bluuebunny/crossref_metadata_2025 huggingface dataset:
Managing the Removal of Radioactive Materials Found in Public Locations
Authors: Lewcock Andrew , Grundy Colette , Shaw Catherine , Copeland Paddy , Jackson Duncan
Published: December 2010

Managing the removal of Radioactive Materials found in public locations. In January 2006 the Environment Agency for England and Wales requested assistance in preparing a project to plan, collect and safely dispose of radium painted aluminium aircraft hatches, discovered in 3 separate business premises in the UK. These World-War Two aircraft hatches had been marked with radium “luminous” paint, to guide crew in the dark to the escape exits if they needed to bale out. The hatches had been stored since the early 1960s in one location, with some of the inventory being moved two other locations in 2003/2004. The North West Region of the Environment Agency appointed a consortium of Enviros, Safeguard International and Aurora Health Physics to undertake the work, and they funded it from the sealed sources disposal budget. The paper will set out how the project was complicated by an assortment of “real world” problems; preliminary estimates of both the activity per hatch and the number of hatches established the potential for a significant quantity of radium to be disposed of safely. The total number of hatches was not known for sure at the start of the work. Access to retrieve the majority of the frames was hazardous due to the poor structural condition of the building roof. Other difficulties included constrained access under a railway line, and bird-related biohazards. The sites involved in the collections were not intended to house radioactive materials, so physical security was another important issue. Some of the hatches were known to be in poor condition, with a very high probability of radium contamination being spread to the surrounding areas. The hatches had to be removed from the sites before the full extent of contamination of other materials could be established. As it was difficult to identify a disposal route in the UK at the time for the estimated inventory, a novel metal recycling option, using a facility in the USA, was proposed as a solution. This was a new approach for dealing with such radioactive materials in the UK. The hatches were successfully recovered, sectioned appropriately and packaged for transport by road and air to the USA for metal recycling in June 2007. Subsequently approximately 0.75 tonnes of other contaminated materials were finally removed from the premises later in the year for authorised waste disposal in the UK.
Remediation of the Site of a Former Active Handling Building in the United Kingdom
Authors: Armitage J. , Cornell R. M. , Staples A.
Published: December 2010

In 2000, Nuvia Limited was contracted to carry out the decommissioning of a former Active Handling Building A59 on the United Kingdom Atomic Energy Authority (UKAEA) site at Winfrith in the UK. This is in support of UKAEA’s mission, which is to carry out environmental restoration of its nuclear sites and to put them to alternative uses wherever possible. Recently UKAEA has been reorganised and responsibility for the site lies with Research Sites Restoration Limited (RSRL) with funding provided by the National Decommissioning Authority (NDA). Following major decommissioning operations the main containment building structure and the two suites of concrete shielded caves were demolished between June 2006 and March 2007 leaving just the base slab for final removal and the site remediation operations undertaken. The base slab contained a quantity of encast, internally contaminated items including more than 100 steel mortuary tubes set up to 6.6m deep into the slab. At the outset it was suspected that some leakage of radioactive contamination had occurred into the ground although the precise location/s of the leakage was unknown. As a result the scope of the work required the underlying soil to be carefully monitored for the presence of radioactive contamination and, if detected, its remediation to an end state suitable for unrestricted use without planning or nuclear regulatory controls. These latter operations form the basis of this paper, which reviews some of the significant tasks undertaken during the process and describes the waste monitoring procedures utilised on the concrete and soil debris. Extensive dewatering was required to support the removal of the deeper mortuary tubes and the impact this had upon the operations and associated excavations will be described. Further, the demolition of an external active effluent tank and excavation and monitoring of the surrounding soils due to the presence of significant local contamination will be a key feature of the paper. A number of significant problems that were encountered during the operations will also be identified with a narrative about how these arose and were subsequently overcome. The use of Nuvia’s Groundhog™ system, a gamma radiation ground surveying and global positioning system, together with a well defined sampling grid enabled the footprint of the base slab to be surveyed and subsequently remediated to an agreed standard by the end of 2008 to allow infilling with non-calcareous soil ahead of final landscaping as the last step for completion of the project. One area of particular significance to the remediation process has been the use of office-based contaminated land assessment tools including ReCLAIM, a Microsoft Excel spreadsheet based tool used to assess current and future impacts of radiological contamination at nuclear licensed sites. This tool is particularly recommended to others working on similar projects.
Operational Experience With a Commercial Plant for Stabilisation of Radioactive Sludge and Other Materials in the United Kingdom
Authors: Hagan M. , Cornell R. M. , Riley B. , Ware B.
Published: December 2010

In 2000, Nuvia Limited was contracted to design, build and commission a waste treatment plant (WETP) to stabilise the active sludge stored in the External Active Storage Tanks (EAST) at UKAEA Winfrith, UK. The sludge was generated during the operational period of the prototype Steam Generating Heavy Water Reactor (SGHWR), which is now in the process of being decommissioned. This work supports UKAEA’s mission, which is to carry out environmental restoration of its nuclear sites and to put them to alternative uses wherever possible. Recently UKAEA has been reorganised and responsibility for the site lies with Research Sites Restoration Limited (RSRL) with funding provided by the Nuclear Decommissioning Authority (NDA). The process of stabilisation of the SGHWR sludge from the EAST tanks within 500 litre stainless steel drums in the Winfrith EAST Treatment Plant (WETP) using ordinary Portland cement (OPC) and blast furnace slag (BFS) is now almost complete. At this stage it was planned to decommission and demolish the WETP facilities but RSRL have introduced a further stabilisation project involving thorium metal waste ahead of the start of the planned decommissioning. As a result, the facilities are to be revised to provide for the encapsulation of bars of thorium metal within modified 500 litre drums together with a number of necessary changes to the plant control system. The cell line used for stabilisation consists of five operational cells separated by shield doors designed to maintain strict contamination control. There is a wet cell where the drums are filled with sludge and powder, a cell with stations for curing and grouting the drums, a cell for lidding, bolting and QA inspection, a maintenance and gamma monitoring cell and a buffer store to hold the completed drums. After completion, drums are moved in a shielded overpack to the Treated Radwaste Store located on a different part of the Winfrith site. In the revised cell line configuration the wet cell will be completely isolated and not used and the drums, after introduction, are to be taken to the original re-work cell located to one side halfway along its length. The modification to this cell to make it suitable for handling the thorium bars will be described together with changes to the drum design to centralize the bars ahead of their stabilisation under revised grouting arrangements using OPC and Pulverised Fly Ash (PFA). The final aspects of the paper will briefly describe the approach that is to be adopted for the final decontamination and demolition of both the EAST facility and WETP plant and associated buildings.
Development of the ENVI Simulator to Estimate Korean SNF Flow and Its Cost
Authors: Hwang Yongsoo , Miller Ian
Published: December 2010

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for managing spent nuclear fuel (SNF) in South Korea. A companion paper (Hwang and Miller, 2009) describes a performance assessment model to address the long-term safety of alternative geological disposal options for different waste streams. The model addresses alternative concepts for storage, transportation, and processing of SNF of different types (CANDU, PWR), leading up to permanent disposal in geological repositories. It uses the GoldSim software to simulate the logistics of the associated activities, including the associated capital and operating costs. The model’s results allow direct comparison of alternative waste management concepts, and predict the sizes and timings of different facilities required. Future versions of the model will also address the uncertainties associated with the different system components in order to provide risk-based assessments.
Statistical Modelling Applied to the Contents of Waste Drums
Authors: Bull Richard K. , Pearman Ian
Published: December 2010

Gamma spectrometry is widely used to determine the radioactive content of waste drums. However, the results of such surveys often result in large numbers of limit-of-detection (LOD) results. In this paper we will show how simple statistical methods can be used to obtain useful information on the average drum activities, even in these unfavourable circumstances. Results from measurements on 60Co, 152Eu, 154Eu activities in drums of waste from decommissioning of the GLEEP reactor suggest that these activities are lognormally distributed with geometric standard deviations (GSD) ranging from 3–4. This statistical model can be used to extract information from 235U, 234mPa and 234Th datasets which show only LOD activity results. In a repository of N drums, each with activity <L, the total activity is clearly <NL. However, we can use the lognormal model to make a much stronger statement about the total waste activity. This model is developed quantitatively in the paper.
Integrated Model of Korean Spent Fuel and High Level Waste Disposal Options
Authors: Hwang Yongsoo , Miller Ian
Published: December 2010

This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21’st century. The model addresses alternative design concepts for disposal of SNF of different types (CANDU, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model’s results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses.
Improvement of the Management of Institutional Radioactive Waste in Slovenia
Authors: Niels Yves , Fabjan Marija , Tavcˇar Gasˇper , Rojc Jozˇe , Stepisˇnik Matjazˇ , Lenie Koen
Published: December 2010

The Central Storage Facility (CSF) in Brinje is the only storage facility for institutional radioactive waste in Slovenia. The storage has been in operation since 1986. Since the year 1999, operation of the CSF in Brinje and managing of institutional radioactive waste in Slovenia has been under the control of Agency for Radwaste Management (ARAO). At the time of taking over the CSF, the waste in store was not fully characterized and in some cases the available data did not match records and inventories. Besides this, some shielded containers and drums were degraded, which creates a potential risk of uncontrolled spread of contamination. In addition, in 1999 the Slovene Nuclear Safety Administration (SNSA) requested the ARAO to perform refurbishing works in the CSF in order to reinforce and tighten the building on the one hand, and characterize and condition radioactive waste in store on the other. In order to improve the existing situation ARAO lunched considerable assistance and know-how transfer through training and other technical cooperation within the IAEA and the EC projects. In this context, several projects have been carried out in the period between 1999 and 2007. However, these projects only addresses one third of the total inventory of radioactive waste. In particular, the radioactive waste in the form of bulky material, which occupies a significant surface of the CSF, will not be processed. Therefore, the implementation of new project in the frame of EC Transition Facility programme entitled “Improvement of the Management of Institutional Radioactive Waste in Slovenia” has been carried out in December 2007. Through a public invitation for tenders, an international Consortium consists of Belgian and Slovene radwaste experts gained the contract. The project was executed in ten months and it was divided into three tasks of treatment and conditioning according to the type of packaging: - waste packed in drums, - contaminated or activated bulky items and - spent sealed radioactive sources. The implementation of the project enabled the ARAO to meet the requirements for improving the institutional radioactive management and the storage facility itself. It also improved knowledge of the ARAO’s staff in treatment and conditioning technologies for the future management of the CSF. The results of the project and operating experience regarding the waste management in the CSF which were obtained through this project are presented in this paper.
Improvement of Institutional Radioactive Waste Management via Implementation of Environmental Management System
Authors: Fabjan Marija , Hertl Bojan , Kralj Metka , Rojc Jozˇe
Published: December 2010

Management of institutional radioactive waste in Slovenia is one of the main processes in the management system of Agency for Radwaste Management (ARAO). The management system integrates the quality management system ISO 9001 and the environmental management system that was certified according to the standard ISO 14001:2004 in October 2007. The certificate represents a permanent commitment of ARAO to improve the environmental management system and implement environmental planning. Environmental planning includes and maintains environmental objectives and targets in all ARAO’s activities, especially in managing the institutional radioactive waste that can have potential environmental impacts and is considered as a risky activity from the public view point. We defined a general ARAO’s register of environmental aspects with seven main groups of aspects. Each aspect was evaluated according to its impacts with the help of multiple criteria and only the aspects evaluated as significant are dealt with in further environmental planning. Environmental planning was implemented in the process of managing of institutional radioactive waste. Our methodology for evaluating the significance of environmental aspects showed three significant aspects. Appropriate environmental objectives and targets were defined for these aspects in a way that they were measurable, if applicable, consistent with the ARAO’s environmental policy. It was assured that they provided adequate environmental protection measures and protection of human health, compliance with national legal requirements and international directives dealing with radioactive waste management and that the production of radioactive waste was minimized. The environmental planning is carried out annually, and each year we are looking for opportunities to minimize potential environmental impacts of our service. In this paper, ARAO’s experience with environmental planning and the realization of the objectives and targets that have already been achieved, implemented and integrated in the process of managing of institutional radioactive waste are presented. The results and the successful completion of environmental programs contribute to optimization of the management system and the quality of ARAO’s process performance. Implementation of environmental management system according to ISO 14001 increased the satisfaction of producers of institutional radioactive waste management that come from medicine, research and industry. It is also part of confidence building by our stakeholders.
UK Surplus Source Disposal Programme
Authors: John Gordon H. , Reeves Nigel , Nisbet Amy C. , Williams Clive R. , Garnet Andrew
Published: December 2010

The UK Surplus Source Disposal Programme (SSDP), managed by the Environment Agency, was designed to remove redundant radioactive sources from the public domain. The UK Government Department for Environment, Food and Rural Affairs (Defra) was concerned that disused sources were being retained by hospitals, universities and businesses, posing a risk to public health and the environment. AMEC provided a range of technical and administrative services to support the SSDP. A questionnaire was issued to registered source holders and the submitted returns compiled to assess the scale of the project. A member of AMEC staff was seconded to the Environment Agency to provide technical support and liaise directly with source holders during funding applications, which would cover disposal costs. Funding for disposal of different sources was partially based on a sliding scale of risk as determined by the IAEA hazard categorisation system. This funding was also sector dependent. The SSDP was subsequently expanded to include the disposal of luminised aircraft instruments from aviation museums across the UK. These museums often hold significant radiological inventories, with many items being unused and in a poor state of repair. These instruments were fully characterised on site by assessing surface dose rate, dimensions, source integrity and potential contamination issues. Calculations using the Microshield computer code allowed gamma radiation measurements to be converted into total activity estimates for each source. More than 11,000 sources were disposed of under the programme from across the medical, industrial, museum and academic sectors. The total activity disposed of was more than 8.5E+14 Bq, and the project was delivered under budget.
Gunnar Uranium Mine Environmental Remediation: Northern Saskatchewan
Authors: Muldoon Joe , Schramm Laurier L.
Published: December 2010

Thirty-six now-abandoned uranium mine and mill sites were developed and operated in Northern Saskatchewan, Canada, from approximately 1957 through 1964. During their operating lifetimes these mines produced large quantities of ore and tailings. The Gunnar Mine is located on the shores of Lake Athabasca, the 22nd largest lake in the world. The Gunnar mine (open pit and underground) produced over 5 million tonnes of uranium ore and nearly 4.4 million tonnes of mine tailings. There is an estimated 2,710,700 m3 of waste rock that abuts the shores of Lake Athabasca. After closure in the 1960’s, the Gunnar site along with all of the other uranium mine and mill sites were abandoned with little remediation and no reclamation being done. The governments of Canada and Saskatchewan are now funding the clean-up of these abandoned northern uranium mine and mill sites and have contracted the management of the project to the Saskatchewan Research Council. The clean-up activity is expected to take about 8 years, followed by 10–15 years of monitoring activity before the sites are to be released into an institutional controls program that will allow government oversight of a long term management and monitoring program. The Gunnar site, because of the magnitude of tailings and waste rock, is subject to an environmental site assessment process regulated by both provincial and federal governments. This process requires a detailed study of the projected environmental impacts resulting from the mining activities and an analysis of projected impacts from remediation efforts. Prescribed environmental and land use endpoints will be made based on the environmental assessment studies and remediation options analyzed and implemented based on expected results. Remediation options range from deep lake disposal of tailings to disposal of tailings in the open pit which is now filled with water and fish (contaminated, but which are reproducing successfully) to covering the tailings with a cap. The site also includes many buildings that are remnants of a community of approximately 800 people who once occupied the site. These buildings, many of which contain asbestos, must be appropriately removed and disposed of. The original mine head frame and mill site buildings, many of which still contain the original machinery and equipment, must also be removed and disposed of. The regulatory requirements include the environmental assessment processes, a complex public involvement strategy and licensing from the Canadian Nuclear Safety Commission. The environmental assessment process, specific site studies and public involvement initiatives are underway with the long term goal of releasing the property in a fully remediated state.
US EPA’s Experiences Implementing Environmental Safety Standards at the Waste Isolation Pilot Plant
Authors: Peake R. Thomas , Byrum Charles , Feltcorn Ed , Lee Raymond , Joglekar Rajani , Ghose Shankar , Eagle Mike
Published: December 2010

The U.S. Environmental Protection Agency (EPA or the Agency) developed environmental standards for the disposal of defense-related transuranic wastes for the U.S. Department of Energy’s (DOE or the Department) Waste Isolation Pilot Plant (WIPP). EPA implements these standards for WIPP, which has been in operation for over ten years. The general environmental standards are set forth in the Agency’s 40 CFR Part 191 Environmental Radiation Protection Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes [1]. These standards are implemented by site-specific compliance criteria [2]. The WIPP Land Withdrawal Act requires DOE to submit a re-certification application every five years after the initial receipt of waste. DOE submitted the latest WIPP re-certification application in March 2009. For re-certification, DOE must identify changes that have occurred over the previous five years and analyze their impact on the potential long-term performance of the repository. Once EPA determines that the re-certification application is complete, the Agency has six months to review the application and make a final decision. During this review, EPA solicits and incorporates public comment where appropriate. During the first re-certification in 2004, several stakeholder groups brought up issues (e.g., karst) that were addressed in the original certification. EPA has received comments again raising some of these same issues for the 2009 re-certification. In addition, DOE must submit proposed changes to the WIPP repository to EPA for review and approval. This paper describes selected issues of concern to WIPP and highlights interactions between EPA as the regulatory authority and DOE as the implementing organization. In general EPA’s experience points out the importance of communication, documentation and the regulator’s responsibility in determining “how much is enough.”
Plasma Arc Cutting Experiments Using Radioactive Materials for Evaluation of Airborne Dispersion Ratio
Authors: Shimada Taro , Takamura Atsushi , Kamiya Atsushi , Sukegawa Takenori , Tanaka Tadao
Published: December 2010

Experiments for airborne dispersion ratio of radionuclides during plasma arc cutting were carried out in a contamination control enclosure, using stored radioactive metal wastes arising from the decommissioning activities of Japan Power Demonstration Reactor, which was a boiling water type reactor. Neutron induced-activated piping and surface contaminated piping were segmented into pieces using air plasma arc cutting, using a current power was 100A. In addition, similar experiments for contaminated piping of the Advanced Thermal Reactor, Fugen were carried out. As a result, dispersion ratios for activated piping were 0.2 to 0.7% of Co-60 and 0.4% of Ni-63 under the condition with a covered cap on the head. And those for surface contaminated piping were from 18 to 23%. In addition, those for vertically segmented piping which simulated flat plate were from 34 to 43%. There was no difference of dispersion ratios between stainless steel and carbon steel base materials. All values obtained were smaller than the Handbook recommended value of 70% for contaminated materials. Filtering collection efficiencies of the coarse dust filter were approximately 40% for activated piping and approximately 55 to 80% for surface contaminated piping. However there was no effect for collection of aerosols smaller than 1 μm. Size distribution analysis indicated a greater concentration of radionuclides in particles smaller than 0.1μm when compared with larger particles. In addition, there was a tendency that the Ni-63 was concentrated to the particles smaller than 0.3 μm compared with the Co-60. The results support data obtained in the previous studies using non-radioactive materials.
Measurement of Radioactive Aerosol Behavior During Dismantling and Reflection to the Exposure Dose Evaluation
Authors: Iguchi Yukihiro , Kato Masami
Published: December 2010

Radioactive aerosol disperses slightly via contamination prevention systems such as control enclosures and filters when the nuclear installation is dismantled, and it might impact the environment. Therefore, when decommissioning is planned, it is necessary to assess the safety such as exposure dose evaluation to the public. For the radioactive aerosol, it is possible that the dispersion ratio is different according to the contamination condition, the dismantlement method of the material, nuclides (elements), etc. The radiation exposure evaluation for the decommissioning plan has been executed by operators in Japan based on a number of experiments (mostly cold tests) and overseas results. The decommissioning is now being carried out at the Tokai Power Station (GCR) and Fugen Decommissioning Engineering Center in Japan. In this study, the results data is acquired at the decommissioning sites, and the methodology and data for the exposure dose evaluation are verified and confirmed. These examination results will lead to the upgrading and improvement of the exposure evaluation methodology. In particular, the dismantlement work of connected piping of the heat exchanger (steam generator) was executed in the Tokai Power Station in 2008. In this study, we paid attention to the radionuclides of Co-60 and Cs-137 that adhered to piping, and the dispersion behavior of aerosol was measured and contamination prevention effect was assured. As a result, the data show that the cesium concentrates about four times higher than cobalt. Moreover, the effects of the prevention measures of contamination were confirmed and the behavior of the radioactive aerosol became clear and the effective findings about the dose evaluation of the dismantling were collected.
Uptake of 14C-Acetic Acid by Rice Plant as Related to Root Function and Microbial Activity on the Root Surface
Authors: Ogiyama Shinichi , Ishii Nobuyoshi , Uchida Shigeo
Published: December 2010

Experiments using rice plants (Oryza sativa L.) were conducted to examine uptake of 14C-acetic acid via the root and 14C behavior on the root surface. For hydroponics, three types of rice plants were cultured with 14C-acetic acid solution: complete plant, half-rooted plant, and non-rooted plant. Also, for the root incubation experiment, sterilized root and non-sterilized root were incubated with 14C-acetic acid solution. The 14C radioactivities in the plant parts and solution were measured. Non- and half-rooted plant had 14C radioactivity in their aerial part, but the complete plant did not. The trends of radioactivity levels in the solution were directly opposite to those of plant root biomass. A high level of 14C radioactivity was observed on the entire root surface of non-sterilized root in the incubation experiment, and 14C radioactivity in the solution also remarkably decreased from 7 h to 96 h after the 14C addition. These results suggest that the amount of 14C-acetic acid absorbed by the plant through the roots is very small. However, the plant absorbs 14C-acetic acid through breaks in the roots. Once 14C-acetic acid is inside the plant, it immediately transfers to the shoots. Degradation of 14C radioactivity in the solution and 14C fixation on the root surface arise from the context of microbial activities.
The Effect of High-Scatter Shielding Geometries in Validating the Dose Inferred by N-Visage™
Authors: Adams Jamie C. , Joyce Malcolm J. , Mellor Matthew
Published: December 2010

The aim of this paper is to further validate the physical capability of N-Visage™ under more challenging shielding geometries, when the number of mean free paths is greater than one. N-Visage™ is a recently established technique developed at REACT Engineering Ltd. The software locates radionuclide sources and contours radiation magnitude. The N-Visage™ software uses a geometric computer model combined with measured spectra. The software is able to estimate source locations through shielding materials by using mass attenuation coefficients to calculate the number of unscattered gamma photons arriving at the detector, and build-up factors to estimate scatter contribution to dose rate. The experiments described in this paper were carried out in a high-scatter environment using cobalt-60 and caesium-137 sources, these two sources are the primary sources of radiological contamination found in the nuclear industry. It is hoped that this will further assist in the identification, characterisation and removal of buried radiologically contaminated waste.
The Radioactivity Depth Analysis Tool (RADPAT)
Authors: Shippen B. Alan , Joyce Malcolm J.
Published: December 2010

The Radioactive Depth Analysis Tool (RADPAT) is a PhD bursary project currently being undertaken at Lancaster University in the UK. The RADPAT project involves the development of nuclear instrumentation capable of ascertaining depth of radioactive contamination within legacy plant materials such as concrete. This paper evaluates the merits of two types of detector; sodium iodide (NaI(Tl)) and cadmium zinc telluride (CZT), both of which have been identified as possible solutions for the final RADPAT detector. A bespoke concrete phantom has been developed to allow a set depth of simulated contamination to be obtained with a low measurement error within a concrete analogue: silica sand. Utilising this phantom, in combination with the selected detectors, a set of measurements have been obtained varied with increasing depth of caesium-137 contamination. By comparing the relative attenuation of the x-ray and γ-ray photo-peaks from the data-set to that suggested by a differential attenuation law, a set of model parameters can be obtained. This model, once calibrated, describes the contact depth of contamination with the relative intensity of the peaks in a measured spectrum with a high degree of accuracy. Thus, this technique allows for a set of measurements across the surface of a given material to obtain the inherent distribution of the depth of caesium-137 contamination. This paper is primarily interested in the ability of each detector type to derive the attenuation model, paying particular attention to the associated statistical uncertainty of the fitted parameters and thus the error in the derived depth. The paper describes the contributing effects of the inherent properties of each detector; effects such as their energy resolution, absolute efficiency as well as peak-to-Compton ratio. Finally a commentary on the applicability of each selected detector type is presented, including a comment on the extension the technique to a more generic, real world solution.
Analytical Methodology for Optimization of Waste Management Scenarios in Nuclear Installation Decommissioning Process
Authors: Zachar Matej , Danisˇka Vladimi´r , Reha´k Ivan , Vasˇko Marek , Necˇas Vladimi´r
Published: December 2010

The nuclear installation decommissioning process is characterized by production of large amount of various radioactive and non-radioactive waste that has to be managed, taking into account its physical, chemical, toxic and radiological properties. Waste management is considered to be one of the key issues within the frame of the decommissioning process. During the decommissioning planning period, the scenarios covering possible routes of materials release into the environment and radioactive waste disposal, should be discussed and evaluated. Unconditional and conditional release to the environment, long-term storage at the nuclear site, near surface or deep geological disposal and relevant material management techniques for achieving the final status should be taken into account in the analysed scenarios. At the level of the final decommissioning plan, it is desirable to have the waste management scenario optimized for local specific facility conditions taking into account a national decommissioning background. The analytical methodology for the evaluation of decommissioning waste management scenarios, presented in the paper, is based on the materials and radioactivity flow modelling, which starts from waste generation activities like pre-dismantling decontamination, selected methods of dismantling, waste treatment and conditioning, up to materials release or conditioned radioactive waste disposal. The necessary input data for scenarios, e.g. nuclear installation inventory database (physical and radiological data), waste processing technologies parameters or material release and waste disposal limits, have to be considered. The analytical methodology principles are implemented into the standardised decommissioning parameters calculation code OMEGA, developed in the DECOM company. In the paper the examples of the methodology implementation for the scenarios optimization are presented and discussed.
Dose Assessment of Personnel Handling Conditioned Radioactive Waste
Authors: Pa´nik Michal , Zachar Matej , Necˇas Vladimi´r
Published: December 2010

Radioactive waste may arise throughout the lifetime of any type of nuclear facility. This waste has to be isolated from the environment using the engineered and natural barriers of near surface or deep geological radioactive waste repository. Before final disposal, the waste volume is reduced in the treatment process and then it is immobilized into the stable matrix. In Slovakia, the treated radioactive waste is conditioned into fibre-reinforced concrete containers using a cementation technology. These containers are the only overpacks approved for near surface disposal in the National Radioactive Waste Repository located at Mochovce nuclearsite (Slovakia). Every filled container has to fulfill the defined limits and conditions for safe transport and disposal. The dose parameters, determining the safety of personnel handling a container, are one of them. Basically, the doses are measured on the container walls’ outer surfaces and on the lid of the container. The dose monitoring in 1 meter distance from the container walls and from the lid of the container is performed before transportation as well. The calculation code VISIPLAN 3D ALARA is a planning tool and it calculates dose parameters also for the above mentioned positions concerning the fibre-reinforced concrete container which contains waste with different physical or radiological characteristics. In the paper, calculated data are compared with in-situ measurements. Using VISIPLAN 3D ALARA planning tool, various scenarios are evaluated. Finally, the optimization leading to the lowest radiation exposure of personnel handling the filled fibre-reinforced containers is discussed.
Financial Risks of Post-Closure Custodial Care for the Barnwell Radioactive Waste Disposal Facility
Authors: Baird Robert D. , Newberry William F.
Published: December 2010

This paper reports evaluations of the adequacy of the Barnwell Extended Care Fund in light of identified risks, with the conclusion that the fund is sufficient to cover the costs and uncertainties associated with planned post-closure care of the Barnwell, South Carolina low-level radioactive waste disposal facility. It reviews background information pertinent to the facility’s post-closure monitoring and maintenance and describes financial responsibility for post-closure activities. It identifies and briefly characterizes the activities planned to be conducted following facility closure and presents the mid-range estimate of planned post-closure costs. The paper identifies and quantifies sources of uncertainty in activities and costs planned for post-closure care and presents 50-, 80-, and 95-percent confidence levels of planned costs. The fund is currently sufficient to cover some but not all of the costs that might be incurred as a result of unplanned events. The paper identifies, characterizes, and quantifies unplanned events, possible consequences, and probabilities of occurrence. The paper presents costs that might be incurred in responding to the unplanned initiating events and identifies levels of confidence that the fund is adequate to cover such costs.
Amendments to the U.S. Environmental Protection Agency’s Public Health and Environmental Radiation Protection Standards for Yucca Mountain, Nevada (40 CFR Part 197)
Authors: Clark Raymond L. , Czyscinski Kenneth , Rosnick Reid J. , Schultheisz Daniel
Published: December 2010

In 2001, as directed by the Energy Policy Act of 1992, the United States Environmental Protection Agency (EPA) issued public health and environmental radiation protection standards for the proposed repository at Yucca Mountain, Nevada. Several parties sued the Agency on numerous aspects of the rule. A Federal Court upheld EPA on all counts except for the compliance period associated with the individual-protection standard, which the Agency had limited to 10,000 years for a number of technical and policy reasons. However, the National Academy of Sciences (NAS) had recommended that the standard be set for the time of peak risk, within the limits imposed by the long-term stability of the geologic environment, which NAS estimated at 1 million years. EPA’s standards required that the Department of Energy (DOE) project doses to the time of peak dose but did not apply a compliance standard to these longer term projections. The Court ruled that EPA’s 10,000-year compliance period was inconsistent with the NAS recommendation. This aspect of the rule was vacated and remanded to the Agency for revision. In 2005, EPA proposed amendments to the standards. Following public hearings and a public review period, the final amendments were issued in September 2008. This paper discusses the new requirements.
Completion of the South Alligator Valley Remediation: Northern Territory, Australia
Authors: Waggitt Peter , Fawcett Mike
Published: December 2010

13 uranium mines operated in the South Alligator Valley of Australia’s Northern Territory between 1953 and 1963. At the end of operations the mines, and associated infrastructure, were simply abandoned. As this activity preceded environmental legislation by about 15 years there was neither any obligation, nor attempt, at remediation. In the 1980s it was decided that the whole area should become an extension of the adjacent World Heritage, Kakadu National Park. As a result the Commonwealth Government made an inventory of the abandoned mines and associated facilities in 1986. This established the size and scope of the liability and formed the framework for a possible future remediation project. The initial program for the reduction of physical and radiological hazards at each of the identified sites was formulated in 1989 and the works took place from 1990 to 1992. But even at this time, as throughout much of the valley’s history, little attention was being paid to the long term aspirations of traditional land owners. The traditional Aboriginal owners, the Gunlom Land Trust, were granted freehold Native Title to the area in 1996. They immediately leased the land back to the Commonwealth Government so it would remain a part of Kakadu National Park, but under joint management. One condition of the lease required that all evidence of former mining activity be remediated by 2015. The consultation, and subsequent planning processes, for a final remediation program began in 1997. A plan was agreed in 2003 and, after funding was granted in 2005, works implementation commenced in 2007. An earlier paper described the planning and consultation stages, experience involving the cleaning up of remant uranium mill tailings and other mining residues; and the successful implementation of the initial remediation works. This paper deals with the final planning and design processes to complete the remediation programme, which is due to occur in 2009. The issues of final containment design and long term stewardship are addressed in the paper as well as some comments on lessons learned through the life of the project.
Characterization of the Radiochemical Activity in CANDU Steam Generators
Authors: Husain Aamir , Verzilov Yury , Suryanarayan Sriram , Antoniazzi Armando B.
Published: December 2010

The scope of mid-life refurbishment activities at some CANDU plants includes replacement of the existing steam generators. Shipment of the discarded steam generators for interim storage, intact disposal or for recycling via metal melting requires an assessment of dose rates and the inventory of radionuclides within the components. Kinectrics was contracted by CANDU utility owners to develop such data. This paper presents the detailed methodology employed to develop dose rate and radionuclide inventory data for radionuclides within both in-service and the out-of-service (in-storage) steam generators. The data were developed as follows: a) archived tube sections from various steam generators were characterized; scaling factors were derived using the detailed alpha and beta activity data obtained, b) in-situ gamma spectrometry (using germanium and cadmium zinc telluride detectors) and dose rate surveys were performed at various steam generators and c) a detailed assessment of the tritium inventory in various primary and secondary side components was performed.
BPEO/BPM in Recycling of Low Level Waste Metal in the UK
Authors: Dodd Kevin , Robinson Joe , Lindberg Maria
Published: December 2010

Best Practicable Environmental Option (BPEO) and Best Practicable Means (BPM) are concepts well established in the nuclear industry to help guide and inform waste management decision making. The recycling of contaminated metal waste in the UK is not well established, with the majority of waste disposed of at the Low Level Waste Repository (LLWR) at Drigg. This paper presents an overview of the Strategic BPEO study completed by Studsvik examining the options for low level metal waste management and a subsequent BPM study completed in support of a proposed metals recycling service. The environmental benefits of recycling metals overseas is further examined through the application of lifecycle analysis to the metals recycling process. The methodologies used for both studies are discussed and the findings of these studies presented. These indicate that recycling contaminated metal is the preferred option, using overseas facilities until UK facilities are available. The BPM for metals recycling is discussed in detail and indicates that a tool box for processing metal waste is required to ensure BPM is applied on a case by case basis. This is supported by effective management of waste transport and waste acceptance criteria. Whilst the transport of contaminated metal overseas for treatment adds to the environmental burden of metals recycling, this when compared with the production of virgin metal, is shown to remain beneficial. The results of the Studsvik studies demonstrate the benefits of recycling metals, the options available for such a service and challenges that remain.
The Role of the National Low Level Waste Repository Operator in Delivering New Solutions for the Management of Low Level Wastes in the UK
Authors: Walkingshaw Martin
Published: December 2010

The UK National Low Level Waste Repository (LLWR) is located near to the village of Drigg in West Cumbria. It is the principal site for disposal of solid Low Level Radioactive Waste (LLW) in the United Kingdom. This paper describes the program of work currently being undertaken by the site’s operators, (LLW Repository Ltd and its newly appointed Parent Body Organisation), to extend the life of the LLWR and reduce the overall cost of LLW management to the UK taxpayer. The current focus of this program is to prevent disposal capacity being taken up at LLWR by waste types which lend themselves to alternative treatment and/or disposition routes. The chosen approach enables consignors to segregate LLW at source into formats which allow further treatment for volume reduction or, (for wastes with lower levels of activity), consignment in the future to alternative disposal facilities. Segregated waste services are incorporated into LLW Disposal commercial agreements between the LLWR operator and waste consignors.