at Savannah River Sites to preserve the D&D information generated and collected by the D&D community. This is an open secured system that can be accessed from https://www.dndkm.org over the web and through mobile devices at https://m.dndkm.org.

This knowledge system serves as a centralized repository and provides a common interface for D&D-related activities. It also improves efficiency by reducing the need to rediscover knowledge and promotes the reuse of existing knowledge. It is a community-driven system that facilitates the gathering, analyzing, storing, and sharing of knowledge and information within the D&D community. It assists the DOE D&D community in identifying potential solutions to their problem areas by using the vast resources and knowledge base available throughout the global D&D community.

The D&D KM-IT offers a mechanism to the global D&D community for searching relevant D&D information and is focused on providing a single point of access into the collective knowledge base of the D&D community within and outside of the DOE. Collecting information from subject matter specialists, it builds a knowledge repository for future reference archiving Lessons Learned, Best Practices, ALARA reports, and other relevant documents and maintains a secured collaboration platform for the global D&D community to share knowledge. With the dynamic nature and evolution of the D&D knowledge base due to multiple factors such as changes in the workforce, new technologies and methodologies, economics, and regulations, the D&D KM-IT is being developed in a phased and modular fashion.
Decommissioning for the Containment Vessel of HTR-PM: Monte Carlo Simulation and Long-Time Predictions
Authors: Li Wenqian , Li Hong , Cao Jianzhu , Li Ming , Xie Feng , Fang Sheng
Published: February 2014

A high temperature gas cooled reactor-pebble bed module (HTR-PM) is currently under design and construction in China. Calculations of induced radioactivity in the concrete containment vessel of HTR-PM were performed in order to predict the decommissioning problems. The main activation products and the ambient dose equivalent rate corresponding to 40 years of operations followed by 30 days/1 year/10 years of decay, were calculated by using the Monte Carlo particle transport code FLUKA. Finally, according to the Chinese National Standard and on the basis of the knowledge gained from this study, the classifications of the radioactive solid wastes were discussed, and some suggestions for the working duration of the employees were given.
On Evaluation of Assessments of Accruals of Future Dismantling Costs
Authors: Labor Bea , Lindskog Staffan
Published: February 2014

A major prerequisite in order for civilian commercial nuclear energy production to qualify as sustainable energy production is that systems for the management of the nuclear waste legacy are in operation. These waste types are present in a range from very low short lived waste (VLLW) to long lived high level waste (HLW) (including the used nuclear fuel).

The second prerequisite is that financial responsibilities or other constraints must not be passed on to coming generations.

The first condition for qualification corresponds to the Polluters Pays Principle (PPP) which demands that the responsibility for the waste management rests solely with the polluter. The second qualification corresponds to the principle of fairness between generations and thus concerns the appropriate distribution of responsibilities between the generations.

It is important to note that these two conditions must be met simultaneously, and that compliance with both is a necessary prerequisite in order for commercial use of nuclear power to qualify as a semi-sustainable energy source.

Financial and technical planning for dismantling and decommissioning of nuclear installations cannot be regarded as successful unless it rests upon a distinctive way to describe and explain the well-founded values of different groups of stakeholders. This cumbersome task can be underpinned by transparent and easy to grasp models for calculation and estimation of future environmental liabilities. It essential that a systematic classification is done of all types of costs and that an effort is done to evaluate the precision level in the cost estimates. In this paper, a systematic and transparent way to develop a parametric approach that rest upon basic accounting standards is combined with data about younger stakeholder’s values towards decommissioning and dismantling of nuclear installation. The former entity rests upon theoretical and practical methods from business administration, whilst the latter is based on current survey data retrieved from 667 personal interviews in one town in Poland and one town in Slovakia with a near 100 % response rate.

The main conclusions from this field study may be summarised as follows:

• Sustainable energy sources are prioritised.

• Around one quarter of the respondents regards nuclear power as a future semi-sustainable commercial energy production mode subject to that the waste is managed in a sustainable, environmental friendly and safe way.

• The values are to a significant degree positioned on health, safety and environmental (HSE) attributes.

• The polluter pays principle is honoured.

• There are doubts regarding the compliance with these principles due to risks for delays in the implementation phase of repositories for disposal of the nuclear residues.

• 1/5th of the respondents expressed an openness to reprocessing (which is linked to the concept of “new nuclear power”).
Behaviour of IPG Waste Forms Bearing BaSO4 as the Dominant Sludge Constituent Generated From the Treatment of Water Used for Cooling the Stricken Power Reactors
Authors: Amamoto Ippei , Kobayashi Hidekazu , Yokozawa Takuma , Yamashita Teruo , Nagai Takayuki , Kitamura Naoto , Takebe Hiromichi , Mitamura Naoki , Tsuzuki Tatsuya
Published: February 2014

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi following the earthquake and tsunami of 11 March 2011 has resulted in accumulation of cooling water so-called the remaining water in some buildings.

As the cooling water is subsequently contaminated by fission products (FPs) and some other radioactive substances, it is necessary to decontaminate this ‘cooling water’ to reduce the volume of liquid radioactive waste and to reuse it again for cooling the affected reactors.

Some methods are applied to remove the radioactive substances from the cooling water. However, after treatments of water, there arises a secondary radioactive waste, the sludge. Steps are now taken to immobilize this sludge.

In this paper, BaSO4, as one of main constituents of the sludge, was chosen as an immobilizing target substance. The appropriate manufacturing condition of glass waste form for loading the sludge (BaSO4) was studied and the chemical durability was evaluated by measuring the dissolution rate. For this experiment, the iron phosphate glass (IPG), which is known to possess a large loading capacity for a variety of chemical substances, was employed as the glass medium.

Based on experimental results, it is evident that BaSO4 can be loaded into the IPG medium when it possesses the appropriate composition and melting temperature. During loading, BaSO4 converted into BaO, acting as a network modifier, which leads to enhanced stability of IPG.
Geological Criteria for Site Selection of an LILW Radioactive Waste Repository in the Philippines
Authors: Aurelio Mario , Taguibao Kristine Joy , Vargas Edmundo , Palattao Maria Visitacion , Reyes Rolando , Nohay Carl , Luna Roy Anthony , Singayan Alfonso
Published: February 2014

In the selection of sites for disposal facilities involving low- and intermediate-level radioactive waste (LILW), International Atomic Energy Agency (IAEA) recommendations require that “the region in which the site is located shall be such that significant tectonic and surface processes are not expected to occur with an intensity that would compromise the required isolation capability of the repository”. Evaluating the appropriateness of a site therefore requires a deep understanding of the geological and tectonic setting of the area. The Philippines sits in a tectonically active region frequented by earthquakes and volcanic activity. Its highly variable morphology coupled with its location along the typhoon corridor in the west Pacific region subjects the country to surface processes often manifested in the form of landslides.

The Philippine LILW near surface repository project site is located on the north eastern sector of the Island of Luzon in northern Philippines. This island is surrounded by active subduction trenches; to the east by the East Luzon Trough and to the west by the Manila Trench. The island is also traversed by several branches of the Philippine Fault System. The Philippine LILW repository project is located more than 100 km away from any of these major active fault systems. In the near field, the project site is located less than 10 km from a minor fault (Dummon River Fault) and more than 40 km away from a volcanic edifice (Mt. Caguas).

This paper presents an analysis of the potential hazards that these active tectonic features may pose to the project site. The assessment of such geologic hazards is imperative in the characterization of the site and a crucial input in the design and safety assessment of the repository.
Waste Handling and Storage in the Decontamination Pilot Projects of JAEA for Environments of Fukushima
Authors: Nakayama S. , Kawase K. , Iijima K. , Kato M.
Published: February 2014

After the Fukushima Daiichi nuclear accident, Japan Atomic Energy Agency (JAEA) was chosen by the national government to conduct decontamination pilot projects at selected sites in Fukushima prefecture. Despite tight boundary conditions in terms of timescale and resources, the projects served their primary purpose to develop a knowledge base to support more effective planning and implementation of stepwise regional remediation of the evacuated zone. A range of established, modified and newly developed techniques were tested under realistic field conditions and their performance characteristics were determined. The results of the project can be summarized in terms of site characterization, cleanup and waste management.

A range of options were investigated to reduce the volumes of waste produced and to ensure that decontamination water could be cleaned to the extent that it could be discharged to normal drainage. Resultant solid wastes were packaged in standard flexible containers, labelled and stored at the remediation site (temporary storage until central interim storage becomes available). The designs of such temporary storage facilities were tailored to available sites, but all designs included measures to ensure mechanical stability (e.g., filling void spaces between containers with sand, graded cover with soil) and prevent releases to groundwater (impermeable base and cap, gravity flow drainage including radiation monitors and catch tanks). Storage site monitoring was also needed to check that storage structures would not be perturbed by external events that could include typhoons, heavy snowfalls, freeze/thaw cycles and earthquakes.
The Spectrometric Determination of the Individual Exposure Rate for Gamma Nuclides From an Environmental Radiation Detector
Authors: Ji Young-Yong , Chung Kun Ho , Lee Wanno , Park Doo-Won , Kang Mun-Ja
Published: February 2014

For making the spectrometric determination of the exposure rate from the environment as well as the radioactive material more practical, an accurate calculation method of the individual exposure rate for the detected gamma nuclides from that spectrum should be suggested without the sophisticated calibration procedure. In this study, the calculation method for the individual exposure rate for detected gamma nuclides from a 3″×3″ NaI(Tl) detector was suggested by introducing the concept of the dose rate spectroscopy and the peak-to-total ratio in the energy spectrum for the exposure rate, which means just a form of multiplied counts and the value of a G-factor in the spectrum.
The Conceptual Solutions Concerning Decomissioning and Dismantling of Russian Civil Nuclear Powered Ships
Authors: Kulikov Konstantin N. , Nizamutdinov Rinat A. , Abramov Andrey N.
Published: February 2014

From 1959 up to 1991 nine civil nuclear powered ships were built in Russia: eight ice-breakers and one lash lighter carrier (cargo ship). At the present time three of them were taking out of service: ice-breaker “Lenin” is decommissioned as a museum and is set for storage in the port of Murmansk, nuclear ice-breakers “Arktika” and “Sibir” are berthing. The ice-breakers carrying radwastes appear to be a possible source of radiation contamination of Murmansk region and Kola Bay because the ship long-term storage afloat has the negative effect on hull’s structures. As the result of this under the auspices of the Federal Targeted Program “Nuclear and Radiation Safety of Russia for 2008 and the period until 2015” the conception and projects of decommissioning of nuclear-powered ships are developed by the State corporation Rosatom with the involvement of companies of United Shipbuilding Corporation. In developing the principal provisions of conception of decommissioning and dismantling of ice-breakers the technical and economic assessment of dismantling options in shiprepairing enterprises of North-West of Russia was performed.

The paper contains description of options, research procedure, analysis of options of decommissioning and dismantling of nuclear ice-breakers, taking into account the principle of optimization of potential radioactive effect to personnel, human population and environment. The report’s conclusions contain the recommendations for selection of option for development of nuclear ice-breaker decommissioning and dismantling projects.
Efficient and Compact Mobile Equipment Based on the New RADEON-NWM Technology to Process Liquid Radioactive Wastes Resulted From the Accidents of the Nuclear Installations
Authors: Martoyan Gagik , Nalbandyan Garik , Gagiyan Lavrenti , Karamyan Gagik , Brutyan Gagik
Published: February 2014

During the operation of nuclear reactors important volume of liquid and solid radioactive wastes are generated, which, in normal conditions, becomes processed by stationary equipment by different methods to minimize their volume and then sent to specially constructed storages. The cases of accidents of Chernobyl and Fukushima showed that the localization of rejected big quantity of radioactive wastes is a prior problem for their further processing by stationary equipment. In this regard it is very important the processing of radioactive wastes on the contaminated areas to localize them by mobile equipment based on the efficient technologies. RADEON-NWM new technology allows resolving this problem. This technology is compact, completely automated, which makes possible to assemble it on a standard 40-ft by 7-ft trailer driven by heavy-duty truck. The new technology is fully elaborated, the necessary tests are conducted.
Calculation of the Dose Load During Dismantling of Large Components in the Process of Decommissioning of Nuclear Installations
Authors: Hornacek Martin , Necas Vladimir , Bezak Peter
Published: February 2014

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters — decay time and decontamination. The obtained results — external exposure of workers and the influence of time and pre-dismantling decontamination — are studied. Also detailed analyses of every dismantling step considered are presented.
Preliminary Waste Management Plan of European Spallation Source
Authors: Ene Daniela
Published: February 2014

The European Spallation Source (ESS) is the European common effort in designing and building a next generation large-scale user facility for studies of the structure and dynamics of materials. The proposed schematic layout of the ESS facility is based on a linear driver (linac) directing the proton beam (5 MW of 2.5 GeV) of 2.8 ms long pulses with a 20 Hz on a tungsten target where neutrons are produced via spallation reactions. Further the neutrons will be moderated to thermal and subthermal energies in a couple of moderators placed around the target. The moderators feed 22 beamlines guiding the neutrons to the scattering instruments, mainly for neutron scattering research, as has been previously mentioned.

The objective of this work is to develop a waste management plan for ESS facility. In this respect two important aspects are analyzed. First the present status of the problem is outlined as follow. Estimate types and quantities of waste that the ESS project will generate at different stages: commission, operation, decommissioning were derived using: i) precise Monte Carlo calculations ii) scaling the activity from the operation experience of the existing spallation source installations for waste such it is difficult to predict level of activation or for components of the facility in stage of the pre-conceptual model.

Associated waste treatment/conditioning options and final disposal route were further analyzed in order to define the waste type and packet descriptions in agreement with Swedish regulations and policy. It was found that the compilation of completely new waste type descriptions for qualification of the ESS waste for disposal will be necessary. Particular attention was devoted to “problematic waste” as Beryllium reflector, C-14 from graphite used as core zone of the beam-dump and collimators or waste arising from the purification systems of both Helium and water cooling circuits. Management of waste on ESS site: collection/segregation systems, characterization system, storage options, is also described.

In the second step, the acquired information is used for planning and implementing actions involving all participants (ESS, treatment facility operator, disposal operator, regulatory body and other authorized authorities).
Ultratrace Determination of Strontium-90 in Environmental Soil Samples From Qatar by Collision/Reaction Cell-Inductively Coupled Plasma Mass Spectrometry (CRC-ICP-MS/MS)
Authors: Al-Meer S. H. , Amr M. A. , Helal A. I. , Al-Kinani A. T.
Published: February 2014

Because of the very low level of 90Sr in the environmental soil samples and its determination by beta counting may take several weeks, we developed a procedure for ultratrace determination of 90Sr using collision reaction cell-inductively coupled plasma tandem mass spectrometry (CRC-ICP-MS/MS, Agilent 8800).

Soil samples were dried at 105 °C and then heated in a furnace at 550 °C to remove any organics present. 500 g of each soil samples were aliquoted into 2000 ml glass beakers. Each Soils samples were soaked in 2 ppm Sr solution carrier to allow determination of chemical yield. The solid to liquid ratio was 1:1. Finally the soil samples were dried at 105 °C. Five hundred milliliters concentrated nitric acid and 250 ml hydrochloric acid volumes were added on 500 g soil samples. The samples were digested on hot plate at 80 °C to prevent spraying with continuous manual mixing. The leachate solution was separated. The solids were rinsed with 500 ml deionized water, warmed on a hot plate and the leachate plus previous leachate were filtered and the total volume was reduced to 500 ml by evaporation. Final leachate volume was transferred to a centrifuge tubes. The centrifuge tubes were centrifuged at 3,500 rpm for 10 min. The leachate was transferred to a 1 L beaker and heated on a hot plate to evaporate the leachate to dryness. The reside was re-dissolved in 100 ml of 2% HNO3 and reduced by evaporation to 10 mL. The solution was measured directly by CRC-ICP-MS/MS by setting the first quadruple analyzer to m/z 90 and introducing oxygen gas into the reaction cell for elimination isobar interference from zirconium-90. The method was validated by measurements of standard reference materials and applied on environmental soil samples. The overall time requirement for the measurement of strontium-90 by CRC-ICP-MS/MS is 2 days, significantly shorter than any radioanalytical protocol currently available.
Treatment of Irradiated Graphite From French Bugey Reactor
Authors: Stevens Howard , Laurent Gerard
Published: February 2014

In 2008, following the general French plan for nuclear waste management, Électricité de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage.

In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as THermal Organic Reduction, “THOR” (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary waste to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite.

In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, ‘Phase I’ of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, ‘Phase II’ testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite.

Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal volumes.
Processing of Solid Low Level Alpha Suspected Radioactive Waste
Authors: Maes Michael , Huys Thomas , Rommes Jeroen , Tuerlinckx Robin , Bloemen Raf
Published: February 2014

Incineration of low active waste is the most efficient way to obtain high volume reduction factors. Although the initial scenario for processing low active alpha suspected waste (α-activity ≤ 400MBq/m3) in Belgium was supercompaction, incineration was eventually chosen for various benefits concerning volume reduction as well as the elimination of organic compounds improving the long term performance. This specific type of waste is often contaminated with for example mercury, requiring special attention in the treatment of both off-gas en waste water. In this paper it is shown that the CILVA incinerator of Belgoprocess N.V. has a high efficiency to combine both incineration and supercompaction of such types of waste.
The Long-Term Safety and Performance Analyses of the Surface Disposal Facility for the Belgian Category A Waste at Dessel
Authors: Cool Wim , Vermariën Elise , Wacquier William , Perko Janez
Published: February 2014

ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, and its partners have developed long-term safety and performance analyses in the framework of the license application for a surface disposal facility for low level radioactive waste (category A waste) at Dessel, Belgium.

This paper focusses on the methodology of the safety assessments and on key results from the application of this methodology.

An overview is given (1) of the performance analyses for the containment safety function of the disposal system and (2) of the radiological impact analyses confirming that radiological impacts are below applicable reference values and constraints and leading to radiological criteria for the waste and the facility. In this discussion, multiple indicators for performance and safety are used to illustrate the multi-faceted nature of long-term performance and safety of the surface disposal.

This contributes to the multiple lines of reasoning for confidence building that a positive decision to proceed to the next stage of construction is justified.
Know-How and Experience of the Decommissioning of the Eurochemic Reprocessing Plant
Authors: Notelteirs W. , Ooms B. , Huys T.
Published: February 2014

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions.

During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements.

• The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects.

• The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris.

• The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris.

As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now.
The Design Methodology and the Safety Concept of the Surface Disposal Facility for the Belgian Category A Waste at Dessel
Authors: Wacquier William , Bastiaens Wim , Cool Wim , Van Cotthem Alain
Published: February 2014

The paper describes the design methodology that ONDRAF/NIRAS, the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, has developed and applied in the framework of the surface disposal facility for category A waste, i.e. solid conditioned low level radioactive waste that contains a limited amount of long lived radionuclides, at Dessel. The proposed disposal facility supporting the submitted construction and operation license application [1, 2] is also presented.
Boom Clay Borehole Water, Home of a Diverse Bacterial Community
Authors: Wouters Katinka , Moors Hugo , Leys Natalie
Published: February 2014

For over two decades, Boom Clay has been studied in the framework of geological disposal of nuclear waste thereby mainly addressing its geochemical properties. Today, also the microbiological properties and the possibility of microbes interacting with radionuclides or repository components including the waste form, in a host formation like Boom Clay are considered [2,3]. In the past, a reference composition for synthetic Boom Clay pore water (BCPW) was derived, based on interstitial water sampled from different layers within the Boom clay [1]. Similarly, the primary aim of this microbiological study was to determine the core BCPW bacterial community and identify representative water samples for future microbial directed lab experiments. In this respect, BCPW was sampled from different Boom Clay layers using the Morpheus piezometer (Fig. 1) and subsequently analysed by microscopy and molecular techniques, in search for overall shared and abundant micro-organisms.
Influence of Cracks in Cementitious Engineered Barriers in a Near-Surface Disposal System: Assessment Analysis of the Belgian Case
Authors: Perko Janez , Seetharam Suresh C. , Jacques Diederik , Mallants Dirk , Cool Wim , Vermariën Elise
Published: February 2014

In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids.
Characterisation and Evaluation of Wastes for Treatment in the Batch Pyrolysis Plant in Studsvik, Sweden
Authors: Lindberg Maria , Österberg Carl , Vernersson Thomas
Published: February 2014

The new batch pyrolysis plant in Studsvik is built primarily for treatment of uranium containing dry active waste, “DAW”. Several other waste types have been identified that are considered or assumed suitable for treatment in the pyrolysis plant because of the possibility to carefully control the atmosphere and temperature of the thermal treatment. These waste types must be characterised and an evaluation must be made with a BAT (Best Available Technology) perspective. Studsvik have performed or plan to perform lab scale pyrolysis tests on a number of different waste types. These include

- Pyrophoric materials (uranium shavings)

- Uranium chemicals that must be oxidised prior to being deposited in repository

- Sludges and oil soaks (this category includes NORM-materials)

- Ion exchange resins (both “free” and solidified/stabilised)

- Bitumen solidified waste

Methodology and assessment criteria for various waste types, together with results obtained for the lab scale tests that have been performed, are described.
Remediation of Uranium In-Situ Leaching Area at Stráž pod Ralskem, Czech Republic
Authors: Vokál Vojtěch , Mužák Jiří , Ekert Vladimír
Published: February 2014

A large-scale development in exploration and production of uranium ores in the Czech Republic was done in the 2nd half of the 20th century. Many uranium deposits were discovered in the territory of the Czech Republic. One of the most considerable deposits in the Czech Republic is the site Hamr na Jezere – Straz pod Ralskem where both mining methods — the underground mining and the acidic in-situ leaching — were used. The extensive production of uranium led to widespread environmental impacts and contamination of ground waters. Over the period of “chemical” leaching of uranium (ca. 32 years), a total of more than 4 million tons of sulphuric acid and other chemicals have been injected into the ground. Most of the products (approx. 99.5 %) of the acids reactions with the rocks are located in the Cenomanian aquifer. The contamination of Cenomanian aquifer covers the area larger then 27 km2. The influenced volume of groundwater is more than 380 million m3. The total amount of dissolved SO42− is about 3.6 million tons. After 1990 a large-scale environmental program was established and the Czech government decided to liquidate the ISL Mine and start the remediation in 1996.

The remediation consists of contaminated groundwater pumping, removing of the contaminants and discharging or reinjection of treated water. Nowadays four main remedial technological installations with sufficient capacity for reaching of the target values of remedial parameters in 2037 are used — the “Station for Acid Solutions Liquidation No. One”, the “Mother liquor reprocessing” station, the “Neutralization and Decontamination Station NDS 6” and the “Neutralization and Decontamination Station NDS 10”. It is expected that the amount of withdrawn contaminants will vary from 80 000 to 120 000 tons per year. Total costs of all remediation activities are expected to be in excess of 2 billion EUR.
Application of PHADEC Method for the Decontamination of Radioactive Steam Piping Components
Authors: Lo Frano R. , Pilo F. , Aquaro D.
Published: February 2014

The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue.

In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc.

The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing.

Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters.

This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line.

Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination.
Destructive Testings: Dry Drilling Operations With TruPro System to Collect Samples in a Powder Form, From Two Hulls Containing Immobilized Wastes in a Hydraulic Binder
Authors: Pombet Denis , Desnoyers Yvon , Charters Grant , Aggarwal Sue
Published: February 2014

The TruPro® process enables to collect a significant number of samples to characterize radiological materials. This innovative and alternative technique is experimented for the ANDRA quality-control inspection of cemented packages. It proves to be quicker and more prolific than the current methodology.

Using classical statistics and geostatistics approaches, the physical and radiological characteristics of two hulls containing immobilized wastes (sludges or concentrates) in a hydraulic binder are assessed in this paper. The waste homogeneity is also evaluated in comparison to ANDRA criterion. Sensibility to sample size (support effect), presence of extreme values, acceptable deviation rate and minimum number of data are discussed.

The final objectives are to check the homogeneity of the two characterized radwaste packages and also to validate and reinforce this alternative characterization methodology.
Preparation Methods of Copper-Ferrocyanide Functionalized Magnetic Nanoparticles for Selective Removal of Cesium in Aqueous Solution
Authors: Yang Hee-Man , Lee Kune Woo , Seo Bum-Kyoung , Moon Jei Kwon
Published: February 2014

Copper ferrocyanide functionalized magnetite nanoparticles (Cu-FC-MNPs) were successfully synthesized by the immobilization of copper and ferrocyanide on the surface of [1-(2 amino-ethyl)-3-aminopropyl] trimethoxysilane modified magnetite nanoparticles. A radioactive cesium (Cs) adsorption test was carried out to investigate the effectiveness of Cu-FC-MNPS for the removal of radioactive Cs. Furthermore, the Cu-FC-MNPs showed excellent separation ability by an external magnet in an aqueous solution.
Evaluating Performance of High Efficiency Mist Eliminators
Authors: Waggoner Charles A. , Parsons Michael S. , Giffin Paxton K.
Published: February 2014

Processing liquid wastes frequently generates off gas streams with high humidity and liquid aerosols. Droplet laden air streams can be produced from tank mixing or sparging and processes such as reforming or evaporative volume reduction. Unfortunately these wet air streams represent a genuine threat to HEPA filters. High efficiency mist eliminators (HEME) are one option for removal of liquid aerosols with high dissolved or suspended solids content.

HEMEs have been used extensively in industrial applications, however they have not seen widespread use in the nuclear industry. Filtering efficiency data along with loading curves are not readily available for these units and data that exist are not easily translated to operational parameters in liquid waste treatment plants.

A specialized test stand has been developed to evaluate the performance of HEME elements under use conditions of a US DOE facility. HEME elements were tested at three volumetric flow rates using aerosols produced from an iron-rich waste surrogate. The challenge aerosol included submicron particles produced from Laskin nozzles and super micron particles produced from a hollow cone spray nozzle. Test conditions included ambient temperature and relative humidities greater than 95%.

Data collected during testing HEME elements from three different manufacturers included volumetric flow rate, differential temperature across the filter housing, downstream relative humidity, and differential pressure (dP) across the filter element. Filter challenge was discontinued at three intermediate dPs and the filter to allow determining filter efficiency using dioctyl phthalate and then with dry surrogate aerosols. Filtering efficiencies of the clean HEME, the clean HEME loaded with water, and the HEME at maximum dP were also collected using the two test aerosols.

Results of the testing included differential pressure vs. time loading curves for the nine elements tested along with the mass of moisture and solid material on each element at final dP. Plots of overall filtering efficiencies for DOP (spherical aerosol) and dry surrogate (aspherical aerosols) at specified dPs were computed for each filter. Filtering efficiencies were determined as a function of particle size. Curves were also generated showing the most penetrating particle size as a function of dP.

A preliminary set of tests was conducted to evaluate spray location, duration, pressure, and wash volume for in-place cleaning the interior surface (reducing dP) of the HEME element. A variety of nozzle designs were evaluated and test results demonstrated the potential to overload the HEME (saturate filter medium) resulting in very high dPs and extensive drain times. At least one combination of spray nozzle design, spray location on the surface of the element, and spray time/pressure was successful in achieving extension of operational life.