A final cap will be emplaced over the disposed waste as part of the closure engineering for the UK’s Low Level Waste Repository (LLWR). Additional profiling material will be required above the waste to obtain the required landform. Consideration has been given to the potential opportunity to reuse Low Specific Activity Material (LSAM, defined as up to 200 Bq g −1 ) imported from other sites as a component of the necessary profiling material for the final repository cap. Justification of such a strategy would ultimately require a demonstration that the solution is optimal with respect to other options for the long-term management of such materials. The proposal is currently at the initial evaluation stage and seeks to establish how LSAM reuse within the cap could be achieved within the framework of an optimised safety case for the LLWR, should such a management approach be pursued. The key considerations include the following: The LSAM must provide the same engineering function as the remainder of the profiling material. The cap design must ensure efficient leachate collection, drainage and control for Low Level Waste (LLW) (and, by extension, LSAM) during the Period of Authorisation. In the longer term the engineering design must passively direct any accumulating waters preferentially away from surface water systems. An initial design has been developed that would allow the placement of around 220,000m 3 of LSAM. The potential impact of the proposal has been assessed against the current Environmental Safety Case.

Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of ‘what if’ scenarios until one is sure that a decision is the right one. In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D&D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task. The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks. The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream. In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner, taking account of the detailed waste stream and specific requirements.

NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated wasteforms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable wasteforms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL’s Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO 2 as a surrogate for PuO 2 . This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of immobilising the identified waste and residues inventory within 3 years. During that period it will also be used to demonstrate the potential of this technology to deliver the immobilisation of a proportion of the UK plutonium stockpile that may not be suitable for use as MOx fuel should that decision be taken.

The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination.

Knowledge of the radiological state of processes and equipment of a nuclear facility is essential to supervise a wide variety of sensitive tasks: building of intervention scenarios in order to optimize maintenance or dismantling operations, optimization of waste categorization, monitoring the effectiveness of decontamination processes, monitoring of nuclear facility decommissioning, etc. In order to meet the diversity of the issues involved, the CEA has developed in situ radiological characterization methods and techniques to acquire reliable radiological data. The data gathered is necessary to build robust radiological models which can be used as input data for dismantling studies. Over the last 30 years, the main nuclear measurement techniques, such as gamma imaging and gamma spectrometry, have been widely deployed by the CEA on many facilities under dismantling and more recently, on the Phénix nuclear power plant. Phénix was a small-scale prototype of a sodium-cooled fast breeder reactor, located at the Marcoule nuclear site. These techniques have been implemented on this reactor in order to meet the increased need for radiological knowledge to prepare for future dismantling operations following its final shutdown in 2009. This paper will focus on the description of three radiological characterization methods which take advantage of advanced nuclear measurement techniques. For each method, an example of a specific application on the Phénix reactor will be presented. Firstly, the so-called “gamma scanning” method will be explained. The objective of this method is to determine the activity profile of equipment based on collimated gamma spectrometry measurements with compact probes like CdZnTe. This method was applied to a neutron shielding of the reactor core to estimate the 60 Co activity profile. Then the measured activities helped to validate the theoretical activities resulting from neutron activation calculations. Secondly, this paper will focus on the interest of combining different measurement techniques such as gamma imaging, gamma spectrometry and collimated/uncollimated dose rate mapping to characterize complex equipment or processes. In this case, a specific methodology was developed to define the radiological state of a shielded cell used for the processing of irradiated nuclear fuels. Finally, an isotopic characterization technique using a high purity germanium detector will be discussed. This technique was applied to a non-irradiated fertile fuel sub-assembly in order to determine the level of uranium enrichment. The processing was carried out by three types of analysis: two automated, with the MGA-U and IGA software, and one manual.

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

This paper deals with the dismantling of the Brennilis NPP plant located in the west of France (Finiste`re). This prototype NPP of Brennilis was the unique reactor of the heavy water developed in France during the 50’ and the 60’. The reactor diverged in December 1966 and the NPP was operated during 9 years from 1972 to 1981, then the permanent shutdown occurred in July 1985. In 2008, the operator and owner of the plant Electricite´ de France (EDF) commissioned the consortium Onet Technologies Grands Projets (France) and Nukem Technologies (Germany) with the dismantling of the reactor block of the NPP. The reactor block essentially contains the reactor vessel including built-in units and biological shields, the peripheral piping as well as systems for controlling the nuclear-related process. In addition to the complete dismantling, the scope of the contractual services also includes their proper handling in accordance with the applicable regulation: safety requirements, waste management, radioprotection optimization and management. The central element of the plant is the reactor pressure vessel filled with heavy water. Each of the 216 horizontal fuel element channels made of zircaloy is at each side connected to a pipe which directs the heat transfer gas to a header mounted in the upper part of the reactor block. The control rods are introduced vertically into the reactor. It should be pointed out that due to this reactor design, the reactor pressure vessel is equipped with a complex pipe system to all sides which makes it difficult to freely access the core area of the reactor block and thus to dismantle the reactor. In this context, the axial and lateral neutron shields should be mentioned, which are situated in close proximity to the reactor as well as the biological shield which protects from ionizing radiation originating from the pressure vessel. The access to the core area is made difficult due to a high local dose rate and the extremely high constructive complexity of the prototype, the interior of which is virtually criss-crossed by complex piping. The elevated local dose rate in the area of the reactor pressure vessel makes manual work in this zone impossible (even after 30 years), so that remote dismantling techniques have to be used. Before starting dismantling, the remote devices are determined with the help of a test stand which representatively simulates the real conditions of the reactor block in respect to dimensions and material.

The French Act voted in 2006 made the choice of deep geological disposal as the reference option for the long term management of high level (HLW) and intermediate level long-lived waste. The Cige´o repository project aims at avoiding or limiting burden to future generations, which could not be achieved by the extension in time of interim storage. The reversibility as provided by the Act will maintain a liberty of choice for waste management on a duration which is comparable to new storage facility. Interim storage is required to accommodate waste as long as the repository is not available. The commissioning of the repository in 2025 will not suppress needs for interim storage. The paper describes the complementarities between existing and future interim storage facilities and the repository project: repository operational issues and planning, HLW thermal decay, support for the reversibility, etc. It shows opportunities to prepare a global optimization of waste management including the utilization at best of storage capacities and the planning of waste emplacement in the repository in such a way to facilitate operational conditions and to limit cost. Preliminary simulations of storage-disposal scenarios are presented. Thanks to an optimal use of the waste management system, provision can be made for a progressive increase of waste emplacement flow during the first operation phase of the repository. It is then possible to stabilize the industrial activity level of the repository site. An optimal utilization of interim storage can also limit the diversity of waste packages emplaced simultaneously, which facilitates the operation of the repository. 60 years minimum interim storage duration is generally required with respect to HLW thermal output. Extending this interim storage period may reduce the underground footprint of the repository. Regarding reversibility, the capability to manage waste packages potentially retrieved from the repository should be analyzed. The 2006 French Act provides for a research program on interim storage to be carried out along with the repository project development. This program has been guided by the complementarities between interim storage and the repository project. The main research issues address the longevity of storage facilities, up to 100 years, their versatility with regard to waste package types and their modularity to match future needs progressively. In parallel the dialogue between Andra and waste producers will continue to propose optimized waste management scenarios.

The UK nuclear industry has in its inventory legacy waste in the form of complex, polydisperse and “polydense” suspensions, slurries and sludges in a variety of storage and transport vessels. This waste has been difficult to characterise because of radioactivity and limited accessibility, and conditioning and disposal of the waste presents a continuing challenge. In addition, the mechanisms by which very dense particles are transported in pipes are not well understood. Our objectives are to investigate the effect of mono- and bidisperse suspensions with a range of particle sizes and densities on the turbulence characteristics, transport and settling behaviour of slurries that are chosen to be analogues of those found on nuclear sites. Two versatile slurry pipe-flow loops of different diameters have been commissioned which can be operated over a large range of Reynolds numbers and are amenable to ultrasonic measurement methods. Details of the flow loops are presented, including optimisation studies. Results are presented for a variety of particle characterisation studies that have been performed on the particle species that form the suspensions, along with mean and RMS (root mean square) velocity profiles over a range of Reynolds number and particle concentration. In particular, the effect of particle concentration on the formation of settled beds, and mean flow velocity and turbulence characteristics has been investigated.

The UK’s Low Level Waste Repository Ltd submitted an Environmental Safety Case (ESC) for the disposal of low-level waste to the Environment Agency on the 1 st of May 2011. The ESC is a major submission that will decide the future use of the Repository and has major implications for the success of the UK’s LLW Strategy and decommissioning programme. This paper provides an overview of the work that has been carried out to support the submission. Key aspects of this ESC include: • detailed investigations of existing disposals, based on careful examination of existing records and other investigations, including interviews with former operational staff; • analysis of uncertainties in future disposals; • modelling of the biogeochemical evolution of the disposal system, which provides understanding of the evolution of pH, Eh and gas generation and thence underpinning for radionuclide releases in groundwater and gas; • development of a 3-D groundwater flow model, calibrated against observed heads and with a detailed representation of the engineered features; • analysis of coastal erosion and its impacts; • a major focus on optimisation based on detailed technical studies; • a conclusion that existing disposals do not require remediation; • the choice of a concrete vault design with permeable side walls designed to avoid bathtubbing after the end of management control; • a comprehensive set of assessment calculations, including thorough analysis of uncertainties, which demonstrate consistency with the Environment Agency’s risk and dose guidance levels; • revision of the LLWR’s WAC, based in part on the use of the ‘sum of fractions’ approach; • the use of a safety case document structure that emphasises key safety arguments in a Level 1 document and provides supporting evidence in a series of Level 2 documents; • the provision of a Level 2 document that describes in detail how each aspect of the regulatory guidance has been addressed. In the future, the 2011 ESC will be maintained using a formal system of change control. It will be used as a tool for decision making concerning the future development of the LLWR and waste acceptance.

The presented methodological study illustrates a geostatistical approach suitable for radiological evaluation in nuclear premises. The waste characterization is mainly focused on floor concrete surfaces. By modeling the spatial continuity of activities, geostatistics provide sound methods to estimate and map radiological activities, together with their uncertainty. The multivariate approach allows the integration of numerous surface radiation measurements in order to improve the estimation of activity levels from concrete samples. This way, a sequential and iterative investigation strategy proves to be relevant to fulfill the different evaluation objectives. Waste characterization is performed on risk maps rather than on direct interpolation maps (due to bias of the selection on kriging results). The use of several estimation supports (punctual, 1 m 2 , room) allows a relevant radiological waste categorization thanks to cost-benefit analysis according to the risk of exceeding a given activity threshold. Global results, mainly total activity, are similarly quantified to precociously lead the waste management for the dismantling and decommissioning project.

This paper deals with information on the radioactive waste cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled and segmented radioactive materials were cemented in concrete containers using tritiated water-cement mixture. The viscosity of water-cement mortar, mechanical tests of solidified mortar’s samples, change of temperature of the samples during solidification time and long time leakage of 137 Cs, 14 C, 60 Co and 3 T radionuclides was studied for different water-cement compositions with additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar’s viscosity and solidified samples mechanical stability. The role of additives — fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81°C up to 62°C. The optimal interval of water ratio in cement mortar is discussed. Radionuclides leakage tests show that the release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that additives can significantly influence on the radionuclides release processes from cemented samples. The optimization of cementation of radioactive wastes in concrete containers was performed using mechanical stability, solidification temperature, radionuclide releases and viscosity of mortar.

The Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and investigations on the siting and design of the new repository upon the French Radioactive Waste Management Agency ( Agence nationale pour la gestion des de´chets radioactifs – Andra), in order for the review of the creation-licence application to start in 2015 and, subject to its approval, the commissioning of the new repository in 2025. In late 2009, Andra submitted to the French government proposals concerning the implementation and the design of Cige´o ( Centre industriel de stockage ge´ologique ). A significant step of the project was completed with the delineation of an interest zone for the construction of the repositor’s underground facilities in 2010. This year, Andra has launched a new dialogue phase with local actors in order to clarify the implementation scenarios on the surface. The selected site will be validated after the public debate that is now scheduled for the first half of 2013. This debate will be organized by the National Public Debate Committee (Commission nationale du de´bat public ). In parallel, the State is leading the preparation of an territorial development scheme, which will be presented during the public debate. The 2009 milestone also constitutes a new step in the progressive design process of the repository. After the 1998, 2001 and 2005 iterations, which focused mainly on the long-term safety of the repository, the Dossier 2009 highlighted its operational safety, with due account of the non-typical characteristics of an underground nuclear facility. It incorporates the first results of the repository-optimisation studies, which started in 2006 and will continue in the future. The reversibility options for the repository constitute proposals in terms of added flexibility in repository management and in package-recovery levels. They orient the design of the repository in order to promote those reversibility components. They contribute to the dialogue with stakeholders in the preparation of the public debate and of the future act on the reversibility conditions of the repository. The development of the repository shall be achieved over a long period, around the century. Hence, the designer will acquire additional knowledge at every new development of the project, notably during Phase 1, which he may reuse during the following phase, in order, for instance, to optimise the project. This process is part of the approach proposed by Andra in 2009 pursuant to the reversibility principle.

Many established nuclear power programmes have learned to their dismay that waste management and disposal are not tasks that can be postponed at will if public and political acceptance is a prerequisite for progress. In fact, some programmes that recognised this back in the 1970s and 1980s moved into leading positions in repository development. This happened, for example, in Sweden and Switzerland where already in the 1970s Laws were passed specifying that safe disposal must be demonstarted before new nuclear plants could opersate. In recent years, it has become recognised that, in order to ensure that the radioactive wastes in any country are managed safely, it is necessary to have an established legislative and regulatory framework and also to create the necessary organizations for implementation and for oversight of waste management operations and facility development. Guidance on these issues is given in the Joint Convention and a number of other IAEA documents. The IAEA, and also the EC, have in addition published key overarching advisory documents for new nuclear programmes. These are useful for strategic planning but, when it comes to actual implementation projects, the advice tends to imply that all nuclear programmes, however large or small, should be pressing ahead urgently towards early operation of geological repositories. In practice, however, in small programmes there are neither economic nor technical drivers for early implementation of deep geological repositories. Constructing simpler facilities for the disposal of the larger volume of low-level wstes has higher priority. Nevertheless, in all countries political decisions have to be taken and policies set in place to ensure that geological disposal will implemented without unjustified delay. This paper distils out a set of key messages for new programmes. Amongst the most critical are the following. Even if disposal is far off, planning and organization should begin at the initiation of the programme; this can help with technical and economic optimization and (importantly) also with public and political acceptance. Important lessons can be learned from advanced programmes — but these must be adapted to allow for the different boundary conditions of new and small programmes. The key differences relate to the timescales involved, and the resources available. There is a range of waste management and waste disposal options open to new programmes. It is not necessary to choose definitive solutions at the outset; options can be kept open, but a minimum level of engagement is required for all open options.

There are many sites in the world, where Environment are still under influence of the contamination related to the Uranium production carried out in past. Author’s experience shows that lack of site characterization data, incomplete or unreliable environment monitoring studies can significantly limit quality of Safety Assessment procedures and Priority actions analyses needed for Remediation Planning. During recent decades the analytical laboratories of the many enterprises, currently being responsible for establishing the site specific environment monitoring program have been significantly improved their technical sampling and analytical capacities. However, lack of experience in the optimal site specific sampling strategy planning and also not enough experience in application of the required analytical techniques, such as modern alpha-beta radiometers, gamma and alpha spectrometry and liquid-scintillation analytical methods application for determination of U-Th series radionuclides in the environment, doesn’t allow to these laboratories to develop and conduct efficiently the monitoring programs as a basis for further Safety Assessment in decision making procedures. This paper gives some conclusions, which were gained from the experience establishing monitoring programs in Ukraine and also propose some practical steps on optimization in sampling strategy planning and analytical procedures to be applied for the area required Safety assessment and justification for its potential remediation and safe management.

Nowadays, nuclear industry is facing a crucial need in establishing radiological characterization for the appraisal and the monitoring of any remediation work. Regarding its experience in this domain, the French Alternative Energies and Atomic Energy Commission (CEA) of Fontenay-aux-Roses, established an important feedback and developed over the last 10 years a sound methodology for radiological characterization. This approach is based on several steps: - historical investigations; - assumption and confirmation of the contamination; - surface characterization; - in-depth characterization; - rehabilitation objectives; - remediation process. The amount of measures, samples and analysis is optimized for data processing using geostatistics. This approach is now used to characterize soils under facilities. The paper presents the radiological characterization of soils under a facility basement. This facility has been built after the first generation of nuclear facilities, replacing a plutonium facility which has been dismantled in 1960. The presentation details the different steps of radiological characterization from historical investigations to optimization of excavation depths, impact studies and contaminated volumes.

The “Commissariat a` l’Energie Atomique” (CEA, French Atomic Energy Commission) has set up over the last 10 years an innovative methodology aiming at characterizing radiological contaminations. The application of the latter relies on various tools such as recently developed software platform called Kartotrak which is used in the expertise vehicles with impressive detection performances (VEgAS). A Geographic Information System tailored to radiological needs constitutes the heart of the platform; it is surrounded by several modules dedicated to sampling optimization, data analysis and geostatistical modeling, real-time monitoring (Kartotrak-RT) and validation of proper clean-up operations. This paper presents the purpose and the performances of the VEgAS which provides exhaustive instruments for the radiological surface characterization of sites.

Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA [1] developed at DECOM, a.s. is described.

Decommissioning a nuclear power plant is a complex project. The project involves the coordination of several different departments and the management of changing plant conditions, programs, and regulations. As certain project Milestones are met, the evolution of such plant programs and regulations can help optimize project execution and cost. This paper will provide information about these Milestones and the plant departments and programs that change throughout a decommissioning project. The initial challenge in the decommissioning of a nuclear plant is the development of a definitive plan for such a complex project. EPRI has published several reports related to decommissioning planning. These earlier reports provided general guidance in formulating a Decommissioning Plan. This Change Management paper will draw from the experience gained in the last decade in decommissioning of nuclear plants. The paper discusses decommissioning in terms of a sequence of major Milestones. The plant programs, associated plans and actions, and staffing are discussed based upon experiences from the following power reactor facilities: Maine Yankee Atomic Power Plant, Yankee Nuclear Power Station, and the Haddam Neck Plant. Significant lessons learned from other sites are also discussed as appropriate. Planning is a crucial ingredient of successful decommissioning projects. The development of a definitive Decommissioning Plan can result in considerable project savings. The decommissioning plants in the U.S. have planned and executed their projects using different strategies based on their unique plant circumstances. However, experience has shown that similar project milestones and actions applied through all of these projects. This allows each plant to learn from the experiences of the preceding projects. As the plant transitions from an operating plant through decommissioning, the reduction and termination of defunct programs and regulations can help optimize all facets of decommissioning. This information, learned through trial in previous plants, can be incorporated into the decommissioning plan of future projects so that the benefits of optimization can be realized from the beginning of the projects. This process of the collection of information and lessons learned from plant experiences is an important function of the EPRI Decommissioning Program.

The uranium refining and conversion plant (URCP) at Ningyo-toge was constructed in 1981 for the purpose of demonstrating on refining and conversion process from yellow cake (or uranium trioxide) to uranium hexafluoride by way of uranium tetrafluoride. For 20 years, 385 tons of natural uranium hexafluoride and 336 tons of reprocessed uranium hexafluoride (approximately) was produced. There are two different type of refining processes in the URCP. One is the wet process by convertig the natural uranium and the other is the dry conversion process for the reprocessed uranium. The dismantling of the dry process facilities began in March, 2008. It was found the large amount of uranium residuals such as wet slurry and powder uranium inside the vessels and pipes. Therefore, we have to take care of the spread of the contamination during dismantling works. The basic strategy concerning plant dismantling were the optimization of the total labor costs and the minimization of the radioactive wastes generated. The dismantling procedure is shown below; i) measuring doserate by using high sensitivity surveymeters, and nuclide identification by using gamma ray spectrometry, ii) estimating uranium mass inventory, iii) planning work force distributions with radiological survey staffs, iv) deciding dismantling methods concretely, v) decontaminating schematically if required, vi) collecting detailed data of working conditions, vi) measuring and classifying contaminated materials, vii) managements of radioactive waste drum and non-contaminated equipment, viii) control for personal exposures. Almost all equipment will be decontaminated except building decontamination it by around 2013FY. In addition, the secondary wastes were also yielded. Few thousands man-days were necessary for this project. The measurement data have not showed the high environmental radiation doserate, generally less than 0.3 μ Sv/h. However, by the trace of the reprocessed uranium, the trans-uranium nuclides such as uranium-232 progenies, Th-228 and Tl-208 were observed. The accumulation of the nuclides which emit high energy gamma rays such as Tl-208 caused radiation exposure. As for the waste disposal, the determination of uranium content must be necessary. We have been now developing the uranium measuring systems with better accuracy. The further tasks imposed by our experiences are summarized the followings; i) minimization and reduction of radioactive wastes, ii) decontamination for the buildings and utilities, iii) wastes disposal. We have to work hard toward the final decommissioning.