bluuebunny/crossref_metadata_2025 huggingface dataset:
Transactive-20 Container for Transporting Large Volumes of Intermediate Level Waste
Authors: Jenkins Jon , Gaffka Tony , Withers Steve
Published: March 2009

AEA Technology has commissioned the Transactive-20 container to transport large quantities of intermediate level waste (ILW). The container was designed and built by RWE NUKEM and has the appearance of a standard ISO freight container approximately 6m long × 2.4m wide × 2.6m high (20’ × 8’ × 8’ 6”). Inside it is a tubular stainless steel pressure vessel, 5.5m long × 1.9m diameter, that enables great flexibility in the payload type, for example: • up to thirty standard 227 litre (55 US gallon) drums of contact-handled ILW (CHILW); • up to five 500 litre drums of CHILW; • multiple packages of remote-handled ILW (RHILW). A variety of other desings of internal furniture can be developed to accommodate inner containers to meet specific customer requirements such as pallets of bulk items, glovebox waste or drums of radioactive liquid. The maximum gross payload is 6.5 tonnes, which is equivalent in practice to about 4 tonnes of CHILW. The maximum laden weight is 26 tonnes. The package is approved to IAEA Type B (U) F standards for transporting up to 3kg of fissile material. This paper describes the design and operation of Transactive-20 and our experience of transporting plutonium-contaminated material (PCM) between UKAEA’s Winfrith and Harwell sites. The loading system enables the container to remain on the transport vehicle during the whole of the load/transport/unload/transport cycle between the sites and therefore minimises turnaround time. For the operation described here, UKAEA lifts six standard 227litre drums into a circular aluminium container, known as a ‘carousel’, which is then closed and turned automatically through 90° for loading into Transactive’s inner vessel. Five such carousels are loaded in this way and locked firmly in place, before bolting on the lid and leak testing the vessel. A hydraulically-driven outer door is then closed and fastened to make the package ready for transport. Unloading is a straightforward reverse sequence of the loading operation. This paper describes in detail the design of the Transactive-20 and our operational experience with PCM and other fissile CHILW outlined above. It also outlines our plans to use the container for remote-handled ILW (RHILW). For example, a carousel has been constructed to move spent 60Co or 137Cs sources from medical, educational or industrial irradiators to a centralised long-term storage facility.
Packaging and Conditioning of Solvent Wastes From Decontamination Operation Cernavoda NPP
Authors: Deneanu Nicoleta , Dianu Magdalena , Teoreanu Ion
Published: March 2009

Solvent wastes produced at Cernavoda NPP consist of miscellaneous acetone, toluene, methanol, chloroform, triclorethan, white spirit and ethylene glycol. These are normally LLW containing only relatively small quantities of beta/gamma emitting radionuclides and varying amounts of tritium with activity below E08Bq/l. This paper is a review of some current innovative work of Waste Management Facility from Institute for Nuclear Research Pitesti in the development of a viable solidification technology to convert solvent wastes into a stable monolithic form, which minimises the probability to release tritium in the environment during interim storage, transportation and final disposal. The paper presents the author’s research on immobilisation of solvent wastes by cementation using aluminium stearate additive. A quality assurance program should accompany the production of waste forms. The goal of all tests should be to obtain a license for a certain process from a competent authority. The process will be clean, which means there will be no secondary waste and low doses to the personnel will be achieved; the product quality will meet any National requirement and the reproducibility of the process meets any QA requirement.
The Importance of Geochemical Characterization of Repository Host Horizons for Radioactive Waste Disposal: Saligny Repository Site for L/ILW, Romania
Authors: Anghel Ioana , Bucur Crina , Ware Stuart D. , Pavelescu Margarit
Published: March 2009

The Saligny Site, situated in the vicinity of the Cernavoda Nuclear Power Plant in Romania, has been chosen as a repository site for placement of future Low and Intermediate Level Nuclear Waste. Scientific investigations are on going to help assess the long-term performance of the nuclear waste repository and include the development of a site conceptual model, geochemical characterization of the proposed liner material and the geologic formations surrounding the repository. Geochemical characteristics of the proposed waste site and of repository liner material are essential information in providing key model parameters needed to determine the mechanisms and rates for the potential release and transport of radionuclides in the environment. A series of laboratory sorption and diffusion experiments in support of the site conceptual model were run at Los Alamos National Laboratory, USA and Institute for Nuclear Research, Romania. These experiments compared responses of cesium-137 and tritiated water as radionuclides, and included natural soil samples from Saligny site and cement liner material. Sorption experiments focused on studying the affinity of cesium-137 for soil samples from: a clay rich loess horizon, a compacted red clay horizon and a Portland type cement, proposed as liner for the repository. A Saligny synthetic water was used for the soil sorption experiments. For the cement sorption and diffusion experiments, NaCl 10mM water was necessary in order to avoid insolubility problems at the high pH (around 12) when in contact with the cement. Preliminary studies of the geochemistry for radionuclide-soil matrix and radionuclide-concrete liner interactions were obtained through the sorption and diffusion experiments. The sorption isotherms were linear for all samples with high Kd values for soil samples and low values for the cement. Mineralogical information obtained through quantitative X-ray diffraction (QXRD) analysis, for the studied soil samples were in agreement with the sorption results. The high amount of clay present in the soil samples was determined to be the main contributor for the strong sorption. Diffusion cells were prepared for comparing hydrologic response characteristics for tritiated water and cesium-137, through the measurement of diffusion coefficients as influenced by the soil and cement matrices. Tritiated water and cesium-137 diffusion coefficients were determined from the change in radionuclide concentration profile over time while diffusing through the soil or cement matrix, and were found in agreement with the values listed in literature findings.
Soil Based Barriers for a Low and Intermediate Level Radioactive Waste Disposal
Authors: Antonov Dimitar
Published: March 2009

The loess terrains near “Kozloduy” NPP are among the prospective areas for the disposal of low and intermediate level radioactive waste. The analysis of the loess properties has shown two main problems: a loess collapsibility and water permeability. Using a soil-cement cushion under the repository foundation and a soil-cement backfill between the containers is a possibility to avoid these disadvantages. In this connection loess-cement mixtures with bentonite and clinoptilolite additives have been investigated. The aim of mixtures is to improve the impermeability and sorbtion properties against radionuclide migration. In the paper strength parameters of two kind of mixtures are discussed. According to their water content some are compacted at the optimum moisture content until the maximum dry density and others are compacted at higher moisture content equal to the liquid limit of loess. For the first type of mixtures the unconfined compressive strength (UCS) varies from 2 to 6 MPa depending on the cement and additives percents. Permeability measurements have shown satisfactory results. The UCS for the second type of mixtures is less than the first type, but is sufficient for a backfill between the waste containers. The conclusion is that the loess-cement mixtures, especially these with clinoptilolite additive, are prospective as barriers of a low and intermediate level radioactive waste repository.
Distribution of Palladium During Spent Fuel Reprocessing
Authors: Pokhitonov Yu. , Romanovski V. , Rance P.
Published: March 2009

The principal purpose of spent fuel reprocessing consists in the recovery of the uranium and plutonium and the separation of fission products so as to allow re-use of fissile and fertile isotopes and facilitate disposal of waste elements. Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants (NPPs,) there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. Given current predictions for nuclear power generation, it is predicted that the quantities of palladium to be accumulated by the middle of this century will be comparable with those of the natural sources, and the quantities of rhodium in spent nuclear fuel may even exceed those in natural sources. These facts allow one to consider spent nuclear fuel generated by NPPs as a potential source for creation of a strategic stock of platinum group metals. Despite of a rather strong prediction of growth of palladium consumption, demand for “reactor” palladium in industry should not be expected because it contains a long-lived radioactive isotope 107Pd (half-life 6,5·105 years) and will thus be radioactive for a very considerable period, which, naturally, restricts its possible applications. It is presently difficult to predict all the areas for potential use of “reactor” palladium in the future, but one can envisage that the use of palladium in radwaste reprocessing technology (e.g. immobilization of iodine-129 and trans-plutonium elements) and in the hydrogen energy cycle may play a decisive role in developing the demand for this metal. Realization of platinum metals recovery operation before HLW vitrification will also have one further benefit, namely to simplify the vitrification process, because platinum group metals may in certain circumstances have adverse effects on the vitrification process. The paper will report data on platinum metals (PGM) distribution in spent fuel reprocessing products and the different alternatives of palladium separation flowsheets from HLW are presented. It is shown, that spent fuel dissolution conditions can affect the palladium distribution between solution and insoluble precipitates. The most important factors, which determine the composition and the yield of residues resulting from fuel dissolution, are the temperature and acid concentration. Apparently, a careful selection of fuel dissolution process parameters would make it possible to direct the main part of palladium to the 1st cycle raffinate together with the other fission products. In the authors’ opinion, the development of an efficient technology for palladium recovery requires the conception of a suitable flow-sheet and the choice of optimal regimes of “reactor” palladium recovery concurrently with the resolution of the problem of HLW partitioning when using the same facilities.
Measurement of Glass Corrosion in Boom Clay Disposal Conditions: First Results of the Experimental Programme 2000-2003 of SCK•CEN
Authors: Lemmens Karel , Aertsens Marc , Pirlet Ve´ra , Maes Norbert , Moors Hugo , Van Iseghem Piere
Published: March 2009

To estimate the lifetime of vitrified high level waste (HLW-glass) in Boom Clay disposal conditions, the dissolution behaviour of waste glass has been studied with experiments performed in surface laboratories and in the HADES underground research facility of SCK·CEN since 1980. We present the main topics and first results of the SCK·CEN programme 2000–2003. This programme focuses on the following items: (1) the diffusion/sorption/precipitation of silica in Boom clay or backfill clay, (2) demonstration of glass dissolution behaviour in realistic test conditions, (3) the effect of presaturation of the clay with silica, and (4) the estimation of near field concentrations of critical isotopes. The experiments have shown so far that Si, released by the glass, is effectively immobilized by Boom Clay, but it can nevertheless diffuse into the clay without immediately precipitating. The dissolution rate of glass SON68 and SM539 is determined in Boom Clay at in situ density and at 30°C (this is the long-term temperature expected near the waste glass packages in a Boom Clay repository). The dissolution rates, based on glass mass losses, are constant during the first year, at ∼ 0.010 g.m−2.day−1 for glass SON68 and ∼ 0.012 g.m−2.day−1 for glass SM539. The addition of glass frit causes a decrease of the glass dissolution rate, both with glass SON68 and SM539, and both in Boom Clay and in FoCa-clay. In FoCa-clay at high density with glass frit, the dissolution rates, based on glass mass losses, after 8 months at 30°C are ∼ 0.001 g.m−2.day−1 (SM539) and ∼0.005 g.m−2.day−1 (SON68). Because the experiments performed in Boom Clay and FoCa-clay with glass frit simulate realistic conditions (high clay density, low temperature), they can be used to estimate the maximum glass dissolution rate in a (Boom) clay repository. The corresponding minimum lifetime of a glass canister, calculated with the SCK·CEN code for lifetime predictions, is of the order of 105 to 106 years, if we neglect the internal glass surface area (due to cracking). In more diluted clay suspensions with glass frit, the glass dissolution rate is 10−4 to 10−5 g.m−2.day−1 or even zero. This would correspond to a lifetime of >>106 years. So far, there is no indication that the addition of glass frit leads to secondary phase formation at low temperature (30–40°C). Leach experiments with doped glasses SON68 and SM539 suggest that the maximum concentrations of most critical radionuclides in near field conditions are lower than the best estimate solubilities used for performance assessment studies in Boom Clay. For Se, relatively high concentrations were measured, though. The research programme for the underground laboratory is not discussed.
Implementation of a Quality Assurance System in Radioactive Waste Management in Cuba
Authors: Salgado M. M. , Navarro J. C. Benitez , Ferna´ndez I. M. , Marrero M. M.
Published: March 2009

The Centre for Radiation Protection and Hygiene is the institution responsible for Radioactive Waste Management Service in Cuba. This Service comprises: centralized collection, transportation, treatment, conditioning, long term storage, and disposal of radioactive waste, as well as dismantling, decontamination and decommissioning of small nuclear facilities. Radioactive waste should be managed in such a way as to ensure adequate safety and protection of both human health and the environment. In order to fulfil this principle a Quality Management System has been implemented for the Radioactive Waste Management Service, based on the ISO 9000 Standards. The Quality System provides, as appropriate: • adequate assurance that the requirements specified by the Regulatory Authority relating to protection and safety are satisfied; • adequate assurance that the customer requirements are satisfied; • control of all kind and inventories of radioactive waste through the implementation of a comprehensive system for record keeping; • continuous improvement to guarantee the cost minimization related with the radioactive waste management; • and quality control mechanisms and procedures for reviewing and assessing the overall effectiveness of the system. In the year 2002, as a conclusion of a certification audit performed by the direction of the Centre for Radiation Protection and Hygiene, the Quality Management System implemented for the Radioactive Waste Management Service was successfully certified according to the requirements of the ISO 9001:2000 Standard.
Approach to Derivation of Waste Acceptance Criteria for Mochovce Disposal Facility
Authors: Hanusˇi´k V. , Kusovska´ Z. , Bala´zˇ J. , Mrsˇkova´ A.
Published: March 2009

In Slovakia, low and intermediate level radioactive wastes are disposed in a near-surface repository at Mochovce site. The repository, which was commissioned in September 2001, has a disposal capacity 22,320 m3. It is a vault-type concrete structure repository with reinforced concrete containers as the final waste packages. The Mochovce repository is designed to receive acceptable radioactive wastes from decommissioned A-1 power plant at Jaslovske´ Bohunice, operational waste from NPPs V-1 and V-2 at Jaslovske´ Bohunice site and NPP Mochovce, as well as institutional radioactive wastes. Generally, calculation endpoint of disposal facilities performance assessment is radiological impact on humans and environment. In that case, starting points of assessment are the waste activity concentrations and inventory activity. The acceptance of radioactive waste in Mochovce repository is one of the many elements that directly contribute to the safety of the disposal system. In Mochovce repository safety analysis, end points are both the concentration per package and total activity values. On the other hand, radiological protection criteria are the starting points of the calculation. This approach was developed and applied because the actual inventory that will be disposed of is highly uncertain. As a result of the accidents, the primary circuit was contaminated by fission products. Some auxiliary circuits and facilities were also contaminated. The complicated problem is the relatively high content of long-lived radionuclides (inclusive transuranic elements) in some waste streams. After two technological incidents at NPP A-1 uncertainties in waste inventory are large because of variability in the types of waste streams and variability in the quality and completeness of the waste characterization data. This paper presents the philosophy of safety analysis, development of scenarios, their modelling and approach that have been used to derive waste acceptance criteria, specifically limits of activity. The approach consists of the determination of radionuclides important for safety, the use of relevant safety scenarios, the setting of dose limits associated with scenarios, the calculation of activity limits and application of the simple summation rule. Finally, information is provided about short operation of the repository.
Innovations Using Ultra-Light Cellular Grout and High Strength, Medium Density Grout for Stabilization of Irradiated Components During Decommissioning
Authors: Peterson Leonard , Heath Danen , Stephens Patrick
Published: March 2009

Injection of cellular grout into tunnels, tanks, reactor vessels, embedded piping and other radioactive components stabilizes the internal contents, fills voids, and fixes loose contamination. Injection of cellular grout also significantly reduces radiation levels in the vicinity of the component. Ultralight cellular grout mix designs (25–35 lbs/ft3) have been used where weight is a critical concern. Medium density cellular grouts (60–105 lbs/ft3) with compressive strengths exceeding 3000 lbs/in2 have been used in a nuclear applications to reduce radiation levels and increase structural strength and stability. Geo-technical grout injection technology has been adapted to nuclear applications with the ability to inject special “water-absorbing” and “water-displacing” mix designs over long distances where remote injection is required.
Decommissioning Software for the Kozloduy NPP
Authors: Ribarski Valentin , Ba¨cker Axel , Schulz Roland , Demireva Emilia
Published: March 2009

The Bulgarian government and the European Commission have agreed on the close down date for units 1 & 2 and for units 3 & 4 the decision is still pending. In order to support the decommissioning activities, planning and licensing projects have been performed for the units 1 & 2 in the frame of PHARE by the EC. Meanwhile the units 1 & 2 are shut down on the end of 2002. The implementation of hardware activities is presently being supported by the EBRD. In the frame of a project financed by the IAEA and the Bulgarian Decommissioning Fund, the development of a Decommissioning Management System (DeManS) and the installation of the necessary hardware are going on. DeManS is based on the most up-to-date Oracle database and application server architecture in order to offer maximum scaling, failsafe security and flexibility. Different integrated modules are available for the Management of the complete decommissioning project. These modules support all the tasks necessary for the: • documentation, • environmental information, • registration of the plant status, • Safe Enclosure, • planning/calculation and • dismantling planning and supervision. The project started in 2002 and will continue until late 2003. The hardware installation has been completed in the beginning of 2002. The first modules for documentation (DMS), Environmental Information (EIS) and Registration (REG) have been developed and installed up to May 2003. The functionality of the DeManS will be described for each module as well as for the interaction between them.
Fixation of Actinides in Zirkonium Based Ceramics
Authors: Barrier D. , Bukaemskiy A. A. , Soe K. S. , Titov M. M. , Modolo G.
Published: March 2009

In recent years, various fuel types or targets, free from U-238, have been proposed for the transmutation of plutonium and the minor actinides. Yttrium-stabilised zirconia is currently under investigation because of its small neutron cross section and is high solubility with PuO2 and AmO2. The aim of the present work is to investigate the behaviour of Yttrium Stabilised Zirconium Oxide in presence of Cerium, as simulate of minor actinides. Several oxide materials composed of Zr, Y, Ce (5–25%), have been synthesised by coprecipitation method at different calcination temperature and the products have been characterised by Optical Microscopy (OM), Thermogravimetry (TG) coupled with Thermal Differential Analysis (DTA), X-ray Line Diffraction (XRD), Scanning Electron Microscopy (SEM) Energy Dispersive X-Ray analyser (EDX) etc. The investigations show that, the structure of the powder, characterised by the phase condition, the lattice parameter, the crystalline size and the lattice distorsion, varies with the CeO2 content, and has an influence on the sinterability. “Mild” methods of powder preparation (drying, low-temperature calcination, wet grinding by attrition) were used before sintering. For all investigated Cerium concentrations, the optimal parameter of pressing and sintering were determined. For a wide Cerium concentration region, from 5 to 20 wt%, the density of compact materials is 92–95% of TD, the microhardness 13.5 GPa and the pellets have a good morphology, grains and pores being homogeneous distributed in material.
Radioactive Contamination of Concrete: Uptake and Release of Radionuclides
Authors: Bath Adrian , Deissmann Guido , Jefferis Stephan
Published: March 2009

Concrete in nuclear installations may become contaminated by various radionuclides. Consequently, decommissioning and dismantling produce considerable quantities of potentially contaminated materials that must be managed safely and cost-effectively. In this paper we present preliminary results from a research project that aims to improve knowledge about release behaviour of radionuclides from contaminated concrete, and that proposes a scientific approach to calculating the source term for radiological dose assessment for the various management options (e.g. direct reuse, recycling, disposal of rubble). The first step is to consider which nuclides are likely to have contaminated concrete, where they might be located in concrete, and the extent to which they are chemically bound to concrete constituents. Relevant radionuclides include 60Co, 63Ni, 90Sr, 137Cs, 129I, U, Pu, Am and other actinide elements. Some nuclides are likely to be bound in specific solid phases and others are sorbed to greater or lesser degrees. The proposed modelling of releases from concrete takes into account the chemical behaviour (speciation, sorption and solubility) of the individual radionuclide contaminants and their binding to concrete phases. Other important factors that will influence release are mechanical and chemical condition of concrete, including cracking, carbonation, sulfate attack and degree of water saturation. Model calculations illustrate the potential release processes of desorption-diffusion, leaching (shrinking core model) and dissolution of discrete solid phases. For example, a scoping calculation suggests that 50-year old concrete may be contaminated with 129I to about 1 cm depth from the surface or more if the concrete is degraded, and that subsequent release will occur slowly by diffusion. Strongly sorbed or particulate nuclides such as Pu are likely to remain at the surface. Predicting the behaviour of some nuclides (e.g. Ni, U) is more uncertain because of uncertainty in the key parameters and their dependence on the local chemical conditions. Release models and source terms have been developed as the starting point for (i) the modelling of radiological consequences (i.e. dose assessments) of disposal options for building materials from nuclear installations and the optimisation of the disposal process (i.e. selection of cost-effective and reasonable disposal options), and (ii) the assessment of recycling/reuse options of slightly contaminated materials in order to reduce the amount of waste for disposal.
Remediating Sellafield: A New Focus for the Site
Authors: Baldwin N. D.
Published: March 2009

The structure of the ownership and management of nuclear liabilities on civil sites in the United Kingdom is undergoing fundamental change. The UK Government will take responsibility for the liabilities on the UKAEA, BNFL Sellafield and Capenhurst sites and the Magnox Generation sites. When fully implemented the accountability for long term strategy will rest with the new Government Nuclear Decommissioning Authority (NDA), and contracts will be placed on M&O contractors to manage the site and implement the liabilities discharge plans. At Sellafield whilst the commercial reprocessing and MOX contracts continue, it is clear that the overall focus of the site has changed to remediation. Until the NDA is established the task of undertaking the planning is the responsibility of BNFL. To address this task the Site Remediation Team has been established. The production of the Sellafield Lifecycle Baseline Plan requires the existing long term decommissioning and waste management plans (primarily produced for provisioning purposes) together with several other specific strategies to be combined and developed into a co-ordinated and optimised plan for the remediation of the Sellafield Site, recognising the ongoing reprocessing, MOX manufacture and long term fuel storage activities. An important principle within the plan is to achieve early hazard reduction whilst demonstrating value for money. The paper will address the scale of the remediation challenge and the process being followed to develop the necessary strategy. The paper will appeal to those involved in managing remediation of large, complex and interdependent nuclear sites.
MONA, Public Participation in the Siting of a LLW Repository in Mol, Belgium
Authors: Meus Bert , Ceulemans Hugo
Published: March 2009

In 1996 it became clear that the approach used by ONDRAF/NIRAS i in selecting a site for a low-level nuclear waste repository, had failed. Selection based solely on technical criteria proved to be totally inadequate. This led to a drastic change in the decision-making process. Methods needed to be developed to integrate a repository project at the local level. Ever since, ONDRAF/NIRAS has concentrated on establishing local partnerships in communities expressing interest in hosting a repository. Mol was one of three communities that expressed interest and in 2000 a local partnership called MONA (Mols Overleg Nucleair Afval) was established between ONDRAF/NIRAS and local stakeholders. The partnership provides experts with an opportunity to enter into dialogue with the local stakeholders. All aspects of a possible near-surface or a deep geological repository for low-level waste are being discussed in different working groups. This paper will focus on the work accomplished by the three technical working groups and the criteria, which, in their view, a repository has to satisfy. The impacts of a repository on the environment (radiological, scenic, aesthetic, logistic, ...) and sustained development are important themes for the local stakeholders.
Developments in Radiological Characterization and Remediation of Non-Nuclear Industrial Sites in the Netherlands
Authors: van Velzen L. P. M.
Published: March 2009

In the Netherlands, like in other countries, many industries process ores that contain small amounts of natural radioactivity e.g. phosphate ore for the production of phosphate-acid and fertilizers. These non-nuclear industries have often started their activities many years ago, sometimes 50 to 60 years ago, and cases are known for even longer periods. In these periods no governmental regulations concerning naturally occurring radioactive materials (NORM) existed and it was generally accepted that remaining materials of the industrial process were used for road construction, site improvement, send to household waste belts etc. In the eighties of the last century one became aware that due to industrial processes enhanced concentrations of NORM (TENORM; Technical Enhanced NORM) could occur and employees could receive doses that had to be observed and regulated. Due to this regulations and R&D new processes were set-up and existing factories were closed and successfully decommissioned and dismantled. What remains is that at these sites, due to actions and disposals in the past, enhanced NORM material can still be present that now will fall under the new established regulations. The radiological characterization is a necessarily action for the official ending of the site license in the case that no activities remain that use NORM or produce TENORM. Radiological characterization is also recommended in the case that the property changes of ownership. The characterization gives the new owner the security that there are no spots left with NORM or TENORM above existing governmental limits. While the advantage for the former owner is that no financial funds have to be reserved for an eventual contamination of a non localized spot with NORM or TENORM that supersedes existing governmental limits. In The Netherlands experience has been gained with the radiological characterization, decommissioning and remediation of non-nuclear industrial sites. Used methods will be presented and especially the process of radiological characterization and the definitions of selection and action criteria including motivations.
Decommissioning Cost Calculation Code Based on Proposed Standardised List of Items for Costing Purposes With Integrated Material and Radioactivity Flow Control and Integrated Cost Allocating System
Authors: Danisˇka Vladimir , Rehak Ivan , Vasko Marek , Pritrsky Jozef , Bezak Peter , Kristofova Kristina
Published: March 2009

In 1999 was issued commonly by EC, IAEA and OECD the document “A Proposed Standardised List of Items for Costing Purposes in the Decommissioning of Nuclear Installations” (PSL). The main aim of this document was to establish a standard platform for structure of presented decommissioning costs. The decommissioning costs calculated in various projects have in general different structure and for presenting the decommissioning costs in PSL structure, additional recalculation or rearrangement of data are needed. This can increase the uncertainty of data or it is not sure if they really reflect the cost defined in the PSL structure. The paper describes a newly developed cost calculation code, which respect the PSL structure. The calculated costs data are compatible with the standardised structure. The code is intended to use for decommissioning planning of a NPP with a non-standard radiological situation, shutdown after an accident For this purpose the code has an integrated material and radionuclide resolved radioactivity flow control system in calculation process which enables a direct calculation link between the input material and radiological inventory data and the calculated costs and other decommissioning parameters for increasing the accuracy of calculated data. The code has an integrated cost allocating system for generating of user defined working breakdown structure of the decommissioning project with allocated costs calculated in PSL structure. The working breakdown structure can be on-line optimised using the standard Microsoft Office Project Planning tool software.
Problems of Radioactive Waste Management at Chornobyl Nuclear Power Plant (ChNPP)
Authors: Oskolkov Borys Ya. , Neretin Yuri A. , Saliy Valeryi P. , Seyda Valeryi A. , Fomin Vyascheslav V.
Published: March 2009

According to the assessments the overall amount of radioactive waste (RAW) to be reprocessed and buried at the Chornobyl NPP site amounts to 1,696,738 m3 (without regard for reactor metal structures, dismantling of building structures and plan constructions, and the Unit Shelter building). The overall activity of radioactive waste are evaluated at 1,640,504.64 ΤBq. The RAW management activities are implemented at the Chornoby1 NPP within the frame of several programs of different hierarchy including the State Comprehensive Program for RAW Management in Ukraine, Integrated Program of RAW Management at the Chornoby1 NPP Shutdown Stage and Transformation of the Unit Shelter into an Ecologically Safe System. At the present time a number of key RAW management facilities are being constructed within the frame of the international aid to Ukraine. They are the Liquid Radioactive Waste Plant and Solid Radioactive Waste Reprocessing Complex. As of now, the issues concerning RAW utilization at the Unit Shelter are resolved at a conceptual level. There has not practical decision in relation to a geologic facility. The complexity and scale of ChNPP RAW management problems will require significant efforts of both Ukraine and the whole world community to solve these issues. The task related to removal and final burial of accumulated and generating radioactive waste is one of the main aspects of decommissioning activities at any nuclear power plant. RAW management work is the most important and complicated work performed at the Chornoby1 NPP. The specific features of ChNPP RAW management are as follows: • Variety of RAW generation sources, their types, physical and chemical properties. • Large amount of radioactive wastes which already exist and those generated in the decommissioning process. • Presence of disorganized RAW characterized by wide spatial distribution within the Unit Shelter and at the plant site. • Need to apply a very wide spectrum of various RAW management techniques depending on their location and type. • Need in developing unique techniques to manage special types of RAW located at the site (fuel containing masses of the Unit Shelter). • Large amount and variety of facilities required for RAW final storage. • Absence of reliable and serviceable instrumental procedures and necessary equipment to define RAW properties for RAW separation and classification. • Ecological peculiarities of RAW management within the Chornoby1 zone. • Multiphase decontamination and restoration processes resulting in RAW formation. • Need in integrating RAW management problems at the ChNPP and within the Chornoby1 Exclusion Zone taken as whole. • Long time period required for implementing the whole program of RAW management at the ChNPP. • Large quantity of people involved in RAW management process (local and foreign participants, different organization operated by various departments).
Should We Discount Low Dose Radiation Risk?
Authors: Peckover Richard , Priest Nick
Published: March 2009

This paper addresses four inter-related topics of current concern: (i) the relationship between man-made and natural radiation; (ii) the uncertainties in predicted consequences following low dose irradiation; (iii) the scope for using a Properly Prioritized Precautionary Principle (P4); and (iv) why unobservable effects should not be a cause of concern.
Risk-Based Radiological Site Investigation and Assessment Methodology for the Sellafield Site, Cumbria, UK
Authors: El-Ghonemy Hamdi , Lean Candida , Lee Alexander G.
Published: March 2009

A preliminary assessment of the Sellafield site in Cumbria was undertaken to provide the information for the hazard identification stage in the environmental risk assessment process. This preliminary investigation has been crucial to understanding the site’s history, the zones of potential risk and for the design of a site investigation. The formalised assessment strategy that has subsequently been developed by BNFL for the Sellafield study has drawn upon UK and international guidance to take into account the size of the site and the anticipated complexity of issues. The approach presented represents an advance on existing working practices in contaminated land investigations. The use of a generic list of Features, Events and Processes (FEPs) has been adopted to ensure the systematic and comprehensive appraisal of all relevant uncertainties and sources of risk. Proforma have also been developed to provide a data management and retrieval system that is transparent for quality assurance purposes. This unique approach facilitates the development of conceptual site models as new data become available during the site investigation. It also aids in the development and justification of alternative site conceptualisations and in the early identification of data and interpretative uncertainty.
Lessons Learned Siting and Successfully Processing U.S. DOE Radioactive Wastes Using a High Throughput Vitrification Process
Authors: Prince Robert E. , Bowan Bradley W.
Published: March 2009

This paper describes actual experience applying a technology to achieve volume reduction while producing a stable waste form for low and intermediate level liquid (L/ILW) wastes, and the L/ILW fraction produced from pre-processing of high level wastes. The chief process addressed will be vitrification. The joule-heated ceramic melter vitrification process has been used successfully on a number of waste streams produced by the U.S. Department of Energy (DOE). This paper will address lessons learned in achieving dramatic improvements in process throughput, based on actual pilot and full-scale waste processing experience. Since 1991, Duratek, Inc., and its long-term research partner, the Vitreous State Laboratory of The Catholic University of America, have worked to continuously improve joule heated ceramic melter vitrification technology in support of waste stabilization and disposition in the United States. From 1993 to 1998, under contact to the DOE, the team designed, built, and operated a joule-heated melter (the DuraMelterTM) to process liquid mixed (hazardous/low activity) waste material at the Savannah River Site (SRS) in South Carolina. This melter produced 1,000,000 kilograms of vitrified waste, achieving a volume reduction of approximately 70 percent and ultimately producing a waste form that the U.S. Environmental Protection Agency (EPA) delisted for its hazardous classification. The team built upon its SRS M Area experience to produce state-of-the-art melter technology that will be used at the DOE’s Hanford site in Richland, Washington. Since 1998, the DuraMelterTM has been the reference vitrification technology for processing both the high level waste (HLW) and low activity waste (LAW) fractions of liquid HLW waste from the U.S. DOE’s Hanford site. Process innovations have doubled the throughput and enhanced the ability to handle problem constituents in LAW. This paper provides lessons learned from the operation and testing of two facilities that provide the technology for a vitrification system that will be used in the stabilization of the low level fraction of Hanford’s high level tank wastes.
Long-Term Prediction of Radionuclide Release From Borehole Type Repositories
Authors: Prozorov L. B. , Korneva S. A.
Published: March 2009

As part of a development program for radioactive waste management, two test boreholes have been constructed at the MosNPO “Radon” Site to evaluate the conditions of LILW storage in such facilities, to observe the behaviour of engineered barriers as a function of time, the ability of the system to contain radionuclides and to test the reliability of a monitoring system. To obtain a licence for operation of such boreholes, MosNPO “Radon” specialists have prepared a package of documents, including a “Report on Safety Assessment of LILW Storage in Large Diameter Boreholes” [1]. On the basis of radiation safety standards and requirements being in force in Russian Federation, a range of possible accidents and emergencies during the operational period has been considered in this study. In this paper, the authors present results of a long term safety assessment of LILW disposal in the boreholes, performed within the framework of the above mentioned report.
Characterization of Current State of SRAW Storage Facilities of the Siberian Group of Chemical Enterprises and Radiological Situation on Adjacent Territory
Authors: Bushuev Anatoly , Petrova Ekaterina , Aleeva Tatiana , Zubarev Victor , Proshin Igor , Kachanovsky Alexander , ZaHarova Elena , Ushakov Sergey , Mescheryakov V. , Nikolaev Angrey , Hvostov V. , Savinyh P.
Published: March 2009

The information is presented on solid radioactive wastes (SRAW) repositories of Siberian Group of Chemical Enterprises, which contain spent graphite sleeves and parts metallic constructions of plutonium production reactors (PUGRs). Some of the repositories are filled with water. The samples of concrete walls, soil and water in vicinity if the repositories were taken as well as samples of water and sediments from ones filled with water. 3H, 14C, 137Cs, 241Am, 60Co and some other radionuclides content in the samples was determined. The affect of storage conditions oa various types of SRAW on their radioactive contamination was studied. Penetartion of some radionuclides from repositories into environment was established. Basing on the results obtained, the conclusions were made about state of various types the repositories and on possibility of their further using for the wastes storage.
Comparative Experiences in Environmental Remediation of LLR Waste Sites in Diverse Canadian Environments
Authors: Case G. G. , Zelmer R. L.
Published: March 2009

A variety of sites contaminated with legacy low-level radioactive (LLR) waste materials have been identified across Canada. Many of these sites, associated with former radium and uranium refining and processing operations, are located in urbanized areas of southern Ontario. However, other sites have been discovered at more remote locations in Canada, including northern Alberta and the Northwest Territories. The diversity of waste froms, ranging from pitchblende ore and processing wastes, to discarded luminescent products, combined with construction and transportation logistical issues encountered at these sites, present ongoing challenges for the Low-Level Radioactive Waste Management Office (LLRWMO) to overcome in meeting its mandate to resolve these legacy problems. Since its establishment in 1982, the federal government’s LLRWMO has operated programs to characterize and delineate contaminated historic waste sites across Canada. These programs have included undertaking property decontaminations, waste consolidation and interim storage projects at many sites, and participating with federal and provincial government departments and local communities to consider long-term storage and disposal opportunities. This paper compares four specific environmental remediation programs conducted by the LLRWMO within diverse Canadian settings found at Port Hope and Toronto (southern Ontario), Fort McMurray (northern Alberta), and Vancouver (west coast of British Columbia). Contaminant characterization and delineation, and remediation plan design and implementation aspects of these individual programs span the time period from the early 1980s through to 2002. The individual programs dealt with a variety of legacy waste forms that contained natural radioactive materials such as radium-226, total uranium, total thorium and thorium-230, as well as coincidental inorganic contaminants including arsenic, barium, cadmium, cobalt, lead, mercury, vanadium and zinc. Application of the lessons learned during these individual programs, as well as the development of new and innovative technologies to meet the specific needs of these programs, have enabled the LLRWMO to effectively and efficiently implement environmental remediation solutions that address the variety of Canada’s legacy LLR wastes.introduction.
Safe and Compliant Nuclear Waste Management: What Does It Really Mean?
Authors: Devarakonda Murthy , Biedscheid Jennifer
Published: March 2009

Nuclear waste management is a complex and contentious issue in all parts of the world, involving social, political, technical, and economic interests, and generating a reaction of public suspicion and mistrust in most cases. Not surprisingly, the end goal for all parties involved in the nuclear waste management debate is identical: the safe disposition of the waste in compliance with governing regulations. The governing regulations, in turn, are intended primarily to protect public health and the environment, not just in the present, but well into the future, given the long-lived nature of many radionuclides in the waste. However, each party in the nuclear waste management debate approaches and defines the end goal differently. The balancing of interests and ideas pursued by the government, regulators, scientific community, local watchdog groups, and the general public regarding the end goal affect the way that policies are determined and by whom. The strength of the various arguments and the environment in which they are asserted also plays a role in policy development. The resolution of a nuclear waste management issue in any given case can never be described unequivocally as the “best,” “safest,” or “optimal” solution simply because the various parties and entities involved will very rarely look at the end point from the same perspective (technical, emotional, or political). However, nuclear waste management programs can be designed and developed so that the disparity of expectations and emotions is minimized by means of open communications and a sound technical basis for all decisions. This paper discusses examples of these concepts in the context of the U.S. Department of Energy’s (DOE) Waste Isolation Pilot Plant (WIPP) and Yucca Mountain Programs. These programs, which address the permanent disposition of transuranic (TRU) waste and high-level radioactive waste (HLW), respectively, provide the opportunity to view policy decisions and associated impacts both within the framework of resulting operational realities in the case of the WIPP Program and within the process of defining a strategy for the progress of the Yucca Mountain Program.