Remote Dismantling of the WWER Reactors in Greifswald

At the Greifswald site, there are 8 units of the Russian pressurised water reactor type WWER 440. After the reunification of Germany, it was decided to decommission all units without a safe enclosure phase. Due to the high activation of the components of the reactors 1 to 4, it was necessary to use mainly remote dismantling techniques. The concept first foresees a model dismantling in the steam generator room of unit 5 to test the transport, cutting and packaging equipment to perform the later dismantling of the activated reactors 1 to 4 safely and optimally. For this purpose the non-activated and non-contaminated equipment of the units 7 and 8 will be used. For the dismantling, a dry cutting area, wet cutting area and a cutting area in the reactor cavity room are foreseen. The installations of the cutting equipment in the steam generator room of unit 5 were executed from June 1997 until September 1999. During model dismantling, several technical problems occurred, especially with the band saws and video technique, which strongly affected or even stopped the tests. For these reasons, only few data are available for the evaluation of the cutting procedures and for the estimation of the time needed as well as for the radiological exposure to be expected for units 1 to 4.

Dismantling of the BR3 Reactor Pressure Vessel

The BR3 PWR is a small nuclear power plant (thermal power 40.9 MWth, net electrical power output 10.5 MWe), designed in the late fifties and started in 1962. It was definitely shut down in 1987. In 1989 the BR3 was selected by the European Union as pilot decommissioning project in the framework of its RTD programme on the decommissioning of nuclear installations. A pre-dismantling decontamination of the reactor primary loop was carried out and allowed to save doses to the operators. The savings are estimated to be up to about 4 to 7 man-Sv. The decommissioning project concerns mainly: • The dismantling of the highly radioactive reactor internals. Different techniques were used and compared on a first actual piece called the thermal shield: from plasma arc torch cutting to mechanical sawing, including also electric discharge machining. Based on the experience gained during this part of the project, the mechanical cutting techniques were promoted for the segmentation of both sets of internals, the desolidarisation and the segmentation of the RPV. • For the dismantling of the reactor pressure vessel, wet and dry dismantling were studied and compared. For economical and feasibility reasons, the wet dismantling was selected. Afterwards, two underwater segmentations were also studied: in-situ segmentation and a segmentation after having removed the RPV out of its cavity. • Mainly for technical reasons, the reactor pressure vessel was removed in one piece out of its cavity in order to be cut in the former refuelling pool. The disconnection of the RPV from the other parts of the plant was followed by the reinstallation of the watertightness of the pool in order to allow remote underwater segmentation. The disconnection, the watertightness reinstallation and the segmentation represented important challenges. The subtasks will be extensively described in the paper: disconnection from the pools floor, removal of the thermal insulation from the legs, decoupling from the primary loop at two levels, from its supporting structure, the reinstallation of the watertightness of the pool and testing, the removal of the RPV out of its cavity, the remote dismantling of its surrounding thermal insulation (which led to an annoying pool water turbidity) and, finally the effective RPV dismantling. • For the segmentation, two main cutting equipments were used: the milling cutter for cutting the RPV into rings and the bandsaw machine for cutting each ring into segments. The bandsaw machine was also used in order to cut the RPV upper flange into pieces vertically as well as horizontally. • The last generated pieces, the highest radioactive ones, were evacuated at the end of 2000. • Waste characterisation, minimization and management is an important part of the task in order to reduce evacuation and storage costs. • ALARA approach was applied from the early beginning of the project. • For each “key operation” cold tests were organized in order to optimize the work and to take benefit of the learning effect of such operation. Results of the operations will be presented, the lessons drawn for the technical choices, dose uptake minimization, waste reduction and the technical problems met will be highlighted. As a pioneering project, the dismantling of the BR3 Reactor Pressure Vessel has shown the technical feasibility of such an operation in a safe and economical way as well.

Operation Experience of HALW Storage Process at the Tokai Reprocessing Plant

Volume of nitric acid solution, which contain the most of fission products (FPs), is concentrated to 0.5–2m3 in an HALW evaporator by reprocessing a spent fuel of ltU. The HALW is stored in HALW storage tanks temporarily till it is transferred to the Tokai Vitrification Facility (TVF). During the storage, the HALW of 1–8m3/y vaporized per storage tank. A shift coefficient of radioactive nuclides from the HALW to off-gas was 4E−11. Through the operation experience, knowledge obtained about storage management of highly active liquid waste (HALW) is reported.

Derivation of Clearance Levels for the New German Radiation Protection Ordinance

The new Radiation Protection Ordinance (RPO) in Germany which transforms the EURATOM Basic Safety Standards into national legislation contains detailed regulations on clearance. For each of the following clearance options, a separate set of clearance levels (CL) exists: unconditional clearance, clearance of building rubble (> 1000 Mg/a), clearance of buildings for demolition and for reuse, clearance of nuclear sites, and clearance for disposal or incineration. This paper outlines the basis for the derivation of these sets of CL which are all based on generic radiological scenarios taking into account all relevant aspects of the materials. The underlying dose criterion is 10 μSv/a individual dose and 1 man·Sv/a collective dose. When deriving sets of CL in Germany care has been taken to be compatible with recommendations of the European Union and the IAEA. It is a common experience that sets of CL which are intended for the same purpose (e.g. general, unconditional clearance) may vary between studies and therefore between countries. This follows directly from differences e.g. in material quantities, boundary conditions, waste management options etc. which may be country specific. German CL are, however, in full agreement with all recommendations issued by the European Commission.

Pollutant Release Level Prognosis: A Major Input Into the Flooding Concept for the Former ISL Uranium Mine at Königstein (Germany)

At the Königstein mine uranium was extracted by an underground in situ leaching method. WISMUT developed a flooding concept which allows the reduction of pollutant concentrations and prevent pollutant migration into the aquifers above and downstream the mine. The development of the concept and the documentation for permit application were based on a multitude of scientific and engineering studies and prognoses on substance output using two different approaches (upscaling of a flooding experiment, Numeric box model). Both modelling tools provide similar prognoses of flooding. With the flooding in progress (start January 2001), the modelling tools will be validated and further improved.

Geological Disposal for High-Level Waste: Ethical Considerations

This paper describes the activities launched at SCK•CEN, intended to explore ethical and other non-technical aspects when dealing with the time scales considered in the high-level waste disposal program. (1) Especially the issues of retrievability and precaution will be focused on philosophically. Many questions will be raised in order to sensitize all stakeholders for the transdisciplinary character of the transgenerational problem at hand.

Hybrid Probabilistic and Possibilistic Safety Assessment: Methodology and Application

This paper presents a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to the safety assessment of geological disposal of high level radioactive waste. Ignorances associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts, while variability was formulated by means of probability density functions (pdfs) based on available data sets. The exercise demonstrated the applicability of the new methodology and, in particular, its advantage in quantifying ignorances based on expert opinion and in providing information on the dependence of assessment results on the level of conservatism. In addition, it was shown that sensitivity analysis can identify key parameters contributing to uncertainties associated with results of the overall assessment. The information mentioned above can be utilized to support decision making and to guide the process of disposal system development and optimization of protection against potential exposure.

Special Waste Generated by the Belgian Nuclear Research Centre

At research centres, test programs generate very particular waste forms, which cannot be treated using the known regular scenario’s and procedures, and therefore special studies need to evaluate safe handling of these special waste forms. This paper describes how a joint effort between the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, the Belgian Waste Treatment and Storage Facility Belgoprocess, and the Belgian Nuclear Research Centre worked out solutions for the 1st beryllium moderator of the BR2 and the highly activated internal parts of the BR3. The current position regarding activated graphite and aluminium is also summarised.

Methodology in Corrosion Testing of Container Materials for Radioactive Waste Disposal in a Geological Clay Repository

The current worldwide trend for the final disposal of conditioned high-level, medium-level and long-lived alpha-bearing radioactive waste focuses on deep geological disposal. During the geological disposal, the isolation between the radioactive waste and the environment (biosphere) is realised by the multibarrier principle, which is based on the complementary nature of the various natural and engineered barriers. One of the main engineered barriers is the metallic container (overpack) that encloses the conditioned waste. In Belgium, the Boom Clay sediment is being studied as a potential host rock formation for the final disposal of conditioned high-level radioactive waste (HLW) and spent fuel. Since the mid 1980’s, SCK•CEN has developed an extensive research programme aimed at evaluating the suitability of a wide variety of metallic materials as candidate overpack material for the disposal of HLW. A multiple experimental approach is applied consisting of i) in situ corrosion experiments, ii) electrochemical experiments (cyclic potentiodynamic polarisation measurements and monitoring the evolution of ECORR as a function of time), and iii) immersion experiments. The in situ corrosion experiments were performed in the underground research facility, the High Activity Disposal Experimental Site, or HADES, located in the Boom clay layer at a depth of 225 metres below ground level. These experiments aimed at predicting the long-term corrosion behaviour of various candidate container materials. It was believed that this could be realised by investigating the medium-term interactions between the container materials and the host formation. These experiments resulted in a change of reasoning at the national authorities concerning the choice of over-pack material from the corrosion-allowance material carbon steel towards corrosion-resistant materials such as stainless steels. The main arguments being the severe pitting corrosion during the aerobic period and the large amount of hydrogen gas generated during the subsequent anaerobic period. The in situ corrosion experiments however, did not allow to unequivocally quantify the corrosion of the various investigated candidate overpack materials. The main shortcoming was that they did not allow to experimentally separate the aerobic and anaerobic phase. This resulted in the elaboration of a new laboratory programme. Electrochemical corrosion experiments were designed to investigate the effect of a wide variety of parameters on the localised corrosion behaviour of candidate overpack materials: temperature, SO42−, Cl−, S2O32−, oxygen content (aerobic - anaerobic),… Three characteristic potentials can be derived from the cyclic potentiodynamic polarisation (CPP) curves: i) the open circuit potential, OCP, ii) the critical potential for pit nucleation, ENP, and iii) the protection potential, EPP. Monitoring the open circuit potential as a function of time in clay slurries, representative for the underground environment, provides us with a more reliable value for the corrosion potential, ECORR, under disposal conditions. The long-term corrosion behaviour of the candidate overpack materials can be established by comparing the value of ECORR relative to ENP and EPP (determined from the CPP-curves). The immersion tests were developed to complement the in situ experiments. These experiments aimed at determining the corrosion rate and to identify the corrosion processes that can occur during the aerobic and anaerobic period of the geological disposal. Also, some experiments were elaborated to study the effect of graphite on the corrosion behaviour of the candidate overpack materials.

Near Field Radionuclide Concentrations: Sorption or Solubility Constrained?

The amount of mobile radionuclides is controlled by the geochemical isolation potential of the repository. Many investigations are available in order to determine the maximal radionuclide concentrations released from different waste forms of specific disposal strategies for disposal in rock salt formations. These investigations result in reaction (dissolution) rates, maximum concentrations, and sorption coefficients. The experimental data have to be applied to various disposal strategies. The case studies presented in this communication cover the selection, the volumes, and the composition of backfill materials used as sorbents for radionuclides. As an example, for brown coal fly ash (BFA) - Q-brine systems, sorption coefficients were measured as well as solublilities of several actinides and other long-lived radionuclides. Dissolved CO32− was buffered to negligible concentration by the presence of high amount of Mg in solution. In the sorption experiments Pu, Th, Np, and U concentrations close or below detection limit were obtained. Concentrations in the same ranges are computed by means of geochemical modeling, if precipitation of “simple” tetravalent hydroxides (An(OH)4(am) phases) is assumed. In the case of U in a Portland cement dominated geochemical environment, measured U(VI) concentration corresponds to the solubility of hexavalent solids, such as Na2U2O7. A similar behavior of U was observed in high-level waste glass experiments. Experiments investigating sorption behavior of corroded cement showed that in the case of application of a sufficient large inventory of actinides, measured concentrations were found to be independent of the inventory. In this case, measured concentrations were controlled by solid phases. If smaller actinide inventories were applied, resulting concentrations were found to be below concentrations constrained by well-known solids. Here, a more or less pronounced sorption of the radioactive elements was observed. The radionuclide concentrations determined in the BFA “sorption” experiments are found to be close to the detection limits. For this reason, it is not possible to extrapolate the radionuclide behavior to lower concentrations. We cannot distinguish, if sorption or precipitation controls measured radionuclide concentrations. However, in the presence of reducing materials such as BFA, solubilities of tetravalent actinides and of Tc(IV) represent a realistic estimation of the maximal element concentrations needed in performance assessment studies. The concentrations of these redox sensitive elements are controlled by precipitation of An(OH)4(am) phases for disposal concepts considered in German salt formations. Under this assumption, quantities such as solid-solution ratios used in (sorption) experiments do not affect the mobilization of the radionuclides. Additional conclusions can be drawn from comparison of the findings for the redox sensitive elements in the BFA / portland cement brine systems: We can assume that expected actinide and technetium concentrations in the near-field of radioactive wastes are affected by the total inventory of radionuclides in the disposal room. Sorption will be relevant, if the total dissolved radionuclide concentration remains below the maximal solubility defined by the solid radionuclide phase which is stable in the geochemical environment. In contrast to the portland cement system, the relevant radionuclide phase are most probably tetravalent hydroxides in the BFA systems. These conclusions are of high importance to performance assessment for the radioactive waste repository systems, because they restrict the applicability of sorption models in the near field of the waste.