bluuebunny/crossref_metadata_2025 huggingface dataset:
Education in Nuclear Decommissioning in the North of Scotland
Authors: Catlow F. , Reeves G. M.
Published: June 2009

This paper describes the work covered and experience gained in the first two years of operation of DERC, a Centre for Decommissioning and Environmental Remediation in the Highlands of Scotland. The Centre is a unique development which was set up to teach nuclear decommissioning as a separate discipline, address the problem of a declining skills base in the field of nuclear technologies and to take advantage of the unique and exceptional innovative, technical and research opportunities offered through the decommissioning of Britain’s fast reactor site at Dounreay. The Centre is an offshoot from North Highland College which is a member of UHI, the University in embryo of the Highlands and Islands. The Centre currently supports ten PhD students completing various diverse projects mainly in the field of nuclear environmental remediation. In addition there are a number of full and part time MSc students who participate in NTEC (Nuclear Technology Education Consortium) a consortium of British Universities set up specifically to engender interest and skills in nuclear technology at postgraduate level. At undergraduate level, courses are offered in Nuclear Decommissioning and related subjects as part of Electrical and Mechanical degree courses. In addition to our relationship with the United Kingdom Atomic Energy Authority (UKAEA) the Dounreay site licensee, we have links with Rolls-Royce and the Ministry of Defence who also share the Dounreay site and with other stakeholders such as, the UK regulator (HSE/NII), the Scottish Environmental Protection Agency (SEPA), local and international contractors and we liaise with the newly formed Nuclear Decommissioning Authority (NDA), who provide some sponsorship and support. We possess our own equipment and laboratories for taking and analysing soil samples and for conducting environmental surveys. Recently we commissioned an aerial survey of contamination in the locality from natural sources, other background levels such as Chernobyl fall out and any local activity from Dounreay.
Decommissioning Considerations at a Time of Nuclear Renaissance
Authors: Devgun Jas S.
Published: June 2009

At a time of renaissance in the nuclear power industry, when it is estimated that anywhere between 60 to 130 new power reactors may be built worldwide over the next 15 years, why should we focus on decommissioning? Yet it is precisely the time to examine what decommissioning considerations should be taken into account as the industry proceeds with developing final designs for new reactors and the construction on the new build begins. One of the lessons learned from decommissioning of existing reactors has been that decommissioning was not given much thought when these reactors were designed three or four decades ago. Even though decommissioning may be sixty years down the road from the time they go on line, eventually all reactors will be decommissioned. It is only prudent that new designs be optimized for eventual decommissioning, along with the other major considerations. The overall objective in this regard is that when the time comes for decommissioning, it can be completed in shorter time frames, with minimum generation of radioactive waste, and with better radiological safety. This will ensure that the tail end costs of the power reactors are manageable and that the public confidence in the nuclear power is sustained through the renaissance and beyond.
Proven Concepts for LLW-Treatment of Large Components for Free-Release and Recycling
Authors: Bergstro¨m Lena , Lindberg Maria , Lindstro¨m Anders , Wirendal Bo , Lorenzen Joachim
Published: June 2009

This paper describes Studsvik’s technical concept of LLW-treatment of large, retired components from nuclear installations in operation or in decommissioning. Many turbines, heat exchangers and other LLW components have been treated in Studsvik during the last 20 years. This also includes development of techniques and tools, especially our latest experience gained under the pilot project for treatment of one full size PWR steam generator from Ringhals NPP, Sweden. The ambition of this pilot project was to minimize the waste volumes for disposal and to maximize the material recycling. Another objective, respecting ALARA, was the successful minimization of the dose exposure to the personnel. The treatment concept for large, retired components comprises the whole sequence of preparations from road and sea transports and the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into the final packages suitable for customer-related disposal. For e.g. turbine rotors with their huge number of blades the crucial moments are segmentation techniques, thus cold segmentation is a preferred method to keep focus on minimization of volumes for secondary waste. Also a variety of decontamination techniques using blasting cabinet or blasting tumbling machines keeps secondary waste production to a minimum. The technical challenge of the treatment of more complicated components like steam generators also begins with the segmentation. A first step is the separation of the steam dome in order to dock the rest of the steam generator to a specially built treatment cell. Thereafter, the decontamination of the tube bundle is performed using a remotely controlled manipulator. After decontamination is concluded the cutting of the tubes as well as of the shell is performed in the same cell with remotely controlled tools. Some of the sections of steam dome shell or turbine shafts can be cleared directly for unconditional reuse without melting after decontamination and sampling program. Experience shows that the amount of material possible for clearance for unconditional use is between 95 – 97% for conventional metallic scrap. For components like turbines, heat exchangers or steam generators the recycling ratio can vary to about 80–85% of the initial weight.
Alpha-Contaminated Solid Waste Sorting and Conditionining at Belgoprocess (Belgium): Lessons Learned From the First Three Years Operation
Authors: Cuchet J. M. , Lahaye J. P. , Luycx P. , Van Nueten E. , De Goeyse A.
Published: June 2009

The alpha-contaminated solid waste generated in Belgium results from past activities in the fuel cycle (R&D + Reprocessing and MOX fabrication pilot plants) and operation of BELGONUCLEAIRE’s MOX fuel fabrication plant. After the main steps in the management of alpha-contaminated solid waste were established, BELGONUCLEAIRE, with the support of BELGOPROCESS and ONDRAF/NIRAS, started the design and construction of the T & C and interim-storage facilities for this alpha waste. The accumulated solid alpha radwaste containing a mixture of combustible and non-combustible material must be sorted and characterized. After sorting, both the accumulated and recently-generated alpha waste will be compacted and the pellets will be embedded in a cement matrix in a 400-1 drum. The commissioning of the sorting unit which includes glove boxes was completed at BP, at the beginning of year 2005; the sorting campaign of 30-1 cans has been achieved in March 2007. The paper describes the project environment and gives a short description of the used facilities; the lessons learned from the sorting campaign and from the first T/C period, will be presented, as well.
Development of a Requirements Management System for Technical Decision-Making Processes in the Geological Disposal Project
Authors: Ueda Hiroyoshi , Ishiguro Katsuhiko , Kitayama Kazumi , Oyamada Kiyoshi , Sato Shoko
Published: June 2009

NUMO (Nuclear Waste Management Organization of Japan) has a responsibility for implementing geological disposal of vitrified HLW (High-Level radioactive Waste) in the Japanese nuclear waste management programme. Its staged siting procedure was initiated in 2002 by an open call for volunteer sites. Careful management strategy and methodology for the technical decision-making at every milestone are required to prepare for the volunteer site application and the site investigation stages after that. The formal Requirement Management System (RMS) is planned to support the computerized implementation of the specific management methodology, termed the NUMO Structured Approach (NSA). This planned RMS will help for comprehensive management of the decision-making processes in the geological disposal project, change management towards the anticipated project deviations, efficient project driving such as well-programmed R&D etc. and structured record-keeping regarding the past decisions, which leads to soundness of the project in terms of the long-term continuity. The system should have handling/management functions for the database including the decisions/requirements in the project in consideration, their associated information and the structures composed of them in every decision-making process. The information relating to the premises, boundary conditions and time plan of the project should also be prepared in the system. Effective user interface and efficient operation on the in-house network are necessary. As a living system for the long-term formal use, flexibility to updating is indispensable. In advance of the formal system development, two-year activity to develop the preliminary RMS was already started. The purpose of this preliminary system is to template the decision/requirement structure, prototype the decision making management and thus show the feasibility of the innovative RMS. The paper describes the current status of the development, focusing on the initial stage including work analysis/modeling and the system conceptualization.
Determination of Optimum Data Points for Scaling Factor Determination
Authors: Fernandez Michael Dennis T. , Lee Sang Chul , Lee Kun Jai
Published: June 2009

Scaling factors are calculated based on a database from radiochemical analyses of representative waste samples. Several data points are needed to derive a reliable scaling factor. The more the number of data points, the better is the correlation, but more costly because of number of needed radiochemical analyses. Therefore, optimization of data points should be considered to minimize the cost without compromising reliability and prediction of the scaling factor. Scaling factors for Ni-63, Sr-90, and C-14 were calculated using Co-60 and Cs-137 as Key nuclides based on the published data in EPRI-4037. Correlation coefficient, percent error and relative standard deviation were plotted against the number of data points used in the estimation of scaling factor. The optimum number of data points was obtained to where there was no significant improvement in the statistical uncertainties by using additional samples. Even though non-segregated (all data points) has greater effect in calculating scaling factor because of its more data points, this study showed that even segregated data points can also give good DTM characterization.
Key Issues Identified From Project TRU-2 on the Generic Co-Location Concept of Transuranic (TRU) Waste and High-Level Radioactive Waste (HLW) Repositories in Japan
Authors: Martin Andrew J. , Kuroda Shigeki , Morimoto Keiji , Kamei Gento , Shiotsuki Masao
Published: June 2009

The Federation of the Electric Power Companies of Japan (FEPC) and the Japan Atomic Energy Agency (JAEA) have been collaborating with relevant organisations to promote the safe geological disposal of transuranic (TRU) waste following the already established disposal policy for high-level radioactive waste (HLW) in Japan. A result of this intensive collaborative effort was the production of a recent progress report (TRU-2) which describes the generic R&D for TRU-waste disposal in Japan. In order to improve feasibility and reduce costs and the burden on siting, the concept of co-locating TRU-waste and HLW repositories in a single complex was assessed in detail and compared with the results from several other countries that have also looked at co-location disposal. Heat from HLW, high pH plume(s) from the large amounts of cementitious materials used in the engineered barrier system (EBS) of TRU waste, and nitrates and organic materials in certain types of TRU waste were identified as critical reciprocal influences that might degrade the performance of the TRU/HLW co-location disposal system over the long-term. It was shown that these reciprocal influences could be avoided by establishing a separation distance between the two repositories of approximately 300 meters.
The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research
Authors: Lemmens Karel , Cachoir Christelle , Valcke Elie , Ferrand Karine , Aertsens Marc , Mennecart Thierry
Published: June 2009

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.
Abrasive Blasting: A Technique for the Industrial Decontamination of Metal Components and Concrete Blocks From Decommissioning to Unconditional Release Levels
Authors: Gilis R. , Lewandowski P. , Ooms B. , Reusen N. , Van Laer W. , Walthe´ry R.
Published: June 2009

When decommissioning nuclear installations, large quantities of metal components are produced as well as significant amounts of other radioactive materials, which mostly show low surface contamination. Having been used or having been brought for a while in a controlled area marks them as ‘suspected material’. In view of the very high costs for radioactive waste processing and disposal, alternatives have been considered, and much effort has gone to recycling through decontamination, melting and unconditional release of metals. In a broader context, recycling of materials can considered to be a first order ecological priority in order to limit the quantities of radioactive wastes for final disposal and to reduce the technical and economic problems involved with the management of radioactive wastes. It will help as well to make economic use of primary material and to conserve natural resources of basic material for future generations. In a demonstration programme, Belgoprocess has shown that it is economically interesting to decontaminate metal components to unconditional release levels using dry abrasive blasting techniques, the unit cost for decontamination being only 30% of the global cost for radioactive waste treatment, conditioning, storage and disposal. As a result, an industrial dry abrasive blasting unit was installed in the Belgoprocess central decontamination infrastructure. At the end of December 2006, more than 1,128 Mg of contaminated metal has been treated as well as 313 Mg of concrete blocks. The paper gives an overview of the experience relating to the decontamination of metal material and concrete blocks at the decommissioning of the Eurochemic reprocessing plant in Dessel, Belgium as well from the decontamination of concrete containers by abrasive blasting.
Flexible Process Options for the Immobilisation of Residues and Wastes Containing Plutonium
Authors: Stewart M. W. A. , Moricca S. A. , Begg B. D. , Day R. A. , Scales C. R. , Maddrell E. R. , Eilbeck A. B.
Published: June 2009

Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and sociopolitical challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residues resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes.
Three Dimensional Reconstruction of Activity Profiles in 220 Liters Radioactive Waste Packages Containing Supercompacted 100 Liters Drums
Authors: van Velzen L. P. M. , Maes J.
Published: June 2009

The 3DRedact project’s main objective is the development of a non-destructive assay (NDA) system that can replace emission computer tomography (ECT) and transmission computer tomography (TCT) for the routine characterization of decayed radioactive waste 220 liters drums. The existing fast NDA scan system has been extended with a transmission system that fulfils the requirements of fast scan measurements. The design parameters and engineering are described. As a consequence of this extension the analyze program HOLIS had to be updated, so that HOLIS can make full advantage of the transmission data generated by the analysis of a 220 liters waste drum, containing different super compacted drums. The achievements of the new HOLIS version are presented. As a first assessment, based on the presented tests results, the accuracy of the calculated coordinates of hotspots can be assessed for all coordinates ± 1 cm and for the activity of the hot-spot ± 5%. These accuracies are within the predefined requirements e.g. coordinates uncertainty ± 2 cm and activity less than 10%. Further, additional safety systems have been installed to improve a healthy and save working environment.
Building Relationships With Foreign Governments in Support of Threat Reduction
Authors: Cajigal George L.
Published: June 2009

The effectiveness of any multinational effort in support of threat reduction depends on the relationship developed between the nation receiving the assistance and the donor nations. The effectiveness of this relationship must be based on a solid legal and cooperative framework that establishes the ground rules for the interaction between all parties involved. The author proposes in this paper to outline major considerations by the donor nation and the nation receiving the assistance as they establish an effective approach to threat-reduction efforts. The legal framework needs to be founded on a well-developed, country-to-country agreement that establishes general ground rules and officially recognizes the collaborative effort in an internationally binding document between signatories. This document normally addresses such issues as liability, tax exemptions, import duties, contracts, applicable environmental regulations, etc. Also, of utmost importance is the establishment of a collaborative framework. The basis for such a collaborative framework must be the buy-in into a common objective, the willingness to share concerns and work toward resolutions, and continuous communications. Only when a proper legal framework and a collaborative approach are established can effective relationships be built to enhance threat-reduction efforts.
The Nuclear Fuel Services, Inc. Program to Support Disposition of Enriched Uranium-Bearing Materials
Authors: Schutt Stephen M. , Jacob Norman P.
Published: June 2009

The disposition of surplus nuclear materials has become one of the most pressing issues of our time [1, 2]. Numerous agencies have invoked programs with the purpose of removing such materials from various international venues and dispositioning these materials in a manner that achieves non-proliferability. This paper describes the Nuclear Fuel Services, Inc (NFS) Nuclear Material Disposition Program, which to date has focused on a variety of Special Nuclear Material (SNM), in particular uranium of various enrichments. The major components of this program are discussed, with emphasis on recycle and return of material to the nuclear fuel cycle.
Management of Radioactive Waste in Belgium: ONDRAF/NIRAS and Belgoprocess as Major Actors of the Waste Acceptance System
Authors: Van Zaelen Gunter , Verheyen Annick
Published: June 2009

The management of radioactive waste in Belgium is undertaken by the national agency for radioactive waste and enriched fissile materials, ONDRAF/NIRAS, and its industrial partner Belgoprocess. ONDRAF/NIRAS has set up a management system designed to guarantee that the general public and the environment are protected against the potential hazards arising from radioactive waste. Belgoprocess is a private company, founded in 1984 and located in Dessel, Belgium. It is a subsidiary of ONDRAF/NIRAS and its activities focus on the safe processing and storage of radioactive waste. The management system of ONDRAF/NIRAS includes two aspects: a) an integrated system and b) an acceptance system. The integrated system covers all aspects of management ranging from the origin of waste to its transport, processing, interim storage and long-term management. The safety of radioactive waste management not only depends on the quality of the design and construction of the processing, temporary storage or disposal infrastructure, but also on the quality of the waste accepted by ONDRAF/NIRAS. In order to be managed safely, both in the short and the long term, the waste transferred to ONDRAF/NIRAS must meet certain specific requirements. To that end, ONDRAF/NIRAS has developed an acceptance system.
Application of Total Uncertainty Theory in Radioactive Waste Disposal Facilities Safety Assessment
Authors: de Lemos Francisco Luiz , Ross Timothy , Sullivan Terry
Published: June 2009

Safety assessment requires the interaction of a large number of disciplines to model the environmental phenomena necessary to evaluate the safety of the disposal system. In this complex process, the identification and quantification of both types of uncertainties, random and epistemic, plays a very important role for confidence building. In this work an application of the concept of total uncertainty to radioactive waste disposal facilities safety assessment is proposed. By combining both types of uncertainty, aleatoric and epistemic, in the same framework, this approach ultimately aims to assess the confidence one can pose in the safety-assessment decisions.
Processing of Historic High Radioactive Waste Coming From Nuclear Applications
Authors: van Velzen L. P. M. , de Vos R. M. , Ijpelaan R. , van Tongeren R. , Roobol L. P.
Published: June 2009

At ECN-NRG irradiations of materials have been performed with the aid of the High Flux Reactor at the site for investigations of their properties under different conditions as well for nuclear isotope productions since 1967 e.g. molybdenum. The high radioactive waste (HRW) coming from these nuclear applications are stored since the start in an interim storage facility located at the site. Due to the site license the HRW has to be transported to COVRA. Therefore a project has been set-up to transport all the HRW to COVRA. However, COVRA accepts a limited number of HLW containers among the CASTOR® MTR-2 container and thus all temporary stored drums have to be over packed. As the existing infra structure at the site is not suited a new facility has to be build. This also creates the opportunity to minimize, by separation of the HRW in low- and intermediate level waste, the amount of waste that has to be classified as HLW. The applied methodology, design and specifications of the HRW-ILW non-destructive assay characterization and separation system will be described.
The Integrated Project NF-PRO: Recent Advances in European Research on the Near-Field System
Authors: Sneyers Alain , Grambow Bernd , Herna´n Pedro , Alheid Hans-Joachim , Aranyossy Jean-Franc¸ois , Johnson Lawrence
Published: June 2009

The Integrated Project NF-PRO (Sixth Framework Programme by the European Commission) investigates key-processes in the near-field of a geological repository for the disposal of high-level vitrified waste and spent fuel. The paper discusses the project scope and content and gives a summary overview of advances made by NF-PRO.
Modeling of Radionuclide Releases From the Geological Repository for RBMK-1500 Spent Nuclear Fuel in Crystalline Rocks in Lithuania
Authors: Poskas Povilas , Brazauskaite Asta
Published: June 2009

During 2002–2005 the assessment of possibilities for disposal of spent nuclear fuel (SNF) in Lithuania was performed with support of Swedish experts. Potential geological formations for disposal of SNF were selected, disposal concept was developed, reference disposal site was defined and preliminary generic safety assessment was performed. Performing safety assessment the analysis of radionuclides migration from the repository as well as their impact to human and environment were also very important issues. In this paper results on the analysis of the radionuclide releases from the reference geological repository site for RBMK-1500 SNF in crystalline rocks in Lithuania are presented. For radionuclide migration in the near field region of the repository integrated finite difference method and the concept of compartments were used. For radionuclide migration in the far field the discrete channel network concept was used. The assessment of radionuclide migration in the near and far field region was performed using computer codes AMRER4.5 [1] and CHAN3D [2]. The results of analysis show that most of safety relevant radionuclides of RBMK-1500 SNF are effectively retarded in the near field region. The exposure due to possible release of the radionuclides from the crystalline rocks would be dominated by 129I firstly while after app. 250 thousand years 226Ra is dominating already.
A Heating Experiment in the Argillites in the Meuse/Haute-Marne Underground Research Laboratory
Authors: Wileveau Yannick , Su Kun , Ghoreychi Mehdi
Published: June 2009

A heating experiment named TER is being conducted with the objectives to identify the thermal properties, as well as to enhance the knowledge on THM processes in the Callovo-Oxfordian clay at the Meuse/Haute Marne Underground Research Laboratory (France). The in situ experiment has being switched on from early 2006. The heater, 3 m length, is designed to inject the power in the undisturbed zone at 6 m from the gallery wall. A heater packer is inflated in a metallic tubing. During the experiment, numerous sensors are emplaced in the surrounding rock and are experienced to monitor the evolution in temperature, pore-water pressure and deformation. The models and numerical codes applied should be validated by comparing the modeling results with the measurements. In parallel, some lab testing have been achieved in order to compare the results given with two different scales (cm up to meter scale). In this paper, we present a general description of the TER experiment with installation of the heater equipment and the surrounding instrumentation. Details of the in situ measurements of temperature, pore-pressure and strain evolutions are given for the several heating and cooling phases. The thermal conductivity and some predominant parameters in THM processes (as linear thermal expansion coefficient and permeability) will be discussed.
Transparent Tools for Uncertainty Analysis in High Level Waste Disposal Facilities Safety
Authors: de Lemos Francisco Luiz , Helmuth Karl-Heinz , Sullivan Terry
Published: June 2009

In this paper some results of a further development of a technical cooperation project, initiated in 2004, between the CDTN/CNEN, The Brazilian National Nuclear Energy Commission, and the STUK, The Finnish Radiation and Nuclear Safety Authority, are presented. The objective of this project is to study applications of fuzzy logic, and artificial intelligence methods, on uncertainty analysis of high level waste disposal facilities safety assessment. Uncertainty analysis is an essential part of the study of the complex interactions of the features, events and processes, which will affect the performance of the HLW disposal system over the thousands of years in the future. Very often the development of conceptual and computational models requires simplifications and selection of over conservative parameters that can lead to unrealistic results. These results can mask the existing uncertainties which, consequently, can be an obstacle to a better understanding of the natural processes. A correct evaluation of uncertainties and their rule on data interpretation is an important step for the improvement of the confidence in the calculations and public acceptance. This study focuses on dissolution (source), solubility and sorption (sink) as key processes for determination of release and migration of radionuclides. These factors are affected by a number of parameters that characterize the near and far fields such as pH; temperature; redox conditions; and other groundwater properties. On the other hand, these parameters are also consequence of other processes and conditions such as water rock interaction; pH and redox buffering. Fuzzy logic tools have been proved to be suited for dealing with interpretation of complex, and some times conflicting, data. For example, although some parameters, such as pH and carbonate, are treated as independent, they have influence in each other and on the solubility. It is used the technique of fuzzy cognitive mapping is used for analysis of effects of variations on one parameter on the others in a system. This technique uses the concept of fuzzy sets to represent the “quality” of the relation between parameters rather then deterministic numbers.
On-Site Characterisation, Re-Packaging and Transport of Luminised, Former Aircraft Escape Hatches
Authors: Reeves Nigel , Grundy Colette , Sutherland Alex , John Gordon , Shaw Cath , Green Lisa , Beadle Ian
Published: June 2009

AMEC NNC, under contract to the UK Environment Agency, recovered a number of redundant aircraft hatches from an insecure location in North Wales. The Environment Agency instigated emergency action under the Radioactive Substances Act 1993 (RSA93), (1), to recover the hatches. Section 30(1) of RSA93 gives the Environment Agency powers to dispose of radioactive waste where it is unlikely the waste will be lawfully disposed of. Funding for this project was provided by the UK Government, within the Surplus Source Disposal Programme. The Environment Agency worked closely with partner regulatory organisations including the Health and Safety Executive (HSE), the Department for Transport (DfT) and the Local Authority to ensure the safe packaging, removal and transport of the material to a part-shielded store pending final disposal. The project comprised a number of technical difficulties that needed to be overcome. These included poor existing characterisation of the waste, insecure premises requiring daily lockdown, construction of a temporary containment facility with associated filtered extract and the inclement weather. AMEC NNC’s initial risk assessment identified the likelihood of high levels of loose, airborne radiological material. In order to provide adequate protection for personnel, and to prevent the spread of any radioactive contamination, the decision was made to implement radiological containment and to equip contractors with appropriate RPE (Respiratory Protective Equipment). Accurate characterisation of the radiological nature of the material was a crucial objective within the project. This was in order to correctly identify the Proper Shipping Name for consignment for transport, and to ensure that suitable transport containers were used. The packaged wastes were then transported to a secure location for temporary storage prior to final disposal. An innovative route was identified for processing of this material. Beneficial recycling and re-use within the nuclear industry was the outcome.
A Study on the Representative Sampling Survey for the Inspection of the Clearance Level for the Radioisotope Waste
Authors: Ahn Hong Joo , Kim Ju Youl , Sohn Se Chul , Jee Kwang Yong , Lee In Koo
Published: June 2009

Utilization facilities for radioisotope (RI) are increasing annually in South Korea, and the total number was 2,723, as of December 31, 2005. The inspection of a clearance level is a very important problem in order to ensure a social reliance for releasing radioactive materials to the environment. Korean regulations for such a clearance are described in Notice No. 2001–30 of the Ministry of Science and Technology (MOST) and Notice No. 2002–67 of the Ministry of Commerce, Industry and Energy (MOCIE). Most unsealed sources in RI waste drums at a storage facility are low-level beta-emitters with short half-lives, so it is impossible to measure their inventories by a nondestructive analysis. Furthermore, RI wastes generated from hospital, educational and research institutes and industry have a heterogeneous, multiple, irregular, and a small quantity of a waste stream. This study addresses a representative (master) sampling survey and analysis plan for RI wastes because a complete enumeration of waste drums is impossible and not desirable in terms of a cost and efficiency. The existing approaches to a representative sampling include a judgmental, simple random, stratified random, systematic grid, systematic random, composite, and adaptive sampling. A representative sampling plan may combine two or more of the above sampling approaches depending on the type and distribution of a waste stream. Stratified random sampling (constrained randomization) is proven to be adequate for a sampling design of a RI waste regarding a half-life, surface dose, undertaking time to a storage facility, and type of waste. The developed sampling protocol includes estimating the number of drums within a waste stream, estimating the number of samples, and a confirmation of the required number of samples. The statistical process control for a quality assurance plan includes control charts and an upper control limit (UCL) of 95% to determine whether a clearance level is met or not.
Derivation of the Korean Radwaste Scaling Factor
Authors: Jee Kwang Yong , Ahn Hong Joo , Sohn Se Chul , Han Su Ho , Choi Ki Seop
Published: June 2009

The concentrations of several radionuclides in low and intermediate level radioactive waste (LILW) drums have to be determined before shipping to disposal facilities. A notice, by the Ministry of Science and Technology (MOST) of the Korean Government, related to the disposal of LILW drums came into effect at the beginning of 2005, with regards to a radionuclide regulation inside a waste drum. MOST allows for an indirect radionuclide assay using a scaling factor to measure the inventories due to the difficulty of nondestructively measuring the essential α and β-emitting nuclides inside a drum. That is, a scaling factor calculated through a correlation of the α or β-emitting nuclide (DTM, Difficult-To-Measure) with a γ-emitting nuclide (ETM, Easy-To-Measure) which has systematically similar properties with DTM nuclides. In this study, radioactive wastes, such as spent resin and dry active waste which were generated at different sites of a PWR and a site of a PHWR type Korean NPP, were partially sampled and analyzed for regulated radionuclides by using radiochemical methods. According to a reactor type and a waste form, the analysis results of each radionuclide were classified. Korean radwaste scaling factor was derived from database of radionuclide concentrations.
Lessons Learned From the Decommissioning Process Affected by an Accident During Operation: The Case of A1 NPP in Slovakia
Authors: Daniska Vladimir , Timulak Jan , Pekar Anton , Niznansky Vojtech , Konecny Ladislav
Published: June 2009

Decommissioning of NPP’s with standard shutdown is currently well known process. The A1 NPP in Slovakia was shutdown in 1977 after the accident in the core which caused the damage of the fuel and contamination of systems. Long period from 1977 to 2008 was needed to manage issues resulting from affecting the systems and structures of A1 NPP and the environment by the accident. Management of the damaged spent fuel, decontamination of the primary circuit and other processes generated large amounts of alpha bearing waste, mostly liquid, having sludge phases with specific physical-chemical and radiological properties. Up to 1994, the approach for eliminating the consequences of the accident was based on safety priorities. The systematic approach, which includes also the rehabilitation of the affected environment, was implemented in the period 1994–2008. The process includes also establishing of the decommissioning infrastructure, legislative and funding system with the aim to implement the standard decommissioning procedures after 2008. However, the specific aspects, especially the level and radio-nuclide composition of contamination of systems and structures will remain. For final decommissioning 2008–2033, the approach was selected which foresees four licensed phases. This approach enables proper planning and performing of individual decommissioning phases.
Flow and Radionuclide Transport From Rock to Surface Systems: Characterization and Modelling of Potential Repository Sites in Sweden
Authors: Werner Kent , Bosson Emma , Berglund Sten
Published: June 2009

The safety assessments of potential geological repositories for spent nuclear fuel in Sweden are supported by modelling of groundwater flow in rock, to predict locations (exit points) where radionuclides from the deep repository may enter land, surface waters and associated ecosystems above the rock. This modelling includes detailed rock descriptions, but simplifies the upper part of the flow domain, including representations of meteorological processes and interactions with hydrological objects at the surface. Using the Laxemar candidate site as example, this paper investigates some potentially important consequences of these simplifications. Specifically, it compares particle tracking results obtained by a deep-rock groundwater flow model (CONNECTFLOW) and by MIKE SHE-MIKE 11, which contains detailed descriptions of near-surface/surface water flow. Overall, the models predict similar exit point patterns, occurring as clusters along streams in valleys, at a lake, and in sea bays. However, on a detailed level there are some prediction differences, which may be of importance for biosphere-focused safety assessments. CONNECTFLOW essentially predicts flow paths through the repository that follow fractures and deformation zones, outcropping in valleys. In comparison, MIKE SHE-MIKE 11 provides more detailed information on near-surface water flow paths, including the associated exit points and inputs to assessments of radionuclide retention.
Effects of Lump Characteristics on Plutonium Self Absorption Correction Methods
Authors: Curtis D. C. , Wormald M. R. , Croft S.
Published: June 2009

An evaluation study has been undertaken to assess the robustness of several published Pu self-absorption correction methods against variation in size, shape, density etc. for use in the gamma assay of nuclear waste. The correction methods studied are a numerical plutonium self absorption correction (PuSAC) technique, the Fleissner 2-line, Fleissner 3-line and Infinite Energy Extrapolation methods with both linear and polynomial extrapolation to 1/E = 0. The performance of these methods has been compared for a limited set of measured encapsulated PuO2 sources plus a range of modelled unencapsulated Pu lumps. An indication of the magnitude of the uncertainties of the numerical PuSAC method has been determined for cases of blind assays where the Pu material, shape and distribution are unknown with the aim of ultimately applying it to real waste. The importance of the range of Pu lumps used in the baseline modelled dataset has been examined. Data are presented to illustrate how the uncertainties in the method are affected by the shape, composition, density, number and mass distribution of Pu particles in a sample for a given modelled base dataset.
ReCLAIM v2.0: Comparison of Calculated Doses With Other Assessment Tools When Emulating Contaminated Land Scenarios
Authors: Willans Mark , Galais Nathalie , Lennon Chris , Trivedi Divyesh
Published: June 2009

ReCLAIM v2.0 is a software tool designed to calculate doses to exposure groups from radioactively contaminated land including UK Nuclear Licensed Sites. When using tools to undertake contaminated land assessments, it is important to understand the functionality of the tool and how the tool should be best used to undertake an assessment. This work describes the results from inter-comparison of ReCLAIM v2.0 with two other radioactively contaminated land tools (Conland and RCLEA). For the majority of cases there was little difference between ReCLAIM v2.0 and the other tools. In all cases where there were significant differences, the cause of the variation could be explained and quantified. In the majority of these cases the main exposure pathway was external irradiation. In these cases, variations in the calculated doses were due to the differing ways in which underpinning shielding calculations were undertaken to calculate external radiation dose coefficients. The effect was most noticeable for radionuclides emitting high energy penetrating radiation such as Co-60, Cs-134 and Cs+137. For these cases, there was a greater difference of up to around 60% in the dose comparisons between ReCLAIM v2.0 and the alternative tools, with ReCLAIM v2.0 being more conservative. These results demonstrate that where the algorithms and parameter values are similar, different tools will produce similar results. The user of such tools should be aware of cases where different results may be produced in order that an appropriate level of confidence can be assigned when making decisions based upon the results.
Open-Air Demolition of Facilities Highly Contaminated With Plutonium
Authors: Lloyd E. R. , Zinsli L. C. , Lackey M. B. , Stevens J. M.
Published: June 2009

The demolition of highly contaminated plutonium buildings usually is a long and expensive process that involves decontaminating the building to near free-release standards and then using conventional methods to remove the structure. It doesn’t, however, have to be that way. Fluor has torn down buildings highly contaminated with plutonium without excessive decontamination. By removing the select source term and fixing the remaining contamination on the walls, ceilings, floors, and equipment surfaces; open-air demolition is not only feasible, but it can be done cheaper, better (safer), and faster. Open-air demolition techniques were used to demolish two highly contaminated buildings to slab-on-grade. These facilities on the Department of Energy’s Hanford Site were located in, or very near, compounds of operating nuclear facilities that housed hundreds of people working on a daily basis. To keep the facilities operating and the personnel safe, the projects had to be creative in demolishing the structures. Several key techniques were used to control contamination and keep it within the confines of the demolition area: spraying fixatives before demolition; applying fixative and misting with a fine spray of water as the buildings were being taken down; and demolishing the buildings in a controlled and methodical manner. In addition, detailed air-dispersion modeling was done to establish necessary building and meteorological conditions and to confirm the adequacy of the proposed methods. Both demolition projects were accomplished without any spread of contamination outside the modest buffer areas established for contamination control. Furthermore, personnel exposure to radiological and physical hazards was significantly reduced by using heavy equipment rather than “hands on” techniques.
An Optimized Strategy for the Management of Spent Steam Generators
Authors: Cantrel Eric , Denissen Luc , Davain Henri , Leveau Jean-Phillipe , Lauwers Johan , Gillet Thierry
Published: June 2009

The decommissioning of the BR3 (Belgian Reactor 3) approaches its final phase. The electro-mechanical dismantling is almost completed and the program related to the decontamination of the building structures has been initiated. The issue of the evacuation of the primary circuit large components, and more specifically of the Steam Generator (SG), has been dealt successfully, applying innovative technologies to lead to remarkable results in terms of waste volume minimization and occupational radiation exposure. The strategy applied for the evacuation of the BR3 SG resulted from the elaboration and comparison of the following scenarios: • Closed loop chemical decontamination prior to dismantling, cutting and unconditional release or release after melting, • Cutting of the components without decontamination and evacuation of the materials in their respective waste categories, • Cutting, decontamination of the SG secondary side and evacuation of the full SG primary side to the melting facility for recycling. While the availability of the in-house developed MEDOC® process made the clearance of the SG bundle technically feasible, nuclear safety requirements and financial aspects were also in favour of the closed loop decontamination: minimization of contamination spreading and staff exposure during all subsequent manipulations, minimization of radwaste costs. For the segmentation of this component, different techniques have been considered: • An abrasive water jet (AWJ) cutting tool, • A prototype diamond wire developed for this application. The diamond wire allowed to cut in a single pass both the carbon steel shell and the stainless steel tube bundle. While the implementation of the diamond wire saw is rather simple, working conditions must be optimised to limit wearing of the wire and secondary waste production. Existing experience can be extrapolated to different legal frameworks in order to propose a financially and technically optimised “all-in” strategy for the management of “spent” SG.