Activated concrete represents the greatest volume of radioactive materials produced during the dismantling operations of a PWR reactor. For heavy barytes concrete, 133 Ba is the dominating γ nuclide whereas in the rebars, it is the 60 Co isotope. During the dismantling of the BR3 PWR reactor, we studied various aspects of the radioactive concrete issue: • the characterization of the activation depth and its modelization; • the efficiency of various demolition techniques and their application on real scale mock-ups; • the active testing and use of various dismantling and demolition techniques among which the remote controlled jack hammer and the diamond cutting techniques were the most extensively used. As alternative to the conditioning of the radioactive concrete as radioactive waste using the classical cement embedding strategy, we started an extensive R&D programme on the the recycling of the radioactive concrete. The basic idea is to perform a pretreatment of the radioactive concrete so that it can be re-used as aggregates for the fabrication of “radioactive grout”. This grout is then used for the conditioning of metallic radioactive waste. We demonstrated that it is technically feasible to prepare a good quality grout using heavy radioactive concrete as raw material mixed with fresh cement.

A definition of Long-Term Stewardship (LTS) is: “all activities required to protect human health and the environment from hazards remaining after cleanup is complete.” “Cleanup” in this sense may mean completion of a prescribed remedy for contaminated soil or buried waste, or it could mean entombment of a nuclear facility or placing nuclear materials in safe, long-term storage. Among the activities included in this definition are long-term monitoring and surveillance, maintenance of engineered barriers, operation and maintenance of long-term remedies (such as groundwater pump and treat operations), institutional controls (e.g., deed restrictions, land use restrictions, permanent markers, etc.), and information management (including intergenerational transfer of data on residual hazards). The magnitude of the U.S. Department of Energy’s (DOE) LTS commitments, in terms of scope, cost, and time, is beginning to be better understood. The Idaho National Engineering and Environmental Laboratory (INEEL) has been chartered to assist DOE’s Idaho Operations Office and the DOE Headquarters Office of Long-Term Stewardship in: 1. planning and management of the National Long-Term Stewardship Program, 2. ensuring the effective transition of sites from cleanup to long-term stewardship, 3. ensuring safe and effective execution of long-term stewardship operations (in conjunction with the DOE Grand Junction Project Office), and 4. developing and implementing improvements to long-term stewardship operations and decision making through advances in science and technology (S&T). An initial step in determining how advances in S&T can be applied in the LTS program is to identify LTS S&T needs. These are needs which, if advances in scientific understanding can be made or technologies developed to address the needs, may result in reduced risk, cost, or uncertainty of LTS activities, or improved reliability of LTS measures. After LTS S&T needs are identified, DOE will coordinate and manage a research and development program to address needs which aren’t already being addressed through current projects. In Fiscal Year (FY) 2000 (October 1999 through September 2000), we completed an Initial Needs Assessment and Technology Baseline Inventory 2000 report (available on the Internet at http://emi-web.inel.gov/lts), and prepared a Conceptual Framework for a Science and Technology Roadmap . The needs were analyzed, sorted, and placed into categories where they seemed to logically group. The greatest number of needs identified were related to monitoring and surveillance. In FY 2001 (October 2000 through September 2001), we will initiate development of a LTS Roadmap. This Roadmap will lay out a plan for prioritizing and funding research and technology development that has the greatest potential for impacting cost and reliability of LTS actions.

The Georgia Tech Research Reactor (GTRR) is a 5-megawatt (MW) heavy-water-cooled nuclear reactor located on the Georgia Institute of Technology (Georgia Tech) campus in downtown Atlanta, Georgia. On July 1, 1997, Georgia Tech administration notified the U.S. Nuclear Regulatory Commission (NRC) of their intent to decommission the GTRR. In the summer of 1999, the NRC issued a license amendment to decommission the GTRR in accordance with NRC’s Regulatory Guide 1.86 . In the spring of 1999, Georgia Tech and the State of Georgia contracted CH2M HILL to serve as the Executive Engineer to manage the decommissioning project. Later in the summer of 1999, the IT Corporation was selected as the Decommissioning Contractor. The Decommissioning Contractor began the dismantlement process at the Georgia Tech site in November, 1999. By February, 2000, reactor support systems such as the primary and secondary cooling water systems, and the bismuth cooling system were removed and packaged for off-site disposal. Reactor internals were removed in April, 2000. Removal of the bioshield occurred from May through November, 2000. Throughout January, 2001, various concrete structures, including the Spent Fuel Storage Hole, were decontaminated. Dismantlement and decontamination activities were completed by April, 2001. The Final Survey Report to the NRC is planned to be submitted to the NRC December, 2001, 2001. Final license termination by the NRC is anticipated in the spring of 2002.

This paper aims to give an overview of the current status of the French R&D program devoted to the understanding of the long term evolution of spent nuclear fuel in conditions of long term storage (100’s years) and deep geological disposal (10’000’s years). This programs aims to get the scientific and technical data allowing to answer the very generic operational questions which are related to the designing of storage or disposal facilities and which are addressed to the R&D. Main results but also scientific issues are presented and discussed in order to enlighten the expected long term evolution scenario for spent fuel in storage and disposal.

The prediction of the source term of actinides, fission and activation products is of major importance in numerous nuclear fields. An experimental programme of radiochemistry (called ARIANE) was carried out on MOX and UO 2 fuels irradiated in PWR and BWR conditions, reaching burnups of 35 up to 55 GWd/tM. Besides the providing of a large set of experimental data on the irradiated fuel inventory, a second objective of the programme was to confer on these experimental results reliable and minimised uncertainties. About 50 isotopes of actinides and fission products were selected. Their contents were measured using techniques as TIMS, ICP-MS, alpha, beta and gamma spectrometry. The measurement results were crosschecked by three highly qualified laboratories and recommended values were provided after extended analysis and confrontation of the results. For most of the measured isotopes, the target uncertainties were reached and even reduced, providing a unique database for irradiated fuel characterisation and for code qualification.

Nowadays at Moscow SIA “RADON” the method of combined conditioning of solid radioactive wastes (SRW) and liquid radioactive wastes (LRW) is being developed. SRW represent radioactive silts of natural and artificial reservoirs. LRW represent water salt solutions of low and intermediate activity level. The developed method consists of mixing SRW with LRW and dry-weighed additives. From the mixture an alkaline or belite binding material is formed. The synthesized binding material can be used in process for hardening of other SRW and/or LRW. In this work it is shown that the developed method allows obtaining an accelerated hardening product, which has high strength characteristics and good water resistance. Its quality of does not concede to quality of a product received with use of a traditional binding material — portland cement. It is also shown that using this method it is possible to achieve a reduction in volume of the final product in comparison with the initial volume of SRW and LRW of 4–10. All secondary wastes formed during processing can be conditioned within the framework of the given technology.

The cost-benefit analysis of the remedial actions for the contaminated Enisey River floodplain, due to the release of radioactive materials at Mining Chemical Combine (MCC) «Krasnoyarsk-26» is carried out. The analysis was carried out for a region within the first 260 km below discharge point, where the Exposure Dose Rate (EDR) in the air ranges between 50 and 400 microR/hr and the concentration of the radionuclide reaches 25,000 Bq/kg. Both the methods of a) a cost-benefit analysis as functions of time and b) a cost justification analysis in the terms of the Action Levels (AL) have been used. Two possible situations have been considered concerning the remediation of various sites on the contaminated floodplain: 1. The spatial and depth distribution of radioactive contamination is known. In this case, it is possible to estimate the cost of removing the contaminated soil as one of the alternatives of remediation. Two contrasting examples are analyzed, which cover the entire spectrum of possibilities for removal of the contaminated soil: a) The “Gorodskoy” Island, situated inside the “Eniseysk” City, at a distance of 260 km from MCC and b) the Islands and coast of the «Kazachenskoe» settlement, at a distance 160 km from MCC, where the impacted area, the volume of contaminated soil to be removed and the number persons impacted differ by an order of magnitude. These situations were analyzed as a cost-benefit in functions of time. 2. The information is limited: only the EDR or surface contamination is known. In this case, remediation by removing the contaminated soil is impossible. In this case, remedial actions result only in limiting the people’s actions (i.e. - closure of the area). This is a typical and frequent occurrence concerning remedial actions for the Enisey River floodplain. These situations were analyzed as “generic”: the doses were analyzed using data concerning surface contamination and resulted in pessimistic estimations of the site’s specific parameters, the level of contamination and information about depth profiles of the radionuclide-specific concentration in the soil of the Enisey River floodplain. Cost justification of closure of the area is analyzed in terms of the AL. Cost-benefit as functions time and analysis in terms of the AL were used to analyze the alternatives of remedial actions: a) no action, b) removal of the contaminated soil without its stabilization, c) stabilization by the injection of silicate of sodium into the soil, followed by the excavation and removal of the firm soil, d) closure of the area. The cost used, in accordance with the cost assigned to the unit collective dose a (alpha)= $20,000–$3,000 per man*Sv, facilitates a comparison of the justification of the cost alternatives of remedial action to suit the different economical conditions in Russia (the numeral values a were chosen by experts of MCC). It has been proven that under current Russian economical conditions (α = $3,000 per man*Sv) “no action” is best for most contaminated sites on the Enisey River floodplain. Removal of contaminated soil (without stabilization) is cost justified action for high contamination of small areas (such as “Gorodskoy” Island) only. Removal of the contamination in large areas (such as the “Kazachenskoe” settlement) may be a cost justified action in the future (for α = $20,000 per man*Sv).

A theoretical expression is described for the oxidative-dissolution rate response for multi-component radioactive materials that have surface adsorption kinetics and radiolysis kinetics when wetted by a multi-component aqueous solution. An application for this type of oxidative-dissolution response is the performance evaluation of multi-component spent nuclear fuels (SNFs) for long term interim storage and for geological disposition. Typically, SNF compositions depend on initial composition, uranium oxide and uranium-metal alloys being most common, and on reactor burnup which results in a wide range of fission product and actinide concentrations that decay by alpha, beta, and gamma radiation. These compositional/burnup ranges of SNFs, whether placed in interim storage or placed in a geologic repository, will potentially be wetted by multi-component aqueous solutions, and these solutions may be further altered by radiolytic aqueous species due to three radiation fields. The solid states of the SNFs are not thermodynamically stable when wetted and will oxidize and dissolve, with or without radiolysis. The following discussion of an oxidative-dissolution theory is based on a non-equilibrium thermodynamic analysis of energy reactions and energy transport across a solid-liquid phase change discontinuity that propagates at a quasi-steady, dissolution velocity. The integral form of the energy balance equation is used for this spatial surface discontinuity analysis. The integral formulation contains internal energy functionals of classical thermodynamics for both the SNFs’ solid state and surface adsorption species, and the adjacent liquid state, which includes radiolytic chemical species. The steady-state concentrations of radiolytic chemical species are expressed by an approximate analysis of the decay radiation transport equation. For purposes of illustration a modified Temkin adsorption isotherm will be assumed for the surface adsorption kinetics on an arbitrary, finite area of the solid-liquid dissolution interface. For the model developed, the propagation velocity of the solid-liquid dissolution interface is assumed proportional to configurational entropy discontinuity across the interface. These analyses of non-equilibrium thermodynamic processes across a propagating discontinuity, along with other idealized dissolution processes that depend on surface adsorption and radiolysis kinetics, provide generic dissolution response functions for empirical and/or regression analysis of data.

This paper gives a survey of the main results of research projects performed by members of ENTRAP to improve gamma-ray scanning techniques for radioactive waste packages. Performance characteristics, advantages and restrictions of different assay procedures and correction techniques investigated in these projects are discussed. Case studies are presented which demonstrate the difference in performance between ‘standard’ and ‘improved’ assay techniques. Consideration is also given to bias effects resulting from a limited knowledge on gamma-ray attenuation and/or source distribution in the waste matrix. Procedures and tools are presented which may help to decrease respective bias effects or to assess the overall uncertainty. The achievements emphasise that quality control of assay results requires, in nearly all applications, that additional assay techniques are implemented in order to verify the validity of conventional gamma-ray scanning methods.

The European Agency for Safety and Health at Work has issued a report, “Occupational Safety and Health in Marketing and procurement” (1) describing new ways to improve occupational safety and health through marketing and procurement initiatives. The report brings together 22 case examples, of voluntary initiatives taken by companies, sector organisations and governments to introduce: • the use of Occupational Safety and Health as a criteria in purchasing products and services from other companies; • the use of Occupational Safety and Health as a marketing element for promoting the sales of their products or services. The aim of this study has been to identify and describe interesting examples of how companies in the Member States use occupational safety and health in their marketing and procurement strategies. Either by making demands on the suppliers’ safety and health performance or by marketing their goods and services emphasizing company safety and health performance or the safety and health properties of their products as illustrated in Fig. 1. Some initiatives in this area have not been initiated by companies, but by trade organisations or industry associations or even with the involvement of national administrations. This can be considered a useful contribution to the development of these new approaches. It is of course efficient to develop these procurement methods jointly at sector level and share experiences. Furthermore it facilitates the access and use of these instruments if they are being taken care of by intermediary organisations. The examples given in this report have been identified in nine different EU Member States. For each of the levels identified — that is marketing or procurement at the company, sector or national level — cases are identified and described. The described initiatives (see Table I) are not selected to be representative for all existing schemes on the European scene but more to show different ways of promoting safety and health in the workplaces. This study does not evaluate the economic aspects of the different schemes in use. There is, however, no doubt that for the companies involved, the economic benefits on the bottom line are key motivators, whether it be a result of a reduction in lost time accidents, higher productivity or increased market shares. The study is based on interviews with key stakeholders, such as representatives of the developers of the schemes, company management, safety and health managers, purchasers, customers, and worker representatives. Written questionnaires are sometimes used instead of interviews.

EWN is performing a major decommissioning project of 5 WWER-440 reactors in Greifswald and a smaller WWER-70 reactor in Rheinsberg. The decommissioning concept foresees the complete dismantling of the reactor units up to green field conditions. The planning and management of such a project, and especially the handling of enormous amounts of material, require a computerised planning and management tool based on a suitable data base. Ideally, this tool will be an interrelated data base, containing the necessary registration planning, calculation, supervision and control modules and procedures as well as standardised data sets. The paper shortly describes the necessary basic hardware and software requirements, which have been developed. The scope of necessary input and output information and the developed modules as well as the interactions between them will be shown. The main parts of the EWN decommissioning management system are the decommissioning information system, the documentation management/service event tracking system and the environmental information system. The modules for the main project tasks — planning, cost and working resource definition as well as mass flow and project supervision will be described as well as the interaction between the data processing system and the practical needs of the EWN decommissioning project will be illustrated.

Under the geological disposal conditions, spent fuel (SF) is expected to evolve during the 10,000 years while being maintained isolated from the biosphere before water comes in. Under those circumstances, several driving forces would lead to the progressive intrinsic transformations within the rod which would modify the subsequent release of radionuclides: the production of a significant volume of He, the accumulation of irradiation defects, the slow migration of radionuclides (RN) within the pellet. However, the current RN source terms for SF never accounted for these evolutions and was based on the existing knowledge on the fresh SF. Two major mechanisms were considered, the leaching of the readily available fraction (one which was supposed to be instantly accessible to water), and the release of RN through alteration of the UO 2 grains. We are now proposing a new RN source term model based on a microscopic description of the system in order to also account for the early evolution of the closed system, the amplitude of which increases with the burnup and is greater for MOX fuels.

The AVK (Waste Flow Tracking and Quality Assurance System) represents a computer-aided system of the German nuclear power plants for the documentation of radioactive waste with negligible heat generation. AVK keeps track of the waste throughout all steps from their first registration up to final disposal. It is not only used by the waste producers (NPP) itself, but also by the waste conditioner GNS and the external interim waste storage facilities. They are all linked to the AVK-network with a tight organizational structure including a central quality control office and an active user support. The AVK developed under the DOS operating system has been under operation since 1991. Due to the fast development of computer hard and software in the past 10 years the PC-program AVK has undergone an extensive modernization resulting in the new program version AVK 3.0 for Windows. This paper describes the basic principles of the AVK, introduces the modernization project and the new data processing program AVK 3.0, and explains the process of converting networked AVK operation to the new program, which will probably be completed by the end of the year 2001.

In Belgium 57% of the electricity is presently generated by 7 nuclear units of the PWR type located in Doel and Tihange. Their total output amounts to 5632 MWe. Part of the spent fuel unloaded from the first three units has been sent till 2000 for reprocessing in the Cogema facility at La Hague. As the reprocessing of the spent fuel produced by the last four units is not covered by the contracts concluded with Cogema, Synatom, the Belgian utilities’ subsidiary in charge of the front- and back-end of the nuclear fuel cycle for all PWR reactors in Belgium, decided to study the possible solutions for a temporary storage of this spent fuel. End of 1993, the Belgian government decided that reprocessing (closed cycle) and direct disposal (open cycle) of spent fuel had to be considered as equal options in the back-end policy for nuclear fuel in Belgium. The resolution further allowed continued execution of a running reprocessing contract (from 1978) and use of the corresponding Pu for MOX in Belgian NPP’s, but requested a reprocessing contract concluded in 1990 (for reprocessing services after 2000) not to be executed during a five-year period. During this period priority was to be given to studies on the once-through cycle as an option for spent fuel management. Figure 1 is a chart showing the two alternatives for the spent fuel cycle in Belgium. In this context, Synatom entrusted Belgatom 1 to develop a dedicated flask (called “bottle”) for direct disposal of spent fuel, to perform a design study of an appropriate encapsulation process and to prepare a preliminary feasibility study of a complete spent fuel conditioning plant. Meanwhile preparation works were made for the construction of an interim storage facility on both NPP sites of Doel and Tihange in order to meet increasing storage capacity needs. For selecting the type of interim storage facility, Belgatom performed a technical-economical analysis. Considerations of design and safety criteria as well as flexibility, reversibility, technical constraints, global economical aspects and construction time led to adopt dry storage with dual purpose casks (in operation since end 1995) for the Doel site and wet storage in a modular pool for the Tihange site (in operation since 1997). In parallel, ONRAF/NIRAS, the Belgian Agency for the management of radioactive waste and enriched fissile materials and the Belgian nuclear research centre, SCK•CEN, conduct underground investigations in view of geological disposal. The paper describes the methodology that Belgatom has developed to provide the utilities with appropriate solutions (reracking, dry storage in casks, wet storage in ponds, etc.) and how Belgatom demonstrated also the feasibility of spent fuel conditioning with a view to direct disposal in clay layers. The spent fuel storage facilities in operation in Belgium and designed and built by Belgatom are then briefly presented.

At the base of establishing criteria’s selection of a host site for final repository of low and intermediate radioactive waste (LILW) are the study of two major components, the radioactivity diffusion in disposal site and the affecting degree of environment. The hydrological characteristics of formation are the main factors that control radionuclides moving (migration), because, in general, the water is the natural way for dissolving and transport of these in environment. For determination of these characteristics of groundwater that influence radionuclid migration process, we had draw and analysed periodically the groundwater sample from test boreholes realised in Saligny site. The analysis consisted in determination of physical-chemical proprieties of groundwater, such as: pH, conductivity, contents of total dissolved salts, hardness, major concentration of anion and cation species.

Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in shallow burial disposal facility as low level radioactive waste (LLW) must be solidified by cement with adequate strength and must extend no harmful openings. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete and metal for mortar to fill openings in waste forms. Performance of a method to pre-place large sized metal or concrete waste and to fill mortar using small sized metal or concrete was tested. It was seen that the improved method substantially increases the filling ratio, thereby decreasing the numbers of waste containers.

In the nuclear sector, elemental alkali metals are being used as coolant in the primary circuit of fast breeder reactors. Classical safety is a very critical issue for the treatment and conditioning of these metals, due to their chemical reactivity, and considering the quite severe acceptance criteria of the final waste. The SCK•CEN has conceived and patented a new and safe process to meet safety requirements and the compatibility with the further conditioning of the waste into an acceptable form. In this process, molten alkali metal reacts with a mixture of oxygen and carbon dioxide, in a fluidized bed reactor containing calibrated sand particles, to yield the metal oxide and carbonate, while avoiding the formation of peroxide. No hydrogen is formed, while the turbulent conditions guarantee a complete conversion of the metal and excellent heat transfer capabilities, eliminating explosion risks and reducing greatly fire risks. Depending on the level of radioactive contamination, the mixture of sand and carbonates may be considered as the final waste, or may be further conditioned as glass.

This paper describes the research work carried out at the NSC KIPT to develop and apply a final waste form in the shape of a monolithic solid block for the containment of spent nuclear fuel. To prepare radioactive waste for long-term storage and final deep geological disposal, investigations into the development of methods of immobilizing HLW simulators in protective solid matrices are being conducted at the NSC KIPT. For RBMK spent nuclear fuel it is proposed and justified to encapsulate the spent fuel bundles into monolithic protective blocks, produced with the help of hot isostatic pressing (HIP) of powder materials. In accordance with this approach, as a material for the protective block made up of the glass-ceramic composition prepared by sintering at isostatic pressure, the powder mixture of such natural rocks as granite and clay has been chosen. Concept approach and characterization of waste form, technological operations of manufacturing and performance assessment are presented. The container with spent fuel for long-term storage and final disposal presents a three barrier protective system: ceramic fuel UO 2 in cladding tube, material of the glass-ceramic block, material of the sealed metal capsule. Investigations showed that the produced glass-ceramic material is characterized by high stability of chemical and phase compositions, high resistance in water medium, low porosity (compared with the porosity of natural basalt). With the help of mathematical calculations it was shown that the absorbed dose of immobilizing material by RBMK spent fuel irradiation for 1000 years of storage in the geological disposal after 10 years of preliminary cooling will be ∼ 3.10 8 Gy, that is 2–3 orders of magnitude less than the values corresponding to preserving radiation resistance and functional parameters of glasses and ceramics. The average value of velocity of linear corrosion in water medium of the protective layer made up of the glass-ceramic composition determined experimentally makes up ∼ 15 mm per year. This allows to use glass-ceramic compositions effectively as an engineering barrier in the system of spent fuel geological disposal and to increase the lifetime of the waste container, in particular, up to 3000 years with the layer thickness ∼ 40 mm. The possible release of radionuclides from the waste container during its interim storage in the open air (near-surface storage) is estimated. The calculations are made by taking into account the possible increase of coefficients of radionuclide diffusion from 10 −16 to 10 −14 m 2 /c as a result of spent fuel radiation affecting the protective layer. The obtained results showed that the protective barrier (about 40 mm) at the base of the glass-ceramic composition, ensures reliable isolation from the environment against the release of radionuclides from the controlled near-surface long-term storage far up to 1000 years. The relatively limited release of radionuclides will make up about 1% for the period of more than 400 years, and 10% - in 1000 years. During this period of time, the radionuclides 90 Sr and 137 Cs will completely turn into stable 90 Zr and 137 Ba and the decay of many transuranium elements will occur. The results from laboratory scale experiments, tests and calculations carried out so far, show that the proposed glass-ceramic materials may be used as basic materials for manufacturing the monolithic protective block in which the spent fuel elements will be embedded with the aim of further long-term storage or final disposal.

Northwest Russia contains large quantities of spent nuclear fuel (SNF) that potentially threaten the fragile environment of the surrounding Arctic region. The majority of the Russian SNF from their decommissioned nuclear submarines and civilian icebreaker fleet is currently stored either onboard submarines or in floating storage vessels in Northwest Russia. Some of the SNF is damaged, stored in an unstable condition, or of a type that cannot currently be reprocessed. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Some of the existing storage facilities being used in Northwest Russia are unsafe both from a health and safety aspect, as well as an environmental perspective. The removal, handling, interim storage, and shipment of the fuel pose technical and ecological challenges. The U.S. Environmental Protection Agency (EPA), with support from the U.S. Department of Defense, U.S. Department of State, and the Department of Energy’s (DOE) Oak Ridge National Laboratory, is working closely with Minatom of Russia in two multilateral cooperative projects which provide assistance to the Russian Federation (RF) in the management of some of their military and civilian SNF. These two projects involve the development of prototype containers and container storage facilities to meet RF military and civilian requirements. Specifically, these projects involve the development of prototype dual-purpose, metal-concrete containers for the transport and storage of RF SNF and the development of suitable storage facilities for the containers. These are the first dual-purpose containers developed for use in Russia. The projects also address the limitations of the existing infrastructure at the various military and civilian facilities. These projects are designed to provide a safe and environmentally sound interim solution for managing the SNF until permanent storage is attained. There are two prototype dual-purpose containers under development. The first is a 40-tonne container for the handling of military SNF from decommissioned nuclear submarines awaiting dismantlement at various naval installations in the Arctic and Far East regions of Russia. This container is limited to a 40-tonne size because of the limitations of the RF military infrastructure. The second is an 80-tonne container intended primarily for transporting and storage of SNF currently being stored in floating vessels in the Murmansk harbor in Northwest Russia. Both activities also involve the development and construction of special prototype/demonstration concrete storage pads/facilities to store the containers on an interim basis for varying periods of time (a few months up to 50 years). These projects are designed to provide a safe and environmentally sound interim solution while increasing the capacity for removal and management of SNF from decommissioned RF submarines until permanent storage is attained.

The Cernavoda Nuclear Power Plant (NPP), in commercial operation since 1996, produces more than 10% of the electricity produced in Romania. Recently, the Romanian Government declared its commitment for completion of a second reactor of the CANDU design, under construction on the Cernavoda site. The annual spent fuel arising from a CANDU reactor is about 100tU. The current policy for spent fuel management as practiced by the plant owner is to store it in the reactor bay for minimum six years and in a dry storage facility for a minimum of 50 years. For geological disposal of spent fuel, the “wait and see” strategy is considered the best approach, as Romania has a relative low scale nuclear program and wants to benefit by the international progress in this field. The construction of a new spent fuel dry storage facility located in the vicinity of the nuclear power reactor site represents a main priority for the next three years. The site of this facility will accommodate two nuclear units’ inventories of spent fuel for the entire planned lifetime. An international public-limited tender was organized to select the supplier of the dry storage technology in early 2001. The tenderer was asked to propose a proven and licensed technology capable of storing CANDU spent fuel according to specified design parameters and safety and environmental requirements. Design, construction, operation or licensing legal specific requirements for such a facility is generally not established and other already existing national requirements are applicable to a limited degree. Taking into account the different approaches and iterative processes required for Romanian authorities to regulate the nuclear activities for different fields, this paper considers the realistic path forward. The current status and main aspects of the development and licensing of the new nuclear facility in Romania is presented in this paper.